NUREG-1757 Volume 2, Rev 2, Consolidated Decommissioning Guidance, Characterization, Survey, and Determination of Radiological Criteria

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NUREG-1757, Vol 2, Rev 2, Consolidated Decommissioning Guidance, Characterization, Survey, and Determination of Radiological Criteria
ML22194A859
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Issue date: 07/31/2022
From: Cynthia Barr, Gregory Chapman, Sheldon Clark, David Esh, Randall Fedors, Anthony Huffert, Laurie Kauffman, Michael Lafranzo, Chris Mckenney, Leah Parks, Schmidt D, Adam Schwartzman, Bruce Watson
Office of Nuclear Material Safety and Safeguards
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Dickey K
References
NUREG-1757 V2 R2
Download: ML22194A859 (579)


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NUREG-1757 Volume 2 Revision 2 Consolidated Decommissioning Guidance Characterization, Survey, and Determination of Radiological Criteria Final Report Office of Nuclear Material Safety and Safeguards

AVAILABILITY OF REFERENCE MATERIALS IN NRC PUBLICATIONS NRC Reference Material Non-NRC Reference Material As of November 1999, you may electronically access Documents available from public and special technical NUREG-series publications and other NRC records at the libraries include all open literature items, such as books, NRCs Library at www.nrc.gov/reading-rm.html. Publicly journal articles, transactions, Federal Register notices, released records include, to name a few, NUREG-series Federal and State legislation, and congressional reports.

publications; Federal Register notices; applicant, licensee, Such documents as theses, dissertations, foreign reports and vendor documents and correspondence; NRC and translations, and non-NRC conference proceedings correspondence and internal memoranda; bulletins and may be purchased from their sponsoring organization.

information notices; inspection and investigative reports; licensee event reports; and Commission papers and their Copies of industry codes and standards used in a attachments. substantive manner in the NRC regulatory process are maintained at NRC publications in the NUREG series, NRC regulations, The NRC Technical Library and Title 10, Energy, in the Code of Federal Regulations Two White Flint North may also be purchased from one of these two sources: 11545 Rockville Pike Rockville, MD 20852-2738

1. The Superintendent of Documents U.S. Government Publishing Office These standards are available in the library for reference Washington, DC 20402-0001 use by the public. Codes and standards are usually Internet: www.bookstore.gpo.gov copyrighted and may be purchased from the originating Telephone: (202) 512-1800 organization or, if they are American National Standards, Fax: (202) 512-2104 from American National Standards Institute
2. The National Technical Information Service 11 West 42nd Street 5301 Shawnee Road New York, NY 10036-8002 Alexandria, VA 22312-0002 Internet: www.ansi.org Internet: www.ntis.gov (212) 642-4900 1-800-553-6847 or, locally, (703) 605-6000 Legally binding regulatory requirements are stated only in A single copy of each NRC draft report for comment is laws; NRC regulations; licenses, including technical available free, to the extent of supply, upon written specifications; or orders, not in NUREG-series publications.

request as follows: The views expressed in contractor prepared publications in this series are not necessarily those of the NRC.

Address: U.S. Nuclear Regulatory Commission The NUREG series comprises (1) technical and Office of Administration administrative reports and books prepared by the staff (NUREG-XXXX) or agency contractors (NUREG/CR-XXXX),

Digital Communications and Administrative (2) proceedings of conferences (NUREG/CP-XXXX),

Services Branch (3) reports resulting from international agreements Washington, DC 20555-0001 (NUREG/IA-XXXX),(4) brochures (NUREG/BR-XXXX), and E-mail: distribution.resource@nrc.gov (5) compilations of legal decisions and orders of the Facsimile: (301) 415-2289 Commission and the Atomic and Safety Licensing Boards and of Directors decisions under Section 2.206 of the NRCs regulations (NUREG-0750).

Some publications in the NUREG series that are posted at the NRCs Web site address www.nrc.gov/reading-rm/ DISCLAIMER: This report was prepared as an account doc-collections/nuregs are updated periodically and may of work sponsored by an agency of the U.S. Government.

differ from the last printed version. Although references to Neither the U.S. Government nor any agency thereof, nor any employee, makes any warranty, expressed or implied, material found on a Web site bear the date the material or assumes any legal liability or responsibility for any third was accessed, the material available on the date cited partys use, or the results of such use, of any information, may subsequently be removed from the site. apparatus, product, or process disclosed in this publication, or represents that its use by such third party would not infringe privately owned rights.

NUREG-1757 Volume 2 Revision 2 Consolidated Decommissioning Guidance Characterization, Survey, and Determination of Radiological Criteria Final Report Manuscript Completed: April 2022 Date Published: July 2022 Prepared by:

C.S. Barr, S. Clark, G.C. Chapman, D.W. Esh, R.W. Fedors, A.M. Huffert*,

L.A. Kauffman*, M.M. LaFranzo, C.A. McKenney, L.L. Parks, D.W. Schmidt*, A.L. Schwartzman, and B.A. Watson

  • retired Cynthia Barr, NRC Project Manager and Senior Risk Analyst Office of Nuclear Material Safety and Safeguards

PAPERWORK REDUCTION ACT STATEMENT This NUREG provides voluntary guidance for implementing the mandatory information collections in 10 CFR 20 that are subject to the Paperwork Reduction Act of 1995 (44 U.S.C. 3501 et. seq.). These information collections were approved by the Office of Management and Budget (OMB), approval number 3150-0014. Send comments regarding this information collection to the FOIA, Library, and Information Collections Branch (T-6A10), U.S.

Nuclear Regulatory Commission, Washington, DC 20555-0001, or by e-mail to Infocollects.Resource@nrc.gov, and to the OMB reviewer at: OMB Office of Information and Regulatory Affairs (3150-0014), Attn: Desk Officer for the Nuclear Regulatory Commission, 725 17th Street, NW Washington, DC 20503; e-mail: oira_submission@omb.eop.gov.

PUBLIC PROTECTION NOTIFICATION The NRC may not conduct or sponsor, and a person is not required to respond to, a request for information or an information collection requirement unless the requesting document displays a currently valid OMB control number.

ABSTRACT The U.S. Nuclear Regulatory Commission (NRC), Office of Nuclear Material Safety and Safeguards (NMSS), previously consolidated and updated numerous decommissioning guidance documents into a three-volume NUREG. This NUREG series is intended for use by the NRC staff, licensees, and others. The three volumes address the following topics:

(1) Decommissioning Process for Materials Licensees (2) Characterization, Survey, and Determination of Radiological Criteria (3) Financial Assurance, Recordkeeping, and Timeliness The staff last updated Volume 2 of the NUREG series, entitled, Consolidated Decommissioning Guidance: Characterization, Survey, and Determination of Radiological Criteria, in September 2006. This volume provides guidance on compliance with the radiological criteria for license termination (License Termination Rule (LTR)) in Title 10 of the Code of Federal Regulations (10 CFR) Part 20, Standards for Protection against Radiation, Subpart E, Radiological Criteria for License Termination. This guidance takes a risk-informed, performance-based approach to the demonstration of compliance. This guidance will help identify the information (subject matter and level of detail) needed to terminate a license and considers the specific circumstances of the wide range of NRC licensees. Licensees should use this guidance in preparing decommissioning plans (DPs), license termination plans, final status surveys, and other technical decommissioning reports for NRC submittal. The NRC staff will use the guidance in reviewing these documents and related license amendment requests. Volume 2 applies to all licensees subject to the LTR (i.e., fuel cycle, fuel storage, materials, and reactor licensees).

Changes made to this revision of Volume 2 include the following:

  • Dose Modelingadds guidance on model abstraction and simplification, consideration of elevated areas, use of distribution coefficients in dose modeling, and consideration of uncertainty.
  • As Low As Is Reasonably Achievable (ALARA) Analysisupdates guidance on the ALARA analysis review particularly for restricted release based on lessons learned from proposed restricted release scenarios.
  • Composite Samplingadds information on methodologies for incorporating composite sampling strategies into final status survey plans and details when it would or would not be appropriate to use composite sampling.
  • Characterizationupdates information in Appendix F on surface water and groundwater characterization.
  • Engineered Barrier Analysisupdates and reorganizes Section 3.5 and Appendix P; includes new information on how ALARA is considered prior to engineered barriers for restricted release and updates bibliography with new references on evaluation of engineered performance and degradation.

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  • Radiological Surveysprovides updated guidance on subsurface radiological surveys, including surveys associated with excavations and re-use of soils; and provides additional information on use and implementation of Scenario B, an alternative for the null hypothesis of statistical tests evaluated in a final status survey.
  • Lessons Learnedremoves Appendix O which contains lessons learned from RIS-2002-02; and archives questions and answers, and lessons learned in ADAMS (ADAMS Accession Number (No.) ML20052C815). Future interim guidance will be placed on the NRC public decommissioning website to allow for more timely updates to guidance between NUREG revisions. Frequently asked questions can also be found in NUREG-1628.
  • Miscellaneous Editorial Changes-corrects typographical and formatting errors; adds clarity to areas of the guidance.

Draft NUREG-1757, Volume 2, Revision 2 Consolidated Decommissioning Guidance, Characterization, Survey, and Determination of Radiological Criteria was issued for public comment in the Federal Register (85 FR 79044) on December 8, 2020, with a 60-day comment period. On January 22, 2021, (86 FR 6683) the NRC extended the comment period until April 8, 2021. Over 200 comments were received on the draft document and the NRC staffs evaluation and resolution of the public comments are documented in Agencywide Documents Access and Management System under Accession No. ML21299A032. NRC staff addressed the public comments in this final version of this document.

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FOREWORD The NRC staff suggests that licensees contact the NRC or the appropriate Agreement State authority to ensure understanding of what actions should be taken to initiate and complete decommissioning at facilities.

In September 2003, the U.S. Nuclear Regulatory Commission (NRC) staff in the Office of Nuclear Material Safety and Safeguards (NMSS) consolidated and updated the policies and guidance of its decommissioning program in a three-volume NUREG series, NUREG-1757, Consolidated NMSS Decommissioning Guidance. This NUREG series provides guidance on planning and implementing license termination under the NRCs License Termination Rule (LTR) in Title 10 of the Code of Federal Regulations (10 CFR) Part 20, Standards for Protection against Radiation, Subpart E, Radiological Criteria for License Termination; complying with the radiological criteria for license termination; and complying with the requirements for financial assurance and recordkeeping for decommissioning and timeliness in decommissioning materials facilities. The staff periodically updates NUREG-1757 to reflect current NRC decommissioning policy.

This document provides guidance for a wide range of situations that may exist at a decommissioned facility. It is expected that no site will have every situation present and that the licensee, and NRC reviewers will identify the applicable guidance through careful decisions on the intended end state of the site, conditions at the site, and the licensees approach to surveys and demonstrating compliance.

In September 2005, the staff issued, for public comment, draft Supplement 1 to NUREG-1757, which contained proposed updates to the three volumes of NUREG-1757. Draft Supplement 1 included new and revised decommissioning guidance that addresses some of the LTR implementation issues, which were analyzed by the staff in two Commission papers (SECY-03-0069, Results of the LTR Analysis, dated May 2, 2003; and SECY-04-0035, Results of the License Termination Rule Analysis of the Use of Intentional Mixing of Contaminated Soil, dated March 1, 2004). These issues include restricted use and institutional controls, onsite disposal of radioactive materials (10 CFR 20.2002, Method for Obtaining Approval of Proposed Disposal Procedures), selection and justification of exposure scenarios based on reasonably foreseeable future land use (realistic exposure scenarios), intentional mixing of contaminated soil, and removal of material after license termination. The staff also developed new and revised guidance on other issues, including engineered barriers.

The staff received stakeholder comments on the draft NUREG and prepared responses to these comments. Stakeholder comments and responses are located in the Agencywide Documents Access and Management System (ADAMS) at Accession No. ML063250385. Comments were addressed and updated sections from Supplement 1 were placed into the appropriate locations in revisions of Volumes 1 and 2 of NUREG-1757 in 2006 (NUREG-1757, Volume 1, Revision 2; and NUREG-1757, Volume 2, Revision 1). The staff revised Volume 3 of this NUREG in 2012, and that revision incorporated the Supplement 1 guidance that is related to Volume 3.

Draft NUREG-1757, Volume 2, Revision 2 Consolidated Decommissioning Guidance, Characterization, Survey, and Determination of Radiological Criteria was issued for public comment in the Federal Register (85 FR 79044) on December 8, 2020, with a 60-day comment v

period. On January 22, 2021, (86 FR 6683) the NRC extended the comment period until April 8, 2021. Over 200 comments were received on the draft document and the NRC staffs evaluation and resolution of the public comments are documented in ADAMS under Accession No. ML21299A032. NRC staff addressed the public comments in this final version of this document.

The NRC continues to increase the use of risk information in its regulation of nuclear materials and nuclear waste management, including the decommissioning of nuclear facilities. The NRCs risk-informed regulatory approach to the decommissioning of nuclear facilities represents a philosophy whereby risk insights are considered, together with other factors, to better focus the attention and resources of both the licensee and the NRC on the more risk-significant aspects of the decommissioning process and on the elements of the facility and the site that will most affect risk to members of the public following decommissioning. This results in a more effective and efficient regulatory process.

The information used to risk inform the decommissioning process typically comes from the results and findings of risk assessments or dose modeling. A risk assessment is a type of systematic analysis used to understand what can happen, how likely it is to happen, and the resulting consequences. Dose modeling is used to estimate potential dose to members of the public who may use the decommissioned site in the future following license termination. The end result of such assessments (e.g., the calculation of predicted doses from decommissioned sites) relates directly or indirectly to public health effects. The NRC staff has developed this guidance consistent with a risk-informed approach.

The primary decommissioning guidance documents used by licensees and the NRC staff are NUREG-1757 and NUREG-1700, Revision 2, Standard Review Plan for Evaluating Nuclear Power Reactor License Termination Plans, issued April 2018. NUREG-1537, Guidelines for Preparing and Reviewing Applications for the Licensing of Non-Power Reactors, issued February 1996, includes a section on decommissioning and license termination for nonpower reactors. The primary guidance used by the NRC staff for complying with the National Environmental Policy Act is NUREG-1748, Environmental Review Guidance for Licensing Actions Associated with NMSS Programs, and for decommissioning surveys is NUREG-1575, Multi-Agency Radiation Survey and Site Investigation Manual (MARSSIM). Table 1 below describes the general applicability of these documents.

Since the last revision of this volume in 2006, which addressed comments on draft Supplement 1, the NRC staff and larger decommissioning community has gained experience on a number of technical issues for which guidance has been developed. Table 2 describes the most significant changes to the guidance in this volume to include new and updated information based on this experience.

NUREG-1757, Volume 2, Revision 2, is applicable to all licensees that are subject to the LTR.

NUREG-1757 is intended for use by applicants, licensees, NRC license reviewers, and other NRC personnel. It is also available to Agreement States and the public.

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Table 1 Contents and Applicability of Key Decommissioning Guidance Documents Licensees to Which the Volume and Status1 Title Guidance Applies NUREG-1757, Vol. 1, Consolidated Decommissioning Fuel cycle, fuel storage, and Rev. 2, September Guidance: Decommissioning materials licensees2; limited 2006 Process for Materials Licensees applicability to reactor licensees (see text below)

NUREG-1757, Vol. 2, Consolidated Decommissioning All licensees that are subject Rev. 2; July 2022 (this Guidance: Characterization, to the LTR (fuel cycle, fuel revision) Survey, and Determination of storage, materials, and reactor Radiological Criteria licensees)

NUREG-1757, Vol. 3, Consolidated Decommissioning Volume 3 is intended to apply Rev. 1, February 2012 Guidance: Financial Assurance, only to the decommissioning Recordkeeping, and Timeliness of materials facilities licensed under Title 10 of the Code of Federal Regulations (10 CFR)

Parts 30, 40, 70, and 72.

NUREG-1700, Rev. 2, Standard Review Plan for Power reactor licensees April 2018 Evaluating Nuclear Power Reactor License Termination Plans NUREG-1537, Guidelines for Preparing and Nonpower reactor licensees February 1996 Reviewing Applications for the Licensing of Non-Power Reactors NUREG-1748, Environmental Review Guidance All licensees August 2003 for Licensing Actions Associated with NMSS Programs NUREG-1575, Rev. 1, Multi-Agency Radiation Survey All licensees August 2000 and Site Investigation Manual (MARSSIM)

Notes:

1 Versions listed are current as of the date of publication of this document in July 2022. The NRCs Library at http://www.nrc.gov/reading-rm/doc-collections/nuregs contains the most up-to-date version.

2 This refers to licensees regulated under 10 CFR Parts 30, 40 (but not uranium recovery facilities), 60, 61, 63, 70, and 72 (for 10 CFR Parts 60, 61, and 63, this NUREG is applicable only to the ancillary surface facilities that support radioactive waste disposal activities). Because uranium recovery facilities are not subject to 10 CFR Part 20, Subpart E, NUREG-1620, Revision 1, Section 5, should be used for decommissioning guidance for uranium recovery facilities that are subject to Appendix A, Criteria Relating to the Operation of Uranium Mills and the Disposition of Tailings or Wastes Produced by the Extraction or Concentration of Source Material From Ores Processed Primarily for Their Source Material Content, to 10 CFR Part 40, Domestic Licensing of Source Material.

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Table 2 Summary of Major Changes to Volume 2, Revision 2 Subject Affected Sections Updated guidance on review of dose modeling used to Chapters 2, 5, demonstrate compliance with radiological criteria for license Appendix I termination including the following topics:

  • revised and added guidance on technical issues associated with model abstraction and simplification,
  • source term abstraction and development of site-specific parameters such as distribution coefficients, and
  • consideration of elevated areas or hot spots.

Updated guidance on engineered barriers to reflect recent Section 3.5 and research; updated guidance on consideration of ALARA prior to Appendix P use of engineered barriers to reduce dose for restricted release.

Updated guidance on review of ALARA analysis including Chapter 6 and experience and lessons learned gained from proposed restricted Appendix N release scenarios; updated regulatory citations and guidance related to discounting and the monetary value of collective dose averted.

Updated guidance on surface water and groundwater Appendix F characterization.

Added guidance on subsurface investigations, survey of Section 3.6 and excavations, survey of back-fill soils, data visualization tools, Appendix G integration of dose modeling and radiological surveys, and implementation of Scenario B, an alternative for the null hypothesis of statistical tests evaluated in a final status survey.

Updated guidance and information related to use of screening Appendix H values and resuspension factors. Cites NUREG/CR-5512, Volume 3 for screening values for additional radionuclides.

Streamlined guidance on consideration of buried radioactivity. Appendix J Added guidance on intrusion events and exposure scenarios for large substructures.

Streamlined guidance on site-specific exposure scenarios. Appendix M Updated guidance on regulatory standards.

Added guidance on use of composite sampling including Appendix O information on when it would and would not be appropriate to use, derivation of modified investigation levels, and methods to incorporate composite sampling into survey designs.

Added guidance on consideration of uncertainty in performance Appendix Q assessment analyses including issues associated with use of generic data sets, unrepresentative data, model integration, risk dilution, and lack of correlation of correlated parameters.

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TABLE OF CONTENTS ABSTRACT ................................................................................................................... iii FOREWORD ................................................................................................................... v LIST OF FIGURES........................................................................................................ xv LIST OF TABLES ....................................................................................................... xvii ACKNOWLEDGMENTS .............................................................................................. xix ABBREVIATIONS........................................................................................................ xxi GLOSSARY ................................................................................................................ xxv 1 PURPOSE, APPLICABILITY, AND ROADMAP ..................................................... 1-1 1.1 Purpose and Applicabilty of this Volume ........................................................................ 1-1 1.2 Roadmap to this Volume ................................................................................................ 1-2 1.3 Roadmap for Guidance on Restricted Use, Alternate Criteria, and Use of Engineered Barriers ....................................................................................................... 1-7 1.4 Iterative Nature of the Compliance Demonstration Process: A Decision-making Framework ..................................................................................................................... 1-7 1.4.1 Contents and General Concepts of the Iterative Approach in Using the Decision Framework ......................................................................................... 1-10 1.4.2 Steps of the Decision Framework..................................................................... 1-11 2 FLEXIBILITY IN DEMONSTRATING COMPLIANCE WITH 10 CFR PART 20, SUBPART E ............................................................................................................ 2-1 2.1 Risk-informed Approach to Compliance Demonstrations and Reviews ......................... 2-1 2.2 Flexibility in Submissions ............................................................................................... 2-4 2.3 Use of Characterization Data for Final Status Surveys .................................................. 2-6 2.4 Choice of Null Hypothesis for Final Status Survey Statistical Analysis .......................... 2-7 2.5 Demonstrating Compliance Using Dose Assessment Methods versus Derived Concentration Guideline Levels and Final Status Surveys ............................................ 2-9 2.5.1 Dose Modeling Approach ................................................................................. 2-10 2.5.2 DCGL and FSS Approach ................................................................................ 2-11 2.6 Merits of Screening versus Site-Specific Dose Assessments ...................................... 2-11 2.7 Sum of Fractions .......................................................................................................... 2-12 2.8 Flexibility for Use of Institutional Controls and Engineered Barriers at Restricted Use Sites ...................................................................................................................... 2-14 3 CROSS-CUTTING ISSUES ..................................................................................... 3-1 3.1 Transparency and Traceabiltiy of Compliance Demonstrations ..................................... 3-1 3.2 Data Quality Objectives Process .................................................................................... 3-2 3.3 Insignificant Radionuclides and Exposure Pathways ..................................................... 3-4 3.4 Considerations for Other Constraints on Allowable Residual Radioactivity ................... 3-6 3.5 Use of Engineered Barriers ........................................................................................... 3-7 xi

3.6 Surveying and Considering Risk Associated with Surface and Subsurface Soils .......... 3-9 4 FACILITY RADIATION SURVEYS ......................................................................... 4-1 4.1 Radiation Survey and Site Investigation Process .......................................................... 4-1 4.1.1 Historical Site Assessment ................................................................................. 4-1 4.1.2 Summary of Survey Types ................................................................................. 4-2 4.1.3 Areas of Review ................................................................................................. 4-4 4.1.4 Release Criteria .................................................................................................. 4-5 4.2 Scoping and Characterization Surveys .......................................................................... 4-6 4.2.1 Scoping Surveys ................................................................................................ 4-6 4.2.2 Characterization Surveys ................................................................................... 4-7 4.2.3 Areas of Review ................................................................................................. 4-9 4.3 Remedial Action Support Surveys ............................................................................... 4-11 4.3.1 Areas of Review ............................................................................................... 4-12 4.4 Final Status Survey Design .......................................................................................... 4-13 4.4.1 Areas of Review ............................................................................................... 4-13 4.5 Final Status Survey Report .......................................................................................... 4-16 4.5.1 Areas of Review ............................................................................................... 4-16 4.6 Issues not Covered in MARSSIM ................................................................................. 4-22 5 DOSE MODELING EVALUATIONS ........................................................................ 5-1 5.1 Introduction .................................................................................................................... 5-1 5.2 General Approach for Dose Modeling ............................................................................ 5-4 5.3 Information to be Submitted ........................................................................................... 5-8 5.3.1 Decommissioning Groups 1 - 3 (Unrestricted Release Using Screening Criteria) ............................................................................................................... 5-9 5.3.2 Decommissioning Groups 4 - 5 (Unrestricted Release Using Site-Specific Information) ........................................................................................................ 5-9 5.3.3 Decommissioning Group 6 (Restricted Release) ............................................. 5-10 5.3.4 Decommissioning Group 7 (Alternate Release Criteria)................................... 5-10 5.4 Acceptance Review ...................................................................................................... 5-10 5.4.1 Decommissioning Groups 1 - 3 (Unrestricted Release Using Screening Criteria) ............................................................................................................. 5-12 5.4.2 Decommissioning Groups 4 - 5 (Unrestricted Release Using Site-Specific Information) ...................................................................................................... 5-12 5.4.3 Decommissioning Group 6 (Restricted Release) ............................................. 5-13 5.4.4 Decommissioning Group 7 (Alternate Release Criteria)................................... 5-14 5.5 Safety Evaluation Criteria and Review ......................................................................... 5-14 5.5.1 Screening Safety Review Evaluation Criteria ................................................... 5-15 5.5.2 Evaluation Criteria for Decommissioning Groups 4 - 5 (Unrestricted Release Using Site-Specific Information) ......................................................... 5-18 5.5.3 Safety Evaluation Criteria for Decommissioning Group 6 (Restricted Release) ........................................................................................................... 5-25 5.5.4 Safety Evaluation Criteria - Decommissioning Group 7 (Alternate Release Criteria) .............................................................................................. 5-26 5.6 Summary Review Criteria ............................................................................................ 5-26 5.7 Additional Regulatory Guidance ................................................................................... 5-33 5.7.1 Regulatory Guidance Related to Decommissioning Groups 1 - 3 (Unrestricted Release Using Screening Criteria).............................................. 5-33 5.7.2 Regulatory Guidance Related to Decommissioning Groups 4 - 5 (Unrestricted Release Using Site-Specific Information) ................................... 5-33 xii

5.7.3 Regulatory Guidance Related to Decommissioning Group 6 (Restricted Release) ........................................................................................................... 5-33 6 ALARA ANALYSES ................................................................................................ 6-1 6.1 Safety Evaluation Review Procedures ........................................................................... 6-1 6.1.1 Areas of Review ................................................................................................. 6-1 6.1.2 Review Procedures ............................................................................................ 6-1 6.2 Acceptance Criteria ........................................................................................................ 6-2 6.2.1 Regulatory Requirements ................................................................................... 6-2 6.2.2 Regulatory Guidance .......................................................................................... 6-2 6.2.3 Information to be Submitted ............................................................................... 6-2 6.3 Evaluation Criteria .......................................................................................................... 6-2 6.3.1 Evaluation of Good Practice Efforts.................................................................... 6-2 6.3.2 Evaluation of Cost-Benefit ALARA Analyses...................................................... 6-2 6.3.3 When Cost-Benefit Analyses are Unnecessary.................................................. 6-3 6.3.4 Calculation of Benefits ........................................................................................ 6-3 6.3.5 Calculation of Costs ........................................................................................... 6-4 6.3.6 Compliance Methods at the Time of Decommissioning ..................................... 6-4 7 BIBLIOGRAPHY AND SUPERCEDED DOCUMENTS ........................................... 7-1 7.1 NRC Decommissioning Documents Referenced in the Main Body of Volume 2 ........... 7-1 7.2 Other NRC Documents Referenced in this Volume ....................................................... 7-3 7.3 Other References in this Volume ................................................................................... 7-4 7.4 Documents Superceded by this Volume ........................................................................ 7-5 IMPLEMENTING THE MARSSIM APPROACH FOR CONDUCTING FINAL RADIOLOGICAL SUREYS ............................................................. A-1 APPENDIX SIMPLE APPROACHES FOR CONDUCTING FINAL RADIOLOGICAL SURVEYS ................................................................................................... B-1 APPENDIX USE OF TWO-STAGE OR DOUBLE SAMPLING FOR FINAL STATUS SURVEYS ................................................................................................... C-1 APPENDIX SURVEY DATA QUALITY AND REPORTING ........................................... D-1 APPENDIX E MEASUREMENTS FOR FACILITY RADIATION SURVEYS ..................... E-1 APPENDIX F SURFACE WATER AND GROUNDWATER CHARACTERIZATION .........F-1 APPENDIX G SPECIAL ISSUES ASSOCIATED WITH DOSE MODELING, CHARACTERIZATION, AND SURVEY ...................................................... G-1 APPENDIX H CRITERIA FOR CONDUCTING SCREENING DOSE MODELING EVALUATIONS ........................................................................................... H-1 APPENDIX I TECHNICAL BASIS FOR SITE-SPECIFIC DOSE MODELING EVALUATIONS ............................................................................................. I-1 APPENDIX J ASSESSMENT STRATEGY FOR BURIED MATERIAL ............................. J-1 xiii

APPENDIX K DOSE MODELING FOR PARTIAL SITE RELEASE .................................. K-1 APPENDIX L WORKSHEET FOR IDENTIFYING POTENTIAL PATHWAYS FOR PARTIAL SITE RELEASE ...........................................................................L-1 APPENDIX M PROCESS FOR DEVELOPING EXPOSURE SCENARIOS USING SITE-SPECIFIC INFORMATION ................................................................ M-1 APPENDIX N ALARA ANALAYSES ................................................................................. N-1 APPENDIX O GUIDANCE FOR THE USE OF COMPOSITE SOIL SAMPLING FOR DEMONSTRATING COMPLIANCE WITH RADIOLOGICAL RELEASE CRITERIA.................................................................................................... O-1 APPENDIX P FRAMEWORK FOR USE OF ENGINEERED BARRIERS AT DECOMMISSIONING SITES ...................................................................... P-1 APPENDIX UNCERTAINTY IN PERFORMANCE AND DOSE ASSESSMENTS ......... Q-1 xiv

LIST OF FIGURES Figure 1.1 Continuum of Site Complexity. ........................................................................... 1-5 Figure 1.2 Decommissioning and License Termination Decision Framework (Modified from NUREG-1549) ............................................................................ 1-9 Figure D.1 MARLAP Road MapKey Terms and Processes ............................................ D-6 Figure D.2 Typical Components of the Radioanalytical Process ........................................ D-7 Figure F.1 General Monitoring Well Cross Section (Adapted from Figure 3

[NRC, 1994]) ......................................................................................................F-8 Figure G.1 Use of GIS and Geostatistical Tools ................................................................. G-9 Figure G.2 Importance of Contaminated Zone (orange color) Thickness and Depth to Various Pathways (Yu et al., 2021) ................................................................. G-10 Figure G.3 Sampling Strategies for Excavations. .............................................................. G-16 Figure I.1 Approach to Considering Multiple Elevated Areas or Hot Spots ....................... I-16 Figure I.2 Approach to Considering Vertical Heterogeneity............................................... I-17 Figure I.3 Conceptual Model Development ....................................................................... I-37 Figure I.4 Decommissioning Decision Framework ............................................................ I-38 Figure I.5 Activity Ratio of Vapor to Particulate as a Function of Contaminated Area (Used to Account for Missed H-3 and C-14 Inhalation Dose from Vapor Phase in DandD Residential Exposure Scenario) ............................................. I-44 Figure I.6 Source Depletion for Solubility and Kd Release Models (for Six Values of Kd Expressed in L/kg) ........................................................................................ I-47 Figure I.7 DandD Conceptual Model of the Groundwater and Surface Water Systems (from NUREG/CR-5621) ..................................................................... I-60 Figure I.8 Conceptualization Modeled by RESRAD-ONSITE (ANL, 1993) ....................... I-64 Figure I.9 RESRAD Near-Field Leaching and Transport Model (from RESRAD Training Workshop) ........................................................................................... I-64 Figure I.10 Treatment of Parameter Uncertainty in DandD Version 1 ................................. I-97 Figure I.11 Treatment of Parameter Uncertainty in DandD Version 2 ................................. I-98 Figure I.12 Application of Peak of the Mean Dose .......................................................... I-100 Figure J.1 Conceptual Illustration of Buried Material in Subsurface Soil at a Site............... J-2 Figure J.2 Conceptual Model for Remaining Basements which have Residual Radioactivity and are Backfilled with Clean Fill .................................................. J-3 Figure J.3 Simplified Conceptual Model of Human Disturbance into Buried Residual Radioactivity. ...................................................................................................... J-5 Figure J.4 Conceptual Model of Human Disturbance into Buried Residual Radioactivity ....................................................................................................... J-6 xv

Figure J.5 Conceptual Model of Buried Radioactivity with a Well Drilled through the Radioactive Material and the Drilling Spoils Spread out on the Surface ............ J-7 Figure J.6 Conceptual Model for Groundwater Exposure Pathway from Remaining Basements which have Residual Radioactivity and are Backfilled with Clean Fill ............................................................................................................ J-9 Figure J.7 Conceptual Model for Well-Driller Scenario...................................................... J-10 Figure J.8 Large-Scale Excavation Conceptual Model for Remaining Basement which has Residual Radioactivity and is Backfilled .......................................... J-11 Figure O.1 Pu-238 Sample Number Comparison.............................................................. O-13 Figure Q.1 Conceptual Overview of the Performance Assessment Process ...................... Q-2 Figure Q.2 Uncertainty Types Associated with Performance Assessment ......................... Q-4 Figure Q.3 Ten Years of Simulated Liquid Saturation Data ................................................ Q-8 Figure Q.4 Contour Plot of Liquid Saturation at a Given Point in Time ............................... Q-8 Figure Q.5 Kd Distribution with Values Plotted from Measurements at Three Different Hypothetical Sites ............................................................................................ Q-11 Figure Q.6 Influence of Different Amounts of Site-Specific Data on Performance Assessment Model Output .............................................................................. Q-12 Figure Q.7 Hypothetical Waste Disposal System Showing Representation of Different Submodels ...................................................................................................... Q-15 Figure Q.8 Model Response as a Function of Kd Using a Level 0 or Level 1 Abstraction ...................................................................................................... Q-16 xvi

LIST OF TABLES Table 1 Contents and Applicability of Key Decommissioning Guidance Documents ....... viii Table 2 Summary of Major Changes to Volume 2, Revision 2 .......................................... ix Table 1.1 Description and Examples of Each Decommissioning Group ............................ 1-3 Table 1.2 Applicability of Volume 2 to Decommissionoing Groups .................................... 1-4 Table 1.3 Cross-References for Restricted Use, Alternate Criteria, and Use of Engineered Barriers ............................................................................................. 1-8 Table 2.1 Comparison of FSS Statistical Test Scenarios ................................................... 2-8 Table 2.2 Comparison of Dose Modeling to DCGL and FSS Approaches to Compliance ...................................................................................................... 2-10 Table 2.3 Attributes of Screening and Site-Specific Analysis ........................................... 2-13 Table 4.1 Cross-References for Principle Steps in the Radiation Survey and Site Investigation Process ......................................................................................... 4-1 Table 4.2 Scope of MARSSIM ......................................................................................... 4-23 Table 5.1 Comparison and Description of Exposure Scenario Terms Used in this Guidance ............................................................................................................ 5-3 Table 5.2 Dose Modeling Review Criteria for Residual Radioactivity in Soils .................. 5-28 Table 5.3 Dose Modeling Review Criteria for Residual Radioactivity in Buildings ........... 5-31 Table 7.1 Documents Superseded by this Report ................................................................ 7-5 Table A.1 Suggested Survey Unit Areas ........................................................................... A-4 Table A.2 Scanning Coverage Fractions and Scanning Investigation Levels.................. A-11 Table A.3 Interpretation of Sample Measurements when a Reference Area is Used...... A-16 Table A.4 Interpretation of Sample Measurements when No Reference Area is Used ... A-17 Table C.1 Critical Points for Two-Stage Test of Normal Mean for a One-Sided Alternative ......................................................................................................... C-9 Table D.1 The Radioanalytical Data Life Cycle ................................................................. D-5 Table G.1 Calculation of 2 for the Example Data ............................................................. G-24 Table G.2 Analysis of Variance for Example Data.............................................................. G-25 Table G.3 Scenario B WRS and Quantile Tests for Class 2 Interior Drywall Survey Unit .... G-27 Table H.1 Acceptable License Termination Screening Values of Common Radionuclides for Building-Surface Contamination ........................................... H-6 Table H.2 Screening Values (pCi/g) of Common Radionuclides for Soil Surface Contamination Levels ........................................................................................ H-7 Table I.1 Summary of Source Abstraction Information Needs for Two Types of Dose-Modeling Approaches (Screening versus Site-Specific) ............................ I-7 Table I.2 Pathways for Generic Exposure Scenarios ....................................................... I-22 xvii

Table I.3 Potential Exposure Scenarios for Use in Dose Assessments ........................... I-25 Table I.4 Comparison of Area Factor Values from Different References ......................... I-32 Table I.5 Site Features and Conditions that may be Incompatible with those Assumed in DandD ........................................................................................... I-43 Table I.6 Issues to be Considered in Developing a Site-Specific Conceptual Model ....... I-45 Table I.7 Site Features and Conditions that may be Incompatible with the Assumptions Made in RESRAD-ONSITE.......................................................... I-49 Table I.8 Distribution Coefficient Model and Model Simplification Considerations ........... I-67 Table I.9 Four Cases Evaluated in RESRADs Nondispersion Model to Calculate Dilution .............................................................................................................. I-68 Table I.10 Summary Attributes and Information Needs of DandD; and RESRAD-ONSITE and RESRAD-BUILD Codes ............................................................................. I-73 Table I.11 Key DandD Deterministic Behavioral and Metabolic Parameters Mapped to RESRAD-ONSITE and RESRAD-BUILD Parameters................................... I-77 Table I.12 Distribution Coefficient Parameter and Uncertainty Evaluation Considerations .................................................................................................. I-89 Table I.13 Maximum Entropy Probability Distributions ..................................................... I-102 Table L.1 Example of Summary Format ........................................................................... L-12 Table M.1 Effects of Salinity of Drinking Water on Livestock ........................................... M-11 Table N.1 Possible Benefits and Costs Related to Decommissioning ............................... N-8 Table N.2 Parameter Values Used in ALARA Examples ................................................. N-14 Table O.1 Composite Sampling Overview ......................................................................... O-3 Table O.2 DCGLEMC for Pu-238........................................................................................ O-12 Table O.3 Composite Result Investigation Level Evaluation............................................ O-14 Table P.1 Examples of Surviving Native American Mounds .............................................. P-9 Table P.2 Summary of the Graded Approach for the Design of Erosion Protection Systems ........................................................................................................... P-27 Table P.3 Examples of Engineered Barriers Used to Solve Erosion Problems ............... P-29 Table P.4 Summary of Existing Key Documents Related to Engineered Barriers ........... P-36 Table Q.1 Example of Alternative Conceptual Models in a Performance Assessment .... Q-13 xviii

ACKNOWLEDGMENTS The preparers would like to thank the following individuals for assisting in the development and review of this revision to the guidance document. These individuals provided valuable contributions, insights, and recommendations. A special thanks to Lifeng Guo for his contributions to Appendix F, Surface Water and Groundwater Characterization; Mark Fuhrmann and Hans Arlt for their contributions to various sections of the document including guidance on the use of engineered barriers in Section 3.5 and Appendix P; and Allen Gross for his contributions to Appendix M, Process for Developing Exposure Scenarios using Site Specific Information. The preparers would also like to acknowledge the contribution of Michael Kunowski for sharing his valuable NRC regional inspector experience, which greatly enhanced the quality of the guidance document; and Fred Schofer for his expert assistance with revisions to Appendix N on ALARA criteria.

Comments received from the States of California, Pennsylvania, Texas and New Jersey provided valuable Agreement State perspectives and greatly improved the clarity and quality of the document. The preparers would also like to recognize the efforts of Tim Vitkus and Nick Altic of Oak Ridge Associated Universities in developing the composite sampling guidance found in Appendix O and addressing technical comments on the guidance to ensure that it could be used by NRC licensees with minimal effort. Finally, we would like to extend a special thanks to Donald Lowman for his early project management of the draft document; and Marcia Pringle, Pamela Menefee-Buzdygon, and Sarah Achten for technical editing and formatting, which was essential to publishing this document.

Revision 2 Experts and Peer Reviewers H. Arlt, M. Fuhrmann, A. Gross, L. Guo, F. Schofer, and M. Kunowski Revision 1 Preparers (not including preparers of Revision 2)

U.S. Nuclear Regulatory Commission Staff K.L. Banovac, J.T. Buckley, R.L. Johnson, J.J. Kottan, T.G. McLaughlin, and S. Schneider Agreement States California, New Jersey, Pennsylvania, and Texas xix

ABBREVIATIONS ACAP Alternative Cover Assessment Program ADAMS Agencywide Documents Access and Management System AEA Atomic Energy Act of 1954, as amended ALARA as low as is reasonably achievable ALCD Alternative Landfill Cover Demonstration Am americium ANSI American National Standards Institute ASME American Society of Mechanical Engineers ASR alkali-silica reaction ASTM American Society for Testing and Materials Bq becquerel Bq/kg becquerel per kilogram Ca calcium CDE Common Data Environment CERCLA Comprehensive Environmental Response, Compensation, and Liability Act CFR Code of Federal Regulations Ci curie CNWRA Center for Nuclear Waste Regulatory Analyses Co cobalt CO2 carbon dioxide cpm counts per minute Cs cesium C-S-H calcium silicate hydrate CSM conceptual site model DandD Decontamination and Decommissioning (software package)

DCGL derived concentration guideline level DCGLEMC derived concentration guideline level (elevated measurement comparison)

DCGLW derived concentration guideline level (wide-area [over the entire survey unit])

DOE U.S. Department of Energy DP decommissioning plan dpm disintegrations per minute DQA data quality assessment DQOs data quality objectives xxi

DQO process data quality objectives process Eh redox potential EMC elevated measurement comparison EML (U.S. Department of Energy) Environmental Measurements Laboratory (formerly the Health and Safety Laboratory)

EPA U.S. Environmental Protection Agency Fe iron FEP features, events, and processes FR Federal Register FSS final status survey FSSP final status survey plan FSSR final status survey report GCL geosynthetic clay liner GIS geographic information system H hydrogen H-3 tritium HDPE high-density polyethylene HSA historical site assessment HTD hard-to-detect IAEA International Atomic Energy Agency ICRP International Commission on Radiological Protection IL investigation level ISO International Organization for Standardization ITRC Interstate Technology and Regulatory Council Kd distribution coefficient kg kilogram km kilometer L liter LBGR lower bound of the gray region L/kg liter/kilgoram LLW low-level waste LTP license termination plan LTR license termination rule xxii

MARLAP Multi-Agency Radiological Laboratory Analytical Protocols Manual (NUREG-1576)

MARSSIM Multi-Agency Radiological Survey and Site Investigation Manual (NUREG-1575)

MCL maximum contaminant level MDC minimum detectable concentration MDCscan scan minimum detectable concentration MDCstatic static minimum detectable concentration Mg magnesium mg milligram mil unit of corrosion (1 mil=0.001 inch)

MIL modified investigation level mrem millirem mSv millisievert Na sodium NAS National Academy of Sciences NCRP National Council on Radiation Protection and Measurements NEI Nuclear Energy Institute NEPA National Environmental Policy Act Ni nickel NIST National Institute of Standards and Technology NMSS Office of Nuclear Material Safety and Safeguards (U.S. Nuclear Regulatory Commission)

NOAA National Oceanic and Atmospheric Administration NORM naturally occurring radioactive material Np neptunium NRC U.S. Nuclear Regulatory Commission NUREG NRC technical report designation Pb lead pCi picocurie pCi/g picocuries per gram PDF probability density function pH hydrogen ion concentration (negative of the log of the hydrogen ion molar concentration)

Po polonium PSR partial site release Pu plutonium PVC polyvinyl chloride QA quality assurance xxiii

QAPP quality assurance project plan QC quality control Ra radium RCRA Resource Conservation and Recovery Act REMP Radiological Environmental Monitoring Program RESRAD RESidual RADdioactive materials family of computer codes RFR (also Rfo) resuspension factor ROC radionuclide(s) of concern RSS residual sum of squares RSSI radiation survey and site investigation (process)

SADA Spatial Analysis and Decision Assistance (computer code)

SDMP Site Decommissioning Management Plan Sr strontium SRM staff requirements memorandum SRP NMSS Decommissioning Standard Review Plan (NUREG-1727)

SSC system, stucture, and component Sv sievert Tc technetium TEDE total effective dose equivalent Th thorium U uranium UF6 uranium hexafluoride UMTRA Uranium Mill Tailings Remedial Action UMTRCA Uranium Mill Tailings Radiation Control Act UO2 uranium dioxide U.S.C. U.S. Code USDA U.S. Department of Agriculture USGS U.S. Geological Survey VSP Visual Sample Plan WRS Wilcoxon Rank Sum (test) y year xxiv

GLOSSARY The following terms are defined for the purposes of this volume of the NUREG report.

Acceptance Review. The evaluation the NRC staff performs upon receipt of a license amendment request to determine if the information provided in the document is sufficient to begin the technical review.

Activity. The rate of disintegration (transformation) or decay of radioactive material. The units of activity are the curie (Ci) and the becquerel (Bq) (see Title 10 of the Code of Federal Regulations (10 CFR) 20.1003, Definitions).

Affected Parties. Representatives of a broad cross section of individuals and institutions in the community or vicinity of a site that may be affected by the decommissioning of the site.

ALARA. The acronym for as low as is reasonably achievable, which means making every reasonable effort to maintain exposures to radiation as far below the dose limits as is practical, consistent with the purpose for which the licensed activity is undertaken, and taking into account the state of technology, the economics of improvements in relation to the state of technology, the economics of improvements in relation to the benefits to public health and safety, and other societal and socioeconomic considerations, and in relation to the use of nuclear energy and licensed materials in the public interest (see 10 CFR 20.1003).

Alternate Criteria. Dose criteria for residual radioactivity that are greater than certain dose criteria described in 10 CFR 20.1402, Radiological Criteria for Unrestricted Use, and 10 CFR 20.1403, Criteria for License Termination under Restricted Conditions, as allowed in 10 CFR 20.1404, Alternate Criteria for License Termination. Alternate criteria must be approved by the Commission.

Aquifer. A geologic formation, a group of formations, or part of a formation capable of yielding a significant amount of groundwater to wells or springs.

Background Radiation. Radiation from cosmic sources, naturally occurring radioactive material, including radon (except as a decay product of source or special nuclear material) and global fallout as it exists in the environment from the testing of nuclear explosive devices or from past nuclear accidents, such as Chernobyl, that contribute to background radiation and are not under the control of the licensee. Background radiation does not include radiation from source, byproduct, or special nuclear materials regulated by the NRC (see 10 CFR 20.1003).

Broad-Scope Licenses. A type of specific license authorizing receipt, acquisition, ownership, possession, use, and transfer of any chemical or physical form of the byproduct material specified in the license but not exceeding quantities specified in the license. Relevant requirements are found in 10 CFR Part 33, Specific Domestic Licenses of Broad Scope for Byproduct Material. Examples of broad scope licensees are large universities and large research and development facilities.

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Byproduct Material.

1. Any radioactive material (except special nuclear material) yielded in, or made radioactive by, exposure to the radiation incident to the process of producing or using special nuclear material.
2. The tailings or wastes produced by the extraction or concentration of uranium or thorium from ore processed primarily for its source material content, including discrete surface wastes resulting from uranium solution extraction processes. Underground ore bodies depleted by these solution extraction operations do not constitute byproduct material within this definition.
3. (i) Any discrete source of radium (Ra)-226 that is produced, extracted, or converted after extraction, before, on, or after August 8, 2005, for use for a commercial, medical, or research activity; or (ii) Any material that
a. Has been made radioactive by use of a particle accelerator, and
b. Is produced, extracted, or converted after extraction, before, on, or after August 8, 2005, for use for a commercial, medical, or research activity.
4. Any discrete source of naturally occurring radioactive material, other than source material, that (i) The Commission, in consultation with the Administrator of the Environmental Protection Agency, the Secretary of Energy, the Secretary of Homeland Security, and the head of any other appropriate Federal agency, determines would pose a threat similar to the threat posed by a discrete source of Ra-226 to public health and safety or the common defense and security and (ii) Before, on, or after August 8, 2005, is extracted or converted after extraction for use in a commercial, medical, or research activity (see 10 CFR 20.1003).

Categorical Exclusion. A category of regulatory actions that do not individually or cumulatively have a significant effect on the human environment and that the Commission has found to have no such effect, in accordance with procedures set out in 10 CFR 51.22, Criterion for Categorical Exclusion; Identification of Licensing and Regulatory Actions Eligible for Categorical Exclusion or Otherwise Not Requiring Environmental Review, and for which, therefore, neither an environmental assessment nor an environmental impact statement is required (see 10 CFR 51.14(a)).

Characterization Survey. A type of survey that includes facility or site sampling, monitoring, and analysis activities to determine the extent and nature of residual radioactivity. Characterization surveys provide the basis for acquiring necessary technical information to develop, analyze, and select appropriate cleanup techniques.

Cleanup. See Decontamination.

Closeout Inspection. An inspection performed by the NRC, or its contractor, to determine if a licensee has adequately decommissioned its facility. Typically, a closeout inspection is performed after the licensee has provided its demonstration that its facility is suitable for release in accordance with NRC requirements.

Confirmatory Survey. A survey conducted by the NRC, or its contractor, to verify the results of the licensees final status survey. Typically, confirmatory surveys consist of measurements at a fraction of the locations previously surveyed by the licensee, to determine whether the licensees results are valid and reproducible.

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Critical Group. The group of individuals reasonably expected to receive the greatest exposure to residual radioactivity for any applicable set of circumstances (see 10 CFR 20.1003).

DandD Code. The Decontamination and Decommissioning (DandD) software package, developed by the NRC, that addresses compliance with the dose criteria of 10 CFR Part 20, Standards for Protection against Radiation, Subpart E, Radiological Criteria for License Termination. Specifically, DandD embodies the NRCs guidance on screening dose assessments to allow licensees to perform simple estimates of the annual dose from residual radioactivity in soils and on building surfaces. The current version of the code is 2.4, as of publication of this NUREG.

Data Quality Objectives (DQOs). Qualitative and quantitative statements derived from the Data Quality Objectives process that clarify study technical and quality objectives, define the appropriate type of data, and specify tolerable levels of potential decision errors that will be used as the basis for establishing the quality and quantity of data needed to support decisions.

Data Quality Objectives Process (DQO process): A series of logical steps that guides managers or staff to plan for the resource-effective acquisition of environmental data. See also data quality objectives.

Decommission. To remove a facility or site safely from service and reduce residual radioactivity to a level that permits (1) release of the property for unrestricted use and termination of the license or (2) release of the property under restricted conditions and termination of the license (see 10 CFR 20.1003).

Decommissioning Groups. For the purposes of this guidance document, the categories of decommissioning activities that depend on the type of operation and the residual radioactivity.

Decommissioning Plan (DP). A detailed description of the activities that the licensee intends to use to assess the radiological status of its facility, to remove radioactivity attributable to licensed operations at its facility to levels that permit release of the site in accordance with the NRCs regulations and termination of the license, and to demonstrate that the facility meets the NRCs requirements for release. A DP typically consists of several interrelated components, including (1) site characterization information, (2) a remediation plan that has several components, including a description of remediation tasks, a health and safety plan, and a quality assurance (QA) plan, (3) site-specific cost estimates for the decommissioning, and (4) a final status survey plan (see 10 CFR 30.36(g)(4)).

Decontamination. The removal of undesired residual radioactivity from facilities, soils, or equipment before the release of a site or facility and termination of a license; also known as remediation, remedial action, and cleanup.

Deterministic Analyses. Refers to analyses that use a single set of parameter values to demonstrate compliance. Results of the calculations are typically reported as a single value (e.g., best estimate or conservative value). Additional information about demonstrating compliance with release criteria using deterministic analyses can be found in Chapters 2 and 5, and Appendix I of Volume 2.

Derived Concentration Guideline Levels (DCGLs). Radionuclide-specific concentration limits used by the licensee during decommissioning to achieve the regulatory dose standard that permits the release of the property and termination of the license. The DCGL applicable to the xxvii

average concentration over a survey unit is called the DCGLW. The DCGL applicable to limited areas of elevated concentrations within a survey unit is called the DCGL EMC (elevated measurement comparison).

Discrete Source. A radionuclide that has been processed so that its concentration within a material has been purposely increased for use for commercial, medical, or research activities (see 10 CFR 20.1003).

Distribution Coefficient or Kd. Ratio of the concentration of an element or chemical associated with the soil to the concentration in the surrounding aqueous solution when the system is at equilibrium. The units are typically expressed in liters per kilogram (L/kg). While there are other types of distribution or partition coefficients such as Henrys law constant (KH) for air/water or the octanol/water coefficient (Kow), only the distribution coefficient or Kd is referenced in this volume.

Dose (or Radiation Dose). A generic term that means absorbed dose, dose equivalent, effective dose equivalent, committed dose equivalent, committed effective dose equivalent, or total effective dose equivalent, as defined in 10 CFR 20.1003. In this report, dose generally refers to total effective dose equivalent (TEDE).

Durable Institutional Controls. Durable institutional controls are reliable and sustainable for the time period needed. An institutional control that involves government ownership or control of the site would be considered a durable institutional control.

Effluent. Material discharged into the environment from licensed operations.

Environmental Assessment. A concise public document for which the Commission is responsible that serves to (1) briefly provide sufficient evidence and analysis for determining whether to prepare an environmental impact statement or a finding of no significant impact, (2) aid the Commissions compliance with the National Environmental Policy Act (NEPA) when no environmental impact statement is necessary, and (3) facilitate preparation of an environmental impact statement when one is necessary (see 10 CFR 51.14(a)).

Environmental Impact Statement. A detailed written document that ensures the NEPA policies and goals are considered in the actions of the Federal government. It discusses significant impacts and reasonable alternatives to the proposed action.

Environmental Monitoring. The process of sampling and analyzing environmental media in and around a facility (1) to confirm compliance with performance objectives and (2) to detect radioactive material entering the environment to facilitate timely remedial action.

Environmental Report. A document submitted to the NRC by an applicant for a license amendment request (see 10 CFR 51.14(a)). The NRC staff uses the environmental report to prepare environmental assessments and environmental impact statements. The requirements for environmental reports are specified in 10 CFR 51.45-69.

Exposure Pathway. The mechanism by which radioactive material is transferred from the environment to the receptor. Three commonly recognized exposure pathways are inhalation, ingestion, and direct radiation. The combination of individual exposure pathways makes up the site-specific exposure scenario.

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Exposure Scenario. A description of the potential future land uses, human activities, and transport of radioactivity in the natural system as it influences a future receptors interaction with (and therefore exposure to) residual radioactivity. In particular, the exposure scenario describes where humans may be exposed to residual radioactivity in the environment, what exposure group habits determine exposure, and how residual radioactivity moves through the environment.

External Dose. That portion of the dose equivalent received from radiation sources outside the body (see 10 CFR 20.1003).

Final Status Survey (FSS). Measurements and sampling to describe the radiological conditions of a site or facility, following completion of decontamination activities (if any) and in preparation for release of the site or facility.

Final Status Survey Plan (FSSP). The description of the final status survey design.

Final Status Survey Report (FSSR). The results of the final status survey conducted by a licensee to demonstrate the radiological status of its facility. The FSSR is submitted to the NRC for review and approval.

Financial Assurance. A guarantee or other financial arrangement provided by a licensee that funds for decommissioning will be available when needed. This is in addition to the licensee's regulatory obligation to decommission its facilities.

Financial Assurance Mechanism. Financial instruments used to provide financial assurance for decommissioning.

Floodplain. The lowland and relatively flat areas adjoining inland and coastal waters, including flood-prone areas of offshore islands (see 10 CFR 72.3, Definitions).

Footprint. The portion of a site undergoing decommissioning, which comprises all the areas of soil containing residual radioactivity, where intentional mixing is proposed to meet the release criteria.

General Licenses. Licenses that are effective without filing applications with the NRC or the issuance of licensing documents to particular persons. The requirements for general licenses are found in Title 10 of the CFR. Examples of items for which general licenses are issued include tritium exit signs and anti-static devices.

Groundwater. Water contained in pores or fractures in either the unsaturated or saturated zones below ground level.

Historical Site Assessment (HSA). The identification of potential, likely, or known sources of radioactive material and radioactive contamination based on existing or derived information for the purpose of classifying a facility or site, or parts thereof, as impacted or nonimpacted (see 10 CFR 50.2, Definitions).

Hydraulic Conductivity. The volume of water that will move through a medium in a unit of time under a unit hydraulic gradient through a unit area measured perpendicular to the direction of flow.

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Hydrology. Study of the properties, distribution, and circulation of water on the surface of the land, in the soil and underlying rocks, and in the atmosphere.

Impact. The positive or negative effect of an action (past, present, or future) on the natural environment (land use, air quality, water resources, geological resources, ecological resources, aesthetic and scenic resources) and the human environment (infrastructure, economics, social, and cultural).

Impacted Areas. The areas with some reasonable potential for residual radioactivity in excess of natural background or fallout levels.

Inactive Outdoor Area. The outdoor portion of a site not used for licensed activities or materials for 24 months or more.

Infiltration. The process of water entering the soil at the ground surface. Infiltration becomes percolation when water has moved below the depth at which it can be removed (to return to the atmosphere) by evaporation or transpiration.

Insignificant Radionuclides and Pathways. Radionuclides and pathways that can be excluded from further detailed consideration, because they cumulatively contribute no more than 10 percent of the dose standard (e.g., for unrestricted release, no more than 0.025 mSv/y or 2.5 mrem/y). The dose contributions of the insignificant radionuclides and pathways should still be considered in demonstrating compliance with release criteria. See Chapter 3, Section 3.3 for additional information on insignificant radionuclides and pathways.

In Situ Recovery (ISR). In situ recovery (ISR) is one of the two primary extraction methods that are currently used to obtain uranium from underground. ISR facilities recover uranium from low-grade ores where other mining and milling methods may be too expensive or environmentally disruptive. The ISR process is as follows:

(1) A solution called lixiviant (typically containing water mixed with oxygen and/or hydrogen peroxide, as well as sodium carbonate or carbon dioxide) is injected through a series of wells into the ore body to dissolve the uranium.

(2) The lixiviant is then collected in a series of recovery wells, through which it is pumped to a processing plant, where the uranium is extracted from the solution through an ion-exchange process.

(3) The uranium extract is then further purified, concentrated, and dried to produce a material, which is called "yellowcake" because of its yellowish color.

(4) Finally, the yellowcake is packed in 55-gallon drums to be transported to a uranium conversion facility, where it is processed through the stages of the nuclear fuel cycle to produce fuel for use in nuclear power reactors.

Institutional Controls. Measures to control access to a site and minimize disturbances to engineered measures established by the licensee to control the residual radioactivity.

Institutional controls include administrative mechanisms (e.g., land use restrictions) and may include, but are not limited to, physical controls (e.g., signs, markers, landscaping, and fences).

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Karst. A type of topography that is formed over limestone, dolomite, or gypsum by dissolution, characterized by sinkholes, caves, and underground drainage.

Leak Test. A test for leakage of radioactivity from sealed radioactive sources. These tests are made when the sealed source is received and on a regular schedule thereafter. The frequency is usually specified in the sealed source and device registration certificate or license.

License Termination Plan (LTP). A detailed description of the activities a reactor licensee intends to use to assess the radiological status of its facility, to remove radioactivity attributable to licensed operations at its facility to levels that permit release of the site in accordance with the NRCs regulations and termination of the license, and to demonstrate that the facility meets the NRCs requirements for release. An LTP consists of several interrelated components, including (1) a site characterization, (2) identification of remaining dismantlement activities, (3) plans for site remediation, (4) detailed plans for the final radiation survey, (5) a description of the end use of the facility, if restricted, (6) an updated site-specific estimate of remaining decommissioning costs, and (7) a supplement to the environmental report, pursuant to 10 CFR 51.33, Draft Finding of No Significant Impact; Distribution, describing any new information or significant environmental change associated with the licensees proposed termination activities (see 10 CFR 50.82, Termination of License).

License Termination Rule (LTR). The License Termination Rule refers to the final rule on Radiological Criteria for License Termination, published by the NRC as Subpart E to 10 CFR Part 20 on July 21, 1997 (62 FR 39058).

Licensee. A person who possesses a license1, or a person who possesses licensable material, whom the NRC could require to obtain a license.

MARSSIM. NUREG-1575, Multi-Agency Radiation Site Survey and Investigation Manual, is a multiagency consensus manual that provides information on planning, conducting, evaluating, and documenting building surface and surface soil final status radiological surveys for demonstrating compliance with dose- or risk-based regulations or standards.

Model. A simplified representation of an object or natural phenomenon. The model can be in many possible forms, such as a set of equations or a physical, miniature version of an object or system constructed to allow estimates of the behavior of the actual object or phenomenon when the values of certain variables are changed. Important environmental models include those estimating the transport, dispersion, and fate of chemicals in the environment.

Monitoring. Monitoring (radiation monitoring, radiation protection monitoring) is the measurement of radiation levels, concentrations, or quantities of radioactive material and the use of the results of these measurements to evaluate potential exposures and doses (see 10 CFR 20.1003).

mrem/y (millirem per year). One one-thousandth (0.001) of a rem per year. (See also Sievert.)

National Environmental Policy Act (NEPA). The National Environmental Policy Act of 1969, as amended, which requires Federal agencies, as part of their decision-making process, to consider the environmental impacts of actions under their jurisdiction. Both the Council on 1 A license issued under the regulations in 10 CFR Parts 30 through 36, 39, 40, 50, 60, 61, 63, 70, or 72 (see definition in 10 CFR 20.1003).

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Environmental Quality and the NRC have regulations to implement NEPA requirements. The Councils regulations are contained in 40 CFR Parts 1500 to 1508, and NRC requirements are provided in 10 CFR Part 51, Environmental Protection Regulations for Domestic Licensing and Related Regulatory Functions.

Naturally Occurring Radioactive Material (NORM). The natural radioactivity in rocks, soils, air, and water. NORM generally refers to materials in which the radionuclide concentrations have not been enhanced by or are as a result of human practices. NORM does not include uranium or thorium in source material.

Non-impacted Areas. The areas with no reasonable potential for residual radioactivity in excess of background radiation levels.

Pathway. See Exposure Pathway.

Performance-Based Approach. Regulatory decision-making that relies upon measurable or calculable outcomes (i.e., performance results) to be met but provides more flexibility to the licensee as to the means of meeting those outcomes.

Permeability. The ability of a material to transmit fluid through its pores when subjected to a difference in head (pressure gradient). Permeability depends on the substance transmitted (e.g., oil, air, water) and on the size and shape of the pores, joints, and fractures in the medium and the manner in which they are interconnected.

Porosity. The ratio of openings, or voids, to the total volume of a soil or rock, expressed as a decimal fraction or as a percentage.

Potentiometric Surface. The two-dimensional surface that describes the elevation of the water table. In an unconfined aquifer, the potentiometric surface is at the top of the water level. In a confined aquifer, the potentiometric surface is above the top of the water level, because the water is under confining pressure.

Principal Activities. Activities authorized by the license that are essential to achieving the purpose(s) for which the license was issued or amended. Storage during which no licensed material is accessed for use or disposal and activities incidental to decontamination or decommissioning are not principal activities.

Probabilistic Analyses. Refers to analyses that use a random sampling method to select parameter values from a distribution. Results of the calculations are also in the form of a distribution of values. Additional information about demonstrating compliance with release criteria using probabilistic analyses can be found in Chapters 2 and 5, and Appendix I of Volume 2.

Quality Assurance Project Plan (QAPP). A planning document that provides comprehensive details regarding the necessary quality assurance and quality control and other technical activities that must be implemented to ensure that the results of the work performed will satisfy stated performance criteria. See Appendix D of this volume for additional information on QAPPs.

Reasonable Alternatives. Those alternatives that are practical or feasible from a technical and economic standpoint.

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Reasonably Foreseeable Land Use. Land use exposure scenarios that are likely within 100 years, considering advice from land use planners and stakeholders on land use plans and trends.

Receptor. The hypothetical exposed individual for whom the dose received is being assessed.

Typically, this would be the average member of the critical group.

rem. The special unit of any of the quantities expressed as dose equivalent. The dose equivalent in rems is equal to the absorbed dose in rads multiplied by the quality factor (1 rem = 0.01 sievert) (see 10 CFR 20.1004, Units of Radiation Dose).

Remedial Action. See Decontamination.

Remediation. See Decontamination.

Residual Radioactivity. Radioactivity in structures, materials, soils, groundwater, and other media at a site resulting from activities under the licensees control. This includes radioactivity from all licensed and unlicensed sources used by the licensee but excludes background radiation. It also includes radioactive materials remaining at the site as a result of routine or accidental releases of radioactive material at the site and previous burials at the site, even if those burials were made in accordance with the provisions of 10 CFR Part 20 (see 10 CFR 20.1003).

RESRAD-BUILD Code. A computer code developed by the U.S. Department of Energy and designed to estimate radiation doses and risks from RESidual RADioactive materials in BUILDings.

RESRAD Family of Codes. A family of computer codes developed by the U.S. Department of Energy and designed to estimate radiation doses and risks from RESidual RADioactive materials on building surfaces and in both onsite and offsite soils.

RESRAD-OFFSITE Code. A computer code developed by the U.S Department of Energy that extends the capabilities of the RESRAD-ONSITE computer code to estimate the radiological consequences to a receptor located either onsite or outside the area of primary contamination.

RESRAD-ONSITE Code. A computer code developed by the U.S. Department of Energy and designed to estimate radiation doses and risks from RESidual RADioactive materials in soils.

Restricted Area. Any area to which access is limited by a licensee for the purpose of protecting individuals from undue risks from exposure to radiation and radioactive materials (see 10 CFR 20.1003).

Risk. Defined by the risk triplet of a scenario (a combination of events or conditions that could occur) or set of scenarios, the probability that the scenario could occur, and the consequence (e.g., dose to an individual) if the scenario were to occur.

Risk-Based Approach. Regulatory decision-making that is based solely on the numerical results of a risk assessment. (Note that the Commission does not endorse a risk-based regulatory approach.)

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Risk-Informed Approach. Regulatory decision-making that represents a philosophy whereby risk insights are considered together with other factors to establish requirements that better focus licensee and regulatory attention on design and operational issues commensurate with their importance to public health and safety.

Risk Insights. Results and findings that come from risk assessments.

Robust Engineered Barrier. A man-made structure that is designed to mitigate the effect of natural processes or human uses that may initiate or accelerate the release of residual radioactivity through environmental pathways. The structure is designed so that the radiological criteria for license termination (10 CFR Part 20, Subpart E) are met. Robust engineered barriers are designed to be more substantial, reliable, and sustainable for the time period needed without reliance on active ongoing maintenance.

Safety Evaluation Report. The NRC staffs evaluation of the licensees proposed action to determine if that action can be accomplished safely.

Saturated Zone. That part of the earths crust beneath the regional water table in which all voids, large and small, are filled with water under pressure greater than atmospheric.

Scoping Survey. A type of survey that is conducted to identify (1) radionuclide contaminants, (2) relative radionuclide ratios, and (3) general levels and extent of residual radioactivity.

Scenario. As specified in draft low-level waste guidance (NUREG-2175, Guidance for Conducting Technical Analyses for 10 CFR Part 61, Draft Report for Comment, issued March 2015), the term scenario refers to the expected (central scenario) or potential (alternative scenario) future dynamic evolution of the disposal site, which might include consideration of disruptive events (e.g., gully erosion, climate change). Typically, detailed consideration of the future evolution of a decommissioning site is unnecessary. However, central and alternative scenarios may need to be considered to demonstrate compliance with the radiological criteria for license termination for some complex decommissioning sites, or sites with relatively long-lived radionuclides (half-lives comparable to or longer than the compliance period). In general, the term scenario pertains to the exposure scenario when used in NUREG-1757, Volume 2, as defined above.

Screening Analysis/Approach/Methodology/Process. The use of (1) predetermined building surface concentration and surface soil concentration values, or (2) a predetermined methodology (e.g., use of the DandD code) that meets the radiological decommissioning criteria without further analysis, to simplify decommissioning in cases where low levels of residual radioactivity are achievable.

Sealed Source. Any special nuclear material or byproduct material encased in a capsule designed to prevent leakage or escape of the material.

sievert (Sv). The SI unit of any of the quantities expressed as dose equivalent. The dose equivalent in sieverts is equal to the absorbed dose in grays multiplied by the quality factor (1 sievert = 100 rem) (see 10 CFR 20.1004).

Site. The area of land, along with structures and other facilities, as described in the original NRC license application, plus any property outside the originally licensed boundary added for the purpose of receiving, possessing, or using radioactive material at any time during the term xxxiv

of the license, as well as any property where radioactive material was used or possessed that has been released before license termination.

Site Characterization. Studies that enable the licensee to sufficiently describe the conditions of the site, separate building, or outdoor area to evaluate the acceptability of the DP.

Site Characterization Survey. See Characterization Survey.

Site Decommissioning Management Plan (SDMP). The program established by the NRC in March 1990 to help ensure the timely cleanup of sites with limited progress in completing the remediation of the site and the termination of the facility license. SDMP sites typically have buildings, former waste disposal areas, large volumes of tailings, groundwater contamination, and soil contaminated with low levels of uranium or thorium or other radionuclides. The SDMP program was discontinued (69 FR 33946), and since then all decommissioning sites have been managed under a comprehensive decommissioning program.

Site-Specific Dose Analysis. Any dose analysis that is performed other than by using the default screening tools.

Smear. A radiation survey technique that is used to determine levels of removable surface contamination. A medium (typically filter paper) is rubbed over a surface (typically of an area 100 square centimeters), followed by a quantification of the activity on the medium. Also known as a swipe.

Source Material. Uranium or thorium, or any combination of uranium and thorium, in any physical or chemical form, or ores that contain by weight one-twentieth of one percent (0.05 percent) or more of uranium, thorium, or any combination of uranium and thorium. Source material does not include special nuclear material (see 10 CFR 20.1003).

Source Term. Release rates associated with residual radioactivity at a site or facility. The source term is related to the inventory, distribution of contamination, and controlling release mechanisms (e.g., solubility-controlled release, diffusion-limited release, or desorption). Note that the working definition of source term in this volume is slightly different than the definition of source term found in the NRC glossary. The definition found in the online NRC glossary, at https://www.nrc.gov/reading-rm/basic-ref/glossary/source-term.html, is specific to accidents involving radioactive materials: types and amounts of radioactive or hazardous material released to the environment following an accident.

Special Nuclear Material. (1) Plutonium, uranium (U)-233, uranium enriched in the isotope 233 or in the isotope 235, and any other material that the Commission, pursuant to the provisions of Section 51 of the Atomic Energy Act of 1954, as amended (AEA), determines to be special nuclear material, but does not include source material; or (2) any material artificially enriched by any of the foregoing but does not include source material (see 10 CFR 20.1003).

Specific Licenses. Licenses issued to a named person who has filed an application for the license under the provisions of 10 CFR Parts 30, 32-36, 39, 40, 61, 70, and 72. Examples of specific licenses are industrial radiography, medical use, irradiators, and well logging.

Surface Soil. The top layer of soil on a site that supports certain exposure pathways such as direct exposure, soil ingestion, and resuspension of particles for inhalation. Surface soil has also been associated with the thickness of soil that can be measured using direct measurement xxxv

or scanning techniques. Typically, this layer is often represented as the top 15 centimeters (6 inches) of soil but will vary depending on the radionuclide, surface characteristics, measurement technique, and dose modeling assumptions.

Survey. An evaluation of the radiological conditions and potential hazards incident to the production, use, transfer, release, disposal, or presence of radioactive material or other sources of radiation. When appropriate, such an evaluation includes a physical survey of the location of radioactive material and measurements or calculations of levels of radiation, or concentrations or quantities of radioactive material present (see 10 CFR 20.1003).

Survey Unit. A geographical area consisting of structures or land areas of specified size and shape at a site for which a separate decision will be made as to whether the unit attains the site-specific reference-based cleanup standard for the designated pollution parameter. Survey units are generally formed by grouping contiguous site areas with similar use histories and having the same contamination potential (classification). Survey units are established to facilitate the survey process and the statistical analysis of survey data.

Timeliness. Specific time periods stated in NRC regulations for decommissioning unused portions of operating nuclear materials facilities and for decommissioning the entire site upon termination of operations.

Total Effective Dose Equivalent (TEDE). The sum of the effective dose equivalent (for external exposures) and the committed effective dose equivalent (for internal exposures) (see 10 CFR 20.1003).

Transmissivity. The rate of flow of water through a vertical strip of aquifer that is one unit wide and that extends the full saturated thickness of the aquifer. Transmissivity is the hydraulic conductivity multiplied by the saturated thickness; see the definition above for the hydraulic conductivity.

Transport Pathway. A route by which radioactivity travels through the environment to reach a point that may expose a receptor to radiation exposure, either through ingestion, inhalation, or direct exposure.

Unrestricted Area. An area, access to which is neither limited nor controlled by the licensee (see 10 CFR 20.1003).

Unsaturated Zone. The subsurface zone in which the geological material contains both water and air in pore spaces. The top of the unsaturated zone typically is at the land surface, otherwise known as the vadose zone.

Vadose Zone. See Unsaturated Zone.

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PURPOSE, APPLICABILITY, AND ROADMAP Purpose and Applicability of this Volume The purpose of this volume is to do the following:

  • Provide guidance to NRC licensees for demonstrating compliance with the radiological criteria for license termination. Specifically, provide guidance relevant to demonstrating compliance with 10 CFR Part 20, Subpart E, for materials and reactor licensees.
  • Provide guidance to the NRC staff on methods and techniques acceptable for compliance with the license termination criteria.
  • Maintain a risk-informed, performance-based, and flexible decommissioning approach.

This NUREG provides guidance on decommissioning leading to termination of a license.

Licensees decommissioning their facilities are required to demonstrate to the NRC that their proposed methods will ensure that the decommissioning can be conducted safely and that the facility, at the completion of decommissioning activities, will comply with the NRCs requirements for license termination. This volume is also intended to be used in conjunction with NRC Inspection Manual Chapter 2602, Decommissioning Oversight and Inspection Program for Fuel Cycle Facilities and Materials Licensees. Licensees who are subject to Subpart E, Radiological Criteria for License Termination, of Title 10 of the Code of Federal Regulations (10 CFR) Part 20, Standards for Protection against Radiation, should use the policies and procedures discussed in this volume to develop and implement a decommissioning plan (DP) or license termination plan (LTP) (note that throughout this volume, when the term DP is used, it may generally be understood to refer to DPs or LTPs). Uranium recovery facilities may find this information useful, but they are not subject to Subpart E. Agreement State licensees should contact the appropriate regulatory authority. Depending on the State, Agreement State licensees may be able to use this guidance with the substitution of Agreement State Authority for NRC.

Additionally, there are several military and former military sites around the country where the responsible Federal agency (e.g., U.S. Department of the Air Force, U.S. Department of the Army, or U.S. Department of the Navy) is implementing site reclamation activities to address the removal or remediation of radiological material under the Comprehensive Environmental Response, Compensation, and Liability Act (CERCLA), as amended (42 U.S.C. §§ 9601 et seq.), also known as Superfund. The NRC has received a number of inquiries into its regulatory jurisdiction at various sites where the responsible Federal agency uses non-Federal entities (i.e., private service providers) to conduct remediation activities involving regulated radioactive materials (byproduct, source, or special nuclear materials) (see Atomic Energy Act of 1954, as amended (AEA), 42 U.S.C. §§ 2011-2297h (2006)) on Federal property located in an Agreement State. To assist it in making future determinations, the NRC staff has developed a decision process that is consistent with the procedures in SA-500, Jurisdiction Determinations, dated March 10, 2011; the decision process is found in FSME-14-039, Clarification on the Determination of Regulatory Jurisdiction of Non-Federal Entities Conducting Cleanup activities on Federal Property in Agreement States, dated April 22, 2014.

This volume of NUREG-1757 is being issued to describe and make available to licensees and the public (1) guidance on technical aspects of compliance with specific parts of the 1-1

Commissions regulations, (2) methods acceptable to the NRC staff in implementing these regulations, and (3) some of the techniques and criteria the NRC staff uses in evaluating DPs and LTPs. Licensees should use this guidance to prepare DPs, LTPs, final status surveys (FSSs), and other technical decommissioning reports for NRC submittal. The NRC staff will use the guidance in reviewing these documents and related license amendment requests. The guidance in this volume is not a substitute for regulations, and compliance with the guidance is not required. Methods and solutions different from those described in this volume will be acceptable, if licensees provide a sufficient basis for the NRC staff to conclude that the licensees decommissioning actions are in compliance with the Commissions regulations.

However, the use of nonstandard methods may require more detailed justification for the NRC staff to determine acceptability. In addition, the increased complexity and detail of nonstandard demonstrations may result in increased NRC staff review time and, therefore, cost to the licensee.

Volume 2 does not address the following:

  • financial assurance for decommissioning
  • public notification and participation
  • recordkeeping and timeliness in decommissioning
  • decommissioning of uranium recovery facilities
  • disposition of solid materials from licensee control1 Roadmap to this Volume The NRCs regulations require a licensee to submit a DP to support the decommissioning of its facility either (1) when it is required by a license condition or (2) when the NRC has not approved the procedures and activities necessary to carry out the decommissioning, and these procedures could increase the potential health and safety impact to the workers or the public.

Chapters 4-6 provide acceptance criteria and evaluation criteria for use in reviewing DPs and other information submitted by licensees to demonstrate that the facility is suitable for release in accordance with NRC requirements.

The approach used in this volume is similar to that in Volume 1 of this NUREG report.

Volume 1 described the categorization of facilities into Decommissioning Groups 1-7, based on the amount of residual radioactivity, the location of that material, and the complexity of the activities needed to decommission the site. Table 1.1 provides a summary description and examples of each decommissioning group (see Part I of Volume 1 of this NUREG series for more details). Table 1.2 shows the potential applicability of the guidance in this volume to each of these groups. Therefore, where possible, the guidance in this volume has been categorized by the decommissioning groups. For most topics in this volume, the guidance applies to more than one decommissioning group, as shown in Table 1.2. Licensees are encouraged to contact the appropriate NRC staff to determine the applicability of the guidance to their facility.

1 Although dose modeling guidance in this NUREG volume may be useful for assessing dose to members of the public from the release of solid materials, unique scenarios and pathways specific to the release of solid materials are not within the scope of this guidance document.

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Table 1.1 Description and Examples of Each Decommissioning Group Group General Description Typical Examples Licensed material was not released into the Licensees who used only sealed 1 environment, did not cause the activation of sources, such as radiographers and adjacent materials, and did not contaminate irradiators work areas.

Licensed material was used in a way that Licensees who used only quantities of resulted in residual radioactivity on building loose radioactive material that they 2

surfaces and/or soils. The licensee is able to routinely cleaned up (e.g., research demonstrate that the site meets the and development facilities) screening criteria for unrestricted use.

Licensed material was used in a way that Licensees who may have occasionally could meet the screening criteria, but the released radioactivity within NRC limits 3

license needs to be amended to modify or (e.g., broad scope) add procedures to remediate buildings or sites.

Licensed material was used in a way that Licensees whose sites released loose resulted in residual radiological or dissolved radioactive material within contamination of building surfaces or soils, or NRC limits and may have had some 4

a combination of both (but not groundwater). operational occurrences that resulted in The licensee demonstrates that the site releases above NRC limits (e.g., waste meets unrestricted use levels derived from processors) site-specific dose modeling.

Licensed material was used in a way that Licensees whose sites released, resulted in residual radiological stored, or disposed of large amounts of 5 contamination of building surfaces, soils, or loose or dissolved radioactive material groundwater. The licensee demonstrates onsite (e.g., fuel cycle facilities) that the site meets unrestricted use levels derived from site-specific dose modeling.

Licensed material was used in a way that Licensees for whom cleaning their site resulted in residual radiological to the unrestricted release limit would 6 contamination of building surfaces, soils, or cause a greater health and safety or groundwater. The licensee demonstrates environmental impact than could be that the site meets restricted use levels justified (e.g., facilities where large derived from site-specific dose modeling. inadvertent release(s) occurred)

Licensed material was used in a way that Licensees for whom cleaning their site resulted in residual radiological to the restricted release limit would contamination of building surfaces, soils, or cause a greater health and safety or 7 groundwater. The licensee demonstrates environmental impact than could be that the site meets alternate restricted use justified (e.g., facilities where large levels derived from site-specific dose inadvertent release(s) occurred) modeling.

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Table 1.2 Applicability of Volume 2 to Decommissioning Groups Groups Group 1 Group 2 Group 3 Group 4 Group 5 Group 6 Group 7 Dose Assessment Screening criteria Site-specific assessment N/A Method (Section 5, Appendix H) (Section 5, Appendices I, J, M, and Q)

Dose Assessment Yes, for licensees electing partial site releases for Partial Site No Release (Appendices K and L)

Site Yes Yes No Characterization (Section 4.2, Appendix E) (Section 4.2, Appendices E, F, G, and O)

Remedial Action Yes, if remediation is required 1-4 Support Surveys No (Section 4.3, Appendix E)

Yes (Sections 4.4 and 4.5, Yes Final Status Survey No Appendices A, B, D, and (Sections 4.4 and 4.5, Appendices A, B, D, and E)

E)

Complex Survey Situations (Not Yes No Addressed in (Section 4.6, Appendices C, G and O)

MARSSIM)

Surface and Yes Groundwater No Characterization (Appendix F)

Yes, Yes ALARA Analysis No good housekeeping only (Chapter 6, Appendix N)

(Section 6.3)

Because of the variability in the amounts, forms, and types of radioactive material used by each decommissioning group, licensees may need to submit a broad range of information types and details to the NRC for approval of decommissioning activities. The types of information required could vary because of the radionuclides involved, whether remediation is required, or the complexity of the site. The amount of detail discussed in this volume is based on the needs of complicated sites. The NRC staff does not suggest that all licensees should provide the same level of detail. Rather, the amount of detail provided on a specific issue should be commensurate with the complexity of the issue for the facility. Thus, licensees and NRC reviewers should generally determine the level of detail and appropriate methods, based on the complexity of the facility as related to a compliance demonstration. Licensees are encouraged to discuss with their NRC license reviewer the appropriate level of detail to be included in the DP, using the checklists of Appendix D of Volume 1 of this NUREG report.

The technical aspects of sites, as related to decommissioning, are often called either simple or complex. The question becomes what defines the technical aspects as simple or complex.

One needs to decide what aspect of the decommissioning one is trying to judge. For example, site characterization may be complex at a site, but the FSS, after remediation, may be simple and straightforward.

Unfortunately, there is no precise definition or list of characteristics that can define the technical aspects as either simple or complex without caveats. That is because simple and complex are not distinct boxes but part of a continuum. For example, sites using screening criteria are relatively simple, technically, and sites proposing both partial release and restricted release with an engineered barrier design along with institutional controls that rely on active maintenance are relatively complex, technically. While there can be exceptions to the site complexity characterization illustrated in Figure 1.1, Decommissioning Groups 1-3 generally have mostly simple technical aspects, and Decommissioning Groups 5-7 generally have mostly complex technical aspects. Group 4 sites, which are sites without initial residual radioactivity in the groundwater, can be of either complexity.

Simple Complex Group 1-3 4 5 6 7 s

Figure 1.1 Continuum of Site Complexity Simple sites are generally easy to assess, because site characterization information, survey methods, and models with NRC-reviewed default parameter sets are readily available. These sites have residual radioactivity generally limited to building surfaces or surface soil at a site with simple geological and hydrological characteristics.

Technically complex sites are generally sites with one or more of the following conditions:

  • existing groundwater or surface water contamination
  • former burials of radioactive material or highly heterogeneous subsurface soil residual radioactivity 1-5
  • diversified and extensive surface or subsurface residual radioactivity that, because of the interactions between sources, may require data and modeling of these multiple sources at the site
  • radionuclides that (1) are hard-to-detect (HTD), (2) lack suitable surrogate radionuclides, or (3) have very low effective derived concentration guideline levels (DCGLs)
  • current offsite releases such that alternate offsite exposure scenario(s) may be required or the use of an onsite resident farmer exposure scenario may be inadequate (e.g., sites with multiple receptors)
  • planned license termination under restricted conditions (10 CFR 20.1403, Criteria for License Termination under Restricted Conditions)
  • physical barriers or vaults
  • unusual physical or lithologic properties, such as a highly fractured formation, karst features, or sinkholes that may significantly affect assumptions of transport models or the overall conceptual model These conditions are not rigid definitions, as other factors are also important. One such important factor would be the locations where radionuclides are present. For example, a site could be called simple because the predominant radionuclide is a short-lived energetic gamma that is in the surface soil; even if the hydrology at the site is complex, the site would be called simple, because the primary exposure pathway is external exposure, which is an uncomplicated pathway.

Technically complex sites may require more advanced remediation, survey planning, or performance assessment modeling and analysis approaches. Specifically, more advanced approaches may be required to select appropriate models or codes, collect characterization data, justify source term assumptions, ensure internal consistencies in the associated complex transport models, and design site- or source-specific survey plans. Because of the complex nature of these sites, the scope of NRC staff review will depend on site-specific conditions and on the degree of site complexity. Therefore, a generic NRC staff review of complex sites cannot be articulated in this volume.

Licensees and the NRC staff should interact early for information and direction on the development of a complete DP. Once the decision has been made to decommission, the next step is to determine what information the licensee needs to demonstrate site conditions successfully. If the submittal of a DP is not necessary, the licensee should follow the guidance in Volume 1 of this NUREG report for the appropriate decommissioning group.

If the licensee is required to submit a DP, it should schedule a meeting with the NRC staff to discuss both the planned decommissioning and the approach that will be used to evaluate the information submitted to support the decommissioning. The NRC staff and the licensee should review the licensed operations, types and quantities of radioactive materials used at the facility, and any other activities (e.g., spills, leaks) that could affect decommissioning operations. The NRC staff should also discuss the decommissioning goal envisioned by the licensee (i.e., license termination under unrestricted versus restricted conditions) and the information required to be submitted for the appropriate decommissioning group (described in Chapters 10, 11, 12, 13, or 14 of Volume 1 of this NUREG report). The NRC staff should then discuss the acceptance criteria for information to be included in the DP. Finally, the NRC staff should prepare a site-specific checklist for evaluating the DP. Appendix D of Volume 1 of this NUREG report provides a generic checklist that may be used to develop this site-specific checklist.

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Thus, before the licensee begins to develop its DP, both the NRC staff and the licensee should have a good understanding of the types of information that should be included in the DP, as well as the criteria that the NRC will use to evaluate the information submitted to support decommissioning. This should help minimize the need for requests for additional information.

Roadmap for Guidance on Restricted Use, Alternate Criteria, and Use of Engineered Barriers The focus of this volume is on guidance for demonstrating compliance with the dose criteria from 10 CFR Part 20, Subpart E. However, there are additional criteria in Subpart E related to license termination under restricted conditions and the use of alternate criteria for license termination. In addition, some licensees may wish to use engineered barriers as part of the compliance strategy. This section describes where guidance on these subjects may be found in this NUREG series (Volumes 1 and 2).

Table 1.3 provides cross-references to sections of Volume 1 and this volume for guidance on aspects of restricted use, use of alternate criteria, and use of engineered barriers.

Iterative Nature of the Compliance Demonstration Process: A Decision-making Framework The NRC staff developed an overall framework for dose assessment and decision-making where the licensee has decided to begin the decommissioning and license termination process at sites ranging from simple ones to the more complex or contaminated sites. This volume summarizes information for using the framework to step through the decommissioning and license termination process; NUREG-1549, Decision Methods for Dose Assessment to Comply with Radiological Criteria for License Termination, issued July 1998, contains a detailed description. Although the framework was developed for demonstrating compliance using the characterization and dose assessment approach (see Section 2.5), the concepts may be extended for use in the DCGL development and the FSS approach.

This framework is designed to assist the licensee, the NRC staff, and other stakeholders in making decommissioning decisions. By doing so, the process allows the licensee to do the following:

  • coordinate its planning efforts with NRC staff input and conduct dose assessments and site characterization activities that are directly related to regulatory decisions
  • optimize cost decisions related to site characterization, remediation, and land use restrictions
  • integrate analyses for requirements that are as low as is reasonably achievable (ALARA)
  • elicit other stakeholder input at crucial points 1-7

Table 1.3 Cross-References for Restricted Use, Alternate Criteria, and Use of Engineered Barriers Applicable Sections of This Report Issue Volume 1 Volume 2 Initial eligibility demonstration for restricted use 17.7 n/a Institutional controls 17.7 n/a Site maintenance and long-term monitoring 17.7 n/a 17.7 and Obtaining public advice Appendix M n/a Dose modeling for restricted use 17.7 5 ALARA analysis for restricted use 17.7 6 and Appendix N Use of alternate criteria 17.8 n/a Dose modeling for alternate criteria 17.8 5 Use of engineered barriers 17.7 3.5 and Appendix P The framework is designed to allow the licensee flexibility in the decision-making process for demonstrating compliance. As such, the framework provides one method that may be useful for licensees in developing the compliance strategy.

The steps and decision points of the decision framework support an assessment of the entire range of dose modeling options from which a licensee may choose, whether it involves using generic screening parameters, changing parameters, or modifying pathways or models.

Figure 1.2 (modified from NUREG-1549) illustrates the decision framework, including its steps and decision points.

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Figure 1.2 Decommissioning and License Termination Decision Framework (Modified from NUREG-1549) 1-9

1.4.1 Contents and General Concepts of the Iterative Approach in Using the Decision Framework To facilitate the preparation and evaluation of the dose assessments, this framework describes an iterative approach to decision-making for license termination. An iterative approach is helpful because of the very wide range of levels of residual radioactivity, complexity of analysis, and potential remediation necessary at NRC-licensed sites. The iterative approach consists of using existing information for generic screening and using site-specific information as appropriate.

These two phases of the compliance assessment are summarized in broad terms below, while NUREG-1549 contains further details:

(1) Generic screening: In this iteration, licensees would demonstrate compliance with the dose criteria of the License Termination Rule (LTR) by using predefined models and generic screening parameters.

(2) Use of site-specific information as appropriate: If compliance cannot be demonstrated using generic screening, then licensees should proceed to the next iteration of analysis in which defensible site-specific values are obtained and applied.

The following general concepts apply to using the iterative approach with the decision framework shown in Figure 1.2:

  • The approach provides a process for screening sites and for directing additional data collection efforts where necessary or where it is most helpful toward demonstrating compliance.
  • The framework is designed such that the level of complexity and rigor of analysis conducted for a given site should be commensurate with the level of risk that the site poses.
  • The licensee would not need to start the process with generic screening but may move directly to the use of site-specific information, as appropriate.
  • For the process to work efficiently, the licensee is encouraged to involve the NRC staff from the very first step through to the end of the decision-making process.

The framework provides the licensee with a variety of options for performing dose assessments, from simple screening to more detailed site-specific analyses. Use of the framework would normally encompass Steps 1-7; however, the amount of work that goes into each of these steps should be based on the expected levels of residual radioactivity and the health risks they pose. Note that, in this framework, while all sites may start at the same level of very simple analyses (not a requirement for successful implementation), it is expected that only certain sites would progress to very complex dose assessment and options analyses. Some sites may not need to conduct any options analyses, as described in Step 8, and some sites may need to evaluate a limited set of relatively simple and inexpensive options. For example, the licensee at a site with a contained source of residual radioactivity that is obviously simple to remove would not spend time analyzing large suites of alternative data collection and remediation options. On the other hand, the licensee at a site with high levels of widely distributed residual radioactivity may use this process to analyze a variety of simple and complex options to define the best decontamination and decommissioning strategy. Therefore, this approach ensures that the licensees efforts and expenses will be commensurate with the level of risk posed by the site.

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1.4.2 Steps of the Decision Framework NUREG-1549 provides three separate discussions to illustrate the iterative nature of assessments as site complexity increases. The following is both a summary of the steps of the decision framework and a set of examples to help users understand most of the features of dose modeling in the context of the decision framework. This discussion has been modified slightly from that in NUREG-1549 to make it applicable to a broader range of compliance demonstrations. A number of the examples refer to the use of the Decontamination and Decommissioning (DandD) and RESidual RADioactive (RESRAD) materials dose assessment codes. Chapter 5 and Appendices H and I of this volume discuss dose modeling codes, including specifics on these two dose assessment codes, while NUREG-1549 contains further details. Figure 1.2 (modified from NUREG-1549) shows the following steps of the dose modeling framework:

(1) The first step in a compliance assessment involves gathering and evaluating existing data and information about the site for the historical site assessment (HSA), including the nature and extent of residual radioactivity at the site. Often, minimal information is all that is needed for an initial screening analysis (e.g., a simple representation of the source of residual radioactivity). Specifically, information is needed to support the decision that the site is simple and is qualified for a screening analysis. However, licensees should use all readily available information about the site. This step also includes the definition of the performance objectives for compliance with decommissioning criteria.

(2) This step involves defining the exposure scenarios and pathways that are important and relevant for the site dose assessment. This step also includes the preliminary determination of whether the licensee plans to adopt an unrestricted use or restricted use option provided for in the LTR. For all assessments using screening concentration tables or DandD, the NRC has already defined the generic exposure scenarios and pathways for screening. For a site-specific analysis, DandD and the RESRAD family of codes (e.g., RESRAD-ONSITE, RESRAD-OFFSITE, RESRAD-BUILD) may be used, in addition to other codes. The codes used should allow the licensee to select and deselect exposure pathways as appropriate for the site-specific conditions.

(3) Once exposure scenarios are defined and exposure pathways identified, the licensee develops a basic conceptual understanding of the system, often based on simplifying assumptions on the nature and behavior of the natural systems. System conceptualization includes conceptual and mathematical model development and an assessment of parameter uncertainty. Uncertainty in conceptual models should be considered. Using DandD for generic screening (and as the basis for screening concentration tables), the NRC has predefined conceptual models for the exposure scenarios, along with default parameter distributions (based on NUREG/CR-5512, Residual Radiation Contamination from Decommissioning, Volumes 1 and 3, issued October 1992). The site-specific analysis can use DandD or the RESRAD family of codes, after verification that the site conceptual model is compatible with the conceptual model of the code selected.

(4) This step involves the dose assessment or consequence analysis, based on the defined exposure scenario(s), exposure pathways, models, and parameter distributions. This step may also involve the evaluation of FSS results. For generic 1-11

screening, reviewers can accept lookup tables and use the generic models and default parameter probability density functions (PDFs) by running DandD with the appropriate site-specific source concentrations and configuration, while leaving all other information in the software unchanged. Site-specific assessments allow the licensee to use other codes and change pathways and parameter distributions based on site-specific data and information. DandD, and the RESRAD family of codes provide various plots and reports of the dose distribution using Monte Carlo sampling of the input distributions.

(5) This is the first major decision point in the license termination decision process. It involves answering the question of whether the dose assessment results and/or FSS results demonstrate compliance with the dose criteria in 10 CFR Part 20, Subpart E. If the results demonstrate compliance, the licensee proceeds with Steps 6 and 7 to meet the ALARA requirements in Subpart E. If the results are ambiguous or clearly exceed the performance objective, then the licensee proceeds to Steps 8 and 9 for the next iteration of the decision-making process.

(6) In this step, the licensee addresses the ALARA criterion of 10 CFR Part 20, Subpart E, if it has not already done so. If the ALARA requirements are satisfied, then the licensee initiates the license termination. Note that the DandD and the RESRAD family of codes do not involve or automate these steps.

(7) This step includes the administrative and other actions necessary to terminate the license and release the site, with Volume 1 of this NUREG containing more details on the specific actions to do so.

(8) Full application of the decision framework involves defining all possible options the licensee might address to defend a final set of actions needed to demonstrate compliance with license termination criteria. Options may include (i) acquiring more data and information about the site and source(s) of residual radioactivity to reduce uncertainty about the pathways, models, and parameters, and thus reduce the calculated dose, (ii) reducing actual contamination through remediation actions, (iii) reducing exposure to radionuclides through implementation of land use restrictions, (iv) performing an FSS, or (v) some combination of these options.

(9) All the options identified in Step 8 are analyzed and compared to optimize the selection of a preferred set of options. This options analysis may consider the cost of implementation, the likelihood of success given uncertainties (and the expected costs associated with success or failure to achieve the desired results when the option is implemented), the timing considerations and constraints, and other quantitative or qualitative selection criteria.

(10) The activities in Steps 8 and 9 provide information for licensees to choose the preferred decommissioning option based on cost, the likelihood of success, timeliness, and other considerations. For example, results of sensitivity analyses performed with DandD, or the RESRAD family of codes, can be used by a licensee to identify one or more parameters that may be modified, based on the acquisition of site-specific information and data. If new data can reduce the uncertainty associated with parameters found to be important to dose, then the licensee may be able to defend a new calculated dose that meets the license termination criteria. This step may include submitting a DP to the NRC, if necessary, to proceed with the preferred option. If the 1-12

licensee believes that no viable options exist at this time, the licensee should confer with the NRC staff (see also Step 13).

(11) Under this step, the preferred option is implemented. The licensee obtains the information necessary to support revisions to the parameters identified in Steps 8 and 9 or performs an FSS.

(12) Once the licensee obtains the data, it revises the affected parameters for the predefined models, as appropriate. Also, data may support the elimination of one or more of the exposure pathways in the predefined exposure scenarios. DandD and the RESRAD family of codes provide very simple and straightforward modifications of the pathways and parameters of interest.

Once the pathways and parameters are revised, the licensee would revisit Steps 4 and 5 to determine the impact of the revisions on demonstrating compliance with the performance objectives. If met, the licensee proceeds to Steps 6 and 7. If the performance objective is still exceeded, the licensee returns to Steps 8 and 9 to analyze the remaining options.

(13) In certain limited circumstances, terminating the license may not be feasible. The licensee should contact the NRC staff for case-specific guidance and for the regulatory approvals that may be necessary to maintain, rather than terminate, the license.

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FLEXIBILITY IN DEMONSTRATING COMPLIANCE WITH 10 CFR PART 20, SUBPART E The NRC and its licensees share a common responsibility to protect public health and safety.

Federal regulations and the NRC regulatory program are important elements in the protection of the public; however, NRC licensees are primarily responsible for safely using nuclear materials.

The agencys safety philosophy explains that although NRC develops and enforces the standards governing the use of nuclear installations and materials, it is the licensee who bears the primary responsibility for conducting those activities safely. This philosophy applies to the decommissioning of licensed facilities. Thus, the licensee has the primary responsibility for compliance with the license termination criteria. The responsibility of the NRC staff is to oversee the process and conclude that there is reasonable assurance that the criteria have been or will be met and then to terminate or amend licenses, as appropriate.

The dose criteria of 10 CFR Part 20, Subpart E, are performance criteria. In this volume, the NRC staff has taken a risk-informed, performance-based approach to demonstrations of compliance with the license termination criteria, using various methods available to licensees.

Regardless of the specific method used, it is important that the licensee provide sufficient justification for its approach. This chapter discusses some of the aspects of flexibility in methodologies for demonstrating compliance with the license termination criteria. One objective of this chapter is to emphasize the flexibility available in demonstrating compliance with the regulations.

Licensees should consider the flexibility available when demonstrating compliance with the license termination criteria. A licensee may determine that the standard methods are not the best for a given site. The benefit of the performance criteria is the flexibility of approaches allowed to demonstrate compliance.

The NRC staff should evaluate any methodology proposed by licensees. However, the use of nonstandard methods may require more detailed justification for the NRC staff to determine acceptability. In addition, the increased complexity and detail of nonstandard demonstrations may result in increased NRC staff review time and, therefore, cost to the licensee.

Risk-informed Approach to Compliance Demonstrations and Reviews This section summarizes the risk-informed approach to regulatory decision-making. The NRC staff requirements memorandum (SRM) for SECY-98-144, White Paper on Risk-Informed and Performance-Based Regulation, issued March 1999, contains additional details.

The NRC has increased the use of risk information and insights in its regulation of nuclear materials and nuclear waste management, including the decommissioning of nuclear facilities.

Risk is defined by the risk triplet of (1) either a scenario or set of scenarios with a combination of events and/or conditions that could occur, (2) the probability that the scenario(s) could occur, and (3) the consequence (e.g., the dose to an individual) if the scenario(s) were to occur. The term risk insights, as used here, refers to the results and findings that come from risk assessments. The end results of such assessments may relate directly or indirectly to public health effects (e.g., the calculation of predicted doses from decommissioned sites).

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A risk-based approach to regulatory decision-making is based solely on the numerical results of a risk assessment. The Commission does not endorse a risk-based regulatory approach but supports a risk-informed approach to regulation. This approach represents a philosophy whereby risk insights are considered, together with other factors in the regulatory process, to better focus licensee and regulatory attention on design and operational issues commensurate with their importance to public health and safety. The staff does not typically use an explicit consideration of the numerical probability that a scenario would occur (i.e., number 2 of the risk triplet) to determine compliance with the LTR. This is a departure from a purely risk-based approach.

The typical deterministic approach to regulatory decision-making establishes requirements for engineering margin and for quality assurance (QA) in design, manufacture, and construction. In addition, it assumes that adverse conditions can exist and establishes a specific set of design-basis events (i.e., What can go wrong?). The deterministic approach involves implied, but unquantified, elements of probability in the selection of the specific design-basis events to be analyzed. Then, it requires that the design include safety systems capable of preventing and/or mitigating the consequences (i.e., What are the consequences?) of those design-basis events to protect public health and safety. Thus, a deterministic analysis explicitly addresses only two questions of the risk triplet.

The risk-informed approach has enhanced the deterministic approach by (1) allowing explicit consideration of a broader set of potential challenges to safety, (2) providing a logical means for prioritizing these challenges based on risk-significance, operating experience, and/or engineering judgment, (3) facilitating consideration of a broader set of resources to defend against these challenges, (4) explicitly identifying and quantifying sources of uncertainty in the analysis (although such analyses do not necessarily reflect all important sources of uncertainty),

and (5) leading to better decision-making by providing a means to test the sensitivity of the results to key assumptions.

Where appropriate, a risk-informed regulatory approach can also be used to reduce unnecessary conservatism in purely deterministic approaches or can be used to identify areas with insufficient conservatism in deterministic analyses and provide the bases for additional requirements or regulatory actions. Risk-informed approaches lie between the risk-based and purely deterministic approaches (SRM-SECY-98-144).

The NRCs risk-informed regulatory approach to the decommissioning of nuclear facilities is intended to focus the attention and resources of both the licensee and the NRC on the more risk-significant aspects of the decommissioning process and on the elements of the facility and the site that will most affect risk to members of the public following decommissioning. While a licensee must comply with all Commission regulations, a licensee whose site (or aspects of a site) have higher risk-significance may need to provide a more rigorous demonstration to support compliance. Furthermore, the NRC staff generally will apply more scrutiny to reviews of such sites or situations with higher risk-significance. This should result in a more effective and efficient regulatory process. The risk-informed regulatory approach to decommissioning is reflected in this volume, as shown by the following examples:

  • The NRC has developed and is applying the concept of decommissioning groups based on (1) the nature and the extent of the radioactive material present at a site and (2) the complexity of the decommissioning process. The groups are generally related to the potential risks associated with the site, in that the less complex sites with limited 2-2

distribution of radioactive material may pose lower risks (i.e., manageable risks) to individuals and populations during and following decommissioning (see Section 1.2).

  • The NRCs framework for decommissioning regulatory decision-making reflects the iterative nature of the compliance demonstration process. The iterative approach to decision-making for license termination provides a process for screening sites and for directing additional data collection effort toward demonstrating compliance. The framework is designed such that the level of complexity and rigor of analysis conducted for a given site should be commensurate with the level of risk posed by the site (see Section 1.4).
  • This volume provides two different approaches for demonstrating compliance with the dose-based decommissioning criteria, using either a dose modeling approach or a DCGL approach. The dose modeling approach uses measurements of the actual residual radioactivity at a site after cleanup to more realistically assess the potential dose, and therefore the risk, associated with a decommissioned site. The DCGL approach allows a licensee to calculate, a priori, a concentration limit (DCGL) for each radionuclide based on the dose criteria of the LTR and to then demonstrate that the residual radionuclide concentrations are below the DCGLs (see Section 2.5).
  • This volume allows either a screening approach or a site-specific approach to demonstrate compliance. The screening approach allows sites that pose lower potential risks to demonstrate compliance through simpler, yet conservative, screening analysis by adopting screening DCGLs developed by the NRC (see Sections 2.6 and 5.1 and Appendix H).
  • The NRC staff recommends using the data quality objectives (DQOs) process for establishing criteria for data quality and developing survey designs. The process uses a graded approach to data quality requirements, based on the type of survey being designed and the risk of making a decision error based on the data collected. This process aligns the resources expended to collect and analyze data with the risk-significance of the data (see Section 3.2).
  • The NRC provides for an approach to dose assessment that accounts for the site-specific risk-significance of radionuclides and exposure pathways. The NRC staff allows a licensee to identify radionuclides and exposure pathways that may be considered insignificant, based on their contribution to risk, and remove them from further consideration (see Section 3.3). The NRC endorses the approach to FSS design and execution in NUREG-1575, Multi-Agency Radiation Survey and Site Investigation Manual (MARSSIM), which results in a site-specific FSS design that is commensurate with potential risks associated with a site, in terms of the likelihood of exceeding the DCGLs at the site (see Section 4.4).
  • The NRC staff supports a risk-informed approach to site-specific dose modeling for compliance demonstration in several ways: (1) allowing for the site-specific selection of risk-significant exposure scenarios, exposure pathways, and critical groups, (2) expecting selection of conceptual models, numerical models, and computer codes that incorporate the more risk-significant elements of a site, (3) expecting site-specific data for the more risk-significant input parameters, and allowing for more generic data for less risk-significant parameters, and (4) encouraging the use of probabilistic techniques to evaluate and quantify the magnitude and effect of uncertainties in the risk 2-3

assessment, and the sensitivity of the calculated risks to individual parameters and modeling assumptions (see Appendix I and Q).

  • The NRC allows for early partial release of a portion of a site before completion of decommissioning for the entire site, based on the risks associated with the early partial site release (PSR) (see Appendix K).
  • The NRC staff supports a risk-informed graded approach for engineered barriers, and this guidance includes an example of how the risk-informed approach is applied to designing erosion protection barriers (see Appendix P). In addition, the staff supports a risk-informed graded approach for selecting institutional controls and for long-term monitoring and maintenance at restricted use sites, which allows licensees to tailor the type of institutional controls and the specific restrictions on future site use based on a risk framework and insights from dose assessments (see Section 17.7 and Appendix M of Volume 1).

Flexibility in Submissions The NRC staff expects that certain information will be included in licensees DPs, including the FSS design (if an FSS will be performed) and a description of the development of DCGLs or the dose assessment, as applicable. Volume 1 of this NUREG provides additional details on the expected content in these submittals.

Some information is required by regulations (e.g., 10 CFR 30.36(g)(4)) and must be provided in the DP; the DP must include all of the following:

  • the conditions of the site, building, or area, sufficient to evaluate the acceptability of the plan
  • the planned decommissioning activities
  • the methods used to ensure protection of workers and the environment against radiation hazards during decommissioning
  • the planned final radiation survey
  • an updated cost estimate for decommissioning, comparison with decommissioning funds, and a plan for ensuring the availability of adequate funds to complete decommissioning In addition, DCGLs are usually submitted in the DP. Therefore, the typical approach is for the licensee to obtain all the detailed information needed and to submit the information in the DP.

Using the DP checklist (Appendix D of Volume 1 of this NUREG report) as a guide, licensees should coordinate with the NRC staff to determine what information should be included in the DP. For example, for a MARSSIM FSS, the licensee may perform sufficient characterization surveys to determine the appropriate number of samples to obtain for each survey unit. In this case, the NRC staff could approve both the survey design and the DP, and the FSS report (FSSR) may focus primarily on the results of the FSS.

In some cases, not all of the desired information will be available during the DP preparation.

For an FSS, the MARSSIM approach requires that certain information needed for the final 2-4

radiological survey be developed as part of the remedial activities at the site; therefore, this information may not be available for the DP. Similarly, some aspects of the DCGL development or dose assessment may not be available before remediation and final surveys are complete.

When some important information is not available at the time of the DP submission, licensees may either (1) make assumptions about the information or (2) commit to following a specific methodology to obtain the information. In the first case, assumptions will be considered by the NRC staff to be commitments to ensure and subsequently demonstrate that the assumption is true. The licensee would then submit the information at the completion of remediation, at the completion of FSS design, with the FSSR, or at some other appropriate time. For example, a facility uses the ratio of concentrations of thorium (Th)-232 to uranium (U)-238, along with measured concentrations of Th-232, in estimating the concentration of U-238. The licensee may have preliminary information about the ratio of concentrations and, if it is reasonable, may assume that that ratio would be valid for the conditions at the time of the FSS. The NRC staff could accept the use of the assumed value for the ratio. The licensee would demonstrate, at a later stage that the assumed value was valid, perhaps based on measurements made during the FSS.

In the second case, a licensees DP commits to following a specific methodology to obtain the information. For example, a facility may not have sufficient information at the time of the DP submission to determine the number of samples to be taken from each survey unit for a MARSSIM FSS. In this case, the licensee may commit to the procedure recommended in MARSSIM for determining the number of samples in a survey unit and document it in the DP.

The licensee then may determine the number of samples for each survey unit as information is obtained. An FSSR could describe the FSS design, including the number of samples, which the NRC staff would evaluate, along with the FSS results.

Depending on the circumstances and the type of information that is not specifically included in the DP, the NRC staff may consider requiring license conditions to formalize the licensees commitments. This can be accomplished by a specific license condition or by reference to the approved DP (i.e., in the tie-down condition). Licensees should contact the NRC staff for the details on implementing these types of licensee commitments.

The licensee could take a similar approach to information needed to complete a dose assessment. One example is a facility for which the fraction of building-surface residual radioactivity that is removable has been determined during scoping surveys, but the licensee does not know whether remediation activities will change the fraction. In this case, the licensee might assume, for its dose assessment, that the measured fraction will remain unchanged. The NRC staff expects the licensee (1) to make measurements or calculations to demonstrate that the removable fraction was representative of the conditions when remediation is complete and (2) to demonstrate that the dose assessment is representative.

The NRC staff normally would not review DPs or FSSs that use assumptions in lieu of specific information that reasonably could be obtained before submission. In general, the NRC staff expects that assumptions used in developing DPs submitted for review would be limited to those parameters that could change as a result of the remediation or the FSS process itself or to those parameters for which information cannot reasonably be obtained at the time of DP submission. The NRC staff should consider other assumptions on a case-by-case basis.

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Cautions on Making Assumptions or Committing to a Methodology If a licensee finds it reasonable to use the flexible approaches discussed here (e.g., making assumptions to be verified later or committing to a methodology to be performed later), the NRC cautions that (1) it may require a more detailed demonstration of compliance and (2) there may be a greater chance that the staff would not approve the facilitys release. This is because, NRC staff would be reviewing some of the overall compliance strategy at the end of the decommissioning process rather than reviewing complete and detailed information earlier as part of the DP approval process. In addition, the licensee may be required to resolve the assumptions and commitments to meet license conditions. The licensee should contact the NRC staff for details on the use of these flexible approaches.

Use of Characterization Data for Final Status Surveys Although the FSS is generally discussed as if it were an activity performed during a single stage of the radiation survey and site investigation (RSSI) process (see Chapter 4 and Table 4.1 for more about the RSSI process), this does not have to be the case. There is no requirement that an FSS be performed at the end of the decommissioning process. Data from other surveys conducted during the RSSI processsuch as scoping, characterization, and remedial action support surveyscan provide valuable information for an FSS, provided the data are of sufficient quality.

In some cases, the data obtained from these other surveys may be sufficient to serve as an FSS. Licensees may plan the different phases of the RSSI to obtain data of sufficient quality to serve as or to supplement the FSS. The DQO process may be applied to all phases of the RSSI, with DQOs being as robust as those typically developed for the FSS. This approach may result in more costly characterization or remedial action support surveys (to support the more stringent DQOs), which may be balanced against the elimination of a separate FSS.

Choice of Null Hypothesis for Final Status Survey Statistical Analysis The default assumption used in the MARSSIM approach to FSSs and followed by the NRC staff is that the survey unit is considered contaminated above the limit, unless survey data show otherwise. Thus, the null hypothesis used for the MARSSIM FSS statistical tests is that the concentrations of residual radioactivity exceed the DCGLs. This assumption and null hypothesis are considered Scenario A. In most all cases, the NRC staff will consider Scenario A to be the appropriate choice. In some limited cases, a different assumption and null hypothesis, Scenario B, may be appropriate. Scenario B is typically used when the DCGL is within the range of background variability making it difficult to distinguish between residual radioactivity and background. This section (and Appendix A and G, Sections A.4.4 and G.6) provide guidance on Scenario B. NUREG-1505, Revision 1, A Proposed Nonparametric Statistical Methodology for the Design and Analysis of Final Status Decommissioning Surveys:

Interim Draft Report for Comment and Use, issued June 1998, forms the basis for most of the guidance on Scenario B and can be referred to for additional details. Table 2.1 summarizes the differences between Scenarios A and B.

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Deciding which scenario to use and the process to make that decision are difficult issues. In most cases, when the area-wide DCGL (DCGLW) for the survey unit is large compared to the measurement variability, Scenario A should be chosen. This is because even residual radioactivity below the DCGLW should be measurable. In some cases, however, it may be more appropriate to demonstrate indistinguishability from background. When the DCGLW is small compared to measurement and/or background variability, Scenario B may be appropriate. This is because residual radioactivity below the DCGLW may be difficult to measure. Background variability may be considered high when differences in estimated mean concentrations measured in potential reference areas are comparable to screening level DCGLs. Appendix G, Section G.6 of this volume provides an example of the use of Scenario B to demonstrate indistinguishability from background when the residual radioactivity consists of radionuclides that appear in background, and the variability of the background is relatively high (see also NUREG-1505 for additional details).

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Table 2.1 Comparison of FSS Statistical Test Scenarios Characteristic Scenario A Scenario B Assumption for statistical The survey unit is assumed to The survey unit is assumed to test fail unless the data show it can pass unless the data show be released.a that further remediation is necessary.

Null hypothesis The concentration of residual The concentration of residual radioactivity in the survey unit radioactivity in the survey unit exceeds the DCGLs. is indistinguishable from background.b Scenario emphasis Compliance with a dose limit. Indistinguishable from background.

What is needed to reject The concentration of residual The concentration of residual the null hypothesis? radioactivity in the survey unit radioactivity in the survey unit is less than the DCGL. is distinguishable from background.

Rejecting the null The survey unit passes. The survey unit fails.

hypothesis means Increasing the number of Increases the probability that Increases the probability that measurements in a survey an adequately remediated an inadequately remediated unit survey unit will pass. survey unit will fail.

When should the scenario Should be used in most cases Should be used in special be used? (i.e., default) when the DCGL cases (i.e., exception). For is fairly large compared to example, when the DCGL is measurement variability. small compared to measurement and/or background variability.

a For both Scenarios A and B, passing the FSS means a conclusion that the survey unit may be released, and failing means a conclusion that the survey unit may not be released.

b While in most cases Scenario B is expected to be used to show indistinguishability from background, a more general use of Scenario B is to show that the survey unit meets the release criteria. Therefore, other applications of Scenario B are available and may be found acceptable if sufficient technical justification is provided.

As mentioned above, the NRC staffs default assumption is that the use of Scenario A is appropriate. The use of Scenario B is expected only for a small number of facilities, and the considerations for any given facility are expected to be site-specific. Therefore, the staff recommends that licensees contact the NRC early in the FSS design process to discuss considerations for their situation.

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Cautions on the Use of Scenario B for FSS Statistical Tests

  • Scenario B should typically only be used when the DCGL is within the range of background variability, making it difficult to distinguish between the residual radioactivity and background below the DCGL.
  • Case-by-case evaluation is required.
  • Licensees considering the use of Scenario B for compliance with 10 CFR Part 20, Subpart E, are strongly encouraged to contact the NRC staff early in the planning process.
  • Additional information about when it is appropriate to use and potential implementation methods for Scenario B can be found in NUREG-1505 and Appendix G of this document.

Demonstrating Compliance Using Dose Assessment Methods versus Derived Concentration Guideline Levels and Final Status Surveys There is flexibility in the general approach to demonstrating compliance with the dose criteria of 10 CFR Part 20, Subpart E. Two major approaches are available, the dose modeling approach and the DCGL and FSS approach. The dose modeling approach involves characterizing the site post-remediation, if remediation is necessary, and performing a dose assessment. The DCGL and FSS approach requires the development of DCGLs and performing a FSS to demonstrate that the DCGLs have been met. Because the second option is more common, emphasis is placed on the use of DCGLs and FSSs as the compliance method in this NUREG.

However, the two approaches are not mutually exclusive; a hybrid of both approaches can be used. In fact, the NRC recommends licensees assess the final estimated dose using data from the FSS even if the DCGL and FSS approach is used. Table 2.2 shows some advantages and disadvantages of the two approaches.

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Table 2.2 Comparison of Dose Modeling to DCGL and FSS Approaches to Compliance Approach Advantages Disadvantages

  • more realistic
  • may still need preliminary cleanup goals or DCGLs to
  • accounts for time of peak dose design surveys or guide for mixes of radionuclides Dose remediation Modeling
  • can use additional data collected
  • greater chance of additional during decommissioning for iterations of remediation and/or site-specific analyses site characterization
  • can guide remediation activities and data collection
  • could be simpler to implement
  • by using sum of fractions, provides level of conservatism
  • lower likelihood of not showing for radionuclide mix, if peak compliance with dose criterion dose from individual after remediation DCGLs and radionuclides occurs at different FSSs times
  • additional modeling data (i.e., to modify DCGLs) collected during decommissioning cannot be used without license amendment 2.5.1 Dose Modeling Approach Calculating the final dose is the most direct approach to show compliance with the dose criteria in Subpart E. Direct calculation of the total dosefrom all radionuclides in a code that correctly accounts for the time of the peak dose for each radionuclideis a more realistic measure of the potential dose from the site. Another advantage of this approach is that a licensee can use dose modeling information during the decommissioning process to guide additional site characterization, remediation, or other decommissioning options. Additional site characterization could be performed to reduce the level of conservatism in the dose model, parameters, or exposure scenario.

An advantage for sites that comply with the Subpart E criteria without any cleanup is that it may be unnecessary to create any DCGLs; however, the quality of the licensees site characterization data should be sufficient for use as an FSS.

A disadvantage of the dose modeling approach is that changes in the dose modeling, between the approval of the DP and the request for license termination, would result in requiring the NRC staff to review the new information before granting approval of license termination. This additional review step could result in further justification, modeling, remediation activities, or site characterization before approval is granted. This additional review step is similar to what can occur for a site that needs no remediation but uses site-specific dose modeling to show compliance as part of the DP.

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Another disadvantage of using the dose modeling approach is that cleanup goals or final concentrations may need to be estimated (1) to provide assurance that the approach will result in compliance and (2) to design quality surveys, guide remediation activities, and perform additional site characterization.

2.5.2 DCGL and FSS Approach For many sites, especially those that need remediation, the DCGL and FSS approach is a simpler system to show compliance with Subpart E. The DCGL and FSS approach is the one most commonly used for compliance with the LTR and is the one MARSSIM recommends. In the DCGL and FSS approach, the licensee commits to a single concentration value for each radionuclide (i.e., DCGLW) that results in a dose equal to the dose criterion. The DCGLW derivation can use either generic screening criteria or site-specific analysis. The licensee then uses FSSs to demonstrate that the DCGLs have been met. For sites with multiple radionuclides or sources, a sum of fractions approach is typically used to ensure that the dose from all radionuclides and all sources complies with the Subpart E criteria (see Section 2.7).1 The disadvantages of this approach include the following:

  • The sum of fractions approach (Section 2.7) has an underlying assumption that the peak dose for every radionuclide occurs at the same time. This can result in an additional level of conservatism, depending on the mix of radionuclides.
  • Any changes in the DCGLs (e.g., because of new site information) may require a license amendment and NRC staff review.

Merits of Screening versus Site-Specific Dose Assessments The advantage of screening level analyses is that they require minimal justification, characterization, and NRC staff review. The disadvantages are that (i) only residual radioactivity associated with buildings surfaces, and surface soils are considered (may not be appropriate for subsurface residual radioactivity), and (ii) in most cases, screening values are expected to be more restrictive than DCGLs derived using site-specific dose modeling. While site-specific analyses allow more flexibility in estimating the risk for a particular site compared to screening-level analyses, site-specific analyses require site-specific information, and therefore, resources must be spent on obtaining data or providing support for especially risk-significant parameter values. Changes in pathways, exposure scenarios, and conceptual models may also require supporting information. Table 2.3 provides a brief summary of the attributes and merits of both approaches.

The models, exposure scenarios, and parameters used in screening are intended to be conservative, because the lack of information about a site warrants the use of conservative models and default conditions to ensure that the derived dose is not underestimated. The screening analysis is intended to overestimate the dose, to ensure that, for 90 percent of the screening cases, the derived dose is not underestimated. In performing a screening analysis, the NRC staff should recognize that the 90th percentile of the dose distribution is used for 1 Licensees provide their strategies and methods for compliance with LTR criteria in their DP or LTP, which typically include DCGLs and a sum of fractions approach. Once the DP or LTP is approved by NRC, it becomes part of the license. There have been cases in which the NRC has approved plans where the licensee has asked for specific criteria that allow certain narrow types of changes without NRC approval (e.g., allowance of higher clean-up levels or DCGLs under specific conditions) and that the staff found to be acceptable. Otherwise, any change to the approved DP or LTP would require NRC approval via a license amendment.

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calculating compliance, whereas in the site-specific analysis, the peak of the mean dose over time (e.g., 1,000 years) may be used. Deterministic analyses may also be used with sufficient support for those parameters that have a significant impact on dose as identified through sensitivity analysis. An analysis is considered to be site-specific when default parameters in the DandD code used for deriving screening values are changed, source term conditions are modified, or different models or codes are used.

Sum of Fractions The sum of fractions is a simple, yet flexible, approach to deal with multiple radionuclides or sources. A source is any discrete material or medium that contains residual radioactivity. For example, a site with residual radioactivity in surface soil, groundwater, and buried piping has at least three sources. The DCGLW is equivalent to the concentration of a single radionuclide from a single source that would provide 0.25 millisieverts per year (mSv/y) (25 millirem (mrem)/y) total effective dose equivalent (TEDE). The dose from each radionuclide and source should be calculated and then added together. If a licensee only complied with the DCGLW for each radionuclide in each source, the resulting total dose could be as high as 0.25 mSv/y (25 mrem/y) multiplied by the number of radionuclides multiplied by the number of sources.

Unless there was only one source and one radionuclide, the resulting dose would not meet the limits detailed in Subpart E. The dose from all the radionuclides and sources must be equal to or less than the appropriate dose limit in Subpart E.

One simple way to calculate the dose from one radionuclide from one source is to calculate the relative ratio of the residual radioactivity concentration over the DCGLW. Then, the ratio is multiplied by 0.25 mSv/y (25 mrem/y). In fact, for multiple sources or radionuclides, the ratios can be added together, and the sum multiplied by the dose limit. Therefore, the sum of the ratios for all the radionuclides and sources may not exceed 1 (i.e., unity). For example, if radionuclides A and B are present at respective concentrations of Conc A and Conc B, and if the respective applicable DCGLs are Limit A and Limit B, then the concentration needs to be limited so that the following relationship exists to meet Subpart E:

Conc A Conc B

+ 1 (2-1)

Limit A Limit B Similarly, for multiple sources, the sum of the ratios resulting from the sum of the radionuclide contributions may not exceed unity. For example, if the site had a second source, also with radionuclides A and B, but in concentrations of Conc A0 and Conc B0, and DCGLs of Limit A0 and Limit B0, the following condition would need to be satisfied to meet Subpart E:

Conc A Conc B Conc A Conc B

+ + + 1 0 0 (2-2)

Limit A Limit B Limit A Limit B 0 0 2-12

Table 2.3 Attributes of Screening and Site-Specific Analysis Attribute Screening Site-Specific Models/Codes DandD Version 2 or later Any model/code compatible versions with site conditions and approved by the NRC staff.

Scope of Application Only for sites that meet the Any site requirements for screening analysis.

Parameters DandD default parameters Site-specific (physical parameters)2 Exposure DandD default exposure Exposure scenarios/pathways Scenarios/Pathways scenarios/pathways may be modified, based on site conditions.

Dose Metric and The dose at the 90th percentile Peak annual dose from a Consideration of of the peak dose distribution deterministic analysis or peak Uncertainty within 1,000 years of the mean annual dose from a probabilistic analysis3. In both cases, adequate support for risk-significant4 parameters and distributions is needed.

Compliance is evaluated over a 1,000-year period.

The deterministic parameter set described in NUREG/CR-5512, Volume 1, and implemented in DandD Version 1 have been superseded by the parameter set described in NUREG/CR-5512, Volume 3, and implemented in DandD Version 2. DandD Version 1 did not support probabilistic analyses and used a default deterministic input parameter set.

In the general form, the relationship of the ratios, commonly known as the sum of the fractions or the unity rule, would be for M sources (s) and N radionuclides (r):

2 With respect to behavioral and metabolic parameters, default values used in DandD Version 2, and listed in NUREG/CR-5512, Volume 3, can be used to describe the average member of the critical group with minimal justification (e.g., when the site-specific exposure scenario is consistent with the screening scenario).

3 The peak of the mean should be used with caution if it is significantly different then the mean of the peaks and there is evidence of risk dilution as described in Appendix I and Q.

4 Risk-significant parameters are identified through sensitivity analysis. Appendix I contains additional details.

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=1 =1 1 (2-3) where Concs,r = the concentration of radionuclide r in source s, and Limits,r = the DCGLW value for radionuclide r in source s.

For sites with a number of radionuclides and sources, it may be easier to partition the acceptable fraction between various sources or radionuclides. For example, a licensee could commit to keeping the ratio from the groundwater to less than 25 percent of the dose limit.

One major, implicit assumption in using the sum-of-fractions approach is that peak doses for each radionuclide and source occur simultaneously. Because radionuclides can be transported through the environment at different rates, and a particular radionuclide may be dominated by a different pathway compared to another radionuclide, there are many radionuclides and contaminated media for which peak doses do not occur simultaneously. For example, radionuclides that result in predominantly external dose and are short-lived, such as cobalt (Co)-60, usually have a peak dose right after license termination. For radionuclides that result in peak dose through irrigation or drinking groundwater, the peak dose may not occur until years after license termination. When peak doses are from different radionuclides or sources occur at different times, the sum-of-fractions approach tends to overestimate that dose. In some situations, the overestimate may be significant and affect the compliance demonstration. The licensee could directly calculate the combined dose using final concentrations from the FSS to more accurately estimate the risk from the site (see Section 2.5).

Flexibility for Use of Institutional Controls and Engineered Barriers at Restricted Use Sites The new guidance developed for restricted use sites includes risk-informed and performance-based approaches to institutional controls, engineered barriers, monitoring, and maintenance. These approaches not only enhance the attention to safety by being risk informed but also provide flexibility to licensees planning restricted use for a site. The approaches described allow licensees to select the most effective and efficient methods for restricting site use, designing engineered barriers to mitigate disruptive processes important to compliance, and planning monitoring and maintenance activities that are tailored to the specific site and indicators of potential disruptive processes and engineered barrier performance.

Section 3.5, and Appendix P of this volume and Section 17.7 and Appendix M to Volume 1 describe these approaches.

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CROSS-CUTTING ISSUES This chapter provides guidance on several cross-cutting issues that relate to multiple aspects of surveys, characterization, and dose modeling. The issues addressed in this chapter include the following:

  • transparency and traceability of compliance demonstrations
  • DQO process
  • insignificant radionuclides and exposure pathways
  • considerations for other constraints on allowable levels of residual radioactivity
  • the use of engineered barriers
  • integration of radiological surveys and dose modeling for surface and subsurface soils Use of the Guidance in this NUREG Report
  • The suggestions in this NUREG report are only guidance, not requirements.
  • Other methods for demonstrating compliance are acceptable.
  • As noted in Section 5.3 of Volume 1 of this NUREG report, licensees are encouraged to have early discussions with the NRC staff in developing DPs. This is especially important when NRC guidance is limited on a specific topic. Early discussions can save licensees from following an approach that the NRC staff may find unacceptable and can clarify this guidance and identify areas where modification may be helpful for the staffs review.
  • This volume refers to a number of other documents for guidance. In some cases, this volume states that the NRC staff has approved the referenced guidance. In other cases, the documents are only referenced as potentially relevant information. In these latter cases, the licensee should contact the NRC staff to determine the specific applicability to a facility, as appropriate.

Transparency and Traceability of Compliance Demonstrations Licensees submit information to justify their conclusions about compliance with 10 CFR Part 20, Subpart E. Because of insufficient justification, the NRC staff has found a number of licensee submittals to be inadequate for NRC to conclude it has reasonable assurance that the license termination rule criteria can be met. This section describes some considerations for improving the thoroughness of licensee submittals. Transparency refers to arguments or calculations with descriptions sufficient for an independent reviewer to replicate. Traceability refers to the sources of information being relatable to the original source. The NRC staff encourages licensees to submit compliance demonstrations that are transparent and traceable. This should result in more efficient and effective staff reviews.

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To help ensure transparency and traceability, licensees should include the following in their justification:

  • describe the sources of data
  • provide only summary data, if appropriate.
  • to the extent that summary data are provided, include references to detailed data and make them available to the NRC staff for review if requested (e.g., on an inspection)
  • clearly describe the data, including units used, in tables and other presentations
  • state the assumptions and ensure that the difference between assumptions and justified data or parameters is clear
  • provide justifications for parameters or arguments, especially when employing nonstandard arguments or nondefault parameters
  • describe uncertainties in data and parameters Data Quality Objectives Process Compliance demonstration is the process that leads to a decision as to whether or not a survey unit meets the release criteria. For most sites, this decision is supported by statistical tests based on the results of one or more surveys. In most cases, the initial assumption used by the NRC staff is that each survey unit is contaminated above the release criteria until proven otherwise (Scenario A). The surveys are designed to provide the information needed to reject this initial assumption. The NRC staff recommends using the Data Life Cycle as a framework for the planning, implementation, assessment, and decision-making phases of final surveys.

Section 2.3 of MARSSIM discusses the major activities associated with each phase of the Data Life Cycle.

One aspect of the planning phase of the Data Life Cycle is the DQO process, which is a series of planning steps for establishing criteria for data quality and developing survey designs. The DQO process consists of seven steps:

(1) statement of the problem (2) identification of the decision (3) identification of inputs to the decision (4) definition of the study boundaries (5) development of a decision rule (6) specification of limits on decision errors (7) optimization of the design for obtaining data 3-2

The output from each step influences steps later in the process. Even though the DQO process is depicted as a linear sequence of steps, it is iterative in practice; the outputs of one step may lead to reconsideration of prior steps.

The DQO process uses a graded approach to data quality requirements, defined according to (1) the type of survey being designed and (2) the risk of making a decision error based on the data collected. This approach provides a more effective survey design, combined with a basis for judging the usability of the data collected. Thus, the DQO process is a flexible planning tool that licensees can use more or less intensively as the situation requires.

DQOs are qualitative and quantitative statements that satisfy all of the following by doing the following:

  • clarifying the study objective
  • defining the most appropriate type of data to collect
  • determining the most appropriate conditions for collecting the data
  • specifying limits on decision errors that will be used as the basis for establishing the quantity and quality of data needed to support the decision Although the DQO process is generally used for surveys and the steps of an RSSI, the general concepts may also be applied to dose assessments. Chapter 5 contains additional information on data requirements related to dose modeling and DCGL development. Data of sufficient quality should be collected to ensure the technical defensibility of the modeling results focusing on risk-significant parameters. The DQO process should be used to guide data collection and analysis. Licensees are encouraged to apply the general concepts of the DQO process to all applicable parts of their compliance demonstration. The use of the DQO process can help ensure that the type, quantity, and quality of data and calculations used in decision-making will be appropriate for the intended application. Additional guidance on the use of the DQO process is in MARSSIM Section 2.3 and Appendix D and in EPA/600/R-96/055, Guidance for the Data Quality Objectives Process, issued August 2000.

Experience has shown that it is helpful for the licensee to identify all appropriate DQOs in planning and designing the final status survey plan (FSSP). The process of identifying the applicable DQOs ensures that the survey plan requirements, survey results, and data evaluation are of sufficient quality, quantity, and robustness to support the decision on whether the cleanup criteria have been met.

In purpose and scope, the DQO process can include a flexible approach for planning and conducting surveys and for assessing whether survey results support the conclusion that release criteria have been met. The DQO process can be an iterative process that continually reviews and integrates, as needed, new information in decision-making and the design of the final survey plan. Finally, the selection and optimization of DQOs should facilitate the later evaluation of survey results and decision-making processes during the data quality assessment (DQA) phase. Licensees have had difficulty developing DQOs, especially during the optimization step, and have not taken full advantage of the DQO process. Experience has shown that licensees often rigidly structure the process by relying too much on characterization data and are not readily open to the possibility of incorporating new information as it becomes available. This rigid approach makes implementing any changes difficult and is an inefficient 3-3

use of resources that imposes time delays (e.g., the additional time required to determine how to implement any changes).

Insignificant Radionuclides and Exposure Pathways Licensees should note that they are required to comply with the applicable dose criteria; nothing in this discussion should be interpreted to allow licensees to exceed the criteria.

This section provides guidance on conditions under which radionuclides or exposure pathways may be considered insignificant and may be eliminated from further consideration. The dose criteria in 10 CFR Part 20, Subpart E, apply to the total dose from residual radioactivity. Thus, demonstrations of compliance should generally address the dose from all radionuclides and all exposure pathways. However, the NRC staff has determined that it is reasonable to eliminate radionuclides or pathways that are insignificant contributors to dose from further detailed consideration, although their dose contributions must be considered in demonstrating compliance with the radiological criteria for license termination.

The NRC staff considers radionuclides and exposure pathways that contribute no more than 10 percent of the dose criteria, considering uncertainty1, to be insignificant contributors to dose.

Because the dose criteria are performance criteria, this 10-percent limit is an aggregate limitation only. That is, the sum of the dose contributions from all radionuclides and pathways considered insignificant should be no more than 10 percent of the dose criteria (e.g., no more than 0.025 mSv/y (2.5 mrem/y) for the unrestricted release limit of 0.25 mSv/y (25 mrem/y)). No limitation on either single radionuclides or individual pathways is necessary. In cases of restricted release, where two dose criteria apply (one for institutional controls in place and one that considers the possibility of restrictions failing), the 10-percent limitation should be met for both dose criteria. In making a determination that radionuclides or pathways are insignificant, licensees should consider both reasonably foreseeable and less likely but plausible exposure scenarios (see Section 5 for more information). Licensees should also consider the presence of elevated areas, and potential in-growth of progeny from postulated insignificant radionuclides, when determining that the radionuclides or pathways contribute no more than 10 percent of the dose criteria and are, therefore, insignificant.

Typically, licensees would use characterization data to show that certain radionuclides or pathways are insignificant before remediation or the FSS. However, if remediation is planned, licensees may also be able to show through analysis that the dose contributions of certain radionuclides following remediation will contribute no more than 10 percent of the dose criteria.

The approach discussed in the preceding paragraph (accounting for removal of radioactivity during remediation) is similar to the use of surrogate radionuclides that implicitly considers the dose contributions of hard-to-detect radionuclides (see Appendix A of this document and Section 4.3.2 of MARSSIM, Revision 1, for additional information on use of surrogate radionuclides). However, unlike surrogate radionuclides, the licensee has an additional burden of showing that the dose contributions of the radionuclides or pathways are insignificant.

1 Uncertainty in exposure scenarios is considered through evaluation of both reasonably foreseeable and less likely but plausible scenarios. Uncertainty in parameter values can be managed through conservative assumptions (e.g., selection of parameter values from parameter distributions that tend to lead to higher doses). Chapter 5, Appendix I and Appendix Q have more information on consideration of uncertainty.

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Characterization data can be used to define the relative activities of significant and insignificant radionuclides to determine the relative dose contributions of the radionuclides present at the site. Ratios should be conservatively selected so that they do not underestimate the potential dose contributions of the insignificant radionuclides. For example, using the minimum detectable concentrations (MDCs) or actual reported concentrations (even when less than MDCs), consistent with the principles in NUREG-1576, Multi-Agency Radiological Laboratory Analytical Protocols (MARLAP), may be appropriate except in cases where more than 40 percent of results are less than the MDC. Use of the 95th percentile ratios of insignificant to significant radionuclides is acceptable, and alternative approaches will be considered on a case-by-case basis depending on the site characterization data. The licensee should also show that the relative dose contributions of the insignificant radionuclides will not increase following remediation because of an increase in concentration or redistribution of residual radioactivity. In general, the NRC does not require post-remediation sampling of the insignificant radionuclides, due to their low risk-significance. However, if there is a valid concern that the dose contributions of the postulated insignificant radionuclides could be significant following remediation, licensees may choose to manage this uncertainty as part of the DQO process (e.g., through post-remediation sampling of the insignificant radionuclides, similar to the approach used for surrogate radionuclides discussed in MARSSIM Section 4.3.2).

Once a licensee has demonstrated that radionuclides or exposure pathways are insignificant, then (1) the dose from the insignificant radionuclides and pathways must be accounted for in demonstrating compliance, but (2) the insignificant radionuclides and pathways may be eliminated from further detailed evaluations. For example, after sufficient site characterization, suppose a licensee shows that the dose from strontium (Sr)-90 at the facility is 0.02 mSv/y (2 mrem/y), which is less than 10 percent of the dose criterion for unrestricted use. In this case, Sr-90 can be considered insignificant and eliminated from the FSS and from detailed consideration in the dose modeling. However, the dose from Sr-90 has to be considered in demonstrating compliance with the dose criterion. In some cases, licensees may be able to show that the dose contributions of the insignificant radionuclides and pathways are much less than 10 percent of the dose criteria. However, it may be less burdensome on the licensee to show that the insignificant radionuclides and pathways contribute less than 0.025 mSv/y (2.5 mrem/y) than it is to show that the insignificant radionuclides and pathways contribute much less than 0.01 mSv/y (1 mrem/y), for example. Thus, it may be preferable to round the dose contributions of the insignificant radionuclides and pathways higher, and thereby provide a greater safety margin in meeting the dose criteria.

It is important for the licensee to document the radionuclides and pathways that it has considered insignificant and eliminated from further consideration and for the licensee to justify the decision to consider them insignificant. However, licensees and the NRC staff should be aware that remediation techniques (or other activities or processes) may result in an increased dose from the postulated insignificant radionuclides or pathways, such that the dose contributions are no longer insignificant. Thus, licensees should also demonstrate that the dose contributions of insignificant radionuclides and pathways deemed insignificant will not increase (or were not underestimated) as a result of other activities.

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Summary of Determining Insignificant Radionuclides and Exposure Pathways

  • Licensees may eliminate insignificant radionuclides and exposure pathways from further detailed consideration. However, the dose from the insignificant radionuclides and pathways must be accounted for in demonstrating compliance with the applicable dose criteria.
  • Insignificant means no greater than 10 percent of applicable dose criteria.
  • Ten percent is an aggregate limit; total dose contributions of all radionuclides and all exposure pathways considered insignificant should not exceed the 10 percent limitation.
  • There is no additional limit on single radionuclides or pathways.
  • Licensees should also address the potential for dose from postulated insignificant radionuclides or pathways to increase during remediation activities.

Considerations for Other Constraints on Allowable Residual Radioactivity Situations or standards other than the dose criteria and ALARA requirements of Subpart E may constrain the final dose below 0.25 mSv/y (25 mrem/y). Two main causes for constraining the Subpart E dose limit are (1) a PSR and (2) other standards or regulations.

A PSR occurs when a licensee releases a portion of its site for unrestricted use before terminating the entire license. While the licensee should demonstrate that the residual radioactivity at the time of unrestricted release of the specific area meets the Subpart E dose limit, the residual radioactivity of the area should also be taken into account during final termination to demonstrate that the entire site has met the appropriate release criteria.

Appendix K of this volume discusses dose modeling considerations for PSR. In general, the comments below are also applicable to PSRs.

Demonstrating compliance with the Subpart E dose limit does not eliminate the licensees requirement for meeting other applicable Federal, State, or local rules and regulations. These regulations from other governmental agencies may conflict with the requirements of Subpart E, as they may allow higher or lower levels of residual radioactivity on the site or may conflict in other ways, such as limiting decommissioning options or final status. Nevertheless, the staff should review a DP for compliance only with NRC requirements, including 10 CFR Part 20, Subpart D, Radiation Dose Limits for Individual Members of the Public, which incorporates, where applicable, the requirements of 40 CFR Part 190, Environmental Radiation Protection Standards for Nuclear Power Operations. For example, in reviewing the appropriateness of proposed DCGLs or the number of samples per survey unit for an unrestricted site, the NRC staff would use the limit of 0.25 mSv/y (25 mrem/y), and not a States limit of 0.2 mSv/y (20 mrem/y). Thus, any requests for additional information would also be based on compliance with the limit of 0.25 mSv/y (25 mrem/y). Licensees should note that use of a lower dose standard does not necessarily lead to a lower cleanup level because of differences in assessment approaches including differences in exposure scenarios, models and parameters used by different agencies setting the dose standards.

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Use of Engineered Barriers This section, and the details provided in Appendix P, provide guidance to licensees considering the use of engineered barriers (e.g., engineered covers, including those designed for erosion protection, stabilizing cementitious materials, and reactive walls) to demonstrate compliance with radiological criteria for license termination. Section 3.5 and Appendix P also support Section 17.7.3 of Volume 1 by giving guidance on the information expected to be submitted in a DP for the engineered barrier analysis and the technical basis for engineered barrier performance.

In the Commissions view, engineered barriers are distinct and separate from institutional controls (see Decommissioning Criteria for the West Valley Demonstration Project (M-32) at the West Valley Site: Final Policy Statement (67 FR 5003)). Generally, institutional controls are designed to restrict access, whereas engineered barriers are usually designed to inhibit water from contacting waste, limit releases, or mitigate doses to intruders.

Engineered barriers are passive, man-made structures or devices intended to enhance a facilitys ability to meet the dose criteria in the LTR.2 Engineered barriers are usually designed to inhibit water from contacting waste and releasing radionuclides to groundwater, thereby reducing exposure from ingestion of contaminated water. Engineered surface barriers may also slow erosion or otherwise decrease the likelihood of the waste being exposed at the surface through human or biotic activity, thereby reducing dose from inadvertent intrusion and direct exposure to the waste. In some cases, engineered barriers may also be used to passively limit access of critical groups to residual radioactivity (e.g., a durable rock barrier that decreases the likelihood of excavation or drilling into residual radioactivity). Used in the general sense, an engineered barrier could be one of a broad range of barriers with varying degrees of durability, robustness, and isolation capability.

On the other hand, institutional controls are active controls that limit access to the site, and the use of it, to minimize disturbances to engineered barriers and ensure that the exposure from the residual radioactivity does not exceed the established criteria. Institutional controls include legal mechanisms (e.g., land use restrictions) and physical controls (e.g., signs, markers, landscaping, and fences) to prevent unauthorized access to the site and minimize disturbances to engineered barriers. Institutional controls may require financial assurance to ensure adequate control and maintenance of the site, legal enforceability, and an entity with the capability, authority, and willingness to enforce the controls.

The licensee determines the functionality and robustness of barriers using the risk-informed graded approach described in Appendix P and evaluated on a site-specific basis for each license application. However, the general framework that a licensee should consider would not vary from licensee to licensee; only the depth and breadth of information supplied to demonstrate the performance of the engineered barriers may vary. Appendix P provides the general framework a licensee should consider for use of engineered barriers in the decommissioning process.

2 In some cases, for restricted-use sites, a licensee can propose active monitoring and maintenance of the engineered barrier; this would be considered an institutional control and could be used to enhance the assumed level of performance or longevity of the engineered barrier beyond its passive performance.

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It is expected that engineered barriers will most frequently be used for restricted use sites.

However, there may be infrequent cases where engineered barriers are used as one component of a decommissioning approach to achieve unrestricted use of a site. These cases should be infrequent because of the uncertainty associated with the long-term performance of engineered systems without monitoring and maintenance and because the goal should be to achieve unrestricted use without relying on engineered barriers. If an engineered barrier is used at an unrestricted use site, only the passive performance of the barrier to mitigate radiological impacts may be credited (i.e., performance of the barrier without monitoring, inspection, and maintenance) in the dose assessment to demonstrate compliance with the LTR dose criteria.

The assessment of engineered barrier performance should consider reasonably foreseeable, as well as less likely but plausible, disruptive conditions from human activities and from natural events and processes. Results of reasonably foreseeable exposure scenarios should be strictly considered when demonstrating compliance with radiological criteria for license termination, while less likely but plausible exposure scenarios should also be considered to help risk-inform the decision-making process.

Proposals to use engineered barriers for unrestricted use sites should be part of an overall plan for decontaminating and decommissioning a site, as presented in a licensees DP or LTP. The licensee should demonstrate that, at the time of license termination, the site meets LTR criteria, including the criterion to reduce residual radioactivity to ALARA levels. Therefore, licensees should consider whether the following would be consistent with ALARA requirements in the absence of engineered barriers: (1) removal and disposal of contaminated components and equipment, (2) decontamination (and demolition, if appropriate) of buildings, (3) removal and disposal of waste streams remaining on site from past operations, and (4) excavation and removal of large areas of soil contamination as waste. Therefore, while engineered barriers may be proposed to assist with meeting the LTR criteria, source removal must first be considered in demonstrating that residual radioactivity has been removed to ALARA levels.

Chapter 6 and Appendix N contain additional information.

Similarly, for restricted use sites, under 10 CFR 20.1403(a) the licensee must show that further removal of residual radioactivity would result in net public or environmental harm or that leaving the residual radioactivity in place is ALARA. Licensees should also include other considerations (e.g., distance to the disposal facility, efficient use of available disposal capacity at the offsite facility, unavailability of required treatment options, lack of disposal options other than leaving the contaminated materials on site, and overall risk reduction including non-radiological risks), if applicable and appropriate, in its determination of whether additional removal of residual radioactivity is reasonably achievable.3 In their proposal to use engineered barriers, licensees should include all relevant information concerning the risks of using the proposed approach versus other remediation alternatives.

Because of the wide range of licensed decommissioning sites, the LTR and the NRCs decommissioning guidance are not prescriptive as to the criteria for, or acceptability of, site-specific engineered barriers. Therefore, the licensee has flexibility in the methods used to demonstrate compliance with the performance-based criteria of 10 CFR Part 20, Subpart E.

Because of this flexibility and because engineered barrier designs are site-specific, it is very important for the licensee to clearly and completely document how it has considered site-specific conditions (e.g., site-specific resources, climate, degradation mechanisms) in its engineered barrier designs and monitoring and maintenance program.

3 Reasonably achievable is judged by considering the state of technology and the economics of improvements in relation to all the benefits of these improvements. See Section N.1 for additional information.

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Appendix P provides the detailed framework for applying engineered barriers to achieve decommissioning at a site, and an example of a graded approach to erosion covers. A summary of existing guidance and reference information on the application of engineered barriers at decommissioning sites is also listed in Appendix P.

In summary, the following points should be considered when applying engineered barriers to achieve decommissioning at a site:

  • Engineered barriers are distinct and separate from institutional controls.
  • Only the passive performance of engineered barriers (i.e., no monitoring and maintenance) can be relied on at unrestricted use sites.
  • Passive performance and (when institutional controls are assumed to be in place) active monitoring and maintenance of engineered barriers can be relied on at restricted use sites. Active monitoring and maintenance are considered institutional controls and may enhance the level of performance assumed for an engineered barrier beyond what is assumed for a barriers passive performance. For most cover systems, long-term, passive performance cannot be validated, although natural analogs (Section P.1.3) and other forms of model support can be used to provide support for the assumed passive performance of some cover systems.
  • For licensees pursuing unrestricted use of their site, residual radioactivity has been reduced to levels that are ALARA before reliance on engineered barriers to meet LTR criteria. Furthermore, for licensees pursuing restricted use of their site, a demonstration is required to show that additional removal of residual radioactivity to meet unrestricted use criteria would result in net public or environmental harm or that further reductions are not being made because levels that meet restricted use conditions are ALARA.
  • Engineered barrier evaluation are reviewed on a case-by-case basis using a risk-informed approach.

Surveying and Considering Risk Associated with Surface and Subsurface Soils Attempts have been made to define surface soils, based on the capability of relatively low-cost scan instrumentation to detect residual radioactivity near the surface of buildings and in soils.4 MARSSIM survey protocols, discussed in more detail in Chapter 4 and Appendices A-G, assume that residual radioactivity is present on the surfaces of buildings and soils. Likewise, certain dose modeling codes such as DandD, discussed in more detail in Chapter 5 and Appendices H and I, also make assumptions about the depth of residual radioactivity that could be important for assessing dose to potential receptors. For example, surface soils are important for assessing dose from certain exposure pathways such as direct radiation exposure, incidental soil ingestion, and inhalation. While surface soils could also support other important exposure pathways, such as plant growth and consumption, contaminated subsurface soils could also contribute significantly to radionuclide uptake and dose dependent on the plant type and 4 For example, surface soil has been associated with the top 15 cm of soil that can typically be measured using scan instrumentation.

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radionuclide. Thus, if subsurface residual radioactivity is ignored, the dose could be significantly underestimated.

Human activities could cause subsurface soils to become surface soils from mixing or redistribution after license termination. For example, tilling soil to promote crop growth or excavating soil for home construction or well drilling could lead to the redistribution of subsurface soils to the surface (see Appendix J). Before license termination, remediation of the site and sorting of relatively contaminated and uncontaminated soils could lead to the redistribution or reuse of contaminated subsurface soils on the surface (see Appendix G). While the definition of surface soil is an important consideration when it comes to surveying and assessing the dose impacts associated with residual radioactivity remaining at a site at the time of license termination, there is no clear line of demarcation of surface and subsurface soils; the final distribution of residual radioactivity should be considered in dose modeling and the FSS design. Potential redistribution of radioactivity following license termination must also be considered when deriving subsurface soil DCGLs or when performing dose modeling to assess the impact of subsurface residual radioactivity (Appendix J).

Despite the linkages to scan instrument capability and dose modeling discussed above, the depth of residual radioactivity in surface and subsurface soils assumed for the purpose of the FSS design should not be arbitrarily selected based on dose modeling assumptions or scan capability. Rather, the expected horizontal and vertical extent of residual radioactivity informed by characterization surveys should be considered in the FSS design and in assessing dose.

Significant heterogeneity in residual radioactivity concentrations should also be considered when designing the FSS and assessing dose. Chapters 4 and Appendices A-G, provide guidance on for performing radiological surveys for surface and subsurface soils; building materials; and surface water and groundwater (Appendix F); Chapter 5, and Appendices I and J have guidance on evaluating the dose impacts associated with surface, subsurface, and heterogeneous residual radioactivity in soils and building surfaces.

If subsurface residual radioactivity is present, soil sampling will likely be necessary to supplement scan surveys to adequately characterize the full vertical extent of residual radioactivity. Depth discrete sampling of soils may also be needed if there is significant vertical heterogeneity, and representation of the vertical heterogeneity is important in assessing dose.

DCGLs derived from dose modeling should be consistent with the actual vertical extent of the residual radioactivity and significant vertical heterogeneity considered. For example, it may be important to differentiate soil layers based on vertical heterogeneity and derive more than one set of DCGLs (e.g., derive surface and subsurface soil DCGLs), if vertical heterogeneity is found to be important to dose. Alternatively, it would always be acceptable to use the most limiting DCGL to simplify the FSS or dose modeling could be used to confirm that radiological criteria for license termination are met based on the final configuration of residual radioactivity in soils, as measured in FSSs. Chapter 5 and Appendix I contain additional information on the dose modeling approach. Appendix J contains information on scenarios that should be considered for buried residual radioactivity. Appendix I, Section I.2 specifically contains additional information on source term abstraction for heterogenous distributions.

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FACILITY RADIATION SURVEYS Radiation Survey and Site Investigation Process As a framework for collecting the information required for demonstrating compliance identified using the DQO process (see Section 3.2 of this volume), the NRC staff recommends using a series of surveys. The RSSI process is an example of a series of surveys designed to demonstrate compliance with the decommissioning regulations of 10 CFR Part 20, Subpart E.

Table 4.1 identifies the steps in the RSSI process and indicates where specific guidance on each step can be found.

Table 4.1 Cross-References for Principle Steps in the Radiation Survey and Site Investigation Process Principal Step Applicable Guidance Site Identification Chapter 16, Volume 1, of this NUREG report Section 2.4 of MARSSIM, Revision 1 Historical Site Assessment Section 4.1.1 of this volume Section 2.4 and Chapter 3 of MARSSIM, Revision 1 Scoping and Characterization Section 4.2 of this volume Survey Sections 2.4, 5.2 and 5.3 of MARSSIM, Revision 1 Remedial Action Support Survey Section 4.3 of this volume Sections 2.4 and 5.4 of MARSSIM, Revision 1 Final Status Survey Section 4.4 of this volume Sections 2.4 and 5.5 of MARSSIM, Revision 1 Note: As of the date of publication of this NUREG, MARSSIM, Revision 2, is in the process of being published. Because MARSSIM, Revision 2, has not yet been published, references to sections of MARSSIM in this volume are with respect to MARSSIM, Revision 1 (i.e., section numbers may be different in Revision 2). NRC staff plans to incorporate MARSSIM, Revision 2, into a future revision of this volume.

4.1.1 Historical Site Assessment The RSSI process uses a graded approach that starts with Site Identification and the HSA and is later followed by other surveys that lead to the FSS. In most cases, the radiological status of a site will already be known based on its prior use and the presence of radioactive material although a records review could identify areas of use, disposal, or spills that have been overlooked. The HSA collects existing information describing a sites complete history from the start of site activities to the present. The necessity for detailed information and the amount of effort to conduct an HSA depend on the type of site, associated historical events, regulatory framework, and availability of documented information. The main purpose of the HSA is to determine the current status of the site or facility, but the data collected may also be used to differentiate sites that need further action from those that pose little or no threat to human health and the environment (see Section 2.3). This screening process can provide a site disposition recommendation or propose additional surveys. Because much of the data collected during 4-1

HSA activities are qualitative or are analytical data of unknown quality, many decisions on a site are the result of professional judgment.

The primary objectives of the HSA include the following:

  • identify potential sources of residual radioactivity
  • determine if sites pose a threat to human health and the environment
  • differentiate impacted from nonimpacted areas
  • provide input to scoping and characterization survey designs
  • assess the likelihood of residual radioactivity migration
  • identify additional potential radiation sites related to the site being investigated The HSA typically consists of three phases: (1) preliminary investigation of the facility or site, (2) site visits or inspections, and (3) an evaluation of the site based on the information collected.

The HSA should identify special survey situations that may need to be addressed, such as subsurface radioactivity; sewer systems, waste plumbing, and floor drains; ventilation ducts; and embedded piping containing residual radioactivity. Appendix G of this volume includes information on special survey situations. Section 2.4.2 and Chapter 3 of MARSSIM contain additional guidance on the HSA.

4.1.2 Summary of Survey Types The NRCs regulations (10 CFR 20.1501(a)) require a licensee to make or cause to be made surveys that may be necessary for the licensee to comply with the regulations in Part 20, including the radiological criteria for license termination found in 10 CFR Part 20, Subpart E.

The licensee would demonstrate compliance with this requirement by performing an FSS. The FSS will demonstrate that the licensees site or facility, or both, can meet the radiological criteria for license termination.

Other surveys (e.g., scoping surveys, characterization surveys, and remedial action support surveys) are used to identify areas with residual radioactivity but are typically not used to demonstrate compliance with the radiological criteria for license termination.

The NRC endorses the FSS methodology described in MARSSIM. The guidance in this chapter does not replace MARSSIM, and users of this chapter should be familiar with and use MARSSIM. Thus, it is intended that licensees will use this chapter and MARSSIM as guidance for acceptable approaches or methodologies to conduct surveys supporting decommissioning and FSSs, in particular. The following text refers to specific sections of MARSSIM, Revision 1, when applicable.

The measurement methods applied in assessing radiation and radioactivity levels can vary according to the objectives of the particular survey. It is expected that different types of surveys would be conducted during the course of decommissioning work, with each having a different emphasis while at the same time, sharing common elements. The sections below summarize six survey types.

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Background SurveyAlthough not specifically identified as a step in the RSSI process, this survey constitutes measurements of sites in areas surrounding the facility to establish the baseline; that is, the normal background levels of radiation and radioactivity. In some situations, historical measurements may be available from surveys performed before the construction and operation of a facility. The background survey takes on added importance if one may ultimately be comparing onsite cleanup units to offsite reference areas. Appendix A of this volume contains guidance on background surveys.

Scoping SurveyThis survey, performed to augment the HSA, provides sufficient information (1) to determine if residual radioactivity is present that warrants further evaluation and (2) to make initial estimates of the level of effort required for remediation and to prepare a plan for a more detailed survey, such as a characterization survey. The scoping survey does not require that all radiological parameters be assessed. Sections 2.4 and 5.2 of MARSSIM and Section 4.2 of this volume contain additional guidance on the scoping survey.

Characterization SurveyThis survey determines the type and horizontal and vertical extent of residual radioactivity on or in structures and environmental media. The survey should be sufficiently detailed to provide data for planning decommissioning actions, including remediation techniques, projected schedules, costs, waste volumes, and health and safety considerations during remediation. Section 4.2 of this volume contains additional guidance on characterization surveys.

Remedial Action Support SurveyThis survey, which could be repetitive in nature, is conducted in what is effectively a real-time mode to guide cleanup efforts and ensure the health and safety of workers and the public. The effectiveness of the remediation efforts can be assessed as they progress. The precision and accuracy of measurements associated with this type of survey are generally not sufficient to determine the final radiological status of the site.1 Section 4.3 of this volume contains additional guidance on remedial action support surveys.

Final Status SurveyThis survey demonstrates that residual radiological conditions satisfy the predetermined criteria for release for unrestricted use or, where appropriate, for use with designated restrictions. It is this survey that provides data to demonstrate that all radiological parameters (e.g., total surface activity, removable surface activity, exposure rate, and radionuclide concentrations in soil and other materials) satisfy the established guidelines and conditions. Section 4.4 of this volume contains additional guidance on FSSs.

Confirmatory SurveyThe regulator performs this survey to obtain data to substantiate the results of the licensees FSS. The objective of this type of survey is to verify that characterization, remediation, and final status actions and documentation, conducted as part of the RSSI process, are adequate to demonstrate that the site is radiologically acceptable, relative to applicable criteria. Section 15.4.5 of Volume 1 of this NUREG report contains additional information on confirmatory surveys.

These types of surveys are performed at various stages of the decommissioning process. Early on, where known residual radioactivity exists, the simplest of measurement approaches can be used to document the need to clean up a specific building surface or parcel of land. In practice, the simpler methods would generally be applicable to the scoping and remedial action support surveys. The more complex methods, which produce data with higher precision and accuracy, 1 In certain cases, it may be prudent to collect data of sufficient quality during the remedial action support survey to support the final status survey during remediation as discussed in Sections 4.3 and G.3.2.

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will be required for background, characterization, final status, and confirmatory surveys. In general, wherever measurements are to be performed at or close to background levels, greater sensitivity in the measurement is required.

The conduct of these surveys and the methods applied have some interchangeable elements.

It is possible that measurements collected in one survey can be used for another. For instance, if measurements sufficient in spatial coverage and with adequate detection limits were taken, the results of the scoping survey in an unaffected area could be used to support the FSS. The emphasis of the guidance in this volume is on the methods that can be applied to meet the requirements of the FSS, although they can be applied to other survey work as well. In performing decommissioning surveys, licensees should be cognizant of the survey methodologies and their limitations, especially where newer technologies are employed such as in situ gamma spectroscopy. Some information on the capabilities of this technology to detect discrete particles in soil can be found in the NRC sponsored study/report by the Oak Ridge Institute for Science and Education (ORISE) titled Spatially-Dependent Measurements of Surface and Near-Surface Radioactive Material Using In situ Gamma Ray Spectrometry (ISGRS) For Final Status Surveys (Chapman et al., 2006).

NRC released the final version of the Multi-Agency Radiological Laboratory Analytical Protocols manual (MARLAP) in 2004. The MARSSIM and the MARLAP manual are complementary guidance documents in support of cleanup and decommissioning activities. The MARSSIM document contains guidance on how to plan and carry out a study to demonstrate that a site meets appropriate release criteria. It describes a methodology for planning, conducting, evaluating, and documenting environmental radiation surveys conducted to demonstrate compliance with cleanup criteria. The MARLAP manual provides guidance and a framework for both project planners and laboratory personnel to ensure that radioanalytical data will meet the needs and requirements of cleanup and decommissioning activities.

The MARLAP manual recommends the use of a directed or systematic planning process. A directed planning process is an approach for setting well-defined, achievable objectives and developing a cost effective, technically sound sampling and analysis design that balances the data users tolerance for uncertainty in the decision process with the resources available for obtaining data to support a decision. For example, the NRC and licensees have determined that side-by-side surveys are more efficient than waiting for a final sitewide confirmatory survey.

The NRC and licensees should plan ahead and coordinate their schedules to implement efficient side-by-side confirmatory surveys. Appendix D contains more details on MARLAP and how it can enhance radiation monitoring.

Appendix D of this volume includes information on survey data quality and reporting, Chapter 5 of MARSSIM provides survey checklists, Appendix E contains information on survey measurements, and Appendix G has information on special survey issues.

4.1.3 Areas of Review The NRC staff should review the results of the radiological characterization survey to determine whether it contains sufficient information to permit planning for site remediation that will be effective and will not endanger the remediation workers, to demonstrate that it is unlikely that significant quantities of residual radioactivity have gone undetected, and to provide information that will be used to design the FSS.

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The purpose of the NRC staff review is to verify that the FSS design is adequate to demonstrate compliance with the radiological criteria for license termination. The FSS review should determine whether the results demonstrate that the site, area, or building meets the radiological criteria for license termination.

The staff should note that NRC regulations require that DPs include a description of the planned final radiological survey. Recognizing the flexible approach discussed in Section 2.2 of this volume and that the MARSSIM approach allows certain information needed to develop the final radiological survey to be obtained as part of the remedial activities at the site, a licensee or responsible party should submit information on facility radiation surveys in one of two ways, as summarized below. Section 2.2 of this volume provides additional relevant guidance.

Method 1:

The licensee or responsible party may submit the information contained in Sections 4.1.4-4.3 of this volume as part of the DP, along with a commitment to use the MARSSIM approach in developing the final status survey. The licensee or responsible party would then submit the information discussed in Section 4.4 at the completion of remediation or design development for the final status survey for the site. The licensee or responsible party will submit the FSSR (Section 4.5) after performing the final status survey.

Method 2:

The licensee or responsible party may submit the information contained in Sections 4.1.4-4.4 of this volume, along with a commitment to calculate the number of sampling points that will be used in the final status survey, in accordance with the procedure described in MARSSIM. The licensee or responsible party would then submit the FSSR (Section 4.5) after performing the final status survey. If this method is used, the licensee or responsible party should include in the FSSR the information contained in the last three bullets under Information to Be Submitted, in Section 4.4 of this chapter.

4.1.3.1 Acceptance Review The review should ensure that the licensees submittal contains the information summarized under the Areas of Review, as appropriate for the particular submittal. The NRC staff should ensure that the level of detail appears to be adequate for it to perform a detailed technical review but should not review the technical adequacy of the information, which it should determine during the detailed review.

4.1.3.2 Safety Evaluation The material to be reviewed is both informational in nature and requires specific detailed technical analysis. The NRC staff should verify that the survey designs and results are adequate for demonstrating compliance with the radiological criteria for license termination.

4.1.4 Release Criteria The NRC staff review is to verify that the licensee has provided appropriate release criteria, referred to as the DCGLs. Generally, the licensee should provide the DCGLW, for the survey unit average concentrations, and the applicable DCGLEMC (elevated measurement comparison) for small areas of elevated concentrations, for all affected media.

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4.1.4.1 Acceptance Criteria 4.1.4.1.1 Regulatory Requirements 10 CFR 20.1402, 20.1403, and 20.1404 4.1.4.1.2 Regulatory Guidance NUREG-1575, Multi-Agency Radiological Survey and Site Investigation Manual (MARSSIM) 4.1.4.1.3 Information to Be Submitted The licensee should list the DCGL(s) that will be used to design the surveys and to demonstrate compliance with the radiological criteria for release, including the following:

  • a summary table or list of the DCGLW for each radionuclide and affected medium of concern
  • the DCGLEMC (or areas factor derived from dose modeling) for each radionuclide and media of concern if Class 1 survey units are present (Appendix A.2 of this volume discusses classification of site areas)
  • the appropriate DCGLW for the survey method to be used if multiple radionuclides are present The information to be submitted is also included as part of the master DP checklist provided in this NUREG report (see Section XIV.a from Appendix D of Volume 1).

4.1.4.2 Evaluation Criteria The NRC staff should verify that, for each radionuclide and affected media of concern, the licensee has provided a DCGLW and, if Class 1 survey units are present, a table of DCGLEMCs (or area factors). The NRC staff should verify that the values presented are consistent with the values developed pursuant to dose modeling, as discussed in Chapter 5 of this volume. If multiple radionuclides are present, MARSSIM Sections 4.3.2, 4.3.3, and 4.3.4 describe acceptable methods to determine DCGLs appropriate for the survey technique.

Scoping and Characterization Surveys 4.2.1 Scoping Surveys Early in the decommissioning process, the licensee identifies the potential residual radioactivity present at the site, the relative ratios of radionuclides, and the general extent of residual radioactivityif anyboth in activity levels and affected area or volume. Although the license and operational history documentation will assist to varying degrees in providing this information, it will often be necessary to supplement it with actual survey data. Therefore, the licensee should perform a scoping survey typically consisting of limited direct measurements (exposure rates and surface activity levels) and samples (smears, soil, water, and material with induced activity) obtained (1) from site locations considered to be the most likely to contain residual activity and (2) from other site locations, including those immediately adjacent to the 4-6

radioactive materials use areas. This survey provides a preliminary assessment of site conditions, relative to guideline values. The scoping survey forms the basis for initial estimates of the level of effort required for decommissioning and for planning the characterization survey.

Measurements and sampling in known areas of residual radioactivity do not need to be as comprehensive or be performed to the same sensitivity level as will be required for the characterization survey or FSS. However, when planning and conducting the scoping survey, the licensee should remember that some of the data, particularly from locations not affected by site operations, may be used as final survey results or to supplement the characterization or final survey results, or both. Similar measuring and sampling techniques as used for those categories of surveys, therefore, may be warranted.

Scoping surveys provide site-specific information based on limited measurements. The following are the primary objectives of a scoping survey:

  • perform a preliminary hazard assessment
  • support classification of all or part of the site as a Class 3 area
  • evaluate whether the survey plan can be optimized for use in either the characterization or final stage
  • perform radiological status surveys
  • provide data to address the requirements of other applicable regulations
  • provide input to the characterization survey design, if necessary Scoping surveys are conducted after the HSA is completed and consist of judgment measurements based on the HSA data. If the results of the HSA indicate that an area is Class 3 and no residual radioactivity is found, the licensee may classify the area as Class 3 and conduct a Class 3 FSS. If the scoping survey locates residual radioactivity, the licensee may consider the area to be Class 1 (or Class 2) for the FSS and typically conduct a characterization survey, collecting sufficient information to identify situations that require immediate radiological attention. Licensees should be aware that requirements of other applicable regulations (e.g., non-radiological constituents) may differ from NRC requirements. Appendix F to MARSSIM contains a comparison of MARSSIM guidance to other requirements.

4.2.2 Characterization Surveys After identifying the affected locations, the licensee conducts a characterization survey to more precisely define the horizontal and vertical extent and magnitude of residual radioactivity. The survey should be sufficiently detailed to provide data for planning the remediation effort, including the remediation techniques, schedules, costs, and waste volumes, as well as necessary health and safety considerations during remediation. The type of information obtained from a characterization survey is often limited to that which is necessary to differentiate a surface or area as containing or not containing residual radioactivity. A high degree of accuracy may not be required for such a decision when the data indicate levels well above the guidelines. On the other hand, when data are near the guideline values, a higher degree of accuracy is usually necessary to ensure the appropriate decision about the true radiological conditions. In addition, one category of radiological data (e.g., soil radionuclide concentration or total surface activity) may be sufficient to determine if there is residual radioactivity, and other 4-7

measurements (e.g., exposure rates or removable residual radioactivity levels) may, therefore, not need to be performed during characterization. As the scoping survey example demonstrates, the choice of survey technique should be commensurate with the intended use of the data, including considerations for possible future use of the results to supplement the FSS data.

Licensees typically submit site characterization summary information as part of their DP. If submitted site characterization information is insufficient to reasonably identify the horizontal and vertical extent and nature of residual radioactivity, the NRC may decline to accept and review the DP until such information is provided. The NRC may ask the licensee to submit site characterization plans or other site characterization information before submitting the DP, or the NRC may elect to meet with the licensee before or during site characterization work. However, licensees are not required to submit a separate site characterization or site characterization report unless required by a license condition. Rather, site characterization information is required as a component of the DP. Therefore, the NRC staff will only request site characterization and reports separate from the DP submittal when necessary to provide assurance the horizontal and vertical extent and nature of residual radioactivity are reasonably identified.

The characterization survey is generally the most comprehensive of all the survey types and generates the most data. This includes preparing a reference grid, systematic as well as judgment measurements, and surveys of different media to include surface soils and interior and exterior surfaces of buildings. Additionally, the characterization survey should identify all activated materials (typically Decommissioning Groups 4-7) and hard-to-detect radionuclides throughout the site. The decision as to which media will be surveyed is a site-specific decision addressed throughout the RSSI process (see MARSSIM).

Characterization surveys may be performed to satisfy a number of specific objectives.

Examples include the following:

  • determining the nature and horizontal and vertical extent of residual radioactivity
  • evaluating remediation alternatives (e.g., unrestricted use; restricted use; or onsite or offsite disposal under the provisions of 10 CFR Part 20, Subpart K, Waste Disposal (onsite and other alternative methods of disposal can be approved on a case-by-case basis under the provisions of 10 CFR 20.20022))
  • developing input to pathway analysis and dose or risk assessment models for determining site-specific DCGLs in becquerel/kilogram (Bq/kg), picocuries/gram (pCi/g),

becquerel/square meter (Bq/m2), or disintegrations per minute/100 square centimeters (dpm/100 cm2), as applicable

  • estimating the occupational and public health and safety impacts during decommissioning 2

Guidance for the NRC staff review of alternative disposal requests under 10 CFR 20.2002 is available in Guidance for the Reviews of Proposed Disposal Procedures and Transfers of Radioactive Material, under 10 CFR 20.2002 and 10 CFR 40.13(a), Agencywide Documents Access and Management System Accession No. ML19295F109.

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  • evaluating remediation technologies
  • developing input to the FSS design
  • complying with requirements of other applicable regulations This volume does not include detailed discussions of characterization survey design for each of these objectives; the user should consult other references for specific characterization survey objectives not covered. For example, U.S. Department of Energys (DOEs) Decommissioning Handbook, issued March 1994, is a good reference for characterization objectives for evaluating remediation technologies or unrestricted or restricted use alternatives. Additionally, ANSI N13.59:2008, Characterization in Support of Decommissioning Using the Data Quality Objectives Process is a useful document which presents characterization strategies based on the DQO process. The licensee should consult other references (EPA, Guidance for Conducting Remedial Investigations and Feasibility Studies Under CERCLA, issued October 1988; EPA, Superfund Removal Procedures, 1988; EPA, Federal Radiation Protection Guidance for Exposure of the General Public, dated December 23, 1994; NUREG-1501, Background as a Residual Radioactivity Criterion for DecommissioningDraft Report, issued August 1994) for planning decommissioning actions (e.g., remediation techniques, projected schedules, costs, and waste volumes) and health and safety considerations during remediation. Also, the specific modeling code documentation should determine the types of characterization data needed to support risk or dose modeling.

4.2.3 Areas of Review The purpose of the NRC staff review is to verify that the licensee determined the radiological condition of the property well enough to permit planning for a remediation that will be effective and will not endanger the remediation workers, to demonstrate that it is unlikely that significant quantities of residual radioactivity have gone undetected, and to provide sufficient information for designing the FSS. Note that some licensees have used, or may request authorization to use, information developed during the characterization survey to support the final radiological survey.

Licensees may use characterization survey data to support the final radiological survey, as long as they can demonstrate that nonimpacted areas at the site have not been adversely affected by decommissioning operations and that the characterization survey data are of sufficient scope and detail to meet the Information to Be Submitted guidance for a final survey.

4.2.3.1 Acceptance Criteria 4.2.3.1.1 Regulatory Requirements 10 CFR 30.36(g)(4)(i), 40.42(g)(4)(i), 70.38(g)(4)(i), and 72.54(g)(1) 4.2.3.1.2 Regulatory Guidance NUREG-1575, Multi-Agency Radiological Survey and Site Investigation Manual (MARSSIM) 4-9

4.2.3.1.3 Information to Be Submitted The information supplied by the licensee should be sufficient to allow the NRC staff to determine whether the characterization survey design is adequate to assess the radiological status of the facility. The licensee should describe the radiation characterization survey design and the results of the survey, including the following:

  • a description and justification of the survey measurements for affected media (for example, building surfaces, building materials (volumetric contamination) contamination, surface soil, subsurface soil, surface water, groundwater, sediments, as appropriate)
  • a description of the field instruments and methods that were used for measuring concentrations and the sensitivities of those instruments and methods
  • a description of the laboratory instruments and methods that were used for measuring concentrations and the sensitivities of those instruments and methods
  • the survey results, including tables or charts of the concentrations of residual radioactivity measured
  • maps or drawings of the site, area, or building showing areas classified as nonimpacted or impacted and visually summarizing residual radioactivity concentrations in impacted areas
  • a justification for classifying areas as nonimpacted
  • a discussion of why the licensee considers the characterization survey to be adequate to demonstrate that it is unlikely that significant quantities of residual radioactivity have gone undetected
  • a discussion of how areas or surfaces in a survey unit were surveyed or why they did not need to be surveyed if considered to be inaccessible or not readily accessible
  • for sites, areas, or buildings with multiple radionuclides, a discussion justifying the ratios of radionuclides that will be assumed in the FSS or an indication that no fixed ratio exists, and each radionuclide will be measured separately (note that this information may be developed and refined during decommissioning, and licensees may elect to include a plan to develop and justify final radionuclide ratios in the DP)

The information to be submitted is also included as part of the DP Checklist provided in this NUREG report (see Section XIV.b from Appendix D of Volume 1).

Licensees should note that, if they elect to dispose of buildings and structures rather than leave them in place (for unrestricted release), the LTR does not apply to the material moved offsite from those buildings and structures. Rather, building and structure deconstruction and dismantlement materials can be released from the site in accordance with existing license conditions. The data from the characterization survey may be sufficient to demonstrate compliance with the conditions of the existing license for releasing material from the site.

However, a characterization survey may not be required to demonstrate compliance with the license condition for releasing material from the site. Section G.2.1 of Appendix G of this volume provides additional guidance on the offsite disposition of materials.

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4.2.3.2 Evaluation Criteria The NRC staff should verify that the licensee has adequately characterized the site, area, or building relative to the location and horizontal and vertical extent of residual radioactivity. An adequate characterization is one that permits planning for a remediation that will be effective and will not endanger the remediation workers, demonstrates that it is unlikely that significant quantities of residual radioactivity have gone undetected, and provides information that will be used to design the FSS. The extent of detail in the information provided by the licensee should be appropriate for the specific site, area, or building.

The NRC staff should verify that the survey design and results demonstrate that the licensee or responsible party has adequately characterized the site. The characterization survey is adequate if it meets the criteria in the following guidance:

  • Section 5.3 of MARSSIM for the characterization survey (the NRC staff may use the Example Characterization Survey Checklist in Section 5.3 of MARSSIM for evaluating the licensees submittal)
  • MARSSIM Chapter 6 and Appendix E for instrument capabilities and sensitivities
  • MARSSIM Section 4.8.4 for preparing areas for survey Remedial Action Support Surveys The effectiveness of remediation efforts in reducing residual radioactivity to acceptable levels is monitored by a remedial action support survey as the remediation effort is in progress. This type of survey activity guides the cleanup in a real-time mode; it also ensures that the remediation workers, the public, and the environment are adequately protected against exposures to radiation and radioactive materials arising from the remediation activities.

The remedial action support survey typically provides a simple radiological parameter, such as direct radiation near the surface being remediated. The level of radiation, below which there is reasonable assurance that the guideline values have been attained, is determined and used for immediate, infield decisions. Such a survey is intended for expediency and does not provide thorough or accurate data describing the final radiological status of the site.

The remedial action support survey is applicable to monitoring surfaces and soils or other bulk materials only if the radionuclides of concern are detectable by field survey techniques. For radionuclides and media that cannot be evaluated at guideline values by field procedures, samples are to be collected and analyzed to evaluate the effectiveness of remediation efforts.

For large projects, the use of mobile field laboratories can provide more timely decisions on the effectiveness of remedial actions. Examples of situations for which remedial action support surveys would not be practicable are (1) when soil contains pure alpha or beta emitting radionuclides and (2) when very low energy beta emitters such as tritium are present on surfaces.

Licensees conduct remedial action support surveys to do the following:

  • support remediation activities 4-11
  • determine when a site or survey unit is ready for the FSS
  • provide updated estimates of site-specific parameters used for planning the FSS The determination that a survey unit is ready for an FSS following remediation is an important step in the RSSI process. Remedial activities may result in changes to the distribution of residual radioactivity within the survey unit. Thus, for many survey units, the site-specific parameters used during FSS planning (e.g., variability in the radionuclide concentration, probability of small areas of elevated activity) may need to be confirmed or reestablished following remediation. Obtaining updated values for these critical parameters should be considered when planning a remedial action support survey. In some cases, where concentrations of some radionuclides after remediation may be very low, it may be useful for licensees to show that certain radionuclides can be considered insignificant; in that case, further detailed evaluation as part of the FSS may not be necessary (see Section 3.3 of this volume).

However, the dose from the insignificant radionuclides must be accounted for in demonstrating compliance with the applicable dose criteria.

Note that this survey does not provide information that can be used to demonstrate compliance with the DCGLs and is an interim step in the compliance demonstration process. The FSS will then survey in detail areas that are likely to satisfy the DCGLs on the basis of the remedial action support survey. Alternatively, the remedial action support survey can be designed to meet the objectives of an FSS.3 DCGLs may be recalculated, based on the results of the remediation process, although a license amendment may be needed to change (increase) previously approved DCGLs in a DP or LTP.

4.3.1 Areas of Review Staff review of the description of the remedial action support surveys should verify that the licensee has designed these surveys appropriately and to assist in determining when remedial actions have been successful, so that it may begin the FSS. In addition, information from these surveys may be used to provide the principal estimate of residual radioactivity variability that will be used to calculate the FSS sample size in a remediated survey unit.

4.3.1.1 Acceptance Criteria 4.3.1.1.1 Regulatory Requirements 10 CFR 30.36(g)(4)(ii), 40.42(g)(4)(ii), and 70.38(g)(4)(ii),

4.3.1.1.2 Regulatory Guidance NUREG-1575, Multi-Agency Radiological Survey and Site Investigation Manual (MARSSIM) 3 In certain cases, it may be prudent to collect data of sufficient quality during remediation to support the final status survey during remediation. For example, for large subsurface soil excavations, it may be more practical to collect samples at the bottom and sides of the excavation and/or perform scanning prior to filling in the excavation. See Section G.3.2 for additional details.

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4.3.1.1.3 Information to Be Submitted The NRC staff should verify the licensees or responsible partys description of the support survey includes the following information:

  • a description of field screening methods and instrumentation
  • a demonstration that field screening should be capable of detecting residual radioactivity at the DCGLW The information to be submitted is also included as part of the DP Checklist provided in this NUREG report (see Section XIV.c from Appendix D of Volume 1).

4.3.1.1.4 Evaluation Criteria The NRC staff should verify that the description of the remedial action support surveys meets (1) the criteria in MARSSIM Section 5.4 for performing remedial action support surveys and (2) the criteria in the applicable MARSSIM chapters listed in this volume for evaluating technical issues, such as appropriate surveys instruments and survey instrument sensitivity.

Final Status Survey Design Professional judgment and biased sampling are important for locating residual radioactivity and characterizing the horizontal and vertical extent of residual radioactivity at a site. However, the MARSSIM focus is on planning the FSS, which uses a systematic approach to sampling.

Systematic sampling is based on rules that try to achieve the representativeness assumed by the statistical tests.

The licensee uses the FSS to demonstrate compliance with regulations. The primary objectives of the FSS are to do the following:

  • verify survey unit classification
  • demonstrate that the potential dose from residual radioactivity is below the release criterion for each survey unit
  • demonstrate that the potential dose from small areas of elevated activity is below the release criterion for each survey unit Data provided by the FSS can demonstrate that all radiological parameters satisfy the established guideline values and conditions.

4.4.1 Areas of Review The purpose of the NRC staffs review is to verify that the FSS design is adequate to demonstrate compliance with the radiological criteria for license termination.

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4.4.1.1 Acceptance Criteria 4.4.1.1.1 Regulatory Requirements 10 CFR 20.1501(a), 30.36(g)(4)(iv), 40.42(g)(4)(iv), 70.38(g)(4)(iv), and 72.54(g)(4) 4.4.1.1.2 Regulatory Guidance

  • Draft NUREG-1505, Revision 1, A Nonparametric Statistical Methodology for the Design and Analysis of Final Status Decommissioning Surveys - Interim Draft Report for Comment and Use
  • NUREG-1575, Multi-Agency Radiological Survey and Site Investigation Manual (MARSSIM)
  • NUREG-1507, Revision 1, Minimum Detectable Concentrations with Typical Survey Instruments for Various Contaminants and Field Conditions 4.4.1.1.3 Information to Be Submitted The information supplied by the licensee should be sufficient to allow the NRC staff to determine whether the FSS design is adequate to demonstrate compliance with the radiological criteria for license termination. The information should include all of the following:
  • a brief overview describing the FSS design
  • a description and map or drawing of affected areas of the site, area, or building classified by residual radioactivity levels (Class 1, 2, or 3) and divided into survey units, with an explanation of the basis for division into survey units (maps should have compass headings indicated)
  • a description of the background reference areas and materials, if they will be used, and a justification for their selection
  • a summary of the statistical and other tests that will be used to evaluate the survey results, including the elevated measurement comparison (EMC), if Class 1 survey units are present; a justification for any test methods not included in MARSSIM; and the values for the decision errors ( and ) with a justification for and values greater than 0.05 for Scenario A and B, respectively
  • a description of scanning instruments, methods, calibration, operational checks, coverage, and sensitivity for each media and radionuclide
  • a description of the instruments, calibration, operational checks, sensitivity, and sampling methods for in situ sample measurements, with a demonstration that the instruments and methods have adequate sensitivity (noting that if a licensee uses an advanced technology (e.g., in situ gamma spectroscopy), it must be shown to perform with sensitivities that allow detection of residual radioactivity at an appropriate fraction of the DCGL and corresponding investigation levels (ILs))

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  • a description of the analytical instruments for measuring samples in the laboratory, including the calibration, sensitivity, and methodology for evaluation, with a demonstration that the instruments and methods have adequate sensitivity
  • a description of how the samples to be analyzed in the laboratory will be collected, controlled, and handled
  • a description of the FSS investigation levels and how they were determined The information to be submitted is also included as part of the DP Checklist provided in this NUREG report (see Section XIV.d from Appendix D of Volume 1). Appendix A provides additional information about demonstrating the appropriate selection of survey instrumentation.

4.4.1.2 Evaluation Criteria The NRC staff review should verify that the FSS design is adequate to demonstrate compliance with the radiological criteria for license termination. The FSS design is adequate if it meets the criteria in the following guidance:

  • Appendix A to this volume, for general guidance on implementing the MARSSIM approach for conducting FSSs
  • Appendix B to this volume, for guidance on alternative methods of FSS for simple situations
  • MARSSIM Sections 4.4 and 4.6 for classifying areas by residual radioactivity levels and dividing areas into survey units of acceptable size
  • MARSSIM Section 4.5 for methods to select background reference areas and materials
  • NUREG-1505, Chapter 13, for a method to account for differences in background concentrations between different reference areas
  • MARSSIM Section 5.5.2 for statistical tests
  • Appendix A to this volume, Section A.8.2, for decision errors
  • MARSSIM Sections 6.5.3 and 6.5.4 for selection of acceptable survey instruments, calibration, and operational checkout methods
  • MARSSIM Section 6.7 for methods to determine measurement sensitivity; NUREG-1507, Revision 1, for instrument sensitivity information
  • MARSSIM Sections 5.5.2.4, 5.5.2.5, 5.5.3, 7.5, and 7.6 for scanning and sampling
  • MARSSIM Section 7.7 for sample analytical methods (Table 7.2 in Section 7.7 for acceptable analytical procedural references)
  • MARSSIM Sections 7.5 and 7.6 for methods for sample collection
  • MARSSIM Section 5.5.2.6 for survey ILs 4-15
  • Appendix G to this volume for surveys for special structural or land situations Final Status Survey Report To the extent possible, the FSSR should stand on its own with minimal information incorporated by reference. Although the FSS is discussed as if it were an activity performed at a single stage of the site investigation process, this does not have to be the case. Data from other surveys conducted during the RSSI processsuch as scoping, characterization, and remedial action support surveyscan provide valuable information for an FSS, provided the data are of sufficient quality.

4.5.1 Areas of Review The purpose of the NRC staff review is to verify that the results of the FSS demonstrate that the site, area, or building meets the radiological criteria for license termination. At a minimum, the FSSR should contain the information listed in Section 4.5.1.1.3. For licensees who have submitted a DP, the FSSR need only include the information described under Section 4.5.1.1 (Acceptance Criteria). A licensee who has not submitted a DP should contact the NRC staff to ensure its FSSR includes not only the information below but also any other relevant information the staff needs to carry out its review.

4.5.1.1 Acceptance Criteria 4.5.1.1.1 Regulatory Requirements 10 CFR 20.1402, 20.1403, 20.1501, 30.36(j)(2), 40.42(j)(2), 70.38(j)(2), and 72.54(l)(2) 4.5.1.1.2 Regulatory Guidance NUREG-1575, Multi-Agency Radiological Survey and Site Investigation Manual (MARSSIM) 4.5.1.1.3 Information to Be Submitted The information submitted by the licensee should be sufficient to allow the staff to determine whether the site, area, or building meets the radiological criteria for license termination. The information should include the following:

  • an overview of the results of the FSS
  • a summary of the DCGLs for the facility (if DCGLs are used)
  • a discussion of any changes that were made in the FSS from what was proposed in the DP or other prior submittals
  • a description of the method by which the number of samples was determined for each survey unit 4-16
  • a summary of the parameter values used to determine the number of samples and a justification for these values (e.g., width of the grey region, estimated variance or standard deviation, decision errors)
  • the survey results for each survey unit, including the following:

o the number of samples taken for the survey unit o a description of the survey unit, including (1) a map or drawing of the survey unit showing the reference system and random start systematic sample locations for Class 1 and 2 survey units, and random locations shown for Class 3 survey units and reference areas, (2) a discussion of remedial actions and unique features, and (3) areas scanned for Class 2 and 3 survey units o the measured sample concentrations, in units that are comparable to the DCGLs, and associated MDCs; summaries of laboratory or field measurement reports could be included to provide useful information on measurement error and uncertainty o the statistical evaluation of the measured concentrations, including selection of the null hypothesis, statistical test and associated parameters (e.g., decision errors, critical values) o judgmental and miscellaneous sample data sets reported separately from those samples collected for performing the statistical evaluation o a discussion of anomalous data, including any areas of elevated direct radiation detected during scanning that exceeded the IL or any measurement locations in excess of DCGLW o a statement that a given survey unit satisfied the DCGLW and that elevated areas above the DCGLW were considered in demonstrating compliance with release criteria (e.g., using MARSSIM, Revision 1, Equation 8-2, or other approved method), as applicable4

  • a description of any changes in initial survey unit assumptions relative to the extent of residual radioactivity (e.g., material not accounted for during site characterization)
  • a description of how ALARA practices were employed to achieve final activity levels The information to be submitted is also included as part of the DP Checklist provided in this NUREG report (see Section XIV.e from Appendix D of Volume 1).

4.5.1.2 Review Procedures After review of the FSSR, the NRC reviewer should have reasonable assurance that the FSSR demonstrates that residual radioactivity at the facility complies with the criteria of 4

It is important to note that all radionuclides and sources that could cumulatively contribute to dose to the average member of the critical group should be considered in assessing whether release criteria are met. If multiple contaminated media or elevated areas could cumulatively contribute to dose, then MARSSIM, Rev. 1, Equation 8-2 or another approved method could be used to demonstrate compliance.

See Section 2.7 for additional information.

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10 CFR Part 20, Subpart E. The following guidance discusses the minimal review that should be performed and how the reviewer should select survey units for more detailed review.

Section 4.5.1.1.3 describes the minimum information to be submitted in each FSSR. Additional information about the recommended level of documentation is in Appendix D of this volume. At individual facilities, the NRC reviewer many need additional information on site-specific issues and complex technical topics to evaluate the FSSR. In addition, the NRC reviewer should obtain previous NRC-generated reports on the FSS, including but not necessarily limited to inspections, confirmatory surveys, and any safety evaluation reports that may have addressed the FSS plan.

4.5.1.2.1 Minimal Technical Review The NRC reviewer should review all of the following:

  • the results of previously conducted in-process inspections and confirmatory surveys to verify that the licensee has properly implemented the FSSP and associated procedures
  • the licensees QA/quality control(QC) program, if it has not been previously reviewed
  • changes made to the DP or LTP, if not previously reviewed, to confirm that the changes are not significant and are technically correct
  • specific parts of the FSS and supporting data that affect the FSS that were not available when the DP or LTP was approved (data that may include supplemental characterization results, basis for final surrogate ratios for multiple radionuclides, or other data collected to specifically support the FSS)
  • issues (1) identified by interveners and stakeholders and (2) raised in allegations, to ensure such issues have been satisfactorily resolved
  • descriptions of the survey units, to determine if any special survey situations are present (see Appendix G of this volume for examples)
  • results of elevated measurement comparisons, to confirm that small areas of residual radioactivity do not exceed the appropriate limits (e.g., DCGLEMC) and
  • the results of the appropriate statistical tests (e.g., Wilcoxon Rank Sum (WRS) and sign tests), to confirm that results indicate compliance The purpose of the NRC staff review of in-process inspections, confirmatory surveys, and licensee procedures is to ensure all of the following:
  • the FSSs were implemented in accordance with the approved FSSP
  • judgmental survey results are not used in the statistical tests and are evaluated separately against the release criteria, and survey results obtained using random start and systematic sampling are statistically treated separately for the purpose of demonstrating compliance
  • the QA/QC program was adequate and implemented for the FSS 4-18
  • inadequacies in the FSS design or implementation were corrected (e.g., the licensee improved the overall FSS design and implementation, using information from survey units for which the release criteria were not initially met and resurvey or further remediation was needed)
  • results of confirmatory surveys, including split samples or independent measurements, are consistent with results of licensee surveys
  • appropriate instrumentation, with sufficient sensitivities, proper calibrations, and adequately trained users, was used for surveys, scans, and measurements, as described in the FSSP 4.5.1.2.2 Detailed Technical Review Along with the minimal review described, the NRC reviewer may perform detailed reviews for a number of survey units. The number initially chosen for detailed review should use a risk-informed approach and the results of the minimal review. The reviewer should consider past inspection history, results of confirmatory surveys, the relative difference between residual radioactivity concentration and the associated DCGLs, the complexity of the FSSP, and the radionuclide mix. The detailed review could include confirming the selection process and location of measurements using survey unit maps or floor plans, checking measurement results from laboratory data sheets with those inserted into tables or figures in the FSSR, using parameters that are specific to the survey methodology, and re-creating the appropriate MARSSIM statistical test results.

4.5.1.2.3 Selecting Survey Units for Detailed Reviews Discriminating factors that may be used to select specific survey units for detailed review are listed below. A survey unit that is characterized by one or more of these factors should be considered for potential detailed review. However, the NRC reviewer should focus on survey units for which there are risk-significant issues, issues that are prevalent across a large number of survey units instead of isolated cases, and issues involving an inadequate basis for conclusions. Reviewers do not have to review every instance of a potential issue, particularly if the same issue is identified in multiple survey units. A representative survey unit or survey units can be selected for review.

These factors include any of the following:

  • inconsistencies in defining survey units, including the following:

o size different from recommended size o multiple areas now combined as one larger Class 1 survey unit o Class 3 survey units that are bordered by Class 1 units o survey units bordered by PSR areas o gerrymandered survey unit boundaries

  • application of nonstandard statistical tests (e.g., other than WRS test or Sign test) 4-19
  • significant inconsistencies between the DP/LTP and implemented FSS, including the following examples:

o use of surface and detector efficiencies that do not match survey methods, surface features, and instrumentation used o type of survey instrumentation o sample collection method o laboratory analytical methods o any survey unit where the scan coverage is less than 100 percent for Class 1 areas or less than the minimum commitment for Class 2 or 3 areas o number of samples per survey unit

  • survey units that were remediated
  • survey units for which confirmatory surveys had results inconsistent with the licensees FSS results
  • any Class 2 survey unit with final measurement results near the DCGLW (e.g., greater than 75 percent) or any Class 3 survey unit with significant residual radioactivity (e.g., concentrations greater than 10-25 percent of the DCGLW)
  • any survey unit that was downgraded in classification (i.e., from Class 1 to 2, Class 2 to 3, or Class 1 to 3, or from impacted to nonimpacted)
  • units surveyed before resolution of QA/QC concerns
  • significance of the variability in concentrations (i.e., heterogeneity) across survey units
  • inconsistent approach or inadequate basis for determining surrogate radionuclide ratios
  • significant changes to DP or LTP that affect the FSS or that were not previously reviewed
  • reclassification schemes not approved by the NRC staff
  • use of MARSSIM survey methods and statistical tests when hot particles are present
  • presence of systems and components, buried and embedded piping, or building foundations slated to remain on the site after license termination
  • survey units that combine, for demonstrating compliance, the results of random start or systematic sampling patterns with biased or judgmental survey results
  • a survey unit that involves surveying or sampling of media other than building surfaces and surface soils (e.g., groundwater, surface water, sediments, or deep subsurface soils) 4-20
  • survey units with areas that are hard to access or have abnormal geometries
  • any survey unit that combines survey results with a dose assessment or area factors to demonstrate compliance (mixed approaches are used)
  • the use of nonstandard sampling methods to establish compliance with release criteria (i.e., DCGLs) 4.5.1.2.4 Detailed Review Topics The detailed review could include confirming the selection process and location of measurements using survey unit maps or floor plans, checking measurement results using parameters that are specific to the survey methodology, and re-creating the appropriate MARSSIM statistical test results. In performing detailed reviews, reviewers should consider, but not necessarily be limited to the following questions:
  • Does the FSSR adequately address the issues previously discussed under the selection criteria for detailed reviews, immediately above? For example, if a survey technique was changed from the approved technique, did the FSSR adequately justify the new technique?
  • Are the probabilities of Type I (false positiveincorrectly rejecting the null hypothesis when it is true) and Type II (false negativeincorrectly retaining the null hypothesis when it is false) errors acceptable?
  • If the MDC values are high, does the licensees analysis rely on a large number of results expressed at minimum detectable activity or MDC values (which may invalidate assumptions)?
  • Are all of the static measurement or sampling locations for a survey unit taken from a single random-start sampling set, without substitution (e.g., in cases where additional remediation was performed)?
  • Plot or visualize the data on a location map and ask: is there a discernible trend in the results within and among survey units? Survey units should be fairly homogenous, and this type of visualization can aid in identifying whether elevated areas may have been split across survey units when they were delineated.
  • If there are discernible trends in the results, are the statistical tests appropriate?
  • Were any outliers identi"ed through box plots, control charts, or other acceptable methods? In general, outliers should be considered as part of the data set when evaluating compliance unless the data are found to be invalid (e.g., through reanalysis and/or review of laboratory QA/QC results). If outliers were removed from the data set when evaluating compliance, appropriate technical basis should be provided for removal of the outliers consistent with a data validation and verification process established before the survey occurred (e.g., quality assurance project plan or QAPP).
  • Are there any assumptions about the variability (variance) of the population? For example, if the estimate of variance is inappropriate, then the survey design could be negatively impacted (e.g., low power or inadequate number of samples).

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  • What analytical tools (e.g., statistical software packages) were used to analyze the data?
  • What is the format of the presentation of results? Is it consistent for the survey units reported? For example, are the measurement units consistent with the survey data, the media measured, and the DCGLs?

The detailed review of the initially selected survey units may indicate issues that are prevalent across many units instead of isolated cases. In this case, the reviewer may decide to evaluate additional survey units in detail.

4.5.1.3 Evaluation Criteria The NRC review should determine whether the FSSR is adequate to demonstrate compliance with the radiological criteria for license termination. The reviewer should verify that the licensees FSS results support the conclusion that each survey unit meets the radiological criteria for license termination. The FSS is adequate if it meets the following criteria:

  • MARSSIM Section 5.5.2 for the acceptable number of samples
  • Appendix D of this volume for information on survey data quality and reporting
  • Section A.10 from Appendix A of this volume for information on determining compliance
  • MARSSIM Sections 8.3, 8.4, and 8.5 for interpretations of sample results Issues not Covered in MARSSIM MARSSIMs main focus is on providing guidance for the design of the FSSs for residual radioactivity in surface soils and on building surfaces and evaluating the collected data.

However, several issues related to releasing sites are beyond the scope of MARSSIM.

MARSSIM does not provide guidance for translating the release criterion into DCGLs.

MARSSIM can be applied to surveys performed at vicinity propertiesthose not under licensee controlbut the decision to apply MARSSIM at vicinity properties is outside the scope of MARSSIM. MARSSIM does not address other media (e.g., subsurface soil, volumetrically contaminated building materials, groundwater, surface water, sediments) containing residual radioactivity. Nor does it address the disposition of components and equipment that are not part of the survey unit. Some of the reasons for limiting the scope of the guidance to surface soils and building surfaces include (1) residual radioactivity is limited to these media for many sites following remediation, (2) since many sites have surface soil and building surfaces as the leading sources of residual radioactivity, existing computer models used for calculating the concentrations based on dose or risk generally consider only sources associated with surface soils or building surfaces, and (3) MARSSIM was written in support of cleanup rulemaking efforts for which supporting data are mostly limited to residual radioactivity in surface soils and on building surfaces. Table 4.2 summarizes the scope of MARSSIM. Although this table was taken from MARSSIM, it has been modified to be specific to the needs of NRC licensees.

This volume contains guidance for some topics beyond the scope of MARSSIM. Appendix F has guidance specific to the characterization of groundwater, surface water, and sediments.

Chapter 5 and Appendices H, I, J, K, L, M, and Q contain other guidance pertaining to dose modeling. Guidance can be found in Appendix G for special characterization and survey issues such as subsurface residual radioactivity, embedded piping, sewer systems, and paved areas.

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Table 4.2 Scope of MARSSIM Within Scope of MARSSIM Beyond Scope of MARSSIM Guidance MARSSIM provides technical guidance on Regulation MARSSIM does not establish new conducting radiation surveys and site investigations. regulations or address nontechnical issues (e.g., legal or policy) for site cleanup. Release criteria will be provided rather than calculated using MARSSIM.

Tool Box MARSSIM can be thought of as an extensive Tool Box Many topics are beyond the scope of tool box with many componentssome within the text of MARSSIM, including public participation programs, MARSSIM, others by reference. packaging and transportation of wastes for disposal, remediation and stabilization techniques, and training.

Measurement The guidance given in MARSSIM is Procedure The approaches suggested in MARSSIM performance-based and directed toward acquiring site- vary depending on the various site data needsthere specific data. are no set procedures for sample collection, measurement techniques, storage, or disposal established in MARSSIM.

Modeling The interface between environmental pathway Modeling Environmental pathway modeling and modeling and MARSSIM is an important survey design ecological endpoints in modeling are beyond the consideration addressed in MARSSIM. scope of MARSSIM.

Soil and Buildings The two main media of interest in Other Media MARSSIM does not cover other media, MARSSIM are surface soil and building surfaces with including subsurface soil, surface or subsurface water, residual radioactivity. biota, air, sewers, sediments, or volumetric building residual radioactivity.

Materials or Equipment MARSSIM does not cover disposition of materials (including construction materials) or equipment (see Appendix G, Section G.2.1, of this volume).

Final Status Survey (FSS) The focus of MARSSIM is on Other Survey Types Although not the focus, the FSS, as this is the deciding factor in judging if the site MARSSIM provides less detailed information on meets the release criterion. scoping, characterization, and remedial action support surveys.

Radiation MARSSIM only considers radiation-derived Chemicals MARSSIM does not cover any hazards hazards. posed by chemical contamination.

Remediation Method MARSSIM assists in determining Remediation Method MARSSIM does not discuss when sites are ready for an FSS and provides guidance selection and evaluation of remedial alternatives, on how to determine if remediation was successful. public involvement, legal considerations, or policy decisions related to planning.

DQO Process MARSSIM presents a systemized DQO Process MARSSIM does not provide approach for designing surveys to collect data needed for prescriptive or default values of DQOs.

making decisions such as whether to release a site.

DQA MARSSIM provides a set of statistical tests for DQA MARSSIM does not prescribe a statistical test evaluating data and lists alternative tests that may be for use at all sites.

applicable at specific sites.

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DOSE MODELING EVALUATIONS Introduction Decommissioning plans typically include estimates of the potential future dose that could be caused by the residual radioactivity remaining on the site after decommissioning activities are completed. Calculating potential doses allows both the licensee and regulator to take site-specific information into account in determining acceptable concentrations of residual radioactivity at the site using dose models and exposure scenarios that are as realistic as necessary for the given facility. This section has been written to maintain this flexibility. It includes the evaluation findings and supporting detailed technical guidance necessary to review the licensees dose and ALARA analyses. The discussion on decommissioning groups in Volume 1 of this NUREG series provides guidance on information to be submitted.

Dose modeling information is typically submitted as part of a DP or LTP, although in some cases it may be submitted separately or as part of a FSSR or other document. This chapter usually refers to DPs, although other types of reports are implied, if appropriate. The NRC staff should review all of the dose modeling information submitted by the licensee. For certain cases, such as screening analyses using default values or a lookup table, most of the review has already been completed in developing these tools and, therefore, the licensee need only submit minimal site information and justification in using these models, parameters, and exposure scenarios. In addition, the NRC staff should review the ALARA analyses, which are based, in part, on the dose modeling. Two general approaches exist to provide reasonable assurance that the final concentrations should meet the requirements of Subpart E:

  • The licensee can commit to the exposure scenario(s), model(s), and parameters to be used to evaluate compliance with the dose criterion using the final concentrations. The licensee should project expected final concentrations in the DP to show that there is reasonable assurance that the dose criterion will be met at the time of license termination.
  • The licensee can derive and commit to meeting nuclide-specific concentration limits equivalent to the dose limit.

The Decommissioning and License Termination Framework (Figure 1.2), which generalizes the entire decommissioning process (e.g., Step 7 includes FSS and other requirements related to license termination), provides licensees with guidance on how to perform iterative dose analyses. The NRC staff review of dose modeling consists of evaluations in four general areas:

  • the source and source release assumptions
  • an exposure scenario considering the site environment
  • the mathematical model/computational method used
  • the parameter values and a measure of their uncertainty The actions taken as part of the loop suggested by Steps 8 through 12 of Figure 1.2 can result in the licensee modifying one or more of the above four parts. Licensees, generally, should not 5-1

and do not need to provide information on dose modeling iterations that are not the final dose analyses.

In some cases, licensees may wish to include the iterative process as part of the DP. This is, generally, because site characterization is not initially complete enough to provide reasonable justification for assumptions used in modeling the site. Usually, such incorporation would be in the form of license conditions that need to be satisfied before license termination can occur.

For example, a site may have initial data on residual radioactivity in the groundwater but not have enough data on hydrological conditions to determine which survey units will be affected by the plume. Based on the limited data available, the licensee designates an area around the plume, and all survey units that involve that area will include the dose from the groundwater as part of the overall dose analyses. For the purposes of this example, the NRC could require the licensee, through a license condition (or other mechanism), to continue to characterize its groundwater. If the information confirms that the area affected by the residual radioactivity in the groundwater is the same or smaller than the assumed area, the licensee can proceed with the decommissioning process. If the licensee wishes to take advantage of the smaller area, or the data points to a larger affected area, the licensee may need to submit a license amendment request to modify the FSSP, the dose modeling, and any other area of the DP affected by the new assumed groundwater contamination-affected area (e.g., adding or subtracting survey units from the list that would consider groundwater contributions in complying with Subpart E).1 As described by Figure 1.2 and the preceding example, the areas of dose modeling, site characterization, and FSSP are interdependent. This is an advantage as judicious use of dose modeling can help guide site characterization. In addition, both site characterization and FSSP can guide development of reasonable exposure scenarios or modeling approaches. For example, the appropriate survey techniques may require more advanced modeling in some areas to make them cost effective to implement.

This chapter and the associated appendices use different terms describing exposure scenarios.

Table 5.1 includes a description and comparison of these exposure scenario terms.

1 Licensees should also consider how groundwater transport may affect other survey units or environmental media (e.g., surface water) and appropriately consider uncertainty in the temporal and spatial distribution of radioactivity in the environment. Source remediation may be an option to reduce future downgradient impacts if the future impacts are found to be unacceptable. Because dispersion and dilution are expected to lead to a decrease in peak concentrations away from the source, calculations performed for the source area, if they capture the peak concentrations, may be used to bound the impacts associated with other areas of the site.

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Table 5.1 Comparison and Description of Exposure Scenario Terms Used in this Guidance Exposure Scenario Type Description Compliance Exposure Scenarios (Results Compared to Dose Standards)2 Screening A predetermined exposure scenario that can be used with very high confidence, for most facilities, to demonstrate compliance with the radiological criteria for license termination without further analysis. It generally includes assumptions about land use or human behaviors that attempt to err on the side of higher doses.

The screening exposure scenario for residual radioactivity on building surfaces is the building occupancy, and the screening exposure Plausible Exposure Scenarios scenario for residual radioactivity in surface soils is the residential farmer.

Bounding An exposure scenario with a calculated dose that bounds the doses from other likely exposure scenarios. The building occupancy and residential farmer screening exposure scenarios would represent bounding exposure scenarios for most site-specific analyses for residual radioactivity on building surfaces and in surface or subsurface soils, respectively.

Reasonably Foreseeable Land use exposure scenarios that are likely within the next 100 years, considering current area land-use plans and trends. These exposure scenarios are site-specific.

Other Exposure Scenarios (Results Used to Inform Decisions)

Less Likely but Plausible Land use exposure scenarios that are possible, based on historical uses or trends, but are not likely within the next 100 years, considering current area land use plans and trends. These exposure scenarios are usually site-specific.

Implausible Exposure Scenarios (No Analysis is Required)

Implausible Land uses that, because of physical or other Implausible Exposure compelling limitations, could not occur (e.g., residential land use for an underwater plot of land).

Scenarios 2 Any or all of the compliance scenarios can be used to demonstrate compliance with the radiological criteria for license termination. In general, greater support is needed to demonstrate compliance when using reasonably foreseeable exposure scenarios that have limited pathways, consumption rates, or occupancy times compared to the screening scenarios .

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General Approach for Dose Modeling The following section discusses the basic components that are involved in a dose modeling assessment. It is meant to provide an overview of how the pieces fit together. This section should give both licensees and reviewers a high-level understanding of the big picture review of the dose assessment evaluation. Section 5.3 provides additional details regarding the information that should be submitted with a DP or LTP request, Section 5.4 provides additional detail regarding the acceptance review performed by NRC staff upon receipt of the request, Section 5.5 provides details related to the safety evaluation review, Section 5.6 summarizes review criteria, and Section 5.7 provides a list of additional guidance documents for use in the preparation of the request by the licensee as well as in the review by NRC staffs review.

Chapter 4 of this volume addresses characterization of the residual radioactivity currently present at the site and radiological surveys. The information is based on measurements and knowledge of the site history. To perform dose modeling, the licensee should use the site information on residual radioactivity expected to be present at the completion of decommissioning to develop a generalized view of the sites expected final source configuration3. In developing the source term4 model, the licensee should consider the site measurements, the intended remedial actions, and the needs of both the conceptual model and the FSS.

For example, a site may have a large number of both historical and current measurements characterizing the residual radioactivity over a 10-hectare (25-acre) site. If the site information shows that residual radioactivity levels do not vary significantly, the licensee may assume that the source is a uniform layer of residual radioactivity over the site. If the site information shows significant variability in residual radioactivity concentration, then the licensee may conceptualize the site as two or more sources of residual radioactivity:

  • one or more hot spot sources that represent the area(s) of elevated concentration
  • a source that represents the larger area of residual radioactivity, which contains residual radioactivity at a lower concentration compared to the elevated area After the source configuration has been determined, the question becomes, How could humans be exposed either directly or indirectly to residual radioactivity? or What is the appropriate exposure scenario? Each exposure scenario should address the following scenario questions:
  • How does the residual radioactivity move through the environment?
  • Where can humans be exposed to the environmental concentrations?

3 Source configuration refers to the geometry of the source (e.g., shape, including thickness), as well as the distribution of residual radioactivity (e.g., homogenous versus nonhomogeneous).

4 The source term characterizes the release rate of radionuclides from the source zone. The source term is a function of the inventory and the release mechanism (e.g., solubility controlled, desorption, or diffusion).

RESRAD-ONSITE, RESRAD-BUILD, and DandD have built-in release mechanisms and models, while RESRAD-OFFSITE offers several options to define the source term. Section 5.5 contains more information. Note that the definition of source term in this volume is slightly different than the definition of source term found in the NRC glossary. The definition found in the online NRC glossary, at https://www.nrc.gov/reading-rm/basic-ref/glossary/source-term.html, is specific to accidents involving radioactive materials: types and amounts of radioactive or hazardous material released to the environment following an accident.

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  • What is/are the likely land use(s) in the future for these areas?
  • What are the exposure groups habits that will determine exposure? (What do they eat and where does it come from? How much? Where do they get water and how much?

How much time do they spend on various activities?)

In most situations, there are numerous possible exposure scenarios in which future human exposure groups could interact with residual radioactivity. The compliance criteria in 10 CFR Part 20 for decommissioning does not require an investigation of all (or many) possible exposure scenarios; its focus is on the dose to members of the critical group for the compliance exposure scenario. The critical group is defined (at 10 CFR 20.1003, Definitions) as the group of individuals reasonably expected to receive the greatest exposure to residual radioactivity for any applicable set of circumstances. The compliance exposure scenario is the exposure scenario that leads to the largest peak dose to the average member of the critical group from the mixture of radionuclides. It may be based on a bounding exposure scenario, a screening exposure scenario, another exposure scenario using conservative assumptions about land uses or behaviors, or a scenario considering reasonably foreseeable future land uses for the area.

If the licensee bases its compliance exposure scenario on reasonably foreseeable land use scenarios which are not clearly bounding, the licensee should also identify less likely but plausible land use scenarios. These are scenarios that could lead to higher doses compared to the reasonably foreseeable land use scenario used to demonstrate compliance with the LTR criteria. The evaluation of less likely but plausible exposure scenarios ensures that, if land uses other than the reasonably foreseeable land use were to occur in the future, unacceptably high risks would not result.

By combining knowledge about the sources of residual radioactivity and the exposure scenario questions, the analyst can develop exposure pathways. Exposure pathways are the routes that residual radioactivity may take in traveling from the source, through the environment, to the receptor. Exposure routes can be fairly simple and direct (e.g., residual radioactivity in surface soil emits gamma radiation, which results in direct exposure of an individual standing on the soil), or exposure routes can be fairly complex and indirect (e.g., residual radioactivity in the surface soil leaches to unsaturated soil layers and is transported to the underlying aquifer and water from the aquifer is extracted for use as drinking water, which results in exposure to individuals ingesting groundwater). Exposure pathways typically fall into three principal categories identified by the manner in which the exposed individual interacts with residual radioactivity present in environmental media: ingestion, inhalation, and external radiation exposure.

The exposure pathways for many of the exposure groups can be bounded by a smaller number of possible exposure groups. For example, at a rural site with surface soil residual radioactivity, two possible exposure groups are (1) a resident gardener who grows a small fraction of his or her fruits and vegetables in the soil and (2) a resident farmer who grows a larger fraction of his or her own food (i.e., the site supplies not only vegetables but also meat and milk). In this case, the resident farmer scenario could bound the gardener exposure scenario because the 5-5

exposure pathways and specific parameter values associated with the gardener scenario are bounded by the considerations for the resident farmer.5 As required by 10 CFR 20.1402, Radiological criteria for unrestricted use, expected doses are evaluated for the average member of the critical group, whose characteristics differ from those of the maximally exposed individual. This is not a reduction in the level of protection provided to the public but is an attempt to emphasize the uncertainty and assumptions needed in calculating potential future doses while limiting boundless speculation on possible future exposure scenarios. While it is possible to actually identify, with confidence, the most exposed member of the public in some operational situations (e.g., through monitoring, time-studies, distance from the facility), identifying the specific individual who should receive the highest dose some time (up to 1,000 years) in the future is impractical, if not impossible. Speculation on his or her habits, characteristics, age, or metabolism could be endless. The use of the average member of the critical group acknowledges that any hypothetical individual used in the performance assessment is based, in some manner, on the statistical results from data sets (i.e., the breathing rate is based on the range of possible breathing rates) gathered from groups of individuals. While bounding assumptions could be used to select values for each of the parameters (e.g., the maximum amount of meat, milk, vegetables, possible exposure time), the result could be an extremely conservative calculation of an unrealistic exposure scenario and could lead to excessively low allowable residual radioactivity levels.

Calculating the dose to the critical group is intended to bound the individual dose to other possible exposure groups. The critical group is a relatively small group of individuals who, due to their habits, actions, and characteristics, could receive among the highest potential doses at some time in the future. By using the hypothetical critical group as the receptor, coupled with prudently conservative models, it is unlikely that any individual would actually receive doses in excess of those calculated for the average member of the critical group. The description of a critical groups habits, actions, and characteristics should be based on credible assumptions, and the information or data ranges used to support the assumptions should be limited in scope to reduce the possibility of adding members of less exposed groups to the critical group. An analysis of the average member of the critical groups potential exposure should also include, in most cases, some evaluation of the uncertainty in the parameter values used to represent physical properties of the environment.

The definitions in 10 CFR Part 20 should be used when calculating dose to demonstrate compliance with the requirements of Subpart E. The intake-to-dose conversion factors from Federal Guidance Report No. 11, Limiting Values of Radionuclide Intake and Air Concentration and Dose Conversion Factors for Inhalation, Submersion, and Ingestion, issued September 1988 (EPA 520/1-88-020) (EPA, 1988b), which are based primarily on adults, should be used when calculating internal exposures. As stated in EPAs Federal Radiation Protection Guidance for Exposure of the General Public, dated December 23, 1994, implementing age and sex dependent limits for the general public would be difficult. Not only would the uncertainty associated with the impact of age and sex be significant, but the detailed consideration of age and sex is generally unnecessary. Nonetheless, more recent dosimetry 5 The statement that the resident farmer scenario bounds the resident gardener scenario is true with respect to plant and animal ingestion or consumption rates and pathways; however, it may not be true with respect to the drinking water or irrigation pathways. For example, higher groundwater pumping rates assumed for the resident farmer may, in certain models, lead to a decrease in groundwater concentrations extracted from a well, due to greater dilution from additional clean water being pulled into the well. Therefore, in certain circumstances, the residential farmer scenario may not bound the residential gardener scenario.

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methods could be acceptable for use and will be evaluated on a case-by-case basis (see Appendix I, Section I.6.3 for additional detail).

Since age-based dose conversion factors are not being used, the same dose conversion factors are applied to all individuals. Only in rare exposure scenarios will a non-adult individual receive a higher dose than an adult individual in a similar exposure scenario. One example is the milk pathway: children generally drink more milk annually than adults. If milk were the only pathway that would expose the individual to a dose, then the child would have a slightly higher dose than the adult. In most situations, especially ones involving multiple pathways, the total intake of the adult is greater than that of a child. Therefore, for most multiple pathway exposure scenarios, such as screening analyses, the average member of the critical group should usually be assumed to be an adult with the proper habits and characteristics of an adult. As the licensee eliminates pathways or modifies the exposure scenario, the behavior and dietary habits of children may become important. In such cases, the licensees should contact the NRC staff for guidance.

By integrating the exposure scenario, source configuration, and knowledge about the applicable environmental transport routes involved in the exposure pathways, a conceptual model of the features and processes at the site can be created. The conceptual model is a qualitative description of the important environmental transport and exposure pathways and their interrelationships. Abstraction is commonly necessary to translate the concepts of a conceptual model into mathematical terms. Not only is model abstraction necessary, it can be useful to an analyst in explaining complex system behavior by reducing the system to its major components, thereby facilitating communication on the most important components of the system being simulated.

The extent of the details provided in the conceptual model can impact the extent to which the features and processes described in the conceptual model can be translated into mathematical terms. In particular, mathematical modeling of groundwater transport can be challenging for a variety of reasons. The characteristics and features that introduce challenges include, but are not limited to, the following:

  • limited availability of site-specific data to describe important features, events, and processes for a specific groundwater system
  • limited availability of site-specific data on the source term
  • natural heterogeneity of the site and changes in site characteristics over time
  • complex interactions of the hydrologic, geologic, physical, and chemical processes associated with the system Model abstraction for hydrogeological systems may require the expertise of a qualified specialist. In some cases, the geology of an area may initially appear to be very complicated; however, after analysis, it may be determined that most of the geological detail need not be abstracted. For example, a bedrock aquifer layer may include smaller faults that could potentially serve as preferred pathways. If characterization studies indicate that the faults are filled with low permeable minerals and the intermittent nature of the faults prevents flow from being slowed or diverted, these faults may not need to be included in the conceptual model(s) and represented in the mathematical model. If, on the other hand, site characterization shows that the faults transport water more quickly than the rock matrix, flow through the faults should 5-7

be represented in the mathematical model (e.g., the hydraulic conductivity of the matrix could be increased to represent the impact of small faults).

Going from a conceptual model to a mathematical model involves a number of assumptions and simplifications. For example, one part of a conceptual model of surface soil residual radioactivity involves the leaching of radionuclides through the soil and into the aquifer. In reality, the soil between the surface and the aquifer is usually formed by numerous layers of different types of soils with varying thickness across a site. For the purposes of dose modeling, the conceptual model is more focused on knowing how much activity is entering (and leaving) each major environmental compartment (such as the aquifer) than on precisely predicting the level of activity in the intervening material (e.g., any single soil layer between the surface and the aquifer). Therefore, the mathematical model may view the intervening soil layers as one layer or just a few layers, depending on the difficulty of justifying effective parameters that will mimic the real behavior. Users of off-the-shelf codes should be aware of and consider the appropriateness of the assumptions made in the computer model they are using.

The selection of parameter values (or ranges) for the features, events, and processes of a specific site depends not only on the site conditions and the exposure scenario(s) but also on the computer code (or mathematical model) being used. Nearly any data set will need to be transformed into one appropriate to the situation. This can be as straightforward as generating a sitewide effective soil density value or as complex as converting resuspension factor data into resuspension rates. The NRC has already factored these issues into the data used in the screening analyses, but licensees using site-specific information should justify their values.

The conversion of data into parameter values for use in these analyses requires consideration of uncertainty. In the past, the most common computer codes were deterministic and did not explicitly consider parameter uncertainty. Although it is not always necessary to use a probabilistic code to evaluate parameter uncertainty for site-specific analyses, licensees should provide some discussion of the level of uncertainty in the results and understand the most important factors influencing site-specific parameter values to ensure that there is sufficient information to support the results of the analysis. It should be noted that the type of uncertainty of prime interest to the NRC staff is uncertainty in the physical parameters (e.g., default behavioral and metabolic parameters found in NUREG/CR-5512, Volume 3, can be used with limited justification). Appendices I and Q include additional guidance on considering uncertainty in dose modeling.

Licensees using probabilistic dose modeling should use the peak of the mean dose for demonstrating compliance with 10 CFR Part 20, Subpart E. Similar to all regulatory guidance, this NUREG report contains one approach for determining compliance with the regulations using probabilistic analyses. Use of mean of the peaks is also acceptable for demonstrating compliance. If the mean of the peaks dose is significantly higher than the peak of the mean dose, then risk dilution may be an issue in the probabilistic model. Appendix Q contains more information on risk dilution. If the licensee intends to use any probabilistic approach to calculate DCGLs, it should discuss its planned approach with the NRC staff.

Information to be Submitted Dose modeling information is typically submitted as part of a DP or LTP, though in some cases it may be submitted separately or as part of an FSSR or other document. This information 5-8

should include the licensees assessment of the potential doses resulting from the residual radioactivity remaining at the end of the decommissioning process, as well as a comparison of the potential doses against radiological criteria for license termination found in 10 CFR Part 20, Subpart E. Information needed for performing reviews of DPs varies depending on the decommissioning group. For certain cases, such as screening analyses using default values or a look-up table, most of the review has already been completed in developing these tools and, therefore, the licensee need only submit minimal site information and justification in using these models, parameters, and exposure scenarios. On the other hand, site-specific dose modeling associated with Decommissioning Groups 4 - 7 may require extensive details regarding the source term, exposure scenarios, analytical methods, and other details related to the proposed action.

The information to be submitted is also included as part of the DP Checklist provided in this NUREG report (see Checklist Section V.b from Appendix D of Volume 1).

5.3.1 Decommissioning Groups 1 - 3 (Unrestricted Release Using Screening Criteria)

For DPs for sites associated with Decommissioning Groups 1 - 3, the licensee should provide a description of the exposure scenarios proposed to show compliance with the regulations. The licensees dose modeling for building surfaces or surface soil using the default screening criteria should include the general conceptual model (for both the radiological contaminants of concern and the building or outside environment) of the site and a summary of the screening method used (i.e., running DandD6 or using the lookup tables [see Appendix H]).

The information to be submitted is also included as part of the DP Checklist provided in this NUREG report (see Checklist Section V.a.1. (Building Surface Residual Radioactivity) or V.a.2.

(Surface Soil Residual Radioactivity) from Appendix D of Volume 1).

5.3.2 Decommissioning Groups 4 - 5 (Unrestricted Release Using Site-Specific Information)

In addition to providing information on the exposure scenarios, the licensee should also provide information on the parameters used in the site-specific analysis. Site-specific is used in a general sense to describe all dose analyses except those based only on the default screening tools. This may be as simple as a few parameter changes from the default values in the DandD computer code to licensees using exposure scenarios, models, and parameter values that are only applicable at the licensees site. The information submitted should include the following:

  • site-specific source term information, including nuclides of interest, configuration of the source, areal variability of the source, release mechanisms, and so forth
  • a description of the compliance exposure scenario, including a description of the critical group
  • a description of any other reasonably foreseeable or less likely but plausible exposure scenarios considered 6 Decommissioning codes such as DandD can be obtained from NRCs RAMP (Radiation Protection Computer Code Analysis and Maintenance Program).

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  • a description of the conceptual model for the specific site, including the source term, the physical features important to modeling the transport pathways, and the critical group
  • identification, description, and justification of the computer code(s) and/or mathematical model(s) used (e.g., hand calculations, DandD Version 2 (or later version), RESRAD family of codes)
  • a description of the parameters and the basis for the parameter values used in the analysis
  • a discussion about the effect of uncertainty on the results
  • input and output files or printouts, if a computer program was used The information to be submitted is also included as part of the DP Checklist provided in this NUREG report (see Checklist Section V.b from Appendix D of Volume 1).

5.3.3 Decommissioning Group 6 (Restricted Release)

The majority of the information needed for Decommissioning Group 6 is the same as what would be needed for Decommissioning Groups 4 - 5. In addition to the information listed above for Decommissioning Groups 4 - 5, details related to potential impacts associated with a loss of institutional controls and an ALARA analyses should be included.

The information to be submitted is also included as part of the master DP Checklist provided in this NUREG report (see Section V.c from Appendix D of Volume 1).

5.3.4 Decommissioning Group 7 (Alternate Release Criteria)

The same information provided for restricted release should also be provided for release using alternate criteria (see Section 5.3.3). Additionally, the licensee should provide information on the basis for alternative criteria, including an ALARA analysis. The types of information provided are included in NUREG-1757, Volume 1, Chapter 17.8 and Appendix M, as well as SECY-03-0069 and Regulatory Issue Summary 2004-08.

The information to be submitted is also included as part of the master DP Checklist provided in this NUREG report (see Section V.d from Appendix D of Volume 1).

Acceptance Review Upon receipt of a DP, NRC staff should perform a high-level review of the contents of the submittal in an effort to organize the review and evaluation process. The NRC staff should organize the review by first looking at the overall scope of the dose modeling provided (possibly for several decommissioning options or critical groups, or both). This review should determine which decommissioning group and corresponding review criteria should be used and which specific dose modeling sections are applicable for the given DP.

One acceptable way to organize this initial review is (1) to identify and confirm the principal sources of residual radioactivity (before and after remediation) and (2) to identify the decommissioning goal of the DP. Coupling these two sets of information, the NRC staff should have a good indication of what sections of the guidance apply. For decommissioning goals 5-10

involving unrestricted release, the NRC staff should quickly evaluate the appropriate decommissioning group to which the licensee belongs (i.e., Decommissioning Groups 1-5).

Sections 5.4 and 5.5 are organized by review step with subheadings in each section organized by decommissioning group. Therefore, it is important to determine the appropriate decommissioning group to effectively use the guidance in this chapter.

The NRC staff should confirm that conditions at the site are consistent with the approach chosen by the licensee and the decommissioning groups requirements (e.g., whether the conditions of the site are consistent with the modeling assumptions inherent in the screening analysis approach). A screening approach is generally inappropriate for sites exhibiting any of the following conditions (excluding those caused by sources of background radiation):

  • subsurface soil residual radioactivity (screening analyses assume only surface soil residual radioactivity)
  • radionuclide residual radioactivity present in an aquifer
  • buildings with volumetrically contaminated material
  • radionuclide concentrations in surface water sediments
  • sites that have an infiltration rate that is greater than the vertical saturated hydraulic conductivity (i.e., resulting in the water running off the surface rather than only infiltrating into the ground)

These conditions are inappropriate because they are inconsistent with the conceptual models used in developing the screening values. In other words, the conceptual model, parameters, and exposure scenarios used by the DandD computer code to conduct screening analyses and develop the screening tables presented in Appendix H, are generally incompatible with such conditions. Situations do exist where licensees may still use the screening analyses even if one or more of the conditions listed in the bulleted list above is applicable. For example, by conservatively assuming buried radioactive material is excavated and spread across the surface, the screening criteria may be applicable for use at the site. However, if exceptions are made, it is important to understand the underlying assumptions used to develop the screening values to ensure that the risk is not underestimated (e.g., if screening values are used for buried residual radioactivity it is important to consider the thickness of residual radioactivity and other factors to ensure that the use of surface soil screening values based on 15 cm of residual radioactivity adequately or conservatively represents the risk from the buried residual radioactivity).

As part of the acceptance review, NRC staff should review the DP to ensure that the licensee or responsible party has included the necessary information and determine if the level of detail appears to be adequate for a detailed technical review. This acceptance review should include a general review of the DP table of contents and the individual sections of the submittal. NRC staff should also review the dose modeling portion of the DP but does not need to assess the technical accuracy or completeness of the information contained therein, which should be determined during the detailed technical review if the DP is accepted for review. The NRC staff should also verify that the licensee provided enough information to allow an independent 5-11

evaluation of the potential dose resulting from the residual radioactivity after license termination and reasonable assurance that the decommissioning option will comply with regulations.

Specific considerations for acceptance reviews vary based on the decommissioning group associated with the site and the proposed path forward. The following subsections provide specific details regarding what should be considered based on whether NRC staff will be performing a review for a site using screening criteria, a review for sites using site-specific information and seeking unrestricted release, a review for sites using site-specific information and seeking restricted release or a review for sites using site-specific information and seeking to use alternate criteria.

5.4.1 Decommissioning Groups 1 - 3 (Unrestricted Release Using Screening Criteria)

Decommissioning sites included in Decommissioning Groups 1 - 3 can be evaluated using screening criteria. Decommissioning Group 1 includes sites in which licensed material was used in a way that precluded the release of radioactivity into the environment and would not be expected to have contaminated areas above the screening criteria. These sites generally include licensees that possessed and used sealed sources. These sites do not require a decommissioning plan and dose modeling is not performed. Decommissioning Groups 2 and 3 include sites where residual radioactive contamination is present on building surfaces and in soils, and licensees are typically able to demonstrate that their facilities meet the provisions of 10 CFR 20.1402 (Radiological criteria for unrestricted use) using a screening approach.

Decommissioning Group 2 sites do not need to submit a DP since specific cleanup activities and procedures consistent with remediating the facility are already included in their current license. Decommissioning Group 3 sites require amendments to their license to incorporate necessary remediation procedures needed to decommission the site. Specific details related to these groups are included in NUREG-1757, Volume 1.

When evaluating a DP for a site in Decommissioning Groups 1 - 3, NRC staff should review the DP to ensure that, at a minimum, it contains the information needed to make conclusions regarding the DPs compliance with 10 CFR 20.1402. For residual radioactivity on surface soils, the residential farmer scenario is considered, while for residual radioactivity on building surfaces, the building occupancy exposure scenario is considered. For screening analyses, the licensee should provide information for NRC staff to evaluate the appropriateness of use of a screening approach (e.g., information on the expected source configuration should be provided to allow staff to evaluate whether screening assumptions related to the source are met). The staff should also perform a high-level review of the dose modeling portion of the DP without assessing the technical accuracy and completeness of the information contained therein.

Specific details will be evaluated during the more detailed technical review.

5.4.2 Decommissioning Groups 4 - 5 (Unrestricted Release Using Site-Specific Information)

Decommissioning Groups 4 - 5 include sites that are being considered for unrestricted release using site-specific dose analyses. Decommissioning Group 4 includes sites that are not found to include residual radioactivity in the groundwater while Decommissioning Group 5 sites include residual radioactivity in the groundwater. Additional details on these decommissioning groups are included in NUREG-1757, Volume 1.

Considerations when assessing the path forward for decommissioning actions with a goal of terminating the license under the unrestricted release requirements of 10 CFR 20.1402, include 5-12

the primary exposure scenarios for the individuals exposed on the site. Exposure scenarios should be consistent with reasonably foreseeable land uses over the next decades, or use a bounding exposure scenario, such as a resident farmer. The residential farmer scenario is typically a bounding exposure scenario for residual radioactivity in the environment because this group includes a nearly comprehensive number of exposure pathways (a footnote in Section 5.2 explains how, in certain cases, a residential gardener scenario may lead to higher doses than a residential farmer scenario). In addition to pathways that may be limited by land use assumptions, site conditions, such as soil type or groundwater quality, may remove potential exposure pathways from consideration with the appropriate level of justification by the licensee.

In rare instances, an exposure scenario involving offsite use of residual radioactivity may be the critical exposure scenario (e.g., buried contamination that is transported offsite via surface water where a critical group with unique exposure pathways or subsistence land use could be exposed).

When performing an acceptance review on a DP for a site in Decommissioning Groups 4 and 5, NRC staff should ensure that, at a minimum, information on the source term 7, exposure scenario(s), conceptual model(s), numerical analyses (e.g., hand calculations or computer models), and uncertainty have been included to allow the NRC staff to make conclusions regarding the ability of the licensee to comply with 10 CFR 20.1402. The NRC staff may also perform a high-level review of the assumptions regarding the source term, the conceptual model of the site or building as appropriate, the exposure scenario(s), the mathematical method employed, and the parameters used in the analysis and their uncertainty.

5.4.3 Decommissioning Group 6 (Restricted Release)

Decommissioning Group 6 sites are considered for restricted release and therefore conclusions are made regarding the ability of the site to comply with 10 CFR 20.1403 at the time of license termination based on the information provided in the DP. In addition to the information listed above for Decommissioning Groups 4 - 5, the DP should consider, at a minimum, two different sets of exposure scenarios. One set of exposure scenarios should evaluate the dependence on the proposed institutional controls or restrictions by assuming the institutional controls are effective. Depending on where the residual radioactivity is and what the proposed restrictions are, the exposure location(s) for the critical group could be either onsite or offsite. The second set of exposure scenarios should assume that institutional controls are no longer in effect in accordance with 10 CFR 20.1403(e) (i.e., institutional controls put in place by the licensee have failed to work properly, or effectively, and that the site will be used without knowledge of the presence of residual radioactivity). Although various times can be evaluated in sensitivity analyses, institutional controls should be assumed to be ineffective immediately after license termination.

Chapter 6 of NUREG-1200, Standard Review Plan for the Review of a License Application for a Low-Level Radioactive Waste Disposal Facility, Revision 3, issued in April 1994, can be used as a guideline on the development of site-specific acceptance review criteria for restricted release as applicable (e.g., radioactive release and transport, intruder protection, erosion protection during the post-closure phase, which may be applicable for restricted release scenarios involving offsite receptors or use of engineered barriers). Additionally, Appendix J of 7 The source term characterizes the release rate of radionuclides from the source zone. The source term is a function of the inventory and the release mechanism (e.g., solubility-controlled, desorption, or diffusion-limited release). RESRAD-ONSITE, RESRAD-BUILD, and DandD have built-in release models, while RESRAD-OFFSITE offers several options to define the source term. Appendix I, Section 2, provides more information.

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this document contains information for use when considering intrusion scenarios for buried radioactivity.

Similar to the acceptance review for Decommissioning Groups 4 - 5, NRC staff should ensure that, at a minimum, information on the source term, exposure scenario(s), conceptual model(s),

numerical analyses (e.g., hand calculations or computer models), and uncertainty have been included. The NRC staff may also perform a high-level review of the assumptions regarding the source term, the conceptual model of the site or building as appropriate, the exposure scenario(s), the mathematical method employed, and the parameters used in the analyses and their uncertainty.

5.4.4 Decommissioning Group 7 (Alternate Release Criteria)

Decommissioning Group 7 sites are evaluated using a licensees proposed alternate criteria.

Section 5.4.3 for Decommissioning Group 6 and Chapter 6 of NUREG-1200, Standard Review Plan for the Review of a License Application for a Low-Level Radioactive Waste Disposal Facility, Revision 3, issued in April 1994, may be used as guidelines on the development of site-specific acceptance review criteria for alternate decommissioning criteria, as applicable.

When evaluating sites in Decommissioning Group 7 NRC staff should review the dose modeling information provided in the DP pertaining to the licensees proposed alternate criteria. It should use the findings and conclusions of the review under this section to evaluate the DPs compliance with 10 CFR 20.1404, Alternate Criteria for License Termination. The staff should ensure that, at a minimum, information on the source term, exposure scenario(s), conceptual model(s), numerical analyses, and uncertainty have been included. The NRC staff may also perform a high-level review of the assumptions regarding the source term, the conceptual model of the site or building as appropriate, the exposure scenarios, the mathematical method employed, and the parameters used in the analyses and their uncertainty. The NRC staff should also review the public health and safety and protection of the environment as the basis for the alternate criteria.

Safety Evaluation Criteria and Review When performing a safety evaluation, NRC staff should review the technical content of the information provided by the licensee to ensure that the licensee used defensible assumptions and models to calculate the potential dose to the average member of the critical group. The staff should also verify that the licensee provided enough information to allow an independent evaluation of the potential dose resulting from the residual radioactivity after license termination and to allow NRC staff to decide whether it has reasonable assurance that the decommissioning option will comply with license termination regulations in 10 CFR Part 20, Subpart E. A general outline and template for the development of a safety evaluation report is provided in NUREG-1757, Volume 1, Appendix G.

Specific considerations for the safety evaluation vary based on the decommissioning group associated with the site and the proposed path forward. The following subsections provide specific details regarding what should be considered based on whether NRC staff will be performing a screening review, a site-specific review resulting in the release of a site for unrestricted use, a site-specific review resulting in the release of a site for restricted use, or the release of a site using alternate criteria.

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5.5.1 Screening Safety Review Evaluation Criteria Evaluation considerations will vary depending on whether the submittal uses only default screening methods and parameters, or if the licensee proposes the use of site-specific parameter values. When performing a safety evaluation of a licensees submittal proposing the use of a screening criteria NRC staff should determine if the screening criteria were used correctly by the licensee and whether the calculations provide reasonable assurance that potential doses would not exceed dose limits. If site-specific parameters are used, the analysis is not considered a screening analysis and Sections 5.5.2 through 5.5.4 should be consulted.

When licensees use the default screening methods and parameters inherent in the DandD code by either running the computer code or using lookup tables (see Appendix H), the NRC staff will have already reviewed and accepted nearly all areas of the analysis in developing the screening tool and should only need to review the source concentrations and distribution of residual radioactivity and the overall applicability of using the screening method with the associated residual radioactivity.

Specific considerations associated with the evaluation of surface soils and building surfaces, discussed below, also need to be reviewed. If the licensee did not directly calculate the dose from residual radioactivity but instead derived, or proposed to use, lookup tables to derive DCGLs, the licensee should provide the basis or some discussion for their use. In cases where sufficient support is provided for a proposed approach, NRC staff would only need to review the information on the configuration of the residual radioactivity and the appropriate screening criteria section, below. For licensees who plan to use screening criteria, residual radioactivity should be reasonably represented by a homogeneous source (i.e., the source should not be overly heterogeneous). In general, there should be no elevated areas of residual radioactivity (e.g., hot spots) above the screening values. DandD is limited in its capability to consider the sensitivity of dose with respect to the area of residual radioactivity. However, if one or more elevated areas or hot spots of residual radioactivity are expected to be present at the site following decommissioning, these areas will also need to be considered in demonstrating compliance with radiological criteria for license termination. Appendix I, Sections I.2.3 and I.3.3.3, contain additional guidance on considering elevated areas of residual radioactivity. If DCGLs are developed for the site, it is possible to consider elevated areas of residual radioactivity through use of DCGLEMCs and address these areas during the FSS. Because it is more difficult to develop DCGLEMCs when screening approaches are used, it is important for the licensee to engage NRC staff as early as possible on acceptable approaches for deriving these DCGLs or considering elevated areas.

When reviewing a screening analysis for building surfaces and surface soils NRC staff should use the following criteria:

  • Source Configuration The NRC staff should confirm that the actual measurements, facility history, and planned remedial action(s) support the source configuration used in the modeling by reviewing the portions of the DP on facility history, radiological status, and planned remedial action(s).

When evaluating building surfaces the NRC reviewer should verify both the areal extent of residual radioactivity and the depth of penetration of the residual radioactivity into the building surfaces. The NRC reviewer should also determine if the physical configuration 5-15

of the residual radioactivity can adequately be assumed to be a thin layer of residual radioactivity on the building surfaces.

Similarly, when evaluating surface soils NRC staff should review both the areal extent of residual radioactivity and the depth of penetration of the residual radioactivity into the soil. The reviewer should determine if the physical configuration of the residual radioactivity can adequately be assumed to be a layer of surface soil containing residual radioactivity (i.e., no subsurface radioactivity).

If, during the review, it is determined that the residual radioactivity is not limited to the building surfaces or the surface soil then use of the default screening criteria is not warranted without additional justification. The NRC reviewer should reclassify the licensee as Group 4 and evaluate the modeling using Section 5.5.2.

  • Residual Radioactivity Spatial Variability The NRC staff should review the licensees information on conditions before and those projected after the decommissioning alternative is complete. Based on this information, the NRC should determine whether it is appropriate to assume homogeneity (1) for the whole facility or (2) for subsections of the facility when evaluating building surfaces.

Similarly, for surface soils, NRC staff should review the licensees information to determine whether it is appropriate to assume homogeneity (1) for the entire affected area or (2) for major subsections of the site. The NRC staff should then review the adequacy of the licensees determination of a representative value (or range of values) for the residual radioactivity concentration representing the source(s). For elevated areas, the reviewer could use the general concepts related to DCGLEMCs and the more detailed guidance on considering elevated areas of residual radioactivity discussed in Sections I.2.3 and I.3.3.3 of Appendix I.

Evaluation of the heterogeneity of soil and building surface residual radioactivity is important because the screening dose modeling code assumes relatively homogeneous sources. Heterogeneity can be assessed through spatial and statistical evaluation of the data, including visualization of the data as described in Section G.3.1 of Appendix G. If heterogeneous sources exist, the risk of elevated areas within the survey unit(s) can be assessed through dose modeling. Other methods to mitigate detrimental impacts associated with variability in source concentrations include selection of the survey unit with more homogenous residual radioactivity and dose modeling to assess the risk associated with the actual distribution of residual radioactivity remaining at the site.

  • Conceptual Models A detailed review of the conceptual model is not necessary as the NRC staff addressed these topics when it established the default screening methods. However, the reviewer should verify that the site and DandDs conceptual models are compatible. Situations that would not allow the use of the DandD code as a screening tool would include those where the source is not predominantly on building surfaces (i.e., volumetric source) or use of the building could lead to higher predicted doses compared to the building occupancy exposure scenario. A list of screening values for beta and gamma emitters can be found in Appendix H, Table H.1.

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In the case of surface soil evaluations, situations that would not allow use of the DandD code as a screening tool would include those where the source is not predominantly present in the surface soil, residual radioactivity is in the aquifer, or sites with infiltration rates higher than the vertical saturated hydraulic conductivity (i.e., resulting in surface runoff or a bathtub effect). Alternatively, additional information could be provided to support a determination that the results tend to overestimate the dose. A list of screening values can be found in Appendix H, Table H.2.

  • Execution of the DandD Computer Code Dose Calculations If the licensee has used the DandD computer code to calculate the dose based on either current concentrations or projected final concentrations, the NRC staff should verify the following items, as applicable:

o The residual radioactivity is limited to building surfaces or surface soil.

o The total dose calculated includes all sources of residual radioactivity.

o The output reports verify that no parameters (other than source concentrations) were modified.

o The licensee has used the 90th percentile (or higher percentile) of the dose distribution to compare with the dose limit.

o If the appropriate annual peak dose is greater than 0.025 mSv (2.5 mrem), the removable fraction of the residual radioactivity is 10 percent or less at the time of license termination, or the removable fraction has been adjusted, as explained in footnote a in Table H.1. Note: This item only applies to residual radioactivity on building surfaces.

  • DCGLs from the DandD Code or Lookup Tables The licensee may use either the DandD computer code or the published lookup tables for building surfaces in Appendix H, Table H.1, or for surface soils in Appendix H, Table H.2, to establish radionuclide-specific DCGLs equivalent to 0.25 mSv/y (25 mrem/y). If the licensee proposes to use radionuclide-specific DCGLs, the NRC staff should verify that the following conditions are true, as applicable:

(1) The residual radioactivity is limited to building surfaces or surface soil.

(2) If more than one radionuclide is involved, there is reasonable assurance that the sum of fractions (concentrations divided by DCGLs) (see Section 2.7) is no greater than 1.

(3) For building surfaces, if the residual radioactivity is greater than 10 percent of the respective screening DCGLs (Table H.1 from Appendix H of this volume), the removable fraction is 10 percent or less at license termination, or the removable 5-17

fraction has been adjusted, as explained in footnote a in Table H.1.8 Note: This item only applies to residual radioactivity on building surfaces.

If the licensee has used the DandD computer code to calculate the radionuclide-specific DCGLs, the NRC staff should also verify that the following two conditions are true:

(1) The output reports verify that no parameters (other than entering unit concentrations) were modified.

(2) The licensee has used the 90th percentile (or higher percentile) of the dose distribution to derive the DCGLs.

  • Compliance with Regulatory Criteria The licensees projections of compliance with regulatory criteria are acceptable provided that the NRC staff has reasonable assurance that at least one of the following is true:

(1) The final concentrations result in a peak annual dose of less than 0.25 mSv (25 mrem) and the licensee has committed to calculating the annual dose using a screening analysis at license termination.

(2) The planned DCGLs are equal to or less than those provided by the screening criteria, and the licensee has committed to ensuring the sum of fractions is no greater than 1, if applicable.

5.5.2 Evaluation Criteria for Decommissioning Groups 4 - 5 (Unrestricted Release Using Site-Specific Information)

The NRC staff should determine the acceptability of the licensees projections of radiological impacts from residual radioactivity on the average member of the critical group during the compliance period. The information in the DP is acceptable if it is sufficient to ensure a defensible assessment of the possible future impacts from the residual radioactivity. The licensees assessment can be either realistic or prudently conservative. The information should allow an independent NRC staff evaluation of the assumptions used (e.g., source configuration, applicable transport pathways) and possible doses to the average member of the critical group.

The NRC staff should review the following information, as necessary, for each dose assessment of residual radioactivity that the licensee has submitted for the various decommissioning options.

  • Source Configuration and Release The NRC staff should review the licensees dose modeling source term assumptions and compare them with the current site information and planned remedial activities. The model should be an appropriate abstraction (or simplification) of this information. Four key areas of review for the source term assumptions are (1) source configuration, 8 The DandD default scenario assumes that only 10 percent of the building surface residual radioactivity is removable and available for resuspension. Only at greater than 10 percent of the dose limit does the assumption become important because, in the extreme case of a 100 percent removable fraction for radionuclides whose dose is dominated by the inhalation pathway, the result could only be at most 10 times higher which corresponds to the dose limit of 0.25 mSv/y (25 mrem/y).

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(2) residual radioactivity spatial variability, (3) release mechanisms, and (4) chemical form(s). Section I.2 from Appendix I of this volume provides additional guidance.

(1) Source Configuration The NRC staff should confirm that the actual measurements, facility history, and planned remedial action(s) support the source configuration used in the modeling by reviewing the information in the facility history, radiological status, and planned remedial action(s) portions of the DP. The review should include both the areal extent of residual radioactivity and the depth (for soil or buildings) or volume (for groundwater or buried material) of the residual radioactivity. The NRC reviewer should also determine if the information provided supports the configuration assumptions used in the exposure scenario(s) as well as the computer and mathematical model(s) (e.g., a thin layer of residual radioactivity on the building surfaces).

(2) Residual Radioactivity Spatial Variability The NRC staff should review the licensees residual radioactivity concentration values for conditions both before and projected after the decommissioning alternative is complete. For this subsection, NRC staff should review the spatial extent and the degree of heterogeneity in the values. Based on this information, the NRC staff should determine whether it is reasonable to assume homogeneity for each source for either (1) the whole site or (2) specific subsections of the site. The staff should then review the adequacy of the licensees determination of a representative value (or range of values) for the residual radioactivity concentration in the dose model.

For elevated areas, the reviewer could use the general concepts related to DCGLEMCs and more detailed guidance on considering the elevated areas of residual radioactivity discussed in Sections I.2.3 and I.3.3.3 from Appendix I of this volume.

If the licensee used dose modeling to develop DCGLs instead of estimating final concentrations and if spatial variability is a concern, then the licensee should develop DCGLEMCs and provide this information in the DP or FSS. The NRC staff should verify that spatial variability is adequately considered in DCGL development and cleanup criteria are compatible with the assumptions made for dose modeling.

(3) Release Mechanisms The NRC staff should review the licensees assumptions on mechanisms controlling release of radioactivity from the source. Commonly used decommissioning dose modeling codes such as DandD and the RESRAD family of codes contain built-in release models. RESRAD-OFFSITE offers additional source release models that are not available in DandD and RESRAD-ONSITE. For additional information, see Section I.2.

(4) Chemical Form The NRC staff should review the licensees assumptions about the adequacy of the chemical form of the residual radioactivity and should determine whether the licensee has considered possible chemical changes that may occur during the time period of interest. Without any justification of possible chemical forms, the analysis should use the bounding chemical form(s) (e.g., the chemical form(s) that give the 5-19

individual the highest dose per unit intake, as described in Federal Guidance Report Number 11 (EPA, 1988b)). The licensee should provide an acceptable rationale for other assumptions. Some acceptable rationales for using other chemical forms are (1) chemical forms that would degrade quickly in the environment (e.g., uranium hexafluoride (UF6)) or (2) the unavailability of an element or conditions to realistically form that molecule (e.g., strontium titanate or high-fired uranium dioxide (UO2)9).

  • Critical Groups, Exposure Scenarios, Pathways, Identification, and Selection In its review, the NRC staff should confirm that the licensee has identified and quantified the most significant exposure scenarios based on available site- or facility-specific information, as well as the basis and justification for the licensees selected critical group. For exposure scenarios in which possible environmental pathways have been modified or eliminated, the NRC staff should review the justifications provided by the licensee. Section I.3 has additional guidance on these subjects.

(1) Exposure Scenario Identification The compliance exposure scenario is based on the location and type of source (e.g., contaminated walls), the reasonably foreseeable land use, the general characteristics and habits of the critical group (e.g., an adult light-industry worker),

and the possible pathways that describe how the residual radioactivity would incur dose in humans. The licensee should provide justification for the exposure scenario(s) evaluated.

The licensee should justify the possible land use(s) the site might experience in the future and create exposure scenarios consistent with these uses. The licensee should also justify its selection of a compliance exposure scenario from the possible exposure scenarios derived from the current and projected land uses. The compliance exposure scenario should result in the greatest exposure to the average member of the critical group for all exposure scenarios given the mixture of radionuclides. A licensee may choose to make a bounding assumption for land use to derive the exposure scenario (e.g., assuming a rural land use for an urban location) or base the exposure scenario on the reasonably foreseeable land use that results in the highest dose. The other, less likely but plausible exposure scenarios are considered to better risk inform the decision.

If the compliance exposure scenario is based on the reasonably foreseeable land use, the licensee should justify the exposure scenario based on discussions with land planners, meetings with local stakeholders, trending analyses of land use for the region, or comparisons with land use in similar alternative locations. The time period of interest for possible land use changes is within 100 years, depending on the rate of change in the region and the peak exposure time. Note that the 100-year timeframe described here is only for estimating future land uses; the licensee must evaluate doses that could occur over the 1,000-year time period specified in the LTR.

The licensee should also identify what land uses are less likely but plausible and evaluate exposure scenarios consistent with these less likely but plausible land uses.

If use of reasonably foreseeable land use exposure scenarios results in eliminating a 9 Strontium titanate and high-fired UO2 are relatively insoluble and, therefore, these chemical forms would be expected to be retained in the pulmonary region for longer periods of time, delivering a greater dose to the lung compared to other chemical forms. However, high-fired uranium dioxide is only expected to be created under extremely high temperatures, above 800 degrees Celsius (C), and strontium titanate is considered to be artificially created, although strontium titanate has been found naturally in remote areas of the world.

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significant number of exposure pathways, the licensee may need to evaluate offsite exposure scenarios to ensure they do not result in greater exposures, when demonstrating compliance with the radiological criteria for license termination.

The licensee should provide a quantitative analysis or a qualitative argument discounting the need to analyze all exposure scenarios generated from the reasonably foreseeable land uses. The level of detail can vary between exposure scenarios, and the licensee is expected to use simple analyses to limit the number of detailed exposure scenarios. The licensee may use screening or generic analyses to assist in determining the critical exposure scenario for compliance. With a mixture of radionuclides, more than one compliance exposure scenario may be needed. The peak dose from the exposure scenario(s) with the highest dose should be used to demonstrate compliance.

Similarly, the licensee may provide either a quantitative analysis or a qualitative argument discounting the need to analyze all exposure scenarios considering less likely but plausible land uses. The staff will use the results of these analyses to evaluate the degree of sensitivity of dose to overall exposure scenario assumptions (and the associated parameter assumptions). The reviewer will consider both the magnitude and time of the peak dose from these exposure scenarios. If peak doses from the less likely but plausible land use exposure scenarios are significantly above the dose standard, the licensee would need to provide greater assurance that the exposure scenario is less likely to occur, especially during the period of unacceptably high dose.

The screening exposure scenarios for building surface residual radioactivity and soil residual radioactivity are described in NUREG-1549 and NUREG/CR-5512, Volumes 1, 2, and 3. Dose evaluations that use these exposure scenarios (i.e., the licensee changes parameter values or mathematical method but does not change the general exposure scenario) are acceptable if the exposure scenario is appropriate for the situation. In DPs where the licensee eliminates certain pathways, with justification, but still maintains the same general exposure scenario category, the NRC staff should find the exposure scenario identification to be acceptable. For example, a licensee may eliminate the use of groundwater from consideration if the near surface aquifer has total dissolved solids concentrations that exceed the water quality standards for drinking water and agricultural use. The licensee still evaluates the impacts from crops grown in the residual radioactivity, but irrigation is provided by a noncontaminated source and therefore, the screening exposure scenario of a residential farmer is maintained.

(2) Critical Group Determination In general, critical groups exposed to multiple exposure pathways result in higher doses than groups that have more limited interaction with residual radioactivity.

NUREG-1549 and the NUREG/CR-5512 series detail the critical group assumptions for the screening exposure scenarios. In DPs where the licensee has used the screening exposure scenarios, the reviewer should verify that the proposed critical group satisfies the assumptions listed in NUREG-1549 and the NUREG/CR-5512 series.

The licensee should provide either qualitative or quantitative justification that the critical group is the highest exposed group for the assumed land use(s). The selection of the critical group may be dependent on the assumption of the relative mixture of radionuclides and sources of residual radioactivity present at the site. The 5-21

licensee should justify its compliance approach in these cases, as well as for the critical exposure scenario for less likely but plausible exposure scenarios.

(3) Exposure Pathways The DP should describe the exposure pathways to which the critical group is exposed, except for cases where the licensee is using the screening exposure scenarios and critical groups without modification. If the licensee has chosen to modify the screening exposure scenario, the changes should be justified. In general, the justification should be based on physical limitations or situations that would not allow individuals to be exposed as described in the exposure scenario. The exposure pathways should therefore be consistent with the land use assumptions, exposure group behavior, and physical site conditions.

For example, acceptable justifications for removing the groundwater pathway include the following: (1) the near surface groundwater is neither potable nor allowed to be used for irrigation, (2) the aquifer volume is insufficient to provide the necessary yields, and (3) there are current (and informed consideration of future) land use patterns that would support elimination of the groundwater pathway (e.g., only short-lived radionuclides are present at the site, which is currently located in an industrial section of an urban area with restrictions on groundwater use). Typically, multiple lines of evidence are needed to eliminate the groundwater pathway, particularly if the justification relies heavily on reasonably foreseeable future land use, trends, or restrictions on groundwater use. Groundwater treatment should be considered in the case that the groundwater pathway is eliminated based on quality. If groundwater use is less likely but plausible, then the pathway should be considered to provide risk information that can be factored into the decision-making process as discussed in Chapter 5. Justification of water quality and quantity of the saturated zone should be based on the classification systems used by EPA or the State, as appropriate. In cases where the aquifer is classified as not being a source of drinking water but is considered adequate for stock watering and irrigation but not a viable source of drinking water, the licensee can eliminate (i.e., does not need to consider) the drinking water pathway (and the fish pathwaydepending on the exposure scenario). The licensee would still maintain the irrigation and meat and milk pathways, if appropriate, for the land use assumptions. Finally, affected State, local and tribal governments should be consulted when evaluating whether to approve the elimination of the groundwater pathway from the compliance scenario.

Another example would be a rural site with a relatively small discrete outdoor area of residual radioactivity (compared to the area assumed in the default exposure scenarios). In this situation, it may be appropriate, based on the area of residual radioactivity, that gardening of some vegetables and fruits would still be an assumption, but the area would not be large enough to allow one to grow grain or raise animals for meat or milk.

  • Conceptual Models The NRC staff should review the adequacy of the conceptual model(s) used by the licensee.

The conceptual model should qualitatively describe the following:

o the relative location and activities of the critical group 5-22

o both the hydrologic and environmental transport processes important at the site o the dimensions, location, and spatial variability of the source represented in the model o the major assumptions made by the licensee in developing the conceptual model The NRC reviewer should verify that the licensee adequately addressed the site conditions in the conceptual model and simplifying assumptions. Section I.4 has additional guidance on these subjects.

  • Calculations and Input Parameters In its review, the NRC staff will confirm that the licensee has used a mathematical model that is an adequate representation of the proposed conceptual model and the exposure scenario. Section I.5 contains additional guidance on this subject.

(1) Execution of DandD Computer Code If the licensee uses the DandD computer code in its analysis, the NRC staff should verify the following points:

o The residual radioactivity is limited to the surface (either building or near surface soil, as appropriate).

o The total dose calculated includes all sources of residual radioactivity.

o The site conceptual model is adequately represented by DandDs inherent conceptual model.

o For residual radioactivity present on building surfaces, the licensee has modified the resuspension factor, as necessary, to account for the removable fraction expected to be present at the time of decommissioning. The default removable fraction assumed in DandD is 10 percent. If the removable fraction is expected to be greater than 10 percent, the licensee should account for higher removable fractions that might increase the resuspension factor.10 If site conditions are consistent with assumptions in NUREG-1720, Re-evaluation of the Indoor Resuspension Factor for the Screening Analysis of the Building Occupancy Scenario for NRCs License Termination Rule, with respect to activities and exposure scenarios, ventilation conditions, and low removable fractions at the time of decommissioning, the NUREG-1720 recommended resuspension factor value or parameter distribution may be used with minimal justification.

o For sites eliminating pathways, the licensee has used the appropriate parameters in the DandD code as switches to turn off the pathways without unintentionally removing others. For example, to remove the groundwater pathways, the licensee should set the drinking water rate, irrigation rate, and pond volume parameter values to zero.

10 The default removable fraction is multiplied by the loose resuspension factor in DandD to derive the resuspension factor. Either the removable fraction or the resuspension factor can be adjusted to account for removable fractions greater than the default value of 10 percent.

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o The licensee has provided adequate support for site-specific parameter values or ranges, has not defaulted to default values, and has appropriately considered parameter correlations.

o For modifications of behavioral parameters, the changes should be based on acceptable changes in the critical group, and the mean values of the behavioral parameters should be used, although use of the ranges is also acceptable.

o If the licensee has randomly sampled the parameter ranges in DandD, it has used the peak of the mean or mean of the peaks dose to calculate the dose or derive the DCGLs.

(2) Other Mathematical Methods If the licensee uses other mathematical methods or codes, the NRC reviewer should verify the following:

o The mathematical conceptual model is compatible with the sites conceptual model (e.g., RESRAD-ONSITE would not be an acceptable mathematical method for sites with building surface residual radioactivity).

o For each parameter or parameter set, the licensee has adequately justified the parameter value or range. For modifications of behavioral parameters, the licensee should base the changes on acceptable changes in the critical group and use the mean value (or full range) of the behavior.

o If the inhalation dose can be significant (e.g., due to the presence of alpha-emitting radionuclides such as uranium or thorium), the NRC staff should review the resuspension factor or rate and the assumptions about the degree of removable residual radioactivity.

o If the licensee is performing a probabilistic analysis, it has used the peak of the mean or mean of the peaks dose to either calculate the dose or derive the DCGLs.

  • Uncertainty Analysis The NRC staff should review the licensees discussion of the uncertainty resulting from the physical parameter values used in the analysis. The review should focus on the uncertainty analysis for the critical pathways or parameters. Reviewers should expect that the degree of uncertainty analysis depend on the level of complexity of the modeling (e.g., generally qualitative discussions for simple modeling to quantitative analyses for more complex sites). The overall acceptability of the uncertainty analysis should be evaluated on a case-by-case basis. Section I.7 of Appendix I and Appendix Q contain additional guidance on these subjects.

If the licensee evaluated exposure scenarios based on reasonably foreseeable land uses, it should provide either a quantitative analysis of or a qualitative argument discounting the need to analyze all exposure scenarios generated from the less likely but plausible land uses. The staff will use the results of these analyses to evaluate the degree of sensitivity of the dose to overall exposure scenario assumptions (and the associated parameter assumptions). The reviewer will consider both the magnitude and time of the peak dose from these exposure scenarios. If peak doses from the less likely but plausible land use exposure scenarios are significantly above the dose standards, the licensee would need to provide greater assurance that the exposure scenarios are unlikely to occur, especially during the period of unacceptably high dose.

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  • Compliance with Regulatory Criteria The licensees projections of compliance with regulatory criteria are acceptable, provided that the NRC staff has reasonable assurance of the following:

(1) The licensee has adequately characterized the source and applied a technically defensible source term.

(2) The licensee has analyzed the appropriate exposure scenario(s) and found that the exposure group(s) adequately represents a critical group.

(3) The mathematical method and parameters used are appropriate for the exposure scenario and parameter uncertainty has been adequately addressed.

(4) For deterministic analyses, the peak annual dose to the average member of the critical group for the appropriate exposure scenario(s) for the option is less than (or equal to) 0.25 mSv (25 mrem) or was used to calculate DCGLW.

(5) For probabilistic analyses, the peak of the mean or mean of the peaks dose to the average member of the critical group for the appropriate exposure scenario(s) for the option is less than (or equal to) 0.25 mSv (25 mrem) or was used to calculate DCGLW.

(6) Either one of the following:

o The licensee has committed to using a specific exposure scenario, model, and set of parameters with the final survey results to show final compliance with the dose limit.

o The licensee has committed to using radionuclide-specific DCGLs and will ensure that the total dose from all radionuclides will meet the requirements of Subpart E by using the sum of fractions.

5.5.3 Safety Evaluation Criteria for Decommissioning Group 6 (Restricted Release)

As discussed previously, sites in Decommissioning Group 6 are being considered for restricted release using site-specific dose analyses. Specific details describing the types of sites that can be considered for Decommissioning Group 6 are included in NUREG-1757, Volume 1.

The majority of the criteria used by the NRC staff to assess the acceptability of a site for restricted release are the same as criteria used for approving Decommissioning Groups 4 - 5.

This includes specific areas of consideration, including the source configuration, release mechanisms, and chemical form of the waste. Similarly, likely exposure scenarios and corresponding exposure pathways as well as the modeling approach and possible uncertainty issues are also evaluated. Specific differences that need to be considered when assessing whether the DP is in compliance with 10 CFR 20.1403 include the following:

  • When assessing whether the licensee has identified and quantified the most significant exposure scenarios based on available site- and facility-specific information, dose considerations will need to be made for a minimum of two sets of exposure scenarios.

One set of exposure scenarios addresses the situation when institutional controls are in place and working properly. The other set of exposure scenarios addresses the possible doses that may occur if institutional controls are assumed to no longer be in effect. For purposes of the evaluation, the licensee should assume the institutional controls fail at 5-25

time = 0 years. The NRC staff should review the basis and justification for the licensees selected critical group for each exposure scenario. Section I.3 of Appendix I and Appendix M contain additional guidance on these specific areas.

  • As discussed in the previous section for Decommissioning Groups 4 - 5, evaluations of the compliance exposure scenarios proposed by the licensee includes consideration of the location and type of source (e.g., contaminated walls), the reasonably foreseeable land use, the general characteristics and habits of the critical group (e.g., an adult light-industry worker), and the possible pathways that describe how the residual radioactivity would incur dose in humans. When considering restricted releases associated with Decommissioning Group 6, potential limitations for the use of a specific compliance exposure scenario should also consider limitations based on the established institutional controls associated with restricted release. This is to ensure that if land uses other than the reasonably foreseeable land use was to occur in the future, significant exposures would not result.
  • When establishing exposure scenarios, the licensee may use proposed restrictions as a basis for eliminating or changing specific exposure pathways. For example, for Decommissioning Groups 4 - 5, acceptable justifications for removing the groundwater pathway include the following: (1) the near surface groundwater is neither potable nor allowed to be used for irrigation, (2) the aquifer volume is insufficient to provide the necessary yields, and (3) there are current (and informed consideration of future) land use patterns that would support elimination of the groundwater pathway. For Decommissioning Group 6, specific site restrictions precluding groundwater use (e.g.,

permits, regulations, etc.) could also be justification for the removal of the groundwater pathway from consideration. However, it is important to note that in the case that institutional controls are assumed to no longer be in effect, it may be necessary to evaluate the exposure scenarios and pathways eliminated for the case when institutional controls are assumed to be effective.

5.5.4 Safety Evaluation Criteria - Decommissioning Group 7 (Alternate Release Criteria)

Decommissioning Group 7 sites are evaluated using alternate criteria proposed by the licensee.

An alternative release proposal in accordance with 10 CFR 20.1404 may allow a dose of up to 1.0 mSv/y (100 mrem/y) for baseline conditions with restrictions in place. However, for restricted release sites and specifically for the case where institutional controls are assumed to no longer be in effect, the dose may not exceed the values in 10 CFR 20.1403(e). Furthermore, the other provisions of 10 CFR 20.1403 must also be met.

The material to be reviewed by NRC staff should ensure that the licensee used defensible assumptions and models to establish and demonstrate compliance with the proposed alternate criteria. The staff should also verify that the licensee provided (1) enough information to allow an independent evaluation of the assessment resulting from the residual radioactivity after license termination and (2) reasonable assurance that the proposed decommissioning option complies with regulations. Each evaluation should be performed on a case-by-case basis.

Summary Review Criteria A summary of the review criteria listed in this chapter is provided in Table 5.2 and Table 5.3.

Table 5.2 provides information on review criteria for residual radioactivity found in soils.

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Table 5.3 provides information on review criteria for residual radioactivity associated with buildings. Both Tables 5.2 and 5.3 provide information for screening (column 1) versus site-specific reviews (column 2); and unrestricted (top) versus restricted (bottom) release scenarios.

Staff should review the information provided by the licensee to ensure an adequate basis is provided to support the modeling assumptions and parameters selected commensurate with their risk-significance as determined through sensitivity and uncertainty analysis. Data of sufficient quality should be collected to ensure the technical defensibility of the modeling results and DCGLs. The DQO process should be used to guide data collection and analysis. Staff should review the DQOs, including QA/QC requirements, to ensure DQOs are met during the DCGL development and the FSS used to demonstrate compliance with release criteria.

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Table 5.2 Dose Modeling Review Criteria for Residual Radioactivity in Soils Screening (Soils) Site-Specific (Soils)

Unrestricted Release Review source assumptions Review modeling assumptions related to the

  • Nuclide(s) of interest source term
  • Only surface soil residual radioactivity is
  • Nuclide(s) of interest present (residual radioactivity is present in
  • Area of residual radioactivity approximately the top 30 cm of soil)
  • Thickness of residual radioactivity
  • Extent of heterogeneity of residual
  • Extent of heterogeneity of residual radioactivity (i.e., survey units should be radioactivity relatively homogeneous)
  • Method of determining average
  • Method of determining average concentration concentration
  • Treatment of elevated areas (and development of DCGLEMCs), as appropriate
  • Chemical form of the radionuclide (to ensure dose conversion factors do not underestimate dose; bounding values could be used if information on chemical form is lacking)

Review exposure scenario assumptions Review exposure scenario assumptions

  • Evaluate whether default critical group
  • Assess whether the exposure scenario(s)

(resident farmer) is appropriate for the site and critical group(s) used to demonstrate and whether there are any other critical compliance are appropriate for the site.

groups which could incur higher doses.

  • Assess whether reasonably foreseeable exposure scenarios are considered based on land use and other data, as well as stakeholder input.
  • Assess the adequacy of information provided to support elimination of or reduction in dose associated with various pathways of exposure.
  • Ensure that less likely but plausible scenarios are identified and considered, as appropriate.
  • If subsurface residual radioactivity is present, whether DCGLs are derived for multiple layers and whether excavation scenarios which would bring the material to the surface are considered.

Review conceptual site model assumptions Review conceptual site model assumptions

  • Check to make sure site conditions are
  • Check to make sure site conditions are consistent with the built-in conceptual consistent with the built-in conceptual model in the DandD screening code (see model in RESRAD-ONSITE (see Appendix I, Table I.5) Appendix I, Tables I.6 and I.7) o groundwater initially free of residual
  • Evaluate the relative location and activities radioactivity of the critical group.

surface water sediments initially free of

  • Evaluate the distribution of residual residual radioactivity radioactivity in the environment (is residual 5-28

Table 5.2 Dose Modeling Review Criteria for Residual Radioactivity in Soils (cont.)

Screening (Soils) Site-Specific (Soils) o infiltration rate is not greater than the radioactivity initially present in the vertical saturated hydraulic conductivity subsurface or in groundwater?)

  • Evaluate the hydrogeological conceptual model developed for the site (e.g., depth to groundwater, subsurface layers or materials, aquifer thickness).
  • Evaluate important environmental transport processes.

Review screening method Review mathematical models

  • Evaluate whether screening or look-up
  • Evaluate the mathematical models used to values (see Appendix H) are being used. assess dose (guidance in this volume
  • If screening values are being used and focuses on use of DandD and probabilistic multiple radionuclides are present, check to RESRAD, which were sponsored by the make sure that the sum of fractions value NRC for use in decommissioning) and (concentrations divided by screening compatibility with the site conceptual model.

values) is not greater than 1.

  • If custom models or non-traditional codes
  • The final concentration should result in a are used, evaluate whether the peak annual dose of less than 0.25 mSv mathematical models are appropriate and (25 mrem/) compatible with the conceptual site model and have proper QA/QC (see Appendix I, Section I.5).

Review Parameter Assumptions Review Parameter Assumptions

  • Ensure that the default parameters
  • Review appropriateness of physical developed for DandD (see parameters developed for the site NUREG/CR-5512, Volume 3) are used. (licensees may use DandD default
  • The only parameters that require user input behavioral and metabolic parameters) are the list of radionuclides and their
  • If changes are made to DandD default associated concentrations. behavioral and metabolic parameters, ensure that the changes are well supported for the exposure scenario and average member of the critical group.
  • Evaluate whether sensitivity and/or uncertainty analysis are performed to identify the most risk-significant parameters affecting dose.
  • Ensure that an adequate level of support is provided for the most risk-significant parameters.

Consideration of Uncertainty Consideration of Uncertainty

  • If modeling is performed to develop Evaluate parameter ranges to ensure that screening values, ensure that DandD and parameter uncertainty is appropriately the 90th percentile of the dose distribution is considered in the dose modeling. Default used. parameter distributions are available in RESRAD-ONSITE and can be used to assist with performance of probabilistic sensitivity analysis.

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Table 5.2 Dose Modeling Review Criteria for Residual Radioactivity in Soils (cont.)

Screening (Soils) Site-Specific (Soils)

  • If a deterministic analysis is used, ensure that the values selected for risk-significant parameters are sufficiently conservative.
  • If a probabilistic approach is used, ensure that the parameter ranges are adequately supported particularly for risk-significant parameters and that overly broad distributions do not lead to risk dilution for parameters that primarily affect the timing of peak dose.
  • Evaluate the dose metric used (peak of the mean or mean of the peak)
  • The final concentrations should result in a peak annual dose of less than 0.25 mSv (25 mrem/)

Restricted Release Not applicable All of the above review criteria apply for restricted release. Additional review criteria are listed below.

Evaluate two sets of exposure scenarios for the case when

  • Institutional controls are in effect
  • Institutional controls are no longer in effect (at time=0 years)

Evaluate whether the licensee has appropriately considered offsite receptors and identified the critical group in the case where restrictions are in place.

In the case of loss of institutional controls, evaluate whether the licensee considers degradation of engineered barriers and only the passive performance of the barriers assuming no active maintenance.

Evaluate the use of dose modeling information to support ALARA evaluations.

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Table 5.3 Dose Modeling Review Criteria for Residual Radioactivity in Buildings Screening (Building) Site-Specific (Building)

Unrestricted Release Review source assumptions Review source assumptions

  • Residual radioactivity is present on building
  • Nuclide(s) of interest surfaces (no volumetrically contaminated
  • Area of residual radioactivity building materials)
  • Thickness of residual radioactivity (thin
  • Extent of heterogeneity of residual surface or volumetric building radioactivity contamination)
  • Method of determining average
  • Extent of heterogeneity of residual concentration radioactivity
  • Method of determining average concentration
  • Treatment of elevated areas (and development of DCGLEMCs), as appropriate
  • Chemical form of the radionuclide (to ensure dose conversion factors do not underestimate dose; bounding values could be used if information on chemical form is lacking)

Review exposure scenario assumptions Review exposure scenario assumptions

  • Evaluate whether default critical group
  • Assess whether the exposure (building occupancy) is appropriate for the scenario(s) and critical group(s) used to site and whether there any other critical demonstrate compliance are appropriate groups which could incur higher doses. for site conditions.
  • If volumetric contamination is present, an evaluation of whether the building renovation scenario is more limiting and should be considered.

Review conceptual site model assumptions Review conceptual site model assumptions

  • Check to make sure site conditions are
  • Evaluate the relative location and consistent with the conceptual model built- activities of the critical group.

in the DandD screening code

  • Evaluate the distribution of residual o only building surface residual radioactivity in the building (surface radioactivity is present and versus volumetric; floors, walls, ceiling, o loose contamination is 10 percent or equipment, piping, or sewer lines).

less or the removable fraction has been

  • Evaluate representation of the building in adjusted the model (geometry and building ventilation of rooms, building materials).

Review mathematical models and screening Review mathematical models method

  • Evaluate the mathematical models used
  • Evaluate whether screening or look-up to assess dose (guidance in this volume values (see Appendix H) are being used. focuses on use of DandD and RESRAD-
  • If screening values are being used and BUILD, which were sponsored by the multiple radionuclides are present, check to NRC for use in decommissioning) and make sure that the sum of fractions compatibility with the site conceptual (concentrations divided by screening model.

values) is no greater than 1.

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Table 5.3 Dose Modeling Review Criteria for Residual Radioactivity in Buildings (cont.)

Screening (Building) Site-Specific (Building)

  • If modeling is performed to develop
  • If custom models or non-traditional screening values, ensure that DandD is codes are used, evaluate whether the used. mathematical models are appropriate and compatible with the conceptual site model and have proper QA/QC (see Appendix I, Section I.5).

Review Parameter Assumptions Review Parameter Assumptions

  • Ensure that the default parameters
  • Review appropriateness of parameters developed for DandD (see NUREG/CR- developed for the site (licensees may 5512, Volume 3) are used. use DandD default behavioral and
  • The only parameters that require user input metabolic parameters).

are the list of radionuclides and their

  • Evaluate whether sensitivity and/or associated concentrations. uncertainty analysis are performed to identify the most risk-significant parameters affecting dose.
  • Ensure that an adequate level of support is provided for the most risk-significant parameters.

Consideration of Uncertainty Consideration of Uncertainty

  • If modeling is performed to develop
  • Evaluate parameter ranges to ensure screening values, ensure that DandD and that parameter uncertainty is th the 90 percentile of the dose distribution is appropriately considered in the dose used. modeling. Default parameter distributions are available in RESRAD-BUILD and can be used to assist with performance of probabilistic sensitivity analysis.
  • If a deterministic analysis is used, ensure that the values selected for risk-significant parameters are sufficiently conservative.
  • If a probabilistic approach is used, ensure that the parameter ranges are adequately supported.
  • Evaluate the dose metric used (peak of the mean or mean of the peak)

Restricted Release Not applicable All of the above review criteria apply for restricted release. Additional review criteria are listed below.

Evaluate two sets of exposure scenarios for the case when

  • Institutional controls are in effect
  • Institutional controls are no longer in effect (at time=0 years) 5-32

Additional Regulatory Guidance In addition to the information provided above, there are numerous other references that the licensee or the NRC reviewer can refer to when developing and reviewing the dose modeling portions of DPs. Some of these references are listed below.

5.7.1 Regulatory Guidance Related to Decommissioning Groups 1 - 3 (Unrestricted Release Using Screening Criteria)

  • Appendix H to this NUREG report

Decommissioning Process for Materials Licensees 5.7.2 Regulatory Guidance Related to Decommissioning Groups 4 - 5 (Unrestricted Release Using Site-Specific Information)

  • Appendix I to this NUREG report
  • NUREG-1549, Decision Methods for Dose Assessment to Comply with Radiological Criteria for License Termination
  • NUREG/CR-5512, Volume 1, Residual Radioactive Contamination from Decommissioning: Technical Basis for Translating Contamination Levels to Annual Total Effective Dose Equivalent
  • NUREG/CR-5512, Volume 2, Residual Radioactive Contamination from Decommissioning: Users Manual, DandD Version 2.1
  • Draft NUREG/CR-5512, Volume 3, Residual Radioactive Contamination from Decommissioning: Parameter Analysis
  • EPA, Federal Guidance Report Number 11, Limiting Values of Radionuclide Intake and Air Concentration and Dose Conversion Factors for Inhalation, Submersion, and Ingestion, September 1988
  • EPA, Federal Guidance Report Number 12, External Exposure to Radionuclides in Air, Water, and Soil, September 1993 5.7.3 Regulatory Guidance Related to Decommissioning Group 6 (Restricted Release)
  • NUREG-1549, Decision Methods for Dose Assessment to Comply with Radiological Criteria for License Termination
  • NUREG-1573, A Performance Assessment Method for Low-Level Waste Disposal Facilities: Recommendations of NRCs Performance Assessment Working Group

ALARA ANALYSES This chapter is applicable to Decommissioning Groups 2-7. Licensees in Decommissioning Groups 2 and 3 may only have to refer to the discussion of good housekeeping practices in Section 6.3.

Safety Evaluation Review Procedures 6.1.1 Areas of Review The NRC staff should review the information supplied by the licensee or responsible party to determine if the licensee has developed a DP that ensures that doses to the average member of the critical group are ALARA. Information submitted should include (1) a cost-benefit analysis (or qualitative arguments) demonstrating that the applicable ALARA requirement(s) for the licensees preferred decommissioning option will be met and (2) a description of the licensees preferred method for showing compliance with the ALARA requirement at the time of decommissioning. If the licensee proposes to use engineered barriers or intentional mixing to meet the LTR criteria for unrestricted use, it should complete an appropriate ALARA analysis, as described in Section 3.5 of this volume (for engineered barriers) or Section 15.13 of Volume 1 (for intentional mixing). Additionally, an ALARA evaluation for restricted use should follow guidance described in Appendix N of this NUREG-1757, Volume 2. The licensee should also follow the guidance in Appendix N of this volume when evaluating the eligibility tests for restricted use (10 CFR 20.1403(a), ALARA test, or test of net public or environmental harm) and use of a higher dose limit for restricted use with institutional controls not in place (10 CFR 20.1403(e)(2)(i), test for prohibitively expensive, or test for net public or environmental harm).

6.1.2 Review Procedures 6.1.2.1 Acceptance Review The NRC staff should review the DP to ensure that, at a minimum, it contains the information summarized under the above Areas of Review. The NRC staff should review the ALARA portion of the DP without assessing the technical accuracy or completeness of the information contained therein, which it should determine during the detailed technical review. The NRC staff should review the DP table of contents and the individual descriptions under Areas of Review (1) to ensure that the licensee or responsible party has included this information in the DP and (2) determine if the level of detail of the information appears to be adequate for a detailed technical review.

6.1.2.2 Safety Evaluation The material supporting the ALARA portion of the DP to be reviewed is technical in nature and specific detailed technical analysis may be necessary. The NRC staff should evaluate the licensees dose estimates for various alternatives using the appropriate guidance in Chapter 5 of this volume and should evaluate the licensees cost estimates using the guidance in Section 4.1 from NUREG-1757, Volume 3.

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Acceptance Criteria 6.2.1 Regulatory Requirements 10 CFR 20.1101(b), 20.1402, 20.1403(a), 20.1403(b), 20.1403(e), and 20.1404(a)(3) 6.2.2 Regulatory Guidance Appendix N of Volume 2 of this NUREG report 6.2.3 Information to be Submitted The information supplied by the licensee should be sufficient to allow the NRC staff to fully understand the basis for the licensees conclusion that projected dose or residual radioactivity concentrations and quantity (hereafter, the decommissioning goal) are ALARA. The decommissioning goal should be established at the point where the incremental benefits equal the incremental costs. The NRC staff review should verify that the following information is included in the description of the development of the decommissioning goal:

  • a description of how the licensee will achieve a decommissioning goal that meets the dose limit and ALARA requirement
  • a quantitative cost-benefit analysis
  • a description of how costs were estimated
  • a demonstration that the doses to the average member of the critical group are ALARA The information to be submitted is also included as part of the master DP Checklist provided in this NUREG report (see Section VII from Appendix D of Volume 1).

Evaluation Criteria 6.3.1 Evaluation of Good Practice Efforts For ALARA during decommissioning, all licensees should use typical good practice, such as floor and wall washing, removal of readily removable radioactivity in buildings or in soil areas, and other good housekeeping practices. In addition, the FSSR should describe how the licensee employed these practices to achieve the final activity levels.

6.3.2 Evaluation of Cost-Benefit ALARA Analyses The NRC staff review should verify that the qualitative descriptions provide reasonable assurance that the activities and decommissioning goal should result in doses to the average member of the critical group that are ALARA. For those situations in which a licensee prepares cost-benefit analyses, the NRC staff should ensure that the analyses are developed using the methodology described in Appendix N and are applied as described in the following text.

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6.3.3 When Cost-Benefit Analyses are Unnecessary In the following circumstances, the results of an ALARA analysis are known on a generic basis and the NRC staff considers that an analysis is not necessary (see Appendix N of this volume for more details):

  • unrestricted use where excavated soil would be shipped to a LLW disposal facility for disposal
  • soil removal at a site to meet the unrestricted use dose criteria of 0.25 mSv/y (25 mrem/y)
  • remediation of building surfaces or surface soil to the NRC default screening levels (see Appendix H for information about the screening levels)
  • that no residual radioactivity distinguishable from background will remain at the site at termination
  • that loose residual radioactivity on building surfaces has been or will be removed 6.3.4 Calculation of Benefits Appendix N of this volume discusses five different possible benefits: (1) collective dose averted, (2) regulatory costs avoided, (3) changes in land values, (4) esthetics, and (5) reduction in public opposition. Numerical estimates will generally only be available for the first three benefits, if they are appropriate. The licensee can make a qualitative analysis of the benefits, especially if the costs are large (e.g., no matter what the change in land value is, the costs will exceed the benefits). In most comparisons between alternatives in the same class (e.g., both alternatives result in unrestricted release), the only important benefit should be the collective dose averted. In comparisons between restricted and unrestricted release, the other benefits can become important.

The collective dose averted is generalized as the incremental dose difference between the licensees approach (hereafter, preferred option) and the alternative under analysis. Therefore, the NRC staff should ensure that the licensee has calculated the benefits correctly by using the correct population density, area, and averted dose. This may require a technical analysis of the dose modeling, and the reviewer should use Chapter 5 for these cases. If the licensee has used discounting, the NRC staff should ensure that the proper rates were used. The licensee is not required to discount because the discount reduces the benefits of averting dose in later time periods.

One acceptable method of compliance with 10 CFR 20.1403(a) is to demonstrate that cleanup to the unrestricted release criteria is beyond ALARA considerations. In this case, a beneficial estimate should include costs that would be avoided if the site were to be released for unrestricted use, such as site control and maintenance costs, as well as an estimate of the additional regulatory costs associated with termination of a restricted site (e.g., development of an environmental impact statement, public meetings). Appendix N of this volume contains more details on compliance with ALARA requirements for restricted use.

The NRC staff should ensure that the licensee has properly documented the basis for any estimates of changes in land values. Acceptable sources of such estimates include 6-3

governmental assessors (e.g., county, State) or real estate agents familiar with the local area and the issues involved.

6.3.5 Calculation of Costs The NRC staff should verify that the licensee has adequately estimated the effective monetary costs of the incremental remediation by using the equations in Appendix N of this volume. To review the calculated monetary costs of the incremental remediation, the NRC staff should use Section 4.1 of NUREG-1757, Volume 3, with the following changes (this may require calculating total cost estimates for the preferred option and each alternative):

  • The cost estimate should be based on actual costs expected to be incurred by decommissioning the facility and should not assume that the work will be performed by an independent third-party contractor.
  • The cost estimate does take credit for (1) any salvage value that might be realized from the sale of potential assets during or after decommissioning or (2) any tax reduction that might result from paying decommissioning costs or site control and maintenance costs.
  • The decommissioning cost estimates should reflect the actual situation rather than maximized assumptions.

For each of the cost terms (e.g., disposal costs, worker fatalities), the incremental difference between the preferred and the alternative options may be negative (i.e., the alternative may cost less than the preferred option).

6.3.6 Compliance Methods at the Time of Decommissioning There are two approaches to demonstrate compliance with the ALARA requirement at the end of decommissioning: (1) a predetermined acceptable dose limit or concentration guideline(s) or (2) an acceptable preferred option and decommissioning goal with organizational oversight and review during decommissioning. Both options have advantages and disadvantages. The licensee establishes the compliance method, with the staff reviewing the applicability, given the site-specific information.

6.3.6.1 Predetermined Compliance Measure Under the predetermined compliance measure, the licensee would agree to meet the dose calculated for the preferred option or the radiological concentrations associated with that dose.

This could be met by either establishing deterministic concentration limits for the site or agreeing to use a specified dose scenario with associated parameters and assumptions. If the licensees final survey results meet the self-imposed concentration limits (or dose limit), the licensee has met the ALARA requirement.

6.3.6.2 Performance-Based Compliance Performance-based compliance allows a licensee to adjust its ALARA assessment during decommissioning to deal with actual site conditions experienced and actual costs incurred. The philosophy behind this compliance measure is very similar to how ALARA is handled during routine operations. The licensees DP needs to meet all of the following criteria to use this approach:

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  • The preferred option, based on valid assumptions, would result in reducing residual activity to ALARA levels, as described above.
  • The licensee has established decommissioning guidelines (either dose or concentrations) based on the DPs analysis.
  • The licensee has a documented method to review the effectiveness of the remediation activities. This method should include all of the following:

o an ALARA committee or radiation safety officer for small licensees, similar to operations requirements o an appropriate review frequency established o an acceptable set of criteria on the scope of activities or commitments that the ALARA committee can change o a commitment to prepare acceptable documentation of ALARA findings that result in the licensee changing its remediation activities or decommissioning guidelines o a commitment to provide the NRC annually with all necessary page changes to the DP because of ALARA findings At the end of remediation, a licensee using the performance-based approach should meet the following criteria:

  • The final survey results satisfy the appropriate dose limit(s).
  • Any substantial weaknesses in the ALARA program that were found during licensee audits or NRC inspections have been resolved.
  • Any deviation from the decommissioning goal presented in the DP is properly justified by the ALARA committee findings.

The NRC license reviewer or inspection staff should review long-term projects annually.

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BIBLIOGRAPHY AND SUPERSEDED DOCUMENTS This section provides the reference list for the main body of this volume, categorized in the following subsections by type of reference document. Chapter 4 of Volume 1 of this NUREG series provided a more general list of decommissioning references, which included statutes, decommissioning regulations, decommissioning inspection manual chapters, and decommissioning inspection procedures.

Use of References Cited in this Volume This volume refers to a number of other documents for guidance. In some cases, this volume will state that the referenced guidance is approved by the NRC staff. However, in some cases, the documents are only referenced for information, and, if so, the licensee should contact the NRC staff to determine the specific applicability to a facility, as appropriate.

NRC Decommissioning Documents Referenced in the Main Body of Volume 2 United States Nuclear Regulatory Commission (US NRC). Manual for Conducting Radiological Surveys in Support of License Termination, Draft Report for Comment, NUREG/CR-5849. NRC: Washington, DC. 1992a.

. Residual Radioactive Contamination from Decommissioning: Technical Basis for Translating Contamination Levels to Annual Total Effective Dose Equivalent, NUREG/CR-5512, Vol. 1. NRC: Washington, DC. 1992b.

. Background as a Residual Radioactivity Criterion for DecommissioningDraft Report, NUREG-1501. NRC: Washington, DC. August 1994a.

. Draft Branch Technical Position on Site Characterization for Decommissioning.

NRC: Washington, DC. 1994b.

. Standard Review Plan for the Review of a License Application for a Low-Level Waste Disposal Facility, NUREG-1200, Rev. 3. NRC: Washington, DC. 1994c.

. Working Draft Regulatory Guide on Release Criteria for Decommissioning:

NRC Staffs Draft for Comment, NUREG-1500. NRC: Washington, DC. 1994d.

. Guidelines for Preparing and Reviewing Applications for Licensing of Non-Power Reactors, NUREG-1537. NRC: Washington, DC. 1996.

. NMSS Handbook for Decommissioning Fuel Cycle and Materials Licensees, NUREG/BR-0241. NRC: Washington, DC. 1997.

. Decision Methods for Dose Assessment To Comply with Radiological Criteria for License Termination, Draft Report for Comment, NUREG-1549. NRC: Washington, DC.

1998a.

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. Minimum Detectable Concentrations with Typical Radiation Survey Instruments for Various Contaminants and Field Conditions, NUREG-1507, Rev. 1. NRC: Washington, DC. 2020.

. A Proposed Nonparametric Statistical Methodology for the Design and Analysis of Final Status Decommissioning Surveys: Interim Draft Report for Comment and Use, NUREG-1505, Rev. 1. NRC: Washington, DC. 1998c.

. Residual Radioactive Contamination from Decommissioning, Parameter Analysis, Draft Report for Comment, NUREG/CR-5512, Vol. 3. NRC: Washington, DC. 1999.

. Multi-Agency Radiation Survey and Site Investigation Manual (MARSSIM),

NUREG-1575, Rev. 1, EPA 402-R-97-016, Rev. 1, DOE/EH-0624, Rev. 1. NRC: Washington, DC, U.S. Department of Defense, U.S. Department of Energy, U.S. Environmental Protection Agency. 2000a.1

. NMSS Decommissioning Standard Review Plan, NUREG-1727. NRC:

Washington, DC. 2000a.

. Draft Staff Guidance for Dose Modeling of Proposed Partial Site Releases.

Memorandum from John T. Greeves to John A. Zwolinski. NRC: Washington, DC. 2001a.

. Residual Radioactive Contamination from Decommissioning, Users Manual, DandD Version 2.1, NUREG/CR-5512, Vol. 2. NRC: Washington, DC. 2001b.

. Decommissioning Criteria for the West Valley Demonstration Project (M-32) at the West Valley Site: Final Policy Statement. Federal Register, Vol. 67, No. 22, pp. 5003-5012. 2002a.

. Re-Evaluation of the Indoor Resuspension Factor for the Screening Analysis of the Building Occupancy Scenario for NRCs License Termiantion RuleDraft Report for Comment, NUREG-1720. NRC: Washington, DC. 2002b.

. Results of the License Termination Rule Analysis, SECY-03-0069. NRC:

Washington, DC. 2003a.

. Staff RequirementsSECY-03-0069Results of the License Termination Rule Analysis, SRM-SECY-03-0069. NRC: Washington, DC. 2003b.

. Results of the License Termination Rule Analysis of the Use of Intentional Mixing of Contaminated Soil, SECY-04-0035. NRC: Washington, DC. 2004a.

. Staff RequirementsSECY-04-0035Results of the License Termination Rule Analysis of the Use of Intentional Mixing of Contaminated Soil, SRM-SECY-04-0035. NRC:

Washington, DC. 2004b.

. Results of the License Termination Rule Analysis, Regulatory Issue Summary 2004-08. NRC: Washington, DC. 2004c.

1 Corrected pages for MARSSIM, Revision 1 (August 2000) with the June 2001 updates, are available at the NRC website: http://www.nrc.gov/reading-rm/doc-collections/nuregs/staff/sr1575/r1/.

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. Consolidated Decommissioning Guidance - Decommissioning Process for Materials Licensees, NUREG-1757, Vol. 1, Rev.2. NRC: Washington, DC. 2006a.

. Stakeholder Comments and Path Forward on Decommissioning Guidance To Address License Termination Rule Analysis Issues, SECY-06-0143. NRC: Washington, DC.

2006b.

. Staff RequirementsSECY-06-0143Stakeholder Comments and Path Forward on Decommissioning Guidance to Address License Termination Rule Analysis Issues, SRM-SECY-06-0143. NRC: Washington, DC. 2006c.

. Decommissioning Oversight and Inspection Program for Fuel Cycle Facilities and Materials Licensees, Inspection Manual Chapter 2602. NRC: Washington, DC. 2008.

. Guidance for Conducting Technical Anlayses for 10 CFR Part 612, Draft Report for Comment, NUREG-2175. NRC: Washington, DC. 2015.

. Standard Review Plan for Evaluating Nuclear Power Reactor License Termination Plans, NUREG-1700, Rev. 2. NRC: Washington, DC. 2018.

Other NRC Documents Referenced in this Volume Code of Federal Regulations (CFR). 10 CFR Part 20, Standards for Protection against Radiation.

. 10 CFR Part 30, Rules of General Applicability to Domestic Licensing of Byproduct Material.

. 10 CFR Part 40, Domestic Licensing of Source Material.

. 10 CFR Part 50, Domestic Licensing of Production and Utilization Facilities.

. 10 CFR Part 70, Domestic Licensing of Special Nuclear Material.

. 10 CFR Part 72, Licensing Requirements for the Independent Storage of Spent Nuclear Fuel and High-Level Radioactive Waste.

NRC. Staff Requirements Memorandum, SECY-98-144, White Paper on Risk-Informed and Performance-Based Regulation. NRC: Washington, DC. March 1999.

. A Performance Assessment Method for Low-Level Waste Disposal Facilities:

Recommendations of NRCs Performance Assessment Working Group, NUREG-1573. NRC:

Washington, DC. 2000.

. Environmental Review Guidance for Licensing Actions Associated with NMSS Programs, NUREG-1748. NRC: Washington, DC. 2001.

. Standard Review Plan for the Review of a Reclamation Plan for Mill Tailings Sites Under Title II of the Uranium Mill Tailings Radiation Control Act, NUREG-1620, Rev. 1.

NRC: Washington, DC. 2003.

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. Jurisdiction Determinations, SA-500, FSME Procedure Approval. ADAMS Accession No. ML110600350. NRC: Washington, DC. 2011.

. Clarification on the Determination of Regulatory Jurisdiction of Non-Federal Entities Conducting Cleanup Activities on Federal Property in Agreement States, FSME-14-039. NRC: Washington, DC. 2014.

Other References in this Volume American National Standards Institute. Characterization in Support of Decommissioning Using the Data Quality Objectives Process, ANSI N13.59:2008. ANSI: New York, NY. 2008.

Atomic Energy Act of 1954, as amended. 42 U.S.C. §§2011 et seq.

Chapman, J.A., A.J. Boerner, E.W. Abelquist. Spatially-Dependent Measurements of Surface and Near-Surface Radioactive Material Using In situ Gamma Ray Spectrometry (ISGRS) For Final Status Surveys. 2006.

Code of Federal Regulations. 40 CFR Part 190, Environmental Radiation Protection Standards for Nuclear Power Operations.

Comprehensive Environmental Response, Compensation, and Liability Act of 1980 (CERCLA).

42 U.S.C. §§9601.

Department of Energy (US DOE). Decommissioning Handbook, DOE/EM-0142P. DOE:

Washington, DC. 1994.

Environmental Protection Agency (US EPA). Guidance for Conducting Remedial Investigations and Feasibility Studies Under CERCLA, EPA/540/G-89/004. EPA: Washington, DC. 1988a.

. Limiting Values of Radionuclide Intake and Air Concentration and Dose Conversion Factors for Inhalation, Submersion, and Ingestion: Federal Guidance Report No.

11, EPA 520/1-88-020. EPA: Washington, DC. 1988b.

. Superfund Removal Procedures, OSWER Directive 9360.0-03B. EPA:

Washington, DC. 1988c.

. External Exposure to Radionuclides in Air, Water, and Soil: Federal Guidance Report No. 12, EPA 402-R-93-081. EPA: Washington, DC. 1993.

. Federal Radiation Protection Draft Guidance for Exposure of the General Public. Federal Register, Vol. 59, p. 66414. US EPA: Washington DC. 1994.

. Guidance for the Data Quality Objectives Process, EPA QA/G-4, EPA/600/R-96/055. EPA: Washington, DC. 2000.

Resource Conservation and Recovery Act, 42 U.S.C. §§ 6901-6991 (1976).

Uranium Mill Tailings Radiation Control Act, 42 U.S.C. §§ 7901-7925 (1978).

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Documents Superseded by this Volume This volume supersedes the guidance documents listed in Table 7.4, and the superseded documents should no longer be used.

Table 7.1 Documents Superseded by this Report Document Title Date NRC Memorandum Draft Staff Guidance for Dose Modeling of Proposed 09/2001 Partial Site Releases NUREG-1727 NMSS Decommissioning Standard Review Plan 09/2000 NUREG/BR-0241 NMSS Handbook for Decommissioning Fuel Cycle and 03/1997 Materials Licensees Branch Technical Draft Branch Technical Position: Screening 10/1996 Position Methodology for Assessing Prior Land Burials of Radioactive Waste Authorized Under Former 10 CFR 20.304 and 20.302 Branch Technical Draft Branch Technical Position on Site 11/1994 Position Characterization for Decommissioning NUREG-1500 Working Draft Regulatory Guide on Release Criteria for 08/1994 Decommissioning: NRC Staffs Draft for Comment NUREG/CR-5849 Manual for Conducting Radiological Surveys in Support 06/1992 of License Termination Branch Technical Disposal of Onsite Storage of Thorium or Uranium 10/1981 Position Wastes from Past Operations 7-5

This Volume 2 of this NUREG report also incorporates and updates numerous portions of NUREG-1727, NMSS Decommissioning Standard Review Plan (SRP), issued September 2000, specifically Chapters 5, 7, and 14; and Appendices C, D, and E of NUREG-1727 on dose modeling, ALARA, and facility radiation surveys. The NUREG-1727 chapters and appendices that have been incorporated into this NUREG are superseded. This three-volume NUREG series (i.e., NUREG-1757, Volumes 1, 2, and 3) supersedes both NUREG/BR-0241, NMSS Handbook for Decommissioning Fuel Cycle and Materials Licensees, issued March 1997, and NUREG-1727 in their entirety.

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APPENDIX A IMPLEMENTING THE MARSSIM APPROACH FOR CONDUCTING FINAL RADIOLOGICAL SURVEYS

A.1 Introduction This appendix is applicable to Decommissioning Groups 2-7.

Regulations of the NRC in Title 10 of the Code of Federal Regulations (10 CFR) 20.1501(a) require licensees to make or cause to be made surveys that may be necessary for the licensee to comply with the regulations in 10 CFR Part 20, Standards for Protection against Radiation.

The FSS is the radiation survey performed after an area has been fully characterized, remediation has been completed, and the licensee believes that the area is ready to be released. The purpose of the FSS is to demonstrate that the area meets the radiological criteria for license termination. The FSS is not conducted for the purpose of locating residual radioactivity; the HSA and the characterization survey perform that function.

The NRC endorses the FSS method described in NUREG-1575, Revision 1, Multi-Agency Radiological Survey and Site Investigation Manual (MARSSIM), issued August 2000, and references to MARSSIM sections within this appendix are specifically referring to MARSSIM Revision 1 (NRC, 2000).1 This appendix (1) provides an overview of the MARSSIM approach for conducting a final radiological survey, (2) includes additional specific guidance on acceptable values for use in the MARSSIM method, (3) explains how to use the MARSSIM method in a way that is consistent with the dose modeling, (4) describes how to use the MARSSIM method to meet the NRCs regulations, and (5) demonstrates how to extend or supplement the MARSSIM method to certain complex situations that may be encountered, such as how to address subsurface residual radioactivity. Note that the guidance in this appendix does not replace MARSSIM, and licensees and reviewers should refer to, and use, MARSSIM for designing final radiological surveys to support decommissioning. This guidance assumes a working knowledge of the MARSSIM approach and terminology and does not attempt to provide a comprehensive overview of the entire MARSSIM. In addition, for Decommissioning Groups 1-3, licensees may also use the alternative, simpler final survey methods described in Appendix B of this volume.

Chapter 5 of MARSSIM contains survey checklists. These checklists are useful in implementing the steps of the RSSI process (Decommissioning Groups 3-7). These checklists are a useful tool for visualizing the sequential steps (i.e., design, performance, and evaluation) of the survey process. Furthermore, the use of these checklists should ensure that the necessary information is collected for each type of survey. Sites not using the RSSI process, such as Decommissioning Groups 1 and 2, should also find these checklists or parts of these checklists useful.

A.2 Classification of Areas by Residual Radioactivity Levels The licensee should classify site areas based on levels of residual radioactivity from licensed activities. The area classification method contained in Section 4.4 of MARSSIM is acceptable to the NRC staff. Its essential features are described below.

The licensee should first classify site areas as impacted or nonimpacted. Impacted areas are areas that may have residual radioactivity from the licensed activities. Nonimpacted areas are 1 As of the publication of this guidance document MARSSIM, Revision 1, is the current version. MARSSIM, Revision 2, was issued for public comment in 2021 but has not yet been finalized. Unless specifically noted, this guidance document is referring to MARSSIM, Rev. 1. The NRCs decommissioning website can be consulted for updates on issuance of decommissioning guidance and technical reports, as well as a listing of any lessons learned between guidance revisions.

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areas without residual radioactivity from licensed activities. Impacted areas should be identified by using knowledge of past site operations together with site characterization surveys. In the FSS, radiation surveys do not need to be conducted in nonimpacted areas. The licensee should classify impacted areas into one of the three classes, listed below, based on levels of residual radioactivity.

(1) Class 1 Areas: Class 1 areas are impacted areas that are expected to have concentrations of residual radioactivity that exceed the derived concentration guideline level or DCGLW (average concentrations over a wide area).

(2) Class 2 Areas: Class 2 areas are impacted areas that are not likely to have concentrations of residual radioactivity that exceed the DCGLW.

(3) Class 3 Areas: Class 3 areas are impacted areas that have a low probability of containing residual radioactivity.

Surveys conducted during operations or during characterization at the start of decommissioning are the basis for classifying areas. If the available information is not sufficient to designate an area as a particular class, the area either should be classified as Class 1 or should be further characterized. Areas that are considered to be on the borderline between classes should receive the more restrictive classification.

The NRC staff recognizes that a licensee may need to reclassify Class 1 areas to Class 2, when insufficient information is available for the initial classification. If more information becomes available to indicate that another classification is more appropriate, the guidance in MARSSIM allows for classifications to be changed at any time before the FSS. For more guidance on criteria for downgrading classifications (e.g., from Class 1 to Class 2), a licensee should refer to MARSSIM, Revision 1; in particular, Sections 2.2, 2.5.2, and 5.5.3. If a licensee plans to make use of reclassification during the RSSI process, it should include in the DP the criteria and methodology it plans to use for reclassification. In addition, a licensee contemplating the use of reclassification is encouraged to contact NRC staff.

For Class 1 and 2 soils, the licensee should determine whether a significant amount of subsurface residual radioactivity is present, based on the HSA and site characterization. In this context significant amounts of subsurface residual radioactivity would be defined as an amount of radioactivity, or contaminated material (such as soil), that could contribute at least 10 percent of the dose criteria (see Section 3.3).

The presence of subsurface soils is important because subsurface soils cannot typically be measured by scan instrumentation and, therefore, scan surveys are not expected to be adequate for the purposes of an FSS. Although surface soils have been associated with the top 15 centimeters (cm) of soil, which can typically be measured by scan instrumentation, the exact depth of residual radioactivity for which scan survey instrumentation can adequately detect residual radioactivity varies, based on several factors (e.g., survey instrument, radionuclide, and soil characteristics).

Determining whether there is a significant amount of subsurface residual radioactivity should not require a complex set of characterization measurements. In most cases, there will be either significant amounts of residual radioactivity or only traces (such as in occasional small pockets or from leaching from surface layers by rainwater). When there is an insignificant amount of subsurface residual radioactivity, the MARSSIM survey methods for surface measurements are A-2

acceptable. When there is a significant amount of subsurface residual radioactivity, the licensee should modify the dose modeling and survey methods to account for it.

The HSA usually determines the presence of subsurface residual radioactivity (see Chapter 3 of MARSSIM), applying knowledge of how the residual radioactivity was deposited.

Characterization surveys to detect subsurface residual radioactivity in soil are not routinely conducted, unless there is reason to expect that subsurface residual radioactivity may be present. The need to survey or sample subsurface soil will depend, in large part, on the quality of the information used to develop the HSA, the environmental conditions at the site, the types and forms (chemical and radiological) of the radioactive material used at the site, the authorized activities and the manner in which licensed material was managed during operations.

The NRC staffs experience has shown that submittal of the DP should occur only after sufficient site characterization has occurred. The staff suggests that the DP provide sufficient information demonstrating the characterization of the radiological conditions of site structures, facilities, surface and subsurface soils, and groundwater. The NRC staff has observed that some DPs have been submitted with incomplete or inadequate characterizations of radiological conditions. A review of such DPs has shown that the lack of information makes it difficult to evaluate the rationale for the proposed classification of survey units. The NRC staff suggests that the following issues related to the use of characterization survey results and classification of survey units be considered when developing a DP:

  • use of operational, post-shutdown scoping, or turnover surveys as characterization surveys
  • reclassification of survey units
  • completeness of characterization survey design and results Regulatory Issue Summary 2002-02, Lessons Learned Related to Recently Submitted Decommissioning Plans and License Termination Plans, issued January 2002, provides a detailed discussion of this issue.

A.3 Selection and Size of Survey Units The licensee should divide the impacted area into survey units based on the classification described above. A survey unit is a portion of a building or site that is surveyed, evaluated, and released as a single unit. The licensee should give the entire survey unit the same area classification. Section 4.6 of MARSSIM contains a method acceptable to the NRC staff for dividing impacted areas into survey units. The important features of this method are summarized here.

For buildings, it is normally appropriate to designate each separate room as either one or two survey units (e.g., floors with the lower half of walls and upper half of walls with ceiling), based on the pattern of potential of residual radioactivity. It is generally not appropriate to divide rooms of normal size (100 square meters (m2) or less) into more than two survey units, because the dose modeling is based on the room being considered as a single unit. However, very large spaces such as warehouses may be divided into multiple survey units.

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For soil, survey units should be areas with similar operational history or similar potential for residual radioactivity to the extent practical. Survey units should be formed from areas with the same classification to the extent practical, but if areas with more than one class are combined into one survey unit, the entire survey unit should be given the more restrictive classification.

Survey units should have relatively compact shapes and should not have highly irregular (gerrymandered) shapes unless the unusual shape is appropriate for the site operational history or the site topography.

Table A.1 contains suggested survey unit areas from MARSSIM. These areas are suggested in MARSSIM, because they give a reasonable sampling density and are consistent with most commonly used dose modeling codes. However, the size and shape of a particular survey unit may be adjusted to conform to the existing features of the particular site area.

Table A.1 Suggested Survey Unit Areas (MARSSIM, Revision 1, Roadmap, Table 1

[NRC, 2000])

Suggested Survey Unit Area Class Structures Land 1 up to 100 m2 up to 2,000 m2 2 100 to 1,000 m 2 2,000 to 10,000 m2 3 no limit no limit A.4 Selection of Background Reference Areas and Materials A.4.1 Need for Background Reference Areas Background reference areas are not needed when radionuclide-specific measurements will be used for concentrations of a radionuclide that is not present in background. Background reference areas are needed for the MARSSIM method if (1) the residual radioactivity contains a radionuclide that occurs in background, or (2) the sample measurements to be made are not radionuclide specific. However, a licensee may find it cost beneficial to consider the background for a particular radionuclide as zero or some other appropriately low value approved by the staff, recognizing that this is a risk-informed approach. The survey unit itself may serve as the reference area when a surrogate radionuclide in the survey unit can be used to determine background. For example, it may be possible to use radium (Ra)-226 as a surrogate for background uranium when the contaminant is processed uranium. Section 4.3.2 of MARSSIM contains more information on the use of surrogate radionuclides.

The licensee may use multiple reference areas if the reference areas have significantly different background levels because of the variability in background between areas (see Section A.4.4 below and Section 13.2 of NUREG-1505, A Proposed Nonparametric Statistical Methodology for the Design and Analysis of Final Status Decommissioning Surveys-Interim Draft Report for Comment and Use, issued June 1998). The licensee may use a derived reference area to extract background information from the survey unit, because a suitable reference area is not readily available. For example, it may be possible to derive a background distribution based on areas of the survey unit where residual radioactivity is not present.

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A.4.2 Characteristics of Soil Reference Areas The objective is to select nonimpacted background reference areas where the distribution of measurements should be the same as would be expected in the survey unit if that survey unit had never been contaminated. Section 4.5 of MARSSIM contains an acceptable method for selecting background areas, as briefly described below.

Reference areas should have a soil type as similar to the soil type in the survey unit as possible.

If there is a choice of possible reference areas with similar soil types, consideration should be given to selecting reference areas that are most similar in terms of other physical, chemical, geological, and biological characteristics. Each reference area should have an area at least as large as the survey unit, if practical, to include the full potential spatial variability in background concentrations. Reference areas may be off site or on site, as long as they are nonimpacted.

NUREG-1506, Measurement Methods for Radiological Surveys in Support of New Decommissioning Criteria, Draft Report for Comment, issued August 1995, provides additional information on reference area selection. Licensees should contact the NRC staff when they are unable to find a reference area that satisfies the above criteria.

A.4.3 Different Materials in a Survey Unit Survey units may contain a variety of materials with markedly different backgrounds. An example might be a room with drywall walls, concrete floor, glass windows, metal doors, wood trim, and plastic fixtures. It is not appropriate to make each material a separate survey unit, because the dose modeling is based on the dose from the room as a whole and because a large number of survey units in a room would require an inappropriate number of samples.

When there are different materials with substantially different backgrounds in a survey unit, the licensee may use a reference area that is a nonimpacted room with roughly the same mix of materials as the survey unit.

If a survey unit contains several different materials, but one material is predominant, or if there is not too great a variation in background among materials, a background from a reference area containing only a single material may still be appropriate. For example, a room may be mostly concrete but with some metal beams, and the residual radioactivity may be mostly on the concrete. In this situation, where the concrete predominates, it would be acceptable to use a reference area that contained only concrete. However, the licensee should demonstrate that the selected reference area will not result in underestimating the residual radioactivity on other materials.

The licensee may also use measured backgrounds for different materials or for groups of similar materials. When the licensee decides to use different measured backgrounds for different materials or for a group of materials with similar backgrounds, it is acceptable to perform a one-sample test on the difference between the paired measurements from the survey unit and from the appropriate reference material. Chapter 12 of NUREG-1505 describes, in detail, an acceptable method to do this.

For onsite materials, present either in buildings or as nonsoil materials present in outdoor survey units (e.g., concrete, brick, drywall, fly ash, petroleum product wastes), the licensee should attempt to find nonimpacted materials that are as similar as possible to the materials on the site. Sometimes such materials will not be available. In those situations, the licensee A-5

should make a good faith effort to find the most similar materials readily available or use appropriate published estimates.

A.4.4 Differences in Backgrounds between Areas When using a single reference area, any difference in the mean radionuclide concentration between the survey unit and the reference area would be interpreted as caused by residual radioactivity from site operations. This interpretation may not be appropriate when the variability in mean background concentrations among different reference areas is a substantial fraction of the DCGLW. When there may be a significant difference in backgrounds between different areas, the licensee can conduct a Kruskal-Wallis test, as described in Chapter 13 of NUREG-1505, to determine whether there are, in fact, significant differences in mean background concentrations among potential reference areas.

While NUREG-1505 does not recommend specific values for the Kruskal-Wallis test, information on the power of the F-test (parametric complement to the nonparametric Kruskal-Wallis test) is provided in Table 13.5 of NUREG-1505 to assist with the selection of the number of reference areas and measurements per reference area, as part of the DQO process. For example, Table 13.5 of NUREG-1505 indicates that 4 reference areas and 10 to 20 measurements per reference area should generally be adequate. Also, NUREG-1505 states that if four reference areas are selected, an alpha error of 0.1 would be a reasonable default, and in some circumstances larger alpha error values could be considered. NUREG-1505 notes that different alpha and beta error values may be justified based on risk considerations. For example, a lower beta error value than alpha error value would reduce the risk of not detecting background variations that are actually present.

If there are significant differences in backgrounds among reference areas, a value of three times the standard deviation of the mean of the reference area background values could be added to the mean of the reference area background to define a concentration that is distinguishable from background. A value of three times the standard deviation of the mean is chosen to minimize the likelihood that a survey unit that contains only background would fail the statistical test for release. This value can be used as the lower bound of the grey region (LBGR) in a two-sample test (WRS test) (see Appendix A and G, Sections A.5 and Section G.6) to test whether the survey unit meets the radiological criteria for license termination. Chapters 6 and 13 of NUREG-1505 describes this method in detail.

A.4.5 Background Survey Design This survey measures nonimpacted areas on and surrounding the site to establish the baseline; that is, the normal background levels of radiation and radioactivity. In some situations, historical measurements may be available from surveys performed before the construction and operation of a facility. The survey should avoid areas such as roads, parking lots, and other large paved surfaces that may have been impacted or disturbed by site-related activities. The background survey takes on added importance, since the licensee may decide to use a statistical test that compares impacted areas to offsite or onsite reference areas to demonstrate compliance with the release criteria in 10 CFR Part 20, Subpart E, Radiological Criteria for License Termination. To minimize systematic biases in the comparison, the same sampling procedure, measurement techniques, and type of instrumentation (e.g., detection sensitivity and accuracy) should be used at both the survey unit and the reference area.

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NUREG-1505 provides additional guidance on survey design, the methods of accounting for background radiation, and the nonparametric statistical methods for testing compliance with the decommissioning criteria in 10 CFR Part 20, Subpart E. Formulas contained in NUREG-1505 can be used to compute the required number of samples (measurement points) that will be needed in both the background reference and survey areas.

A.5 Methods to Evaluate Survey Results All survey units should be evaluated to determine whether the average concentration in the survey unit as a whole is below the DCGLW. If the radionuclide is not present in background, and the measurement technique is radionuclide-specific so that a comparison with a reference area is not necessary, the licensee should use a one-sample test, the Sign test. Section 8.3 of MARSSIM describes this test.

When the residual radioactivity contains a radionuclide present in the environment or when the measurements are not radionuclide specific, the survey unit should be compared to a reference area. When comparing the survey unit to a reference area, the licensee should use a two-sample test, the WRS test. Section 8.4 of MARSSIM describes this test.

A.5.1 A Case for not Subtracting Background An exception to using a two-sample test when a radionuclide is present in background is when the licensee plans to assume that all the radionuclide activity in the survey unit is caused by licensed operations and none is from background. This could be the case for cesium (Cs)-137, for example, because the levels in the environment are often so much less than the DCGLW that background concentrations may be ignored.

A.5.2 Elevated Measurements Comparison Class 1 survey units that pass the Sign test or WRS test but have small areas with concentrations exceeding the DCGLW should also be tested to demonstrate that those small areas meet the dose criteria for license termination. This test is called the elevated measurement comparison (EMC). It is described in Section 8.5.1 of MARSSIM and summarized here.

To perform the EMC, the licensee first determines the size of the area in the survey unit with a concentration greater than the DCGLW, then determines the DCGLEMC for an area of that size.

(The DCGLEMC is the concentration permitted in a limited area of a survey unit; see Section A.8.6.) The average concentration in the area is calculated for comparison against the DCGLEMC. The EMC is acceptable if the following condition is met, as shown in Equation A-1 (modified from MARSSIM Equation 8-2):

+ <1 (A-1)

Where = the average residual radioactivity concentration for all sample points in the survey unit only.

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If there is more than one elevated area, a separate term should be included for each one. As an alternative to the unity rule expressed in Equation A-1, the licensee can calculate the dose from the actual distribution of residual radioactivity, if an appropriate exposure pathway model is available.

A.6 Instrument Selection and Calibration To demonstrate that the radiological criteria for license termination have been met, the measurement instruments should have an adequate sensitivity, be calibrated properly, and be checked periodically for proper response.

A.6.1 Calculation of Minimum Detectable Concentrations The licensee should determine the MDC for the instruments and techniques to be used. The MDC is the concentration that a specific instrument and technique can be expected to detect 95 percent of the time under actual conditions of use.

For scanning building surfaces for beta and gamma emitters, the MDCscan should be determined from the following equation (obtained by combining MARSSIM Equations 6-8, 6-9, and 6-10 and using a value recommended in this appendix for the index of sensitivity d of 1.38, which is for 95 percent detection of a concentration equal to MDCscan with a 60 percent false-positive rate).

270,000 138 . B MDCscan (building surfaces) = (A-2) p i s A t where MDCscan = minimum detectable concentration for scanning building surfaces in picocuries per square meter (pCi/m2) 270,000 = conversion factor to convert to pCi/m2 1.38 = index of sensitivity d B = number of background counts in time interval t p = surveyor efficiency i = instrument efficiency for the emitted radiation s = source efficiency in emissions/disintegrations A = probes sensitive area in cm2 t = time interval of the observation while the probe passes over the source, in seconds Based on the measurements described in NUREG/CR-6364, Human Performance in Radiological Survey Scanning, issued August 2000, a surveyor efficiency p of 0.5 represents a mean value for normal field conditions and its use is generally acceptable. If the licensee wants to determine a value appropriate for particular measurement techniques, the information in NUREG/CR-6364 describes how to determine the value. NUREG-1507, Revision 1, Minimum Detectable Concentrations with Typical Radiation Survey Instruments for Various Contaminants and Field Conditions, issued August 2020, contains additional information on the interpretation of results reported in NUREG/CR-6364.

For scanning soil with a sodium iodide gamma detector, the MDCscan values given in Table 6.7 of MARSSIM provide an acceptable estimate of MDCscan.

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For static measurements of surface concentrations, the MDCstatic may be calculated using the following equation (from Equation 3-10 in NUREG-1507, Revision 1).

3+4.65

=

(A-3) where MDCstatic = minimum detectable concentration in pCi/m2 or pCi/gram (g)

B = background counts during measurement time interval t t = counting time in seconds K = a calibration constant (best estimate) to convert counts/second to pCi/m2 or pCi/g, discussed further in NUREG-1507 Section 6.7.1 of MARSSIM shows an example using this equation.

The instruments used for sample measurements at the specific sample locations should have an MDCstatic less than 50 percent of the DCGLW, as recommended in Section 4.7.1 of MARSSIM. There is no specific recommendation for the MDCscan, but the MDCscan will determine the number of samples needed, as discussed in Section A.8.6 of this appendix.

The licensee should record all numerical values measured, even values below the MDC or critical level, including values that are negative (when the measured value is below the average background). Entries for measurement results should not be nondetect, below MDC, or similar entries, because the statistical tests can only tolerate a maximum of 40 percent nondetects.

A.6.2 Instrument Calibration and Response Checks NRC regulations at 10 CFR 20.1501(c) require that the licensee periodically calibrate radiation measurement instruments used in surveys such as the FSS.

For in situ gamma measurements, the detector efficiency (count rate per unit fluence rate) should be determined for the gamma energies of interest and the assumed representative depth distribution. The surface and volumetric distributions should be explicitly considered to evaluate potential elevated areas. To calibrate for the representative depth distribution, acceptable methods are to (1) use a test bed with radioactive sources distributed appropriately or (2) use primarily theoretical calculations that are normalized or verified experimentally using a source approximating a point source. The calibration of the source used for the verification source should be traceable to a recognized standards or calibration organization, for example, the National Institute of Standards and Technology.

Some modern instruments are very stable in their response. Thus, as long as the licensee periodically performs instrument response checks to verify that the detector is operating properly, it may be acceptable to calibrate only initially without periodic recalibrations. The initial calibration may be performed by either the instrument supplier or the licensee, but in either case, 10 CFR 20.2103(a) requires that a record describing the calibration be available for NRC inspection.

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A.6.3 Instrument Response Checks Licensees should check the response of survey instruments with a check source each day before use to confirm constancy in instrument response and establish criteria for acceptable response. If the response is not acceptable, the licensee should consider the instrument as not responding properly and should not use it until the problem has been resolved. Measurements made after the last acceptable response check should be evaluated and discarded, if appropriate.

The check source should emit the same type of radiation (i.e., alpha, beta, gamma) as the radiation being measured and should give a similar instrument response, but the check source does not have to use the same radionuclide as the radionuclide being measured.

A.7 Scanning Coverage Fractions and Investigation Levels Scanning is performed to locate small areas of elevated concentrations of residual radioactivity to determine whether they meet the radiological criteria for license termination. The licensee should perform scanning in each survey unit to detect areas of elevated concentrations. The licensee should establish ILs for investigating significantly elevated concentrations of residual radioactivity. Table A.2 shows acceptable scanning coverage fractions and scanning ILs for buildings and land areas. This table is based on MARSSIM Roadmap Tables 2 and 5.8.

Systematic scans are those conducted according to a preset pattern. Judgmental scans are those conducted to include areas with a greater potential for residual radioactivity. In Class 2 areas, a 10 percent scanning coverage would be appropriate when there is high confidence that all locations would be below the DCGLW. A coverage of 25 percent to 50 percent would be appropriate when there may be locations with concentrations near the DCGLW. A coverage of 100 percent would be appropriate, if there is any concern that the area should have had a Class 1 classification rather than a Class 2 classification. In Class 3 areas, scanning coverage is usually less than 10 percent. If any location exceeds the scanning IL, scanning coverage in the vicinity of that location should be increased to delineate the elevated area.

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Table A.2 Scanning Coverage Fractions and Scanning Investigation Levels Class Scanning Coverage Fraction Scanning Investigation Levels 1 100% >DCGLEMC 2 10 to 100% for soil and for floors and >DCGLW or >MDCscan if MDCscan is lower walls of buildings, 10 to 50% for greater than DCGLW upper walls and ceilings of buildings, systematic and judgmental 3 Judgmental >DCGLW or >MDCscan if MDCscan is greater than DCGLW Sometimes the sensitivity of static measurements at designated sample points is high enough to detect significantly elevated areas between sample points. If the sensitivity is high enough, only this single set of measurements is necessary. For example, both scanning and sampling for cobalt-60, which emits an easily detectable gamma, can be done with a single set of in situ measurements in some cases.

A.8 Determining the Number of Samples Needed A minimum number of samples are needed to obtain sufficient statistical confidence that the conclusions drawn from the samples are correct. The method described below from Chapter 5 of MARSSIM is acceptable for determining the number of samples needed.

A.8.1 Determination of the Relative Shift The required number of samples will depend on a ratio involving the concentration to be measured relative to the variability in the concentration. The ratio to be used is called the relative shift, /s. The relative shift, /s, is defined in Section 5.5.2.2 of MARSSIM as:

= (A-4) where DCGLW = derived concentration guideline LBGR = concentration at the lower bound of the gray region. The LBGR is the concentration to which the survey unit must be cleaned to have an acceptable probability of passing the test (i.e., 1).

s = an estimate of the standard deviation of the concentration of residual radioactivity in the survey unit (which includes real spatial variability in the concentration as well as the precision of the measurement system)

The value of s is determined either from existing measurements or by taking limited preliminary measurements of the concentration of the residual radioactivity in the survey unit at about 5 to 20 locations, as recommended in Section 5.5.2.2 of MARSSIM. If a reference area will be used and the estimate of the standard deviation in the reference area, r, is larger than the estimate A-11

of the standard deviation in the survey unit, s, then the larger value should be used in the equation.

The LBGR should be set at the mean concentration of residual radioactivity that is estimated to be present in the survey unit. However, if no other information is available on the survey unit, the LBGR may be initially set equal to 0.5 DCGLW, as recommended by MARSSIM. If the relative shift, /s, exceeds 3, the LBGR should be increased until /s is equal to 3. The licensee may refer to Section 5.5.2.2 of MARSSIM for additional details and information.

A.8.2 Determination of Acceptable Decision Errors A decision error is the probability of making an error in the decision on a survey unit by failing a survey unit that should pass or by passing a survey unit that should fail. When using the statistical tests, larger decision errors may be unavoidable when encountering difficult or adverse measuring conditions. This is particularly true when trying to measure residual radioactivity concentrations close to the variability in the concentration of those materials in natural background.

The decision error is the probability of passing a survey unit where the actual concentration exceeds the release criterion. A decision error of 0.05 is acceptable under the more favorable conditions when the relative shift, /s, is large (about 3 or greater). Larger values of may be considered when the relative shift is small, to avoid an unreasonable number of samples. The decision error is the probability of failing a survey unit where the actual concentration is equal to LBGR. Any value of is acceptable to the NRC.

A.8.3 Number of Samples Needed for the Wilcoxon Rank Sum Test The minimum number of samples, N, needed in each survey unit for the WRS test may be determined from the following equation (adapted from MARSSIM Equation 5-1 with N redefined as the number of samples in the survey unit):

( )

2 1 Z1 + Z1 N=

( )

2 (A-5) 2 3 Pr 0.5 where N = the number of samples in the survey unit Z1 = the percentile represented by the decision error Z1 = the percentile represented by the decision error Pr = the probability that a random measurement from the survey unit exceeds a random measurement from the background reference area by less than the DCGLW when the survey unit median is equal to the LBGR concentration above background 1/2 = a factor added to MARSSIM Equation 5-1 because N is always defined in this guide as the number of samples in the survey unit Tables 5.1 and 5.2 of MARSSIM contain values of Pr, Z1, and Z1. N is the minimum number of samples necessary in each survey unit. An additional N samples will also be needed in the reference area. If N is not an integer, the number of samples is determined by rounding up. In addition, the licensee should consider taking some additional samples (MARSSIM recommends 20 percent) to protect against the possibility of lost or unusable data. Fewer samples increase A-12

the probability of an acceptable survey unit failing to demonstrate compliance with the radiological criteria for release.

A.8.4 Number of Samples Needed for Sign Test The number of samples N needed in a survey unit for the Sign test may be determined from the following equation (adapted from MARSSIM Equation 5-2):

(Z )

2 1

+ Z1 N= (A-6) 4( Sign p 0.5) 2 where N = number of samples needed in a survey unit Z1 = percentile represented by the decision error Z1 = percentile represented by the decision error Sign p = estimated probability that a random measurement for the survey unit will be less than the DCGLW when the survey unit median concentration is actually at the LBGR.

Tables 5.2 and 5.4 of MARSSIM contain the values of Z1, Z1, and Sign p. In addition, the licensee should consider taking some additional samples (MARSSIM recommends 20 percent) to protect against the possibility of lost or unusable data. Fewer samples increase the probability of an acceptable survey unit failing to demonstrate compliance with the radiological criteria for release. If a survey unit fails to demonstrate compliance because there were not enough samples taken, a totally new sampling effort may be needed unless resampling was anticipated.

A.8.5 Use of Two-Stage or Double Sampling It may be desirable for a licensee to sample a survey unit a second time to determine compliance. Two-stage sampling and double sampling are two methods by which additional survey unit data can be acquired. Two-stage sampling refers to survey designs specifically intended to be conducted in two stages. Double sampling refers to the case when the survey unit design is a one-stage design, but allowance is made for a second set of samples to be taken if the retrospective power of the test using the first set of samples does not meet the design objectives. Use of either method should be considered as part of the DQO process when developing the design of the FSS. Appendix C of this volume contains information on the use of two-stage or double sampling.

A.8.6 Additional Samples for Elevated Measurement Comparison in Class 1 Areas Additional samples may be needed when the concentration that can be detected by scanning, MDCscan, is larger than the DCGLW. The licensee should determine whether additional samples are needed in Class 1 survey units for the EMC when the concentration that can be detected by scanning, MDCscan, is larger than the DCGLW. The method in Section 5.5.2.4 of MARSSIM to determine whether additional samples are needed is acceptable to the NRC staff and is described here.

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The area factor is the multiple of the DCGLW that is permitted in a limited portion of the survey unit. In Equation A-7, the ratio of the MDCscan to the DCGLW establishes the area factor (the multiple of the DCGLW) that can be detected by scanning (adapted from MARSSIM Equation 5-4):

MDCscan area factor = (A-7)

DCGLW Using the methods in NUREG-1549, Decision Methods for Dose Assessment To Comply with Radiological Criteria for License Termination, Draft Report for Comment, issued July 1998, the licensee can determine the size of the area corresponding to the area factor, AEC. The number of sample points that may be needed to detect this area of elevated measurement concentration, NEMC, in a survey unit is:

A N EMC = (A-8)

AEC where A = the area of the survey unit AEC = the area of concentration greater than DCGLW If NEMC is larger than N, additional samples may be needed to demonstrate that areas of elevated concentrations meet the radiological criteria for license termination. However, the number of samples needed is not necessarily NEMC. The licensee can use the HSA and site characterization to determine how many additional samples may be needed. Based on what is known about the site, it may be possible to estimate a concentration that is unlikely to be exceeded. If there is a maximum concentration, the size of the area corresponding to this area factor for this concentration may be used for AEC in Equation A-8. Similarly, based on knowledge of how the radioactive material was handled or dispersed on the site, it may be possible to estimate the smallest area likely to have elevated concentrations. If this is so, that area can be used in Equation A-8. Likewise, the licensee could use the knowledge of how the residual radioactivity would be likely to spread or diffuse after deposition to determine an area AEC for Equation A-8.

Figure D-7 of Appendix D to MARSSIM and Section 3.7.2 of NUREG-1505 show that a triangular grid is slightly more effective in locating areas of elevated concentrations. Therefore, a triangular grid generally should be used if NEMC is significantly larger than N and if areas similar in size or smaller than the grid spacing are expected to have concentrations at or above the area factor.

A.9 Determining Sample Locations For the impacted areas, the licensee should establish a reference coordinate system, which is a set of intersecting lines referenced to a fixed site location or benchmark. Reference coordinate systems are established so that the locations of any point in the survey unit can be identified by coordinate numbers. A reference coordinate system does not establish the number of sample points or determine where samples are taken. A single reference coordinate system may be used for a site, or different coordinate systems may be used for each survey unit or for a group of survey units. Section 4.8.5 of MARSSIM describes an acceptable method to establish a reference coordinate system.

A-14

In Class 1 and Class 2 areas, the sampling locations are established in a regular pattern, either square or triangular. The method described below is from Section 5.5.2.5 of MARSSIM.

After the number of samples needed in the survey unit has been determined, and the licensee has decided whether to use a square or triangular grid, sample spacings, L, are determined from Equations A-9 and A-10 (adapted from MARSSIM Equations 5-5, 5-6, 5-7, and 5-8).

A L= for a triangle grid (A-9) 0.866 N A

L= for a square grid (A-10)

N where A = the survey unit area N = the number of samples needed (in Class 1 areas, the larger of the number for the statistical test or the EMC).

The calculated value of L is then often rounded downward to a shorter distance that is easily measured in the field.

A random starting point should be identified for the survey pattern. The coordinate location of the random starting point should be determined by a set of two random numbers, with one representing the x axis and the other, the y axis. The random numbers can be generated by calculator or computer or can be obtained from a table of random numbers. Each random number should be multiplied by the appropriate survey unit dimension to provide a coordinate relative to the origin of the survey unit reference coordinate system.

Beginning at the random starting point, a row of points should be identified parallel to the x axis at intervals of L. For a square grid, the additional rows should be parallel to the first row at a distance of L from the first row. For a triangular grid, the distance between rows should be 0.866 L, and the sample locations in the adjacent rows should be midway on the x axis between the sample locations in the first row. Sample locations selected in this manner that either do not fall within the survey unit area or cannot be surveyed because of site conditions should be replaced with other sample locations determined using the same random selection process that was used to select the starting point. MARSSIM Figure 5.5 contains an example illustrating the triangular grid pattern.

In Class 3 survey units and in reference areas, all samples should be taken at random locations.

Each sample location should be determined by a set of two random numbers, one representing the x axis and the other the y axis. Each set of random numbers should be multiplied by the appropriate survey unit dimension to provide coordinates relative to the origin of the survey unit reference coordinate system. Coordinates identified in this manner that do not fall within the survey unit area or that cannot be surveyed because of site conditions should be replaced with other sample locations determined in the same manner. MARSSIM Figure 5.4 illustrates a random sample location pattern.

A-15

A.10 Determining Compliance The licensee should first review the measurement data to confirm that the survey units were properly classified. MARSSIM Section 8.2.2 contains methods for this review that are acceptable to the NRC staff. If the FSS shows that an area was misclassified with a less restrictive classification, the area should receive the correct classification and the FSS for the area should be repeated. A pattern of misclassifications that are not restrictive enough indicates that the characterization was not adequate. In this case, the site or portions of the site in question should be characterized again, reclassified, and resurveyed for the new classification. The licensee should then determine whether the measurement results demonstrate that the survey unit meets the radiological criteria for license termination.

Tables A.3 and A.4, below, summarize an acceptable way to interpret the sample measurements. MARSSIM Section 8.4 describes the WRS test, while MARSSIM Section 8.3 describes the Sign test and MARSSIM Section 8.5 describes the EMC. The elevated measurement is applied to all sample measurements and all scanning results that exceed the DCGLW.

In some cases, licensees may choose to use scanning or fixed measurement techniques that assess 100 percent of the population of potential direct measurements or samples within the survey unit. For these cases, it may be reasonable to demonstrate compliance by directly comparing the average radionuclide concentrations determined from the survey with the appropriate DCGLW, without the need to perform statistical tests. Guidance has not yet been developed for using such techniques without performing statistical tests; therefore, licensees should discuss such techniques with the NRC staff on a case-by-case basis.

Table A.3 Interpretation of Sample Measurements when a Reference Area is Used Measurement Results Conclusion Difference between maximum survey unit concentration and Survey unit meets release minimum reference area concentration is less than DCGLW. criterion.

Difference between survey unit average concentration and Survey unit fails.

reference area average concentration is greater than DCGLW.

Difference between any survey unit concentration and any Conduct WRS test and EMC.

reference area concentration is greater than DCGLW and the difference of survey unit average concentration and reference area average concentration is less than DCGLW.

A-16

Table A.4 Interpretation of Sample Measurements when No Reference Area is Used Measurement Results Conclusion All concentrations are less than DCGLW. Survey unit meets release criterion.

Average concentration is greater than DCGLW. Survey unit fails.

Any concentration is greater than DCGLW and average Conduct Sign test and EMC.

concentration less than DCGLW.

Some facilities may have residual radioactivity composed of more than one radionuclide. When there are multiple radionuclides rather than a single radionuclide, the licensee should consider the dose contribution from each radionuclide. Section 2.7 of this volume contains information about using the sum of fractions approach for compliance when multiple radionuclides are present.

When there is a fixed ratio among the concentrations of the nuclides, a DCGLW for each nuclide can be calculated. Compliance with the radiological criteria for license termination may be demonstrated by comparing the concentration of the single surrogate radionuclide that is easiest to measure with its DCGLW (which has been modified to account for the other radionuclides present). For example, if Cs-137 and Sr-90 are present, using measured concentrations of Cs-137 as a surrogate for the mix of Cs-137 and Sr-90 may be simpler than separately measuring Cs-137 and Sr-90, and may thus save labor and analytical expenses.

When using a surrogate radionuclide to represent the presence of other radionuclides, a sufficient number of measurements, spatially distributed throughout the survey unit, should be used to establish a consistent ratio between the surrogate and the other radionuclides.

Section 4.3.2 of MARSSIM provides additional information on the use of surrogate radionuclides for surveys.

When there is no fixed ratio among the concentrations of the nuclides, the licensee must evaluate the concentration of each nuclide via sampling. Compliance with the radiological criteria for license termination is then demonstrated by considering the sum of the concentration of each nuclide relative to its DCGLW, followed by an evaluation of all radionuclides of concern via the unity rule. Chapter 11 of NUREG-1505 describes an acceptable method for performing the evaluation. When there is no fixed ratio among the concentrations of the nuclides, and a large number of discrete samples are required, it may be possible to utilize a composite sampling strategy to increase the probability of elevated area detection and as a means to reduce analytical cost. However, this approach requires an evaluation of the exposure scenario related to the hard-to-detect radionuclide and should be performed on a case-by-case basis along with discussions with the regulator. Additional information on composite sampling can be found in Appendix O of this volume.

In some cases in which multiple nuclides are present with no fixed ratio in their concentrations, the dose contribution from one or more of the nuclides in the mixture will dominate the total dose, and the dose from other radionuclides will be insignificant. For example, at a nuclear power plant, many different radionuclides could be present with no fixed ratio in their A-17

concentrations, but almost all of the dose would come from just one or two of the nuclides.

Section 3.3 of this volume contains guidance on elimination of radionuclides or pathways from consideration.

If a survey unit fails, the licensee should evaluate the measurement results and determine why it failed. MARSSIM, in Sections 8.2.2 and 8.5.3, and in Appendix D, provides acceptable methods for reviewing measurement results. If it appears that the failure was caused by the presence of residual radioactivity in excess of that permitted by the radiological release criteria, the survey unit should be re-remediated and resurveyed. However, some failures may not be caused by the presence of residual radioactivity. If it can be determined that this is the case, the survey unit may be released.

A.11 References Nuclear Regulatory Commission (U.S.) (NRC). Measurement Methods for Radiological Surveys in Support of New Decommissioning Criteria, Draft Report for Comment, NUREG-1506. NRC: Washington, DC. August 1995.

. Decision Methods for Dose Assessment to Comply with Radiological Criteria for License Termination, Draft Report for Comment, NUREG-1549. NRC: Washington, DC.

1998a.

. Human Performance in Radiological Survey Scanning, NUREG/CR-6364.

NRC: Washington, DC. 1998b.

. Minimum Detectable Concentrations with Typical Radiation Survey Instruments for Various Contaminants and Field Conditions, NUREG-1507, Rev. 1. NRC: Washington, DC. 2020.

. A Proposed Nonparametric Statistical Methodology for the Design and Analysis of Final Status Decommissioning Surveys-Interim Draft Report for Comment and Use, NUREG-1505, Rev. 1. NRC: Washington, DC. 1998d.

. Multi-Agency Radiation Survey and Site Investigation Manual (MARSSIM),

NUREG-1575, Rev. 1. EPA 402-R-97-016, Rev. 1; DOE/EH-0624, Rev. 1; U.S. Department of Defense; U.S. Department of Energy; U.S. Environmental Protection Agency; and NRC:

Washington, DC. August 2000.

. Regulatory Issue Summary 2002-02, Lessons Learned Related to Recently Submitted Decommissioning Plans and License Termination Plans. NRC: Washington, DC.

January 2002.

A-18

APPENDIX B SIMPLE APPROACHES FOR CONDUCTING FINAL RADIOLOGICAL SURVEYS

B.1 Introduction A large number of licensees may use a simplified method to demonstrate regulatory compliance for decommissioning, and thereby avoid complex FSSs. For Decommissioning Groups 1-3, licensees may use the simplified FSS method described in Appendix B to NUREG-1575, Revision 1, Multi-Agency Radiological Survey and Site Investigation Manual (MARSSIM),

issued August 2000, or the alternative protocol described in this appendix (Section B.3 below).

B.2 MARSSIM Simplified Method The simplified method in Appendix B of MARSSIM may be used by Decommissioning Group 1 and some Decommissioning Group 2 licensees. These are sites where radioactive materials have been used or stored only in the form of (1) nonleaking, sealed sources, (2) short half-life radioactive materials (e.g., T1/2 120 days) that have since decayed to insignificant quantities, (3) small quantities exempted or not requiring a specific license, or (4) a combination of the above. MARSSIM, Revision 1, Appendix B gives the details of this simplified method.

B.3 Alternative Simplified Method This alternative method may be used by Decommissioning Groups 1-3 and is applicable only for surfaces of building structures and for surface soils. The following conditions are prerequisite to the use of this method:

  • screening values are applicable and being used to demonstrate compliance with release criteria (i.e., site meets underlying assumptions in screening analyses discussed in more detail in Appendix H)
  • removable residual radioactivity on building surfaces 10 percent or less; or adjusted to account for higher removable fractions as discussed in Appendix H of this volume
  • no sources requiring complex or special surveys are present (e.g., no (i) volumetric building structure residual radioactivity, (ii) duct work, (iii) embedded piping, (iv) groundwater residual radioactivity, (v) subsurface soil residual radioactivity, (vi) buried conduit, (vii) sewer pipes, or (viii) prior onsite disposals)
  • not to be applied to land areas where soil has been previously remediated
  • minimum detectable concentrations between 10 and 50 percent of the DCGLW for scans, static or direct measurements, and sampling and analysis (using guidance in NUREG-1507, Revision 1, Minimum Detectable Concentrations with Typical Radiation Survey Instruments for Various Contaminants and Field Conditions, issued August 20201)

If the above conditions are met, then the following simplified method may be used to design and conduct the FSS for each survey unit.

  • Size is limited to 2000 m2 for land areas and 100 m2 for structures.

1 Revision 1 of NUREG-1507 was issued in August 2020.

B-1

  • Scanning and sampling are to be performed as follows:

o 100-percent scan o 30 samples

  • The hot spot criterion is three times the DCGLW, applied to any sampling location.
  • A quality control program ensures results are accurate and sources of uncertainty are identified and controlled.
  • The average concentration for the survey unit is compared to the DCGLW.
  • Statistical tests that may be used include the parametric Students t-test, or non-parametric WRS test when the radionuclide(s) of concern are in background, or the non-parametric Sign test when the radionuclide(s) of concern are not in background assuming an alpha () or false positive error of 5 percent. No statistical tests are needed if all measurements are less than the DCGLW. MARSSIM (NUREG-1575) can be consulted for additional information on the statistical tests.

The FSSR should provide a complete and unambiguous record of the radiological status of the site and should stand on its own with minimal information incorporated by reference (see Appendix D of this volume for additional information on reporting survey results).

B.4 References Nuclear Regulatory Commission (U.S.) (NRC). Minimum Detectable Concentrations with Typical Radiation Survey Instruments for Various Contaminants and Field Conditions, NUREG-1507, Rev. 1. NRC: Washington, DC. 2020.

. Multi-Agency Radiation Survey and Site Investigation Manual, (MARSSIM),

NUREG-1575. NRC: Washington, DC. 2000.

B-2

APPENDIX C USE OF TWO-STAGE OR DOUBLE SAMPLING FOR FINAL STATUS SURVEYS

C.1 Introduction This appendix contains information on the assessment of survey data when double or two-stage sampling in a survey unit is used to determine compliance. These sampling strategies utilize the initial sample data and a second round of supplemental sampling. This approach may be desirable when a survey unit fails the hypothesis test (i.e., the decision is made that the survey unit does not meet the release criterion) due to insufficient power as described in more detail below,1 but the mean of the measured data is below the release criterion. Further information on survey unit failures and possible remedies using the DQO process are discussed in Chapter 8 of NUREG-1575, Revision 1, Multi-Agency Radiological Survey and Site Investigation Manual (MARSSIM), issued August 2000.

In this appendix two-stage sampling denotes survey designs specifically intended to be conducted in two stages, whereas double sampling refers to one-stage survey designs that allow for the collection of a second set of samples if the retrospective power of the test using the first set of samples does not meet the design objectives. Double or two -stage sampling may be acceptable if it is incorporated into the final status survey design. Thus, before the initial round of sampling occurs, allowances for these sampling approaches should be discussed in the DQOs for the final status survey.

In general, adequate initial sampling to achieve the desired statistical power and error rates, or data collection in two stages, are both preferable to double sampling. As discussed in the next sections of this appendix, there are several factors to consider when additional random samples are collected. For example, double sampling is also generally not appropriate for Class 1 survey units having confirmed areas of elevated activity, or Class 2 or Class 3 survey units because the need for a second set of samples raises the issue of survey unit misclassification.

C.2 Double Sampling As noted in the preceding section C.1, situations can occur where a survey unit might have passed the final status statistical test, had the initial sampling design been powerful enough to reject the null hypotheses in Scenario A. That is, a retrospective examination of the power of the statistical tests reveals that the probability of detecting that the survey unit actually meets the release criterion was lower than that planned for during the DQO process. This could occur if the spatial variability in residual radioactivity concentrations was larger than anticipated. The power of the test specified during the DQO process depends on an estimate of the uncertainty.

The power of the statistical test will be less than planned if the standard deviation is higher than expected. If samples were lost, did not pass analytical QA/QC, or are otherwise unavailable for inclusion in the analysis, the power will also be lower than was planned. In these situations, it might be desirable to take additional samples in the survey unit to improve the power of the statistical test.

Draft NUREG/CR-5849, Manual for Conducting Radiological Surveys in Support of License Termination, issued June 1992, allowed the licensee to take additional samples in a survey unit if, after the first sampling, the mean was less than the DCGLW, and the desired upper 1 Power is the probability of rejecting the null hypothesis when it is false (i.e., in Scenario A, concluding that the site is clean when it is clean). The power is equal to one minus the Type II (false negative) error rate (i.e., (1-)). The power of the sampling design is important to decision making and can be determined prospectively when planning the survey and retrospectively when interpreting survey results. Additional information on prospective and retrospective power analyses is provided in Chapter 8 and Appendix I of MARSSIM, Revision 1.

C-1

confidence level on the mean was greater than the DCGLW. Because a 95-percent confidence interval is constructed using Students t-statistic rather than using a hypothesis test, the survey design does not consider Type II errors. The second set of samples was taken so that a Students t-test on the combined set of samples would have 90-percent power at the mean of the first set of samples, given the estimated standard deviation from the first set of samples.

Such double sampling was to be performed only once.

Increasing the probability that a clean survey unit passes (power in Scenario A) using double sampling will also tend to increase the probability that a survey unit that is not clean will pass (Type I error). In addition, the two tests are not independent, because the data from the first set of samples is used in both. The increase in the Type I error rate is expected to be less than a factor of two based on the analysis in the following paragraphs of this appendix. The potential increase in the Type I error rate should be considered when designing the survey and determining acceptable error rates as part of the DQO process.

Two-stage or double sampling is not usually expected (nor is it encouraged) when the DQO process is used, as in MARSSIM, Revision 1. This is because the Type II error and the power desired are explicitly considered in the survey design process. If higher power in the test is desired, it should be specified as such. Sufficient samples should be taken to achieve the specified power. The value of this approach lies in the greater objectivity and defensibility of the decision made using the data. Nonetheless, it is recognized that there may be instances when some sort of double sampling is considered desirable. As discussed above, an example is when it is difficult to estimate the standard deviation of the concentrations in a survey unit. A first set of data may be taken with an estimated standard deviation that is too low, and thus, the power specified in the DQO process may not be achieved.

For planning purposes, it is important to understand the circumstances under which it is appropriate to combine this data with additional sample data to be used in the test of the final status survey. Consider the Sign test, as indicated in NUREG-1505, A Proposed Nonparametric Statistical Methodology for the Design and Analysis of Final Status Decommissioning Surveys-Interim Draft Report for Comment and Use, issued June 1998.

Suppose N1 samples are taken. For the Sign test in Scenario A, the test statistic, S1, is equal to the number of survey unit measurements below the DCGLW. If S1 exceeds the critical value k1, then the null hypothesis that the median concentration in the survey unit exceeds the DCGLW is rejected (i.e., the survey unit passes this test). The probability that any single survey unit measurement falls below the DCGLW is found from 2

1 ()

( ) = ( ) = 22 = ( ) (C-1) 2 C is the true, but unknown, mean concentration in the survey unit. Assuming the data are normally distributed, when C = DCGLW, then p = 0.5. More generally, if C is the true, but unknown, median concentration in the survey unit, p is also equal to 0.5.

The probability that more than k1 of the N1 survey unit measurements fall below the DCGLW is simply the following binomial probability:

C-2

N1 t k N1 t

( ) ( )

N1 N t 1 N t

=

1 1 p 1 p 1 p 1 p (C-2) t= k +1 t 1 t=0 t This is the probability that the null hypothesis will be rejected, and it will be concluded that the survey unit meets the release criterion. When the median concentration in the survey unit is at the DCGLW, this probability is just the Type I error rate, . When C = DCGLW, p = (1 p) = 0.5, so 1 1

= (1 )(0.5) (0.5)1 = (0.5)1

1 +1 1 +1 (1) (C-3)

Suppose it is decided to allow the licensee to take a second set of samples of size N2. The test statistic, S, is equal to the number of the total of N = N1 + N2 survey unit measurements below the DCGLW. If S exceeds the critical value k, then the null hypothesis that the median concentration in the survey unit exceeds the DCGLW is rejected (i.e., the survey unit passes this test). In this case, the overall probability that the null hypothesis is rejected (i.e., the survey unit passes) is equal to the sum of the probabilities of the two events, labeled (Event 1) and (Event

2) below, that are mutually exclusive:

(Event 1) the probability that more than k1 of the N1 survey unit measurements fall below the DCGLW, and (Event 2) the probability that fewer than k1 of the first N1 survey unit measurements fall below the DCGLW but that more than k of the N total survey unit measurements fall below the DCGLW.

The test statistic, S, is then equal to the sum of S1 and S2, (S = S1 + S2), where S2 is the number of the second set of N2 survey unit measurements that fall below the DCGLW. S1 and S2 are independent, but S1 and S = S1 + S2 are not.

The covariance of S1 and S using E() to denote expected values, is Cov ( S1 , S ) = E ( S1 , S ) E ( S1 ) E ( S )

= E ( S1 ( S1 + S2 )) E ( S1 ) E ( S )

= E ( S 2 ) + E ( S1 S2 ) E ( S1 ) E ( S )

1 (C-4)

= ( N 12 p(1 p) + N 12 p 2 ) + N 1 N 2 p 2 N 1 p( N 1 + N 2 ) p

= N 1 p(1 p)

C-3

Therefore, the correlation coefficient between S1 and S is N1 (1)

(1 , S) =

1 (1)(1 +2 )(1)

N1

=

1 (1 +2 )

(C-5) 1

=

(1 + 2 1

=

To calculate the overall probability that the survey unit passes, one requires the joint probability of S1 and S, Pr( S1 = s1 , S = s) = Pr ( S1 = s1 ) Pr ( S2 = s s1 )

N1 N2 s s

= p s (1 p) p (1 p)

N s N 2 ( s s1 )

1 1 1 1 s1 s s1 (C-6)

N1 N2 s p (1 p)

Ns

=

s1 s s1 Therefore, the overall probability that the survey unit passes is

( )

Pr S1 k1 or S k = Pr(S1 k1 ) + Pr(S1 k1 and S k )

N1 N1 s1

= p (1 p ) N1 s1 s1 = k1 +1 s1 (C-7)

N1 N1

+ p s1 + s 2 (1 p )( N1 + N 2 ) ( s1 + s 2 )

s1 k1 s 2 k s1 s1 s2 The first term is equal to (or slightly less than) the Type I error rate specified during the DQO process. The second term is the additional probability of a Type I error introduced by allowing double sampling.

C-4

Note that

( ) N1 N 2 N k1 Pr S1 k1 and S k = p s (1 p ) N s sk s1 = 0 s1 k s1 N k N N p s (1 p) N S 1 2 (C-8) sk s1 = 0 s1 k s1 N

N

= p s (1 p) N s = Pr(S k )

sk s Thus, the Type I error rate would be at most doubled when double sampling is allowed.

For example, if a survey is designed so that N1 = 30, and = 0.05, then the critical value for the Sign test is k1 = 19. Suppose the first survey results in 19 or fewer measurements that are less than the DCGLW. In addition, suppose the survey unit is sampled again, taking an additional N2 = 30 samples. Then the total number of samples is N = N1 + N2 = 60. The critical value for the Sign test with = 0.05 and N = 60 is k = 36. When the survey unit concentration is equal to the DCGLW, p = 0.5, one has Pr ( S1 19 or S 36) = Pr ( S1 19) + Pr ( S1 19 and S 36) 30 s (0.5) (1 0.5) 30 30 s1

=

s1 s1 = 20 1 (C-9) 30 30 (1 0.5) ( 30 + 30 ) ( s1 + s2 )

19 30

+ (0.5) s1 + s2 s = 0 s1 1 s2 = ( 37 s ) s2 1

= 0.049 + 0.027 = 0.076 Thus, the total Type I error rate is about 50 percent greater than originally specified.

In conclusion, double sampling should not be used as a substitute for adequate planning. If it is to be used, it should be considered part of the DQO process. The procedure for double sampling (i.e., the size of the second set of samples, N2,) should be specified, recognizing that the Type I error rate could be up to twice that specified for the Sign test when only one set of samples is taken.

Similar considerations apply for the WRS test; however, the calculation of the exact effect on the Type I error rate is considerably more complex and is not discussed in this appendix.

Finally, double sampling should never be necessary for Class 2 or Class 3 surveys, which are not expected to have concentrations above the DCGLW. These classes of survey unit should always pass after the first set of samples, because every measurement should be below the DCGLW. The need for a second set of samples (i.e., failure to reject the null hypothesis) in Class 2 or Class 3 survey units would raise an issue of survey unit misclassification. In addition, double sampling is generally not appropriate for Class 1 survey units where elevated areas have been found.

C-5

In lieu of double sampling, a preferred approach is to plan for data collection in two stages and design the final status survey accordingly, as is discussed in the remainder of this appendix.

C.3 Two-Stage Sequential Sampling In some cases, two-stage sampling may be used instead of a single-stage sample design. For example, if there are a large number of survey units of a similar type to be tested, a two-stage sampling procedure may result in substantial savings of time and money by reducing the average number of samples required to achieve a given level of statistical power.

An example of a two-stage sampling design using the Sign test is summarized here. In this example, N1 is the size of the first set of samples taken, and S1 is the number of these samples that are less than the DCGLW. Similarly, N2 is the size of the second set of samples taken, and S2 is the number of these samples that are less than the DCGLW. Let N = N1 + N2, and let S = S1 + S2. The procedure is as follows:

  • If S1 > u1 then the survey unit passes (reject H0).
  • If S1 < l1 then the survey unit fails.
  • If l1 S1 u1 then the second set of samples is taken.
  • If S = S1 + S2 > u2 after the second set of samples is analyzed, then the survey unit passes.

An advantage of two-stage sampling is that it can reduce the total number of samples if there are many similar survey units of similar design. For given error rates and , the number of samples, N1, taken in the survey unit during the first stage of sampling will be less than the number, N0, required in the MARSSIM, Revision 1, tables. Unless the result is too close to call, this will be the only sampling needed. When the result is too close to call, l1 S1 u1, a second sample of size N2 is taken, and the test statistic S2 is computed using the combined data set, N1 + N2. While the size of the combined set, N = N1 + N2, will generally be larger than the number, N0, from the MARSSIM, Revision 1, tables, the expected sample size over many survey units will still be lower. Thus, a two-stage sampling scheme will be especially useful when there are many similar survey units for which the FSS design is essentially the same.

Two-stage sampling may be used whether or not a reference area is needed (i.e., it may be used with either the Sign or the WRS test).

The remaining major issue is how to choose the critical values l1, u1, and u2. Hewett and Spurrier (1983) suggest three criteria:

(1) Match the power curve of the two-stage test to that of the one-stage test. The curves are matched at three points. The points with power equal to , 1, and 0.5 are generally well enough separated to ensure a good match over the entire range of potential survey unit concentrations.

(2) Maximize the power at the LBGR for given values of and average sample size.

(3) Minimize the sample size for given values of , and 1.

C-6

While any one of these criteria could be used, the first has received more attention in the literature. Thus, it may be more readily applied to the case of final status survey design. The other criteria would require further development.

Spurrier and Hewett (1975) initially developed a two-stage sampling methodology using criteria 1 assuming the data are normally distributed. They matched the two-stage power curve to the one-stage power curve at three points: the first at values of either 0.05 or 0.01; the second in the gray region where power is equal to 0.5; and the third at the lower bound of the gray region with a value of 0.1 and power (1- value) of 0.9. Table C.1 shows the values of l1, u1, and u2 obtained for six different sets of sample sizes, N1/N0, N2/N0, expressed as fractions of the sample size, N0, which would be required for the one-stage test with equivalent power. The term E(N)/N0, is the maximum expected combined sample size for the two-stage test relative to the sample size, N0, which would be required for the one-stage test with equivalent power. This number is almost always less than one, but it depends on how close the actual concentration in the survey unit is to the DCGLW. Clearly, if the concentration is over the DCGLW, the survey unit is likely to fail on the first set of samples. If the concentration is much lower than the DCGLW, the survey unit is likely to pass on the first set of samples. It is only when the true concentration in the survey unit falls within the gray region that there will be much need for the second set of samples. The fact that the maximum E(N)/N0 is almost always less than one indicates that the overall number of samples required for a two-stage FSS will almost never exceed the number required for a one-stage test, even if the true concentration of the survey unit falls in the gray region between the LBGR and the DCGLW.

The power to distinguish clean from dirty survey units is relatively low when the true concentration is in the gray region because the power falls from 1 at the LBGR to at the DCGLW. Thus, when the true concentration is in the gray region, a larger number of cases will require a second set of samples. The gray region is exactly where the results are too close to call. However, if the true concentration of the survey unit is below the LBGR or above the DCGLW, the actual average number of samples will be closer to N1, because the second set of samples will seldom be needed.

In 1976, Spurrier and Hewett dropped the assumption of normality and extended their methodology to two-stage Wilcoxon Signed Rank and WRS tests. The procedure depends on an extension of the Central Limit Theorem to the joint distribution of the test statistics S1 and S =

S1 + S2. Spurrier and Hewett suggest that the approximation works reasonably well for sample sizes as small as nine.

In this appendix, their method is also applied to the Sign test.

For the Sign test, one computes S1+ N 1 2 S1 = (C-10)

N1 4 where 1+ is the usual Sign test statistic (i.e., the number of measurements less than the DCGLW).

Using Table C.1,

  • if S1 > u1 then reject the null hypothesis (the survey unit passes),

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  • if S1 < l1 then do not reject the null hypothesis (the survey unit fails),
  • if l1 S1 u1 then take the second set of samples.

If a second set of samples is taken, then compute

( S1+ + S2+ ) ( N 1 + N 2 ) 2 S + N 2 S= = (C-11)

( N1 + N 2 ) 4 N 4 Using Table C.1,

  • if S > u2 then reject the null hypothesis (the survey unit passes),
  • if S u2 then do not reject the null hypothesis (the survey unit fails).

This test relies on a large sample approximation. That is, one is assuming that the sample size is large enough that the joint distribution of S1 and S is bivariate standard normal with correlation coefficient 1

(1 , ) = (C-12)

Some simulation studies may be done to determine quantitative bounds on the accuracy of this approximation.

The choice of which set of sample sizes should be used is dependent on how confident one is of passing.

For Class 2 and Class 3 survey units (discussed in Appendix A of this volume), case 3 with N1/N0 = 0.2 and N2/N0 = 1.0 might be reasonable. In these classes of survey units, no individual sample concentrations in excess of the DCGLW are expected. The probability of passing on the first set of samples should be close to one. Therefore, it makes sense to choose a design with the minimum number of samples required in the first set.

For Class 1 survey units (discussed in Appendix A of this volume), case 2 with N1/N0 = 0.4 and N2/N0 = 0.8 might be more appropriate. There is some chance that the survey unit will not pass on the first set of samples, so it may be desirable to reduce Max E(N)/N0 from 0.999 to 0.907 by taking more samples in the first set.

If the gray region has been expanded to increase /, case 1 or 4 would be a more conservative choice. In this situation, statistical power has been compromised somewhat, so it may be important to reduce the risk of having a larger average total number of samples (as indicated by the potential Max E(N)/N0 even further).

Scan sensitivity will also affect the ability to use two-stage designs in Class 1 survey units. It would have to be determined if the DCGLEMC can be detected when only N1 samples are taken.

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If not, the sample size would have to be increased until the MDCscan is lower than the DCGLEMC.

In this situation, the choice of N1, and the average savings possible with two-stage sampling may be severely limited.

Table C.1 Critical Points for Two-Stage Test of Normal Mean for a One-Sided Alternative

= 0.05 = 0.01 N1/N0 N2/N0 u1 l1 u2 Max u1 l1 u2 Max E(N)/N0 E(N)/N0 1 0.6 0.6 1.886 0.71 1.783 0.866 2.499 1.259 2.493 0.879 2 0.4 0.8 1.984 0.179 1.782 0.907 2.558 0.635 2.496 0.931 3 0.2 1 2.073 0.482 1.784 0.999 2.6 0.146 2.502 1.03 4 0.55 0.55 2.05 0.438 1.716 0.869 2.635 0.966 2.411 0.878 5 0.66 0.66 1.781 0.95 1.868 0.882 2.415 1.52 2.6 0.897 6 0.7 0.7 1.749 1.045 1.909 0.893 2.39 0.628 2.651 0.908 Source: Spurrier and Hewett (1975).

For the WRS test, at each stage, one sets the number of measurements required in the survey unit, n1 and n2, and in the reference area m1 and m2, relative to the number required for the one-stage test n0 = m0 = N0/2 specified in Table 5.3 of MARSSIM, Revision 1. There is an additional requirement that n1/n2 = m1/m2, which should be satisfied with sufficient accuracy for most MARSSIM, Revision 1 designs. Minor departures due to small differences in sample size caused by filling out systematic grids or the loss of a few samples should not severely affect the results.

One now computes W1R m1 (n1 + m1 + 1) / 2 S1 = (C-13) n1 m1 (n1 + m1 + 1) / 12 where 1 is the usual WRS test statistic (i.e., the sum of the ranks of the adjusted reference area measurements).

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Using Table C.1,

  • if S1 > u1 then reject the null hypothesis (the survey unit passes),
  • if S1 < l1 then do not reject the null hypothesis (the survey unit fails),
  • if l1 S1 u1 then take the second set of samples.

If a second set of samples is taken, then compute the following.

S=

(W + W ) (m + m )(m + m + n + n + 1) / 2 = W m(m + n +1) / 2 1

+

2

+

1 2 1 2 1 2 R

(C-14)

(m + m ) (n + n ) (m + m + n + n + 1) / 12 1 2 1 2 1 2 1 2 mn (m + n + 1) / 12 Using Table C.1,

  • if S > u2 then reject the null hypothesis (the survey unit passes),
  • if S u2 then do not reject the null hypothesis (the survey unit fails).

This test relies on a large sample approximation. That is, one is assuming that the sample size is large enough that the joint distribution of S1 and S is bivariate standard normal with correlation coefficient

( S1 , S ) = (m + n ) / (m + n) 1 1 (C-15)

Some simulation studies would be needed to determine some quantitative bounds on the accuracy of this approximation.

C.4 An Alternative Two-Stage, Two-Sample Median Test A different approach to this testing problem has been suggested by Wolfe (1977). In this procedure, a specific number of sample measurements are made in a reference area, and the median, M, is calculated and the DCGLW added. Survey unit samples are then analyzed until r of them are found to be below M. The test statistic, nr, is the number of survey unit samples that have been analyzed. Smaller values of nr indicate that the survey unit meets the release criterion. For Class 2 and Class 3 survey units, in particular, one would expect that nr = r. In that case, the number of reference area measurements, m, and the value of r are chosen to meet the DQOs for the Type I error rate. In each survey unit, r samples are taken. If all are less than M, one rejects the null hypothesis that the survey unit exceeds the release criterion. If any one of them exceeds M, the null hypothesis will not be rejected. Thus, the total number of samples needed in each survey unit may be relatively small. In addition, as soon as one sample is measured above M, the result of the test is known. Thus, it may not be necessary to analyze every survey unit sample. Of course, the need to identify elevated areas may preclude the use of this method in some circumstances. However, the potential savings when the analytical costs are high may make this procedure attractive. As stated previously, if a licensee C-10

is considering the use of approaches discussed in this appendix, contact with the NRC staff is strongly encouraged early in the planning process.

C.5 References Gogolak, C.V. Use of Two-Stage or Double Sampling in Final Status Decommissioning Surveys. ADAMS Accession No. ML20035F387. Forty-Sixth Annual Meeting of the Health Physics Society: Cleveland, OH. June 13, 2001.

Hewett, J.E. and J.D. Spurrier. A Survey of Two Stage Tests of Hypothesis: Theory and Application. Communications on StatisticsTheory and Methods. Vol. 12, No. 20:

pp. 2307-2425. 1983.

Nuclear Regulatory Commission. A Proposed Nonparametric Statistical Methodology for the Design and Analysis of Final Status Decommissioning Surveys-Interim Draft Report for Comment and Use, NUREG-1505, Rev. 1. NRC: Washington, DC. 1998.

Nuclear Regulatory Commission. Manual for Conducting Radiological Surveys in Support of License Termination, NUREG/CR-5849, Draft Report for Comment. NRC: Washington, DC.

1992.

Spurrier, J.D., and J.E. Hewitt. Double Sample Tests for the Mean of a Normal Distribution.

Journal of the American Statistical Association. Vol. 70, No. 350: pp. 448-450. 1975.

Spurrier, J.D., and J.E. Hewitt. Two-Stage Wilcoxon Tests of Hypotheses. Journal of the American Statistical Association. Vol. 71, No. 356: pp. 982-987. 1976.

Wolfe, D.A. Two-Stage Two Sample Median Test. Technometrics. Vol. 19, No. 4:

pp. 459-501. 1977.

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APPENDIX D SURVEY DATA QUALITY AND REPORTING

D.1 Introduction The MARSSIM and the MARLAP manual are complementary guidance documents in support of cleanup and decommissioning activities. The MARSSIM document provides guidance on how to plan and carry out a study to demonstrate that a site meets appropriate release criteria. It describes a methodology for planning, conducting, evaluating, and documenting environmental radiation surveys conducted to demonstrate compliance with cleanup criteria. Chapter 4 and Appendix A provide more details on MARSSIM. The MARLAP manual provides guidance and a framework for both project planners and laboratory personnel to ensure that radioanalytical data will meet the needs and requirements of cleanup and decommissioning activities.

Radioanalytical data are commonly generated to support activities such as characterization and survey of radiologically contaminated sites, effluent and environmental monitoring of nuclear facilities, emergency response to accidents involving radiological materials, cleanup and decommissioning of nuclear facilities, and radioactive waste management. Numerous significant decisions, affecting the health and safety of the public and the environment, are frequently based on the available radioanalytical data. Considering these activities, the decisions associated with the radioanalytical data may involve issues pertaining to the extent and depth of residual radioactivity and associated remedial actions; demonstration of compliance with the cleanup criteria; demonstration of compliance with the effluent release criteria; assessment of effluent radiological releases and corrective measures; assessment of actions in response to incidents or accidental releases of radiological materials; and issues involving waste storage, transport, and disposal. In addition, radioanalytical data commonly influence decisions related to the cost of remedial actions as well as decisions involving environmental monitoring strategies and designs.

The MARLAP manual was developed to provide guidance and a framework for project planners, managers, technical reviewers, and laboratory personnel to ensure that the radioanalytical data produced by surveys will meet the needs and requirements for cleanup and decommissioning activities. The MARLAP manual addresses the need for a nationally consistent approach to producing radioanalytical laboratory data that meet a projects or programs data requirements.

The guidance provided by MARLAP is both scientifically rigorous and flexible enough to be applied to a diversity of projects and programs. The MARLAP manual (NRC document NUREG-1576 and U.S. Environmental Protection Agency (EPA) document EPA 402-B-04-001A-C) is issued in three volumes (printed version and CD-ROM) and is found on the Internet at: http://www.nrc.gov/reading-rm/doc-collections/nuregs/staff/sr1576.

The NRC staff encourages licensees to follow the recommendations in MARLAP.

D.2 An Overview of MARLAP The MARLAP manual is divided into two main parts. Part I provides guidance on using a performance-based approach for the three phases of radioanalytical projects, including (1) the planning phase, (2) the implementation phase, and (3) the assessment phase. These three main phases and associated processes should result in analytical data of known quality appropriate for the intended use. Table D.1 provides an overview of the three main phases, the processes associated with each phase, and the anticipated outputs for each process.

Figure D.1 illustrates an overview of MARLAP terms and processes and interactions of the radioanalytical project manager with the laboratory performing the analysis. The MARLAP manual processes and terms described in Table D.1 and Figure D.1 are consistent with standard practices of the American Society for Testing of Materials (ASTM) for the generation of D-1

environmental data. Chapters 3 through 9 of the MARLAP manual provide a detailed description of MARLAP phases and specific processes. It should be noted that it is not a regulatory requirement to follow or use MARLAP processes as described in Figure D.1; however, these processes are believed to be flexible and scientifically rigorous to be applied for the generation of radioanalytical data of the desired quality for the intended use.

Part II of MARLAP provides technical information on the laboratory analysis of radionuclides.

Specifically, Part II highlights common radioanalytical problems and how to correct them. It also provides options for analytical protocols and discusses the pros and cons of these options. It should be noted that Part II does not provide step-by-step instructions on how to perform certain laboratory procedures or tasks. However, Part II provides guidance to assist laboratory personnel in selecting the best approach for a particular laboratory task. For example, Chapter 13 does not contain a step-by-step instruction on how to dissolve a soil sample; however, it does provide information on acid digestion, fusion techniques, and microwave digestion, to help the analyst select the most appropriate technique or approach for particular sample characteristics and project needs. Part II presents detailed technical information on (1) field and sampling issues that affect laboratory measurements, (2) sample receipt, inspection, and tracking, (3) laboratory sample preparation, (4) sample dissolution, (5) separation techniques, (6) quantification of radionuclides, (7) data acquisition, reduction, and reporting for nuclear counting instrumentation, (8) waste management in a radioanalytical laboratory, (9) laboratory quality control (QC), (10) measurement uncertainty, and (11) detection and quantification capabilities. MARLAP adopted the International Organization for Standardization (ISO) processes, terms, and expressions for analytical measurements, quantifications, and estimation of uncertainty.

MARLAP also presents technical details on specific topics outlined in Parts I and II.

Appendices A through E support Part I for the following specific topics: Appendix A, Directed Planning Approaches; Appendix B, The Data Quality Objective Process; Appendix C, Measurement Quality Objectives for Method Uncertainty and Detection and Quantification Capability; Appendix D, Content of Project Plan Documents; and Appendix E, Contracting Laboratory Services. Appendix F supports Part II for the specific topic on laboratory subsampling, and Appendix G provides a compilation of statistical tables.

D.3 Use of MARLAP in Decommissioning and Cleanup Projects MARLAP presents a useful approach and methodology applicable to radioanalytical projects for cleanup and decommissioning activities. The major processes of the Data Life Cycle are described briefly below for application in cleanup and decommissioning activities.

D.3.1 The Planning Phase As illustrated in Table D.1 and Figure D.2, planning documents could include Quality Assurance Project Plans (or QAPPs), Work Plans, Sampling and Analysis Plans, Data Validation Plans, and Data Quality Assessment Plans. Different organizations may use different terms for these documents but typically the set of documents include common elements. As provided in ANSI/ASQC E-4, QAPPs or other planning documents should detail the quality assurance (QA),

QC, and other technical requirements that must be implemented to ensure that the results of the work will meet stated performance criteria. MARLAP selected EPAs QAPP as a model for project plan documents because (i) it is closely associated with the DQO planning process, and (ii) widely accepted guidance on content (EPA, 2004; EPA, 2002). Chapter 4 and Appendix D of MARLAP contain additional information on the scope and content of planning documents.

D-2

MARLAP Appendix C contains information on QC acceptance criteria for comparisons of (1) onsite analytical results to offsite analytical results, (2) onsite split or duplicate samples to standard samples, and (3) replicate static measurements to standard static measurements.

Furthermore, Section C, Regulatory Position, (subsection 6, Quality Control in Radioanalytical Laboratory) of Regulatory Guide 4.15, Quality Assurance for Radiological Monitoring Programs (Inception through Normal Operations to License Termination)Effluent Streams and the Environment, cites MARLAP Chapter 18 as providing guidance on monitoring key laboratory performance indicators to determine whether a laboratorys measurement processes are in control. MARLAP Chapter 18 describes a numerical performance indicator for the analysis of a certified reference material in Equation 18.5. Also, the MARLAP Chapter 18 protocol could be used for split samples sent to two different laboratories.

The directed planning process for cleanup and decommissioning typically involves the following radioanalytical aspects:

  • Stating the cleanup problem: Identify the analytes of concern, matrix of concern, regulatory requirements, sampling constraints, primary decisionmakers, available resources, and existing data and their reliability.
  • Identifying the cleanup decision: Assess different analytical protocols, identify items of the analytical protocol specifications, and determine how sample collection will affect the measurement quality objectives (MQOs).
  • Identifying the inputs to the cleanup decisions: Define characteristics of the analytes and matrix, assess the concentration range for the analyte of interest, and define action levels.
  • Defining the decision boundaries: Identify background and temporal and spatial trends of data and determine limitations of current analytical protocols.
  • Developing a decision rule and tolerable decision error rates: For example, the decision rule may be defined as, If the mean concentration of analyte x in the upper 15 cm of the soil is greater than z Bq/g, then an action would be taken to remove the soil from the site. Estimates should be made of uncertainties in the data considering action levels and/or derived concentration guidelines.
  • Specifying limits on decision error rates: Evaluate the range of possible parameter values and the allowable difference between the action level and the actual value.
  • Optimizing the strategy for obtaining data: This process may involve optimization of the design for data collection through coordination with the different team members. The process also involves developing analytical protocols specifications and establishing performance measures of the MQOs.

D.3.2 The Implementation Phase The radioanalytical process is a compilation of activities, starting from the time a sample is collected and ending with the data reduction and reporting. Figure D.2 illustrates the typical components of an analytical process used for radiological characterization and survey of contaminated sites. Certain cleanup or decommissioning projects may not include all of the components listed in Figure D.2. The analytical protocols usually comprise a compilation of D-3

specific procedures or methods and are performed in succession, depending on the particular analytical process. Using a performance-based approach, a number of alternative protocols might be appropriate for a particular analytical process. A major component of the analytical protocol is the analytical method. The radioanalytical process should also include analytical uncertainty, analytical error, precision, bias, and accuracy of the method used.

D.3.3 The Assessment Phase The assessment phase focuses on three major steps:

(1) Data verification: This step ensures that the laboratory conditions and operations are in compliance with the statement of work and the projects QA project plan. The verification process would examine the laboratory standard operating procedures. It would also check for consistency and comparability of the data, correctness of the data calculations, and completeness of the results and data documentation.

(2) Data validation: This step addresses the reliability of the radioanalytical data. It addresses the analyte and matrix types, as well as the uncertainty of the measurement to support the intended use. Validation flags (qualifiers) are typically applied to data that do not meet the acceptance criteria established to meet the project DQOs and MQOs.

(3) Data quality assessment (DQA): This step represents the scientific and statistical data evaluation aspects to determine if data are of the right type, quality, and quantity to support the intended use. The DQA is more global in its purview, such that it considers the combined impacts of all project activities on data quality and its usability.

D.4 Benefits of Using MARLAP in Decommissioning and Cleanup Projects MARLAP is an extensive document that presents comprehensive guidance and information on the three phases of the radioanalytical Data Life Cycle. MARLAP emphasizes the importance of establishing the proper linkages among these phases. Use of MARLAP in decommissioning and cleanup projects can benefit the user in the following respects:

  • MARLAP ensures the generation of radioanalytical data of acceptable quality for the intended use.
  • MARLAP minimizes time and effort expended in generating unacceptable data.
  • MARLAP enhances public trust in the radioanalytical data generated by licensees and regulators.
  • MARLAP minimizes efforts applied to justifying data and may limit any litigation costs.
  • Because MARLAP uses an early coordinated approach to develop the radioanalytical data DQOs and MQOs, this approach would require early coordination and inputs from the decisionmakers, the project manager, shareholders, concerned team members, and the analyst (see Figure D.1). Therefore, this approach should resolve issues or difficulties related to sampling, sample tracking, sample preservation, analysis, data quality, time, and costs early in the process.

D-4

  • MARLAP provides flexibility in selecting the appropriate analytical method, using a performance-based approach that considers the DQOs, the MQOs, and the available resources.
  • MARLAP enhances regulatory reviews of radioanalytical data and saves time and effort for site characterization, environmental monitoring, decommissioning, and remediation.

Table D.1 The Radioanalytical Data Life Cycle PHASE PROCESS PROCESS OUTPUTS Directed Planning Process Development of DQOs and MQOs, including Optimized Sampling and Analytical Designs Plan Documents Project Plan Documents, including Quality PLANNING Assurance Project Plan (QAPP), Work Plan, or Sampling and Analysis Plan, Data Validation Plan, Data Quality Assessment Plan Contracting Services Statement of Work and Other Contractual Documents Sampling Laboratory Samples IMPLEMENTATION Analysis Laboratory Analysis, including QC Samples and Complete Data Package Verification Verified Data and Data Verification Report ASSESSMENT Validation Validated Data and Data Validation Report DQA Assessment Report Data of Known Quality Appropriate for the Intended Use D-5

Figure D.1 MARLAP Road MapKey Terms and Processes D-6

Sample Tracking Field Sample Preparation Quality Assurance Quality Control Sample Receipt and Inspection Laboratory Sample Preparation Sample Dissolution Chemical Separation of Radionuclides of Concern Preparation of Samples for Instrument Measurements Instrument Measurement Data Reduction and Reporting From MARLAP (NUREG-1576)

Figure D.2 Typical Components of the Radioanalytical Process D-7

D.5 References American Society for Quality Control (ANSI/ASQC) E-4. Specifications and Guidelines for Quality Systems for Environmental Data Collection and Environmental Technology Programs.

American Society for Quality Control: Milwaukee, Wisconsin. 1995.

American Society for Testing and Materials (ASTM). Standard Practice for Generation of Environmental Data Related to Waste Management Activities: Development of Data Quality Objectives, ASTM D5792. 1995.

. Standard Guide for Data Assessment for Environmental Waste Management Activities, ASTM D6233. 1998.

Environmental Protection Agency (EPA). Guidance for Data Quality Assessment: Practical Methods for Data Analysis, (EPA QA/G-9). EPA/600/R-96/084. EPA: Washington, DC.

2000a.

. Guidance for the Data Quality Objective Process (EPA QA/G-4),

EPA/600/R-96/055. EPA: Washington, DC. 2000b.

. Guidance for Quality Assurance Project Plans. EPA QA/G-5, EPA/240/R-02/009. EPA: Office of Environmental Information. Washington, DC. 2002.

. U.S. Department of Defense, U.S. Department of Energy, U.S. Department of Homeland Security, U.S. Nuclear Regulatory Commission, U.S. Food and Drug Administration, U.S. Geological Survey, and National Institute of Standards and Technology. Multi-Agency Radiological Laboratory Analytical Protocols (MARLAP), NUREG-1576, EPA402-B-04-001A, NTIS PB2004-105421, Vol. I, II, and III and Supp. 1. Washington, DC. 2004.

International Organization for Standardization (ISO)/International Electrotechnical Commission (IEC). International Vocabulary of Basic and General Terms in Metrology. ISO/IEC Guide 99:2007. ISO: Geneva, Switzerland. 2007.

. Guide to the Expression of Uncertainty in Measurement. ISO:

Geneva, Switzerland. 1995.

Nuclear Regulatory Commission (U.S.) (NRC). Multi-Agency Radiation Survey and Site Investigation Manual, (MARSSIM), NUREG-1575. NRC: Washington, DC. August 2000.

D-8

APPENDIX E MEASUREMENTS FOR FACILITY RADIATION SURVEYS

E.1 Introduction This appendix is applicable to all decommissioning groups. All surveys, whether simple or complex FSSs, require information on the reasons for instrument selection, the nature of the radionuclides, measurement techniques and procedures, MDCs of the instruments (measurement systems), and instrument calibration. Therefore, the information presented in this appendix would apply to a simple survey used to demonstrate compliance with regulatory decommissioning criteria, as well as a complex FSS.

This appendix contains limited, general information on survey techniques and survey measurements. The information presented here is related to the process of implementing a survey plan and refers to the appropriate sections of the MARSSIM (NRC, 2000)1, MARLAP, and various NUREGs for more detailed information. These are important areas for the conduct of surveys in the RSSI process and include the basic modes for determining levels of radiation and radioactivity at a site, instrument and scanning detection limits, instrument calibration, and laboratory measurements for samples. The data from the FSS are the deciding factor in judging if the site meets the release criteria.

Radiological conditions that should be determined for license termination purposes include any combination of total surface activities, removable surface activities, exposure rates, radionuclide concentrations in soil, or induced activity levels. To determine these conditions, field measurements and laboratory analyses may be necessary. For certain radionuclides or radionuclide mixtures, the licensee may have to measure both alpha and beta radiation. In addition to assessing the average radiological conditions, the licensee should identify small areas with elevated levels of residual radioactivity and determine their extent and activities.

There are three basic modes in which one can operate in determining the levels of radiation and radioactivity at a site. They are scanning with hand-held survey instruments, direct measurements with these same or larger instruments, and sample collection at the site followed by analysis in the laboratory. In many cases, the licensee will use some combination of these modes to obtain data, although the exact mix would be expected to vary according to the application.

In practice, the licensee uses the DQO process to obtain a proper balance among the uses of various measurement techniques. In general, there is an inverse correlation between the cost of a specific measurement technique and the detection levels being sought. Depending on the survey objectives, important considerations include survey costs and choosing the optimum instrumentation and measurement mix.

The decision to use a measurement method as part of the survey design is determined by the survey objectives and the survey unit classification. Scanning is performed to identify areas of elevated activity that other measurement methods may not detect. Direct measurements are analogous to collecting and analyzing samples to determine the average activity in a survey unit.

1 It is important to note that as of the date of publication of this draft volume, MARSSIM, Revision 1, is the current version of MARSSIM, while MARSSIM, Revision 2, is currently being drafted. Any changes to MARSSIM guidance in the future that affects the guidance in this volume will be reflected in future revisions to this volume. The decommissioning website should be consulted for issuance of interim guidance and lessons learned between guidance revisions.

E-1

E.2 Direct Measurements (Fixed Measurements)

To conduct direct measurements of alpha, beta, and photon surface activity, instruments and techniques providing the required detection sensitivity are selected. The selection of the type of instrument and method of performing the direct measurement depends on the type of residual radioactivity present, the measurement sensitivity requirements, and the objectives of the radiological survey.

Direct measurements may be collected at random locations in the survey unit. Alternatively, direct measurements may be collected at systematic locations and supplement scanning surveys to identify small areas of elevated activity. Direct measurements may also be collected at locations identified by scanning surveys as part of an investigation to determine the source of the elevated instrument response. Professional judgment may also be used to identify locations for direct measurements to further define the areal extent of residual radioactivity and to determine maximum radiation levels within an area, although these types of direct measurements are usually associated with preliminary surveys (i.e., scoping, characterization, remedial action support). Licensees should document all direct measurement locations and results.

If the equipment and methodology used for scanning are capable of providing data of the same quality required for direct measurement (e.g., detection limit, location of measurements, ability to record and document results), then scanning may be used in place of direct measurements.

Similarly, the usage of in situ measurement instrumentation may be possible if sufficient data quality can be achieved. In both cases, proposed approaches should be developed using the DQO process and should be communicated to NRC staff. Results should be documented for at least the number of locations required for the statistical tests. In addition, some direct measurement systems may be able to provide scanning data, provided they meet the objectives of the scanning survey.

Chapter 6 of MARSSIM includes information on radiation measurements. Specifically, Section 6.4.1 of MARSSIM contains information on direct measurements for alpha-, beta-, and gamma-emitting radionuclides.

E.3 Scanning Measurements Scanning is the process by which the operator uses portable radiation detection instruments to detect the presence of radionuclides on a specific surface (i.e., ground, wall, floor, equipment).

The term scanning survey describes the process of moving portable radiation detectors across a suspect surface with the intent of locating residual radioactivity. Investigation levels for scanning surveys are determined during survey planning to identify areas of elevated activity.

Scanning surveys are useful in locating radiation anomalies indicating residual gross activity that might require further investigation or action.

Areas of elevated activity typically represent a small portion of the site or survey unit. Thus, random or systematic direct measurements or sampling on the commonly used grid spacing may have a low probability of identifying these areas. Scanning surveys are often relatively quick and inexpensive to perform. For these reasons, the licensee typically performs them before direct measurements or sampling. This avoids spending time fully evaluating an area that may quickly prove to contain residual radioactivity above the IL during the scanning process. Based on the HSA, surfaces to be surveyed, and survey design objectives, licensees conduct scans that would indicate all radionuclides potentially present, using surrogate E-2

measurements where appropriate. Documenting scanning results and observations from the field is very important. For example, licensees should document a scan that identified relatively sharp increases in instrument response or identified the boundary of an area of increased instrument response. This information is useful when interpreting survey results.

Chapter 6 of MARSSIM includes information on radiation measurements. Specifically, Section 6.4.2 of MARSSIM contains information on scanning measurements for alpha-, beta-,

and gamma-emitting radionuclides.

E.4 Sampling For certain radionuclides that cannot be effectively measured directly in the field, the licensee should collect samples of the medium under investigation (e.g., soil) and then analyze them with a laboratory-based procedure. On the simplest level, this would include the analysis of a smear sample using a gross alpha-beta counter. More involved analyses would include gamma spectrometry, beta analysis using liquid scintillation counting, or alpha spectrometry following separation chemistry.

Samples from a variety of locations may be required, depending upon the specific facility conditions and the results of scans and direct measurements. Inaccessible surfaces cannot be adequately evaluated by direct measurements on external surfaces alone; therefore, those locations that could contain residual radioactive material should be accessed for surveying.

Residue can be collected from drains using a piece of wire or plumbers snake with a strip of cloth attached to the end; deposits on the pipe interior can be loosened by scraping with a hard-tipped tool that can be inserted into the drain opening. Particular attention is paid to low points or traps where activity would likely accumulate. The need for further internal monitoring and sampling is determined on the basis of residue samples and direct measurements at the inlet, outlet, cleanouts, and other access points to the pipe interior.

Residual activity will often accumulate in cracks and joints in the floor. These are sampled by scraping the crack or joint with a pointed tool such as a screwdriver or chisel. Samples of the residue can then be analyzed; positive results of such an analysis may indicate possible subfloor residual radioactivity. Checking for activity below the floor will require accessing a crawl space (if one is present) or removing a section of the flooring. Coring, using a commercially available unit, is a common approach to accessing the subfloor soil. After removing the core (where the diameter may range from a few centimeters to up to 20 centimeters), direct monitoring of the underlying surface can be performed and samples of soil collected.

Coring is also useful for collecting samples of construction material that may contain activity that has penetrated below the surface or activity induced by neutron activation. This type of sampling is also applicable to roofing material, which may contain embedded or entrapped contaminants. The profile of the distribution and the total radionuclide content can be determined by analyzing horizontal sections of the core.

If residual activity has been coated by paint or some other treatment, the underlying surface and the coating itself may contain residual radioactivity. If the activity is a pure alpha or low-energy beta emitter, measurements at the surface will probably not be representative of the actual residual activity level. In this case, the licensee can remove the surface layer from a known area, usually 100 cm2, using a commercial stripping agent or by physically abrading the surface.

The removed coating material is analyzed for activity content and the level converted to units of E-3

disintegrations per minute (dpm)/100 square centimeters (cm2) for comparison with guidelines for surface activity. The licensee takes direct measurements on the underlying surface, after removing the coating.

MARSSIM and MARLAP contain information on sampling and laboratory analysis for decommissioning. Chapter 10 of MARLAP discusses field and sampling issues that affect laboratory measurements.

E.5 Minimum Detectable Concentrations Detection limits for field survey instrumentation are important criteria in the selection of appropriate instrumentation and measurement procedures. For the most part, the licensee determines detection limits to evaluate whether a particular instrument and measurement procedure is capable of detecting residual activity at the regulatory release criteria. One may demonstrate compliance with decommissioning criteria by performing surface activity measurements and directly comparing the results to the surface activity DCGLs. However, before any measurements are performed, the survey instrument and measurement procedures to be used should be evaluated to ensure they possess sufficient detection capabilities relative to the surface activity DCGLs.

The measurement of residual radioactivity during surveys in support of decommissioning often involves measuring residual radioactivity at near-background levels. Thus, the licensee should determine the minimum amount of radioactivity that may be detected by a given survey instrument and measurement procedure. In general, the MDC is the minimum activity concentration on a surface or within a material volume that an instrument is expected to detect (i.e., activity expected to be detected with 95 percent confidence). It is important that this activity concentration, the MDC, is determined a priori (i.e., before survey measurements are conducted).

As generally defined, the detection limit, which may be a count or count rate, is independent of field conditions such as scabbled, wet, or dusty surfaces. That is, the detection limit is based on the number of counts and does not necessarily equate to measured activity under field conditions. These field conditions do, however, affect the instruments detection sensitivity or MDC. Therefore, the licensee should not use the terms MDC and detection limit interchangeably.

In MARSSIM, MARLAP, and other NRC NUREGs, the MDC corresponds to the smallest activity concentration measurement that is practically achievable with a given instrument and type of measurement procedure. That is, the MDC depends not only on the particular instrument characteristics (e.g., instrument efficiency, background, integration time) but also on the factors involved in the survey measurement process (U.S. Environmental Protection Agency, EPA 520/1-80-012, Upgrading Environmental Radiation Data, issued August 1980), which include surface type, source-to-detector geometry, and source efficiency (e.g., backscatter and self-absorption).

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MARLAP Section 3.3.7, Method Performance Characteristics and Measurement Quality Objectives, and Chapter 20, Detection and Quantification Capabilities, discuss MDCs.

E.6 Survey Minimum Detectable Concentrations During radiological surveys in support of decommissioning, scanning is performed to identify the presence of any locations of elevated direct radiation. The probability of detecting residual radioactivity in the field is affected not only by the sensitivity of the survey instrumentation when used in the scanning mode of operation but also by the surveyors ability. The surveyor will decide whether the signals represent only the background activity, or whether they represent residual radioactivity in excess of background.

The MDC of a scan survey, referred to as scan MDC or MDCscan, depends on the intrinsic characteristics of the detector (e.g., efficiency, window area), the nature (e.g., type and energy of emissions) and the relative distribution of the residual radioactivity (e.g., point versus distributed source and depth of residual radioactivity), scan rate, and other characteristics of the surveyor. Some factors that may affect the surveyors performance include the costs associated with various outcomese.g., cost of missed residual radioactivity versus cost of incorrectly identifying areas as containing residual radioactivityand the surveyors a priori expectation of the likelihood of residual radioactivity being present. For example, if the surveyor believes that the potential for residual radioactivity is very low, as in an unaffected area, then a relatively large signal may be needed for the surveyor to conclude that residual radioactivity is present. NUREG/CR-6364, Human Performance in Radiological Survey Scanning, issued March 1998, contains a complete discussion of the human factors as they relate to the performance of scan surveys.

Signal detection theory provides a framework for the task of deciding whether the audible output of the survey meter during scanning was due to background or signal plus background levels.

An index of sensitivity (d) that represents the distance between the means of the background and background plus signal, in units of their common standard deviation, can be calculated for various decision errorsType I error () and Type II error (). As an example, for a correct detection or true positive rate of 95 percent (1) and a false positive rate () of 5 percent, d is 3.29 (similar to the static MDC for the same decision error rates). The index of sensitivity is independent of human factors, and therefore, the ability of an ideal observer (i.e., theoretical construct) may be used to determine the minimum d that can be achieved for particular decision errors. The ideal observer makes optimal use of the available information to maximize the percent of correct responses and thus provides an effective upper bound against which to compare actual surveyors. Computer simulations and field experimentation can then be performed to evaluate the surveyor efficiency (p) relative to the ideal observer. The resulting expression for the ideal observers minimum detectable count rate (MDCR), in counts per minute (cpm), can be written:

MDCR = d / bi (60 / i ) = si (60 / i ) (E-1) where MDCR = minimum detectable (net) count rate in cpm, bi = background counts in the observation interval, si = minimum detectable number of net source counts in the observation interval, and i = observational interval (in seconds), based on the scan speed and areal extent of the residual radioactivity.

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Scan MDCs are determined from the MDCR by applying conversion factors to obtain results in terms of measurable surface activities and soil concentrations. As an example, the scan MDC for a structure surface can be expressed as:

MDCR Scan MDC = (E-2) probe area p i s 100 cm2 Chapter 6 of NUREG-1507, Revision 1, Minimum Detectable Concentrations with Typical Radiation Survey Instruments for Various Contaminants and Field Conditions, issued August 2020, contains a discussion of survey MDCs. This discussion includes scan MDC equations for both building/structure surfaces and land areas.

E.7 Survey Instrument Calibration Before determining the MDC for a particular instrument and survey procedure, it is necessary to introduce the expression for total alpha or beta surface activity per unit area. In the ISO Guide 7503-1, Evaluation of Surface Contamination, 1988, the ISO recommends calculating the total surface activity, AS, as in the following expression:

RS + B RB s = (E-3)

(i )(W )( S )

where RS+B = the gross count rate of the measurement in cpm, RB = the background count rate in cpm, i = the instrument or detector efficiency (unitless),

s = the efficiency of the residual radioactivity source (unitless), and W = the area of the detector window (cm2).

(For instances in which W does not equal 100 cm2, probe area corrections are necessary to convert the detector response to units of dpm per 100 cm2.)

This expression clearly distinguishes between instrument (detector) efficiency and source efficiency. The product of the instrument and source efficiency yields the total efficiency, tot.

Currently, surface residual radioactivity is assessed by converting the instrument response to surface activity using one overall total efficiency. This is not a problem, provided that the calibration source exhibits characteristics similar to the surface residual radioactivityincluding such characteristics as radiation energy, backscatter effects, source geometry, and self-absorption. In practice, this is hardly the case; more likely, total efficiencies are determined with a clean, stainless steel source, and then those efficiencies are used to measure residual radioactivity on a dust-covered concrete surface. By separating the efficiency into two components, the surveyor has a greater ability to consider the actual characteristics of the surface residual radioactivity.

The instrument efficiency is defined as the ratio between the net count rate of the instrument and the surface emission rate of a source for a specified geometry. The surface emission rate, q2, is defined as the number of particles of a given type above a given energy emerging from the front face of the source per unit time (ISO 7503-1). The surface emission rate is the 2 E-6

particle fluence that embodies both the absorption and scattering processes that affect the radiation emitted from the source. Thus, the instrument efficiency is determined by RS + B R S = (E-4)

(i )(W )( S )

The instrument efficiency is determined during calibration by obtaining a static count with the detector over a calibration source that has a traceable activity or surface emission rate or both.

In many cases, it is the source surface emission rate that is measured by the manufacturer and certified as National Institute of Standards and Technology traceable. The source activity is then calculated from the surface emission rate, based on the assumed backscatter and self-absorption properties of the source. The theoretical maximum value of instrument efficiency is one.

The source efficiency, s, is defined as the ratio between the number of particles of a given type emerging from the front face of a source and the number of particles of the same type created or released within the source per unit time (ISO 7503-1). The source (or surface) efficiency takes into account the increased particle emission due to backscatter effects, as well as the decreased particle emission due to self-absorption losses. For an ideal source (no backscatter or self-absorption), the value of s is 0.5. Many real sources will exhibit values of s less than 0.5, although values greater than 0.5 are possible, depending on the relative importance of the absorption and backscatter processes. Source efficiencies may either be determined experimentally or simply selected from the guidance contained in ISO 7503-1.

Some of the factors that affect the instrument efficiency, I, include detector size (probe surface area), window density thickness, geotropism, instrument response time, and ambient conditions such as temperature, pressure, and humidity. The instrument efficiency also depends on the radionuclide source used for calibration and the solid angle effects, which include source-to-detector distance and source geometry.

Some of the factors that affect the source efficiency, s, include the type of radiation and its energy, source uniformity, surface roughness and coverings, and surface composition (e.g., wood, metal, concrete).

The licensee assesses surface activity levels by converting detector response, through the use of a calibration factor, to radioactivity. Once the detector has been calibrated and an instrument efficiency (i) established, several factors still need to be carefully considered when using that instrument in the field. These factors involve the background count rate for the particular surface and the surface efficiency (s), which addresses the physical composition of the surface and any surface coatings. Ideally, the surveyor should use experimentally determined surface efficiencies for the anticipated field conditions. The surveyor needs to know how and to what degree these different field conditions can affect the sensitivity of the instrument. A particular field condition may significantly affect the usefulness of a particular instrument (e.g., wet surfaces for alpha measurements or scabbled surfaces for low-energy beta measurements).

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One of the more significant implicit assumptions commonly made during instrument calibration and subsequent use of the instrument in the field is that the composition and geometry of residual radioactivity in the field is the same as that of the calibration source. This may not be the case, considering that many calibration sources are fabricated from materials different from those that comprise the surfaces of interest in the field (e.g., activity plated on a metallic disc (Walker, Proper Selection and Application of Portable Survey Instruments for Unrestricted Release Surveys, issued 1994)). This difference usually manifests itself in the varying backscatter characteristics of the calibration and field surface materials.

Generally, it will not be necessary to recalculate the instrument MDC to adjust for the field conditions. The instrument detection limit (in net counts or net count rate) remains the same, but the surface activity MDC may be different (due to the varying s).

It is important to note that the preceding discussion on source efficiency and the calculation of surface activity are based on guidance in ISO 7503-1:19882. The ISO 7503 series was updated in 2016, prior to Revision 2 of this NUREG report. Therefore, a comparison of the 1988 and 2016 versions of ISO 7503 was performed to understand any differences in approach to surface or source efficiency, and to determine if the methodology/terminology in NUREG-1507, Revision 1, should be updated. A detailed discussion of this analysis is provided in Revision 1 of NUREG-1507, while a brief overview of the comparison is provided below.

Several of the basic concepts and methods were compared between the 1988 and 2016 ISO 7503 series. The concepts of source efficiency and an ideal source were compared, and it was concluded that they were presented as essentially the same (noting that a different P-Factor terminology was used for the 2016 series). The equations to evaluate contamination measurement data, and the associated usage of s and the P-Factor, were compared between the two series, and these equations were essentially the same (while different terminology has been used). A comparison was performed between recommended default or conservative s values and P-Factors in the two series, and these values utilized the same assumptions and made the same recommendations for default source efficiencies (again noting that different terminology was presented). Based upon the ISO 7503 series comparison, it has been concluded that the 2016 revision of ISO 7503-1 presented no compelling reason to update the usage of the surface efficiency concept and terminology for Revision 1 of NUREG-1507.

Therefore, the equations for MDC and surface activity measurement, which were originally developed using ISO 7503-1:1988 remain valid and are not changed from the previous NUREG-1757 version.

However, the comparison between the 1988 and 2016 ISO 7503 series identified that there is a need to consider weighted detection efficiencies for use with multiple radionuclides or with complex decay series, as ISO 7503-3:2016, in particular, presented many new concepts in this area. To address this need, Revision 1 to NUREG-1507 will include weighted efficiency calculations that utilize the concepts of instrument efficiency, source efficiency, and emission intensity, while also considering the relative fraction of radionuclides and branching ratios.

Chapter 4 of NUREG-1507 covers survey instrument calibration and the effects of efficiency changes on MDC. Chapter 5 of NUREG-1507 discusses variables affecting efficiencies in the 2 MARSSIM, Revision 1, and NUREG-1507, Revision 1, reference ISO 7503-1:1988. The NRC has found this standard acceptable for use by NRC licensees to calculate MDCs and surface activity measurements.

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field. Chapter 20 of MARLAP discusses instrument efficiency and the minimum detectable net instrument signal.

E.8 Laboratory Measurements Frequently during surveys in support of decommissioning, it is not feasible, or even possible, to detect the residual radioactivity with portable field instrumentation; thus arises the need for a laboratory analysis of media samples. This is especially the case for such media samples as soil, which result in significant self-absorption of the radiation from the residual radioactivity.

Another common situation that necessitates the use of laboratory analyses occurs when the residual radioactivity is difficult to detect even under ideal conditions. This includes residual radioactivity that emits only low-energy beta radiation (e.g., H-3 and Ni-63) or x-ray radiation (e.g., Fe-55). Laboratory analyses for radionuclide identification, using spectrometric techniques, are often performed during scoping or characterization surveys. Here, the principal objective is to simply determine the specific radionuclides present in the residual radioactivity, without necessarily having to assess the quantity of residual radioactivity. Once the licensee identifies the residual radioactivity, it selects sufficiently sensitive field survey instrumentation and techniques to demonstrate compliance with the DCGLs.

Samples collected during surveys for decommissioning purposes should be analyzed by trained individuals using the appropriate equipment and procedures at a well-established laboratory, which uses either in-house or contractor laboratory services. There should be written procedures that document both (1) the laboratorys analytical capabilities for the radionuclides of interest and (2) the QA/QC program that ensures the validity of the analytical results. Many of the general types of radiation detection measuring equipment used for survey field applications are also used for laboratory analyses, usually under more controlled conditions that provide for lower detection limits and greater delineation between radionuclides. Laboratory methods often also involve a combination of both chemical and instrumental techniques to quantify the low levels expected to be present in samples from decommissioning facilities.

To reemphasize, a thorough knowledge of the radionuclides present, along with their chemical and physical forms and their relative abundance, is needed to select appropriate laboratory methods. With this information, it may be possible to substitute certain gross (i.e., nonradionuclide-specific) measurement techniques for the more costly and time-consuming wet chemistry separation procedures and relate the gross data back to the relative quantities of specific contaminants. The individual responsible for the survey should be aware that radiochemical analyses require lead times that will vary according to the nature and complexity of the request. For example, a laboratory could provide a fairly quick turnaround on gamma spectrometry because sample preparation is likely limited to the media being dried and homogenized before being measured in a standard geometry. In comparison, alpha spectrometry usually involves sample preparation that also includes chemical separation and will typically require a longer lead time. Some factors influencing the analysis time include (1) the nuclides of concern, (2) the type of samples to be analyzed, (3) the QA/QC considerations required, (4) the availability of adequate equipment and personnel, and (5) the required detection limits.

For relatively simple analyses, such as gross alpha and gross beta counting of smears and water samples, liquid scintillation spectrometry for low-energy beta emitters in smear and water samples, and gamma-spectrometry of soil, it is usually practical to establish in-house laboratory capabilities. The more complicated and labor-intensive procedures, such as alpha E-9

spectrometry, Sr-90 and low-energy beta emitters (e.g., H-3, Ni-63) in soil samples, should be considered candidates for contract laboratory analyses.

Analytical methods should be capable of measuring levels below the established release guidelines; detection sensitivities of 10 to 25 percent of the guideline should be the target.

Although laboratories will state detection limits, these limits are usually based on ideal situations and may not be achievable under actual measurement conditions. Also, remember that detection limits are subject to variation from sample to sample, instrument to instrument, and procedure to procedure, depending upon sample size, geometry, background, instrument efficiency, chemical recovery, abundance of the radiations being measured, counting time, self-absorption in the prepared sample, and interference from other radionuclides present.

MARSSIM and MARLAP contain information on sampling and laboratory analysis for decommissioning. MARLAP Sections 12, 13, 14, and 15 discuss laboratory sample preparation, sample dissolution, separation techniques, and quantification of radionuclides.

E.9 References Nuclear Regulatory Commission (U.S.) (NRC). Human Performance in Radiological Survey Scanning, NUREG/CR-6364. NRC: Washington, DC. 1997.

. Minimum Detectable Concentrations with Typical Radiation Survey Instruments for Various Contaminants and Field Conditions, NUREG-1507, Revision 1. NRC:

Washington, DC. 2020.

. Multi-Agency Radiation Survey and Site Investigation Manual, (MARSSIM, Rev. 1), NUREG-1575. NRC: Washington, DC. 2000.

Environmental Protection Agency (U.S.) (EPA). Upgrading Environmental Radiation Data, EPA 520/1-80-012. EPA: Washington, DC. 1980.

International Organization for Standardization. Evaluation of Surface ContaminationPart 1:

Beta Emitters and Alpha Emitters (first edition). ISO 7503-1. ISO: Geneva, Switzerland.

1988.

Walker, E. Proper Selection and Application of Portable Survey Instruments for Unrestricted Release Surveys. Bechtel Environmental, Inc. Presented at 1994 International Symposium on D&D. April 24-29, 1994.

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APPENDIX F SURFACE WATER AND GROUNDWATER CHARACTERIZATION

F.1 Introduction This appendix includes guidance on surface and groundwater characterization to support development of conceptual site models, development of hydrologic inputs to dose assessment models, and radiological surveys to estimate dose associated with residual radioactivity in the groundwater. Surface water and groundwater characterization is important for Decommissioning Group 4 sites with residual radioactivity in the surface water and Decommissioning Groups 5-7 that have existing residual radioactivity in the groundwater.1 Characterization is also important for sites that have a medium to high potential for residual radioactivity in the groundwater (see Section F.4) based on historical site activities even in the absence of evidence of residual radioactivity in the groundwater.

If the groundwater contains radioactive constituents, characterization of groundwater is an essential component of the dose modeling used in the estimation of doses to demonstrate compliance with the license termination requirements in Title 10 of the Code of Federal Regulations (10 CFR) Part 20, Standards for Protection against Radiation, Subpart E, Radiological Criteria for License Termination. In these cases, unmodified screening derived concentration guideline levels (DCGLs) for soil are inappropriate to use, because the screening levels assume surface water and groundwater are initially uncontaminated. Likewise, if site-specific dose modeling is used to demonstrate compliance with radiological criteria for license termination, the dose contributions of residual radioactivity in the surface water or groundwater should be taken into consideration in the dose assessment.

This appendix references reports, methods and software that should assist in characterization of surface water and groundwater and associated dose modeling. However, it should be noted that other reports, methods, and software may also be appropriate. If other approaches are utilized, it may be prudent for the licensee to contact NRC to ensure that the approaches are appropriate prior to any significant resource input by the licensee.

F.2 Planning for Surface Water and Groundwater Characterization The licensee should plan surface water and groundwater characterization in a manner that maximizes the utility of the information to be collected during the various stages of radiological surveys. This planning should take place early in the decommissioning process to facilitate development of the license termination plan (LTP). For example, for a particular site, a licensee may show that the surface water pathway is not likely to be significant in terms of existing and potential future exposure to the public. In such a case, the need for detailed characterization of the surface water system decreases. On the other hand, the identification of residual radioactivity in the groundwater during the preliminary scoping survey may warrant installing and sampling additional monitoring wells to define the nature and extent of residual radioactivity in the groundwater. For sites with groundwater contamination identified prior to beginning decommissioning, existing groundwater monitoring programs may need to be modified to better accomplish the objectives for decommissioning. For decommissioning, the objectives are focused on characterizing the extent and magnitude of residual radioactivity in the groundwater to provide either the (1) input to dose modeling for the development of DCGLs in the LTPs, or (2) information the licensee needs to determine whether remediation may be needed to demonstrate compliance with the decommissioning requirements. An alternative to modifying 1 For the purposes of this volume, groundwater refers to water below the land surface in a zone of saturation, which can theoretically be used for drinking water and irrigation although arguments can be presented to eliminate these pathways of exposure (e.g., insufficient yield, water quality).

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the groundwater monitoring program to meet the objectives of decommissioning includes the selection and justification of conservative estimates of residual radioactivity across the site (e.g., modeling).

In some instances, groundwater may be unsuitable for specific uses, such as human and livestock consumption, but may be acceptable for crop irrigation. In addition, some aquifers may not have the yield to support crop irrigation but may produce enough water for human consumption. In some instances, the EPA or a State agency may have declared that the aquifer in question is unfit for human or livestock use. Accordingly, the licensee should address this type of information, because it will support site exposure scenario development and dose modeling. The State agency may also have rules that apply to groundwater resource classifications. Section I.3.3.3 from Appendix I and Appendix M to this volume provides guidance on modification of waterborne exposure pathways.

F.3 Development of CSMs and Mathematical Models A conceptual site model (CSM) provides a hypothetical framework for contaminant source, geologic, hydrologic (including water usage), chemical, biologic, and demographic characteristics for the site (NRC, 2007). The CSM provides the basis for understanding flow and transport at the site for abstraction into a dose assessment model and is the starting point for numerical models if contamination is present in the surface water and groundwater. In general, CSMs should be updated as new information becomes available. The complexity of CSMs should be commensurate with site risk and at an appropriate level to demonstrate radiological criteria for license termination can be met. ASTM E1689-95 (2014) provides information on the development of CSMs that may be useful to licensees.

The model abstraction process involves the representation of major components of a complex system in a conceptual model. Conceptual models should represent or be able to describe the important features and processes of the system. A simplified representation of the conceptual model is developed so that the conceptual model can be more easily represented by a mathematical model. Mathematical models translate the assumptions of a conceptual model into the formalism of mathematics (IAEA, 2004). Mathematical models are represented by equations and, if these equations are solved, the output can include information such as radionuclide concentrations in media, doses to humans, and temporal evolution of the system.

Numerical codes can solve such equations, and constructed numerical models are used to solve the equations represented in the mathematical models, thereby simulating the properties of features and processes represented in a conceptual model. Model simplification is the process for reducing the complexity of a complex numerical model into a simpler numerical model while still maintaining the validity of the simulation results and is another form of model abstraction.

Compared to model simplification (i.e., reducing a complex numerical model to a less complex numerical model), abstracting the physical reality that is represented in a conceptual model to a mathematical model is typically more vulnerable to error. In addition, it can be difficult to validate how successfully the physical system is abstracted into a mathematical model whereas it is usually easier to make sure that a simplified numerical model is adequately representing a more complex numerical model. However, unlike model simplification, model abstraction is a required and necessary step in the performance assessment methodology. Not all features and processes can or should be included in a mathematical model. As an analogy, news reporters must abstract facts from a real set of complex circumstances or else the news story becomes too lengthy to be useful for most readers or viewers. Not only is model abstraction necessary, it F-2

can be useful to an analyst in explaining complex system behavior by reducing the system to its major components and thereby improve analyst communications with stakeholders regarding the system being simulated. Additional information on model abstraction and simplification can be found in NUREG-7026 and NUREG-6884.

Model abstraction builds on insights gained from characterizing the system and developing exposure scenarios. In a dose assessment, several model abstractions typically support an overall assessment of a facilitys ability to demonstrate compliance with release criteria. These abstractions usually include models of projected climate and infiltration, source term release, transport through environmental media including the groundwater, and potential exposures to a receptor in the biosphere. If surface water or groundwater are viable pathways, a CSM of the hydrologic system is needed to adequately model these systems. Thus, adequate characterization of the subsurface is necessary to construct defensible hydrogeological conceptual models to perform dose modeling to estimate future impacts to surface water and groundwater, as well as to estimate doses from existing residual radioactivity in the groundwater. Characterization of the subsurface should also include descriptions of construction activities and buried infrastructure at the site that may have breached confining layers or otherwise modified the natural flow system, thereby modifying transport pathways including creation of pathways between the uppermost aquifer and deeper aquifers. Examples of infrastructure that may affect the flow system include building foundations, buried pipes and tunnels, cofferdams, sheet piling, dewatering systems, and any fill material associated with or surrounding the infrastructure.

Although monitoring data can be useful in determining whether groundwater is contaminated and to determine the potential risk associated with groundwater dependent pathways, deficiencies in the monitoring well network and monitoring approach may compromise the ability to use existing monitoring well data for this purpose. In these cases, the licensee can manage uncertainty in exposure point concentrations through collection of additional data, and through additional modeling, including uncertainty/sensitivity analysis. Licensees should discuss acceptable approaches to assessing the risk and dose contributions associated with surface water and groundwater dependent pathways.

Environmental impact assessments for decommissioning may also benefit from the development of a CSM. The information needs of environmental impact assessments may differ slightly from a dose assessment for a safety case as non-radiological and other impacts associated with surface water and groundwater use must also be considered. Although specific to information needs to support preparation of environmental impact assessments, Regulatory Guide 4.2, Section 2.2, contains useful information regarding the collection of hydrological data for the purpose of environmental assessment, but may also be useful for creation of CSMs and development of dose assessment models (NRC, 2018).

It is also important to note that transients and spatial and temporal variability may need to be considered when simulating contaminant flow and transport if important to decision-making. For example, river stage may control groundwater flow direction under a site with changes in flow direction occurring over smaller timeframes, but net flow may be towards the river considering longer timeframes. A determination will need to be made if representation of short-term behavior is important to demonstrating compliance or will help provide a better understanding of system behavior and potential uncertainties in assessment of risk.

A critical aspect in the development of CSMs, numerical models, and abstractions is the need to make it an iterative process as information is obtained from characterization and evolving F-3

monitoring programs. Additionally, information on uncertainty in model abstraction and alternative conceptual models is provided in Appendix Q.

F.4 Indicators for Potential Residual Radioactivity in the Groundwater The level of effort associated with subsurface characterizations is dependent in large part on the extent of residual radioactivity in the subsurface soils or bedrock and the transport capability of that residual radioactivity in the groundwater system. Section G.3.1, of Appendix G, Subsurface Residual Radioactivity Surveys, discusses the development of a contamination of concern map, which is a useful tool for updating conceptual site models and guiding groundwater characterization.

Analysis of Historical Site Surveys and site infrastructure may provide indicators of potential residual radioactivity in the groundwater. As described in Table 1.1 of Volume 1 of this NUREG report, Decommissioning Groups 5-7 are sites that have the potential for residual radioactivity in groundwater. Based on the experience gained from operational and decommissioning NRC licensed sites, the following is a list of potential indicators for residual radioactivity in the groundwater at decommissioning sites (NUREG-1496, Appendix C, Attachment E, Table C.E.1).2 They are illustrative only and are not intended to constitute a complete list:

  • High Potential: if a site has a history of, or currently has the following:

o unlined lagoons, pits, canals, or surface-drainage ways that received radioactively contaminated liquid effluent o lined lagoons, pits, canals, or surface drainage ways that received radioactively contaminated liquid effluent, where the lining has leaked or ruptured, or where overflow has occurred o septic systems, dry wells, or injection wells that received radioactively contaminated liquid effluent o storage tanks, waste tanks, and/or piping (above or below ground) that held or transported radioactively contaminated fluids and are known or suspected to have leaked o liquid or wet radioactive waste buried on site (i.e., burial under previous regulations found in 10 CFR 20.302 Method for Obtaining Approval for Proposed Disposal Procedures, or 10 CFR 20.304 Disposal by Burial in Soil, (or the current 10 CFR 20.2002, Method for Obtaining Approval of Proposed Disposal Procedures))

o an accident or spill on site where radioactive material was released exterior to a building o wet bulk waste (e.g., sludge or tailings) stored exterior to buildings or used as backfill 2 MARSSIM, Rev. 1 (Sections 3.6.3.4 and 5.3.3.3) also provides guidance on evaluating the likelihood for release of residual radioactivity to groundwater, as well as characterization and sampling of groundwater (NRC, 2000).

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o containerized-liquid waste, stored exterior to buildings, that has leaked

  • Medium Potential: if a site has a history of or currently has the following:

o surface water or atmospheric discharge of radioactive effluents including authorized releases and spills (e.g., releases in compliance with 10 CFR Part 20, Appendix B effluent concentrations or spills) o radioactive contamination detected on the roof of a building o radioactive contamination detected in the floor cracks or sump of a building o an accident or spill on site, where liquid radioactive material was released to the interior of a building o the presence of aging unmonitored underground storage tanks or underground piping that held radioactively contaminated fluids, and are not known to have leaked (e.g., unmonitored tanks or piping which could have released radioactivity due to corrosive environmental or service conditions, use of material types potentially susceptible to corrosion [carbon steel], or design flaws) o a history of incineration of radioactive waste exterior to onsite buildings o dry bulk waste (e.g., sludge or tailings) stored exterior to buildings or used as backfill o solid containerized waste, stored exterior to buildings, that has leaked

  • Low Potential: if a site has a history of or currently has the following:

o underground storage tanks or underground piping early in their service life that held radioactively contaminated fluids and are not known to have leaked o dry bulk waste stored inside the buildings o a sealed-source-only license The potential for residual radioactivity in the groundwater at any of these sites is conditioned by certain site characteristics, such as depth of groundwater, amount of yearly precipitation, and hydraulic conductivity, and by certain source characteristics such as half-life, solubility, and distribution coefficient.

F.5 Groundwater Characterization The level of groundwater characterization should be commensurate with site complexity and potential for subsurface contamination. If existing groundwater contamination is known to be present, or if it is determined that residual radioactivity has a medium to high likelihood of reaching groundwater, radiological surveys of groundwater are expected. Approaches for characterizing residual radioactivity in subsurface soils and bedrock, such as development of a contaminants of concern map, is discussed further in Appendix G, Section G.3.1.

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Characterization of groundwater contamination, including all significant radiological constituents, along with inorganic and organic constituents and related parameters, should be adequate to determine the following:

  • extent and concentration distribution of contaminants
  • source (known or postulated) of radioactive contaminants to groundwater
  • background groundwater quality
  • rate(s) and direction(s) of contaminated groundwater migration
  • location of groundwater plume and concentration profiles (i.e., maximum concentration in the vertical and lateral extent)
  • assessment of present and potential future effects of groundwater withdrawal on the migration of groundwater contaminants
  • potential safety and environmental issues associated with remediating the surface water and groundwater
  • chemical form/speciation of the radionuclides; effect of the non-radiological constituents on the mobility of the radionuclides
  • whether the remediation activities and radiation control measures proposed by the licensee are appropriate for the type and amount of radioactive material present in the surface water and groundwater
  • whether the licensees waste management practices are appropriate
  • whether the licensees cost estimates are plausible Besides licensee process discharges, other mechanisms may affect groundwater. For example, sumps that capture infiltrating groundwater may affect the local groundwater elevation during pumping. In some situations, sumps collect groundwater at the lowest elevation of a building, with pumping going on continuously. Such pumping has been shown to affect the local groundwater elevation (i.e., cone of depression).

Typical analytical parameters include gross alpha particle activity, gross beta particle activity, specific radionuclide concentrations, gamma spectrum analysis for all gamma-emitting radionuclides suspected to be present, sulfate, chloride, carbonate, alkalinity, nitrate, total dissolved solids, total organic carbon, Eh, pH, calcium, sodium, potassium, iron, and dissolved oxygen. Additional analytical parameters may be necessary to characterize any suspected contamination. Other regulatory agencies that have jurisdiction over the decommissioning effort may require characterization of the non-radiological constituents and related parameters including or in addition to those listed above. Therefore, licensees should contact Federal, State, or local government bodies responsible for regulating groundwater.

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The licensee should determine the extent of contamination and background groundwater quality based on groundwater monitoring data from a suitable monitoring well network. The reference section lists the following guidance documents on acceptable groundwater monitoring techniques: Korte and Ealey (1983), Korte and Kearl (1984), NUREG-6948, NUREG/CR-7221, and NUREG/CR-1388, USGS (1977, 1996 and 2018), EPA (1977, 1980, 1985, 1986, and 2009a), and ASTM D5092/D5092M (2016a). The actual number, location, and design of monitoring wells depend on the size of the contaminated area, the type and extent of contaminants, the background groundwater quality, the hydrogeologic system, and the objectives of the monitoring program. For example, if the objective of monitoring is only to indicate the presence of residual radioactivity in the groundwater, the licensee will need relatively few downgradient and upgradient monitoring wells. In contrast, if the objective is to develop a detailed characterization of the distribution of constituents within a complex aquifer as the design basis for a corrective action program, a large number of suitably designed and installed monitoring wells and well points may be necessary.

Planned site characterization activities should be flexible enough to allow for the installation of additional monitoring wells during the characterization effort if either (1) preliminary characterization indicates contamination where previously unanticipated, or (2) there is a need to delineate the vertical or lateral extent of contaminant plumes.

Monitoring well locations, contaminant concentrations, and contaminant sources should be plotted on a map (or a series of maps for multiple contaminants) to show the relationship among contamination, sources, hydrogeologic features and boundary conditions, and property boundaries. At sites with significant vertical migration of contaminants, the Decommissioning Plan (DP) should also provide hydrogeologic cross sections that depict the vertical distribution of contaminants in groundwater. The extent of likely breaches in confining layers, such as those caused by construction activities and buried infrastructure, that could provide preferred pathways for the vertical migration of contaminants should also be depicted in the cross-sections. At sites with groundwater remediation, such as pump and treat programs, pre-and post-treatment sampling plans and transport pathways (plan view and cross-section, as appropriate) should be provided. The vertical exaggeration of the cross-sections should not exceed a factor of 10.

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generic completion design.

GAS VENT TUBE WELL CAP 1/4" GAS VENT STEEL PROTECTOR CAP WITH LOCKS SURVEYOR'S PIN (FLUSH MOUNT)

CONCRETE WELL APRON (MINIMUM RADIUS OF 3 FEET AND 4 INCHES THICK)

CONTINUOUS POUR CONCRETE CAP AND FROST WELL APRON (EXPANDING CEMENT)

ZONE CEMENT AND SODIUM BENTONITE MIXTURE WELL DIAMETER = 4" VADOSE ZONE BOREHOLE DIAMETER = 10" TO 12" (NOMINAL DIMENSION)

ANNULAR SEALANT FILTER PACK (2 FEET OR LESS ABOVE SCREEN)

--- POTENTIOMETRIC SURFACE SATURATED ZONE SCREENED INTERVAL SUMP/SEDIMENT TRAP BOTTOM CAP ZONE OF LESSER PERMEABILITY decom-016.ppt 052902 Figure F.1 General Monitoring Well Cross Section (Adapted from Figure 3 [NRC, 1994])

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The DP should also describe the groundwater characterization program used to characterize the extent and distribution of contaminants in the groundwater. Depending on the complexity of the site, the DP can include the detailed information as described below or can summarize this information and then reference the documents containing the supporting details. The description should provide monitoring well completion diagrams explaining elevation, internal and external dimensions, type of casings, type of backfill and seal, type of the screen and its location and size, borehole diameter and elevation and depth of hole, and type and dimension of riser pipe and other necessary information on the wells. Figure F.1 illustrates an acceptable generic design for well completion. The reference section lists the following documents that may contain useful information on monitoring well installation for specific objectives and geohydrologic conditions: ASTM (2016a, 2016b, and 2018a), EPA (1991a, and 2013), and USGS (1996). ASTM (2018b) may provide useful information on how to remove monitoring or remediation wells from service without creating a conduit for contaminant migration from the surface or between subsurface geologic units.

The DP should document or reference the sampling techniques, methodology, and procedures.

Site characterization procedures and methods should generally adhere to national practices and standards (e.g., American Society for Testing and Materials (ASTM), the USGS, EPA, U.S. Department of Energy Environmental Monitoring Laboratory (DOE/EML), and National Institute of Standards and Technology (NIST)). The DP should identify specific analytical methods that conform to generally accepted protocols and methods, such as those endorsed by NIST and DOE/EML, or other methods established through a comprehensive peer review and recommendation process (e.g., American National Standards Institute (ANSI)/American Society of Mechanical Engineers (ASME), Quality Assurance Program Requirements for Nuclear Facilities, 1986). Korte and Kearl, Procedures for the Collection and Preservation of Groundwater and Surface Water Samples and for the Installation of Monitoring Wells, 1984, provides forms for documenting well summary information, samples, chain of custody, QA information for field chemical analyses, and sample location and identifier.

The site characterization program should include sufficient sampling and analysis of groundwater samples collected upgradient from the site to develop a representative characterization of background groundwater quality. Background groundwater quality should not exhibit any influence from contaminants released by the site and should be representative of the quality of groundwater that would exist if the site had not been contaminated. The site characterization should also assess any temporal or spatial variations in background groundwater quality. If sources of contamination other than the site are present, the licensee should evaluate the potential impact of such sources to determine the degree of their groundwater contamination.

F.6 Monitoring Practices and Procedures Depending on the complexity of the site, the DP can include the detailed information as described below, or the DP can summarize this information and then reference the documents containing the supporting details.

The site characterization should include a description of all surface water and groundwater characterization activities, methods, and monitoring installations sufficient to demonstrate that the methods and devices provided data that are representative of site conditions. It should also describe the monitoring practices, procedures, and QA programs used to collect water quality data. Monitoring well descriptions, for example, should include location, elevation, screened interval(s), depth, construction and completion details, and the hydrologic units monitored.

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Aquifer test descriptions should include testing configuration, test results, and a discussion of the assumptions, analytical techniques, test procedures, pretesting baseline conditions, limitations, errors in measurements, and final results. The description of the water quality sampling and analysis program should include or reference the procedures for sampling, preserving, storing, and analyzing the samples, including QA/QC protocols implemented. All methods should be consistent with current standard methods and practices (e.g., ASTM, USGS, EPA, NIST, and ANSI/ASME). The reference section lists the following documents containing additional guidance on applicable methods for sampling and analyzing water quality samples:

Korte and Ealey (1983); Korte and Kearl (1984); DOE (1988 and 1993); ANSI/ASME (1986);

EPA (1977, 1985, 1986, 1987a, 1991b); and NUREG-1293, NUREG-1383, and Regulatory Guide 4.15 (NRC, 1979). The licensee should document and explain any deviations from the standard methods used.

F.7 Monitoring Network and Sampling Frequencies The monitoring wells are usually placed in the following critical locations (NRC, 2007):

(1) Source areas, (2) Within and immediately down-gradient of the source area; (3) Fringe portions and boundary of the plume; (4) Regulatory compliance boundaries.

To address the effect of physical/chemical heterogeneity on contaminant fate and transport, certain monitoring wells should be placed in relatively high transmissive zones with highest contaminant concentrations in the targeted monitoring areas; monitoring wells may need to be placed in locations that will reduce uncertainties with respect to geochemical conditions, and aquifer hydraulic properties. NUREG/CR-6948, NUREG/CR-7221, EPA (2009a), and Barcelona et al. (1989) provide information on how to design and implement a more optimal groundwater monitoring well network to refine site conceptual and numerical models and improve the capabilities for detection of contamination.

It is important to note that environmental monitoring of groundwater during the period of reactor operation, although adequate for its intended purpose, may not be adequate for site characterization and to support dose assessments for decommissioning. To support site characterization and dose assessments, information supplied by licensees may need to address the types and movement of radioactive contamination in groundwater at the facility, as well as the magnitude and extent of this contamination. The actual number, location, and design of monitoring wells depend on the size of the contaminated area, the nature and extent of contamination, the characteristics relating to background water quality, the hydrogeologic system, and the objectives of the monitoring program. For example, if the only objective of monitoring is to indicate the presence of residual radioactivity in the groundwater, relatively few downgradient and upgradient monitoring wells are needed. In contrast, if the objective is to develop a detailed characterization of the distribution of constituents within a complex aquifer as the design basis for a corrective action program, a relatively large number of suitably designed and installed monitoring wells may be necessary. Power reactors normally have groundwater monitoring programs as part of their radiological environmental monitoring programs (REMPs).

Although data derived from a REMP may provide useful information, the data still tend to be insufficient to allow the staff to fully understand the fate and transport of radioactive materials in F-10

the subsurface environment at the facility, as well as the magnitude and extent of this contamination (see for example, Regulatory Guide 1.185, Post-Shutdown Decommissioning Activities Report, lessons learned on groundwater monitoring). Therefore, a licensee may need to gather additional data to address this lack of understanding.

If remediation is necessary, remedial performance confirmation monitoring may be conducted to demonstrate that remediation is occurring according to expectations. Although remedial expectations and, consequently, appropriate performance monitoring analyses are site-specific in nature, reduction in contaminant concentrations to specified levels is generally expected as selected remedies are being performed (Pope et al., 2004). Data analysis is useful in assessing progress toward contaminant reduction objectives include evaluation of temporal trends in contaminant concentrations or mass, comparisons of observed contaminant distributions with predictions or predefined targets, among other approaches. Evaluations of adequate progress toward restoration objectives are sometimes difficult due, in large measure, to subsurface spatial variability and to a lesser extent, measurement variability. This will often necessitate optimization of the employed remedial remedy, with relatively dense monitoring networks to reduce uncertainty to acceptable levels.

The remedial action objective of attaining specified criteria (such as groundwater protection standards or DCGLs) should be identified before monitoring and the license are terminated to ensure that the required standards are actually achieved in the long term (Pope et al., 2004).

The demonstration of the attainment of cleanup objectives should include sufficient verification monitoring once the standards are achieved to evaluate the effects of variations caused by the active remedy. The length of the verification period should be based on site-specific conditions and on objective statistical analyses of the data. Site-specific conditions to consider include the response times of the hydrogeological system to seasonal or annual variations. Statistical methods useful in these evaluations include analyses of temporal trends in contaminant concentrations and comparisons with the specified concentration standards (e.g., Cohen et al.,

1994; EPA, 1992; and ITRC, 2013).

The licensee should establish concentrations of surface water and groundwater constituents and water levels on a set frequency, based on site-specific considerations. For sites with extensive residual radioactivity in the groundwater, the licensee should design and install a network of monitoring wells to provide a high probability of detecting and characterizing existing contamination and determining concentrations of background groundwater constituents.

Whereas the uppermost aquifer is generally the prime focus, at some sites it may be necessary to determine both the vertical gradient between the aquifers and the horizontal hydraulic gradient in aquifers below the uppermost aquifer. Flow between aquifers may be due to natural hydrogeological features or to construction activities and buried infrastructure breaching confining layers and providing direct pathways for residual radioactivity to travel to deeper aquifers. For temporal aspects of hydrologic conditions, measurements of groundwater levels in piezometers and monitoring wells should provide a sufficiently accurate characterization of changes in hydraulic head to determine the horizontal hydraulic gradient within the uppermost aquifer and vertical hydraulic gradient with adjacent units. It should measure water levels at least on a quarterly basis for a minimum of 1 year to determine temporal variations in the hydraulic conditions. After this period, it should adjust the frequency of water level measurements to reflect the anticipated impact on hydraulic heads by site-specific events and conditions (e.g., tides, rises in river stage and bank storage, increased precipitation, water year variations). The initial monitoring program should be designed so that it can be integrated into the programs for operational and post-operational periods. The reference section lists the following documents containing applicable methods for groundwater sampling and for F-11

measuring water levels: EPA (1977, 1985, 1986, and 1987a), USGS (1977, 2018), and Korte and Kearl (1984), and NUREG-1388.

The licensee should determine the sampling frequency for evaluating spatial and temporal variations in groundwater quality, including radiological and associated non-radiological constituents, based on the site-specific temporal variations in flow directions and hydrogeochemical conditions. After an initial sampling period in which each monitoring well is sampled at a frequency to establish site-dependent temporal variations throughout a year, it should collect and analyze representative samples generally on a quarterly basis from key monitoring wells to obtain representative estimates of the temporal variation of water quality in the uppermost aquifer and adjacent units. After this initial period, it should adjust the sampling frequency to reflect variations in the hydraulic gradient and hydrochemistry. If concentrations of principal radiological constituents change by more than about 10-20 percent between sampling events, it should increase the frequency of sampling to better characterize the temporal variability of groundwater quality. For unconfined groundwater systems, less frequently than biannual sampling schedules may not capture important temporal variations. For most sites, it should be sufficient to sample on a quarterly basis (i.e., one sample per well per calendar quarter) to characterize temporal changes in water quality. More frequent sampling, such as bimonthly may be necessary, however, especially at sites involving offsite or potential offsite contamination of groundwater resources with more reactive, geochemical constituents in a dynamic hydrological system (Barcelona et al., 1989). The sampling frequency needs to be evaluated with respect to the monitoring objective and timeframe over which the monitoring network will be conducted, and the basis documented by the licensee. NUREG/CR-6948, NUREG/CR-7221, EPA (2009a), and Barcelona et al. (1989) include acceptable approaches for assessing frequency of groundwater sampling.

Quarterly sampling of surface water and sediments should be sufficient at most sites. This sampling should be supplemented by additional sampling to characterize the surface water system at representative low- or high-stage flow conditions (e.g., minimum annual, 7-day average low flow or maximum annual, 7-day average high flow). The licensee should use this information to bound the existing and projected impacts of the release of contamination on adjacent surface water bodies.

Note that the incremental benefit of sampling decreases with increasing sampling frequency when there is an autocorrelation in the data. If this autocorrelation is large, a relative low sampling frequency is necessary to avoid sampling redundancy, and the total length of the sampling period must increase to achieve a sufficient return on sampling. The adequacy of the sampling frequency needs to be interpreted or viewed in terms of the time horizon of the sampling program and monitoring objectives (Barcelona et al., 1989).

Licensees have substantial flexibility to modify their monitoring programs for practical and technical purposes. However, some aspects of monitoring programs may be explicitly laid out in the license (e.g., technical specifications, license conditions), changes to which require NRC approval. In addition, all significant changes to a monitoring program should be communicated to the NRC staff. Prior discussions with the NRC staff are recommended when the level of significance is not obvious. Changes to a monitoring program are typically documented (e.g., in annual reports) for later NRC staff review, and help inform inspections.

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F.8 Surface Water and Sediments Surface water can include ponds, creeks, streams, rivers, lakes, coastal tidal waters, oceans, and other bodies of water. Note that certain ditches and intermittently flowing streams qualify as surface water. The licensee should evaluate the need for surface water samples on a case-by-case basis. It should base its surveys for water on appropriate environmental standards for water sampling. If the body of water is included in a larger survey unit, then it should take sediment samples at sample locations selected by the normal method, without taking the body of water into consideration. In addition to the discussion below, MARSSIM (Sections 3.6.3.3 and 5.3.3.3) provides some guidance on evaluating the likelihood for the release of radionuclides into surface water and sediments and on concerns related to characterization and sampling.

Onsite and offsite normal effluent discharges to surface water may be permitted under the license or by wastewater and stormwater permits and registrations. Radioactive effluent reports track these releases and provide dose assessment results for the appropriate scenario pathway.

For effluent discharges to onsite lakes or ponds at power plants, leakage to the groundwater may occur, though it need not be reported again during operations (NRC, 2017b). However, at decommissioning, the residual radioactivity in the lake or pond sediments and any leakage to the groundwater must be treated as residual radioactivity that needs to be dispositioned in accordance with the guidance in this NUREG. For offsite effluent discharges, offsite decommissioning activities are generally not required by NRC regulations.3 REMP reports record radiological results of offsite sampling surrounding the site, including water and sediments at discharge locations. Other agencies, such as the U.S. Environmental Protection Agency (EPA) and State and local entities, may require different treatment of offsite areas that are contaminated by normal effluent discharges. For NRC, however, characterization may be needed to assess environmental impacts as part of the environmental assessment or impact statement. Similarly, groundwater seepage of onsite residual contamination to offsite surface waters must be incorporated into environment assessments or impact statements. Any onsite residual radioactivity in the groundwater during decommissioning must be incorporated into the dose assessment and site release criteria.

For sites that are located near surface water streams and could reasonably affect surface water pathways, the site characterization program should establish background surface water quality by sampling upstream of the site being studied or areas unaffected by any known activity at the site. Water should be collected as grab samples from the stream in a well-mixed zone, focusing on locations in proximity to potential sources or pathways from the facilities. Depending on the significance and the potential for surface water contamination, it may be necessary for certain sites to collect stratified samples from the surface water to determine the distribution of contaminants within the water column. Surface water quality sampling should be accompanied by at least one round of stream sediment quality sampling to assess the relationship between the composition of the dissolved solids, the suspended sediment, and the bedload sediment fractions. The licensee should determine water levels and discharge rates of the stream at the time samples are collected and should also consider the effects of variability of the surface water flow rate. Based on the results of the HSA and preliminary investigation surveys, it should conduct surface scans for gamma activity in areas likely to contain residual activity (e.g., along 3 The license terms of any authorization for offsite effluent discharges vary and site-specific conditions may change over time. Thus, for example, concentrating mechanisms may be present at some sites and should be considered.

In some cases, offsite surveys have been conducted to evaluate the public health and safety risk of accumulated radioactivity in the environment from licensed operations.

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the banks). Korte and Kearl (1984) and USGS (1977, 2018) describe applicable methods for surface water and sediment sampling. In addition, Fleischhauer and Engelder (1984) present suggested procedures for stream sediment sampling. The EPA guidance documents mentioned above are also applicable. In some cases, the REMP data from a facilitys operating period may provide useful information to support the characterization program, although there may be limitations of the use of REMP data to support site characterization.

The licensee should conduct surface water sampling in areas of runoff from active operations.

In case of direct discharge into a stream, it should monitor and sample the outfall and the stream upstream and downstream from the outfall. It should conduct preliminary characterization of the contamination levels by measuring gross alpha and total beta particle activity (total and dissolved) and by obtaining a gamma spectrum for surface water samples. It should be noted that determination of gross alpha activity (and low-energy beta emitters, as well) may be of limited value for samples containing elevated total or dissolved solids concentrations because of sample attenuation. In such instances, gamma spectroscopy might be the only recourse. Specific radionuclide analysis may be needed, depending on the level of activities and type of radionuclides. Non-radiological parameters, such as specific conductance, pH, and total organic carbon, may be used as surrogate indicators of potential contamination, provided a clear relationship is established between radionuclide concentration and the level of the surrogate. Additional analysis for other parameters like volatile and semi-volatile compounds, chelating agents, pesticides, and polychlorinated biphenyls may also be necessary if they affect the mobility of radiological constituents and to evaluate the potential environmental effects of decommissioning.

The licensee should carefully record each of the surface water and sediment sampling locations on the appropriate survey form. It can also use surface water flow models to assist in estimating contaminant concentrations or migration rates.

F.9 Geochemical Conditions The licensee should also describe geochemical conditions at the site and their association with groundwater and contaminants, giving consideration specifically to geochemical conditions that enhance or retard contaminant transport. Geochemical data should include information on solid composition, buffering capacity, redox potential, pH, sorption (represented as a range of distribution coefficients (Kd) for each radiological constituent), and other relevant geochemical data. Piper and Stiff diagrams may be useful for visualizing the geochemistry of the water. In some cases, it may be necessary for licensees (or responsible parties) to use appropriate geochemical codes to understand and quantify geochemical mechanisms that significantly affect transport of radiological and non-radiological contaminants and their potential fate (e.g., geochemical speciation codes such as PHREEQC (Parkhurst and Appelo, 2013) are typically used for this purpose)).

In general, licensees or responsible parties should estimate the values of Kd through laboratory column or batch sorption measurements considering site-specific conditions, if found to be important to estimating dose. If not found to be important to dose, site-specific Kd measurements may not be necessary and generic ranges may be used from published tables (e.g., Sheppard and Thibault, 1990; EPA, 1999, 2004, 2005; and Yu et al., 2015). Licensees (or responsible parties) may use appropriate geochemical codes to understand and quantify geochemical mechanisms that significantly affect transport of radiological and non-radiological contaminants and their potential fate (e.g., MINTEQ (EPA 1984); EQ3/6 (Daveler and Woolery, 1992); PHREEQC (Parkhurst and Appelo, 2013)). For complex sites with the F-14

possibility of transient variable geochemical conditions, the use of surface complexation models in geochemical speciation models may be needed to adequately represent the variability of K d (e.g., NRC, 2003a).

Appendix I of this volume discusses additional information on groundwater parameters necessary for dose modeling.

F.10 Surface Water and Groundwater Models to Support Dose Modeling A graded approach may be used for selecting groundwater models to support estimates of dose at a site. As site complexity and magnitude of residual contamination increase, more sophisticated models are recommended. Models with simplified flow and transport approaches may be sufficient for sites with straightforward hydrogeology and source areas, such as the DandD4 code (NRC, 2001) that uses generic and conservative inputs. As site complexities increase, a code that allows more sophisticated flow and transport equations and site-dependent inputs may be needed, such as RESRAD-ONSITE (Kamboj et al., 2018).

Appendix I contains additional information on hydrogeological conceptual models found in commonly used decommissioning dose modeling codes DandD and RESRAD-ONSITE. The inherent assumptions and limitations of these models are addressed. RESRAD-ONSITE does not consider existing groundwater radionuclide concentrations and only addresses the potential, future transport of residual radioactivity and contamination of ground and surface water and associated doses. 5 At complex sites and in cases where existing residual radioactivity in the groundwater is present, it is likely necessary to construct a more complicated groundwater model to support development of DCGLs. Additionally, in these cases, some allowance for the existing residual radioactivity in the groundwater must be made (e.g., fractionation of the release criteria to account for multiple contaminated media) to ensure the potential dose is not underestimated.

Section F.10.1 discusses several flow and transport numerical models to illustrate the types of models that may be used for complex sites. Although a licensee always has the option of assessing the peak dose within the 1,000-year compliance period based on more realistic modeling, in many cases it may be more straightforward to perform more bounding type analyses to simplify the decommissioning process if compliance can be easily demonstrated.

As a joint effort, the NRC, EPA, and DOE have developed specific guidance on selecting and applying surface water and groundwater models (EPA 1994a, b, c). The reference section lists other documents containing supporting details (NUREG/CR-6805 and NUREG/CR-6948; ASTM D5447-17 (2018c), NCRP 1985, 1996; EPA 1987a, 1987b, 2009b; NAS 1999).

As previously discussed, model simplification is the process for reducing the complexity of a numerical model into a simpler numerical model while still maintaining the validity of the simulation results. Although specific computer codes may be discussed or referenced in this 4 Decommissioning codes, such as DandD, can be obtained from the NRCs RAMP (Radiation Protection Computer Code Analysis and Maintenance Program).

5 More recent versions of RESRAD-OFFSITE (3.2 and 4.0) and RESRAD-ONSITE 7.2 consider sources located below the water table, although NRC guidance on the potential use of RESRAD-OFFSITE to consider existing residual radioactivity in the groundwater has not yet been developed. A future revision to this volume may include evaluation of this tool and its efficacy in considering existing residual radioactivity in groundwater. It is important to note that the tool does not consider existing groundwater plumes not associated with the source area. If existing groundwater plumes are present, a separate evaluation would be needed to assess the dose associated with those plumes.

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guidance, the NRC does not endorse the use of any particular code or modeling software package for analyzing the performance of a decommissioning or disposal site. However, it would be useful to discuss commonly used codes for developing groundwater models by NRC licensees and to know which codes are better suited for constructing complex models and which codes usually provide good results when used to construct simple models. A discussion of commonly used codes and a comparison of the model results obtained by codes used to construct complex models and the results from codes used for simple models should provide for a better understanding of the model simplification process and information on how model simplifications can affect the modeling results.

F.10.1 Codes for Development of More Complex Models The MODFLOW and MT3DMS codes are commonly used to simulate multi-dimensional flow and transport in the subsurface. These codes can simulate three-dimensional flow and transport, and therefore can generally better project contaminant concentrations at various points of exposure compared to codes that are only able to simulate flow and transport in one-dimension. Of course, sufficient data must be available to construct these more complex models, or the additional code capabilities are of limited value. In some cases, it may be preferable or necessary to use simpler codes to simulate groundwater flow and transport when adequate safety margin is available and more conservative assumptions can be made, or when data or resources are limited making it difficult for an analyst to construct a more complex model of the site. However, when additional complexity is warranted there are several off the shelf codes available to assist with constructing these more complex three-dimensional groundwater flow and transport models.

The MODFLOW code is a groundwater flow modeling code developed by the USGS that solves the groundwater flow equation in up to three dimensions using finite-difference approximations.

The code is public domain free software written primarily in FORTRAN and can be compiled and run on Microsoft Windows or Unix-like operating systems. The first version of MODFLOW was published in 1984. It has a modular structure and can be modified to address different applications or problems. MODFLOW-2005 has many new capabilities compared to older versions (Harbaugh, 2005). MODFLOW 6 is the current core MODFLOW version distributed by the USGS (Langevin, et al., 2017; Hughes, et al., 2017). The previous core version, MODFLOW-2005, is actively maintained, supported, and able to handle supplemental MODFLOW features not yet updated for MODFLOW 6.

The MODFLOW-2005 code simulates steady and non-steady flow in an irregularly shaped flow system in which aquifer layers can be confined, unconfined, or a combination of confined and unconfined. Flow from external stresses, such as flow to wells, areal recharge, evapotranspiration, flow to drains, and flow through river beds, can be simulated. Hydraulic conductivities or transmissivities for any layer may differ spatially and be anisotropic (restricted to having the principal directions aligned with the grid axes), and the storage coefficient may be heterogeneous. Specified head and specified flux boundaries can be simulated as can a head dependent flux across the model's outer boundary. MODFLOW 6 uses a new format of blocks and keywords for input of model data and is written using an object-oriented design.

MODFLOW 6 uses a control-volume finite difference approach that can utilize structured or unstructured grids; MODLFOW-2005 requires a Cartesian grid system. MODFLOW 6 presently supports one type of process model the Groundwater Flow Model (GWF). Multiple GWF models can be coupled and concurrently simulated, such as nesting of models at different scales. Other models may be added in the future that can seamlessly couple with the GWF, such as a groundwater transport model, a surface-water model, and a pipe network model. The F-16

new MODFLOW 6 framework will enable the ability to solve multiple, tightly coupled, numerical models in a single system of equations, or simplify the computational effort by loosely coupling the different models. For continuity, licensees may choose to continue using MODFLOW-2005; but if starting fresh or the site has hydrogeologically-significant features that exhibit as complex geometries, licensees should consider using MODFLOW 6.

Other options exist for sites with complex geometries. Codes developed using finite volume or finite element representations may be needed to better represent the geometries of the system.

A finite volume code in the public domain that can handle a wide range of site characteristics and geometries is TOUGH2 (Pruess, et al., 2012). The PORFLOWTM 6 finite-element code has been used by DOE due to its ability to consider radioactive decay, can solve problems involving transient or steady state fluid flow, and mass transport in multi-phase, variably saturated, porous or fractured media. The porous/fractured media may be anisotropic and heterogeneous, arbitrary sources (injection or pumping wells) may be present and, chemical reactions or radioactive decay may take place. The geometry may be 2D or 3D, Cartesian or Cylindrical and the mesh may be structured or unstructured.

The MT3DMS code is a modular three-dimensional transport model for the simulation of advection, dispersion, and chemical reactions of dissolved constituents in groundwater systems (Bedekar, et al., 2016). The MT3DMS code uses a modular structure similar to the structure utilized by MODFLOW. The MT3DMS code is used in conjunction with MODFLOW in a two-step flow and transport simulation. Heads and cell-by-cell flux terms are computed by MODFLOW during the flow simulation and are written to a specially formatted file. This file is then read by MT3DMS and utilized as the flow field for the transport portion of the simulation.

The MT3DMS code differs from MT3D in that it allows for multi-species transport, supports additional solvers, and allows for cell-by-cell input of all model parameters.

F.10.2 Codes Typically Used in Decommissioning The RESRAD-ONSITE (Kamboj et al., 2018) and RESRAD-OFFSITE (NRC, 2020) codes are commonly used for calculating groundwater concentrations and groundwater-dependent dose for decommissioning dose assessments. Details of the RESRAD-ONSITE groundwater models are discussed in Appendix I.

The RESRAD-ONSITE code includes a mass balance and non-dispersion model to compute groundwater concentrations at an onsite well. Neither the RESRAD-ONSITE mass balance, nor the non-dispersion model, considers dispersion, while RESRAD-OFFSITE can consider advection and dispersion in calculating groundwater concentrations at a receptor well. The groundwater transport model in RESRAD-OFFSITE considers 1-D advection (straight or curved flow path), along with 3-D dispersive transport in the saturated zone. Likewise, while only 1-D advection is considered in RESRAD-ONSITE, RESRAD-OFFSITE considers 1-D advection, and 1-D dispersive transport in the unsaturated zone. Furthermore, the unsaturated zone, saturated zone, and contaminated zone7 can be subdivided into smaller zones to increase the accuracy of transport simulations.8 6 PORFLOW was developed by Analytic & Computational Research, Inc. (ACRi, 2008).

7 The capability to more accurately simulate transport through the contaminated zone through use of sub-zones was added in RESRAD-OFFSITE, Version 3.0 8 It is important to note that RESRAD-OFFSITE also has the capability of mimicking the RESRAD-ONSITE code for calculation of doses to an onsite receptor. However, reference to RESRAD-OFFSITE models and calculations in this section pertain to just the offsite capabilities, and not the onsite dose calculations.

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Both RESRAD-ONSITE and RESRAD-OFFSITE have the capability to consider variable transport rates of progeny created during transport in groundwater. The RESRAD-OFFSITE code has two groundwater transport algorithms: the first algorithm considers variable transport rates of parents and progeny, and the second algorithm models longitudinal dispersion 9. When either variable transport rates or dispersion is clearly dominant, the RESRAD-OFFSITE user should choose the transport algorithm that is most important to increase computational efficiency. When both the longitudinal dispersion and the variable transport rates are important, the user has the option of subdividing the transport pathway into a number of subzones to more accurately simulate the transport of progeny in transport, although this approach may significantly increase computation times. Only the zone where the progeny atoms are created would not consider both processes (longitudinal dispersion and variable transport rates of parents and daughters).

Benchmarking of RESRAD-ONSITE with RESRAD-OFFSITE (onsite simulation)

Code developers benchmarked RESRAD-OFFSITE against RESRAD-ONSITE (Yu, 2006) using an earlier version of the RESRAD-OFFSITE code (i.e., benchmarking was conducted prior to the 2007 release of RESRAD-OFFSITE Version 2.0). The results of the benchmarking exercises showed that RESRAD-OFFSITE could mimic the results of RESRAD-ONSITE when certain parameters were changed consistent with the RESRAD-ONSITE conceptual model.

Notable differences between initial simulations run with RESRAD-ONSITE and RESRAD-OFFSITE included travel times to the point of compliance that were attributable to differences in use of porosity in the transport calculations. The current versions of RESRAD-ONSITE (7.2) and RESRAD-OFFSITE Version 4.0 use the same retardation factor expressions by default, although the option in the prototype version of RESRAD-OFFSITE is still available as a user-selectable option in RESRAD-OFFSITE version 4.0. Context specific help on retardation factor in RESRAD-OFFSITE indicates that the key command ALT-F will bring up an option to change the retardation factor expression on the groundwater transport screen. Section H.2.1 of NUREG/CR-7268, Volume 1, Users Manual for RESRAD-OFFSITE Code Version 4, Volume 1, Methodology and Models Used in RESRAD-OFFSITE Code, issued March 2021, provides additional information about options for the retardation factor expression. Another noteworthy difference in results was observed for the water-dependent pathways due to differences in accumulation periods of radioactivity in soil from application of contaminated irrigation water that is considered in RESRAD-OFFSITE over multiple years but is only considered over a single growing season in RESRAD-ONSITE.

9 Both RESRAD-OFFSITE transport algorithms account for transverse dispersion.

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Benchmarking of RESRAD-OFFSITE with Other Codes The RESRAD-OFFSITE10 code was also benchmarked with several other codes as detailed in Gnanapragasam et al. (2000). The benchmarking study is instructive as it provides extensive discussion on the differences in modeling results for more complex and less complex models.

Specifically, the model comparison study illustrates how differences between codes in their treatment of longitudinal and transverse dispersion and differences in ability to model variable transport rates of parents and daughters impact the results. The radionuclides considered in the benchmarking study included relatively short-lived Sr-90, and U-234 decay chain members (U-234, Th-230, Ra-226, Pb-210, and Po-210). Source loading spanned 200 years for Sr-90 and 100 years for U-234 (at a constant rate of 0.88 Ci/y). The RESRAD-OFFSITE, PRESTO, MMSOILS, and MEPAS codes were the four primary codes tested in this study. They have different capabilities with respect to simulation of dispersion and variable transport rates of progeny. Increasing dispersivity with distance, and variable pore water velocity along the flow path were also considered in this study.

The results of the comparison showed markedly different results for PRESTO at increasing distance from the source, and at longer times after source loading ceases (when longitudinal dispersion is important). The PRESTO longitudinal profiles and breakthrough curve results are different from other codes because PRESTO is the only code of the four codes tested that does not consider longitudinal dispersion. However, at shorter distances and for constant sources, the differences between PRESTO and the other codes are less significant.

Another important difference between the codes tested is related to the calculation of retardation. Because the prototype version of RESRAD-OFFSITE is the only code to consider immobile pore water in the calculation of the retardation factor, the calculated retardation factors can be significantly higher for radionuclides with low retardation factors and if the effective and total porosity are significantly different (e.g., in the Gnanapragasam et al. (2000) study RESRAD-OFFSITE calculated retardation factors are higher by a value of 0.56 (unitless) for all radionuclides due to the difference in total and effective porosity at 0.39 and 0.25, respectively).

Differences in retardation factors also have a more significant impact on relatively short-lived11 radionuclides. As stated above, it is important to note that the current version of RESRAD-OFFSITE 4.0 uses the same default value for the retardation expressions as RESRAD-ONSITE 7.2. The option to use the retardation factor expression in the prototype version of RESRAD-OFFSITE is still available as a user-selectable option in RESRAD-OFFSITE 4.0.

All of the codes are slightly different when it comes to important processes for progeny created in transport (i.e., longitudinal dispersion and consideration of variable transport rates of parents and progeny). The PRESTO code considers neither longitudinal dispersion, nor variable transport rates of progeny. The version of RESRAD-OFFSITE tested in the study only considered variable transport rates of daughters but not longitudinal dispersion of progeny (additional features were added to RESRAD-OFFSITE in subsequent versions as described above). When longitudinal dispersion is not considered, a sharp front is observed in the breakthrough curves of progeny; however, if variable transport rates of progeny are considered, as is the case for RESRAD-OFFSITE, higher retardation factors of progeny (e.g., Th-230 has a retardation factor of 198 compared to a retardation factor of 48 for U-234) can lead to long tails 10 The Gnanapragasam et al. (2000) reference refers to what is now called RESRAD-OFFSITE as just RESRAD.

Reference to RESRAD-OFFSITE in this section is to an earlier prototype version of RESRAD-OFFSITE.

11 Short-lived with respect to the observation time or transport time at the point of observation.

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in the breakthrough curves. RESRAD-OFFSITEs capability of modeling variable transport rates of progeny also leads to lower peak concentrations of radionuclides such as Th-230, which has a higher retardation factor than its parent. On the other hand, higher peak concentrations of radionuclides such as Ra-226, Pb-210, and Po-210 result due to (1) the high retardation factor of Th-230, and (2) the lower retardation factors of the radionuclides compared to their immediate parents. Improvements to the accuracy of the breakthrough curves were observed when the aquifer was broken up into smaller segments and longitudinal dispersion and variable transport rates of progeny (compared to parents) was considered in all segments with the exception of the segment of transformation (where either longitudinal dispersion or variable transport rates of the progeny were considered). Presumably, this study led to improvements in the progeny transport models for some of the evaluated codes, including more recent versions of RESRAD-OFFSITE.

The various codes studied also differ with respect to the consideration of dilution in a pumping well (i.e., some codes consider just aquifer concentrations, while other codes such as RESRAD-OFFSITE and PRESTO can consider dilution in a well). Depending on the location of the well and other factors, the impact of pumping may be more or less significant. For example, the study concludes that at exposure points far away from the source, transverse dispersion can be expected to distribute contamination nearly uniformly near the well given the size of the plume in relation to the capture zone. On the other hand, if the well is located close to the source and the plume is narrow or not well-developed, then pumping may lead to dilution of the narrow contaminant plume via mixing with clean water in the larger (than plume) volume of water pulled in by the well.

A prototype version of RESRAD-OFFSITE was also evaluated in a study conducted by the Biospheric Model Validation Study II Working Group on Uranium Mill Tailings (BIOMOVS II 1996). The predictions of well water concentrations and offsite soil accumulation made by the prototype version of RESRAD-OFFSITE were stated by code developers to compare well with the predictions of other codes participating in the study (Yu, 2006).

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Athens, Georgia. 2013.

Fleischhauer, C. and T. Engelder. Procedures for Reconnaissance Stream-Sediment Sampling. Bendix Field Engineering Corporation. GJ/TMC-14. 1985.

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Harbaugh, A.W., MODFLOW-2005, The U.S. Geological Survey Modular Ground-water Modelthe Ground-Water Flow Process: U.S. Geological Survey Techniques and Methods 6-A16. USGS: Reston, Virginia. 2005.

Hughes, J.D., Langevin, C.D., and Banta, E.R. Documentation for the MODFLOW 6 Framework. U.S. Geological Survey Techniques and Methods, Book 6, Chapter A57, 40 pages. USGS: Reston, VA. https://doi.org/10.3133/tm6A57. 2017.

International Atomic Energy Agency (IAEA). "Safety Assessment Methodologies for Near Surface Disposal Facilities (ISAM), Vol. I - Review and Enhancement of Safety Assessment Approaches and Tools; Vol. II - Tests Cases." IAEA: Vienna, Austria. 2004.

Interstate Technology & Regulatory Council (ITRC). Groundwater Statistics and Monitoring Compliance, Statistical Tools for the Project Life Cycle. GSMC-1. ITRC: Washington, DC.

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Kamboj, S., E. Gnanapragasam, and C. Yu. Users Guide for RESRAD-ONSITE Code Version 7.2. Argonne National Laboratory ANL/EVS/TM-18/1. ANL, Environmental Science Division: Argonne, IL. 2018.

Korte, N. and D. Ealey. Procedures for Field Chemical Analyses of Water Samples. Bendix Field Engineering Corporation. Technical Measurements Center Report GJ/TMC-07. DOE:

Grand Junction, CO. 1983.

Korte, N. and P. Kearl. Procedures for the Collection and Preservation of Groundwater and Surface Water Samples and for the Installation of Monitoring Wells. Bendix Field Engineering F-23

Corporation. Technical Measurements Center Report GJ/TMC-08. DOE: Grand Junction, CO.

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Langevin, C.D., Hughes, J.D., Banta, E.R., Niswonger, R.G., Panday, Sorab, and Provost, A.M.

Documentation for the MODFLOW 6 Groundwater Flow Model. U.S. Geological Survey Techniques and Methods, Book 6, Chapter A55, 197 pages. USGS: Reston, VA.

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National Academy of Sciences (NAS). Risk Assessment of Radon in Drinking Water. National Academy Press: Washington, DC. 1999.

National Council on Radiation Protection and Measurements (NCRP). NCRP Report No. 76, Radiological Assessment: Predicting the Transport, Bioaccumulation and Uptake by Man of Radionuclides Released to the Environment. NCRP Publications: Bethesda, Maryland. 1985.

. NCRP Report No. 123, Vol. 2, Screening Models for Releases of Radionuclides to Atmosphere, Surfaces, Water, and Ground. NCRP Publications: Bethesda, Maryland.

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Nuclear Regulatory Commission (U.S.) (NRC). Regulatory Guide 4.15, Rev. 1, Quality Assurance for Radiological Monitoring Programs (Normal Operations)Effluent Streams and Environment. NRC: Washington, DC. February 1979.

. NUREG-1293, Quality Assurance Guidance for Low-Level Radioactive Waste Disposal Facility. NRC: Washington, DC. January 1989a.

. NUREG-1383, Guidance on the Application of Quality Assurance for Characterizing a Low-Level Radioactive Disposal Site. NRC: Washington, DC.

November 1989b.

. NUREG-1388, Environmental Monitoring of Low-Level Radioactive Waste Disposal Facility. NRC: Washington, DC. December 1989c.

. Draft Branch Technical Position on Site Characterization for Decommissioning.

NRC: Washington, DC. November 1994.

. NUREG-1496, Generic Environmental Impact Statement in Support of Rulemaking on Radiological Criteria for License Termination of NRC-Licensed Nuclear Facilities. NRC: Washington, DC. July 1997.

. NUREG-1575, Rev.1. Multi-Agency Radiation Survey and Site Investigation Manual (MARSSIM). EPA 402-R-97-016, Rev.1. DOE/EH-0624, Rev.1. NRC: Washington, DC. 2000.

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. NUREG/CR-6805, A Comprehensive Strategy of Hydrogeologic Modeling and Uncertainty Analysis for Nuclear Facilities and Sites. NRC: Washington, D.C. July 2003b.

. NUREG/CR-6884, Model Abstraction Techniques for Soil-Water Flow and Transport. NRC: Washington, D.C. December 2006.

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APPENDIX G SPECIAL ISSUES ASSOCIATED WITH DOSE MODELING, CHARACTERIZATION, AND SURVEY

G.1 Introduction There are several special situations during the decommissioning process that are not, or are only minimally, addressed in the regulatory guidance of the NRC and in NUREG-1575, Revision 1, Multi-Agency Radiological Survey and Site Investigation Manual (MARSSIM), issued August 2000. In these situations, licensees may need to perform characterization and a final status survey (FSS) to demonstrate compliance with the license termination criteria in Title 10 of the Code of Federal Regulations (10 CFR) Part 20, Standards for Protection Against Radiation, Subpart E, Radiological Criteria for License Termination. As part of its review and approval of DPs and LTPs, the NRC staff at this time evaluates these special situations on a case-by-case basis. The NRC may develop additional guidance in the future that covers these special situations and will include them in revisions to the consolidated guidance.

This Appendix G applies, either in total or in part, to Decommissioning Groups 4-7.

G.2 Surveys for Special Situations in Buildings The survey method described thus far in this volume (e.g., Chapter 4 and Appendix A) applies to simple ideal geometries in a straightforward manner; however, there are likely to be some additional special situations at actual sites that will need further consideration. For each situation discussed below, it is assumed that the HSA and minimal site characterization have located and given a rough estimate of the concentration of residual radioactivity present.

G.2.1 Structures Versus Equipment G.2.1.1 Background The NRC staff acknowledges that the relationship between the LTR for unrestricted use of a site (dose criteria of 0.25 mSv/y (25 mrem/y) and ALARA found in 10 CFR 20.1402), and existing guidance for unrestricted releases of solid materials from a site on a case-by-case basis under 10 CFR 20.2002 may have been unclear. In particular, the criteria for the LTR and for releases of solid materials with small or no amounts of residual radioactivity before license termination are different. Consistent with the LTR, once a site meets the radiological criteria for unrestricted use and the NRC terminates the license, solid material may be removed from a site. However, before license termination, material cannot be removed from the site for unrestricted use unless it meets either (1) criteria already approved for the licensed facility (e.g., in a license condition),

for surficially contaminated materials, or (2) the few mrem/y criterion for the case-by-case approach for volumetrically contaminated materials (see Section 15.11 in Volume 1 of this NUREG). One rationale for the difference in criteria is that the technical basis for the LTR assumes that individuals are generally exposed to residual radioactivity at a single location (the site), while, for releases of solid material, an individual may be exposed to materials through several exposure scenarios at offsite locations. For more information about the relationship between the LTR and the case-by-case approaches to release of solid materials from a site, see the LTR Analysis Commission Paper, SECY-03-0069, Results of the License Termination Rule Analysis, dated May 2, 2003, and the associated Regulatory Issue Summary 2004-08, Results of the License Termination Rule Analysis, issued May 2004.

This section focuses on compliance with the LTR, in particular, the building structure-related materials that may be left on site at license termination and the criteria that should apply.

Section 15.11 of Volume 1 of this NUREG provides more information about current approaches to releases of solid material before license termination.

G-1

G.2.1.2 Implementation The LTR applies to building structures that remain in place after decommissioning and does not apply to releases of equipment from the facility before license termination. If licensees elect to dismantle building structures and dispose of the associated materials off site (in accordance with applicable regulatory requirements), rather than leave the building structures in place (for unrestricted use), the LTR does not apply to the associated materials moved off site before license termination. Materials licensees may release equipment and building structure deconstruction and dismantlement materials in accordance with existing license conditions.

Reactor licensees (licensed under 10 CFR Part 50, Domestic Licensing of Production and Utilization Facilities) may release equipment and building structure deconstruction and dismantlement materials in accordance with the guidance in Inspection and Enforcement Circular 81-07, Control of Radioactively Contaminated Material, dated May 14, 1981; Information Notice 85-92, Surveys of Wastes Before Disposal from Nuclear Reactor Facilities, dated December 2, 1985; and Information Notice 88-22, Disposal of Sludge from Onsite Sewage Treatment Facilities at Nuclear Power Stations, dated May 12, 1988. Licensees should refer to Section 15.11 of Volume 1 of this NUREG report and should contact the NRC staff for further guidance on equipment and solid material releases.

When the LTR was developed, the NRC assumed that decommissioning generally would include the removal of systems and components from onsite buildings before license termination. However, with experience, it has become clear that each licensee uses a different approach for decommissioning, and these approaches are not necessarily consistent with the original assumptions of the LTR. Differences are the result of factors such as (1) the potential for reuse of systems and components, (2) cost of recycling and price of scrap metal and concrete, and (3) cost and availability of disposal options.

It is clear from the LTR technical basis in NUREG-1496, Generic Environmental Impact Statement in Support of Rulemaking on Radiological Criteria for License Termination of NRC-Licensed Nuclear Facilities, issued July 1997, and NRC draft Regulatory Guide DG-4006, Demonstrating Compliance with the Radiological Criteria for License Termination, issued July 1998, that the LTR was not intended to apply to releases of equipment from the facility.

Equipment includes anything not attached to, or not an integral part of, the building structure.

On the other hand, previous guidance (the previous version of Section G.1.1 of NUREG-1757, Volume 2, Revision 0) was not prescriptive enough to provide a definitive answer about whether systems and components must be considered building structures or equipment. The previous guidance considered doors, windows, sinks, lighting fixtures, utility lines, built-in laboratory hoods and benches, and other types of built-in furniture to be part of the structure.

Under that guidance, those items could be included in the FSS and left in place at license termination. It could be argued that, based on the examples provided, many plant systems and components also could be considered building structures, and, therefore, left in place at license termination. This previous guidance may have been inconsistent with the discussion in the LTR Analysis Commission Paper, SECY-03-0069, which described an expectation that removable materials and equipment would generally not be present at the time of license termination.

The staff has identified a number of acceptable approaches to clarify what building structure-related materials may be left on site at the time of license termination and what criteria should be applied to those materials.

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For this discussion only, the NRC staff uses the following descriptions of building structures, systems and components, and equipment:

  • Building structures include floors, walls, and roofs; components embedded in floors, walls, and roofs (e.g., embedded piping); and items that are attached to and are an integral part of the buildings (e.g., doors and windows).
  • Systems and components include items attached to a building structure that are not an integral part of the building but provide important functions to the building (e.g., utility lines, sinks, lighting fixtures, built-in laboratory hoods and benches, polar cranes (in power reactors), and major process equipment).
  • Equipment includes items not attached to the building structure that are generally readily removable from the building. Examples of equipment include furniture or appliances that are not built into or attached to the structure; stocks of chemicals, reagents, metals, and other supplies; motor vehicles; and any other items that normally would not be conveyed with a building when it is sold.

G.2.1.3 Building Structures, and Systems and Components that may be Left in Place at License Termination The NRC staff finds the following approaches acceptable to determine what materials may be left in buildings at license termination.

  • Materials Left Onsite Meet Previously Approved Release CriteriaBuilding structures and systems and components may be left in place if residual radioactivity in all materials is within the licensees previously approved criteria for releases of solid materials for unrestricted use. Such criteria may have been approved in license conditions, technical specifications, or generic NRC guidance. The criteria could include use of the no-detect policy for reactor licensees, or Policy and Guidance Directive FC 83-23, Termination of Byproduct, Source and Special Nuclear Material Licenses, 1983, for materials licensees (see also Section 15.11 of Volume 1 of this NUREG report for more information about the current approaches to releases of solid materials).
  • Materials Left Onsite Meet Few Millirem per YearBuilding structures and systems and components may be left in place if residual radioactivity in all materials is volumetrically distributed (not surficial) and if the potential dose from offsite use exposure scenarios is no greater than a few hundredths of a mSv per year (few mrem per year).
  • Materials Left Onsite Meet 0.25 mSv/y (25 mrem/y)Building structures may be left in place if the potential dose from the residual radioactivity in or on the structures is within the applicable dose criteria of the LTR (for unrestricted use, no greater than 0.25 mSv/y (25 mrem/y) and ALARA).
  • Alternative ApproachesLicensees also may propose alternative approaches, which the staff will review on a case-by-case basis. Before submitting such alternative approaches, licensees should contact the NRC staff.

For all approaches, the residual radioactivity in building structures, systems and components, and all other media at the site (e.g., soils or groundwater) must be in compliance with the G-3

applicable criteria of the LTR (e.g., for unrestricted use, doses must not exceed 0.25 mSv/y (25 mrem/y) and must be ALARA).

Licensees will perform dose assessments (or use NRC-approved screening dose assessments) to demonstrate compliance with the dose criteria. Typically, licensees may not need to evaluate potential offsite future use exposure scenarios, such as removal of soil for fill material or road base or reuse of concrete as roadbed material, because such offsite use exposure scenarios are usually bounded by onsite use exposure scenarios. However, for some of the dose assessments needed for the above approaches, when less conservative and more realistic exposure scenarios are selected, the onsite exposure scenarios may no longer bound potential offsite use exposure scenarios. Thus, in these cases, the licensee should evaluate offsite use exposure scenarios. For additional guidance, see the Offsite Exposure Scenarios subsection in Section I.3.3.3 of Appendix I of this volume.

G.2.1.4 Equipment Not Covered by the License Termination Rule The LTR does not apply to equipment, so equipment should not be left on the site at license termination. Equipment should be released under the current approaches for releases of solid materials, as discussed in Section 15.11 of Volume 1 of this NUREG report or could be disposed of as radioactive waste.

G.2.2 Residual Radioactivity Beneath the Surface The HSA and characterization surveys may indicate that residual radioactivity is present beneath the surface. In the dose modeling, direct exposure, inhalation, and ingestion pathways may be important for residual radioactivity on the surface. However, if the residual radioactivity is located in subsurface soils, additional pathways may become important to dose, and subsurface soil derived concentration guideline levels (DCGLs) may be derived. In these cases, surveys may be conducted, and results may be interpreted in a manner consistent with the dose modeling (see Section G.3 for additional details).

The FSS surveys cracks and crevices in the same manner as other building surfaces, except that these areas should receive judgmental scans when scanning coverage is less than 100 percent.

For painted-over residual radioactivity, the licensee should use the HSA and characterization surveys to determine whether residual radioactivity was fixed in place by being painted over. If so, it may consider the process for its removal in developing the parameters for the dose modeling and may interpret the survey results in a manner consistent with the dose modeling.

G.2.3 Sewer Systems, Waste Plumbing Systems and Floor Drains The HSA and characterization surveys determine whether there are unusual or unexpected levels of residual radioactivity in sewer systems and floor drains. Residual radioactivity in sewer systems and floor drains generally does not contribute to the dose pathways in the building occupancy exposure scenario or the residential exposure scenario. Thus, the licensee should calculate the dose from residual radioactivity in sewer pipes using a site-specific exposure scenario and conduct the FSS in a manner consistent with that scenario. If the sewer water is sent to an onsite drainage field or cesspool, the licensee should evaluate and survey any residual radioactivity as subsurface residual radioactivity. If unusual or unexpected results are G-4

found during the characterization survey, the licensee should handle the situation on a case-by-case basis.

If sewage is sent to an onsite drainage field, any residual radioactivity is subsurface, and the survey methods discussed in Section G.3.1 are appropriate.

G.2.4 Ventilation Ducts The HSA and characterization surveys indicate whether residual radioactivity may be present.

External duct surfaces of ventilation ducts are surveyed as if they are a part of the building surface. The licensee should survey internal duct surfaces in a manner consistent with the dose modeling assumptions.

G.2.5 Piping and Embedded Piping Embedded piping is piping embedded in a durable material, typically concrete, that cannot be easily removed without significant effort and tools. The HSA and characterization surveys indicate whether residual radioactivity is present in piping. The normal room surveys will adequately account for direct (external gamma) radiation from the pipes when the pipes are in place and undisturbed. The direct (external gamma) dose from the pipes will be in addition to the dose from the residual radioactivity on surfaces in the room. It may also be necessary consider building renovation that would disturb the piping, as described in NUREG/CR-5512, Volume 1, Residual Radioactive Contamination from Decommissioning, issued October 1992.

If this is done, the survey should be consistent with the dose modeling assumptions.

NRC staff experience has shown that some DPs have not adequately described the methods the licensee plans to use when surveying the embedded piping planned to be left behind.

Often, licensees have not discussed the methodology for conducting surveys of embedded pipe planned to be left behind, nor have they provided sufficient justification for the assumptions considered in the dose modeling analysis. Regulatory Issue Summary 2002-02, Lessons Learned: Related to Recently Submitted Decommissioning Plans and License Termination Plans, issued January 2002, contains a detailed discussion of this issue.

G.3 Surveys for Special Situations on Land G.3.1 Subsurface Residual Radioactivity Surveys MARSSIM addresses radiological surveys of surface soils only (i.e., subsurface radiological surveys are not within the scope of MARSSIM). Because the MARSSIM FSS method was designed specifically for residual radioactivity in surface soils, if significant amounts of residual radioactivity are located at depth (e.g., significant quantities of residual radioactivity in soils deeper than approximately 15 centimeters), the presence of subsurface residual radioactivity should be taken into consideration in designing the FSS. The licensee should first determine whether it needs surveys of subsurface residual radioactivity. The HSA and other surveys will play an important role in determining whether there is likely to be residual radioactivity in the subsurface. Modeling can also be used to supplement survey data to determine the potential for residual radioactivity to be present in significant quantities in subsurface soils or groundwater due to environmental transport. If the survey data and supplemental modeling indicate that G-5

there is little likelihood of significant subsurface residual radioactivity, then subsurface surveys are likely unnecessary.

If the survey data indicates that there is substantial subsurface residual radioactivity, and the licensee plans to terminate the license with some subsurface residual radioactivity in place, the FSS should consider the subsurface residual radioactivity to demonstrate compliance with the radiological criteria for license termination. To prepare for the FSS, the characterization survey determines the depth of the residual radioactivity. In addition to conventional drilling, the licensee may consider the use of exploratory trenches and pits, where the patterns, locations, and depths are determined using prior survey results or HSA data.

Performing radiological surveys at sites with significant quantities of subsurface residual radioactivity is more complex compared to surveys of surface soils given the relative inaccessibility of the subsurface regions (e.g., subsurface soils cannot be scanned for elevated areas without the extraction of subsurface materials). Additionally, heterogeneous materials are often encountered in the subsurface, and the presence of contaminated groundwater also presents challenges to subsurface radiological surveys (see Appendix F). Because the MARSSIM methodology relies heavily on scanning to identify elevated areas of concern, alternative or supplemental methods are needed when residual radioactivity is present in the subsurface. Modeling may help inform and supplement collection of radiological survey data and help alleviate the challenge of adequately characterizing the subsurface when scanning is not a viable option. NUREG/CR-7021 (A Subsurface Decision Model for Supporting Environmental Compliance) presents a framework focused on development of a conceptual site model referred to as a contamination concern map (CCM). The CCM describes the extent, location, and significance of residual radioactivity relative to the decision criteria. The CCM can be developed with the aid of visualization, geographic information system (GIS), and geostatistical software. As additional data are collected, the CCM transitions from a more qualitative description to a more quantitative and detailed map. Subsurface concentration estimates and uncertainty measures serve as surrogates to scanning to facilitate more optimal sampling designs and decision-making. The approach laid out in NUREG/CR-7026 (Application of Model Abstraction Techniques to Simulate Transport in Soils) presents one potentially acceptable method that may be used in conjunction with radiological survey data to demonstrate compliance. For complex decommissioning cases where subsurface residual radioactivity and groundwater contamination are present, it is important to work with NRC early in the process to discuss acceptable approaches for demonstrating compliance with radiological criteria for license termination.

As discussed above, GIS and geostatistical software are available to assist with designing, performing, and evaluating the results of radiological investigations. GIS tools can be used to help with creation of conceptual models (e.g., provides spatial context and a better understanding of site features that may control or enhance radionuclide transport in the environment). Figures created with GIS software can also assist with identifying relatively homogeneous areas of residual radioactivity for delineation of survey units. Examples of features that can be captured on a figure using GIS tools include the following:

  • Study area and property boundary
  • Buildings where residual radioactivity may be present
  • Roads
  • Surface water features (streams, ponds, runoff basins, ditches, culverts)

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  • Underground features (underground storage tanks, piping)
  • Topography, surface geology, and outcrop locations
  • Hydrostratigraphic surfaces and isopach maps
  • Water table and potentiometric surfaces
  • Sampling locations
  • Monitoring well locations
  • Contaminant distributions Geostatistical tools can be used to create figures showing contaminant distributions, predict radionuclide concentrations in areas where no data exists, and identify areas with a higher probability of residual radioactivity above levels of concern. This information can be beneficial in designing the scoping, characterization, and remediation surveys to define the nature and extent of residual radioactivity (e.g., optimizing the number and locations of samples).

Figure G-1 presents an example use of geostatistical software.

If subsurface residual radioactivity is present, dose modeling may be conducted for both surface (if present) and subsurface soils and DCGLs developed for each. In these cases, the MARSSIM methodology will need to be supplemented or an alternative methodology will need to be developed to demonstrate compliance with radiological criteria for license termination (MARSSIM only addresses residual radioactivity at the surface). Because the depth and thickness of residual radioactivity are correlated to dose, the modeling should reflect the actual distribution of radioactivity in the survey unit. For example, for certain radionuclides (e.g., those whose risk is dominated by the plant ingestion pathway), the thickness of residual radioactivity is strongly correlated to dose. If the modeling assumes a thinner layer of residual radioactivity than is present, then the risk could be significantly underestimated. If the modeling assumes a thicker layer of residual radioactivity than is present, then the risk could be significantly overestimated. Additionally, for some radionuclides (e.g., those whose risk is dominated by the external dose pathway), the surface concentration may drive the risk as radiation emitted from residual radioactivity located at greater depth may be attenuated in the soil column and not contribute to dose. Therefore, if vertical heterogeneity is an issue, then it may be necessary to take discrete samples to ensure that higher concentration residual radioactivity at the surface is not diluted in cleaner materials at depth. Dose modeling can be used to determine the sensitivity of dose to these parameters, and the soil sampling design should ultimately be consistent with the modeling used to develop the DCGLs. Ideally, sufficient resolution in the sampling data would be available to evaluate vertical heterogeneity and calculate appropriate concentrations for comparison against DCGLs derived for specific depths and thicknesses and/or for the total thickness of residual radioactivity to ensure dose is not underestimated.

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GIS tools can be used to assist with designing and interpreting results of radiological surveys.

Figure G.1 shows a map that includes the location of two hypothetical tanks (a). Leaks are known to have occurred near the tanks. GIS information on the location of important features and topography of surficial or subsurface structure can be used to identify areas where residual radioactivity may be present and more likely to have been transported (e.g., surface runoff direction). GIS information and geostatistical tools can be helpful in designing survey plans and identifying areas most likely to be above risk-based thresholds. For example, the geostatistical tools available in codes such as Visual Sample Plan and Spatial Analysis and Decision Assistance (SADA) can be used to analyze data and extend data in areas where no data are available. Figure G.1 illustrates the use of SADA (Version 5) in creating a display of the results of sampling (b) and a 3D visualization of the volume of soil most likely to be impacted based on the sampling results and use of geostatistical tools available in the code (c).

Dose modeling and DCGL development considerations for subsurface residual radioactivity were some of several topics at a July 14-15, 2021, subsurface soils survey workshop (Agencywide Documents Access and Management System (ADAMS) Accession No. ML21208A206). Workshop presentations stressed the importance of studying the impact of area, thickness, and depth of residual activity on dose through sensitivity analysis. The results of the sensitivity analysis should be considered in radiological survey design and considered as part of the data quality objectives process to ensure the key question can be answered: does the site meet the radiological criteria for license termination? For example, significant pathways vary by soil depth and the sensitivity to soil depth would be important for determining whether multiple DCGLs need to be developed and whether depth-discrete sampling is needed (Barr, 2021). Dominant pathways for surface soils would include incidental ingestion and inhalation, as well as external radiation exposure. The plant pathway may be important for intermediate soil depths, while deeper areas will involve groundwater-dependent pathways. Buried radioactivity could come closer to the surface with time due to erosion; therefore, erosion rates could also be an important model parameter. Intrusion scenarios where buried radioactivity is brought to the surface were also discussed due to their importance to DCGL development.

Yu et al. (2021) presented specific results of sensitivity analyses conducted with four radionuclides (cobalt-60, plutonium, technetium- (Tc)-99, and strontium-90) and varying thickness, area and cover (i.e., soil cover thickness or depth of residual radioactivity), as well as results where the impact of distribution coefficients is studied. The geometry of residual radioactivity can be defined using survey data. The sensitivity analysis results show that the DCGL becomes more restrictive as the thickness of soil increases with changes in the slope of the DCGL versus thickness curve occurring at 0.15 cm or 0.9 m, or both, corresponding to the mixing (till) depth and plant root depth depending on the key radionuclide being studied and associated dominant pathway. Area factors are used to consider the impact of area on dose and DCGLs (generally smaller areas lead to lower doses and higher DCGLs). RESRAD-OFFSITE has the capability of evaluating the impact of geometry (location of the receptor in relation to the elevated area) on dose, or RESRAD-ONSITE can be used to determine the impact of area on dose with the simplifying (and conservative) assumption that the receptor is located on the elevated area. Results were also presented showing the impact of area and thickness of cover (corresponding to depth of residual radioactivity). The cover thickness was unimportant for Tc-99, with the Tc-99 dose being strongly related to area of contamination (total inventory drives the dose from the groundwater pathway). Figure G.2 depicts the depth and thickness of residual radioactivity that are important to various pathways. Similar to area G-8

factors, depth factors could be used to account for the impact of depth of residual radioactivity on dose.

(a)

(b)

(c)

Figure G.1 Use of GIS and Geostatistical Tools (a) to Identify Potential Areas where Residual Radioactivity could be Present, (b) to Visualize Borehole Sampling Results and (c)

Interpolate Data and Determine Probability of Exceeding a Threshold following Scoping (top) and Create a more Refined Map following Characterization (bottom)

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Figure G.2 Importance of Contaminated Zone (orange color) Thickness and Depth to Various Pathways (Yu et al., 2021)

When the licensee establishes the appropriate DCGLs and mixing volumes, based on an acceptable site-specific dose assessment, the FSS takes core samples to the measured depth of the residual radioactivity. The number of cores to be taken is initially the number (N) required for the WRS or Sign test, as appropriate. The adjustment to the grid spacing for an EMC is more complicated than for surface soils, because scanning is not applicable. The core samples should be homogenized over a soil thickness that is consistent with assumptions made in the dose assessment, typically not exceeding 1 meter in depth, although finer vertical resolution is always acceptable, and results may be averaged over an appropriate thickness. Note that it is not acceptable to average radionuclide concentrations over an arbitrary soil thickness. The appropriate test (WRS or Sign) is then applied to the sample results. Triangular grids are recommended, because they are slightly more effective in locating areas of elevated concentrations. Site-specific EMCs may also need to be developed to demonstrate regulatory compliance and should take into consideration key radionuclides, pathways, and exposure scenarios important to dose. For subsurface residual radioactivity at depth, the groundwater pathway and total inventory may drive the risk (i.e., small, elevated areas of concentration may not be important to dose). Most intrusion scenarios assume some minimum degree of mixing of excavated soils; therefore, mixing arguments can be presented when determining the minimum volume of soil of interest in developing EMCs. The NRC has not yet developed generic guidance for performing an EMC for subsurface samples; therefore, licensees should discuss this matter with the NRC staff on a case-by-case basis.

The sampling approach described above may not be necessary if sufficient data to characterize the subsurface residual radioactivity are available from other sources. For example, for some burials conducted under prior NRC regulations, the records on the material buried may be sufficient to demonstrate compliance with the radiological criteria for license termination if the total inventory consisted of short-lived radionuclides that have since decayed away or concentrations are sufficiently low that release to deep subsurface soils and groundwater is not a concern.

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As noted above, the NRC held a workshop July 14-15, 2021, to discuss radiological surveys and dose modeling associated with subsurface residual radioactivity. The NRC staff will consider the information presented and discussed at the workshop and a workshop held on May 11, 2022, in developing guidance on subsurface radiological surveys. Some conclusions and findings from the 2021 workshop include the following (see presentations at ADAMs Accession No. ML21208A206):

  • Electric Power Research Institute (EPRI) Report 1013511, Connecticut Yankee Decommissioning Experience Report: Detailed Experiences 1996-2006, issued November 2006, discusses a case study for a site with significant subsurface contamination. Direct push samples and core drilling of bedrock were used to extract cores or to facilitate down hole gamma logging. Areas above the DCGLs were remediated through excavation. Portable gamma spectroscopy equipment was used to survey the bottom of the bedrock excavation. There was concern about missed activity in bedrock, so the licensee used a graded sampling approach approved by the NRC.

Hard-to-detect radionuclides were present.

  • EPRI Report 3002007554, Guidance for Using Geostatistics in Developing a Site Final Status Survey Program for Plant Decommissioning, issued May 2016 provides a review of geostatistics software, including Visual Sample Plan (VSP) and Spatial Analysis and Decision Assistance (SADA), as well as a summary table. The EPRI report provides a roadmap for applying geostatistics, including major phases of geostatistical analysis, steps within each phase, and key questions associated with each step.
  • Geostatistical approaches have a long history in non-radiological applications and have been developed in the mining industry for characterization, for example. However, one example of a radiological application is at the Fontenay-aux-Roses site in France (see Section 4.3 of EPRI Report 3002007554. The site had a relatively thin layer of contamination along a vertical gradient along the bank of a former moat. Initial drill hole campaigns for cesium-137 were used to develop a three-dimensional kriging map, which in turn informed additional sampling campaigns as well as development of a remediation plan.
  • A geostatistical approach has been used at both excavated and nonexcavated sites. Any area with three-dimensional contamination, such as contaminated concrete, could benefit from geostatistics. It may also be useful for designing sampling plans and to guide remediation.
  • Geostatistics is just one tool to address subsurface problems. Supplemental information about physical boundaries and contaminant transport should be considered in developing a spatial model to ensure that unrealistic results do not occur. Leveraging expertise in multiple disciplines and relying on more than one tool will help limit decision errors and lead to more stable decision-making.
  • In the past, characterization was more reactive. If there was a concern about impacts to the environment or exposure to members of the public, a team would mobilize to the site, dig things up, and take soil borings that were then sent off to a laboratory. The laboratory results would come back and the results would be assessed. Likely the G-11

extent of contamination was not bounded or the source had not been identified.

Perhaps some remediation would take place and remobilization and resampling would occur. This process would be repeated until the site was found to be clean. While often effective for certain project goals, the process was drawn out and expensive.

  • Regulations that required better recordkeeping and advancements in technologies made the decommissioning process more efficient. For example, advancements in field measurement technologies have allowed more measurements and decision-making in the field during assessment and remediation. This is formalized in the U.S.

Environmental Protection Agencys Triad approach with (1) systematic planning (i.e.,

historical site assessment, development of the conceptual site model), (2) dynamic work strategies, and (3) real-time measurements (using mobile laboratories and instrumentation, remote sensing, and GPS/GIS data to create a digital twin of the site).

  • The Common Data Environment (CDE) represents an investment in characterization to reduce remediation and waste disposal costs. It includes a living conceptual site model that contains GIS information/models (e.g., risk ranking of systems, structures, and components (SSCs), land use, hydrogeological data, LIDAR) integrated with a building information model that has architectural, mechanical, and structural facility data embedded in a three-dimensional digital twin. The Nuclear Energy Institute (NEI) SSC risk ranking guidance (NEI 09-14, Guideline for the Management of Underground Piping and Tank Integrity, and NEI 07-07, Industry Ground Water Protection InitiativeFinal Guidance Document) provide the data to input into the GIS and building information model. Nuclear facilities have design controls and extensive documentation that makes them well suited to use the CDE, which can be employed to know what was happening at different plant locations even after the structures themselves are gone. Once set up, the site model can be used when needed to investigate issues even before decommissioning and can be modified rapidly when needed.
  • The CDE approach is data intensive, and the software tools can be expensive, so it may not be a solution for smaller sites but data from larger sites, such as reactor decommissioning sites, could be leveraged to construct a CDE.

G.3.2 Surveys of Excavations and Use of Backfill Soil for Excavated Land Areas G.3.2.1 Surveys of Excavations In cases where a licensee must remediate a site through excavation of subsurface residual radioactivity above cleanup levels, several options are available to the licensee to demonstrate compliance with radiological criteria for license termination. Although a backfilled excavation represents the final configuration of the site, it is a reasonable expectation that the licensee will perform the FSS on the open excavation prior to backfilling, if the survey can be performed safely. This is due to the potential cost and difficulty associated with adequately sampling a backfilled survey unit and the fact that scanning the entire depth of backfill would likely not be possible for most situations. Sampling and scanning of the open excavation are also beneficial by helping to ensure that residual radioactivity above levels that would lead to an exceedance of the dose criteria are removed and appropriately disposed. When a FSS is performed on an open excavation, it is important to document the locations and depth range below final grade represented by sampling as well as the general topographical layout of the excavation relative to final grade to understand the final distribution of residual radioactivity at the site and to facilitate comparison to release criteria. Additionally, it is important for the licensee to G-12

communicate with the NRC staff to allow the NRC to perform a confirmatory survey that provides independent evaluation of radiological conditions before the licensee backfills the excavation.

Survey instruments should be used that are appropriate for evaluating the radioactive contamination of interest and all accessible surfaces of Class 1 survey units should be evaluated. Specialized equipment such as extended instrument cables/poles, trench boxes, or other tools and equipment may be needed to perform surface scans and to obtain samples from locations within the excavation and sidewalls suspected of contamination. Deep excavations may require significant preparatory work both to perform the excavation and to make an excavation safe to access. Examples of this include pumping to lower the water table, using sheet pilings to shore up excavation sidewalls (likely making the sidewalls inaccessible for direct surveying/sampling efforts), and implementing appropriate respiratory protection if toxic chemicals are present or could displace oxygen. In all cases, the health and safety of workers and the public should take priority.1 If the health and safety of personnel may be compromised by sampling or scanning, the NRC staff has, in the past, accommodated the use of in-situ gamma spectroscopy as a substitute for sampling or scanning described in MARSSIM. In-situ gamma spectroscopy poses its own technical challenges for adequately characterizing a site and should be discussed with the NRC staff if a licensee feels it presents the best solution for an unsafe working environment.

There are several approaches a licensee can take for sampling an excavation. Although MARSSIM does not apply to subsurface soils, a MARSSIM-based approach may be extended to subsurface problems. The survey classification of an excavation should consider whether the entire excavated area, including the floor and the sidewalls, has the same contamination potential. Also, if remediation of soil took place to meet the release criteria, then a MARSSIM Class 1 classification would be appropriate. In this case, if the floor and sidewalls of the excavation both possess the same potential for contamination, then the same level of survey effort should be applied to both areas. However, there may be cases where sloped sidewalls into an excavation are created to provide safe access for remediation activities, and the contamination potential of the sidewalls could be lower than the bottom of the excavation. In this example, the new sidewalls may be a Class 2 buffer area, whereas the floor of the excavation should be a Class 1 area. As such, it is important to evaluate excavation practices and sampling strategies during the DQO process to ensure that classification and survey methodologies account for actual site conditions. It is also important to consider the potential for ongoing decommissioning activities to contaminate buffer areas in an excavation. Although the sidewalls or bottom of the excavation may not be planar surfaces, sampling locations could be established based on a flat, final graded area (see Figures G.2 (a) and (b)). The depth of sampling of the excavation should be considered during the DQO process. Sampling of sloped excavation sidewalls that are excavated to a depth shallower than the bottom of the excavation pose additional considerations due to their difference in elevation below grade. Different sample collection strategies such as sampling more horizontally or vertically into the sidewall, may also be considered during the DQO process. In addition, a sampling strategy may need to take into account the slope of sidewalls, sample density, thickness of strata, and multiple DCGLs developed to represent different soil intervals. Licensees are encouraged to discuss with the NRC staff the appropriate level of detail and technical features of their sampling 1 Proper excavation safety practices should always be applied in accordance with Occupational Safety and Health Administration (OSHA), such as the use of support systems, and/or sloping and benching, to stabilize the excavation site. Refer to the US Army Corps of Engineers, Safety and Health Requirements Manual EM 385-1-1 for additional details.

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strategy. In many cases, licensees have used a layered approach in which multiple subsurface layers or strata are considered individually and then the cumulative risk from the multiple layers or strata are assessed. If the sidewalls of the excavation are considered the same Class 1 survey unit consistent with the bottom of the excavation, then the cumulative risk from all of the strata (below and above the bottom of the excavation) should be considered. Alternative approaches are available to assess the data.

As noted above, in some cases multiple sets of DCGLs may have been approved for different strata or layers below final grade based on dose modeling. For example, surface soil is important to the external dose and inhalation pathways, intermediate depth soil corresponding to the root depth is potentially important to the plant ingestion pathway, and subsurface soil may be important to the groundwater pathway. Intrusion scenarios are considered to develop subsurface DCGLs for otherwise inaccessible subsurface soils at depth (see Appendix J). In cases where different sets of DCGLs are developed for different strata, it is important to ensure the average contaminant concentration in each designated stratum is lower than the applicable DCGL and that any elevated areas are appropriately investigated and addressed. A sum of fractions approach can be used to assess the cumulative risk associated with multiple strata.

For example, a sum of fractions approach may entail calculating the sum of fractions for each sampling location/ strata, and then calculating the average sum of fractions for all the locations in each strata. The average sum of fractions for each strata are then summed to achieve a sum of fractions for the survey unit (assuming DCGLs for each strata are separately based on 0.25 mSv/y (25 mrem/y)). Alternatively, an analyst could calculate a sum of fractions for each borehole location (considering all strata) and average the sum of fractions across the borehole locations. In all cases, it is important to understand the basis for development of the DCGLs (thickness, depth, distribution, and area of residual radioactivity) for each stratum to ensure that risk is not underestimated. It is also important to understand the dose modeling assumptions with respect to the reuse of soil in the excavation (i.e., use of clean or slightly contaminated soils) to ensure the risk is not underestimated. In complex cases involving subsurface residual radioactivity, it is always prudent to calculate the final estimated dose for the compliance scenario(s) based on the final configuration and measured radionuclide concentrations of residual radioactivity at the site through dose modeling.2 As stated above, it is important to adequately assess the radiological characteristics of residual radioactivity in soil used for backfilling an excavation. The NRC staff expects survey of reused soil to receive the same rigor of characterization, scanning and sampling, as the area(s) of the site from which it originated. For each of the strata for which DCGLs are developed, radioanalytical results of backfill soil sampling should be analyzed and appropriately included in the sum of fractions calculations discussed in the preceding paragraph. In some cases, DCGLs developed for strata below the excavation bottom could be significantly greater than DCGLs developed for more shallow strata (i.e., strata associated with backfill soil). Therefore, it is important to ensure that the backfill soil concentrations are compared to the appropriate DCGL to ensure that the risk is not underestimated. In many cases, the backfill soil is scanned and sampled while being excavated or immediately after the soil is excavated and placed in a laydown area.

2 Although commonly used codes such as RESRAD-ONSITE only consider one average soil concentration as input to the code, the code can be run multiple times and the doses summed to assess the contributions of multiple strata.

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In some cases, it may be appropriate and more straightforward to derive a single set of DCGLs based on the entire thickness of residual radioactivity, rather than develop multiple sets of DCGLs for different strata thicknesses. However, sensitivity analyses may need to be conducted to ensure that the risk is not underestimated using this approach. For example, for radionuclides dominated by the external dose pathway, a similar DCGL may be developed for a small thickness of residual radioactivity (e.g., 15 cm) compared to a larger thickness of residual radioactivity, as residual radioactivity at depth may not contribute to the external dose pathway due to attenuation in the soil column. The concentration of the surface layer would then be appropriate to compare to the DCGL. If the plant pathway is important, soil concentrations down to 1 m may be important to dose (depending on the plant types and depth of roots).

When a mix of radionuclides are present, certain radionuclides may be transported to greater depths than others. DCGLs derived for all of the radionuclides based on the depth of residual radioactivity for one radionuclide may lead to underestimate of risk for other radionuclides. In these cases, care should be taken to ensure that residual radioactivity in the subsurface is adequately characterized to determine the lateral and vertical extent of contamination and elevated concentrations are appropriately considered (i.e., depth discrete surveys or sampling may be needed to capture heterogeneity in soil concentrations and ensure that elevated areas above DCGLs are not diluted by averaging concentrations with clean soils).

Excavated land areas at sites undergoing decommissioning will often require backfill soil.

Multiple options exist for backfill sources, such as offsite areas, nonimpacted areas on site, or impacted areas on site that have been appropriately surveyed.

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A Class 1 or 2 Survey Unit-Survey sidewalls (see B)

Class 1 Survey Unit Survey bottom of the excavation Strata 1 Strata 2 B Strata 3 Figure G.3 Sampling Strategies for Excavations (A) Use of MARSSIM Survey Design Approach for Excavation Bottom and Sidewalls [Plan View Map on top] and (B)

Depiction of Potential Options for Sampling into Excavation Sidewalls to Assess Residual Radioactivity in Remaining Soils (Various Strata for which DCGLs may be Derived are Depicted) [Cross-Section Map on bottom]

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Challenges Associated with Surveys of Open Excavations In some cases, excavation activities may be needed to remediate subsurface contamination to demonstrate compliance with release criteria. A technical presentation on challenges associated with subsurface surveys was provided at the July 14-15, 2021, subsurface soil survey workshop (Watson, 2021). In most cases, bottoms of open excavations remediated to remove soil above DCGLs to below the release criteria should be classified as Class 1 survey areas. Class 1 survey units should be 100 percent scanned per MARSSIM guidance, although exceptions and alternative approaches will be considered on a case-by-case basis until additional guidance can be developed for subsurface surveys. Challenges and lessons learned associated with remediation and radiological surveys associated with open excavations include the following and should be considered in project planning:

  • seasonal flooding
  • contingencies for delays due to weather and the creation of mud
  • groundwater intrusion (e.g., shallow ground water may need to be pumped to reach contamination creating a secondary waste stream)
  • flooding of buildings when pumps are removed
  • instability of sidewalls of excavation
  • inadequate excavation or sampling depth
  • inadequate sampling of backfill material (the decommissioning plan or license termination plan should include information regarding reuse of onsite materials or soils and associated survey)
  • lack of consideration of dose associated with reused soil
  • issues associated with survey of sidewalls in addition to excavation bottom
  • hard to access areas due to standing water, utility lines and active rail lines, and rock substrate, may need to be addressed by revising survey maps or using new technologies. Licensees can also use new technologies to reduce risk to workers during surveys.
  • lack of support of for non-standard technologies for survey
  • coordination with the NRC on timing of FSS to allow NRC to perform confirmatory studies G-17

G.3.2.2. Backfill from Nonimpacted Onsite and Offsite Areas Licensees have typically proposed to use backfill from non-impacted areas onsite or from offsite locations. If the licensee is assuming there is no added residual radioactivity in the backfill, support should be provided for this assumption (i.e., that the backfill soils do not contain residual radioactivity). Residual radioactivity, as defined in this NUREG series, includes radioactivity from all licensed and unlicensed sources used by the licensee, but excludes background radiation. If there is uncertainty that backfill soils are non-impacted, one potential method to support this assumption of no added residual radioactivity would be to use a two-sample statistical test such as a Scenario B type analysis to show indistinguishability from background, as described in Chapter 13 of NUREG-1505, Revision 1, A Nonparametric Statistical Methodology for the Design and Analysis of Final Status Decommissioning SurveysInterim Draft Report for Comment and Use, issued June 1998. See Section G.6 for more information about Scenario B. If Scenario B is used, the licensee should discuss appropriate values for the lower bound of the gray region (LBGR), and upper bound of the gray region (UBGR) (or width of the gray region) with the NRC. Although some form of radiological survey is expected, other approaches proposed by the licensee may also be acceptable.

G.3.2.3. Backfill from Impacted Onsite Areas Re-use of soil from radiologically impacted areas as backfill at a site undergoing decommissioning has been allowed by the NRC on a case-by-case basis. Licensees should continue to discuss proposed soil reuse plans with the NRC, as there are potentially complex issues associated with radiological measurement capabilities and site-specific dose assessments. The following guidance may assist licensees in developing reuse plans, though site-specific conditions may lead to additional issues for consideration.

If licensees plan to reuse soil from impacted areas onsite, characterization and radiological surveys should be completed to the rigor of an FSS. The MARSSIM survey unit classification (i.e., Class 1, 2, or 3) should be considered to ensure that the number of measurements and scanning coverage are adequate. Licensees should also consider the depth at which surface soil can be measured via surface scanning and adjust excavation methodologies accordingly (e.g., a Class 1 area should receive 100 percent surface scanning followed by systematic soil sampling). In this case, if the entire depth of reuse soil cannot be adequately surveyed it may be necessary to excavate and survey soil via lift depths consistent with surface soil dose modeling and instrument capabilities. Similarly, licensees should ensure that the sampling and scanning performed for Class 2 and 3 areas satisfy the survey objectives. Additionally, the effects of reusing soil onsite should be evaluated in the context of the final site dose assessment. The licensee should consider the final configuration of reused soil and may take credit for clean cover materials in estimating dose for residual radioactivity remaining at the site.

The reuse of impacted soil, depending on the methods used, may include intentional mixing of soil containing residual radioactivity. Licensees should refer to Volume 1, Section 15.13, of this NUREG for additional guidance on intentional mixing of soil to meet license termination criteria.

It is recommended that the licensee fully consider the implications of reusing impacted soil during the planning and DQO processes and discuss plans with the NRC as needed. Additional details regarding radiological surveys for reused soil are discussed in the following paragraphs.

Excavated soil may be segregated at a site, based on the level of residual radioactivity for potential reuse (e.g., impacted soil may be stockpiled for possible reuse as backfill or grading materials on a site). If the residual radioactivity includes radionuclides that can be measured in a scan survey (i.e., contaminants that are not hard-to-detect radionuclides and can be detected G-18

in a gamma detecting scan survey 3), then the licensee could use scanning as a first method to evaluate the soil. If scanning indicates that the soil contains residual radioactivity above background when it is expected to be free of residual radioactivity, or above release criteria when DCGLs are developed for reused soil, then the licensee should determine whether it is acceptable to reuse the soil at the site. In some cases, the soil may not be suitable for stockpiling and reuse4. The stockpiling of soil should be limited so that only scanned strata are excavated before the newly exposed soil is scanned. The thickness of the excavated lifts should be consistent with the scanning instrument capability and scanning coverage consistent with the survey unit classification. Alternatively, soil can be excavated, transported to a suitable laydown area, and then scanned in the laydown area similar to a FSS prior to stockpiling. In these cases, the thickness of soil in the laydown area for scanning should be consistent with the scanning instrument capability and survey unit classification (e.g., in a Class 1 area, the areal coverage should be 100 percent and the survey instrument should be able to detect residual radioactivity at the bottom of a soil layer in the laydown area). Stockpiled soil and materials from impacted areas that are heterogeneous may result in local areas of increased radioactivity levels. In these situations, it may be useful to rescan the soil and materials after emplacement to determine whether additional evaluations or spot remediation is warranted, especially if the soil is used for final grade material.

In some cases, licensees have used conveyor belt scanning and sorting systems to address excavated materials. Soil characteristics such as soil type, density, and moisture content, and measurement sensitivity are factors to consider in the DQO process when using these systems.

Also, as previously mentioned, soil reused from onsite areas should be characterized to the rigor of an FSS. DQOs should address the sampling and analysis needed to satisfy this in addition to scanning processes.

Maintaining a database of sampling data used to characterize stockpiled soil can be useful to demonstrate that the average residual radioactivity concentration in soil in the stockpile/backfill meets the release criteria. This may be especially useful if the eventual backfill is of considerable depth (e.g., > 1 meter). If more than one stockpile is being generated, then the licensee should take care to properly isolate the soil stockpiles from each other using appropriate barriers (e.g., jersey barriers, hay bales, silt screens, etc.). It could also be beneficial to verify that the surface of a stockpile hasnt been impacted by resuspended materials from decommissioning operations or adjacent stockpiles (e.g., by taking composite samples prior to using stockpiled materials for backfill).

In summary, when stockpiled soil is planned for use as backfill or grading material and remains onsite at the time of license termination, it is important to develop a survey plan that takes into account the characteristics of the soil, measurement or characterization methods to be used before and after emplacement, and applicable radiological criteria, which may be based on site-specific dose modeling.

Although DCGLs may have been developed to guide site remediation activities, it is also prudent to perform dose modeling using the final configuration and concentrations of residual radioactivity at the time of license termination.

3 See Appendix A of this document and Section 4.3.2 of MARSSIM, Revision 1 for additional information on use of surrogate radionuclides for hard-to-detect radionuclides.

4 Soil that is deemed unacceptable for reuse, may be disposed of in accordance with 10 CFR Part 20, Subpart K.

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G.3.3 Rubble, Debris, and Rocks Rubble, debris, and rocks can include naturally occurring rocks (either in place or in piles),

pieces of concrete or rubble from buildings that have been razed, sheet metal disposed of as trash, asphalt, fly ash, and similar material. The HSA and characterization surveys determine the volumetric extent and residual radioactivity concentration. If the materials are contaminated, they would be disposed of as radioactive waste. If the radioactivity is not substantially elevated, the licensee may evaluate the rubble, debris, and rocks as part of a larger survey unit. When these materials will be evaluated as part of a larger survey unit and when they are found on a relatively small fraction of the area of a survey unit, the volumetric soil DCGL should be used uniformly throughout the survey unit. However, the licensee should justify the reasonableness of modeling rocks and rubble as soil.

G.3.4 Paved Parking Lots, Roads, and Other Paved Areas The HSA and characterization surveys determine whether the residual radioactivity is on or near the surface of the paving and whether there are significant concentrations of residual radioactivity beneath the paving. If the residual radioactivity is primarily on top of the paving, then the licensee should take the measurements as if the area were normal soil. Depending on how large the paved area is, the licensee may include it as part of a larger survey unit, or it may be its own survey unit. If the residual radioactivity is primarily beneath the paving, the licensee should survey it as subsurface residual radioactivity, as discussed above.

G.3.5 Recontamination Finally, measures should be employed to avoid recontamination of previously remediated areas and the need for additional surveys or analyses to demonstrate compliance with release criteria.

Recontamination of cleaned survey units or areas on site can occur particularly at decommissioning sites where (i) decommissioning is conducted in phases, (ii) partial site releases are conducted, and (iii) remediation occurs over long timeframes. Contamination of cleaned areas from residual radioactivity in unremediated areas can occur due to environmental and human-induced transport such as through surface water runoff, air transport; and soil moving or demolition activities. Should NRC staff suspect recontamination of previously surveyed areas, additional surveys, dose modeling, or remediation may be necessary to demonstrate compliance with release criteria.

G.4 Surveys Associated with Multiple Radionuclides and Media In cases where residual radioactivity is present in multiple media and multiple radionuclides are present, it may be necessary to adjust the dose modeling approach used to derive DCGLs, as well as the survey design approach used to demonstrate compliance with radiological criteria for license termination to account for the cumulative dose from multiple radionuclides and media.

For example, consider a site with residual radioactivity present on the surfaces of a building, surface and subsurface soils, and nearby stream bed sediments. Multiple key radionuclides contributing significantly to the dose are present in each of the media and different key radionuclides are important for each of the media. Given the previous industrial use of the site and presence of residential areas in close proximity to the site, reasonably foreseeable future land use scenarios include industrial and residential use of the site within 100 years of license termination. Exposure to stream bed sediments may also occur due to recreational use of the stream such as swimming, boating, and fishing. Therefore, a recreational scenario may also be considered for the streambed sediments. DCGLs are typically derived for each of the G-20

contaminated media based on the release criteria (e.g., 0.25 mSv/y [25 mrem/y] for unrestricted release). However, given limitations in the dose modeling codes used to derive the DCGLs, receptors are typically assumed to be exposed to just one of each of the media. Likewise, given use of different DCGLs for each of the different media, the survey design does not automatically consider the contributions of multiple media in demonstrating compliance with release criteria.

With respect to consideration of multiple radionuclides, the sum of fractions approach can be used to account for multiple radionuclides in the same survey unit. Chapter 11 in NUREG-1505 (NRC, 1998) provides guidance and an example on how to use the sum of fractions approach to design the survey to determine the appropriate number of samples to demonstrate compliance with release criteria while ensuring decision criteria are met (e.g., acceptable Type I [false positive] and Type II [false negative] errors). However, this example is only applicable to a single survey unit type (e.g., soil, building, streambed sediments).

While surface and subsurface soils may have separate DCGLs, similar to the approach used for multiple radionuclides, surface and subsurface soils can be considered together when demonstrating compliance because they can be assumed to be located in the same survey unit(s). However, soil survey units are typically separate from building survey units and streambed sediment survey units. Therefore, if the demonstration of compliance is made for each media survey unit type based on cleanup levels derived at the compliance limit, if a member of the public could be exposed to multiple media and survey unit types in the same year, the annual release limit could be exceeded.

Licensees can take a conservative approach to addressing cumulative dose associated with each of the media by re-calculating the DCGLs for each of the media assuming a portion of the compliance limit (e.g., 45 percent of the 0.25 mSv/y [25 mrem/y] unrestricted release limit for (i) the soils, and (ii) building; and 10 percent for (iii) streambed sediments). It is important to note that re-calculation of the DCGL is necessary prior to designing the survey to ensure that the survey design is based on the correct width of the gray region (e.g., additional samples may be needed if the width of the gray region is smaller due to the lower DCGL or UBGR used in the survey design). While assuming both a residential and industrial scenario occur at the same site may be overly conservative with occupancy factors approaching 100 percent or greater, this would be an acceptable approach to consideration of cumulative dose associated with all sources/media and radionuclides.

Similar to methods used to consider elevated areas or hot spots (see Appendix I), the licensee may also develop DCGLs based on more reasonable exposure assumptions if a receptor is assumed to be exposed to multiple sources or media. For example, if the dose contributions from both the building and the soils are considered in demonstrating compliance, occupancy factors may be adjusted lower to account for the time spent in the building and time spent outside in an industrial scenario when developing DCGLs for the building and soils. Codes such as RESRAD-BUILD and RESRAD-ONSITE allow for occupancy factor parameter values to be adjusted such that the total time fraction is not greater than 1 (e.g., the residential farmer scenario assumes time fractions of 0.12 (outdoors) and 0.66 (indoors), and building occupancy scenario assumes a time fraction of 0.25 for a total time fraction greater than 1.0) . Other acceptable approaches for adding realism to the dose calculations may be available and should be discussed with the NRC staff early in the process.

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G.5 Consideration of Elevated Areas in Survey Designs Using the FSS approach, sites are typically broken up into a number of survey units without regard to the presence of elevated areas. In cases where multiple elevated areas exist in relatively close proximity to one another, the licensee should consider restructuring the survey units so that the elevated areas are evaluated in as few survey units as practical and which meet the guidance for survey unit size and have a shape similar to typical local property lots for housing or industry.5 This may be unnecessary if physical barriers exist such that it would be unlikely that future land use would result in a single occupant having potential exposure from all elevated areas. Alternatively, if the elevated areas do not contribute significantly to the potential exposure, it may be possible to simply show that the contributions of the elevated areas do not exceed the criteria for any of the adjoining survey units when considered as an additional source of exposure.

With regard to delineation of elevated areas, sampling data and scanning data are typically used to define the boundaries of the elevated area. When defining the boundaries of the elevated area, the area should not be intentionally established to incorporate multiple sample results that are significantly less than the DCGLs and which would, therefore, intentionally lower the average activity level of the elevated area. In cases where the survey unit boundaries split the elevated area, the entirety of the elevated area should be considered as being in the confines of both survey units or else the survey units should be reconfigured so that it exists in its entirety in just one survey unit. Other approaches may be proposed by the licensee that may be acceptable and should be discussed with the NRC staff early in the process.

When elevated areas are encountered, good housekeeping practices consistent with ALARA criteria should be considered. For example, if it is practical or cost-beneficial to remove a small area of elevated activity (e.g., a couple buckets of contaminated soil or caulk in a floor seam),

removal activities would likely be considered consistent with ALARA criteria. Minor removal activities would also help alleviate complications associated with consideration of elevated area contributions to dose and interpretation of anomalous survey results.

G.6 Additional Information and Example on Conducting Scenario B As noted in Section 2.4, the default assumption for FSSs is that the survey unit is considered contaminated above the limit (i.e., the null hypothesis is that the concentrations of residual radioactivity exceed the DCGLs). This assumption and null hypothesis are considered Scenario A. Typically, statistical tests are used to demonstrate that the average or median of the measurements in the survey unit are not above the DCGL, with the burden of proof on the Iicensee to show that the survey unit meets the release criteria. In special cases (i.e., exceptions) it may be appropriate to assume an alternative scenario in which the survey unit is considered clean (i.e., the null hypothesis is that the concentrations of residual radioactivity meet the release criterion). This alternative, Scenario B, is typically used when the cleanup levels are within the range of background variability, making it difficult to distinguish between the residual radioactivity and background below the cleanup levels. For example, sites with highly variable background and comparatively small DCGLs may desire to demonstrate that measurements in the survey unit are indistinguishable from measurements in reference areas. Other Scenario B applications may include the use of gross alpha or gross beta measurements, or nuclide-specific measurements when the nuclide is in background, similar to 5 Novel dose modeling approaches to addressing potential gross over-estimations in dose when considering multiple elevated areas is found in Appendix I.

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Scenario A. MARSSIM also points to a situation in which the release criteria are based on no added radioactivity above background or the radionuclide is not present in background. This results in an action level of zero added radioactivity and makes the selection of the LBGR in Scenario A difficult, if not impossible. Licensees considering the use of Scenario B for compliance with 10 CFR Part 20, Subpart E, are strongly encouraged to contact the NRC staff early in the planning process. In all cases, the licensee should discuss its plans with the NRC to determine the acceptability of using Scenario B, as well as determining appropriate values for the test parameters.

As a Scenario B example, this section describes five major steps to demonstrate indistinguishability from background when the Wilcoxon Rank Sum and Quantile tests are used:

(1) assess background variability using the Kruskal-Wallis test; (2) determine a concentration of radioactivity that is indistinguishable from background; (3) perform the Wilcoxon Rank Sum test; (4) perform the Quantile test if the survey unit passes the Wilcoxon Rank Sum test; and (5) perform the elevated measurement comparison test. Information on these methods is contained in Chapters 6, 7 and 13 of NUREG-1505, and summarized below. Scenario B simply denotes that the null hypothesis is defined as the survey unit meeting the release criteria.

Therefore, other applications of Scenario B are available and may be found acceptable if sufficient technical justification is provided. For example, NUREG-1575, Supplement 1, Multi-Agency Radiation Survey and Assessment of Materials and Equipment Manual (MARSAME), issued January 2009, and MARSSIM, Revision 2, discuss a scenario in which the LBGR is set to zero (or zero added radioactivity above background) and the UBGR is set to some background discrimination level.

Prior to performing statistical tests, measurements are made in several different background reference areas. The number of samples needed depends on the probability of Type I error ()

and Type II error () that are considered acceptable. The null hypothesis is that there is no significant variability between different background reference areas, so a Type I error would incorrectly conclude there is a significant difference between the background reference areas when there is no significant difference. Using the DQO process, a value of is selected for an acceptable frequency of a Type I error rate. Table 13.5 of NUREG-1505 provides results of calculations of the power of the F-test (parametric complement to the nonparametric Kruskal-Wallis test) as a function of the number of reference areas and the number of measurements in each reference area for values of 0.05, 0.10 and 0.20. Based on the results of these calculations, NUREG-1505, Chapter 13, indicates that 4 reference areas with between 10 and 20 measurements in each reference area should generally be adequate, and that an value of 0.10 is a reasonable default value. Other values, number of reference areas and number of measurements in each reference area could be found to be reasonable during the DQO process. For planning purposes, it may be convenient to select multiples of five measurements for performing the subsequent Quantile test because Table A.7a of NUREG-1505 contains values as multiples of five survey measurements.

In this Scenario B example, measurement data from 10 measurements in 4 reference areas are provided in Table G.1. The Kruskal-Wallis test is performed on this data to determine if significant variability exists in the background reference areas. The measurements for the reference areas are pooled and ranked, and then the sum of the ranks for the individual reference areas (Ri) and the mean measurement are determined. A test statistic (K) is obtained using NUREG-1505 equation 13-3 (see equation below), where N is the total number of measurements in all the reference areas i=1 to k reference areas; is the number of measurements in a given reference area; and is the sum of the ranks of the measurements in a given reference area:

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12 2

= (=1 ) 3( + 1) (G-1)

(+1)

The value of K is then compared with the critical value (Kc) found in Table 13.1 of NUREG-1505.

If K is greater than Kc, the null hypothesis is rejected, and it is concluded that there is significant variability between the reference areas. For the example data in Table G.1, the K value of 14.0 is compared to Kc for three reference areas (four reference areas minus one reference area).

The value of Kc is lower than K, which ranges from 11.3 (for an value of 0.01) to 4.6 (for an value of 0.2). Therefore, the null hypothesis can be rejected with high confidence. It is concluded that there is significant variability between the reference areas, which helps justify use of Scenario B for determining indistinguishability from background.

As mentioned above, an alternative to the Kruskal-Wallis test is the F-test to determine if variability between the means of potential reference areas is statistically significant. Chapter 13 of NUREG-1505 contains information on the use of the F-test for this purpose. Although the Kruskal-Wallis test (or F-test) is used to determine if it is appropriate to consider reference area variability in applying Scenario B, it may be acceptable to not conduct either test if it is decided that there are significant differences among the potential reference areas for a survey unit.

Thus, if it is considered appropriate to give background variability the benefit of the doubt, the Kruskal-Wallis and F-test test need not be conducted. Not conducting these tests is essentially the same as setting equal to 1.0.

2 Table G.1 Calculation of for the Example Data Measurements Measurement Ranks Measurements Squared Area 1 Area 2 Area 3 Area 4 Area 1 Area 2 Area 3 Area 4 Area 1 Area 2 Area 3 Area 4 1 0.27 1.04 2.45 3.77 6 13 27 39 0.07 1.08 6.00 14.21 2 1.87 0.39 0.34 2.63 20 9 8 31 3.50 0.15 0.12 6.92 3 0.97 2.07 3.06 4.05 10 23 37 40 0.94 4.28 9.36 16.40 4 1.01 -0.57 2.83 1.72 11 2 35 19 1.02 0.32 8.01 2.96 5 2.08 1.97 1.09 1.50 24 21 6 14 17 4.33 3.88 1.19 2.25 6 1.62 -0.22 0.26 2.47 18 3 5 29 2.62 0.05 0.07 6.10 7 0.30 1.39 2.80 1.42 7 15 34 16 0.09 1.93 7.84 2.02 8 1.98 0.05 2.77 2.47 22 4 533 33 28 3.92 0.00 7.67 6.10 9 2.18 -0.75 2.42 2.76 25 1 26 32 4.75 0.56 5.86 7.62 10 1.02 2.50 2.86 3.35 12 30 36 38 1.04 6.25 8.18 11.22 sum 13.30 7.87 20.88 26.14 155 121 255 289 22.28 18.50 54.30 75.80 average 1.33 0.79 2.09 2.61 ave: sqd 1.77 0.62 4.36 6.83 The next step is to determine the LBGR. The LBGR is the concentration level above background that may be considered distinguishable from background and can be established using a measure of the variability among the background reference areas. The mean square between reference areas, sb2, and mean square within reference areas, sw2, are used to calculate the component of the variance (2). These test parameters can be calculated using methods described in Chapter 13 of NUREG-1505 or can be found in analysis of variance

[ANOVA] output from commonly used statistical software (see Table G.2). The difference in concentration that is distinguishable from background may be expressed in terms of an G-24

appropriate multiple of . The equation below is based on equation 13-13 in NUREG-1505, where M is a multiplier selected using the DQO process and no is equal to the number of measurements per reference area when the number of measurements in each reference area is the same.

(2 2)

= 2 = M (G-2) 0 Equation 13-13 in NUREG-1505 can be used when the number of measurements in the reference areas are not the same. Using the DQO process, a value for M needs to be determined. Based on information contained in Section 13.4 of NUREG-1505, a value of 3 is characterized as a reasonable default. Depending on the distribution, other values may also be appropriate including lower factors of down to a value of zero added radioactivity for more stringent release criteria or when the radionuclide is not present in background. Using the ANOVA output in Table G-2, the LBGR = 3 x = 2.22 in this example. Note that the difference in means between reference areas 2 and 4 in Table G-1 is 1.82, which is consistent with the LBGR calculated based on 3 .

Table G.2 Analysis of Variance for Example Data Source of Sum of Degrees of Mean Square F Statistic Variation Squares Freedom Between Groups 19.56 3 6.52 6.69 Within Groups 35.08 36 0.97 Total 54.65 39 Section 9.5 of NUREG-1505 contains information on calculating the sample size for the Wilcoxon Rank Sum test under Scenario B, and is essentially the same method used in Scenario A. Also, the Wilcoxon Rank Sum test is performed similarly as it is used in Scenario A. The data for an example Wilcoxon Rank Sum test using Scenario B are shown in column A of Table G.3. In column B, the label R is inserted to denote a reference area measurement, and the label S to denote a survey unit measurement.

In Scenario B, the survey unit measurements are adjusted by subtracting the LBGR (e.g., 3 )

from each survey unit measurement, whereas in Scenario A the reference area measurements are adjusted. Column C of Table G.3 contains the adjusted data obtained by subtracting the LBGR (142) from the survey unit measurements. The ranks of the adjusted data in Column C are listed in Column D. Next, the adjusted survey unit measurements are ranked, which range from 1 to 24, since there is a total of 12 + 12 = 24 measurements. The sum of all the ranks is N(N + 1)/2 = (24)(25)/2 = 300. Column E contains only the ranks belonging to the adjusted survey unit measurements. The sum of the ranks of the adjusted survey unit data, Ws, is 194.5.

When using the DQO process to determine an acceptable probability of a Type I error, it should be noted that the total probability of a Type I error in assessing the survey unit data is the sum G-25

of the probabilities of Type I errors for both the Wilcoxon Rank Sum and Quantile tests.

Therefore, the acceptable probabilities of Type I errors in the Wilcoxon Rank Sum and Quantile tests is half the total acceptable probability. For this example, the acceptable total probability of a Type I error is set at 0.05, so WRS is 0.025. For comparing the statistic Ws with the critical value, Table A.4 of NUREG-1505 is used. However, for Scenario B, the meaning of m and n are the reverse of those in Scenario A. For Scenario B, m is the number of survey unit measurements and n is the number of reference area measurements in this table. From Table A.4 of NUREG-1505, for values of for the Wilcoxon Rank Sum test = 0.025 and n = m = 12, the critical value is 184. Because the sum of the adjusted survey unit ranks, 194.5, is greater than the critical value, 184, the null hypothesis that the survey unit concentrations do not exceed the LBGR is rejected (i.e., the site is determined to be dirty). In this Scenario B example, the true survey unit residual radioactivity is judged to be in excess of 142 above background.

The Quantile test is only used in Scenario B and if the Wilcoxon Rank Sum test fails to reject the null hypothesis. Whereas the Wilcoxon Rank Sum test is best at detecting excess radioactivity that is uniformly distributed, the Quantile test is intended to recognize excess radioactivity in only a fraction of the survey unit (i.e., higher concentrations of radioactivity in a fraction of the survey unit). Although the null hypothesis was rejected and the Quantile test need not be performed, for this Scenario B example the Quantile test is presented for illustrative purposes using the data in Table G.3. The data are adjusted and ranked in the same manner as in the Wilcoxon Rank Sum test, then the ranks are sorted, and the area associated with the rank is identified. The null hypothesis of the Quantile test is that there is no residual radioactivity above the LBGR in any part of the survey unit.

Table A.7 of NUREG-1505 provides for the Quantile test the critical value, k, of the largest r measurements for different values of n, the number of measurements from the survey unit, and m, the number of measurements from the reference area. Different sub-tables are provided in Table A.7 of NUREG-1505 for different values. The same rankings in Column D of Table G.3 for the Wilcoxon Rank Sum test can be used for the Quantile test. If k or more of the r largest measurements in the combined ranked data set are from the survey unit, the null hypothesis is rejected.

Columns F and G of Table G.3 show the sorted ranks of the adjusted data, and the location associated with each rank (i.e., R for reference area and S for survey unit). From Table A.7b of NUREG-1505, the closest entry to n = m = 12 is for n = m = 10, with corresponding values of r

= 7, k = 6 and = 0.029. Thus, the null hypothesis is rejected if six of the seven largest adjusted measurements are from the survey unit. From Table G.3, we find that only five of the seven largest adjusted measurements are from the survey unit, so the null hypothesis is not rejected based on the Quantile test. The values of n and m that were used are close to, but not equal to, the actual values, so the value will be different from that listed in the table. It is prudent to check a few other entries in Table A.7b of NUREG-1505 that are near the actual sample size.

Additionally, Chapter 7 in NUREG-1505 provides equations to calculate exact and approximate values of for the Quantile test as a function of n, m, k, and r.

It is recommended that an elevated measurement comparison is conducted, regardless of the outcome of the WRS and Quantile test. This consists of determining if any measurement in the remediated survey unit exceed a specified investigation level. If so, then additional investigation of the data is required to determine if there are elevated measurements that were not identified by the statistical tests.

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Table G.3 Scenario B WRS and Quantile Tests for Class 2 Interior Drywall Survey Unit A B C D E F G Location Survey Adjusted Sorted Associated 1 Data Area Ranks Unit Data Ranks with Sorted Ranks Rank 2 47 R 47 18 - 1 R 3 28 R 28 1 - 2 R 4 36 R 36 6 - 3 R 5 37 R 37 7 - 4.5 R 6 39 R 39 9.5 - 4.5 S 7 45 R 45 13 - 6 R 8 43 R 43 11 - 7 R 9 34 R 34 3 - 8 S 10 32 R 32 2 - 9.5 R 11 35 R 35 4.5 - 9.5 R 12 39 R 39 9.5 - 11 R 13 51 R 51 21 - 13 R 14 209 S 67 24 24 13 S 15 197 S 55 23 23 13 S 16 188 S 46 16 16 16 S 17 191 S 49 19 19 16 S 18 193 S 51 21 21 16 S 19 187 S 45 13 13 18 R 20 188 S 46 16 16 19 S 21 180 S 38 8 8 21 R 22 193 S 51 21 21 21 S 23 188 S 46 16 16 21 S 24 187 S 45 13 13 23 S 25 177 S 35 4.5 4.5 24 S 26 Sum= 300 194.5 (Measurements from the reference area and the survey unit are denoted by R and S, respectively)

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G.7 Integration of Dose Modeling and Radiological Surveys Pathway dose or risk modeling is oftentimes used to determine cleanup levels or DCGLs used as decision criteria in statistical tests discussed in Chapter 8 of MARSSIM and Appendix A.

Because DCGLs are an integral part of the survey design, model integration with the survey design is an important topic discussed in various sections of this guidance.

Assuming the pathway dose or risk modeling is representative of actual survey unit conditions, the survey design should be compatible with modeling used to derive cleanup levels or DCGLs.

Because the distribution, thickness, depth, and area of residual radioactivity are directly related to dose (or risk) in many commonly used decommissioning modeling codes, the FSS should be designed consistent with the modeling assumptions related to these key parameters to the extent practical. For example, if vertical heterogeneity is an issue and DCGLs are sensitive to the distribution of residual radioactivity within the soil column, DCGLs could be developed for different soil intervals (e.g., 0 to 5 cm and 5 to 15 cm). Depth discrete sampling may then be necessary for comparison against the DCGLs developed for each soil interval. For certain radionuclides and pathways (e.g., Cs-137 and external dose pathway), depth discrete sampling may also be important to ensure that higher concentration residual radioactivity located near the surface is not diluted with clean or cleaner radioactivity located deeper in the soil column, which could lead to an underestimate of risk6. Likewise, the thickness of residual radioactivity can also be important to risk. If residual radioactivity is located deeper in the soil column then assumed in the dose modeling, then the risk could be underestimated. Subsurface residual radioactivity should be adequately characterized and if present, appropriate methods should be developed to evaluate whether the site is clean (the MARSSIM methodology summarized in Appendix A was developed for surficial soils and building surfaces only; other methods may be necessary to make decisions regarding release for sites with subsurface or volumetric residual radioactivity).

Additionally, the dose and risk pathway modeling typically assume that residual radioactivity is relatively homogeneous across the model domain and most models only accept a single, average concentration for each radionuclide as input. If, in contrast, the survey unit is heterogeneous (e.g., spotty or elevated areas of radioactivity), the assumptions in the modeling may be violated, and the effectiveness of the statistical tests may be reduced. In these cases, the survey design team may need to consider other methods to mitigate or assess the impact of heterogeneous distributions such as more careful delineation of survey units or added emphasis and reliance on EMC tests. For Scenario B, the quantile test used in conjunction with the WRS test is also useful in evaluating whether a portion of the survey unit has unacceptable levels of residual radioactivity.

Finally, it is important to note that dose or risk modeling typically uses mean concentrations of radionuclides from survey unit measurements, while the Sign and WRS tests are tests on the median. While the median concentration is a good approximation for the mean concentration if the radionuclide distributions are symmetric (or not highly skewed), in some cases the mean concentration could exceed the median concentration. The average of the survey unit measurements (or the difference between the average survey unit and average reference area) 6 While most dose modeling codes only accept a single concentration, lateral and vertical heterogeneity can be represented by running the model more than once. For example, to represent vertical heterogeneity, initial concentrations at the surface (e.g., 0 to 5 cm) can be run separate from deeper residual radioactivity (e.g., 5 to 15 cm).

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should always be compared to the DCGL before a site is released (Scenario A) to help ensure that a site with residual radioactivity above the DCGL is not released.

G.8 References Barr, C.S. Development of Derived Concentration Guideline Levels (DCGLs) for Subsurface Residual Radioactivity. Presented at the NRC Subsurface Soil Surveys Public Workshop, July 14-15, 2021. U.S. NRC: Washington, DC. July 15, 2021. ADAMS Accession No. ML21208A218.

Electric Power Research Institute (EPRI). Connecticut Yankee Decommissioning Experience, Detailed Experiences 1996-2006. EPRI Report 1013511. EPRI: Palo Alto, CA. November 2006.

Nuclear Energy Institute (NEI). Guideline for the Management of Underground Piping and Tank Integrity. NEI 09-14, Rev. 3. NEI: Washington, DC. April 2013.

. Industry Groundwater Protection Initiative, Final Guidance Document, Rev. 1.

NEI 07-07, Rev. 1. NEI: Washington, DC. March 2019.

NRC. Inspection and Enforcement Circular No. 81-07, Control of Radioactively Contaminated Material. U.S. NRC: Washington, DC. May 14, 1981.

. Fuel Cycle Policy and Directive, FC 83-23, Termination of Byproduct, Source and Special Nuclear Material Licenses. U.S. NRC: Washington, DC. 1983.

. Information Notice No. 85-92, Surveys of Wastes Before Disposal from Nuclear Reactor Facilities. U.S. NRC: Washington, DC. December 2, 1985.

. Information Notice No. 88-22, Disposal of Sludge from Onsite Sewage Treatment Facilities at Nuclear Power Stations. U.S. NRC: Washington, DC. May 12, 1988.

. NUREG/CR-5512, Vol. 1, Residual Radioactive Contamination from Decommissioning: Technical Basis for Translating Contamination Levels to Annual Total Effective Dose Equivalent. U.S. NRC: Washington, DC. October 1992.

. Generic Environmental Impact Statement in Support of Rulemaking on Radiological Criteria for License Termination of NRC-Licensed Nuclear Facilities, NUREG-1496. U.S. NRC: Washington, DC. November 1996.

. NUREG-1496, Generic Environmental Impact Statement in Support of Rulemaking on Radiological Criteria for License Termination of NRC-Licensed Nuclear Facilities. Division of Regulatory Applications, Office of Nuclear Regulatory Research, U.S.

NRC: Washington, DC. July 1997.

. NUREG 1505, Revision 1. A Nonparametric Statistical Methodology for the Design and Analysis of Final Status Decommissioning SurveysInterim Draft Report for Comment and Use. U.S. NRC: Washington, DC. June 1998.

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. Draft Regulatory Guide DG-4006, Demonstrating Compliance with the Radiological Criteria for License Termination. Office of Nuclear Regulatory Research, U.S.

NRC: Washington, DC. August 1998.

. NUREG-1575, Rev. 1, Multi-Agency Radiation Survey and Site Investigation Manual (MARSSIM). EPA 402-R-97-016, Rev. 1; DOE/EH-0624, Rev. 1; U.S. Department of Defense; U.S. Department of Energy; U.S. Environmental Protection Agency; and the U.S.

NRC: Washington, DC. August 2000.

. Regulatory Issue Summary 2002-02, Lessons Learned Related to Recently Submitted Decommissioning Plans and License Termination Plans. U.S. NRC:

Washington, DC. January 2002.

. SECY-03-0069, Results of the License Termination Rule Analysis. Policy Issue Memorandum from W.D. Travers to the Commissioners. U.S. NRC: Washington, DC.

May 2, 2003.

. Regulatory Issue Summary 2004-08, Results of the License Termination Rule Analysis. U.S. NRC: Washington, DC. May 2004.

. NUREG/CR-7026, Application of Model Abstraction Techniques to Simulate Transport in Soils, NRC: Washington, D.C. March 2011.

. NUREG/CR-7021, A Subsurface Decision Model for Supporting Environmental Compliance, NRC: Washington, D.C. January 2012.

. Subsurface Soil Surveys Public Workshop Materials. U.S. NRC. Washington, DC. July 2021.

Watson, B. Survey Issues with Excavations from Recent Decommissionings. Presented at NRC Subsurface Soil Surveys Public Workshop, July 14-15, 2021. U.S. NRC:

Washington, DC. July 15, 2021. ADAMS Accession No. ML21208A222.

Yu et al., Subsurface DCGL: Effects of Thickness, Area, and Cover. Presented at the NRC Subsurface Soil Surveys Public Workshop, July 14-15, 2021. U.S. NRC: Washington, DC.

July 15, 2021. ADAMS Accession No. ML21208A219.

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APPENDIX H CRITERIA FOR CONDUCTING SCREENING DOSE MODELING EVALUATIONS

H.1 Introduction This appendix consists of the technical guidance for the use of the screening criteria, applicable to Decommissioning Groups 1-3. Section I.8 of Appendix I contains details of the references cited below.

This section pertains to NRC staffs review of a licensees demonstration of compliance with the dose criteria in 10 CFR Part 20, Standards for Protection against Radiation, Subpart E, Radiological Criteria for License Termination, using a screening approach dose analysis. The NRC staff should review the screening analysis using one or more of the currently available screening tools:

  • a lookup table for common beta- and gamma-emitting radionuclides for building surface residual radioactivity (Volume 63 of the Federal Register (FR), page 64,132 (63 FR 64132); November 18, 1998)
  • a lookup table for common radionuclides for soil surface residual radioactivity (64 FR 68395; December 7, 1999)
  • screening levels derived using DandD Version 2.1, or the most current version, for the specific radionuclide(s) and using the DandD codes default parameters Other tools for performing a screening analysis might become available in the future, or the NRC staff may modify current lookup tables or develop additional lookup tables (e.g., the NRC staff may develop lookup tables for the common alpha-emitters for building surfaces using modified parameter values in current versions of DandD1). In addition, the NRC staff may consider the use of other screening tools (e.g., other lookup tables or other conservative codes/models) after evaluating and comparing these screening tools with the current screening codes.

A licensee usually conducts a screening analysis for simple sites with building surface (i.e., nonvolumetric) and/or with surficial soil (approximately 15 centimeters (cm) (6 inches (in.))

residual radioactivity. The analysis usually employs simple and conservative models/codes and parameters, under generic exposure scenarios and default site conditions, to define the screening derived concentration guideline levels (DCGLs) equivalent to the dose criteria.

Because of the conservative nature of the screening analysis approach, the screening DCGLs are typically more restrictive than the site-specific DCGLs. A screening analysis may save licensees time and effort by reducing the amount of site characterization, modeling analysis, and reviews needed, versus those needed when using a site-specific analysis approach.

To review a screening analysis, the NRC staff first needs to make a generic assessment and evaluation of a licensees justification that the site is qualified for screening. In addition, the NRC staff should be familiar with the tools (e.g., models, codes, and calculations) and embedded assumptions used to derive the screening DCGLs. This section addresses the major 1 One sensitive parameter identified by the NRC staff for the building occupancy exposure scenario is the resuspension factor. NUREG-1720, Re-Evaluation of the Indoor Resuspension Factor for the Screening Analysis of the Building Occupancy Scenario for NRCs License Termination Rule, issued June 2002, documents the NRC staffs evaluation of the default resuspension factor used in DandD Version 1. If site conditions are consistent with the assumptions made in NUREG-1720, the recommended resuspension factor and parameter distribution in NUREG-1720 can be used with minimal justification in a site-specific analysis.

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issues that the NRC staff may encounter in the generic screening analysis reviews and includes recommendations of approaches for addressing and resolving these issues.

H.1.1 Issues in Performing Screening Analysis The NRC staff may encounter issues with the screening analysis, including (1) the definition of screening and the transition from a screening to a site-specific analysis, (2) the qualification of the site for screening, in terms of site physical conditions and compatibility with the modeling codes assumptions and default parameters, and (3) the acceptable screening tools (e.g., code, lookup tables), approaches, and parameters that the staff can use to translate the dose into equivalent screening concentration levels. Each one of these issues is the subject of discussion in the following subsections:

H.1.2 Screening Definition and Approaches for the Transition from Screening to Site-Specific Analysis The NRC staff may encounter some inconsistencies about the definition of the term screening in dose analysis, which may cause confusion about the transition from a screening to a site-specific analysis. These inconsistencies become more apparent when dividing screening approaches into multiple levels (NCRP Report No. 123, Screening Models for Releases of Radionuclides to Atmosphere, Surface Water, and Ground, dated January 22, 1996; NCRP Report No. 129, Recommended Screening Limits for Contaminated Surface Soil and Review of Factors Relevant to Site-Specific Studies, dated January 29, 1999). In some cases, screening and site-specific terms are mixed, and the term site-specific screening is used (NUREG/CR-5512, Volume 1, Residual Radioactive Contamination from Decommissioning:

Technical Basis for Translating Contamination Levels to Annual Total Effective Dose Equivalent, issued October 1992). In certain cases, screening is categorized by the type of models used (e.g., simple and conservative models versus more advanced and complex models) and the extent of data and information needed to support the dose analysis.

Within the context of NUREG-1757, the NRC staff should consider the definition of screening as the process of developing DCGLs at a site using either (1) the NRCs lookup tables in 63 FR 64132 and 64 FR 68395, or (2) the latest version of the DandD code developed by the NRC to perform the generic screening analysis.

When licensees either (1) select other approaches or models for the dose analysis or (2) modify the DandD code default parameters, exposure scenarios, or pathways, the NRC staff considers licensees to be performing site-specific analyses. With regard to footnote a of Table H.1, the use of values of the fraction of removable surface contamination other than 0.1 or 1.0 (as described in the footnote) in the DandD code is considered a site-specific analysis and the staff should use Section 5.2 to review it.

While there is no requirement that licensees consider the use of screening criteria, they should recognize the advantages and disadvantages of selecting a screening approach for demonstrating compliance with the dose criteria. Section 2.6 of this volume discusses the merits of using screening versus using site-specific analysis.

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H.1.3 Qualification of the Site for Screening The NRC staff should be aware that a screening analysis, for demonstrating compliance with the dose criteria in 10 CFR Part 20, Subpart E, may not be applicable for certain sites because of the status of contaminants (e.g., location and distribution of radionuclides), or because of site-specific physical conditions. Therefore, the staff should assess the source characteristics (e.g., spatial distribution of residual radioactivity) to ensure consistency with the source configuration assumptions in the DandD code. Further, the NRC may determine that there could be conditions at the specific site that cannot be handled by the simple screening model, because of the complex nature of the site or because of the simple conceptual model in the DandD screening code.

When using the screening approach for demonstrating compliance with the dose criteria in 10 CFR Part 20, Subpart E, licensees need to demonstrate that the particular site conditions (e.g., physical and source conditions) are compatible and consistent with the DandD model assumptions (NUREG/CR-5512, Volume 1). In addition, the default parameters, exposure scenarios, and pathways must also be used in the screening dose analysis. Therefore, reviewers should examine the site conceptual model, the generic source term characteristics, and other attributes of the site to ensure that it is qualified for screening.

The NRC staff should verify that the following site conditions exist for each of the residual radioactivity conditions:

  • Building Surface Residual Radioactivity:

o The residual radioactivity on building surfaces (e.g., walls, floors, ceilings) should be surficial and nonvolumetric (e.g., 10 millimeters (mm) (0.39 inches (in.)) of penetration).

o Residual radioactivity on surfaces is mostly fixed (not loose), with the fraction of loose (removable) residual radioactivity no greater than 10 percent of the total surface activity. Note, for cases when the fraction of removable contamination is undetermined or higher than 0.1, licensees may assume, that 100 percent of surface contamination is removable, and therefore the screening values should be decreased by a factor of 10 (see footnote a to Table H.1).

o The screening criteria are not being applied to surfaces such as buried structures (e.g., drainage or sewer pipes) or equipment within the building without adequate justification; such structures, buried surfaces, and clearance of equipment should be treated on a case-by-case basis.

  • Surface Soil Residual Radioactivity:

o The initial residual radioactivity (after decommissioning) is contained in the top layer of the surface soil (e.g., approximately 15 cm (6.0 in.)).

o Subsurface soil (e.g., approximately 15 cm (6.0 in.) or greater below the surface) in the unsaturated zone and the groundwater are initially free of residual radioactivity.

o The vertical saturated hydraulic conductivity at the specific site is greater than the infiltration rate (e.g., there is no ponding or surface runoff).

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Questions have also been raised about the appropriateness of using a screening analysis at sites with contaminated areas larger than the DandD Version 1 default cultivated area (e.g., 2,400 square meters (m2) (25,800 square feet (ft2)). Initially, the NRC staff evaluated the effect of a large, contaminated area on the derived screening dose and determined that this effect is trivial for sites with the dominant dose arising from direct exposure or inhalation. As modeled by DandD with its default parameter set, this effect could be appreciable for sites with a significant dose contribution associated with the ingestion pathway (specifically ingestion associated with the drinking water and irrigation pathways). The staff determined that, for sites with contaminated areas of 6,000-7,200 m2 (64,600-77,500 ft2), the dose may be underestimated under worst-case conditions by a factor of 2 to 3. However, further staff analysis showed that, because of the conservative assumptions of the DandD code, it is more likely that the derived dose (based on the use of other codes or the use of a site-specific analysis) would be far less than the derived dose using these default conditions. Therefore, for sites with areas larger than 7,200 m2 (77,500 ft2), the change in actual risk due to this effect is not appreciable. In summary, assuming that the site is qualified for screening based on the above-listed criteria, the NRC would accept the screening approach for sites with areas larger than the default cultivated area (i.e., 2,400 m2 (25,800 ft2)).

It should be noted that the NRC staff should also evaluate complex site conditions that may disqualify the site for screening. Examples of such complex site conditions may include highly fractured formation, karst conditions, extensive surface water contamination, and highly nonhomogeneous distribution of residual radioactivity. Therefore, reviewers should ensure that the site meets the definition of a simple site to qualify for screening (see Section 1.2 of this document for additional details).

H.1.4 Acceptable Screening Tools In the past, it may not have been clear what screening tools the NRC has determined to be acceptable. Some may believe that using simple, common codes (other than DandD), with their deterministic default parameters, may be acceptable to derive the desired screening derived concentration guideline levels (DCGLs). Others may believe that the use of any lookup tables published by certain scientific committees or authorities may be used to convert concentration levels directly into doses for purposes of complying with Subpart E. Questions have also been raised on the use of the DandD code for screening, particularly whether modifying input default parameters is acceptable for screening.

The NRC staff should accept, for screening analyses, the following currently available screening tools:

  • a lookup table (Table H.1) for common beta- and gamma-emitting radionuclides for building-surface residual radioactivity (63 FR 64132; November 18, 1998)
  • a lookup table (Table H.2) for common radionuclides for soil surface residual radioactivity (64 FR 68395; December 7, 1999)
  • screening levels derived using the latest version of DandD for the specific radionuclide and using code default parameters and parameter ranges The screening values in Tables H.1 and H.2 are intended for single radionuclides. For radionuclides in mixtures, the sum of fractions rule should be used (see Section 2.7 of this volume). Table H.1 values that provide screening values for beta and gamma emitters H-4

associated with building surfaces were derived using DandD Version 1 and its deterministic, default parameter set, before the development of probabilistic DandD Version 2. Table H.2 values were derived using a version of DandD that was similar to and the predecessor of DandD Version 2. The screening values in Table H.2 are based on the selection of the 90th percentile of the output dose distribution for each specific radionuclide or radionuclide with the specific decay chain. Behavior parameters were set at the mean of the distribution of the assumed critical group. The metabolic parameters were set either at the Reference Man or at the mean of the distribution for an average human.

For a radionuclide with its progeny present at equilibrium, the +C values of Table H.2 should be interpreted carefully. As described in footnote c to Table H.2, these +C values are concentrations of the parent radionuclide only but account for dose contributions from the complete chain of progeny in equilibrium with the parent radionuclide. For example, U-238+C lists the soil screening value as 18.5 Bq/kg (0.5 pCi/g). This means that it also assumed the presence of 18.5 Bq/kg (0.5 pCi/g) of U-234, 18.5 Bq/kg (0.5 pCi/g) of Th-230, and so forth.

The current NRC staff position is to limit screening to its lookup tables or the execution of the latest version of DandD code with the default parameters and distributions. As indicated above, the NRC staff may develop additional lookup tables or modify the screening tables based on refining certain sensitive parameters in the future. For example, NUREG-1720 documents the NRC staffs evaluation of the default resuspension factor used in DandD Version 1.0.

NUREG-1720 provides specific recommendations related to the use of the resuspension factor in screening analyses for the building occupancy exposure scenario. If site conditions are consistent with assumptions in NUREG-1720 (with respect to activities and scenarios, ventilation conditions, and low removable fractions at the time of decommissioning), the NRC staff has determined that it is acceptable to use the NUREG-1720-recommended resuspension factor value or parameter distribution with minimal justification. However, until such time as the NUREG-1720-recommended parameter value or distribution is used in developing an updated screening table for the building occupancy exposure scenario, the NRC staff considers use of the NUREG-1720 values a site-specific analysis (because the underlying assumptions in NUREG-1720 must be verified by the licensee before use of the recommended resuspension factor values). The NUREG-1720-recommended resuspension factor is approximately 10 times less than the default value used in DandD Version 1 and would, thus, lead to significantly higher (or less restrictive) screening values for radionuclides dominated by the inhalation pathway (e.g., alpha-emitting radionuclides), had such screening values been published in 63 FR 641322.

ORISE (2017) provides screening values for sites potentially contaminated with discrete sources of Ra-226 developed using the DandD code and three exposure scenarios including (i) industrial building occupancy, (ii) residential building occupancy, for residual radioactivity on building surfaces; and (iii) resident farmer for residual radioactivity associated with soils. Site-specific dose modeling is required for sites with residual radioactivity in the groundwater or with site conditions that are otherwise inconsistent with the dose modeling assumptions. The screening values are stated to be inappropriate for use for other types of Ra-226 contaminated decommissioning sites.

2 The NRC staff only published beta- and gamma-emitting radionuclide screening values in 63 FR 64132, while indicating it was continuing its assessment of screening approaches for sites with alpha-emitting radionuclides.

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Table H.1 Acceptable License Termination Screening Values of Common Radionuclides for Building-Surface Contamination Radionuclide Symbol Acceptable Screening Levelsa for Unrestricted Release (dpm/100 cm2)b Hydrogen-3 (Tritium) 3H 120000000 Carbon-14 14C 3700000 Sodium-22 22Na 9500 Sulfur-35 35S 13000000 Chlorine-36 36Cl 500000 Manganese-54 54Mn 32000 Iron-55 55Fe 4500000 Cobalt-60 60Co 7100 Nickel-63 63Ni 1800000 Strontium-90 90Sr 8700 Technetium-99 99Tc 1300000 Iodine-129 129I 35000 Cesium-137 137Cs 28000 Iridium-192 192Ir 74000 Notes:

a. Screening levels are based on the assumption that the fraction of removable surface contamination is equal to 0.1. For cases when the fraction of removable contamination is undetermined or higher than 0.1, users may assume for screening purposes that 100 percent of surface contamination is removable, and therefore the screening levels should be decreased by a factor of 10. Users may calculate site-specific levels using available data on the fraction of removable contamination in DandD.3
b. Units are dpm/100 cm2. One dpm is equivalent to 0.0167Bq. To convert to units of Bq/m2, multiply each value by 1.67. The screening values represent surface concentrations of individual radionuclides deemed to be in compliance with the 0.25 mSv/y (25 mrem/y) unrestricted release dose limit in 10 CFR 20.1402. For radionuclides in a mixture, the sum of fractions rule applies; see Section 2.7 of this volume. A description of sum of fractions is also provided in note 4, of footnote 3, to 10 CFR Part 20, Appendix B. An unofficial copy of the note is available on the NRC website at https://www.nrc.gov/reading-rm/doc-collections/cfr/part020/appb/footnotes.html.
c. NUREG/CR-5512, Vol. 3, Table 5.19 (see values in column labeled Pcrit = 0.90), contains screening values for additional radionuclides not found in Table H.1. The Table 5.19 screening values are also acceptable for use provided the underlying assumptions of the screening approach are met (65 FR 37186).

3 The original footnote to this table, published in 1998 (63 FR 64132), referenced use of DandD Version 1, which was the version of the code used to develop the screening values in this table. However, after publication of the screening values in the FR in 1998, probabilistic DandD Version 2 was developed and was used to create the screening values reported in Table H.2. Therefore, licensees should use more current versions of DandD (Version 2 and later) to develop screening values.

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Table H.2 Screening Valuesa (pCi/g) of Common Radionuclides for Soil Surface Contamination Levels Radionuclide Symbol Surface Soil Screening Valuesb Hydrogen-3 3H 110 Carbon-14 14C 12 Sodium-22 22Na 4.3 Sulfur-35 35S 270 Chlorine-36 36Cl 0.36 Calcium-45 45Ca 57 Scandium-46 46Sc 15 Manganese-54 54Mn 15 Iron-55 55Fe 10000 Cobalt-57 57Co 150 Cobalt-60 60Co 3.8 Nickel-59 59Ni 5500 Nickel-63 63Ni 2100 Strontium-90 90Sr 1.7 Niobium-94 94Nb 5.8 Technetium-99 99Tc 19 Iodine-129 129I 0.5 Cesium-134 134Cs 5.7 Cesium-137 137Cs 11 Europium-152 152Eu 8.7 Europium-154 154Eu 8 Iridium-192 192Ir 41 Lead-210 210Pb 0.9 Radium-226 226Ra 0.7 Radium-226+Cc 226Ra+C 0.6 Actinium-227 227Ac 0.5 Actinium-227+C 227Ac+C 0.5 Thorium-228 228Th 4.7 H-7

Table H.2 Screening Values (pCi/g) of Common Radionuclides for Soil Surface Contamination Levels (cont.)

Radionuclide Symbol Surface Soil Screening Valuesa,b Thorium-228+Cc 228Th+C 4.7 Thorium-230 230Th 1.8 Thorium-230+C 230Th+C 0.6 Thorium-232 232Th 1.1 Thorium-232+C 232Th+C 1.1 Protactinium-231 231Pa 0.3 Protactinium-231+C 231Pa+C 0.3 Uranium-234 234U 13 Uranium-235 235U 8 Uranium-235+C 235U+C 0.29 Uranium-238 238U 14 Uranium-238+C 238U+C 0.5 Plutonium-238 238Pu 2.5 Plutonium-239 239Pu 2.3 Plutonium-241 241Pu 72 Americium-241 241Am 2.1 Curium-242 242Cm 160 Curium-243 243Cm 3.2 Notes:

a. These values represent surficial surface soil concentrations of individual radionuclides that would be deemed in compliance with the 0.25 mSv/y (25 mrem/y) unrestricted release dose limit in 10 CFR 20.1402. For radionuclides in a mixture, the sum of fractions rule applies; see Section 2.7 of this volume. A description of sum of fractions is also provided in note 4, of footnote 3, to 10 CFR Part 20, Appendix B. An unofficial copy of the note is available on the NRC website at https://www.nrc.gov/reading-rm/doc-collections/cfr/part020/appb/footnotes.html.
b. Screening values are in units of (pCi/g) equivalent to 0.25 mSv/y (25 mrem/y). To convert from pCi/g to units of Bq/kg, divide each value by 0.027. These values were derived using DandD screening methodology (NUREG/CR-5512, Volume 3 (NRC 1999)). They were derived based on selection of the 90th percentile of the output dose distribution for each specific radionuclide (or radionuclide with the specific decay chain). Behavioral parameters were set at the mean of the distribution of the assumed critical group. The metabolic parameters were set at Reference Man or at the mean of the distribution for an average human.
c. Plus Chain (+C) indicates a value for a radionuclide with its decay progeny present in equilibrium. The values are concentrations of the parent radionuclide but account for contributions from the complete chain of progeny in equilibrium with the parent radionuclide (NUREG/CR-5512, Volumes 1, 2, and 3).
d. NUREG/CR-5512, Vol. 3, Table 6.91 (see values in column labeled Pcrit = 0.10), contains screening values for additional radionuclides not found in Table H.2. The Table 6.91 screening values are also acceptable for use provided the underlying assumptions of the screening approach are met (65 FR 37186).

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H.2 References Code of Federal Regulations (CFR). 10 CFR Part 20, Standards for Protection against Radiation.

NRC. Residual Radioactive Contamination from Decommissioning: Technical Basis for Translating Contamination Levels to Annual Total Effective Dose Equivalent, NUREG/CR-5512, Volume 1. NRC: Washington, DC. 1992.

. Residual Radioactive Contamination from Decommissioning, Users Manual, DandD Version 2.1, NUREG/CR-5512, Vol. 2. NRC: Washington, DC. 2001.

. Residual Radioactive Contamination from Decommissioning, Parameter Analysis, Draft Report for Comment, NUREG/CR-5512, Vol. 3. NRC: Washington, DC. 1999.

. Supplemental Information on the Implementation of the Final Rule on Radiological Criteria for License Termination, Federal Register: Vol 63, p. 64132. November 18, 1998.

. Supplemental Information on the Implementation of the Final Rule on Radiological Criteria for License Termination, Federal Register: Vol 64, p. 68395. December 7, 1999.

. Use of Screening Values to Demonstrate Compliance with the Final Rule on Radiological Criteria for License Termination, Federal Register: Vol 65, p. 37186. June 13, 2000.

. Re-Evaluation of the Indoor Resuspension Factor for the Screening Analysis of the Building Occupancy Scenario for NRCs License Termiantion RuleDraft Report for Comment, NUREG-1720. NRC: Washington, DC. 2002.

National Council on Radiation Protection and Measurements (NCRP). Screening Models for Releases of Radionuclides to the Atmosphere, Surface Water, and Ground, NCRP Report No. 123. NCRP: Bethesda, MD. 1996.

. Recommended Screening Limits for Contaminated Surface Soil and Review of Factors Relevant to Site-Specific Studies, NCRP Report No. 129. NCRP: Bethesda, MD.

1999.

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APPENDIX I TECHNICAL BASIS FOR SITE-SPECIFIC DOSE MODELING EVALUATIONS

I.1 Introduction This appendix consists of the technical guidance for the use of site-specific dose modeling, applicable to Decommissioning Groups 4-7.

I.1.1 Background On July 21, 1997, the NRC published a final rule on Radiological Criteria for License Termination, in the Federal Register (62 FR 39058), which was incorporated as Subpart E, Radiological Criteria for License Termination, to 10 CFR Part 20, Standards for Protection against Radiation. In 1998, the NRC staff developed a draft regulatory guide (DG),

Demonstrating Compliance with the Radiological Criteria for License Termination (DG-4006)

(NRC, 1998e), and a draft document, NUREG-1549, Decision Methods for Dose Assessment to Comply with Radiological Criteria for License Termination (NRC, 1998d), in support of the final rule. In addition, the staff developed a screening code DandD for demonstrating compliance with the dose criteria in 10 CFR Part 20, Subpart E.

On July 8, 1998, the Commission approved publication of the draft guidance, DG-4006, the draft NUREG-1549, and the DandD screening code for interim use for a 2-year period (i.e., from July 8, 1998, through July 7, 2000), in Staff Requirements SECY-98-051Guidance in Support of Final Rule on Radiological Criteria for License Termination, dated July 8, 1998 (NRC, 1998c). In addition, the Commission directed the staff to develop an SRP for decommissioning and provide the Commission with a timeline for developing it, maintain a dialogue with the public during the interim period, address areas of excessive conservatism, particularly in the DandD screening code, develop a more user-friendly format for the guidance, and use a probabilistic approach to calculate the TEDE to the average member of the critical group (NRC, 1998c).

The NRC staff published NUREG-1727, NMSS Decommissioning Standard Review Plan, in September 2000. Chapter 5 of the SRP (which is incorporated into Chapter 5 of this volume) addresses the NRC staff review of the licensees dose modeling to demonstrate compliance with the criteria in 10 CFR Part 20, Subpart E. The NRC staff developed Appendix C of the SRP (Appendix I of this volume) as a technical information support document for performing staff evaluations of the licensees dose modeling. It presents detailed technical approaches, methodologies, criteria, and guidance to the staff reviewing dose modeling for demonstrating compliance with the dose criteria in 10 CFR Part 20, Subpart E. To develop Appendix C of the SRP, the NRC used an iterative process with the public, which included licensees, Federal agencies, States, and other interested individuals. To support this process, the staff conducted seven public workshops and gave several presentations to national and international professional groups, stakeholders, the Interagency Steering Committee on Radiation Standards, and the Conference of Radiation Control Program Directors, as well as to the NRCs Advisory Committee on Nuclear Waste. In addition, the NRC posted the draft Appendix C (formerly the Technical Basis Document) on the NRCs website and invited interested individuals to comment.

Since the publication of the license termination rule (LTR), the NRC staff has tested the DandD code for complex sites and addressed the issue of excessive conservatism in the DandD code.

In addition, the NRC developed a new probabilistic DandD code (i.e., DandD Version 2.1) to reduce the excessively conservative approach in the initial version. Further, the staff developed the RESRAD-ONSITE and RESRAD-BUILD probabilistic codes for site-specific analysis, which also responds to the Commissions direction to use a probabilistic approach to calculate the I-1

TEDE to the average member of the critical group. Later, RESRAD-OFFSITE was developed and also included probabilistic capabilities.

Licensees using probabilistic dose modeling should use the peak of the mean dose (see Section I.7.3.2 from Appendix I of this volume) for demonstrating compliance with 10 CFR Part 20, Subpart E. Similar to all regulatory guidance, this NUREG contains one approach acceptable to the NRC staff for determining compliance with the regulations using probabilistic analyses. Use of mean of the peaks is also acceptable for demonstrating compliance. If the mean of the peaks dose is significantly higher than the peak of the mean dose, risk dilution may be an issue in the probabilistic model.

Consult Appendix Q for more information on the potential for and impacts of risk dilution. To use any probabilistic approach to calculate DCGLs, the licensee should discuss its approach with the NRC staff.

I.1.2 Brief Description and Scope This section is divided into the following different topic areas, as summarized below.

  • Section I.2 presents NRC approaches for reviewing the conceptual representation of the distribution and release of residual radioactivity from soils and building materials. This section describes areas of review for conceptualization and representation of the source and source term in dose modeling used to demonstrate compliance with radiological criteria for license termination
  • Section I.3 focuses on areas of review and criteria for modifying the two generic exposure scenarios used in screening-level analyses: modifications to (1) the resident farmer and (2) the building occupancy exposure scenarios. Section I.3, and Appendices L and M, discuss the type of information a licensee should provide to support the modification of default (screening) exposure scenarios. Section I.3 also presents approaches for establishing site-specific exposure scenarios and pathways reflecting the activities and behaviors of identified critical group(s), based on current and reasonably foreseeable future land use, site restrictions, and other physical conditions associated with a decommissioning site.
  • Section I.4 provides approaches for developing conceptual and mathematical models in dose modeling. This section presents approaches for collecting and using characterization data to develop conceptual and mathematical models of the site and issues associated with model simplification and abstraction. This section also includes information on the underlying conceptual models in the DandD and RESRAD family of codes and associated limitations.
  • Section I.5 presents approaches and criteria for NRC staff acceptance of computer codes/models. This section discusses review aspects pertaining to specifications, testing, verification, documentation, and QA/QC of the licensees codes and models.

This section also addresses reviews applicable to embedded numerical models for the source term, fate and transport, and biosphere (exposure) modeling.

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  • Section I.6 describes approaches for the selection and modification of input parameters used in dose modeling.
  • Section I.7 provides information on evaluating uncertainty and identifying parameters most important to dose in analyses used to demonstrate compliance with LTR criteria.
  • Section I.8 compiles the references used throughout the appendix.

I.2 Source and Source Term Abstraction1 I.2.1 Introduction Source abstraction is the process of developing a conceptual representation of the residual radioactivity present at a site, focusing on the geometry and distribution of contamination in the environment. Typically, the radiological conditions at a site undergoing decommissioning are relatively complex. Source abstraction is necessary to allow the detailed radiological characterization of the site to be incorporated into the mathematical and computer models that are used to demonstrate compliance with the dose-based criteria in 10 CFR Part 20, Subpart E.

The abstraction process involves generalizing the radiological characteristics across the site to produce a simplified representation, which should facilitate the modeling of radiological impacts.

This guidance makes a distinction between parameters that characterize the source, including parameters related to the concentration of radionuclides, thickness, and area of residual radioactivity; and those parameters that characterize the release of radioactivity in the environment (i.e., that help define the source term). The source term considers parameters that define the source concentrations and configuration of residual radioactivity, as well as the release mechanism (e.g., solubility-controlled, desorption, or diffusion-limited release).

Although source and source term abstraction are a necessary part of the dose modeling process, the licensee should take care to ensure that the conceptual representation of the source and source term developed in the abstraction process are not oversimplified in a manner that results in the underestimation of potential radiological impacts.

As discussed in Chapter 5 of this volume, source configuration and source term abstraction serve as the starting point for the dose modeling process. The conceptual abstraction of the source and source term is combined with (1) the physical characteristics of the site, and (2) characteristics of the average member of the critical group to develop a conceptual model for the site being studied. Thus, the conceptual model also provides a representation of the natural environment through which radioactivity can be transported, as well as applicable exposure scenarios and pathways to members of the public who may be exposed to the radioactivity.

The conceptual model can be used to determine acceptability of use of computer codes that 1 Source abstraction or configuration takes into consideration the geometry of the source (e.g., areal extent and thickness), as well as the distribution of residual radioactivity within the extent of contamination (e.g., homogenous versus nonhomogeneous). On the other hand, the source term characterizes the release rate of radionuclides from the source zone. The source term is a function of the inventory, physical and chemical characteristics of the contaminated materials, and surrounding environment, as well as the release mechanism (e.g., solubility-controlled, desorption, or diffusion-limited release). RESRAD-ONSITE, RESRAD-BUILD, and DandD have built-in release mechanisms and models, while RESRAD-OFFSITE offers several additional options to define the source term. Note that the working definition of source term in this volume is slightly different than the definition of source term found in the NRC glossary. The NRC glossary definition of source term, which is specific to accidents involving radioactive materials, indicates the following: types and amounts of radioactive or hazardous material released to the environment following an accident.

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contain their own built-in conceptual models with respect to how residual radioactivity is transported in the environment and how humans can be exposed to radioactivity.

Volume 1 of NUREG-1757 and Chapter 4 of this volume discuss the information the licensee is expected to provide about the existing radiological characterization of the site. The licensee should describe the types, levels, and extent of residual radioactivity in contaminated materials at the site. This should include residual radioactivity in all media (including buildings, systems and components, and equipment that will remain after license termination; surface and subsurface soil; and surface water and groundwater). The source configuration and source term abstraction should be based on the characterization of the radiological status (e.g., historical site assessment; records of leakage or disposal). The licensee should explicitly relate the information provided in the discussion of radiological status of the site with its assumptions on the source configuration in dose modeling. The reviewer should be able to clearly interpret the relationship.

Generally, in the source abstraction process, the licensee may focus on several specific elements, which include the following:

  • The licensee should identify the radionuclides of concern, taken directly from the description of the sites radiological status. The licensee should identify the radionuclides based on the pre-remediation radiological status. It should include all radionuclides potentially present at the site, so that their presence or absence may be verified during the FSS, except as noted in Chapter 4, Facility Radiation Surveys, and Section 3.3, Insignificant Radionuclides and Exposure Pathways, of this volume.
  • The licensee should describe the physical/chemical form(s) of the contaminated media anticipated at the time of FSS and site release. The licensee should indicate whether the residual radioactivity will be limited to building surfaces or surface soil, or both, or whether the residual radioactivity will involve other media, such as subsurface soil, debris or waste materials (e.g., sludge, slag, tailings), or ground and surface water.

Information on the physical/chemical form(s) of the contaminated media will also help determine whether source term assumptions are appropriate.

  • The licensee should delineate the spatial extent of the residual radioactivity anticipated at the time of FSS and site release. The delineation of the spatial extent should include descriptions of (1) the areal extent of radionuclides throughout the site and (2) the vertical extent of soil residual radioactivity of radionuclides below the ground surface.

The delineation of spatial extent and depth should establish the source areas and volumes. Source areas and volumes may differ for individual radionuclides.

  • The licensee should define the distribution of each radionuclide throughout the delineated source areas and volumes anticipated at the time of FSS and site release.

The distribution of a radionuclide through the source should be defined in terms of representative volumetric or areal concentrations. In addition, for volumetrically contaminated soil, the licensee may provide an estimate of total activity of each radionuclide.

  • The licensee should define sources in surface water or groundwater, if any, based on environmental monitoring and sampling of aquifers and surface water bodies. A site with groundwater or surface water contamination may be categorized as complex and may I-4

require more advanced dose modeling analysis (see Appendix F for additional information on surface water and groundwater characterization).

In the source abstraction process, the licensee should identify the radionuclides of concern and have sufficient information to determine if residual radioactivity is surficial and relatively homogeneous. Depending on the dose modeling approach, the licensee may or may not need to address the other elements, as discussed in more detail later in this section.

I.2.2 Issues Associated with Source Term Abstraction The level of effort that a licensee expends to develop a conceptualization of the source and source term should be commensurate with the complexity of the site and the licensees approach to demonstrating compliance with the release criterion (i.e., screening versus site-specific). Also, the focus should be on the source and source term characteristics anticipated to exist at the site at the time of FSS and release, after any planned remediation.

If a licensee plans to use the screening DCGLs published by the NRC in the Federal Register (see Appendix H), a licensee should only have to identify the radionuclides present at the site and demonstrate that the conditions at the site meet the prerequisites for using the screening values (i.e., residual radioactivity is limited to building surfaces or approximately the uppermost 15 centimeters (cm) (6 inches (in.)) of surface soil and show there is no contamination of groundwater or surface water2), as discussed further in Section I.2.3 of this appendix.

If a licensee anticipates that residual radioactivity will be limited to building surfaces or surface soils at the time of the FSS but considers the published DCGLs overly restrictive, it may develop site-specific DCGLs. In this case, the licensee would most likely have to delineate the anticipated areal extent and depth of residual radioactivity. However, the licensee would not have to discuss the anticipated spatial variability of radionuclide concentrations within the anticipated area of residual radioactivity in developing the DCGL, because variability is considered in the survey design and in EMC tests, as discussed in more detail below.3 The licensee should provide a site-specific dose assessment if the residual radioactivity is not limited to building surfaces or surface soil. In this case, the licensee would have to delineate the spatial extent (laterally and vertically) of the residual radioactivity and discuss the spatial variability of the physical, chemical, and radiological characteristics of the contaminated media.

Ideally, the source characteristics at a site would be relatively uniform, justifying simplified source abstraction. However, this is often not the case. Issues may arise when the residual radioactivity projected at a site at the time of release is inconsistent with the ideal case. These issues may include the following:

2 If surface water or groundwater are contaminated, it may still be possible to use screening values if the dose contributions from the residual radioactivity in the surface water or groundwater are separately considered.

3 It is important to note that if insignificant radionuclides or modified DCGLs (e.g., surrogate DCGLs) are developed to account for multiple radionuclides, it would be important for the licensee to discuss spatial variability in radionuclide ratios used to develop surrogate relationships to ensure that the modified DCGLs do not lead to an underestimate of risk. However, this information is unnecessary for the development of site-specific DCGLs. See Appendix A, and Section 4.3.2 of MARSSIM, Revision 1, for more information on the use of surrogate radionuclides.

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(1) Spatial extent o limited areal extent of residual radioactivity o irregular areal shape o varying depth of residual radioactivity in soil (2) Spatial variability o nonuniform distribution of radioactivity throughout a site o limited areas of relatively elevated radionuclide concentrations o multiple noncontiguous areas of residual radioactivity o nonuniform physical and chemical characteristics The sections below discuss approaches to addressing these technical issues.

I.2.3 Approach to Source Abstraction Source abstraction data needs will depend on the approach used to demonstrate compliance with radiological criteria for license termination presented in the licensees DP. Generally, the licensee will use one of the two following approaches to dose modeling:

(1) Develop DCGLs for each radionuclide that would lead to a dose at the release criterion, and then demonstrate through the FSS that median residual radioactivity concentrations at the site are equal to or below the DCGLs with a certain specified level of confidence.

(2) Assess dose associated with the actual distribution of residual radioactivity at the site to determine whether the residual radioactivity will result in a dose equal to or below the dose-based release criterion.

In the first approach, the licensee intends to demonstrate at the time of the FSS that residual radionuclide concentrations across the site are below a prespecified concentration limit with some prespecified degree of confidence. The design of the FSS would be based on the proposed DCGLs, in accordance with NUREG-1575, Multi-Agency Radiological Survey and Site Investigation Manual (MARSSIM) (NUREG, 2000a). The MARSSIM process considers variability in determining the number of samples needed to demonstrate compliance with the radiological criteria for license termination. DCGLEMCs are used to account for the dose contributions of smaller elevated areas of residual radioactivity. Knowledge about the characteristics (e.g., area and thickness) of elevated areas may also assist with the development of DCGLEMCs.

In the second approach, the licensee intends to assess potential radiation doses that may result from specified levels of radioactive material. The contaminated material may not be limited to building surfaces or surface soils but may include contaminated subsurface soil, debris, and waste. The licensees dose modeling should demonstrate that the residual radioactivity should not result in radiation doses in excess of applicable regulatory limits. Most likely, this modeling I-6

approach would require that the licensee incorporate information on both the spatial extent and spatial variability of radioactivity in the source abstraction.

Table I.1 summarizes source abstraction information needs, depending on the licensees dose modeling approach and whether the licensee is providing screening or site-specific analyses.

This table can serve as an index for the reviewer of the licensees source abstraction.

Table I.1 Summary of Source Abstraction Information Needs for Two Types of Dose-Modeling Approaches (Screening versus Site-Specific)

Approach to Screening Site-Specific Demonstrating Compliance DCGLs No source term abstraction is Delineate proposed lateral and necessary beyond radionuclide vertical extent of residual identification. radioactivity. Smaller areas of elevated activity can be (assume unit radionuclide considered through DCGLEMCs.

concentrations to calculate DCGLs (i.e., to calculate what (assume unit radionuclide concentration leads to a dose at concentrations to calculate DCGLs the limit)) (i.e., to calculate what concentration leads to a dose at the limit))

Dose Modeling Use actual concentrations with Site-specific source abstraction DandD Version 2 (or later incorporates spatial extent and version) and ensure that residual variability.

radioactivity is surficial and spatial variability is minimal.

I.2.3.1 Dose Modeling Approach One: Develop DCGLs The MARSSIM approach, as documented in NUREG-1575 (NRC, 2000a) and discussed in Chapter 4 of this volume, requires that a licensee establish a set of DCGLs before conducting an FSS. In fact, the design of the FSS should be based on the identified DCGLs. DCGL is defined in MARSSIM as:

a derived, radionuclide-specific activity concentration within a survey unit corresponding to the release criterionDCGLs are derived from activity/dose relationships through various exposure pathway scenarios.

The DCGLW is the concentration of a radionuclide that, if distributed uniformly across a survey unit, would result in an estimated dose equal to the applicable dose limit. The DCGL EMC is the concentration of a radionuclide that, if distributed uniformly across a smaller limited area within a survey unit, would result in an estimated dose equal to the applicable dose limit. DCGLEMC is specific to the size of the area for which it is derived.

Two approaches are possible for developing DCGLs: screening and site-specific analysis.

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SCREENING DCGLs The NRC has published radionuclide-specific screening DCGLs in the Federal Register for residual building-surface radioactivity and residual surface-soil radioactivity (see Appendix H, Table H-1 and H-2; or Table 5.19 and 6.91 in NUREG/CR-5512, Volume 3, for additional radionuclides). The DCGLs in the Federal Register are intended to be concentrations that, if distributed uniformly across a building or soil surface, would individually result in a dose equal to the dose criterion. The licensee may adopt these screening DCGLs without additional dose modeling, if the site is suitable for screening analysis. Alternatively, the licensee may use the DandD computer code to develop screening DCGLs. The licensee would use the code to determine the dose attributable to a unit concentration of a radionuclide and scale the result to determine the DCGLW for the radionuclide. Either of these methods for identifying screening DCGLs requires the licensee (1) to identify the radionuclides of concern for the site and (2) to demonstrate that the source and model screening assumptions are satisfied. Thus, this approach requires essentially no source abstraction.

Typically, before designing an FSS, the licensee identifies a DCGLEMC for each radionuclide for smaller limited areas of radioactivity (e.g., for areas between sampling locations). However, the underlying assumption for use of screening analyses is that the residual radioactivity is homogeneous. By default, DandD calculates dose based on the assumption that the size of the contaminated area is unlimited. Although DandD provides an option to allow the user to enter a limited area of residual radioactivity, due to the simplistic manner in which DandD corrects the dose for smaller, limited areas of residual radioactivity, DandD is not ideally suited for calculating DCGLEMCs. Therefore, it is recommended that licensees use other codes or approaches to develop DCGLEMC values, if there is a need for allowing higher concentrations above the DCGLW in smaller areas between sampling locations. These would be considered site-specific analyses in that they would not be using the DandD code with the default screening values. Section I.3.3.3.5 of this appendix contains additional information on this topic.

If the licensee can show that residual radioactivity is relatively homogeneous and concentrations of radionuclides averaged over relatively small exposure areas (areas that are consistent with the exposure pathway assumptions in DandD) are less than the screening values or DandD-derived DCGLs, then development of DCGLEMC values should not be needed. Consult the DandD online help for more information on calculating the average concentrations for use in the DandD code.

SITE-SPECIFIC DCGLs The licensee may choose to identify site-specific DCGLs if (1) the site conditions are not consistent with screening criteria or (2) the licensee believes the screening DCGLs are unnecessarily restrictive. As defined in MARSSIM, Rev. 1 (2000a) the licensee may derive site-specific DCGLs from activity or dose relationships through various exposure pathway scenarios.

Site-specific in this context may refer to the selection of conceptual models/computer models, physical (site) input parameter values, or behavioral or metabolic input parameter values.

These aspects of site-specific analyses are discussed in other sections of this document. Site-specific may also refer to the source or source term abstraction.

From the MARSSIM perspective, identifying a site-specific DCGLW still begins with assuming a uniform radionuclide concentration across some source area (building surface) or volume of I-8

surface soil4. The site-specific DCGLW for a particular radionuclide may be identified by evaluating the dose resulting from a unit concentration and then scaling the result to the dose limit. Spatial variability of the radionuclide concentration within the area or volume is not evaluated in calculating the DCGLs but is taken into account in the statistical analysis of the data collected during the FSS. When developing the site-specific DCGLs, the licensee should, however, take the spatial extent into account, including the horizontal and vertical extent of residual radioactivity. While some examples in MARSSIM use the top 15 cm of soil as surface soil, for the purpose of compliance measurements, licensees should consider the actual depth of residual radioactivity which should be consistent with the dose model used to generate site DCGLs. Soil sampling should be representative of the chosen surface soil depth, and surface scanning methods should be capable of detecting residual radioactivity within the entire depth of surface soil. Accordingly, scan MDCs should be determined based upon the depth of contaminated soil associated with the dose model used to generate surface soil DCGLs. If subsurface residual radioactivity is present, dose modeling may be conducted for both surface and subsurface soils and DCGLs developed for each. In these cases, the MARSSIM methodology will need to be supplemented or an alternative methodology will need to be developed to demonstrate compliance with radiological criteria for license termination.

Through the FSS, the licensee would have to demonstrate that the DCGLW is satisfied within the specified exposure area for each survey unit. The licensee should consider the area assumed in the dose modeling for consistency against the area of the survey unit and the area over which concentrations are averaged for comparison against DCGLs. The licensee should also develop DCGLEMC values for smaller areas within the larger area of residual radioactivity as part of the FSS design (e.g., develop DCGLEMC values for smaller areas in the survey unit between soil sampling locations that may be missed by sampling). If it is certain that the residual radionuclide concentration is limited to a specific lateral extent, the licensee may also incorporate the area of residual radioactivity into the identification of DCGLs. Computer modeling codes, such as RESRAD-ONSITE, allow the user to directly specify the area of residual radioactivity and adjust the dose as a function of the specified area in a unique manner for each exposure pathway.

The NRC recommends that the licensee use dose modeling to calculate DCGLEMCs. For example, it can use RESRAD-ONSITE and RESRAD-BUILD to calculate dose for the larger area of concern (i.e., the entire contaminated area) and the dose from the smaller areas of potentially elevated concentration, separately. It can then use dose modeling results to calculate the cleanup levels for the larger area of concern and for the smaller elevated areas (DCGLW and DCGLEMCs, respectively). If only a single elevated area is present, the EMC is acceptable if Equation I-1 is met (modified from MARSSIM, Equation 8.2)

+ <1 (I-1)

Where = the average residual radioactivity concentration for all sample points in the survey unit.

4 MARSSIM only considers surface soils; residual radioactivity in subsurface soils is not addressed by MARSSIM.

Surface soil is defined in MARSSIM as the top layer of soil onsite that is available for direct exposure, growing plants, resuspension of particles for inhalation, and mixing from human disturbances.

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In cases where there is more than one elevated area, a separate term should be included in the calculation for each area of elevated activity. The unity rule is satisfied when radionuclide mixtures yield a combined fractional concentration limit that is less than or equal to one. In situations where there is more than one radionuclide at a single source (elevated area or hot spot), the sum of the individual ratios also cannot be greater than or equal to 1. The sum of fractions rule applies in situations where there are multiple radionuclides and sources. For example, consider a site where residual radioactivity is present in two elevated areas or hot spots, areas A and B, as well as in a larger area of concern. The concentrations of radionuclides 1, 2, and 3, in elevated or hot spot areas A and B, and the larger area, W, must all be considered and the sum of fractions must be less than or equal to 1 (see Equation 2). Note that the general area concentrations CXW for each radionuclide, x=1, 2, or 3, can be subtracted from the elevated area concentrations, CXA and CXB for each radionuclide, in Equation (I-2) consistent with Equation 1.

1 2 3 1 2 3

+ + + + + +

1 2 3 1 2 3 (I-2) 1 2 3

+ + 1 1 2 3 The following examples present different options for considering the contributions of multiple elevated areas within a single larger area. In general, two or more smaller areas modeled independently and combined will result in a higher dose then if contaminant concentrations are averaged across a single area. Although the higher dose may be considered conservative, it is often unrealistic, because it may assume, for example, that an individual spends all of his or her time in multiple locations simultaneously. The examples below demonstrate how the assumptions made about the residual radioactivity on a site can affect whether or not the site passes the cleanup criteria.

Example 1Base Case For this example, a survey unit of 10,000 m2 is uniformly contaminated with americium (Am)-241 at a concentration of 9 pCi/g. The DCGL is 10 pCi/g. Because the concentration (9 pCi/g) is below the DCGL, the survey unit will pass.

  • A 10,000 m2 survey unit with uniform Am-241 contamination (no elevated areas)
  • Am-241 concentration = 9 pCi/g
  • DCGLW = 10 pCi/g 9

= 0.9 0.9 < 1.0 Survey unit passes cleanup criteria 10 Example 2Single Elevated Area in the Survey Unit Example 2 considers a 10,000 m2 survey unit with a single 300 m2 elevated area contaminated with Am-241 at a concentration of 9 pCi/g. The DCGLW for the site is 10 pCi/g and the DCGLEMC for the elevated area (300 m2) is 12. As expected, the smaller contaminated area I-10

within the larger survey unit results in a smaller sum of fraction (0.75), which is less than 1.

Therefore, the survey unit with these characteristics also passes the cleanup criteria.

  • A 10,000 m2 survey unit with no contamination except for a 300 m2 elevated area
  • Am-241 concentration = 9 pCi/g
  • DCGLW = 10 pCi/g
  • DCGLEMC = 12 9

= 0.75 0.75 < 1.0 Survey unit passes cleanup criteria 12 Example 3Multiple Elevated Areas or Hot Spots in a Single Survey Unit As shown in Example 3, evaluating multiple smaller elevated areas often results in a higher dose than if the areas were combined and modeled as a single area. Instead of the single 300 m2 elevated area used in Example 2, Example 3 evaluates three 100 m2 elevated areas within the same 10,000 m2 survey unit. The DCGLW for the site is still 10 pCi/g, and the DCGLEMC (100 m2) is 18.6 pCi/g. While if a single 100 m2 elevated area was present, the dose criterion could be easily met, summing the dose contributions of the three individual elevated areas together results in a value greater than 1 and therefore, exceeds the cleanup criteria.

  • A 10,000 m2 survey unit with no contamination except for 3 X 100 m2 elevated areas
  • Am-241 concentration = 9 pCi/g
  • DCGLW = 10 pCi/g
  • DCGLEMC = 18.6 A 100 m2 elevated area:

9

= 0.48 18.6 Combining the results of each of the three 100 m2 elevated areas:

9 9 9

+ + = 1.45 18.6 18.6 18.6 I-11

In this case the higher dose is unrealistic, because it assumes that a receptor spends all of his or her time on each contaminated area simultaneously. The licensee can determine a more reasonable estimate of potential dose by combining the individual elevated areas into a single larger elevated area within the survey unit, as in Example 2.

In addition to specifying a limited area of residual radioactivity in developing the site-specific DCGLs for soil, the licensee should appropriately represent the vertical extent of residual radioactivity within the area. The screening DCGLs and the DandD code assume that residual radioactivity is contained within the uppermost 15 cm of soil. If the licensee intends to leave residual radioactivity at depths below approximately 15 cm, the calculation of the DCGLW should reflect a greater thickness of residual radioactivity. Otherwise, leaving residual radioactivity below 15 cm may not be acceptable.

For subsurface residual radioactivity (i.e., residual radioactivity at depths greater than approximately 15 cm), the NRC staff should evaluate whether the licensee has reviewed existing historical site data (including previous processes or practices) and site characterization data to establish an adequate conceptual model of the subsurface source specifically about the horizontal and vertical extent of residual radioactivity. The licensee should evaluate lateral and vertical trends of variation in concentration for each specific radionuclide. Because certain radionuclides have higher mobility than others, radionuclide ratios may not be maintained as constant across subsurface soil. In other words, radionuclide concentrations within the unsaturated zone may vary, depending on the original source location and the time since the source existed. The NRC staff should evaluate whether the licensee has evaluated the physical and chemical properties of the source and the attenuating properties of the subsurface materials to assess the potential for radionuclide leaching and transport. In this context, the reviewer should evaluate the selected physical parameters and conceptual model of the site against the actual subsurface hydrostratigraphy to evaluate the acceptability of the parameter selections if found to be important to dose. The reviewer should also consider (1) heterogeneity in subsurface soils, and (2) the depth to the water table if found to be important to dose.

If the thickness of residual radioactivity that will remain at a site is generally uniform across the site, the licensee may choose to use an upper bounding value for modeling the thickness. The NRC reviewer should evaluate the representative thickness value proposed by the licensee to ensure that the value selected does not lead to a significant underestimate of dose, particularly for sites where the vadose thickness is quite variable. For example, given the variable depth to the water table, a licensee may propose to use an area-weighted approach to assign the vadose zone thickness in dose modeling. However, if the timing of peak dose from groundwater dependent pathways is important to the compliance demonstration, then representation of the area with a thinner vadose zone may be important, for example.

If appropriate, the licensee should provide maps and cross sections detailing the proposed lateral and vertical extent of residual radioactivity left on the site.

I.2.3.2 Dose Modeling Approach Two: Assess Dose Based on Actual Concentrations An alternative objective that a licensee may have for performing and submitting dose modeling may be to assess doses attributable to specific quantities of radioactive material. Although the development of DCGLs focuses on the determination of radionuclide concentrations corresponding to a specified dose, the dose assessment objective focuses on the determination of doses corresponding to specified radionuclide concentrations.

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In this situation, the licensee should give much more attention to the source abstraction and address all elements of the source term abstraction:

  • identify the radionuclides of concern
  • delineate the spatial extent of residual radioactivity
  • represent the spatial variability of residual radioactivity
  • incorporate the spatial variability of physical and chemical characteristics of the contaminated media The licensee should focus on the distribution of radioactive material expected to be present at the time of FSS and subsequent site release. The licensee may assess doses attributable to existing radiological conditions at the site, if it can demonstrate that the existing radiological conditions reasonably bound conditions expected at the FSS, from a dose perspective.

The first two elements of source abstractionradionuclides of concern and spatial extentwere considered in the discussion of source abstraction for the development of DCGLs. Spatial variability was not considered, since it is statistically evaluated as part of the FSS. If the dose modeling approach using actual radionuclide concentrations is used, however, spatial variability should be factored into the source abstraction before modeling.

Assuming that the licensee has identified the radionuclides of concern and delineated the spatial extent of residual radioactivity, it should project the residual radionuclide concentration distribution and total residual radionuclide inventory across the site. The licensee should tie this projection directly to the characterization of existing radiological conditions at the site. The site may then be divided into relatively large areas that are radiologically distinct, based on radionuclide concentration or depth of residual radioactivity. The licensee should statistically demonstrate that the radionuclide concentrations or depth within an area may be relatively uniform, taking into account the spatial distribution of the data. Similarly, within the larger areas, the licensee should statistically delineate relatively small areas of projected elevated radionuclide concentrations or increased depth. (The licensee should discuss the reason for leaving the elevated concentrations in place as residual radioactivity.)

When complete, the licensees source term abstraction should define a site divided into relatively large areas of statistically uniform radionuclide concentrations and residual radioactivity depth. Within these areas may be relatively small areas of elevated concentration or increased depth. Assuming that the physical and chemical conditions across the site are relatively uniform, the licensee may use this source abstraction for modeling and proceed with the dose assessment.

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The following is a suggested approach:

  • Consider each relatively large area independently and initially ignore the relatively small elevated areas within each large area.
  • Assess dose based on the properties of a large area, taking the areal extent into account.
  • Repeat the dose assessment but assume an essentially infinite areal extent. The specific approach will depend on the computer modeling code used. This should quantify the impact of dividing the site into artificial modeling areas.
  • Assess dose attributable to each limited area of elevated concentration, assuming no residual radioactivity exists outside the limited area. This may then be combined with the dose attributable to the surrounding larger area, to assess the impact of leaving the elevated concentrations.

In some cases, it may not be practical to separate a site into areas with relatively uniform radionuclide concentrations; sometimes areas to be evaluated will have nonuniform distributions of concentrations. In such cases, for performing the second step above, there may be a question about what statistical value best represents the radionuclide concentration for the large area. Log-normal distributions occur frequently in nature and are not unexpected when surveying contaminated sites. For log-normal distributions, the geometric mean is often used as a descriptor of the distribution. However, the geometric mean concentration should not be used as the average value for the source concentration for dose calculations, as use of the geometric mean could lead to a significant underestimate of the dose compared to use of the arithmetic mean. Because (1) the dose rate is proportional to radionuclide concentration, (2) it is reasonable to assume that the receptor spends an equal amount of time in each area of the site, and (3) each characterization data point represents an equal area, use of the arithmetic mean (and not the geometric mean) is more technically defensible for calculating average concentrations for use in dose modeling (second step above). If samples are not taken randomly or systematically (and thus data points represent unequal areas), weighted means may be appropriate, with application of weighting factors consistent with the assumptions of receptor exposures.

The above discussion does not specifically address the determination of relatively significant large or small areas. This designation will depend on the areal assumptions underlying the computer modeling code used. For example, the DandD code considers the area of cultivation to be uniformly contaminated and irrigated. The area of cultivation depends on the cultivation requirements defined by the specific exposure scenario. While DandD allows the user to specify either (1) an unlimited area of residual radioactivity consistent with the underlying conceptual model, or (2) smaller areas of residual radioactivity, the relationship between area and dose is defined in a simplistic manner. Furthermore, the area of residual radioactivity is not treated as an uncertain parameter and, therefore, an automated sensitivity analysis on the area of residual radioactivity is not available. With respect to the RESRAD family of codes, ingestion pathway doses are largely dependent on the area of the source; however, the impact of area on dose varies depending on the specific pathway being considered. For example, soil and plant I-14

ingestion doses scale directly with areas up to 1,000 m2,5 while animal pathway doses scale directly with areas up to 20,000 m2. The licensee should discuss and justify the designation of relatively large and relatively small areas, based on the computer code used. The licensee can provide additional information (e.g., results of alternative scenarios evaluating the sensitivity of the dose modeling results to area) to lend more support for the compliance demonstration.

As discussed above, the licensee may also have to consider the impact of multiple areas of elevated concentration within a single larger area. In general, modeling two small areas independently and combining the results of the two dose assessments should result in a higher dose than if the two areas were combined and modeled as a single area (see Examples 1-3 in Section I.2.3.1). The higher dose may be unrealistic if it assumes that the receptor location relative to each contaminated area is such that the dose is maximized from each contaminated area independently. For a more reasonable estimate of potential dose, these smaller areas may be combined into a single larger area, if the concentrations within the smaller areas are comparable (e.g., see Example 3 in Section I.2.3.1). If this is not the case, then the licensee may model each smaller area individually and modify the exposure scenario and critical group assumptions for each area (e.g., time spent on each area) and combine the results.

The example illustrated in Figure I.1 presents an acceptable method for considering the contributions of multiple elevated areas or hot spots within a larger contaminated area, when the concentrations of the elevated areas are variable with each other, as well as with the larger area of residual radioactivity. Consider a site with two relatively small areas of elevated radioactivity in comparison to levels of radioactivity over a much larger area. Because most dose modeling codes, including DandD and RESRAD-ONSITE, assume that the receptor is located in the center of the contaminated area, the conceptual model in Figure I.1 depicts overlapping contaminated areas. However, to avoid overestimating the dose when the contaminated areas are overlaid and summed, the larger contaminated area is run first with an average concentration of 3 pCi/g. Next, the second most elevated area or hot spot, which has an average concentration of 10 pCi/g, is simulated with a concentration of 3 pCi/g less than the average concentration of 10 pCi/g (or with a value of 7 pCi/g), so as not to double-count the activity in the larger area to which it is summed. The second most elevated area or hot spot is assumed to be 200 m2 instead of 100 m2, to account for the fact that 100 m2 of the simulated contaminated area overlaps the most elevated 100 m2 area that is considered separately.

Using this approach, the second most elevated area or hot spot with an average concentration of 10 pCi/g is conceptually assumed to surround the most elevated 100 m2 area, (see yellow doughnut surrounding the orange, most elevated, hot spot in Figure I.1). This approach is acceptable, as it is impossible for the receptor to occupy two different 100 m2 areas at the same time, and assuming that the receptor is located in the center of the hottest area on site for the entire exposure period is conservative. Finally, the most elevated area or hot spot, which has an average concentration of 16 pCi/g, is simulated with a concentration 10 pCi/g less than the average concentration of 16 pCi/g (or with a value of 6 pCi/g), so as not to double-count the activity assigned in the larger area simulation of 3 pCi/g, and not double-count the activity assigned in the second most elevated area simulation of 7 pCi/g. Although the geometry and locations of the elevated areas or hot spots differ in the conceptual model versus the actual 5 When a value of -1 is input into the field for the contaminated fraction for plant food, and the size of the contaminated zone is equal to or greater than 1,000 m2, RESRAD-ONSITE assumes that 50 percent of the crops consumed by the receptor come from a garden grown in contaminated soil (i.e., no more than 50 percent of the produce comes from the contaminated garden and 2,000 m2 is needed to support 100-percent home grown produce ingestion rates). For areas less than 1,000 m2, RESRAD-ONSITE and RESRAD-OFFSITE linearly scale the consumption rates of contaminated produce down from 50 percent for 1,000 m2 areas to 0 percent for 0 m2 areas. Contaminated fraction values between 0 and 1 can otherwise be specified as a user input.

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configuration depicted in Figure I.1, the assumed geometry and elevated area location tends to overestimate the dose with the receptor standing directly on top of the hottest contaminated area on the site and in relatively close proximity to the second most contaminated area on site.

Depending on the actual size and geometry of the elevated areas being simulated, this method may produce overly conservative results. If less conservative methods are needed to demonstrate compliance, the licensee may propose alternative methods that will require approval by NRC reviewers on a case-by-case basis.

Figure I.2 illustrates how vertical heterogeneity can be considered in dose modeling. In the example, variability in concentration for the first 5 cm of soil and the next 10 cm of soil is considered in the dose modeling. In the example, two simulations are run and the doses from the simulations added together. In the first simulation, the entire 15 cm of soil is considered but using the lower concentration of the deeper 5 to 15 cm soil interval. In the second simulation, the top 5 cm of soil is considered using the difference in concentration between the shallower 0 to 5 cm and the deeper 5 to 15 cm soil intervals. The doses from the two simulations are summed to provide a dose estimate for the two soil intervals represented in the model.

Figure I.1 Approach to Considering Multiple Elevated Areas or Hot Spots I-16

Figure I.2 Approach to Considering Vertical Heterogeneity I-17

Case Study 1: Evaluation of the Impact of Variability of Exposure Area Concentrations Using the dose modeling approach, analysts can evaluate the uncertainty in dose modeling predictions based on spatial variability in exposure area concentrations using a code such as SADA. A case study by the NRC staff used soil sampling data to calculate average concentrations for various exposure areas at a thorium-contaminated site. The default exposure area first considered was a 2,000 m2 area, which is of sufficient size to support the assumed external exposure pathway occupancy factors for a resident gardener, as well as the assumed plant ingestion rates derived from regional data. The NRC used SADA to calculate a spatially moving average concentration or the possible set of concentrations for 2,000 m2 exposure areas. The NRC then used the distribution of 2,000 m2 exposure area concentrations for comparison against an area-wide DCGL derived from dose modeling. The staff also performed sensitivity analyses that considered a range of exposure area sizes to provide additional information with respect to other types of uncertainty, such as institutional, land use, or behavioral uncertainties (e.g., uncertainty in future land use, including the size and location of parcels of land a resident may reside on following license termination and redevelopment of the released site). It is important to note that, as the size of the exposure area changes, so may the exposure pathways, occupancy factors, and other exposure scenario-related parameters. For example, if one were to consider relatively small exposure areas appropriate for an apartment or condominium resident, which will result in higher maximum exposure area concentrations, then it may be (1) unreasonable to assume certain pathways are viable, such as homegrown produce ingestion, and (2) shielding factors for the apartment or condominium residence may need to be adjusted to more accurately calculate external pathway dose. In these cases, dose modeling can estimate dose for the smaller exposure areas using more representative parameters and pathways appropriate for the exposure scenario being considered, with appropriate justification. Care should also be taken to ensure that the statistics are not skewed due to the presence of large areas of unimpacted soils (e.g., only Class 1-3 areas should be included in the analysis for particularly large sites with significant portions of the site that are not radiologically affected by facility operations).

Evaluation of uncertainty or variability in exposure area concentrations will lead to a more comprehensive evaluation of risk to the average member of the critical group. Analysis of a range of exposure areas, including smaller exposure areas that may be appropriate for alternative exposure scenarios, will better inform the decisionmaker with respect to less likely but plausible exposure scenarios that may be more limiting. As the size of the exposure areas decreases and approaches the size of true elevated areas or hot spots, SADA can also provide valuable information on the distribution of elevated area concentrations at a site. Consideration could then be given to the number and location of elevated areas on the site to ensure that potential risk is well understood. In some cases, it may be reasonable to assume the receptor spends his time at only one elevated area or to consider some form of time or spatial weighting if the elevated areas are located a considerable distance apart and the exposure area is rather large in relation to the elevated areas.

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I.3 Criteria for Selecting and Modifying Exposure Scenarios, Pathways, and Critical Groups I.3.1 Introduction After the source term has been evaluated, the question becomes, How could humans be exposed either directly or indirectly to residual radioactivity? or What is the appropriate exposure scenario? Each exposure scenario should address the following questions:

(1) How does the residual radioactivity move through the environment?

(2) Where can humans be exposed to the environmental concentrations?

(3) What is the likely land use(s) in the future for these areas?

(4) What are the exposure groups habits that will determine exposure? (For example, what do they eat and where does it come from? How much? Where do they get water and how much? How much time do they spend on various activities?)

The ultimate goal of dose modeling is to estimate the dose to a specific receptor. Broad generalizations of the direct or indirect interaction of the affected receptors with the residual radioactivity can be identified for ease of discussion among the licensee, regulator, public, and other interested parties. Exposure scenarios are defined as reasonable sets of activities related to the future use of the site. Therefore, exposure scenarios describe future land uses, human activities, and the behavior of the natural system.

In most situations, there are numerous possible exposure scenarios of how future human exposure groups could interact with residual radioactivity. The compliance criteria in 10 CFR Part 20 for decommissioning do not require an investigation of all (or many) possible exposure scenarios; their focus is on the dose to members of the critical group. The critical group is defined (at 10 CFR 20.1003) as the group of individuals reasonably expected to receive the greatest exposure to residual radioactivity for any applicable set of circumstances.

By combining knowledge about the answers to Questions 1 and 2, the licensee can develop exposure pathways. These are the routes that residual radioactivity travels through the environment, from its source, until it interacts with a human. They can be fairly simple (e.g., surface-soil residual radioactivity emits gamma radiation, which results in direct exposure to the individual standing on the soil) or they can be fairly involved (e.g., the residual radioactivity in the surface soil leaches through the unsaturated soil layers into the underlying aquifer, and the water from the aquifer is pumped out by the exposed individual for use as drinking water, which results in the exposed individual ingesting the environmental concentrations). Exposure pathways typically fall into three principal categories, identified by the manner in which the exposed individual interacts with the environmental concentrations resulting from the residual radioactivity: ingestion, inhalation, or external (i.e., direct) exposure pathways.

As required under Subpart E, the licensee evaluates the dose from residual radioactivity for the average member of the critical group, which is not necessarily the same as the maximally exposed individual. This is not a reduction in the level of protection provided to the public but an attempt to emphasize the uncertainty and assumptions needed in calculating potential future doses, while limiting boundless speculation on possible future exposure scenarios. Although it I-19

is possible to actually identify with confidence the most exposed member of the public in some operational situations (e.g., through monitoring, time studies, distance from the facility),

identification of the specific individual who may receive the highest dose some time (up to 1,000 years) in the future is impractical, if not impossible. Speculation on his or her habits, characteristics, age, or metabolism could be endless. The use of the average member of the critical group acknowledges that any hypothetical individual used in the performance assessment is based, in some manner, on the statistical results from data sets (e.g., the breathing rate is based on the range of possible breathing rates) gathered from groups of individuals. Although bounding assumptions could be used to select values for each of the parameters (e.g., the maximum amount of meat, milk, vegetables, possible exposure time), the result could be an extremely conservative calculation of an unrealistic exposure scenario and may lead to excessively low allowable residual radioactivity levels, compared to the actual risk.

Calculating the dose to the critical group is intended to bound the individual dose to other possible exposure groups, because the critical group is a relatively small group of individuals, who, because of their habits, actions, and characteristics, could receive among the highest potential doses at some time in the future. By using the hypothetical critical group, coupled with prudently conservative models, it is highly unlikely that any individual would actually receive doses in excess of that calculated for the average member of the critical group. The licensee should base the description of a critical groups habits, actions, and characteristics on credible assumptions, and the information or data ranges used to support the assumptions should be limited in scope to reduce the possibility of adding members of less exposed groups to the critical group.

As low as is reasonably achievable (ALARA) analyses should use the dose based on the reasonably foreseeable land use for any cost-benefit calculations performed.

I.3.2 Issues in Selecting and Modifying Exposure Scenarios, Pathways, and Critical Groups The definition of exposure scenarios, identification of a critical group with its associated exposure pathways, and the dose assessment based on that definition can be generic or site-specific. Licensees might do the following:

  • Use screening exposure scenarios, screening groups, and pathway parameters as described in NUREG-1549 (NRC, 1998d) and the NUREG/CR-5512 series, Residual Radioactive Contamination from Decommissioning, Volumes 3 and 4, issued October 1999 (e.g., NRC, 1999c, 1999d). This can be used for either screening or site-specific analyses.
  • Use the default screening exposure scenarios as a starting point to develop more site-specific pathway analyses or critical group habits.
  • Develop site-specific exposure scenarios and critical groups and identify associated exposure pathways from scratch.

To establish site-specific exposure scenarios, critical groups, or sets of exposure pathways, the licensee may need to justify its selections. For some licensees, this may require minimum amounts of site-specific data to support the assumptions inherent in the existing default screening exposure scenarios or for removing specific exposure pathways. For others, the licensee may need to thoroughly investigate and justify the appropriateness of the selected I-20

exposure scenarios or critical groups, which may include evaluation of alternative exposure scenarios or critical groups. If a licensee creates the exposure scenario and associated critical group based on site-specific conditions (e.g., at a site that is grossly different than the assumptions inherent in the default exposure scenarios), it should include documentation that provides a transparent and traceable audit trail for each of the assumptions used in developing the exposure scenario and critical group (e.g., justify the inclusion (or exclusion) of a particular exposure pathway).

I.3.3 Recommended Approaches I.3.3.1 Screening Analyses In the case of screening, the decisions involved in identifying the appropriate exposure scenario and critical group, with their corresponding exposure pathways, have already been made.

Exposure scenario descriptions acceptable to the NRC staff for use in generic screening are developed and contained in NUREG/CR-5512, Volume 1. NUREG/CR-5512, Volume 3, and NUREG-1549 provide the rationale for applicability of the generic exposure scenarios, critical groups, and pathways at a site; the rationale and assumptions for exposure scenarios and pathways included (and excluded); and the associated parameter values or ranges (NUREG/CR-5512, Volume 3, contains detailed information on data to support development of parameter distributions). A description of the screening exposure scenarios and associated pathways is provided in Table I.2.

I.3.3.2 Site-Specific Analyses Site-specific analyses give licensees greater flexibility in developing the compliance exposure scenario. The licensee should justify its selection of this scenario based on reasonably foreseeable land use at the site. This scenario should result in an exposure to the public, such that no other exposure scenario, using reasonably foreseeable land use assumptions, will result in higher doses to its exposure group(s). The level of justification and analysis provided by the licensee will depend on how much credit is taken by adding realism to the exposure scenario based on site-specific information (i.e., how much credit is being taken for elimination of pathways or reduction in pathway contributions relative to more conservative assumptions made in screening exposure scenarios). As the analysis becomes more realistic, greater degrees of justification and, potentially ancillary analyses, will be required. For example, a site is currently zoned as industrial, and the local area is a mix of suburban, commercial, and industrial uses.

Rural uses of the property are less likely but plausible for the foreseeable future. If it chose to use the generic screening exposure scenario, the licensee would need to provide limited justification. If the licensee proposed to use a maintenance worker exposure scenario assuming industrial land use as the compliance exposure scenario, it would need to provide quantitative analyses of, or a qualitative argument discounting the need to analyze, other competing exposure scenarios (based on industrial land use and on suburban or commercial land use) to justify the selection of the compliance exposure scenario. In addition, the licensee would need to provide analyses of the rural use of the land to show what impacts would occur from the less likely but plausible exposure scenario.

Site-specific analyses can use the generic screening exposure scenario(s) with little justification.

The licensee may need to justify that the site contains neither physical features nor locations of residual radioactivity (other than those assumed in the screening analyses), which would invalidate the assumptions made in developing the exposure scenarios. If site or source features are found to be incompatible with exposure scenario assumptions, the licensee should I-21

justify why the generic exposure scenario are nonetheless appropriate for use in the dose modeling. A site can fail to meet the requirements of the conceptual model (see Section I.4 of this appendix) without invalidating the generic exposure scenario, and situations can arise where the default exposure scenario is no longer the limiting case. For example, the site may have preexisting groundwater contamination, which is counter to the assumptions in the conceptual model inherent in the screening models. However, this may not require any change in the exposure scenario, because the residential farmer scenario may still be an appropriate scenario, as it contains all of the appropriate exposure pathways, including groundwater use for drinking, irrigation, and for animals. Alternately, if the residual radioactivity were a volumetric source in the walls of a building, rather than on the building surfaces, the generic exposure scenario of an office worker may not be the exposure scenario leading to the critical group. For certain sets of radionuclides, a building renovation exposure scenario may be more limiting because of the exposure to the airborne concentration of material as the walls are disturbed.

Table I.2 Pathways for Generic Exposure Scenarios Building Occupancy Exposure Scenario This exposure scenario accounts for exposure to fixed and removable residual radioactivity on the walls, floor, and ceiling of a decommissioned facility. It assumes that the building may be used for commercial or light industrial activities (e.g., an office building or warehouse).

Pathways include the following:

  • external exposure from building surfaces
  • inhalation of (re)suspended removable residual radioactivity
  • inadvertent ingestion of removable residual radioactivity Resident Farmer Exposure Scenario This exposure scenario accounts for exposure involving residual radioactivity that is initially in the surficial soil. A farmer moves onto the site and grows some of his or her diet and uses water tapped from the aquifer under the site.

Pathways include the following:

  • external exposure from soil
  • inhalation to (re)suspended soil
  • ingestion of soil
  • ingestion of drinking water from aquifer
  • ingestion of plant products grown in contaminated soil and using aquifer to supply irrigation needs
  • ingestion of animal products grown on site (using feed and water derived from potentially contaminated sources)
  • ingestion of fish from a pond filled with water from the aquifer I-22

The licensee can develop site-specific exposure scenarios, critical groups, and pathways for any situation. Some cases where changes to the default exposure scenarios or modification to exposure pathways are or are likely to be appropriate include the following:

(1) Major pathways (e.g., the groundwater pathway or agricultural pathways) associated with the default screening scenarios could be eliminated, for either physical or site-use reasons.

(2) The location of the residual radioactivity and the physical features of the site are outside the major assumptions used in defining the default critical group and exposure scenarios.

(3) Restricted use is proposed for a site.

The second case listed above can be ambiguous, as a number of assumptions key to the development of the DandD screening tool do not affect the exposure scenario description, and the NRC reviewer may need to evaluate whether the initial generic exposure scenario would still be appropriate for the site.

Modifying exposure scenarios or developing a site-specific critical group requires information on plausible uses of the site and demographics. Such information might include considerations of the prevailing (and future) uses of the land and physical characteristics of the site that may constrain site use. The licensee should categorize potential land uses as reasonably foreseeable, less likely but plausible, or implausible. Any land uses that similar property in the region currently has, or may have in the near future (e.g., approximately 100 years), should be characterized as reasonably foreseeable. The licensee should consider trends and area land use plans in determining the likelihood of potential land use. Land uses that are plausible, generally because similar land was historically used for similar purposes, could be characterized as less likely but plausible if found to be counter to the current trends or regional experience, (e.g., rural use of property currently in an urban setting). Implausible land uses are those that, because of physical limitations, could not occur (e.g., residential land use for an underwater plot of land). It may be necessary to evaluate several potential critical groups, based on different combinations of site-specific exposure scenarios developed from expected land use, pathways, and demographics, to determine the group receiving the highest exposure.

Depending on the resulting exposure scenarios, considerations of offsite exposure by either transport (e.g., through groundwater) or material transfer may be necessary to identify the critical group. Thus, the licensee should consider if offsite uses are reasonably foreseeable. If offsite uses are found to be reasonably foreseeable, such offsite uses should be analyzed to determine if the offsite user receives a higher dose compared to an onsite user and if offsite users should be identified as the critical group.

Similar considerations apply for restricted release. Thus, when analyzing the dose under restricted conditions, the nature of the critical group is likely to change because of these restrictions and controls. Site restrictions and institutional controls can restrict certain kinds of activities and land or water uses associated with the physical features of the site. The detailed definition of the exposure scenarios considered for restricted release need to include the impact of the control provisions on the location and behavior of the average member of the appropriate critical group.

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For restricted use, licensees must also evaluate doses assuming institutional controls are no longer in effect. This evaluation should address (1) the associated degradation of engineered barriers assuming there is no active maintenance, and (2) exposure scenarios assuming a loss of institutional controls immediately following license termination (i.e., time=0 years).

Section 3.5 of this volume contains additional information on modeling the performance and degradation of engineered barriers in dose modeling analyses.

The NRC license reviewer should evaluate the licensees justifications for its exposure scenarios using the following appropriate guidance. The guidance is characterized by the general approach used in developing the exposure scenarios, either (1) modifying existing generic exposure scenarios or (2) developing site-specific scenarios from scratch.

MODIFICATION OF GENERIC EXPOSURE SCENARIOS First, the NRC license reviewer should evaluate whether the generic exposure scenario was applicable to the site before the licensee started modifying the exposure scenario, based on physical features or restrictions, and should identify the modifications and evaluate the licensees justification for those changes. Table I.3 lists some common exposure scenarios but is by no means comprehensive. The Sandia Letter Report, Process for Developing Alternate Scenarios at NRC Sites Involved in D&D and License Termination, issued January 2000 (Thomas, et al., 2000)), which is included, in part, as Appendix M in this volume, provides information to assist a licensee or reviewer with respect to the modification of default exposure scenarios using site-specific information. Specific guidance on acceptable justifications for modifying the default exposure scenarios is provided below, based on different types of site-specific information. Additionally, if the licensees intent is restricted release, the NRC should review the final exposure scenario for the case where restrictions are in place. Based on either site restrictions or site-specific data, the licensees justifications should support the elimination from the analysis of exposure scenarios and pathways. The NRC should focus the review on the most risk-significant pathways and model components.

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Table I.3 Potential Exposure Scenarios for Use in Dose Assessments

  • urban construction (contaminated soil, no suburban or agricultural uses), meant for small urban or industrial sites cleared of all original buildings, with only contaminated land and/or buried waste remaining
  • residential (residential farmer exposure scenario with eliminated exposure pathways appropriate for those urban or suburban sites where farming is not a realistic projected future use of the land)
  • recreational user (where the site is preserved for recreational uses only)
  • maintenance worker (tied to the recreational user exposure scenario but involves the grounds keepers maintaining or building on the site)
  • hybrid industrial building occupancy (adds contaminated soil, while building may or may not be contaminated)
  • offsite drinking water (e.g., no onsite use of groundwater; offsite impacts from the contaminated plume)

The licensee may need to evaluate whether the final modified exposure scenario is still the limiting reasonable representation of the critical group at the site. This may involve investigating exposure pathways not covered in the default exposure scenarios.

DEVELOPMENT OF ALTERNATIVE EXPOSURE SCENARIOS In some decommissioning cases, either the location of the residual radioactivity, the physical characteristics of the site, or planned institutional restrictions may make the default exposure scenarios inappropriate. In other cases, the licensee may wish to provide a transparent and traceable development of the compliance and other exposure scenarios, starting with the potential land use and the site conditions. Development (and review) of alternative exposure scenarios may involve iterative steps to create the conceptual model of the site. For example, the licensee may (1) develop a generic list of exposure pathways, (2) develop the site conceptual model to screen the generic list, (3) aggregate or reduce the remaining exposure pathways to the major exposure pathways, and (4) reevaluate the conceptual model to verify that all the necessary processes are included.

A brief summary of the NRC-recommended pathway analysis process follows. Appendix K contains an example of exposure scenarios developed for PSR.

  • Compile a list of exposure pathways applicable to any contaminated site. A number of existing sources of information can be used, for example, NUREG/CR-5512, Volume 1 (NRC, 1992). Another source, although the guidance is more focused on offsite exposures, is NUREG/CR-5453, Background Information for the Development of a Low-I-25

Level Waste Performance Assessment Methodology, Volumes 1 and 2, issued December 1989 (NRC, 1989).

  • Categorize the general types of residual radioactivity at the site (e.g., sediment or soil, deposits in buildings, surface residual radioactivity, surface water, groundwater, or industrial products such as slag).
  • Screen out pathways, for each contaminant type, that do not apply to the site.
  • Identify the physical processes pertinent to the remaining pathways for the site.
  • Separate the list of exposure pathways into unique pairs of exposure media (e.g., source to groundwater, groundwater to surface water). Determine the physical processes that are relevant for each exposure media pair and combine the processes with the pathway links.
  • Reassemble exposure pathways for each source type, using the exposure media pairs as building blocks, thus associating all the physical processes identified with the individual pairs with the complete pathway.

The licensees documentation of the decisions made about inclusion (or exclusion) of the various pathways should be transparent and traceable. An international working group established a methodology for developing models to analyze radionuclide behavior in the biosphere and associated radiological exposure pathways (i.e., the Reference Biospheres Methodology). BIOMOVS II published the methodology in its Technical Report No. 6, Development of a Reference Biospheres Methodology for Radioactive Waste Disposal, issued September 1996 (SSI, 1996), and included a list of international biosphere features, events, and processes.6 The report may be useful as a guide for additional information on a logical method to complete the pathway analysis above and include proper justification. Generally, the Reference Biospheres Methodology is more useful for complex sites that may have numerous physical processes that interact in such a way that a number of different exposure groups may need to be investigated to identify the critical group. Additional work has been done on implementing the Reference Biospheres Methodology by a working group of the International Atomic Energy Agencys (IAEAs) Biosphere Modeling and Assessment (BIOMASS) program (IAEA, 1999a, 1999b, 2001). Specifically, IAEA Working Document BIOMASS/T1/WD03, Guidance on the Definition of Critical and Other Hypothetical Exposed Groups for Solid Radioactive Waste Disposal, may provide additional information on developing a site-specific critical group for situations where the generic critical group is inappropriate (IAEA, 1999b).

6 Additional features, events, and processes lists are presented and discussed in Appendix C of NUREG-2175, draft Guidance for Conducting Technical Analyses for 10 CFR Part 61, issued March 2015, that may be appropriate for more complex decommissioning sites (NRC, 2015a). In additional to considering alternative exposure scenarios, for some complex decommissioning sites, the licensee may need to consider central and alternative scenarios, as defined in this volume (see definition of scenario). For example, if long-lived residual radioactivity is present at an actively eroding site, alternative scenarios related to future landscape evolution of the site may need to be evaluated to adequately assess the long-term risk associated with residual radioactivity remaining at a site. Consideration of this type of alternative scenario (i.e., an alternative scenario that considers future landscape evolution) differs fundamentally from consideration of alternate exposure scenarios, which primarily focus on assumptions related to future human behavior and expected land use. In some cases, however, an alternative scenario may cause the exposure pathways to change and require consideration of a new alternative exposure scenario as well (e.g., gully erosion may expose buried residual radioactivity that could expose a recreational user of the site).

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I.3.3.3 Guidance on Specific Issues Land Use A licensees assumptions for land use should focus on current practice in the region. The region of concern can be as large as an 80-kilometer (km) (50-mile) radius. To narrow the focus of current land practices, the licensees can use information on how land use has been changing in the region and should give more weight to land use practices either close to the site or in similar physical settings. This can be very important for semirural sites that are being encroached upon by suburban residential development. Reviewers may wish to involve State and local land use planning agencies in discussions, if the licensee has not already requested their involvement.

Potential land uses should be categorized as reasonably foreseeable, less likely but plausible, or implausible. Any land uses that similar properties in the region currently have, or may have in the near future (e.g., approximately 100 years7), should be characterized as reasonably foreseeable. Consideration should be given to trends and area land use plans in determining the likelihood of potential land use. The time frame of interest for exposure scenario development could be less than 100 years in certain cases and would depend on such factors as the rate of change in land use patterns in the area, radionuclides of interest, and the time of peak dose. For example, a site with residual cobalt-60, which has approximately a 5-year half-life, would not likely need to explore possible land uses that may exist at the site beyond a few decades, because of the natural decay of the residual material.

Land uses that are plausible, generally because similar land historically was used for the purpose but are counter to the current trends or regional experience should be characterized as less likely but plausible (e.g., rural use of property currently in an urban setting). Implausible land uses are those that, because of generally physical limitations, could not occur (e.g., residential land use for an underwater plot of land).

Land use justifications by licensees often rely on State or local codes, in building or well development to constrain future use. In general, licensees requesting unrestricted release should not rely solely on these factors as reasons to remove pathways or justify the exposure scenario unless (1) the radionuclides have a relatively short half-life (approximately 10 years or less) or (2) the dose from long-lived radionuclides reaches its peak before 100 years. Similarly, licensees requesting unrestricted release should not limit land use exposure scenarios based on commitments or require the enforcement of limitations by the licensee or another party (e.g., a licensee reiterates that the land will remain industrial by stating that the land will not be sold by the licensee after the license is terminated).

Licensees should base justifications of land use on (1) the nature of the land and reasonable predictions based on its physical and geologic characteristics, and (2) societal uses of the land, based on past historical information, current uses of it and similar properties, and what is reasonably foreseeable in the near future. The societal uses of the site in the future should be based on advice from local land planners and other stakeholders on what possible land uses are likely within a time period of around 100 years. The level of justification for the final land uses is inversely proportional to the level of realism assumed by the licensee. Limited justification may be required for bounding analyses, while much more detailed justification, 7 Note that the 100-year timeframe described here is only for estimating future land uses; the licensee must evaluate doses that could occur over the 1,000-year time period specified in the LTR.

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including alternative reasonably foreseeable and less likely but plausible exposure scenario analyses, may be needed for a situation with a smaller degree of conservatism in the analyses.

Additional guidance is available on potential sources of land use information in Appendix M.

WATERBORNE EXPOSURE PATHWAYS Removal of waterborne exposure pathways can range from global (e.g., all groundwater pathways) to specific (e.g., no drinking water but agricultural/fish pond use remains).

Acceptable justifications are generally based on physical conditions at the site rather than local codes. The licensee should base its justification of water quality and quantity of the saturated zone on the classification systems used by the U.S. Environmental Protection Agency (EPA) or the State, as appropriate. Arguments involving depth to water table, or well production capacity, should have supporting documentation from either the USGS, an appropriate State agency, or an independent consultant.

NRC license reviewers should evaluate the reasons for the classification (e.g., information on water quality used as a basis for eliminating pathways). Where the aquifer is classified as not being a source of drinking water but is adequate for stock watering and irrigation, the licensee can eliminate the drinking water pathway but should still maintain the irrigation and meat/milk pathways. Aquifers may exceed certain constituents and still be able to be used for various purposes, because those constituents may easily be treatable (e.g., turbidity). In cases where the water may be treatable or because the degree of connection between the aquifer and surface water may make the use of the aquifer questionable, the reviewer should involve the EPA or the State, or both, as appropriate, in discussions on reasonable assumptions for the aquifer use.

AGRICULTURAL PATHWAYS Agricultural pathways may be removed or modified for reasons such as (1) land use patterns, (2) poor-quality soil, (3) topography, and (4) size of contaminated area. Many justifications may result in modification of the pathways, rather than complete elimination. For example, the poor quality of the soil may make intensive farming activities impractical, but residential gardening may still be reasonable.

Licensees using poor-quality soil as a justification for modifying the agricultural pathways should provide the reviewer with supporting documentation from the Soil Conservation Service, appropriate State or local agency, or an independent consultant. Reviewers should carefully consider whether the state of the soil would reasonably preclude all activities (e.g., because of high salinity of soil) or only certain activities. In most cases, soil quality can reasonably preclude activities such as intensive farming but could allow grazing or small gardens.

When reviewing justifications involving topography, the NRC reviewer should limit speculation of future topographical changes from civil engineering and evaluate the reasonableness of the critical group performing its activities on the current topography, for example, a slope. The licensee should provide supporting documentation in the form of pictures, USGS or similar topographic maps, hand-drawn maps, or a detailed description of how the topography would limit farming. NRC reviewers may wish to visit a site to evaluate the topography firsthand.

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AGE-DEPENDENT CRITICAL GROUPS The definitions in 10 CFR Part 20 should be used when demonstrating compliance with the requirements of Subpart E. EPAs Federal Guidance Report No. 11, Limiting Values of Radionuclide Intake and Air Concentration and Dose Conversion Factors for Inhalation, Submersion, and Ingestion, issued September 1988, should be used when calculating internal exposures by using the intake-to-dose conversion factors, which are based primarily on adults.

EPAs Federal Radiation Protection Draft Guidance for Exposure of the General Public, Volume 59, dated December 23, 1994 indicates that implementing age and sex dependent limits for the general public is difficult and due to uncertainty in the impact of these factors on dose that detailed consideration of age and sex is generally unnecessary.

Since age-based dose conversion factors are not being used, the same dose conversion factors are applied to all individuals. Only in rare exposure scenarios will a nonadult individual receive a higher dose (i.e., intake more radioactive material) than an adult individual in a similar exposure scenario. One example may be related to the milk pathway. Children generally drink more milk annually than adults. If milk were the only pathway that would expose the individual to a dose, then the child would have a slightly higher dose than the adult. But in most situations, especially ones involving multiple pathways, the total intake of the adult is greater than that of a child. Therefore, for most multiple pathway exposure scenarios, such as screening analyses, the average member of the critical group should usually be assumed to be an adult, with the proper habits and characteristics of an adult. As the licensee eliminates pathways or modifies the exposure scenario, the behavior and dietary habits of children may become important. In such cases, the licensees should contact the NRC staff for guidance.

IMPACT OF AREA ON EXPOSURE PATHWAYS AND SCENARIOS As discussed above, the default exposure scenario for surface soil assumes large areas of homogeneous surface residual radioactivity. If the area of residual radioactivity is smaller than that needed to support the exposure pathway (e.g., an insufficient area to support the production of the quantity of contaminated crops assumed to be consumed by the receptor), the licensee may propose modifying the exposure pathways to account for the effect area has on the critical groups activities. The licensee can use different approaches to account for smaller areas of elevated activity, including the following two methods:

(1) Reduce the calculated dose by modifying the exposure time or usage parameters accordingly.

(2) Modify the exposure scenario and pathways and/or modify the calculation method to account for the size of the residual radioactivity.

These methods may be built into some dose assessment codes for surficial soil (e.g., RESRAD-ONSITE), but the user should understand how the codes consider the area of contamination in adjusting pathway doses. For example, RESRAD-ONSITE and RESRAD-BUILD are commonly used to perform site-specific dose modeling and to calculate DCGLs. These codes uniquely consider the size of the area of the contaminated zone in calculating the dose for each pathway.

It should be noted, however, that RESRAD-ONSITE and RESRAD-BUILD do not adjust the occupancy factors for smaller areas of contamination. As the size of the elevated areas decrease, full occupancy assumptions may become increasingly unrealistic and overly conservative (e.g., by default, RESRAD-ONSITE assumes that the receptor is located in the I-29

center of the contaminated zone). In these cases, the licensee may present arguments for modified occupancy assumptions or alternative exposure scenario assumptions for smaller areas of residual radioactivity, based on the expected habits and characteristics of members of critical groups, as well as the characteristics of the site being evaluated. Certain exposure pathways, such as meat and plant ingestion, are also affected by the size of the elevated area.

Therefore, the licensee can present arguments for why the dose from certain pathways may be limited due to the area of the elevated concentration and adjust other parameters, as necessary, to avoid overly conservative dose calculations and associated overly conservative cleanup levels. For example, if the contaminated plant fraction is set to -1, RESRAD-ONSITE adjusts the total amount of contaminated plant products ingested based on the size of the contaminated area. If other dose modeling codes are used, the licensee may justify using lower consumption rates of contaminated homegrown produce or animal products, based on the size of the elevated area.

Table I.4 summarizes the radionuclide-specific factors above the DCGLW that are allowable for smaller areas of residual radioactivity (i.e., area factors).8 These factors are (1) provided in lookup tables included in various regulatory documents, (2) calculated using the formula provided for field analysis in the Users Manual for RESRAD Version 6, and (3) calculated using RESRAD-ONSITE, Version 6.5. Differences between area factor values listed for a specific radionuclide can be attributed to a variety of factors, including the use of different models or methods, modeling parameter values (e.g., length parallel to aquifer flow in the RESRAD nondispersion model), and modeling assumptions. The EPA Soil Screening Guidance (EPA, 1996) and RESRAD Field Formula methods are generally more limiting for radionuclides where the dose is attributable primarily from ingestion pathways and comparable to other methods for radionuclides where the dose is attributable primarily to the external exposure pathway.

Some general observations related to cleanup levels for smaller areas and general characteristics of area factors include the following:

  • The DCGLEMC value will always be greater than or equal to the DCGLW value (i.e., area factors will always be larger than 1). In other words, higher concentrations or cleanup levels are allowable for smaller elevated areas compared to larger areas of residual radioactivity to meet the same radiological criteria for license termination.
  • With respect to water-independent pathways:

o RESRAD-calculated area factors for gamma-emitting radionuclides where the risk is dominated by the external exposure pathway generally have more restrictive or lower area factors that vary nonlinearly with the area of contamination (see for example cesium (Cs)-137 and Co-60 in Table I.4).

o RESRAD-calculated area factors for the inhalation exposure pathway also vary nonlinearly with the size of the contaminated area.

8 It is important to note that DCGLEMCs should be based on dose modeling and not calculated based on area factors reported in this volume or default area factors provided in other reference material. Area factors reported in this volume are provided to facilitate discussion of differences in the impact of area on dose for different radionuclides and pathways only. They should not be interpreted as being acceptable for use in developing DCGLEMCs.

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o RESRAD-calculated area factors for ingestion pathways such as soil ingestion, animal product ingestion, and plant ingestion generally have area factors that scale directly to the size of the contaminated area.

  • With respect to water-dependent pathways:

o RESRAD-calculated area factors for the drinking water and fish ingestion pathways generally have more limiting area factors compared to other water-dependent pathways and can be more limiting than water-independent pathways.

o RESRAD-calculated area factors for ingestion pathway doses incurred from use of contaminated irrigation water are generally lower and DCGLs are generally higher or less restrictive compared to other water dependent and water-independent pathways, and therefore, area factors for these pathways are generally not limiting and are not as risk significant.

The NRC staff should review the following information provided by the licensee:

  • summary table or list of the DCGLW(s) for each radionuclide and impacted medium of concern
  • the DCGLEMC for each radionuclide and medium of concern, if Class 1 survey units9 are present
  • the appropriate DCGLW for the survey method to be used if multiple radionuclides are present The NRC should review licensee-calculated DCGLEMCs to ensure that the values are developed based on dose modeling (i.e., default area factors found in the literature should not be used to assign DCGLEMCs). Consideration can be given to site-specific conditions, including the contributions of individual exposure pathways to the overall dose, since exposure scenarios and pathways can vary from one site to another. Additionally, the NRC should carefully review the licensees approach for considering multiple radionuclides and elevated areas.

9 Class 1 survey units are impacted areas that are expected to have concentrations of residual radioactivity that exceed the DCGLW.

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Table I.4 Comparison of Area Factor Values from Different References 10000 m2 1000 m2 100 m2 10 m2 1 m2 NUREG-1505 1 1.01 1.86 13.4 109 Am-241 NUREG-1575 1 1.3 13.4 96.3 208.7 EPA Soil Screening Guidance 1 1.1 1.3 2.3 N/A RESRAD Field Formula 3.2 10 RESRAD-ONSITE, Version 6.5 1 1.01 1.87 14.1 124 NUREG-1505 1 1.1 1.41 2.41 11 Cs-137 NUREG-1575 1 1.1 1.4 2.4 11 EPA Soil Screening Guidance 1 1.1 1.3 2.6 N/A RESRAD Field Formula 3.2 10 RESRAD-ONSITE, Version 6.5 1 1.14 1.41 2.41 11 NUREG-1505 1 1.1 1.23 2.12 9.81 NUREG-1575 1 1.1 1.2 2.1 9.8 Co-60 EPA Soil Screening Guidance 1 1.1 1.3 2.6 N/A RESRAD Field Formula 3.2 10 RESRAD-ONSITE, Version 6.5 1 1.06 1.23 2.12 9.81 NUREG-1505 1 1.03 1.75 3.12 12.3 Th-232 NUREG-1575 1 1.1 1.8 3.2 12.5 EPA Soil Screening Guidance 1 1.1 1.3 2.3 N/A RESRAD Field Formula 3.2 10 RESRAD-ONSITE, Version 6.5 1 1.05 1.82 3.31 15 NUREG-1505 1 1.04 2.27 11.1 30.5 NUREG-1575 1 1.3 6.7 11.1 30.6 U-238 EPA Soil Screening Guidance 1 1.1 1.3 2.4 N/A RESRAD Field Formula 3.2 10 RESRAD-ONSITE, Version 6.5 1 1.15 2.32 15.4 80.2 Notes:

  • NUREG-1505 (NRC 1998b), Table 8.1, reports radionuclide-specific area factors calculated using RESRAD-ONSITE, Version 5.7.
  • NUREG-1575 (NRC 2000a), Table 5.6, reports radionuclide-specific area factors calculated using RESRAD-ONSITE, Version 5.6.
  • Area factors associated with the EPA Soil Screening Guidance (EPA, 1996) are the reciprocals of the area correction factors reported in Table 5.2 of the document. The area correction factors are based on the external exposure pathway only.

As indicated above, when the extent of residual radioactivity becomes smaller, some of the activities are no longer viable as reasonable assumptions for exposure. Generally, the first pathways affected are animal husbandry activities, because of the larger area needed for grazing and growing fodder. As a general rule, as the area gets smaller, the more the exposure scenario transforms into a residential gardener exposure scenario, so long as the initial residual radioactivity begins in the surface soil. For cases where the residual radioactivity is not in the I-32

surficial soil, the original area of residual radioactivity may not be as important in exposure scenario development, because some of the primary transport mechanisms result in redistribution of the radionuclides over larger areas (i.e., groundwater used as irrigation).

One common mistake in licensee submittals is that DCGLEMCs are typically not provided for residual radioactivity on building surfaces. When the screening DCGLW values were published in the Federal Register (see Appendix H), associated DCGLEMCs were not published. Although newer versions of DandD (Version 2) allow specification of a limited area of contamination, DandD adjusts dose based on area in a simplistic manner. Therefore, the licensee may wish to calculate DCGLEMCs for building surfaces using the RESRAD-BUILD computer code.

OFFSITE EXPOSURE SCENARIOS As discussed above, in rare situations, the exposure scenario resulting in the highest exposures from the residual radioactivity will be an offsite use exposure scenario. For these evaluations, the dose limits in 10 CFR Part 20, Subpart E, remain applicable, even though the situation may seem similar to the clearance of materials before license termination. In these scenarios, the exposure to the radioactive material will occur, because it has been removed from the current location, and this results in either new or enhanced exposure pathways. For example, a site has poor groundwater characteristics (thereby, allowing the licensee to remove the groundwater pathway from any applicable exposure scenarios), and the reasonably foreseeable land use is either commercial or industrial. The primary contaminant is technetium (Tc)-99, which primarily results in dose through either the groundwater or vegetable pathways, both of which are not applicable to the physical characteristics of the site or land use assumptions. The residual radioactivity is present in the sites topsoil. A possible offsite exposure scenario is where, during construction of any commercial interest on the site after license termination, the removed topsoil is sold for use in a residential setting. In this case, it is likely that the topsoil with residual radioactivity will be unintentionally mixed with other topsoil at the offsite location. Licensees can use generic analyses to screen the importance of offsite uses with such sources as NUREG-1640, Radiological Assessments for Clearance of Materials from Nuclear Facilities.

(NRC 2003b)

Even if offsite use is not considered reasonably foreseeable, offsite exposure scenarios may be less likely but plausible scenarios and should be analyzed as exposure scenarios, to understand the robustness of the analysis.

DETERMINING THE COMPLIANCE EXPOSURE SCENARIO In many situations, a licensee will be faced with selecting a compliance exposure scenario from a potentially large suite of exposure scenarios and exposure groups. The licensee is expected to base its demonstration of compliance on the reasonably foreseeable exposure scenario resulting in the highest peak dose during the compliance period, consistent with the definition of the critical group. Licensees may find it advantageous to use an iterative approach to screen all the potential exposure scenarios. This will allow the licensees to focus their more detailed analyses on the important exposure scenarios. Licensees may be able to use information from NUREG/CR-5512 (NRC 1992), NUREG-1640 (NRC 2003b), and NUREG-1717 (NRC 2001), as well as other licensees analyses to screen their potential exposure scenarios with quantitative methods. Licensees also may be able to provide qualitative arguments to demonstrate that the dose from certain exposure scenarios is bounded by the dose of higher level exposure scenarios (e.g., a residential gardening exposure scenario will bound the dose for the residential I-33

nongardening exposure scenario). The licensees should provide justifications on the basis, method, and results of their exposure scenario screening in their DP.

Even after screening the exposure scenarios, a licensee will likely be left with a few exposure scenarios that may require detailed analyses to determine which will result in the critical group.

For licensees with multiple radionuclides, determining the compliance exposure scenario commonly depends on the final mixture of radionuclides. This can provide a dilemma for licensees creating DCGLs. The licensee must show that the final concentrations at the site meet the dose criteria of 10 CFR Part 20, Subpart E. Two possible approaches that the licensee may use to show compliance are, but are not limited to, the following:

(1) Use the most limiting DCGL for each radionuclide, regardless of the exposure scenario, and use the sum of fractions, ignoring the exposure scenario basis for each DCGL. This approach requires limited justification. It will always either estimate the same dose as the individual exposure scenarios or overestimate the dose. Generally, it will greatly overestimate the dose for the individual exposure scenarios.

(2) Commit to demonstrating the final dose for each of the important exposure scenarios in the FSS reports. This approach will require the licensee to establish operational DCGLs to fully use MARSSIM (see Section 2.5).

The licensee needs to provide either a quantitative analysis of, or a qualitative argument discounting the need to analyze, all the exposure scenarios generated from the less likely but plausible land uses. The results of these analyses will be used by the staff to evaluate the degree of sensitivity of dose to overall exposure scenario assumptions (and the associated parameter assumptions). Analyses of less likely but plausible exposure scenarios are not meant to be worst-case analyses and should not use a set of worst-case parameters.

Selection of parameters for less likely but plausible exposure scenarios should be consistent with the guidance in this appendix. The reviewer will consider both the magnitude and time of the peak dose from these exposure scenarios. If the peak dose from the less likely but plausible land use exposure scenarios is significant, the licensee would need to provide greater assurance that the exposure scenario is unlikely to occur, especially during the period of peak dose. The licensee may be able to show that the compliance exposure scenario bounds the results of all or many of the exposure scenarios associated with the less likely but plausible land uses.

I.3.4 Generic Examples The following examples describe situations where the default pathways may be removed or modified. Note that the examples assume that an adequate level of justification has been provided by the licensee.

I.3.4.1 Removal of Groundwater Pathways A licensee has extensive contamination of the upper soil horizon and the upper aquifer, which is unconsolidated, and it wishes to remove the groundwater pathway because the upper aquifer would not be used as a water source. The aquifer shows relatively high levels of microbial activity, turbidity, and nitrates. In addition, adjacent to the site is a small patch of wetlands that shows a great deal of communication with the upper aquifer. The potential yield rate of the upper aquifer is sufficient for domestic use, but there is a better quality, confined aquifer, with a horizon at a depth of approximately 30 meters (100 feet). The licensee has also demonstrated I-34

that the deeper aquifer will not become contaminated from the upper aquifer. Considering all of these reasons in combination, it is questionable whether the upper aquifer would actually be used. Although it may be possible for someone to treat the contaminants and use the aquifer, there are better sources of water easily available. After consultation with the EPA and the State, it is agreed that it would be unreasonable to assume someone would use the upper aquifer as a water source. Therefore, the licensee is allowed to remove the groundwater pathway from the exposure scenario.

I.3.4.2 Exposure Scenario Development for Buried Residual Radioactivity EXAMPLE 1: SUBSURFACE SOIL A site has residual radioactivity buried at a few feet below the surface and the licensee is requesting unrestricted release. The residual radioactivity does not have enough highly energetic gamma emitters to result in an external dose in the current configuration. Two scenarios can be developed (without any other site-specific information): (1) leaching of the radionuclides to the groundwater, which is then used by a residential farmer, and (2) exposure to the buried residual radioactivity by house construction for a resident farmer with the displaced soil, which includes part of the residual radioactivity, spread across the surface. Exposure scenario 2 encompasses all the exposure pathways and, although not all of the source term is in the original position, leaching may occur both from the remaining buried residual radioactivity and the surface soil. In certain cases, alternative intrusion events could be more limiting and should be considered particularly if home construction into the buried residual radioactivity is precluded due to the depth of the residual radioactivity or presence of a cover. Appendix J of this volume describes in greater detail the consideration of residual radioactivity in subsurface soils.

EXAMPLE 2: EMBEDDED PIPING At another site, the licensee is requesting unrestricted release of its site. It is removing the buildings but is evaluating the need to remove the concrete pads, which have embedded piping that contains the residual radioactivity. Two exposure scenarios can be reasonably envisioned.

The first involves a resident farmer onsite, who builds a house on the concrete pad, without disturbing the embedded piping. Possible exposure pathways would be external dose from the piping and exposure to leached materials from the piping through groundwater use (e.g., drinking, irrigation). The second exposure scenario is similar to the building renovation exposure scenario discussed in Example 1, where the concrete pad and piping are removed from the site during a future construction project. The licensee should investigate both exposure scenarios to find the limiting scenario.

I.3.4.3 Exposure Scenario Development for Restricted Release For this example, the site restrictions planned for an alternate site include a restriction on the deed that the property can be used only for parkland, and an engineered cover is placed over the residual radioactivity. The engineered cover is contoured for use as parkland and has a vegetative cover (i.e., not a mound covered in rip-rap). Three exposure scenarios are easily envisioned for the restricted release analysis. The first is recreational use of the property as a city park or golf course, which would limit exposure scenarios to possible external exposure.

The second would involve offsite use of groundwater that contains radionuclides leached from the buried residual radioactivity. The default offsite user would be a resident farmer using the I-35

groundwater for all water needs. The third exposure scenario would be a worker maintaining the park.

The doses assuming the loss of institutional control (i.e., the deed restriction) immediately following license termination (or time=0 years) and degradation of the engineered cover also must be evaluated. Scenarios should consider how critical groups could be exposed to the residual radioactivity through disruptive events.

Consider a residential farmer who uses groundwater from the aquifer located under the site. An engineered cover may become compromised by the placement of buildings. The cover may still perform in some degraded function for some period of time. Whether buried residual radioactivity is transported to the surface by the construction of a basement under the resident farmers house would depend on the thickness of the engineered cover. If typical basement depths are deeper than the engineered covers thickness, some portion of residual radioactivity would be transported to the surface, mixed with the clean cover material, and spread over the site. If the typical basement depth is shallower than the engineered cover thickness, other disruptive events such as well construction or large-scale excavation of material may need to be considered to evaluate a scenario where residual radioactivity is brought to the surface (see Appendix J).

In another scenario, the engineered cover may become degraded from erosion and residual radioactivity redistributed through hydrological processes. The reasonableness of this scenario would depend on the thickness and erosion-resistance of the engineered cover (see Section 3.5 for additional information on performance assessments for engineered barriers).

I.4 Criteria to Establish Conceptual and Mathematical Models I.4.1 Introduction Analyzing the release and migration of radionuclides through the natural environment and engineered systems, at a specific site, requires the licensee to interpret the nature and features of the site so that the site can be represented by mathematical equations (i.e., mathematical models). This simplified representation of the site is commonly referred to as the conceptual model of the site.

Figure I.3 depicts the process of conceptual model development. In dose assessments, developing a conceptual model involves making an abstraction of site data into a form that is capable of being modeled. This development should generally involve making simplifying assumptions, including simplification of the appropriate governing equations, to reflect the physical setting. These simplifying assumptions are usually made in describing the geometry of the system, the spatial and temporal variability of parameters, the isotropy of the system, and the influence of the surrounding environment. The conceptual model should provide an illustration or description of site conditions, to show, or explain, contaminant distributions, release mechanisms, exposure pathways and migration routes, and potential receptors. In other words, the conceptual model should explain or illustrate how radionuclides enter, move through and/or are retained in, and leave the environment.

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Site Data Engineering Scenario Designs Mental Construct Conceptual Model Mathematical Models Computer code Figure I.3 Conceptual Model Development As shown in Figure I.4, developing a conceptual model at a site is Step 3 of the decommissioning decision framework (see Figure 1.2 of this volume). Conceptual model development follows after assimilation of site data (Step 1) and definition of scenarios (Step 2),

because information from these two steps feeds into its development. In other words, the conceptual model should be based on what is known about the site from data and information gathered as part of Step 1, and how the site evolves during the period covered by the analysis, based on the assumed land use defined under Step 2.

Mathematical models are a quantitative representation of the conceptual model. Because the conceptual model provides the linkage between site conditions and features (Steps 1 and 2) and the computer code(s) (with its associated mathematical models) used in the dose analysis (Step 4 of the decommissioning framework), it is a key step in a dose assessment and should not be taken lightly.

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Figure I.4 Decommissioning Decision Framework I.4.2 Technical Issues Uncertainties in conceptual models can be large and possibly even larger than uncertainties in parameters used in the analysis (James and Oldenburg, Linear and Monte Carlo Uncertainty Analysis for Subsurface Contaminant Transport Simulation, issued November 1997). Thus, conceptual model uncertainties can be a significant source of uncertainty in the overall dose assessment. Uncertainties in the conceptual model(s) are generally caused by incomplete knowledge about the natural system being analyzed and differing views about how to interpret data representing the system.

Development of conceptual models is a subjective process based on the interpretation of limited (or in most cases, sparse) site data. From these limited data, the licensee should determine the key processes and features at the site and how they are likely to affect the movement of radionuclides through the environment. Because the conceptual model of the site is based on incomplete information, it is possible that multiple interpretations of the same data can be derived. A licensee should also determine the appropriate level of simplification acceptable for representing the site. An overly simplified conceptual model may leave out key site features or conditions that are important in estimating where radionuclides are likely to be transported (thus, where people might be exposed) and when they might get there (thus, the radionuclide concentration when it arrives). On the other hand, an overly complex conceptual model may introduce unnecessary uncertainty and costs into the analyses. As a broad example, simple I-38

models contained in screening codes may oversimplify features and processes at a specific site.

The licensee also should ensure that the model provides the appropriate level of detail. It is important that the conceptual model have sufficient detail and scope for a license reviewer to be able to assess the appropriateness of the computer codes used in the analysis and the defensibility of the assumptions made. In summary, key issues in developing and presenting the conceptual model are (1) identifying the important site features and processes that need to be included in the conceptual model, (2) deciding among possible competing interpretations of the site data, and (3) determining the level of detail needed to describe those features and processes.

Some important staff insights gained from reviewing decommissioning dose assessments and performance assessments are summarized below. Insights are grouped into those related to (1) model abstraction and (2) model simplification. Issues associated with model abstraction and simplification are generally important for more complex sites, while screening models and codes discussed in the next section I.4.3 may be appropriate for simpler sites provided the underlying assumptions of the conceptual models for those models and codes are consistent with site conditions.

I.4.2.1 Model Abstraction Some level of abstraction is required to translate the concepts of a conceptual model into mathematical terms. An abstracted model can be something as simple as a data value or lookup table. The methods to produce abstracted models are often not particularly rigorous and can introduce quantified uncertainties and biases. Important aspects can be lost in the abstraction process, and, for this reason, the reduction is often undertaken in such a way that it produces a conservative result. It is therefore important to clearly document the model abstraction and recognize the potential impacts due to the abstraction.

Insights gained on abstracting models of hydrogeological systems include the following:

  • Very complicated sites may not need to have all geological and hydrogeological features and processes represented in the model, so that model abstraction may be less problematic than originally estimated.
  • Many of the processes that govern transport of radionuclides in the unsaturated zone are essentially the same as those that govern transport in the saturated zone. However, effective hydraulic properties have a nonlinear dependence on soil moisture content, so that flow in the unsaturated zone can be strongly influenced by extreme, but not necessarily uncommon, conditions.

Additional important insights gained from the NRC review of examples of model abstraction include the following:

  • Code selection for a particular modeling exercise should be judicious to ensure that code limitations do not lead to nonconservative or unrealistic dose modeling predictions (e.g., lack of consideration of complex source terms, important transport processes such as diffusion and dispersion, or complex flow systems).
  • Model abstractions can successfully represent the essential elements of the system being simulated, increase computational efficiency, enable a more complete evaluation I-39

of model and parameter uncertainty, and lead to a better understanding of the system being simulated.

  • Additional complexity can and should be added to a model if certain processes or parameters are found to be important (i.e., if the additional complexity significantly influences the results).
  • Important pathways of exposure and scenarios should be evaluated to ensure that the most limiting pathways and scenarios are considered. In some cases, it is not intuitive to determine the most risk-significant exposure scenario.
  • Source term assumptions may have a significant impact on modeling results (e.g., source orientation and geometry, source distribution, and elevation of release).

I.4.2.2 Model Simplification Model simplification is the process for reducing the complexity of a numerical model into a simpler numerical model while still maintaining the validity of the simulation results.

NUREG/CR-6884, Model Abstraction Techniques for Soil-Water Flow and Transport, issued December 2006 (NRC 2006), presented a systematic and objective approach to model simplification relevant to subsurface flow and transport modeling. This approach included (1) justifying the need for the model simplification, (2) reviewing the context of the modeling problem, (3) selecting applicable model simplification techniques, (4) determining model simplification directions, and (5) simplifying the complex model in each direction. When performing a model simplification, various categories of techniques are relevant to the subsurface flow and transport modeling. These categories include selecting from a predefined hierarchy of models, changes in spatial dimensionality (e.g., three dimensional to two dimensional), and scale change, including upscaling, aggregation, and metamodeling.

The model simplification process starts with an existing complex model that can be calibrated and used in simulations. Justifying the need for the model simplification is usually associated with unsatisfactory results from the complex model in some way. This can either be related to the complex model being too expensive, too large, and too difficult to run and calibrate; input is too hard to obtain; or output is difficult to understand. After the model simplification process, the resulting model output should provide information that is both necessary and sufficient to make a decision on the issue(s) of interest.

Some important insights gained from the NRC review of previous examples of model simplifications include the following:

  • While in many cases, simpler models and codes developed to demonstrate compliance with regulatory criteria are purposefully constructed to err on the side of higher dose, the use of simpler models and codes does not guarantee conservative results. Irrespective of the level of complexity of the model used to facilitate decision-making, the analyst should ensure that the model is adequately supported and that the impact of model simplification is well understood.
  • The necessary level of complexity of a model is dependent on many factors, including the complexity of the site, uncertainty in site parameters and processes, the questions that are being addressed by the models, and the safety margin.

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  • With respect to physical dispersion, models that consider dispersion can generally provide more realistic estimates of concentrations and dose compared to simpler models that do not consider dispersion. For example, more complex three-dimensional groundwater flow and transport models can explicitly consider longitudinal and transverse dispersion, while one-dimensional models may consider no dispersion at all or longitudinal dispersion, or only implicitly consider dispersion. Consideration of physical dispersion can lead to higher or lower concentrations (and the resultant dose),

or to no significant impact at all, depending on such factors as the radionuclides driving the results, the presence of multiple sources, receptor locations, numerical model construction decisions, and site complexity (e.g., heterogeneity). Because, in many cases, dispersion assumptions can significantly affect the results, the treatment of dispersion in a model should be adequately studied to ensure that concentrations and dose are not underestimated.

  • The essential elements of a complex, three-dimensional groundwater flow and transport model can be identified and considered in a simpler flow and transport model. However, data of sufficient quality needed to calibrate the complex model must be available and a sufficient number of intermediate outputs and observations extracted to ensure that the simpler model accounts for risk-significant processes. If this approach is used, the model abstraction can sometimes be helpful in identifying and communicating key parameters and processes that are difficult to identify or explain using the complex model alone.

For complex sites, when relatively simple models are used to demonstrate compliance with regulatory criteria, the analyst should provide sufficient information (e.g., relatively more and less complex model comparisons) to show that the model simplifications do not lead to a significant loss of fidelity in the results that could be important to decision-making. Appendix F contains additional information about conceptual site model development, model abstraction and simplification pertaining to hydrological models.

I.4.3 Recommended Approach I.4.3.1 Screening An acceptable dose assessment analysis need not incorporate all the physical, chemical, and biological processes at the site. The scope of the analysis, and accordingly, the level of sophistication of the conceptual model, should be based on the overall objective of the analysis.

A performance assessment conceptual model can be simple, if it still provides satisfactory confidence in site performance. For an initial screening analysis, little may be known about the site from which to develop a conceptual model. Computer codes used for screening analyses are generally intended to provide a generic and conservative representation of processes and conditions expected for a wide array of sites. Accordingly, the generic conceptual model in such codes may not provide a close representation of conditions and processes at a specific site.

Such a generic representation is still acceptable, as long as it provides a conservative assessment of the performance of the site.

In general, the conceptual models within DandD are expected to provide a conservative representation of site features and conditions. Therefore, for screening analyses, the NRC staff should consider such generic conceptual models to be acceptable, provided it is acceptable to assume that the initial radioactivity is contained in the top layer (building surface or soil) and the remainder of the unsaturated zone and groundwater are initially free of residual radioactivity. In I-41

using DandD for site-specific analyses, it is important to ensure that a more realistic representation of the site that is consistent with what is known about the site would not lead to higher doses. Table I.5 lists some site features and conditions that may be incompatible with the generic conceptual models within DandD. The relative importance of the incompatibilities varies with the exposure scenario and radionuclides involved. More information on the assumptions of the model is available in the development documentation (e.g., the NUREG/CR-5512 series).

For any site where it is known that one or more of these conditions or features are present, the licensee should provide an appropriate rationale on why the use of DandD should not result in an underestimation of potential doses at the specific site.

As an example, DandD considers only the inhalation dose from particulates in the air and does not consider the loss of H-3 and C-14 from the soil to the air as a gas or vapor. To adjust results from the DandD resident farmer exposure scenario for analyzing sites that contain either H-3 or C-14 (Haaker, Upper bound for inhalation dose from carbon-14 vapor and tritium vapor, published June 1999) (Haaker, 1999), (1) determine the area of the contaminated zone, (2) run DandD for the site with only H-3 or C-14, (3) read the associated activity ratio factor for the given area from Figure I.5, and (4) estimate the potential missed dose by multiplying the inhalation dose calculated from DandD by the activity ratio factor to account for the dose associated with the gas or vapor phase. Due to the low risk-significance of the inhalation pathway dose for the resident farmer scenario used to derive the screening values, no change was made to the published screening values in Table H.2 to address the missed dose for C-14 and H-3.

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Table I.5 Site Features and Conditions that may be Incompatible with those Assumed in DandD

  • sites with highly heterogeneous radioactivity
  • sites with wastes other than soils (e.g., slags and equipment)
  • sites that have multiple source areas
  • sites that have contaminated zones thicker than approximately 15 cm (6 in.)
  • sites with chemicals or a chemical environment that could facilitate radionuclide releases (e.g., colloids)
  • sites with soils that have preferential flow conditions that could lead to enhanced infiltration
  • sites with a perched water table, surface ponding, or no unsaturated zone
  • sites where the groundwater discharges to springs or surface seeps
  • sites with existing groundwater contamination
  • sites where the potential groundwater use is not expected to be located immediately below the contaminated zone
  • sites with significant transient flow conditions
  • sites with significant heterogeneity in subsurface properties
  • sites with fractured or karst formations
  • sites where the groundwater dilution would be less than 2,000 m3 (70,000 ft3)
  • sites where the overland transport of contaminants is of potential concern
  • sites with radionuclides in soil that may generate gases (i.e., H-3 or C-14)
  • sites with stacks or other features that could transport radionuclides to result in a higher concentration offsite than onsite I-43

Activity Ratio of Vapor to Particulate 1E+06 Activity Ratio 1E+05 1E+04 1E+03 10 100 1000 10000 100000 1000000 Contaminated area (m2)

Figure I.5 Activity Ratio of Vapor to Particulate as a Function of Contaminated Area (Used to Account for Missed H-3 and C-14 Inhalation Dose from Vapor Phase in DandD Residential Exposure Scenario)

I.4.3.2 Site-Specific Analyses For site-specific analyses, the intent is to provide a more realistic assessment of doses based on more site-specific information and data. Presumably for such analyses, more is known about the site from which to develop a conceptual model. For site-specific analyses, the licensee should provide a schematic or verbal description of the problem that it is attempting to analyze.

Even when using a computer code that has a predefined conceptual model, it is important for the licensee to identify any site features or conditions that may differ from those assumed in the code. In developing a site-specific conceptual model or identifying potential limitations with a predefined conceptual model, the licensee should consider the issues listed in Table I.6.

Because conceptual models are developed based on limited data, in most cases, more than one possible interpretation of the site can be justified based on the existing data. The licensee should address this uncertainty by developing multiple alternative conceptual models and proceeding with the conceptual model(s) that provides the most conservative estimate of the dose and yet is consistent with the available data. Consideration of unrealistic and highly speculative conceptual models should be avoided. Consistent with the overall dose modeling framework of starting with simple analyses and progressing to more complex modeling, as warranted, it may be advisable for the analyst to begin with a simple, conservative analysis that incorporates the key site features and processes and progress to more complexity only as I-44

merited by site data. It is important to stress that a simple representation of the site, in itself, does not mean that the analysis is conservative. It is incumbent on the licensee to demonstrate that its simplification is justified, based on what is known about the site and the likelihood that alternative representations of the site would not lead to higher calculated doses.

Table I.6 Issues to be Considered in Developing a Site-Specific Conceptual Model

  • whether a more realistic representation of the site would lead to higher doses
  • whether the conceptual model accounts for the most important physical, chemical, and biological processes at the site
  • whether the conceptual model adequately represents responses to changes in stresses
  • whether the conceptual model includes consistent and defensible assumptions In general, there are two primary areas of the dose analysis where the conceptual model is expected to change from one site to another; these are related to the source term and environmental transport. Aspects of the analysis related to the exposure pathways in the biosphere and dosimetry are largely determined by the exposure scenario and the assumed behavior of the critical group. Accordingly, models related to the exposure pathways in the biosphere and dosimetry should not change from one site to another, unless there is a significant change in the exposure scenario and associated critical group. The principal environmental transport pathways that should have to be considered in a dose assessment are groundwater (including transport through the unsaturated zone), surface water, and air.

The conceptual model of the source area should describe the contaminants and how they are likely to be released into the environment. Specifically, it should describe key features and processes such as the infiltration of water into the source area, the geometry of the source zone, the distribution of contaminants, release mechanisms, the physical form of the contaminants, near-field transport processes, and containment failure. If the contaminants are assumed to be uniformly distributed, this is an important assumption that needs to be justified because, in general, contaminants may not be uniformly distributed (see discussion under Section I.2 of this appendix). The source description should clearly identify how the contaminants are assumed to be released from the media. Common release mechanisms are diffusion, dissolution, surface release, and gas generation. The source description should also identify key processes and features that may retain or limit the release of contaminants from the source area (e.g., solubility and sorption). In addition, the description of near-field transport should state assumptions made about the dimensionality. In general, the assumption of one-dimensional vertical flow should be appropriate, unless there is some type of barrier present that may hinder flow in the vertical direction. The description of the source term should also I-45

describe failure mechanisms for any containment (e.g., corrosion, concrete degradation, or cover degradation), if containers or other forms of containment are present.

With respect to source term models, release mechanisms or models include simple wash off, solubility-limited release due to waste form dissolution, and diffusion-controlled release.10 Release rates are typically expressed as a fractional release rate. Dissolution/precipitation is more likely to be the key process in locations such as at a point source, an area where high contaminant concentrations exist, or where steep pH or oxidation-reduction (redox) gradients exist. Adsorption/desorption will likely be the key process controlling inorganic contaminant migration in areas where the naturally present constituents are already in equilibrium and only the anthropogenic constituents (contaminants) are out of equilibrium, such as in areas far from the point source (EPA, 1999a; and EPA, 1999b).

Use of a Kd model results in the release of residual radioactivity decreasing over time as the source is depleted. Figure I.6 illustrates the depletion of a uranium source for two different release models: (1) solubility-limited release at a value of 3.5x10-05 moles per liter (mol/L), and (2) desorption from a solid for six different assumed Kds (0.1, 1, 10, 50, 100, 200, 500, and 1,000 L/kilogram (kg)). The total source is depleted more rapidly when solubility control is assumed (at a solubility of 3.5x10-05 mol/L) compared to a leaching model using a Kd greater than around 50 L/kg, although the release rates may be higher initially for the Kd model. At lower Kds (i.e., less than about 50 L/kg), Figure I.6 shows the sorption model would deplete the total source more quickly than the solubility model. The Kd below which source depletion is more rapid than solubility limited release is a function of the bulk soil density, effective porosity, assumed solubility limit, and source inventory. Important characteristics of the curves shown in Figure I.6 are the following:

  • Release rates (i.e., mass released per unit time) in a solubility-limited model are constant over time and the source decay curve is linear.
  • The fractional release rate (i.e., fraction of source released per unit time) in a Kd model is constant over time (i.e., quantity released decreases over time) and the release curve is similar to an exponential, first order decay curve.

10 DUST-MS (Sullivan, 1996) has several models available to characterize the source term or rate of release of residual radioactivity from the source to infiltrating groundwater. The release rate depends upon the physical and chemical form of the radionuclides in the wastes. DUST-MS has four process models including, rinse with partitioning, diffusion, uniform degradation, and solubility-limited release. RESRAD-OFFSITE 4.0 also offers more complex source term models compared to RESRAD-ONSITE including many of the models in DUST-MS.

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Figure I.6 Source Depletion for Solubility and Kd Release Models (for Six Values of Kd Expressed in L/kg)

The conceptual model of the groundwater pathway should describe how contaminants could migrate through the unsaturated and saturated zones to potential receptors (e.g., a well, spring, or surface water). Essential features that should be included in the conceptual model include hydrostratigraphic units; boundary conditions; the physical and chemical form of the residual radioactivity (e.g., dissolved, suspended sediment, gas, speciation, complexation), the structural features of the geology (i.e., those that influence contaminant transport such as fractures, faults, and intrusions), and geochemical conditions and gradients important to contaminant transport.

Important processes that should be characterized include the dimensions and state conditions (e.g., steady state) of flow, dimensions and state conditions of transport (e.g., dispersion),

chemical and mass transfer processes (e.g., sorption, precipitation, complexation), and transformation processes (e.g., radioactive ingrowth and decay). Although contaminant migration through both the unsaturated and saturated zones is best represented in three dimensions, it may be appropriate to assume only one or two dimensions, if this provides a more conservative representation of contaminant migration, or if it can be demonstrated that migration in one or more other directions is not expected to result in exposure to potential receptors.

The conceptual model of the surface water pathway should describe potential contaminant migration to potential receptors through surface water bodies, such as lakes, streams, channels, or ponds. Essential features that should be included in the conceptual model include: the geometry of the surface water body (i.e., boundaries and boundary conditions), the physical I-47

form of the contaminants (e.g., dissolved or solid), and physical and chemical properties. Key processes that should be described include: the dimensions and state conditions of flow and transport, chemical and mass transfer processes (e.g., sorption, precipitation, volatilization), and transformation. One key boundary condition that should be described is how the contaminants are expected to initially mix or interact with the surface water.

The conceptual model of the air pathway should describe potential contaminant migration through the air to potential receptors. Essential features that should be included in the conceptual model are similar to those for the other environmental pathwaysnamely, the geometry (i.e., boundaries and boundary conditions), form of contaminants (e.g., particulates or gases), and physical and chemical properties. Key processes that should be described include the dimensions and state conditions of flow and transport and the transformation processes.

SITE-SPECIFIC COMPUTER CODES Three common computer codes used for site-specific analyses are RESRAD-ONSITE, RESRAD-OFFSITE, and RESRAD-BUILD.11 All have predefined conceptual models.

Therefore, in using these codes, it is important for the licensee to demonstrate that key site features and conditions are consistent with the modeling assumptions within the codes or, where they are not consistent, the analysis may not result in an underestimation of potential doses. Additional information is provided in Section I.5.3 regarding the built-in conceptual models in these codes.

LIMITATIONS OF SITE-SPECIFIC COMPUTER CODES In general, the conceptual models within the RESRAD family of codes are expected to provide an acceptable generic representation of site features and conditions. Table I.7 lists some specific site features and conditions that may be incompatible with this generic representation.

At any site where it is known that one or more of these conditions or features are present, the licensee should provide appropriate justification for use of the computer code.

11 The RESRAD family of codes are commonly used by decommissioning licensees to perform dose modeling to support license termination under 10 CFR Part 20, Subpart E. It is important to note that most of the information on conceptual models in this section pertains to RESRAD-ONSITE and RESRAD-BUILD. While less information is provided about RESRAD-OFFSITE, RESRAD-OFFSITE typically has the same functionality as RESRAD-ONSITE and much more.

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Table I.7 Site Features and Conditions that may be Incompatible with the Assumptions Made in RESRAD-ONSITE

  • sites with highly heterogeneous radioactivity
  • sites with wastes other than soils (e.g., slags and equipment)
  • sites with multiple source areas
  • sites that have chemicals or a chemical environment that could facilitate radionuclide releases
  • sites with soils that have preferential flow conditions that could lead to enhanced infiltration
  • sites where the groundwater discharges to springs or surface seeps
  • sites where the potential groundwater use is not expected to be located in the immediate vicinity of the contaminated zone
  • sites with significant transient flow conditions
  • sites with significant heterogeneity in subsurface properties
  • sites with fractured or karst formations
  • sites where overland transport of contaminants is of potential concern
  • sites with stacks or other features that could transport radionuclides off the site at a higher concentration than on site I.4.4 Generic Examples I.4.4.1 Screening A hypothetical research and development facility is authorized to use radiological chemicals through an NRC license. Because the research and development facility plans to discontinue its use of radioactive material, it wants to decommission the facility and terminate its license. An HSA reveals that the use of radioactive material was limited to a single building within the facility. The floor area of the facility is estimated to be 560 m2 (6,000 ft2). The wall area is 430 m2 (4,600 ft2). In addition, an outside area of roughly 930 m2 (10,000 ft2) was used for dry storage of chemicals. A preliminary characterization program has determined that approximately 10 percent of the building floor area and 5 percent of the wall area are contaminated with Cs-137 and Co-60. Surficial soils covering an area of approximately 2,500 m2 (27,000 ft2) are contaminated from windblown dust and runoff from spills in the storage area. The soils are also contaminated with Cs-137 and Co-60.

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The licensee proposes to use a screening analysis, using DandD, to demonstrate compliance with the LTR. A building occupancy exposure scenario is assumed for the building and a residential farmer exposure scenario is assumed for the contaminated soils. Based on what is known about the site, the licensee certifies that the use of the generic conceptual models within DandD is appropriate for the analysis.

I.4.4.2 Site-Specific A hypothetical manufacturing facility has a former radioactive waste burial area that may be decommissioned for unrestricted release. Radioactively contaminated trash was previously buried in 0.2-m3 (55-gallon) drums, in trenches covering an area of roughly 2,000 m2 (22,000 ft2). The trenches, which are roughly 0.9 m (3 ft) deep, are covered with 1.2 m (4 ft) of native soil. A review of site operating records shows that the radionuclides of concern are natural uranium, enriched uranium, and natural thorium.

Based on information from the local county agricultural extension office and published reports, the geology and hydrogeology at the site are described as follows. This description shows that no site features or conditions in Table I.7 are present at this site.

The surface geology at the site contains 14 to 27 m (46 to 89 ft) of till consisting primarily of fine, silty sand to sandy silt with narrow, discontinuous sand lenses.

Sandstone bedrock underlies the unconsolidated till. A shallow unconfined aquifer occurs in the unconsolidated till. The average depth to the water table ranges between 3 and 4 m below the land surface. The mean horizontal hydraulic conductivity is roughly 60 m/y (197 ft/y). The average vertical hydraulic conductivity of the till is estimated to be an order of magnitude less. The hydraulic gradient is estimated to range between 0.006 and 0.021. The mean precipitation at the site is roughly 0.8 m/y (30 in./y). The site is located in the reach of a surface water drainage basin that has a drainage area of approximately 500,000 m2 (5.4 million ft2).

The licensee assumes a residential farmer exposure scenario as a reasonable future land use and proposes to use the RESRAD-ONSITE computer code for the dose analysis. Because the contaminated media is trash, an assumption is made that the trash degrades and becomes indistinguishable from soil. In addition, the metal drums are assumed to have degraded away.

Given the relative short lifespan for metal drums and the long half-life of the radionuclides, this should be a reasonable assumption. The cover is also assumed to be breached through the construction of a basement for the house. The contaminated soil is assumed to be uniformly mixed with the excavated cover. Because the trash is assumed to be indistinguishable from soil, it is also assumed that, once the cover is breached, the future hypothetical farmer may not recognize the contaminated material as contaminated. The licensee also assumes that the hypothetical future well is located at the center of the residual radioactivity because of limited bases for assuming otherwise.

The licensee determines that the other aspects of conceptual models within RESRAD-ONSITE are acceptable for analyzing the problem.

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I.5 Criteria for Selecting Computer Codes/Models I.5.1 Introduction Dose assessment commonly involves the execution of numerical model(s) that mathematically represent the conceptual model of the contaminated site. The numerical models used to implement the mathematical equations are usually linked via the conceptual model and codified in a software package known as the code. The words code and model are frequently used to express the software package, including the embedded numerical models or the specific models contained in the code. For example, DandD code may refer to the software package, including the associated exposure models (e.g., the water use model, food-ingestion pathway model, inhalation-exposure model) embedded in the code. The DandD model may also refer to DandD software, the DandD conceptual model, or any of the numerical models, or the group of models used in the code (e.g., DandD groundwater model). Within the context of this volume, the word code will refer to the software package and the associated numerical models. However, the word model will refer to the mathematical representation of the conceptual model, including representation of the specific exposure scenario and pathways.

This section describes the process and criteria used in the selection of codes and models for dose assessments.

The codes and models used in the dose assessment can be either generic screening codes/models or site-specific codes/models. Regardless of the intent of the use of the code/model (e.g., for screening or site-specific analysis), the reviewer should ensure that the licensee properly documents and verifies the dose assessment codes/models and the associated databases in accordance with rigorous QA/QC criteria acceptable to the NRC.

Currently, the only acceptable generic screening code is DandD. If the licensee uses site-specific models/codes, it should justify the conceptual model used (see Section I.4.3.2 of this appendix). The NRC staff should also review the source term model(s), the transport models, the exposure models, and the overall dose models and assess the QA/QC documentation and level of conservatism of any alternative code/model.

This section describes the generic issues associated with the selection of the screening and site-specific codes/models that the NRC staff may encounter and recommends approaches and criteria for its acceptance of the codes/models. In addition, this section presents a generic description of common dose assessment codes, DandD, RESRAD-ONSITE and RESRAD-BUILD. The NRC developed or funded (in part) the development of these codes. In addition, the NRC staff and licensees have used these codes to demonstrate compliance with the dose criteria in Subpart E.

I.5.2 Issues in Selection of Computer Codes/Models The major issues associated with the selection of computer codes/models include the following:

  • Generic criteria for the selection of computer codes/models: This issue pertains to the NRC staffs review criteria of code aspects related to QA/QC requirements, specifications, testing, verification, documentation, interfacing, and other features related to uncertainty treatment approaches.
  • Acceptance criteria for the selection of site-specific codes/models: This issue pertains to the NRC staffs review of additional specific requirements for justifying the I-51

use of the conceptual model, the numerical mathematical models, the source term model and its abstraction, and the transport and exposure pathway models.

  • Options for the selection of deterministic or probabilistic site-specific codes: This issue pertains to the NRC staffs review of the justification to support the decision to use either of these two approaches.

A generic description of the DandD Version 2 is presented below to familiarize users with this code. Further, the rationale for the development of DandD Version 2 and the issue of excessive conservatism in DandD Version 1 are addressed. A description of the approaches to minimize such excessive conservatism, using DandD Version 2, site-specific input data, or use of other models/codes is included.

For a site-specific analysis, the NRC staff should accept any model or code that meets the criteria described below in Generic Criteria for the Selection of Codes/Models. However, the staff is expected to conduct a more detailed and thorough review of less common codes/models (e.g., codes other than DandD and the RESRAD family of codes), specifically those developed by licensees. The NRC sponsored development of the probabilistic RESRAD-ONSITE (Version 6) and RESRAD-BUILD (Version 3) codes for site-specific analysis12. These have already been reviewed for QA/QC and are acceptable.

The selection of appropriate models/codes for complex sites may also present challenges. For example, sites with multiple source terms, with significant groundwater or surface water contamination, or sites with existing offsite releases, may require more advanced codes/models than commonly used codes such as DandD or RESRAD-ONSITE13. Complex sites may also include sites with engineered barrier(s), or with complex hydrogeological conditions such as highly fractured geologic formations. Because of site complexity and variability, there are no standard dose analysis review criteria for these sites.

I.5.3 Recommended Approach I.5.3.1 Generic Criteria for Selection of Codes/Models The generic criteria under this subsection pertain to the NRC staff review of codes/models other than commonly used codes; specifically, those developed or modified by the NRC staff (i.e., other than DandD, RESRAD-ONSITE, and RESRAD-BUILD). The reviewer should use the generic criteria when the codes/models have no readily available documentation of testing, verification, and QA/QC review. In this context, the reviewer should use the following generic criteria in reviewing the codes/models selected for the dose assessment:

  • The NRC staff should review the adequacy and completeness of the database available on QA/QC aspects of the code/model. The QA/QC database should be comparable to the NRCs QA/QC requirements (NUREG/BR-0167, Software Quality Assurance Program and Guideline, issued February 1993 (NRC, 1993b) and NUREG-0856, Final Technical Position on Documentation of Computer Codes for High Level Waste Management, issued June 1983 (NRC, 1983)). The QA/QC should include information on mathematical formulation, code/model assumptions, consistency of the pathways 12 Newer versions of the code are also acceptable for use.

13 RESRAD-OFFSITE 4.0 was released in 2020 and contains more complex source term, hydrogeological, air, and surface water transport models compared to RESRAD-ONSITE.

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with the assumed conceptual model(s) used in the code, and accuracy of the software to reflect the models mathematical formulation and correct representation of the process or system for which it is intended.

  • The NRC staff should ensure that the software used for the code is in conformance with the recommendations of the Institute of Electrical and Electronics Engineers (IEEE) Standard 830-1984, IEEE Guide for Software Requirement Specifications.
  • The NRC staff should review the adequacy and appropriateness of the code/model documentation with regard to (1) software requirements and intended use, (2) software design and development, (3) software design verification, (4) software installation and testing, (5) configuration control, (6) software problems and resolution, and (7) software validation.
  • For uncommon codes/models, the NRC staff should review code data, including (1) a software summary form, (2) a software problem/change form, (3) a software release notice form, and (4) a code/model users manual, which covers code technical description, software source code, functional requirements, and external interface requirements (e.g., user interface, hardware interface, software interface, and communication interface), if necessary.
  • The NRC staff should review the conceptual model of the selected code to ensure compatibility with the specific site conceptual model, including the pathways and the exposure scenario. The source term assumptions of the selected code should also be compatible with the site-specific source term. The staff may accommodate minor modifications in the source term conceptual model, as long as the basic model assumptions are not violated.
  • The NRC staff should verify that the exposure scenario of the selected code is compatible with the intended scenario for the site. For example, models/codes designed for the onsite exposure scenario may not be appropriate for assessment of an offsite exposure scenario.
  • The NRC staff should review the selected model/code formulation to account for radionuclide decay and progenies. The code should have proper and timely formulation, as well as linkages of decay products with the receptor location and the transport pathways, via corresponding environmental media.
  • The NRC staff should examine the documentation of the selected code/model performance; specifically, test and evaluation, as well as a code comparison with commonly used (accepted) codes and models (e.g., DandD and the RESRAD family of codes). The staff should also review documentation on code/model verification, if available, to support decisions for code acceptance.
  • The NRC staff should review code/model features of the sensitivity/uncertainty analysis to account for variability in the selection of input parameters and uncertainty in the conceptual model and multiple options for the interpretation of the system.

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I.5.3.2 Acceptance Criteria for Selection of Site-Specific Codes/Models This issue involves the NRC staffs review of additional requirements supporting the justification for using the conceptual model, the numerical mathematical models, the source term model and its abstraction, and the transport and exposure pathway models.

CONCEPTUAL MODELS The NRC staff review should compare the conceptual model for the site with the conceptual model(s) in the selected code, to ensure compatibility with site-specific physical conditions and pathway assumptions for the critical group.

NUMERICAL MATHEMATICAL MODELS The staff should review the equations used in the code to implement the conceptual model and the numerical links between mathematical models to ensure correctness and consistency. For codes developed or modified by the NRC staff (e.g., DandD, RESRAD-ONSITE, and RESRAD-BUILD), minimal review is needed, because the NRC revised these codes and examined them early for consistency with its QA/QC requirements. For less commonly used codes, or codes developed locally by user(s), the NRC staff should verify the numerical mathematical models, including the numerical links between these models. In this context, the reviewer may examine, if necessary, each mathematical model used for the specific transport-exposure pathway, to ensure that the code is designed for its intended use.

SOURCE TERM MODELS The NRC staff should review the source term model(s) used for the specific site. In this context, the review should include the following source term aspects:

  • Building Occupancy Exposure Scenario Source Term: The NRC staff should review the HSA and other relevant data on the extent of the source and its depth (e.g., within 1 to 10 millimeters (mm) (0.04 to 0.39 in.) deep into the building surface or more). Based on this review, the reviewer should identify the source as surficial or volumetric and examine the assumptions made for the loose/fixed fractions of the source. The review should address the sources of residual radioactivity on surfaces that are not integral parts of the building (e.g., equipment, pipes, and sewer lines) separately, because the applicable model and exposure scenario could be different. Therefore, source term model assumptions for such surfaces should be reviewed on a case-by-case basis.

The NRC staff should also review the radionuclide mixture comprising the source and whether a constant ratio is assumed in the dose analysis, as well as determine if surrogate radionuclides are being used. The latter two situations may require additional NRC staff verification of the source definition and review of consistency with the intended final survey methodology.

The review should also include the use of multiple sources (e.g., multiple rooms).

Certain codes may provide an option to define multiple sources in various configurations, such as two to three rooms, with multiple-story buildings. The source term under these conditions allows for source depletion due to open air circulation and common ventilation. For example, the RESRAD-BUILD code model uses two- or three-room models with two- or three-story buildings, allowing for air exchange within the rooms, I-54

and source depletion. The review should include the indoor air-quality model (e.g., building ventilation and infiltration), and the indoor air-concentration model, as well as the adaptation of the air-quality model in the RESRAD-BUILD code, to ensure consistency with the site-specific conditions. The review should verify the input parameters associated with these models. The NRC staff may accept such site-specific source term models after an assessment of the compatibility of the source term model with the conceptual model of the site. The review should also include the physical parameters defining the source term, to ensure consistency with site-specific conditions, and the occupancy parameters, to ensure consistency with the exposure scenario.

  • Resident Farmer Exposure Scenario Source Term: The NRC staff should examine the licensees source information to determine if the nature and extent of residual radioactivity is consistent with the model/assumptions in the selected code. The review should include evaluation of the vertical and horizontal extent of residual radioactivity, to verify model assumptions on the area and thickness of the contaminated zone, and to determine if there is residual radioactivity in the subsurface and/or groundwater at the site. The DandD model assumes residual radioactivity is located in surface soils only (approximately 15 cm or less). For contaminated zone thicknesses significantly greater than 15 cm, DandD may underestimate the dose and justification would be needed to use DandD to assess the dose from subsurface residual radioactivity. The contaminated area and its shape should also be assessed to check for possible correction of the area and/or for geometry of the source. Additionally, the NRC staff should determine if credit is taken for a clean cover or a barrier over the contaminated zone and the basis for the modeling assumption. Support for the assumed level of performance of the engineered cover or barrier should be evaluated within the context of institutional control assumptions (e.g., active maintenance for the case where institutional controls are in effect; and only passive performance in the case when institutional controls are assumed to no longer be in effect). Although sensitivity analysis on the timing of failure of institutional controls can be performed, institutional controls should be assumed to fail at time=0 years. The evaluation should consider assumptions regarding the performance and degradation of the engineered covers and barriers during the compliance period (e.g., 1,000 years).

The NRC staff should also review the physical and chemical form of the source to evaluate the adequacy of the underlying soil leaching model(s) available in the selected code. This review should help assess the source mass-balance model and the transport model within the concerned environmental media. In addition, a review of these source term aspects would help establish consistencies for the selection of relevant parameters.

The review should include the horizontal distribution and homogeneity of the source and the variation of source concentration with depth. The NRC staff should use either an upper-bounding value for modeling the thickness or an area-weighted approach to calculate the representative thickness, if this approach does not significantly underestimate the potential dose (e.g., it may not be appropriate to represent an area with a thin vadose zone with a thicker vadose zone based on source thickness averaging, if the thicker vadose zone assumption could lead to longer transport times and lower exposure concentrations due to radiological decay). In certain cases, multiple sources or more advanced subsurface source term modeling may be needed to adequately assess dose; these would be evaluated on a case-by-case basis.

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TRANSPORT MODELS The transport models simulate transport mechanisms of contaminants from the source to the receptor. The NRC staff should review transport models for consistency and compatibility with respect to: (1) the source term, (2) the exposure scenario defined for the critical group, and (3) the simplified conceptual model, which describes site-specific physical conditions. The transport models may include the diffusive and advective transport of contaminants via air, surface water, and groundwater. The transport models can be overly simplified, using simple conservative assumptions such that minimal characterization data would be required to execute the model(s). Transport models can also be very complex, requiring advanced mathematical derivation and extensive site-specific, or surrogate data about the site.

For the building occupancy exposure scenario, the associated transport models (e.g., transport models for ingestion, inhalation, and direct exposure pathways) of the DandD code are simple and conservative. For example, the ingestion pathway depends on the effective transfer rate of the removable surface residual radioactivity from surfaces to hands and from hands to mouth.

The inhalation transport model depends largely on mechanical disturbance of the contaminated surface, resuspension of residual radioactivity in the air, and subsequent breathing of contaminated air. The external dose formulation assumes exposure from a nonuniform source of residual radioactivity on the walls, ceiling, and floor of a room. This model was found to be comparable to the infinite plane source for the building occupancy exposure scenario (NRC, 1992).

For the resident farmer exposure scenario, the associated DandD transport models include models of contaminants transport to groundwater, to surface water (e.g., three-box model that relies on transfer of contaminate through leaching), and to air (e.g., through dust mass loading and indoor resuspension). Transport models of contaminants via the air include dust loading, resuspension of contaminated soil, and use of mass loading factor for deposition. Transfer of contaminants from the soil/water to plants, fish, animals, and animal products are calculated using a water use model, along with transfer factors, translocation factors, and bio-accumulation factors. Separate models were used for C-14, and H-3, as described in NUREG/CR-5512, Volumes 1, 2, and 3. The RESRAD-ONSITE model can consider residual radioactivity in surface soils (approximately 15 cm) or thicker contaminated zones, with an idealized cylindrical shape of the contaminated zone14, and allows for a cover at the top of the contaminated zone, if appropriate.

In general, the NRCs review of the selected code should include transport models and the appropriateness of such models with respect to the site-specific conditions (e.g., area, source, unsaturated zone, and aquifer conditions). In addition, the staff should review, for compatibility and consistency, the transport model assumptions and the generic formulation pertaining to the applicable pathways of the critical group exposure scenario. The extent of the transport model review depends on the familiarity of the NRC staff with these models. Because they developed or modified certain commonly used codes/models (e.g., DandD, RESRAD-ONSITE, and RESRAD-BUILD), the NRC staff is more familiar with them, and they would require less of a review than for less common codes/models developed by users or other parties. The NRC review should also include updated new models or code versions and studies on code/models testing, comparison, and verifications.

14 RESRAD-ONSITE considers a circular source by default; however, RESRAD-ONSITE has the capability of considering more complicated source geometries.

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The RESRAD-BUILD code is more advanced than the DandD code because it employs multiple sources and more advanced particulate air transport models. In other words, each contaminated location may be considered a distinct source. Depending on its geometric appearance, the source can be defined either as a volume, an area, or a point source. The RESRAD-BUILD code depends on erosion of the source and transport of part of its mass into the indoor air environment, resulting in airborne residual radioactivity. The RESRAD-BUILD model differs from DandD because it assumes air exchange among all compartments of the building. In other words, the model assumes that the airborne particulates are being loaded into the indoor air of the compartment and then transported to the indoor air of all compartments of the building. In addition to air exchange between compartments, the indoor air model simulates air exchange between compartments and the outdoor air. Descriptions of models pertaining to indoor air quality, air particulate deposition, inhalation of airborne dust, and ingestion of removable materials and deposited dust, were documented in an Argonne National Laboratory report ANL/EAD/03-1 (ANL 2003). The exposure pathways in the RESRAD-BUILD code include (1) the external exposure to radiation emitted directly from the source and from radioactive particulates deposited on the floors and exposure caused by submersion from radioactive particulates, (2) inhalation of airborne radioactive particulates, and (3) ingestion of contaminated material directly from the source, as well as airborne particulates deposited onto the surface of the building.

EXPOSURE PATHWAY MODELS The exposure pathway models pertain to the formulation of the links between the radiological source, the transport of contaminants within environmental media, the critical group location, and behaviors that lead to exposure to residual radioactivity through direct exposure, inhalation, and ingestion of contaminated water, soil, plants, crops, fish, meat, milk, and other dairy products. The NRC staff should review the conceptual model(s) that describe the human behaviors that lead, or control, the amount of exposure. Therefore, the occupational, behavioral, and metabolic parameters describing these models should be reviewed and compared with the default model exposure scenarios and associated parameters. The NRC staff should review exposure model(s) and associated parameters to ensure conservatism, consistency, and comparability with site-specific conditions and exposure scenario assumptions.

NUREG/CR-5512, Volumes 1, 2, and 3, provide detailed information on default parameters and approaches for changing parameters in dose modeling analysis.

I.5.3.3 Option for Selection of Deterministic or Probabilistic Site-Specific Codes Licensees may select either a deterministic analysis or a probabilistic approach for demonstrating compliance with the dose criteria in 10 CFR Part 20, Subpart E. A deterministic analysis uses single parameter values for each variable in the code. In contrast, parameter distributions are specified for uncertain variables in a probabilistic analysis, sets of parameter values are selected via sampling from the specified parameter distributions, many sets of parameter values called realizations are run through the model, and a distribution of results is generated and evaluated. Although deterministic sensitivity analysis can be conducted, a deterministic analysis gives more limited information on the uncertainty in the results, based on uncertainty in the parameter values, and is generally less efficient in identifying important model sensitivities. Therefore, the deterministic approach may require a more elaborate justification of code input parameter values and may require further analysis of doses using upper or lower bounding conditions.

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NRC-approved data sets for both DandD and the RESRAD family of codes are for the probabilistic calculation and not the deterministic mode.

Section I.7.3.2 of this appendix contains a detailed description of an NRC staff review for both deterministic and probabilistic analyses.

I.5.3.4 Modeling of Subsurface Residual Radioactivity For subsurface residual radioactivity (residual radioactivity deeper than approximately 15 cm (6 in.)), the NRC staff should review existing historical site data (including previous processes or practices) and site characterization data, to establish an adequate conceptual model of the subsurface source; specifically, the horizontal and vertical extent of residual radioactivity.

Section I.2.3.1 describes approaches for subsurface source term abstraction for dose modeling analysis. In some cases, the licensee may wish to develop multiple DCGLs for surface and subsurface DCGLs. Additional information on integration of dose modeling with radiological surveys for license termination is found in Appendix G.

I.5.3.5 Generic Description and Development of DandD The DandD code has two default land use exposure scenarios: a building occupancy and a resident farmer exposure scenario. The building occupancy exposure scenario is intended to account for exposure to both fixed and removable residual radioactivity within a building.

Exposure pathways included in the building occupancy exposure scenario include external exposure to penetrating radiation, inhalation of resuspended surface residual radioactivity, and inadvertent ingestion of surface residual radioactivity. The resident farmer exposure scenario is intended to account for exposure to residual radioactivity in soil. Resident farmer exposure scenario pathways include the following: external exposure to penetrating radiation; inhalation exposure to resuspended soil; ingestion of soil; and ingestion of contaminated drinking water, plant products, animal products, and fish. The predefined conceptual models within DandD are geared to assessing releases of radioactivity, transport to, and exposure along, these pathways.

For the building occupancy exposure scenario, DandD models external exposure to penetrating radiation as an infinite area source, using surface source dose rate factors from Federal Guidance Report No. 12: External Exposure to Radionuclides in Air, Water and Soil, issued September 1993 (EPA 1993). Exposure to inhalation of resuspended surface residual radioactivity is modeled as a linear static relationship between surface residual radioactivity and airborne concentrations. The model accounts for ingrowth and decay. Exposure to incidental ingestion of surface residual radioactivity is modeled with a constant transfer rate.

The generic conceptual models for the resident farmer exposure scenario are more complicated because of the large number of exposure pathways and considerations of release of radioactivity from the source area and transport of radionuclides in the environment. DandD models external exposure from volume soil sources when the person is outside as an infinite slab of residual radioactivity 15 cm (6 in.) thick, using dose rate factors from Federal Guidance Report No. 12 for volume residual radioactivity. When the person is indoors, exposure from external radiation is modeled in a similar manner, except the exposure is assumed to be attenuated through the use of a shielding factor (note: the higher the shielding factor, the lower the assumed attenuation). Exposure through ingestion of contaminated animal and plant products is modeled simply through the use of transfer factors. Instantaneous equilibrium is assumed to occur between radionuclide concentration in the soil and the concentration in plants, and between animal feed and animal products.

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The generic source term conceptual model in DandD assumes a constant release rate of radionuclides into the water and air pathways. Release of radionuclides by water is assumed to be downward and a function of a constant infiltration rate, constant contaminant zone thickness, constant moisture content, and equilibrium adsorption. DandD assumes that there are no radioactive gas or vapor releases. Release of radioactive particulates is assumed to be upward, instantaneous, uniform, and a function of a constant particulate concentration in the air and the radioactivity within the soil. Radionuclides in the contaminant zone are assumed to be uniformly distributed in a single soil layer, 15 cm (6 in.) thick. No transport is assumed to occur within the source zone, but radioactive decay is taken into account. In terms of containment, DandD assumes that there are no containers (or that they have failed), and that there is no cover over the contaminated zone.

The DandD generic conceptual model for the groundwater pathway assumes a single hydrostratigraphic layer for each of the unsaturated and saturated zones. The unsaturated zone (vadose zone) can be broken into multiple layers within DandD; however, each layer is assumed to have the same properties. For radionuclides entering the vadose zone, DandD accounts for adsorption-limited leaching by considering the vadose zone to behave as a well-mixed chemical reactor with a constant water inlet and outlet rate set at the infiltration rate.

Accordingly, it is assumed that the vertical saturated hydraulic conductivity of the unsaturated zone is greater than or equal to the infiltration rate (i.e., there is neither ponding nor runoff on the surface). The outlet concentration from one unsaturated zone layer to another is assumed to be a function of the constant infiltration rate, equilibrium partitioning, the thickness of the layer, a constant moisture content, and radioactive decay. Radionuclides entering the saturated zone are assumed to be instantaneously and uniformly distributed over a constant volume of water equivalent to the larger of either the volume of infiltrating water (i.e., the infiltration rate times the contaminated area) or the sum of the water assumed to be removed for domestic use and irrigation. Based on the default parameters in DandD, dilution in the groundwater pathway is based on the water use. No retardation is assumed to occur in the aquifer; however, radioactive decay is taken into account. A volume of contaminated water equivalent to the irrigation volume is assumed to be returned annually to the source zone. The concentration of radionuclides in the irrigation water is assumed to remain constant during the year.

Radionuclides deposited on the vegetation are assumed to be removed at a constant rate. The DandD groundwater model should generally provide a conservative representation of the groundwater system, because it allows very little dilution and nominal attenuation.

The generic surface water conceptual model in DandD assumes that radionuclides are uniformly mixed within a finite volume of water representing a pond. Radionuclides are assumed to enter the pond at the same time and concentration as they enter the groundwater.

Accordingly, there is assumed to be no transport of radionuclides through the groundwater to the pond and thus no additional attenuation (besides the initial groundwater dilution) is assumed for transport in the groundwater. The surface water model within DandD should provide a conservative dose estimate as long as a small volume is assumed for the surface water pond.

Because the parameters in DandD are selected to provide a conservative dose estimate, the generic conceptualization of the surface water pathway should generally provide a conservative representation of transport of radionuclides through it. Figure I.7 shows the generic groundwater and surface water conceptual model within DandD.

The generic conceptual model of the air pathway in DandD assumes an equilibrium distribution between radionuclides in the air and soil. The concentration in air is assumed to be a function of the soil concentration and a constant dust loading in the air. Accordingly, all radionuclides in the air are assumed to be in a particulate form. The air pathway model within DandD is very I-59

simple and should generally allow a conservative dose estimate, as long as a conservative particulate concentration is assumed. Because the default parameters in DandD are geared to be conservative, the air pathway in DandD should generally allow a conservative dose estimate.

Precipitation/Evaporation Irrigation Domestic Use I = Infiltration Box 1 Surface-Soil Ground-Water Well Thickness = H1 Layer Box 2 Unsaturated-Soil Thickness = H2 Layer Ground-Water Aquifer Surface-Water Pond Box 3 Figure I.7 DandD Conceptual Model of the Groundwater and Surface Water Systems (from NUREG/CR-5621)

PROBABILISTIC DANDD VERSION 2 The NRC staff developed a probabilistic DandD Version 2, which updates, improves, replaces and significantly enhances the capabilities of Version 1.0. In particular, Version 2 allows full probabilistic treatment of dose assessments, whereas Version 1.0 embodied constant default parameter values and only allowed deterministic analyses. DandD implements the methodology and information contained in NUREG/CR-5512, Volume 1 (NRC, 1992), as well as the parameter analysis in Volume 3 (NRC, 1999c), which establishes the PDFs for all of the parameters associated with the exposure scenarios, exposure pathways, and models embodied in DandD.

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Finally, DandD Version 2 includes a sensitivity analysis module that assists licensees and the NRC staff to identify those parameters in the screening analysis that have the greatest impact on the results of the dose assessment. Identification of risk-significant parameters helps licensees make informed decisions on the allocation of resources needed to gather site-specific information important to the compliance demonstration.

I.5.3.6 Generic Description of the RESRAD Family of Codes Argonne National Laboratory developed the RESRAD family of computer codes under th