NUREG-1757

Volume 2

Revision 2

Consolidated 

Decommissioning 

Guidance 

Characterization, Survey, 

and Determination of 

Radiological Criteria 

Draft Report for Comment

Office of Nuclear Material Safety and Safeguards
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SR-CR 10/2017
Consolidated 

Decommissioning 

Guidance 

Characterization, Survey, and 

Determination of Radiological 

Criteria

Draft Report for Comment

Manuscript Completed: September 2020 

Date Published: November 2020 

Prepared by: 

C.S. Barr, S. Clark,

G.C. Chapman, J.P. Clements,

D.W. Esh, R.W. Fedors,

A.M. Huffert, L.A. Kauffman,

M.M. LaFranzo, D.B. Lowman,

C.A. McKenney, L.L. Parks,

D.W. Schmidt*, A.L. Schwartzman,

and B.A. Watson

*retired

Office of Nuclear Material Safety and Safeguards

Cynthia Barr, NRC Project Manager

NUREG-1757

Volume 2

Revision 2
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iii

1 ABSTRACT

2 The U.S. Nuclear Regulatory Commission (NRC), Office of Nuclear Material Safety and 

3 Safeguards (NMSS), previously consolidated and updated numerous decommissioning 

4 guidance documents into a three-volume NUREG. This NUREG series is intended for use by 

5 the NRC staff, licensees, and others. The three volumes address the following topics:

6 (1) Decommissioning Process for Materials Licensees

7 (2) Characterization, Survey, and Determination of Radiological Criteria

8 (3) Financial Assurance, Recordkeeping, and Timeliness

9 The staff last updated Volume 2 of the NUREG series, entitled, “Consolidated Decommissioning 

10 Guidance: Characterization, Survey, and Determination of Radiological Criteria,” in September

11 2006. This volume provides guidance on compliance with the radiological criteria for license 

12 termination (License Termination Rule (LTR)) in Title 10 of the Code of Federal Regulations

13 (10 CFR) Part 20, “Standards for Protection against Radiation,” Subpart E, “Radiological Criteria 

14 for License Termination.” This guidance takes a risk-informed, performance-based approach to 

15 the demonstration of compliance. This guidance will help identify the information (subject 

16 matter and level of detail) needed to terminate a license and considers the specific 

17 circumstances of the wide range of NRC licensees. Licensees should use this guidance in 

18 preparing decommissioning plans (DPs), license termination plans, final status surveys, and 

19 other technical decommissioning reports for NRC submittal. The NRC staff will use the 

20 guidance in reviewing these documents and related license amendment requests. Volume 2 

21 applies to all licensees subject to the LTR (i.e., fuel cycle, fuel storage, materials, and reactor 

22 licensees).

23

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

25

26 • Dose Modeling—adds guidance on model abstraction and simplification, consideration

27 of elevated areas, use of distribution coefficients in dose modeling, and consideration of

28 uncertainty.

29 • “As Low As Is Reasonably Achievable (ALARA)” Analysis—updates guidance on

30 the ALARA analysis review particularly for restricted release based on lessons learned

31 from proposed restricted release scenarios.

32 • Composite Sampling—adds information on methodologies for incorporating composite

33 sampling strategies into final status survey plans and details when it would or would not

34 be appropriate to use composite sampling.

35 • Characterization—updates information in Appendix F on surface water and

36 groundwater characterization.

37 • Engineered Barrier Analysis—updates and reorganizes Section 3.5 and Appendix P;

38 includes new information on how ALARA is considered prior to engineered barriers for

39 restricted release and updates bibliography with new references on evaluation of

40 engineered performance and degradation.
iv

1 • Radiological Surveys—provides updated guidance on subsurface radiological surveys, 

2 including surveys associated with excavations and re-use of soils; and provides 

3 additional information on use and implementation of Scenario B. 

4 

5 

6 

7

8 

9 

• Lessons Learned—removes 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. 

10
v

1 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.

2

3 In September 2003, the U.S. Nuclear Regulatory Commission (NRC) staff in the Office of 

4 Nuclear Material Safety and Safeguards (NMSS) consolidated and updated the policies and 

5 guidance of its decommissioning program in a three-volume NUREG series, NUREG-1757, 

6 “Consolidated NMSS Decommissioning Guidance.” This NUREG series provides guidance on 

7 planning and implementing license termination under the NRC’s License Termination Rule 

8 (LTR) in Title 10 of the Code of Federal Regulations (10 CFR) Part 20, “Standards for Protection 

9 against Radiation,” Subpart E, “Radiological Criteria for License Termination”; complying with 

10 the radiological criteria for license termination; and complying with the requirements for financial 

11 assurance and recordkeeping for decommissioning and timeliness in decommissioning 

12 materials facilities. The staff periodically updates NUREG-1757 to reflect current NRC 

13 decommissioning policy.

14

15 In September 2005, the staff issued, for public comment, draft Supplement 1 to NUREG-1757, 

16 which contained proposed updates to the three volumes of NUREG-1757. Draft Supplement 1 

17 included new and revised decommissioning guidance that addresses some of the LTR 

18 implementation issues, which were analyzed by the staff in two Commission papers 

19 (SECY-03-0069, “Results of the LTR Analysis,” dated May 2, 2003; and SECY-04-0035, 

20 “Results of the License Termination Rule Analysis of the Use of Intentional Mixing of 

21 Contaminated Soil,” dated March 1, 2004). These issues include restricted use and institutional 

22 controls, onsite disposal of radioactive materials (10 CFR 20.2002, “Method for Obtaining 

23 Approval of Proposed Disposal Procedures”), selection and justification of exposure scenarios 

24 based on reasonably foreseeable future land use (realistic exposure scenarios), intentional 

25 mixing of contaminated soil, and removal of material after license termination. The staff also 

26 developed new and revised guidance on other issues, including engineered barriers. 

27

28 The staff received stakeholder comments on the draft NUREG and prepared responses to these 

29 comments. Stakeholder comments and responses are located in the Agencywide Documents 

30 Access and Management System at Accession Number ML062370521. Comments were 

31 addressed and updated sections from Supplement 1 were placed into the appropriate locations 

32 in revisions of Volumes 1 and 2 of NUREG-1757 (NUREG-1757, Volume 1, Revision 2; and 

33 NUREG-1757, Volume 2, Revision 1). The staff plans to revise Volume 3 of this NUREG series 

34 at a later date, and that revision will incorporate the Supplement 1 guidance that is related to 

35 Volume 3. 

36

37 The NRC has increased the use of risk information in its regulation of nuclear materials and 

38 nuclear waste management, including the decommissioning of nuclear facilities. The NRC’s 

39 risk-informed regulatory approach to the decommissioning of nuclear facilities represents a 

40 philosophy whereby risk insights are considered, together with other factors, to better focus the 

41 attention and resources of both the licensee and the NRC on the more risk-significant aspects of 

42 the decommissioning process and on the elements of the facility and the site that will most 
vi

1 affect risk to members of the public following decommissioning. This results in a more effective 

2 and efficient regulatory process. 

3

4 The information used to “risk inform” the decommissioning process typically comes from the 

5 results and findings of risk assessments or dose modeling. A risk assessment is a type of 

6 systematic analysis used to understand what can happen, how likely it is to happen, and the 

7 resulting consequences. Dose modeling is used to estimate potential dose to members of the 

8 public who may use the decommissioned site in the future following license termination. The 

9 end result of such assessments (e.g., the calculation of predicted doses from decommissioned 

10 sites) relates directly or indirectly to public health effects. The NRC staff has developed this 

11 guidance consistent with a risk-informed approach.

12

13 The primary decommissioning guidance documents used by licensees and the NRC staff are 

14 NUREG-1757 and NUREG-1700, Revision 2, “Standard Review Plan for Evaluating Nuclear 

15 Power Reactor License Termination Plans,” issued April 2018. NUREG-1537, “Guidelines for 

16 Preparing and Reviewing Applications for the Licensing of Non-Power Reactors,” issued 

17 February 1996, includes a section on decommissioning and license termination for nonpower 

18 reactors. Table 1 below describes the general applicability of these documents. 

19

20 Since the last revision of this volume in 2006, which addressed comments on draft Supplement 

21 1, the NRC staff and larger decommissioning community has gained experience on a number of 

22 technical issues for which guidance has been developed. Table 2 describes the most significant 

23 changes to the guidance in this volume to include new and updated information based on this 

24 experience. 

25

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

27 NUREG-1757 is intended for use by applicants, licensees, NRC license reviewers, and other 

28 NRC personnel. It is also available to Agreement States and the public. 

29

30 This NUREG is not a substitute for NRC regulations, and compliance with it is not required. The 

31 NUREG describes approaches that are acceptable to the NRC staff. However, methods and 

32 solutions different than those in this NUREG will be acceptable, if they provide a basis for 

33 concluding that the decommissioning actions are in compliance with NRC regulations.

34

35 _____________________

36 Patricia Holahan, Director

37 Division of Decommissioning, Uranium Recovery, and Waste Programs

38 Office of Nuclear Material Safety and Safeguards
vii

1 Table 1 Contents and Applicability of Key Decommissioning Guidance Documents

Volume and Status1 Title Licensees to Which the 

Guidance Applies 

NUREG-1757, Vol. 1, 

Rev. 2; September 

2006

Consolidated Decommissioning 

Guidance: Decommissioning 

Process for Materials Licensees 

Fuel cycle, fuel storage, and 

materials licensees2; limited 

applicability to reactor 

licensees (see text below)

NUREG-1757, Vol. 2, 

Rev. 2; November

2020 (this revision) 

Consolidated Decommissioning 

Guidance: Characterization, 

Survey, and Determination of 

Radiological Criteria 

All licensees that are subject 

to the LTR (fuel cycle, fuel 

storage, materials, and reactor 

licensees)

Draft NUREG-1757, 

Vol. 3, Rev. 1, 

February 2012

Consolidated Decommissioning 

Guidance: Financial Assurance, 

Recordkeeping, and Timeliness 

Volume 3 is intended to apply 

only to the decommissioning 

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; 

April 2018 

Standard Review Plan for 

Evaluating Nuclear Power Reactor 

License Termination Plans 

Power reactor licensees 

NUREG-1537, 

February 1996

Guidelines for Preparing and 

Reviewing Applications for the 

Licensing of Non-Power Reactors 

Nonpower reactor licensees 

Notes: 

1 Versions listed are current as of the date of publication of this document in 2020. The 

NRC’s 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, 60, 61, 63, 70, and 72 (for 

10 CFR Parts 60, 61, and 63, only 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.” 

2
viii

1 Table 2 Summary of Major Changes to Volume 2, Revision 2

Subject Affected Sections

Updated guidance on review of dose modeling used to 

demonstrate compliance with radiological criteria for license 

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”.

Chapters 2, 5, 

Appendix I 

Updated guidance on engineered barriers to reflect recent 

research; updated guidance on consideration of ALARA prior to 

use of engineered barriers to reduce dose for restricted release.

Section 3.5 and 

Appendix P

Updated guidance on review of ALARA analysis including 

experience and lessons learned gained from proposed restricted 

release scenarios; updated regulatory citations and guidance 

related to discounting and the monetary value of collective dose

averted.

Chapter 6 and

Appendix N

Updated guidance on surface water and groundwater 

characterization.

Appendix F

Added guidance on subsurface investigations, survey of 

excavations, survey of back-fill soils, data visualization tools, 

integration of dose modeling and radiological surveys, and 

implementation of Scenario B.

Section 3.6 and 

Appendix G

Updated guidance and information related to use of screening 

values and resuspension factors. Cites NUREG/CR-5512, Volume 

3 for screening values for additional radionuclides.

Appendix H

Streamlined guidance on consideration of buried radioactivity. 

Added guidance on intrusion events and exposure scenarios for 

large substructures.

Appendix J

Streamlined guidance on site-specific exposure scenarios. 

Updated guidance on regulatory standards.

Appendix M

Added guidance on use of composite sampling including 

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.

Appendix O

Added guidance on consideration of uncertainty in performance 

assessment analyses including issues associated with use of 

generic data sets, unrepresentative data, model integration, risk 

dilution, and lack of correlation of correlated parameters.

Appendix Q

2
ix

TABLE OF CONTENTS 

ABSTRACT ................................................................................................................... iii

FOREWORD................................................................................................................... v

LIST OF FIGURES ....................................................................................................... xiii 

LIST OF TABLES ......................................................................................................... xv

ACKNOWLEDGMENTS ............................................................................................. xvii

ABBREVIATIONS........................................................................................................ xix

GLOSSARY ................................................................................................................ xxiii

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 Decisionmaking 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 ....................2-6

2.5 Demonstrating Compliance Using Dose Assessment Methods versus Derived 

Concentration Guideline Levels and Final Status Surveys ......................................2-8

2.5.1 Dose Modeling Approach .............................................................................2-9

2.5.2 DCGL and FSS Approach ..........................................................................2-10

2.6 Merits of Screening versus Site-Specific Dose Assessment..................................2-10

2.7 Sum of Fractions ....................................................................................................2-11

2.8 Flexibility for Use of Institutional Controls and Engineered Barriers at 

Restricted Use Sites...............................................................................................2-13

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
x

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 ...........................................................................................4-8

4.3 Remedial Action Support Surveys..........................................................................4-10

4.3.1 Areas of Review .........................................................................................4-11

4.4 Final Status Survey Design....................................................................................4-12

4.4.1 Areas of Review .........................................................................................4-13

4.5 Final Status Survey Report ....................................................................................4-14

4.5.1 Areas of Review .........................................................................................4-14

4.6 Issues not Covered in MARSSIM...........................................................................4-21

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 SiteSpecific 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 SiteSpecific 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 .............................................5-14

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.............................................................................5-32

5.7.1 Regulatory Guidance Related to Decommissioning Groups 1 – 3 

(Unrestricted Release Using Screening Criteria)........................................5-32
xi

5.7.2 Regulatory Guidance Related to Decommissioning Groups 4 – 5 

(Unrestricted Release Using Site-Specific Information) .............................5-32

5.7.3 Regulatory Guidance Related to Decommissioning Group 6 

(Restricted Release)...................................................................................5-32

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 Vol. 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
xii

APPENDIX J ASSESSMENT STRATEGY FOR BURIED MATERIAL .............................J-1

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
xiii

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

LIST OF FIGURES IN APPENDICES

Figure D.1 MARLAP Road Map—Key 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-7

Figure G.1 Use of GIS and Geostatistical Tools .................................................................G-8

Figure G.2 Sampling Strategies for Excavations...............................................................G-12

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
xiv

Figure J.4 Conceptual Model of Human Disturbance into Buried Residual 

Radioactivity ....................................................................................................... J-6

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
xv

LIST OF TABLES

Table 1 Contents and Applicability of Key Decommissioning Guidance Documents ........vii

Table 2 Summary of Major Changes to Volume 2, Revision 2.........................................viii

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...................................................2-7

Table 2.2 Comparison of Dose Modeling to DCGL and FSS Approaches to 

Compliance ........................................................................................................2-9

Table 2.3 Attributes of Screening and Site-Specific Analysis...........................................2-12

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..................5-27

Table 5.3 Dose Modeling Review Criteria for Residual Radioactivity in Buildings ...........5-30

Table 7.1 Documents Superseded by this Report................................................................7-5

LIST OF TABLES IN APPENDICES

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 ώ2for the Example Data .............................................................G-19

Table G.2 Analysis of Variance for Example Data..............................................................G-20

Table G.3 Scenario B WRS and Quantile Tests for Class 2 Interior Drywall Survey Unit ....G-22

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
xvi

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

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 RESRAD’s 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-10

Table N.1 Possible Benefits and Costs Related to Decommissioning ............................... N-8

Table N.2 Acceptable Parameter Values for Use in ALARA Analyses ............................ 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-26

Table P.3 Examples of Engineered Barriers Used to Solve Erosion Problems ............... P-28

Table P.4 Summary of Existing Key Documents Related to Engineered Barriers........... P-35

Table Q.1 Example of Alternative Conceptual Models in a Performance Assessment....Q-13
xvii

1 ACKNOWLEDGMENTS

2 The preparers would like to thank the following individuals for assisting in the development and 

3 review of this revision to the guidance document. These individuals provided valuable 

4 contributions, insights, and recommendations. A special thanks to Lifeng Guo for his

5 contributions to Appendix F, “Surface Water and Groundwater Characterization”; Mark 

6 Fuhrmann and Hans Arlt for their contributions to various sections of the document including 

7 guidance on the use of engineered barriers in Section 3.5 and Appendix P; and Allen Gross for 

8 his contributions to Appendix M, “Process for Developing Exposure Scenarios using Site 

9 Specific Information”. The preparers would also like to acknowledge the contribution of Michael 

10 Kunowski for sharing his valuable NRC regional inspector experience, which greatly enhanced 

11 the quality of the guidance document. The preparers would like to acknowledge the very 

12 thorough review performed by Sheldon Clark in the NRC’s Office of General Counsel whose 

13 edits and comments substantially improved the defensibility and readability of the document. 

14 Comments received from the States of California, Pennsylvania, and Texas were also 

15 invaluable to obtain Agreement State perspectives and greatly improved the clarity and quality 

16 of the document. Finally, a special thanks goes out to Marcia Pringle and Sarah Achten for 

17 technical editing and formatting of the document without which this document could not be 

18 published.

19

20 Revision 2 Experts and Peer Reviewers

21 H. Arlt, M Fuhrmann, A. Gross, 

22 L. Guo, and M. Kunowski

23

24 Revision 1 Preparers (not including preparers of Revision 2)

25 U.S. Nuclear Regulatory Commission Staff

26 K.L. Banovac, J.T. Buckley,

27 R.L. Johnson, J.J. Kottan,

28 T.G. McLaughlin, and S. Schneider

29

30 Agreement States

31 California, Pennsylvania, and Texas 

xix

1 ABBREVIATIONS

2 ACAP Alternative Cover Assessment Program

3 ADAMS Agencywide Documents Access and Management System

4 AEA Atomic Energy Act of 1954, as amended

5 ALARA as low as is reasonably achievable

6 ALCD Alternative Landfill Cover Demonstration

7 Am americium

8 ANSI American National Standards Institute

9 ASME American Society of Mechanical Engineers

10 ASR alkali-silica reaction

11 ASTM American Society for Testing and Materials

12 Bq becquerel

13 Bq/kg becquerel per kilogram

14 Ca calcium

15 CERCLA Comprehensive Environmental Response, Compensation, and Liability 

16 Act

17 CFR Code of Federal Regulations

18 Ci curie

19 CNWRA Center for Nuclear Waste Regulatory Analyses

20 Co cobalt

21 CO2 carbon dioxide

22 cpm counts per minute

23 Cs cesium

24 C-S-H calcium silicate hydrate

25 CSM Conceptual site model

26 DandD Decontamination and Decommissioning software package

27 DCGL derived concentration guideline level

28 DCGLEMC derived concentration guideline level (elevated measurement

29 comparison)

30 DCGLW derived concentration guideline level (wide-area [over the entire survey 

31 unit])

32 DP decommissioning plan

33 dpm disintegrations per minute

34 DQA data quality assessment

35 DQOs data quality objectives

36 DQO process data quality objectives process

37 Eh redox potential

38 EMC elevated measurement comparison

39 EML DOE Environmental Measurements Laboratory (formerly the Health and 

40 Safety Laboratory)

41 EPA U.S. Environmental Protection Agency
xx

1 Fe iron

2 FEP feature, event, and process

3 FR Federal Register

4 FSS Final status survey

5 FSSP Final status survey plan

6 FSSR Final status survey report

7 GCL geosynthetic clay liner

8 GIS geographic information system

9 H hydrogen

10 H-3 tritium

11 HDPE high-density polyethylene

12 HSA historical site assessment

13 HTD hard-to-detect

14 IAEA International Atomic Energy Agency

15 ICRP International Commission on Radiological Protection

16 IL investigation level

17 ITRC Interstate Technology and Regulatory Council

18 ISO International Organization for Standardization

19 Kd distribution coefficient

20 kg kilogram

21 km kilometer

22 L liter

23 L/kg liter/kilgoram

24 LBGR lower bound of the gray region

25 LLW low-level waste

26 LTP License termination plan

27 LTR License termination rule

28 MARLAP Multi-Agency Radiological Laboratory Analytical Protocols Manual

29 (NUREG-1576)

30 MARSSIM Multi-Agency Radiological Survey and Site Investigation Manual 

31 (NUREG-1575)

32 MCL maximum contaminant level

33 MDC minimum detectable concentration

34 MDCscan scan minimum detectable concentration

35 MDCstatic static minimum detectable concentration

36 Mg magnesium

37 mg milligram

38 mil unit of corrosion (1 mil=0.001 inch)

39 MIL modified investigation level

40 mrem millirem

41 mSv millisievert
xxi

1 Na sodium

2 NAS National Academy of Sciences

3 NCRP National Council on Radiation Protection and Measurements

4 NEPA National Environmental Policy Act

5 Ni nickel

6 NIST National Institute of Standards and Technology

7 NMSS Office of Nuclear Material Safety and Safeguards (U.S. Nuclear 

8 Regulatory Commission)

9 NOAA National Oceanic and Atmospheric Administration

10 NORM naturally occurring radioactive material

11 Np neptunium

12 NRC U.S. Nuclear Regulatory Commission

13 NUREG NRC technical report designation

14 Pb lead

15 pCi picocurie

16 pCi/g picocuries per gram

17 PDF probability density function

18 pH hydrogen ion concentration (negative of the log of the hydrogen ion molar 

19 concentration)

20 Po polonium

21 PSR partial site release

22 Pu plutonium

23 PVC polyvinyl chloride

24 QA quality assurance

25 QAPP quality assurance project plan

26 QA/QC quality assurance and quality control

27 Ra radium

28 RCRA Resource Conservation and Recovery Act

29 REMP Radiological Environmental Monitoring Program

30 RESRAD RESidual RADdioactive materials computer code

31 RFR (also Rfo) resuspension factor

32 ROC radionuclide(s) of concern

33 RSS residual sum of squares

34 RSSI radiation survey and site investigation (process)

35 SADA Spatial Analysis and Decision Assistance (computer code)

36 SDMP Site Decommissioning Management Plan

37 Sr strontium

38 SRM staff requirements memorandum

39 SRP NMSS Decommissioning Standard Review Plan (NUREG-1727)

40 Sv sievert

41 TEDE total effective dose equivalent
xxii

1 Th thorium

2 U uranium

3 UF6 uranium hexafluoride

4 UMTRA Uranium Mill Tailings Remedial Action 

5 UMTRCA Uranium Mill Tailings Radiation Control Act

6 UO2 uranium dioxide

7 USDA U.S. Department of Agriculture 

8 U.S.C. U.S. Code

9 USGS U.S. Geological Survey

10 WRS Wilcoxon Rank Sum (test)

11 xLPR Extremely Low Probability of Rupture (project)

12 y year
xxiii

1 GLOSSARY

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

2

3 Affected Parties. Representatives of a broad cross section of individuals and institutions in the 

4 community or vicinity of a site that may be affected by the decommissioning of the site.

5

6 Acceptance Review. The evaluation the NRC staff performs upon receipt of a license 

7 amendment request to determine if the information provided in the document is sufficient to 

8 begin the technical review.

9

10 Activity. The rate of disintegration (transformation) or decay of radioactive material. The units 

11 of activity are the curie (Ci) and the becquerel (Bq) (see Title 10 of the Code of Federal 

12 Regulations (10 CFR) 20.1003, “Definitions”).

13

14 ALARA. The acronym for “as low as is reasonably achievable,” which means making every 

15 reasonable effort to maintain exposures to radiation as far below the dose limits as is practical, 

16 consistent with the purpose for which the licensed activity is undertaken, and taking into account 

17 the state of technology, the economics of improvements in relation to the state of technology, 

18 the economics of improvements in relation to the benefits to public health and safety, and other 

19 societal and socioeconomic considerations, and in relation to the use of nuclear energy and 

20 licensed materials in the public interest (see 10 CFR 20.1003).

21

22 Alternate Criteria. Dose criteria for residual radioactivity that are greater than certain dose 

23 criteria described in 10 CFR 20.1402, “Radiological Criteria for Restricted Use,” and 

24 10 CFR 20.1403, “Criteria for License Termination under Restricted Conditions,” as allowed in 

25 10 CFR 20.1404, “Alternate Criteria for License Termination.” Alternate criteria must be 

26 approved by the Commission.

27

28 Aquifer. A geologic formation, a group of formations, or part of a formation capable of yielding a 

29 significant amount of groundwater to wells or springs.

30

31 Background Radiation. Radiation from cosmic sources, naturally occurring radioactive material, 

32 including radon (except as a decay product of source or special nuclear material) and global 

33 fallout as it exists in the environment from the testing of nuclear explosive devices or from past 

34 nuclear accidents, such as Chernobyl, that contribute to background radiation and are not under 

35 the control of the licensee. Background radiation does not include radiation from source, 

36 byproduct, or special nuclear materials regulated by the NRC (see 10 CFR 20.1003).

37

38 Broad-Scope Licenses. A type of specific license authorizing receipt, acquisition, ownership, 

39 possession, use, and transfer of any chemical or physical form of the byproduct material 

40 specified in the license but not exceeding quantities specified in the license. Relevant 

41 requirements are found in 10 CFR Part 33, “Specific Domestic Licenses of Broad Scope for 

42 Byproduct Material.” Examples of broad scope licensees are large universities and large 

43 research and development facilities.
xxiv

1 Byproduct Material. 

2 1. Any radioactive material (except special nuclear material) yielded in, or made radioactive 

3 by, exposure to the radiation incident to the process of producing or using special 

4 nuclear material.

5 2. The tailings or wastes produced by the extraction or concentration of uranium or thorium 

6 from ore processed primarily for its source material content, including discrete surface 

7 wastes resulting from uranium solution extraction processes. Underground ore bodies

8 depleted by these solution extraction operations do not constitute “byproduct material” 

9 within this definition.

10 3. (i) Any discrete source of radium (Ra)-226 that is produced, extracted, or converted after 

11 extraction, before, on, or after August 8, 2005, for use for a commercial, medical, or 

12 research activity; or

13 (ii) Any material that—

14 a. Has been made radioactive by use of a particle accelerator, and

15 b. Is produced, extracted, or converted after extraction, before, on, or after 

16 August 8, 2005, for use for a commercial, medical, or research activity.

17 4. Any discrete source of naturally occurring radioactive material, other than source 

18 material, that—

19 (i) The Commission, in consultation with the Administrator of the Environmental 

20 Protection Agency, the Secretary of Energy, the Secretary of Homeland Security, and 

21 the head of any other appropriate Federal agency, determines would pose a threat 

22 similar to the threat posed by a discrete source of Ra-226 to public health and safety 

23 or the common defense and security and

24 (ii) Before, on, or after August 8, 2005, is extracted or converted after extraction for use 

25 in a commercial, medical, or research activity (see 10 CFR 20.1003).

26

27 Categorical Exclusion. A category of regulatory actions that do not individually or cumulatively 

28 have a significant effect on the human environment and that the Commission has found to have 

29 no such effect, in accordance with procedures set out in 10 CFR 51.22, “Criterion for 

30 Categorical Exclusion; Identification of Licensing and Regulatory Actions Eligible for Categorical 

31 Exclusion or Otherwise Not Requiring Environmental Review,” and for which, therefore, neither 

32 an environmental assessment nor an environmental impact statement is required (see 

33 10 CFR 51.14(a)). 

34

35 Characterization Survey. A type of survey that includes facility or site sampling, monitoring, and 

36 analysis activities to determine the extent and nature of residual radioactivity. Characterization 

37 surveys provide the basis for acquiring necessary technical information to develop, analyze, and 

38 select appropriate cleanup techniques.

39

40 Cleanup. See Decontamination.

41

42 Closeout Inspection. An inspection performed by the NRC, or its contractor, to determine if a 

43 licensee has adequately decommissioned its facility. Typically, a closeout inspection is 

44 performed after the licensee has demonstrated that its facility is suitable for release in 

45 accordance with NRC requirements.

46

47 Confirmatory Survey. A survey conducted by the NRC, or its contractor, to verify the results of 

48 the licensee’s final status survey. Typically, confirmatory surveys consist of measurements at a 

49 fraction of the locations previously surveyed by the licensee, to determine whether the 

50 licensee’s results are valid and reproducible.
xxv

1 Critical Group. The group of individuals reasonably expected to receive the greatest exposure 

2 to residual radioactivity for any applicable set of circumstances (see 10 CFR 20.1003).

3

4 DandD Code. The Decontamination and Decommissioning (DandD) software package, 

5 developed by the NRC, that addresses compliance with the dose criteria of 10 CFR Part 20, 

6 “Standards for Protection against Radiation,” Subpart E, “Radiological Criteria for License 

7 Termination.” Specifically, DandD embodies the NRC’s guidance on screening dose 

8 assessments to allow licensees to perform simple estimates of the annual dose from residual 

9 radioactivity in soils and on building surfaces. The current version of the code is 2.4, as of 

10 publication of this NUREG.

11

12 Data Quality Objectives (DQOs). Qualitative and quantitative statements derived from the Data 

13 Quality Objectives process that clarify study technical and quality objectives, define the 

14 appropriate type of data, and specify tolerable levels of potential decision errors that will be 

15 used as the basis for establishing the quality and quantity of data needed to support decisions.

16

17 Data Quality Objectives Process (DQO process): A series of logical steps that guides 

18 managers or staff to plan for the resource-effective acquisition of environmental data. See also 

19 data quality objectives.

20

21 Decommission. To remove a facility or site safely from service and reduce residual radioactivity 

22 to a level that permits (1) release of the property for unrestricted use and termination of the 

23 license or (2) release of the property under restricted conditions and termination of the license 

24 (see 10 CFR 20.1003). 

25

26 Decommissioning Groups. For the purposes of this guidance document, the categories of 

27 decommissioning activities that depend on the type of operation and the residual radioactivity.

28

29 Decommissioning Plan (DP). A detailed description of the activities that the licensee intends to 

30 use to assess the radiological status of its facility, to remove radioactivity attributable to licensed 

31 operations at its facility to levels that permit release of the site in accordance with the NRC’s

32 regulations and termination of the license, and to demonstrate that the facility meets the NRC’s 

33 requirements for release. A DP typically consists of several interrelated components, including 

34 (1) site characterization information, (2) a remediation plan that has several components, 

35 including a description of remediation tasks, a health and safety plan, and a quality assurance 

36 (QA) plan, (3) site-specific cost estimates for the decommissioning, and (4) a final status survey

37 plan (see 10 CFR 30.36(g)(4)).

38

39 Decontamination. The removal of undesired residual radioactivity from facilities, soils, or 

40 equipment before the release of a site or facility and termination of a license; also known as 

41 remediation, remedial action, and cleanup.

42

43 Derived Concentration Guideline Levels (DCGLs). Radionuclide-specific concentration limits 

44 used by the licensee during decommissioning to achieve the regulatory dose standard that 

45 permits the release of the property and termination of the license. The DCGL applicable to the 

46 average concentration over a survey unit is called the DCGLW. The DCGL applicable to limited 

47 areas of elevated concentrations within a survey unit is called the DCGLEMC (elevated 

48 measurement comparison).
xxvi

1 Distribution Coefficient or Kd. Ratio of the concentration of an element or chemical associated 

2 with the soil to the concentration in the surrounding aqueous solution when the system is at 

3 equilibrium. The units are typically expressed in L/kg. 

4

5 Dose (or Radiation Dose). A generic term that means absorbed dose, dose equivalent, 

6 effective dose equivalent, committed dose equivalent, committed effective dose equivalent, or 

7 total effective dose equivalent, as defined in 10 CFR 20.1003. In this report, dose generally 

8 refers to total effective dose equivalent (TEDE). 

9

10 Durable Institutional Controls. Durable institutional controls are reliable and sustainable for the 

11 time period needed. An institutional control that involves government ownership or control of 

12 the site would be considered a durable institutional control.

13

14 Effluent. Material discharged into the environment from licensed operations.

15

16 Environmental Assessment. A concise public document for which the Commission is 

17 responsible that serves to (1) briefly provide sufficient evidence and analysis for determining 

18 whether to prepare an environmental impact statement or a finding of no significant impact, 

19 (2) aid the Commission’s compliance with the National Environmental Policy Act (NEPA) when 

20 no environmental impact statement is necessary, and (3) facilitate preparation of an 

21 environmental impact statement when one is necessary (see 10 CFR 51.14(a)).

22

23 Environmental Impact Statement. A detailed written document that ensures the NEPA policies 

24 and goals are considered in the actions of the Federal government. It discusses significant 

25 impacts and reasonable alternatives to the proposed action.

26

27 Environmental Monitoring. The process of sampling and analyzing environmental media in and 

28 around a facility (1) to confirm compliance with performance objectives and (2) to detect 

29 radioactive material entering the environment to facilitate timely remedial action.

30

31 Environmental Report. A document submitted to the NRC by an applicant for a license 

32 amendment request (see 10 CFR 51.14(a)). The NRC staff uses the environmental report to 

33 prepare environmental assessments and environmental impact statements. The requirements 

34 for environmental reports are specified in 10 CFR 51.45–69.

35

36 Exposure Pathway. The route by which radioactivity travels through the environment to 

37 eventually cause radiation exposure to a person or group.

38

39 Exposure Scenario. A description of the potential future land uses, human activities, and 

40 transport of radioactivity in the natural system as it influences a future human receptor’s 

41 interaction with (and therefore exposure to) residual radioactivity. In particular, the exposure 

42 scenario describes where humans may be exposed to residual radioactivity in the environment, 

43 what exposure group habits determine exposure, and how residual radioactivity moves through 

44 the environment.

45

46 External Dose. That portion of the dose equivalent received from radiation sources outside the 

47 body (see 10 CFR 20.1003).

48

49 Final Status Survey (FSS). Measurements and sampling to describe the radiological conditions 

50 of a site or facility, following completion of decontamination activities (if any) and in preparation 

51 for release of the site or facility. 
xxvii

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

2

3 Final Status Survey Report (FSSR). The results of the final status survey conducted by a 

4 licensee to demonstrate the radiological status of its facility. The FSSR is submitted to the NRC 

5 for review and approval.

6

7 Financial Assurance. A guarantee or other financial arrangement provided by a licensee that 

8 funds for decommissioning will be available when needed. This is in addition to the licensee's 

9 regulatory obligation to decommission its facilities.

10

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

12 decommissioning.

13

14 Floodplain. The lowland and relatively flat areas adjoining inland and coastal waters, including 

15 flood-prone areas of offshore islands. Areas subject to a one-percent or greater chance of 

16 flooding in any given year are included (see 10 CFR 72.3, “Definitions”).

17

18 Footprint. The portion of a site undergoing decommissioning, which comprises all the areas of 

19 soil containing residual radioactivity, where intentional mixing is proposed to meet the release

20 criteria.

21

22 General Licenses. Licenses that are effective without filing applications with the NRC or the 

23 issuance of licensing documents to particular persons. The requirements for general licenses 

24 are found in 10 CFR Part 30, “Rules of General Applicability to Domestic Licensing of Byproduct 

25 Material;” 10 CFR Part 31, “General Domestic Licenses for Byproduct Material;” 10 CFR Part 

26 40, “Domestic Licensing of Source Material;” and 10 CFR Part 70, “Domestic Licensing of 

27 Special Nuclear Material.” Examples of items for which general licenses are issued include 

28 tritium exit signs and anti-static devices.

29

30 Groundwater. Water contained in pores or fractures in either the unsaturated or saturated 

31 zones below ground level.

32

33 Historical Site Assessment (HSA). The identification of potential, likely, or known sources of 

34 radioactive material and radioactive contamination based on existing or derived information for 

35 the purpose of classifying a facility or site, or parts thereof, as impacted or nonimpacted (see

36 10 CFR 50.2, “Definitions”).

37

38 Hydraulic Conductivity. The volume of water that will move through a medium in a unit of time 

39 under a unit hydraulic gradient through a unit area measured perpendicular to the direction of 

40 flow.

41

42 Hydrology. Study of the properties, distribution, and circulation of water on the surface of the 

43 land, in the soil and underlying rocks, and in the atmosphere.

44

45 Impact. The positive or negative effect of an action (past, present, or future) on the natural 

46 environment (land use, air quality, water resources, geological resources, ecological resources, 

47 aesthetic and scenic resources) and the human environment (infrastructure, economics, social, 

48 and cultural).

49

50 Impacted Areas. The areas with some reasonable potential for residual radioactivity in excess 

51 of natural background or fallout levels (see 10 CFR 50.2).
xxviii

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

2 for 24 months or more.

3

4 Infiltration. The process of water entering the soil at the ground surface. Infiltration becomes 

5 percolation when water has moved below the depth at which it can be removed (to return to the 

6 atmosphere) by evaporation or transpiration.

7

8 Insignificant Radionuclides and Pathways. Radionuclides and pathways that can be excluded 

9 from further detailed consideration, because they cumulatively contribute no more than 10 

10 percent of the dose standard (e.g., for unrestricted release, no more than 0.025 mSv/y or 2.5 

11 mrem/y). The dose contributions of the insignificant radionuclides and pathways should still be 

12 considered in demonstrating compliance with release criteria. See Chapter 3, Section 3.3 for 

13 additional information on insignificant radionuclides and pathways.

14

15 In Situ Recovery (ISR). In situ recovery (ISR) is one of the two primary extraction methods that 

16 are currently used to obtain uranium from underground. ISR facilities recover uranium from low17 grade ores where other mining and milling methods may be too expensive or environmentally 

18 disruptive. The ISR process is as follows:

19

20 (1) A solution called lixiviant (typically containing water mixed with oxygen and/or hydrogen 

21 peroxide, as well as sodium carbonate or carbon dioxide) is injected through a series of 

22 wells into the ore body to dissolve the uranium.

23 (2) The lixiviant is then collected in a series of recovery wells, through which it is pumped to 

24 a processing plant, where the uranium is extracted from the solution through an 

25 ion-exchange process.

26 (3) The uranium extract is then further purified, concentrated, and dried to produce a 

27 material, which is called "yellowcake" because of its yellowish color.

28 (4) Finally, the yellowcake is packed in 55-gallon drums to be transported to a uranium 

29 conversion facility, where it is processed through the stages of the nuclear fuel cycle to 

30 produce fuel for use in nuclear power reactors.

31 Institutional Controls. Measures to control access to a site and minimize disturbances to 

32 engineered measures established by the licensee to control the residual radioactivity. 

33 Institutional controls include administrative mechanisms (e.g., land use restrictions) and may 

34 include, but are not limited to, physical controls (e.g., signs, markers, landscaping, and fences).

35

36 Karst. A type of topography that is formed over limestone, dolomite, or gypsum by dissolution, 

37 characterized by sinkholes, caves, and underground drainage.

38

39 Leak Test. A test for leakage of radioactivity from sealed radioactive sources. These tests are 

40 made when the sealed source is received and on a regular schedule thereafter. The frequency 

41 is usually specified in the sealed source and device registration certificate or license.

42

43 License Termination Plan (LTP). A detailed description of the activities a reactor licensee 

44 intends to use to assess the radiological status of its facility, to remove radioactivity attributable 

45 to licensed operations at its facility to levels that permit release of the site in accordance with the 

46 NRC’s regulations and termination of the license, and to demonstrate that the facility meets the 

47 NRC’s requirements for release. An LTP consists of several interrelated components, including 
xxix

1 (1) a site characterization, (2) identification of remaining dismantlement activities, (3) plans for 

2 site remediation, (4) detailed plans for the final radiation survey, (5) a description of the end use 

3 of the facility, if restricted, (6) an updated site-specific estimate of remaining decommissioning 

4 costs, and (7) a supplement to the environmental report, pursuant to 10 CFR 51.33, “Draft 

5 Finding of No Significant Impact; Distribution,” describing any new information or significant 

6 environmental change associated with the licensee’s proposed termination activities (see 

7 10 CFR 50.82, “Termination of License”).

8

9 License Termination Rule (LTR). The License Termination Rule refers to the final rule on 

10 “Radiological Criteria for License Termination,” published by the NRC as Subpart E to 

11 10 CFR Part 20 on July 21, 1997 (62 FR 39058).

12

Licensee. A person who possesses a license1 13 , or a person who possesses licensable material, 

14 whom the NRC could require to obtain a license.

15

16 MARSSIM. NUREG-1575, “Multi-Agency Radiation Site Survey and Investigation Manual,” is a 

17 multiagency consensus manual that provides information on planning, conducting, evaluating, 

18 and documenting building surface and surface soil final status radiological surveys for 

19 demonstrating compliance with dose- or risk-based regulations or standards.

20

21 Model. A simplified representation of an object or natural phenomenon. The model can be in 

22 many possible forms, such as a set of equations or a physical, miniature version of an object or 

23 system constructed to allow estimates of the behavior of the actual object or phenomenon when 

24 the values of certain variables are changed. Important environmental models include those 

25 estimating the transport, dispersion, and fate of chemicals in the environment.

26

27 Monitoring. Monitoring (radiation monitoring, radiation protection monitoring) is the 

28 measurement of radiation levels, concentrations, or quantities of radioactive material and the 

29 use of the results of these measurements to evaluate potential exposures and doses (see 

30 10 CFR 20.1003).

31

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

33

34 National Environmental Policy Act (NEPA). The National Environmental Policy Act of 1969, 

35 which requires Federal agencies, as part of their decision-making process, to consider the 

36 environmental impacts of actions under their jurisdiction. Both the Council on Environmental 

37 Quality and the NRC have regulations to implement NEPA requirements. The Council’s 

38 regulations are contained in 40 CFR Parts 1500 to 1508, and NRC requirements are provided in 

39 10 CFR Part 51, “Environmental Protection Regulations for Domestic Licensing and Related 

40 Regulatory Functions.”

41

42 Naturally Occurring Radioactive Material (NORM). The natural radioactivity in rocks, soils, air, 

43 and water. NORM generally refers to materials in which the radionuclide concentrations have 

44 not been enhanced by or are as a result of human practices. NORM does not include uranium 

45 or thorium in source material.

46

47 Non-impacted Areas. The areas with no reasonable potential for residual radioactivity in excess 

48 of natural background or fallout levels (see 10 CFR 50.2).

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).
xxx

1 Pathway. See Exposure Pathway.

2

3 Performance-Based Approach. Regulatory decision-making that relies upon measurable or 

4 calculable outcomes (i.e., performance results) to be met but provides more flexibility to the 

5 licensee as to the means of meeting those outcomes.

6

7 Permeability. The ability of a material to transmit fluid through its pores when subjected to a 

8 difference in head (pressure gradient). Permeability depends on the substance transmitted 

9 (e.g., oil, air, water) and on the size and shape of the pores, joints, and fractures in the medium 

10 and the manner in which they are interconnected.

11

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

13 decimal fraction or as a percentage.

14

15 Potentiometric Surface. The two-dimensional surface that describes the elevation of the water

16 table. In an unconfined aquifer, the potentiometric surface is at the top of the water level. In a 

17 confined aquifer, the potentiometric surface is above the top of the water level, because the 

18 water is under confining pressure.

19

20 Principal Activities. Activities authorized by the license that are essential to achieving the 

21 purpose(s) for which the license was issued or amended. Storage during which no licensed 

22 material is accessed for use or disposal and activities incidental to decontamination or 

23 decommissioning are not principal activities (see 10 CFR 30.4, “Definitions”).

24

25 Probabilistic. Refers to computer codes or analyses that use a random sampling method to 

26 select parameter values from a distribution. Results of the calculations are also in the form of a 

27 distribution of values. The results of the calculation do not typically include the probability of the 

28 exposure scenario occurring.

29

30 Quality Assurance Project Plan (QAPP). A planning document that provides comprehensive 

31 details regarding the necessary quality assurance and quality control and other technical 

32 activities that must be implemented to ensure that the results of the work performed will satisfy 

33 stated performance criteria. See Appendix D of this volume for additional information on 

34 QAPPs.

35

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

37 economic standpoint.

38

39 Reasonably Foreseeable Land Use. Land use exposure scenarios that are likely within 

40 100 years, considering advice from land use planners and stakeholders on land use plans and 

41 trends.

42

43 Receptor Scenario. See Exposure Scenario.

44

45 rem. The special unit of any of the quantities expressed as dose equivalent. The dose 

46 equivalent in rems is equal to the absorbed dose in rads multiplied by the quality factor 

47 (1 rem = 0.01 sievert) (see 10 CFR 20.1004, “Units of Radiation Dose”).

48

49 Remedial Action. See Decontamination.

50

51 Remediation. See Decontamination.
xxxi

1 Residual Radioactivity. Radioactivity in structures, materials, soils, groundwater, and other 

2 media at a site resulting from activities under the licensee’s control. This includes radioactivity 

3 from all licensed and unlicensed sources used by the licensee but excludes background 

4 radiation. It also includes radioactive materials remaining at the site as a result of routine or 

5 accidental releases of radioactive material at the site and previous burials at the site, even if 

6 those burials were made in accordance with the provisions of 10 CFR Part 20 (see 

7 10 CFR 20.1003).

8

9 RESRAD Code. A computer code developed by the U.S. Department of Energy and designed 

10 to estimate radiation doses and risks from RESidual RADioactive materials in soils.

11

12 RESRAD-BUILD Code. A computer code developed by the U.S. Department of Energy and 

13 designed to estimate radiation doses and risks from RESidual RADioactive materials in 

14 BUILDings.

15

16 RESRAD-OFFSITE Code. A computer code developed by the U.S Department of Energy that 

17 extends the capabilities of the RESRAD (onsite) computer code to estimate the radiological 

18 consequences to a receptor located either onsite or outside the area of primary contamination.

19

20 Restricted Area. Any area to which access is limited by a licensee for the purpose of protecting 

21 individuals from undue risks from exposure to radiation and radioactive materials (see 

22 10 CFR 20.1003).

23

24 Risk. Defined by the “risk triplet” of a scenario (a combination of events or conditions that could 

25 occur) or set of scenarios, the probability that the scenario could occur, and the consequence 

26 (e.g., dose to an individual) if the scenario were to occur.

27

28 Risk-Based Approach. Regulatory decision-making that is based solely on the numerical results 

29 of a risk assessment. (Note that the Commission does not endorse a risk-based regulatory 

30 approach.)

31

32 Risk-Informed Approach. Regulatory decision-making that represents a philosophy whereby 

33 risk insights are considered together with other factors to establish requirements that better 

34 focus licensee and regulatory attention on design and operational issues commensurate with

35 their importance to public health and safety.

36

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

38

39 Robust Engineered Barrier. A man-made structure that is designed to mitigate the effect of 

40 natural processes or human uses that may initiate or accelerate the release of residual 

41 radioactivity through environmental pathways. The structure is designed so that the radiological 

42 criteria for license termination (10 CFR Part 20, Subpart E) are met. Robust engineered 

43 barriers are designed to be more substantial, reliable, and sustainable for the time period 

44 needed without reliance on active ongoing maintenance.

45

46 Safety Evaluation Report. The NRC staff’s evaluation of the licensee’s proposed action to 

47 determine if that action can be accomplished safely.

48

49 Saturated Zone. That part of the earth’s crust beneath the regional water table in which all 

50 voids, large and small, are filled with water under pressure greater than atmospheric.
xxxii

1 Scoping Survey. A type of survey that is conducted to identify (1) radionuclide contaminants, 

2 (2) relative radionuclide ratios, and (3) general levels and extent of residual radioactivity. 

3

4 Scenario. As specified in draft low-level waste guidance (NUREG-2175, “Guidance for 

5 Conducting Technical Analyses for 10 CFR Part 61, Draft Report for Comment,” issued 

6 March 2015), the term “scenario” refers to the expected (“central scenario”) or potential 

7 (“alternative scenario”) future dynamic evolution of the disposal site, which might include

8 consideration of disruptive events (e.g., gully erosion, climate change). Typically, detailed 

9 consideration of the future evolution of a decommissioning site is unnecessary. However, 

10 central and alternative scenarios may need to be considered to demonstrate compliance with 

11 the radiological criteria for license termination for some complex decommissioning sites, or sites 

12 with relatively long-lived radionuclides (half-lives comparable to or longer than the compliance 

13 period). In general, the term “scenario” pertains to the “exposure scenario” (or receptor 

14 scenario), when used in NUREG-1757, Volume 2, as defined above.

15

16 Screening Analysis/Approach/Methodology/Process. The use of (1) predetermined building 

17 surface concentration and surface soil concentration values, or (2) a predetermined 

18 methodology (e.g., use of the DandD code) that meets the radiological decommissioning criteria 

19 without further analysis, to simplify decommissioning in cases where low levels of residual 

20 radioactivity are achievable.

21

22 Sealed Source. Any special nuclear material or byproduct material encased in a capsule 

23 designed to prevent leakage or escape of the material.

24

25 sievert (Sv). The SI unit of any of the quantities expressed as dose equivalent. The dose 

26 equivalent in sieverts is equal to the absorbed dose in grays multiplied by the quality factor 

27 (1 sievert = 100 rem) (see 10 CFR 20.1004).

28

29 Site. The area of land, along with structures and other facilities, as described in the original 

30 NRC license application, plus any property outside the originally licensed boundary added for 

31 the purpose of receiving, possessing, or using radioactive material at any time during the term 

32 of the license, as well as any property where radioactive material was used or possessed that 

33 has been released before license termination.

34

35 Site Characterization. Studies that enable the licensee to sufficiently describe the conditions of 

36 the site, separate building, or outdoor area to evaluate the acceptability of the DP.

37

38 Site Characterization Survey. See Characterization Survey.

39

40 Site Decommissioning Management Plan (SDMP). The program established by the NRC in 

41 March 1990 to help ensure the timely cleanup of sites with limited progress in completing the 

42 remediation of the site and the termination of the facility license. SDMP sites typically have 

43 buildings, former waste disposal areas, large volumes of tailings, groundwater contamination, 

44 and soil contaminated with low levels of uranium or thorium or other radionuclides.

45

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

47 default screening tools.

48

49 Smear. A radiation survey technique that is used to determine levels of removable surface 

50 contamination. A medium (typically filter paper) is rubbed over a surface (typically of an area 
xxxiii

1 100 square centimeters), followed by a quantification of the activity on the medium. Also known 

2 as a swipe.

3

4 Source Material. Uranium or thorium, or any combination of uranium and thorium, in any 

5 physical or chemical form, or ores that contain by weight one-twentieth of one percent 

6 (0.05 percent) or more of uranium, thorium, or any combination of uranium and thorium. Source 

7 material does not include special nuclear material (see 10 CFR 20.1003).

8

9 Source Term. Release rates associated with residual radioactivity at a site or facility. The 

10 source term is related to the inventory, distribution of contamination, and controlling release 

11 mechanisms (e.g., solubility-controlled release, diffusion-limited release, or desorption). Note

12 that the working definition of source term in this volume is slightly different than the definition of 

13 source term found in the NRC glossary. The definition found in the online NRC glossary, at 

14 https://www.nrc.gov/reading-rm/basic-ref/glossary/source-term.html, is specific to accidents

15 involving radioactive materials: “types and amounts of radioactive or hazardous material 

16 released to the environment following an accident”.

17

18 Special Nuclear Material. (1) Plutonium, uranium (U)-233, uranium enriched in the isotope 233 

19 or in the isotope 235, and any other material that the Commission, pursuant to the provisions of 

20 Section 51 of the Atomic Energy Act of 1954, as amended (AEA), determines to be special 

21 nuclear material, but does not include source material; or (2) any material artificially enriched by 

22 any of the foregoing but does not include source material (see 10 CFR 20.1003).

23

24 Specific Licenses. Licenses issued to a named person who has filed an application for the 

25 license under the provisions of 10 CFR Parts 30, 32–36, 39, 40, 61, 70, and 72. Examples of 

26 specific licenses are industrial radiography, medical use, irradiators, and well logging.

27

28 Surface Soil. The top layer of soil on a site that supports certain exposure pathways such as 

29 direct exposure, soil ingestion, and resuspension of particles for inhalation. Surface soil has 

30 also been associated with the thickness of soil that can be measured using direct measurement 

31 or scanning techniques. Typically, this layer is often represented as the top 15 centimeters 

32 (6 inches) of soil but will vary depending on the radionuclide, surface characteristics, 

33 measurement technique, and dose modeling assumptions.

34

35 Survey. An evaluation of the radiological conditions and potential hazards incident to the 

36 production, use, transfer, release, disposal, or presence of radioactive material or other sources 

37 of radiation. When appropriate, such an evaluation includes a physical survey of the location of 

38 radioactive material and measurements or calculations of levels of radiation, or concentrations 

39 or quantities of radioactive material present (see 10 CFR 20.1003).

40

41 Survey Unit. A geographical area consisting of structures or land areas of specified size and 

42 shape at a site for which a separate decision will be made as to whether the unit attains the 

43 site-specific reference-based cleanup standard for the designated pollution parameter. Survey 

44 units are generally formed by grouping contiguous site areas with similar use histories and

45 having the same contamination potential (classification). Survey units are established to 

46 facilitate the survey process and the statistical analysis of survey data.

47

48 Timeliness. Specific time periods stated in NRC regulations for decommissioning unused 

49 portions of operating nuclear materials facilities and for decommissioning the entire site upon 

50 termination of operations.
xxxiv

1 Total Effective Dose Equivalent (TEDE). The sum of the deep-dose equivalent (for external 

2 exposures) and the committed effective dose equivalent (for internal exposures) (see 

3 10 CFR 20.1003). 

4

5 Transmissivity. The rate of flow of water through a vertical strip of aquifer that is one unit wide 

6 and that extends the full saturated depth of the aquifer.

7

8 Unrestricted Area. An area, access to which is neither limited nor controlled by the licensee 

9 (see 10 CFR 20.1003).

10

11 Unsaturated Zone. The subsurface zone in which the geological material contains both water 

12 and air in pore spaces. The top of the unsaturated zone typically is at the land surface, 

13 otherwise known as the vadose zone.

14

15 Vadose Zone. See Unsaturated Zone.
1-1

1 PURPOSE, APPLICABILITY, AND ROADMAP

2 1.1 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.

3

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

5 Licensees decommissioning their facilities are required to demonstrate to the NRC that their 

6 proposed methods will ensure that the decommissioning can be conducted safely and that the 

7 facility, at the completion of decommissioning activities, will comply with the NRC’s 

8 requirements for license termination. This volume is also intended to be used in conjunction 

9 with NRC Inspection Manual Chapter 2602, “Decommissioning Oversight and Inspection 

10 Program for Fuel Cycle Facilities and Materials Licensees.” Licensees who are subject to 

11 Subpart E, “Radiological Criteria for License Termination,” of 10 CFR Part 20, “Standards for 

12 Protection against Radiation,” should use the policies and procedures discussed in this volume 

13 to develop and implement a decommissioning plan (DP) or license termination plan (LTP) (note 

14 that throughout this volume, when the term “DP” is used, it may generally be understood to refer 

15 to DPs or LTPs). Uranium recovery facilities may find this information useful, but they are not 

16 subject to Subpart E. Agreement State licensees should contact the appropriate regulatory 

17 authority. Depending on the State, Agreement State licensees may be able to use this 

18 guidance with the substitution of “Agreement State Authority” for “NRC.” 

19

20 Additionally, there are several military and former military sites around the country where the 

21 responsible Federal agency (e.g., U.S. Department of the Air Force, U.S. Department of the 

22 Army, or U.S. Department of the Navy) is implementing site reclamation activities to address the 

23 removal or remediation of radiological material under the Comprehensive Environmental 

24 Response, Compensation, and Liability Act (CERCLA), as amended (42 U.S.C. §§ 9601 et 

25 seq.), also known as Superfund. The NRC has received a number of inquiries into its 

26 regulatory jurisdiction at various sites where the responsible Federal agency uses non-Federal 

27 entities (i.e., private service providers) to conduct remediation activities involving regulated 

28 radioactive materials (byproduct, source, or special nuclear materials) (see Atomic Energy Act 

29 of 1954, as amended (AEA), 42 U.S.C. §§ 2011–2297h (2006)) on Federal property located in 

30 an Agreement State. To assist it in making future determinations, the NRC staff has developed 

31 a decision process that is consistent with the procedures in SA-500, “Jurisdiction 

32 Determinations,” dated March 10, 2011; the decision process is found in FSME-14-039, 

33 “Clarification on the Determination of Regulatory Jurisdiction of Non-Federal Entities Conducting 

34 Cleanup activities on Federal Property in Agreement States,” dated April 22, 2014.

35

36 This volume of NUREG-1757 is being issued to describe and make available to licensees and 

37 the public (1) guidance on technical aspects of compliance with specific parts of the 

38 Commission’s regulations, (2) methods acceptable to the NRC staff in implementing these 
1-2

1 regulations, and (3) some of the techniques and criteria the NRC staff uses in evaluating DPs 

2 and LTPs. Licensees should use this guidance to prepare DPs, LTPs, final status surveys 

3 (FSSs), and other technical decommissioning reports for NRC submittal. The NRC staff will use 

4 the guidance in reviewing these documents and related license amendment requests. The 

5 guidance in this volume is not a substitute for regulations, and compliance with the guidance is 

6 not required. Methods and solutions different from those described in this volume will be 

7 acceptable, if licensees provide a sufficient basis for the NRC staff to conclude that the 

8 licensees’ decommissioning actions are in compliance with the Commission’s regulations. 

9 However, the use of nonstandard methods may require more detailed justification for the NRC 

10 staff to determine acceptability. In addition, the increased complexity and detail of nonstandard 

11 demonstrations may result in increased NRC staff review time and, therefore, cost to the 

12 licensee.

13

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

14

15 1.2 Roadmap to this Volume

16 The NRC’s regulations require a licensee to submit a DP to support the decommissioning of its 

17 facility either (1) when it is required by a license condition or (2) when the NRC has not 

18 approved the procedures and activities necessary to carry out the decommissioning, and these 

19 procedures could increase the potential health and safety impact to the workers or the public. 

20 Chapters 4–6 provide acceptance criteria and evaluation criteria for use in reviewing DPs and 

21 other information submitted by licensees to demonstrate that the facility is suitable for release in 

22 accordance with NRC requirements.

23

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

25 Volume 1 described the categorization of facilities into Decommissioning Groups 1–7, based on 

26 the amount of residual radioactivity, the location of that material, and the complexity of the 

27 activities needed to decommission the site. Table 1.1 provides a summary description and 

28 examples of each decommissioning group (see Part I of Volume 1 of this NUREG series for 

29 more details). Table 1.2 shows the potential applicability of the guidance in this volume to each 

30 of these groups. Therefore, where possible, the guidance in this volume has been categorized 

31 by the decommissioning groups. For most topics in this volume, the guidance applies to more 

32 than one decommissioning group, as shown in Table 1.2. Licensees are encouraged to contact 

33 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.
1-3

1 Table 1.1 Description and Examples of Each Decommissioning Group

Group General Description Typical Examples

1

Licensed material was not released into the 

environment, did not cause the activation of 

adjacent materials, and did not contaminate 

work areas. 

Licensees who used only sealed 

sources, such as radiographers and 

irradiators

2

Licensed material was used in a way that 

resulted in residual radioactivity on building 

surfaces and/or soils. The licensee is able to 

demonstrate that the site meets the 

screening criteria for unrestricted use. 

Licensees who used only quantities of 

loose radioactive material that they 

routinely cleaned up (e.g., research 

and development facilities) 

3

Licensed material was used in a way that 

could meet the screening criteria, but the 

license needs to be amended to modify or 

add procedures to remediate buildings or 

sites. 

Licensees who may have occasionally 

released radioactivity within NRC limits 

(e.g., broad scope)

4

Licensed material was used in a way that 

resulted in residual radiological 

contamination of building surfaces or soils, or 

a combination of both (but not groundwater). 

The licensee demonstrates that the site 

meets unrestricted use levels derived from 

site-specific dose modeling. 

Licensees whose sites released loose 

or dissolved radioactive material within 

NRC limits and may have had some 

operational occurrences that resulted in 

releases above NRC limits (e.g., waste 

processors)

5

Licensed material was used in a way that 

resulted in residual radiological 

contamination of building surfaces, soils, or 

groundwater. The licensee demonstrates 

that the site meets unrestricted use levels 

derived from site-specific dose modeling. 

Licensees whose sites released, 

stored, or disposed of large amounts of 

loose or dissolved radioactive material 

onsite (e.g., fuel cycle facilities)

6

Licensed material was used in a way that 

resulted in residual radiological 

contamination of building surfaces, soils, or

groundwater. The licensee demonstrates 

that the site meets restricted use levels 

derived from site-specific dose modeling.

Licensees for whom cleaning their site

to the unrestricted release limit would 

cause a greater health and safety or 

environmental impact than could be 

justified (e.g., facilities where large 

inadvertent release(s) occurred) 

7

Licensed material was used in a way that 

resulted in residual radiological 

contamination of building surfaces, soils, or

groundwater. The licensee demonstrates 

that the site meets alternate restricted use 

levels derived from site-specific dose 

modeling.

Licensees for whom cleaning their site

to the restricted release limit would 

cause a greater health and safety or 

environmental impact than could be 

justified (e.g., facilities where large 

inadvertent release(s) occurred) 

2
Table 1.2 Applicability of Volume 2 to Decommissioning Groups

Groups Group 1 Group 2 Group 3 Group 4 Group 5 Group 6 

Dose Assessment 

Method

N/A

Screening criteria 

(Section 5, Appendix H)

Site-specific assessment

(Section 5, Appendices I, J, M, a

Dose Assessment 

for Partial Site 

Release

No

Yes, for licensees electing partial site releases

(Appendices K and L)

Site 

Characterization

No

Yes

(Section 4.2, Appendix E)

Yes

(Section 4.2, Appendices 

Remedial Action 

Support Surveys

No

Yes, if remediation is required

(Section 4.3, Appendix E)

Final Status Survey No

Yes

(Sections 4.4 and 4.5, 

Appendices A, B, D, and 

E)

Yes

(Sections 4.4 and 4.5, Appendices A, B

Complex Survey 

Situations (Not 

Addressed in 

MARSSIM)

No

Yes

(Section 4.6, Appendices C, G a

Surface and 

Groundwater 

Characterization

No

Yes

23

(Appendix F)

Yes

(Appendix F

ALARA Analysis No

Yes,

good housekeeping only 

(Section 6.3)

Yes

(Chapter 6, Appendix N)

2 Group 4 sites with residual radioactivity in surface water will benefit from Appendix A on surface water characterization.

1-4
1-5

1 Because of the variability in the amounts, forms, and types of radioactive material used by each 

2 decommissioning group, licensees may need to submit a broad range of information types and 

3 details to the NRC for approval of decommissioning activities. The types of information required 

4 could vary because of the radionuclides involved, whether remediation is required, or the 

5 complexity of the site. The amount of detail discussed in this volume is based on the needs of 

6 complicated sites. The NRC staff does not suggest that all licensees should provide the same 

7 level of detail. Rather, the amount of detail provided on a specific issue should be 

8 commensurate with the complexity of the issue for the facility. Thus, licensees and NRC

9 reviewers should generally determine the level of detail and appropriate methods, based on the 

10 complexity of the facility as related to a compliance demonstration. Licensees are encouraged 

11 to discuss with their NRC license reviewer the appropriate level of detail to be included in the 

12 DP, using the checklists of Appendix D of Volume 1 of this NUREG report. 

13

14 The technical aspects of sites, as related to decommissioning, are often called either “simple” or 

15 “complex.” The question becomes what defines the technical aspects as “simple” or “complex.” 

16 One needs to decide what aspect of the decommissioning one is trying to judge. For example, 

17 site characterization may be complex at a site, but the FSS, after remediation, may be simple 

18 and straightforward. 

19

20 Unfortunately, there is no precise definition or list of characteristics that can define the technical 

21 aspects as either simple or complex without caveats. That is because simple and complex are 

22 not distinct boxes but part of a continuum. For example, sites using screening criteria are 

23 relatively “simple,” technically, and sites proposing both partial release and restricted release 

24 with an engineered barrier design along with institutional controls that rely on active 

25 maintenance are relatively “complex,” technically. While there can be exceptions to the site 

26 complexity characterization illustrated in Figure 1.1, Decommissioning Groups 1–3 generally 

27 have mostly simple technical aspects, and Decommissioning Groups 5–7 generally have mostly 

28 “complex” technical aspects. Group 4 sites, which are sites without initial groundwater 

29 contamination, can be of either complexity.

30

31

32

33

34 Figure 1.1 Continuum of Site Complexity

35 Simple sites are generally easy to assess, because site characterization information, survey 

36 methods, and models with NRC-reviewed default parameter sets are readily available. These 

37 sites have residual radioactivity generally limited to building surfaces or surface soil at a site 

38 with simple geological and hydrological characteristics.

39

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

41

42 • existing groundwater or surface water contamination

43 • former burials of radioactive material or highly heterogeneous subsurface soil residual 

44 radioactivity 

Simple Complex

Group

s

1 - 3 4 5 6 7
1-6

1 • diversified and extensive surface or subsurface residual radioactivity that, because of the 

2 interactions between sources, may require data and modeling of these multiple sources 

3 at the site

4 • radionuclides that (1) are hard-to-detect (HTD), (2) lack suitable surrogate radionuclides, 

5 or (3) have very low effective derived concentration guideline levels (DCGLs)

6 • current offsite releases such that alternate offsite exposure scenario(s) may be required 

7 or the use of an onsite resident farmer exposure scenario may be inadequate (e.g., sites 

8 with multiple receptors)

9 • planned license termination under restricted conditions (10 CFR 20.1403, “Criteria for 

10 License Termination under Restricted Conditions”) 

11 • physical barriers or vaults 

12 • unusual physical or lithologic properties, such as a highly fractured formation, karst 

13 features, or sinkholes that may significantly affect assumptions of transport models or 

14 the overall conceptual model

15 These conditions are not rigid definitions, as other factors are also important. One such 

16 important factor would be the locations where radionuclides are present. For example, a site 

17 could be called simple because the predominant radionuclide is a short-lived energetic gamma 

18 that is in the surface soil; even if the hydrology at the site is complex, the site would be called 

19 simple, because the primary exposure pathway is external exposure, which is an uncomplicated 

20 pathway.

21

22 Technically complex sites may require more advanced remediation, survey planning, or 

23 performance assessment modeling and analysis approaches. Specifically, more advanced 

24 approaches may be required to select appropriate models or codes, collect characterization 

25 data, justify source term assumptions, ensure internal consistencies in the associated complex 

26 transport models, and design site- or source-specific survey plans. Because of the complex 

27 nature of these sites, the scope of NRC staff review will depend on site-specific conditions and 

28 on the degree of site complexity. Therefore, a generic NRC staff review of complex sites cannot 

29 be articulated in this volume. 

30

31 Licensees and the NRC staff should interact early for information and direction on the 

32 development of a complete DP. Once the decision has been made to decommission, the next 

33 step is to determine what information the licensee needs to demonstrate site conditions 

34 successfully. If the submittal of a DP is not necessary, the licensee should follow the guidance 

35 in Volume 1 of this NUREG report for the appropriate decommissioning group.

36

37 If the licensee is required to submit a DP, it should schedule a meeting with the NRC staff to 

38 discuss both the planned decommissioning and the approach that will be used to evaluate the 

39 information submitted to support the decommissioning. The NRC staff and the licensee should 

40 review the licensed operations, types and quantities of radioactive materials used at the facility, 

41 and any other activities (e.g., spills, leaks) that could affect decommissioning operations. The 

42 NRC staff should also discuss the decommissioning goal envisioned by the licensee 

43 (i.e., license termination under unrestricted versus restricted conditions) and the information 

44 required to be submitted for the appropriate decommissioning group (described in Chapters 10, 

45 11, 12, 13, or 14 of Volume 1 of this NUREG report). The NRC staff should then discuss the 

46 acceptance criteria for information to be included in the DP. Finally, the NRC staff should 

47 prepare a site-specific checklist for evaluating the DP. Appendix D of Volume 1 of this NUREG 

48 report provides a generic checklist that may be used to develop this site-specific checklist. 
1-7

1 Thus, before the licensee begins to develop its DP, both the NRC staff and the licensee should 

2 have a good understanding of the types of information that should be included in the DP, as well 

3 as the criteria that the NRC will use to evaluate the information submitted to support 

4 decommissioning. This should help minimize the need for requests for additional information.

5

6 1.3 Roadmap for Guidance on Restricted Use, Alternate Criteria, and Use of

7 Engineered Barriers

8 The focus of this volume is on guidance for demonstrating compliance with the dose criteria 

9 from 10 CFR Part 20, Subpart E. However, there are additional criteria in Subpart E related to 

10 license termination under restricted conditions and the use of alternate criteria for license 

11 termination. In addition, some licensees may wish to use engineered barriers as part of the 

12 compliance strategy. This section describes where guidance on these subjects may be found in 

13 this NUREG series (Volumes 1 and 2).

14

15 Table 1.3 provides cross-references to sections of Volume 1 and this volume for guidance on 

16 aspects of restricted use, use of alternate criteria, and use of engineered barriers.

17

18 1.4 Iterative Nature of the Compliance Demonstration Process: A Decision19 making Framework

20 The NRC staff developed an overall framework for dose assessment and decision-making 

21 where the licensee has decided to begin the decommissioning and license termination process 

22 at sites ranging from simple ones to the more complex or contaminated sites. This volume 

23 summarizes information for using the framework to step through the decommissioning and 

24 license termination process; NUREG-1549, “Decision Methods for Dose Assessment to Comply 

25 with Radiological Criteria for License Termination,” issued July 1998, contains a detailed 

26 description. Although the framework was developed for demonstrating compliance using the 

27 characterization and dose assessment approach (see Section 2.5), the concepts may be 

28 extended for use in the DCGL development and the FSS approach.

29

30 This framework is designed to assist the licensee, the NRC staff, and other stakeholders in 

31 making decommissioning decisions. By doing so, the process allows the licensee to do the 

32 following:

33

34 • coordinate its planning efforts with NRC staff input and conduct dose assessments and

35 site characterization activities that are directly related to regulatory decisions

36 • optimize cost decisions related to site characterization, remediation, and land use

37 restrictions

38 • integrate analyses for requirements that are as low as is reasonably achievable (ALARA)

39 • elicit other stakeholder input at crucial points
1-8

1 Table 1.3 Cross-References for Restricted Use, Alternate Criteria, and Use of Engineered

2 Barriers

Issue

Applicable Sections of This Report

Volume 1 Volume 2

Initial eligibility demonstration for restricted use 17.7.2 n/a

Institutional controls 17.7.3 n/a

Site maintenance and long-term monitoring 17.7.4 n/a

Obtaining public advice

17.7.5 and 

Appendix M n/a

Dose modeling for restricted use 17.7.6 5

ALARA analysis for restricted use 17.7.3.5 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 3.5 and Appendix P

3

4

5 The framework is designed to allow the licensee flexibility in the decision-making process for 

6 demonstrating compliance. As such, the framework provides one method that may be useful for 

7 licensees in developing the compliance strategy.

8

9 The steps and decision points of the decision framework support an assessment of the entire 

10 range of dose modeling options from which a licensee may choose, whether it involves using 

11 generic screening parameters, changing parameters, or modifying pathways or models. 

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

13 and decision points.
1-9

1

2 Figure 1.2 Decommissioning and License Termination Decision Framework (Modified 

3 from NUREG-1549)
1-10

1 1.4.1 Contents and General Concepts of the Iterative Approach in Using the Decision 

2 Framework

3 To facilitate the preparation and evaluation of the dose assessments, this framework describes 

4 an iterative approach to decision-making for license termination. An iterative approach is helpful 

5 because of the very wide range of levels of residual radioactivity, complexity of analysis, and 

6 potential remediation necessary at NRC-licensed sites. The iterative approach consists of using 

7 existing information for generic screening and using site-specific information as appropriate. 

8 This approach provides assurance that obtaining additional site-specific information is 

9 worthwhile, because it ensures that a more realistic dose assessment will generally result in an 

10 estimated dose no greater than that estimated using screening. These two phases of the 

11 compliance assessment are summarized in broad terms below, while NUREG-1549 contains

12 further details:

13

14 (1) Generic screening: In this iteration, licensees would demonstrate compliance with the 

15 dose criteria of the License Termination Rule (LTR) by using predefined models and 

16 generic screening parameters.

17 (2) Use of site-specific information as appropriate: If compliance cannot be demonstrated 

18 using generic screening, then licensees should proceed to the next iteration of analysis 

19 in which defensible site-specific values are obtained and applied.

20 The following general concepts apply to using the iterative approach with the decision 

21 framework shown in Figure 1.2:

22

23 • The approach provides a process for screening sites and for directing additional data 

24 collection efforts where necessary or where it is most helpful toward demonstrating 

25 compliance. 

26 • The framework is designed such that the level of complexity and rigor of analysis 

27 conducted for a given site should be commensurate with the level of risk that the site 

28 poses. 

29 • The licensee would not need to start the process with generic screening but may move 

30 directly to the use of site-specific information, as appropriate.

31 • For the process to work efficiently, the licensee is encouraged to involve the NRC staff 

32 from the very first step through to the end of the decision-making process.

33 The framework provides the licensee with a variety of options for performing dose assessments, 

34 from simple screening to more detailed site-specific analyses. Use of the framework would 

35 normally encompass Steps 1–7; however, the amount of work that goes into each of these 

36 steps should be based on the expected levels of residual radioactivity and the health risks they 

37 pose. Note that, in this framework, while all sites may start at the same level of very simple 

38 analyses (not a requirement for successful implementation), it is expected that only certain sites 

39 would progress to very complex dose assessment and options analyses. Some sites may not 

40 need to conduct any options analyses, as described in Step 8, and some sites may need to 

41 evaluate a limited set of relatively simple and inexpensive options. For example, the licensee at 

42 a site with a contained source of residual radioactivity that is obviously simple to remove would 

43 not spend time analyzing large suites of alternative data collection and remediation options. On 

44 the other hand, the licensee at a site with high levels of widely distributed residual radioactivity 
1-11

1 may use this process to analyze a variety of simple and complex options to define the best 

2 decontamination and decommissioning strategy. Therefore, this approach ensures that the 

3 licensee’s efforts and expenses will be commensurate with the level of risk posed by the site.

4

5 1.4.2 Steps of the Decision Framework 

6 NUREG-1549 provides three separate discussions to illustrate the iterative nature of 

7 assessments as site complexity increases. The following is both a summary of the steps of the 

8 decision framework and a set of examples to help users understand most of the features of 

9 dose modeling in the context of the decision framework. This discussion has been modified 

10 slightly from that in NUREG-1549 to make it applicable to a broader range of compliance 

11 demonstrations. A number of the examples refer to the use of the Decontamination and 

12 Decommissioning (DandD) and RESidual RADioactive (RESRAD) materials dose assessment 

13 codes. Chapter 5 and Appendices H and I of this volume discuss dose modeling codes, 

14 including specifics on these two dose assessment codes, while NUREG-1549 contains further

15 details. Figure 1.2 (modified from NUREG-1549) shows the following steps of the dose 

16 modeling framework:

17

18 (1) The first step in a compliance assessment involves gathering and evaluating existing 

19 data and information about the site for the historical site assessment (HSA), including 

20 the nature and extent of residual radioactivity at the site. Often, minimal information is 

21 all that is needed for an initial screening analysis (e.g., a simple representation of the 

22 source of residual radioactivity). Specifically, information is needed to support the

23 decision that the site is “simple” and is qualified for a screening analysis. However, 

24 licensees should use all readily available information about the site. This step also 

25 includes the definition of the performance objectives for compliance with 

26 decommissioning criteria. 

27 (2) This step involves defining the exposure scenarios and pathways that are important 

28 and relevant for the site dose assessment. This step also includes the preliminary 

29 determination of whether the licensee plans to adopt an unrestricted use or restricted 

30 use option provided for in the LTR. For all assessments using screening concentration 

31 tables or DandD, the NRC has already defined the generic exposure scenarios and 

32 pathways for screening. For a site-specific analysis, DandD and the RESRAD family 

33 of codes (e.g., RESRAD-ONSITE, RESRAD-OFFSITE, RESRAD-BUILD) may be 

34 used, in addition to other codes. The codes used should allow the licensee to select 

35 and deselect exposure pathways as appropriate for the site-specific conditions.

36 (3) Once exposure scenarios are defined and exposure pathways identified, the licensee 

37 develops a basic conceptual understanding of the system, often based on simplifying 

38 assumptions on the nature and behavior of the natural systems. System 

39 conceptualization includes conceptual and mathematical model development and an 

40 assessment of parameter uncertainty. Using DandD for generic screening (and as the 

41 basis for screening concentration tables), the NRC has predefined conceptual models 

42 for the exposure scenarios, along with default parameter distributions (based on 

43 NUREG/CR-5512, “Residual Radiation Contamination from Decommissioning,” 

44 Volumes 1 and 3, issued October 1992). The site-specific analysis can use DandD or 

45 the RESRAD family of codes, after verification that the site conceptual model is 

46 compatible with the conceptual model of the code selected.
1-12

1 (4) This step involves the dose assessment or consequence analysis, based on the 

2 defined exposure scenario(s), exposure pathways, models, and parameter 

3 distributions. This step may also involve the evaluation of FSS results. For generic 

4 screening, reviewers can accept lookup tables and use the generic models and default 

5 parameter probability density functions (PDFs) by running DandD with the appropriate 

6 site-specific source concentrations and configuration, while leaving all other 

7 information in the software unchanged. Site-specific assessments allow the licensee 

8 to use other codes and change pathways and parameter distributions based on

9 site-specific data and information. DandD, and the RESRAD family of codes provide 

10 various plots and reports of the dose distribution using Monte Carlo sampling of the 

11 input distributions.

12 (5) This is the first major decision point in the license termination decision process. It 

13 involves answering the question of whether the dose assessment results and/or FSS 

14 results demonstrate compliance with the dose criteria in 10 CFR Part 20, Subpart E. If 

15 the results demonstrate compliance, the licensee proceeds with Steps 6 and 7 to meet 

16 the ALARA requirements in Subpart E. If the results are ambiguous or clearly exceed 

17 the performance objective, then the licensee proceeds to Steps 8 and 9 for the next 

18 iteration of the decision-making process.

19 (6) In this step, the licensee addresses the ALARA criterion of 10 CFR Part 20, Subpart E, 

20 if it has not already done so. If the ALARA requirements are satisfied, then the 

21 licensee initiates the license termination. Note that the DandD and the RESRAD 

22 family of codes do not involve or automate these steps.

23 (7) This step includes the administrative and other actions necessary to terminate the

24 license and release the site, with Volume 1 of this NUREG containing more details on 

25 the specific actions to do so.

26 (8) Full application of the decision framework involves defining all possible options the 

27 licensee might address to defend a final set of actions needed to demonstrate 

28 compliance with license termination criteria. Options may include (i) acquiring more 

29 data and information about the site and source(s) of residual radioactivity to reduce 

30 uncertainty about the pathways, models, and parameters, and thus reduce the 

31 calculated dose, (ii) reducing actual contamination through remediation actions, 

32 (iii) reducing exposure to radionuclides through implementation of land use restrictions,

33 (iv) performing an FSS, or (v) some combination of these options.

34 (9) All the options identified in Step 8 are analyzed and compared to optimize the 

35 selection of a preferred set of options. This options analysis may consider the cost of 

36 implementation, the likelihood of success (and the expected costs associated with 

37 success or failure to achieve the desired results when the option is implemented), the 

38 timing considerations and constraints, and other quantitative or qualitative selection 

39 criteria.

40 (10) The activities in Steps 8 and 9 provide information for licensees to choose the 

41 preferred decommissioning option based on cost, the likelihood of success, timeliness, 

42 and other considerations. For example, results of sensitivity analyses performed with 

43 DandD, or the RESRAD family of codes, can be used by a licensee to identify one or

44 more parameters that may be modified, based on the acquisition of site-specific 

45 information and data. If new data can reduce the uncertainty associated with 
1-13

1 parameters found to be important to dose, then the licensee may be able to defend a 

2 new calculated dose that meets the license termination criteria. This step may include 

3 submitting a DP to the NRC, if necessary, to proceed with the preferred option. If the 

4 licensee believes that no viable options exist at this time, the licensee should confer 

5 with the NRC staff (see also Step 13).

6 (11) Under this step, the preferred option is implemented. The licensee obtains the 

7 information necessary to support revisions to the parameters identified in Steps 8 

8 and 9 or performs an FSS.

9 (12) Once the licensee obtains the data, it revises the affected parameters for the 

10 predefined models, as appropriate. Also, data may support the elimination of one or 

11 more of the exposure pathways in the predefined exposure scenarios. DandD and the 

12 RESRAD family of codes provide very simple and straightforward modifications of the 

13 pathways and parameters of interest.

14 Once the pathways and parameters are revised, the licensee would revisit Steps 4 

15 and 5 to determine the impact of the revisions on demonstrating compliance with the 

16 performance objectives. If met, the licensee proceeds to Steps 6 and 7. If the 

17 performance objective is still exceeded, the licensee returns to Steps 8 and 9 to 

18 analyze the remaining options.

19 In certain limited circumstances, terminating the license may not be feasible. The 

20 licensee should contact the NRC staff for case-specific guidance and for the regulatory 

21 approvals that may be necessary to maintain, rather than terminate, the license.

2-1

1 FLEXIBILITY IN DEMONSTRATING COMPLIANCE WITH 

2 10 CFR PART 20, SUBPART E

3 The NRC and its licensees share a common responsibility to protect public health and safety. 

4 Federal regulations and the NRC regulatory program are important elements in the protection of 

5 the public; however, NRC licensees are primarily responsible for safely using nuclear materials. 

6 The agency’s safety philosophy explains that “although NRC develops and enforces the 

7 standards governing the use of nuclear installations and materials, it is the licensee who bears 

8 the primary responsibility for conducting those activities safely.” This philosophy applies to the 

9 decommissioning of licensed facilities. Thus, the licensee has the primary responsibility for 

10 compliance with the license termination criteria. The responsibility of the NRC staff is to 

11 oversee the process and conclude that there is reasonable assurance that the criteria have 

12 been or will be met and then to terminate or amend licenses, as appropriate.

13

14 The dose criteria of 10 CFR Part 20, Subpart E, are performance criteria. In this volume, the 

15 NRC staff has taken a risk-informed, performance-based approach to demonstrations of 

16 compliance with the license termination criteria, using various methods available to licensees. 

17 Regardless of the specific method used, it is important that the licensee provide sufficient 

18 justification for its approach. This chapter discusses some of the aspects of flexibility in

19 methodologies for demonstrating compliance with the license termination criteria. One objective 

20 of this chapter is to emphasize the flexibility available in demonstrating compliance with the 

21 regulations.

22

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.

23

24 2.1 Risk-informed Approach to Compliance Demonstrations and Reviews

25 This section summarizes the risk-informed approach to regulatory decision-making. The NRC 

26 staff requirements memorandum (SRM) for SECY-98-144, “White Paper on Risk-Informed and 

27 Performance-Based Regulation,” issued March 1999, contains additional details. 

28

29 The NRC has increased the use of risk information and insights in its regulation of nuclear 

30 materials and nuclear waste management, including the decommissioning of nuclear facilities. 

31 Risk is defined by the “risk triplet” of (1) either a scenario or set of scenarios with a combination 

32 of events and/or conditions that could occur, (2) the probability that the scenario(s) could occur, 

33 and (3) the consequence (e.g., the dose to an individual) if the scenario(s) were to occur. The 

34 term risk insights, as used here, refers to the results and findings that come from risk 

35 assessments. The end results of such assessments may relate directly or indirectly to public 

36 health effects (e.g., the calculation of predicted doses from decommissioned sites). 
2-2

1 A risk-based approach to regulatory decision-making is based solely on the numerical results of 

2 a risk assessment. The Commission does not endorse a risk-based regulatory approach but 

3 supports a risk-informed approach to regulation. This approach represents a philosophy 

4 whereby risk insights are considered, together with other factors in the regulatory process, to 

5 better focus licensee and regulatory attention on design and operational issues commensurate 

6 with their importance to public health and safety. The staff does not typically use an explicit 

7 consideration of the numerical probability that a scenario would occur (i.e., number 2 of the risk 

8 triplet) to determine compliance with the LTR. This is a departure from a purely risk-based 

9 approach.

10

11 The typical deterministic approach to regulatory decision-making establishes requirements for 

12 engineering margin and for quality assurance (QA) in design, manufacture, and construction. In 

13 addition, it assumes that adverse conditions can exist and establishes a specific set of 

14 design-basis events (i.e., What can go wrong?). The deterministic approach involves implied, 

15 but unquantified, elements of probability in the selection of the specific design-basis events to 

16 be analyzed. Then, it requires that the design include safety systems capable of preventing 

17 and/or mitigating the consequences (i.e., What are the consequences?) of those design-basis 

18 events to protect public health and safety. Thus, a deterministic analysis explicitly addresses 

19 only two questions of the risk triplet.

20

21 The risk-informed approach has enhanced the deterministic approach by (1) allowing explicit 

22 consideration of a broader set of potential challenges to safety, (2) providing a logical means for 

23 prioritizing these challenges based on risk significance, operating experience, and/or 

24 engineering judgment, (3) facilitating consideration of a broader set of resources to defend 

25 against these challenges, (4) explicitly identifying and quantifying sources of uncertainty in the 

26 analysis (although such analyses do not necessarily reflect all important sources of uncertainty), 

27 and (5) leading to better decision-making by providing a means to test the sensitivity of the 

28 results to key assumptions.

29

30 Where appropriate, a risk-informed regulatory approach can also be used to reduce 

31 unnecessary conservatism in purely deterministic approaches or can be used to identify areas 

32 with insufficient conservatism in deterministic analyses and provide the bases for additional 

33 requirements or regulatory actions. Risk-informed approaches lie between the risk-based and 

34 purely deterministic approaches (SRM-SECY-98-144).

35

36 The NRC’s risk-informed regulatory approach to the decommissioning of nuclear facilities is 

37 intended to focus the attention and resources of both the licensee and the NRC on the more 

38 risk-significant aspects of the decommissioning process and on the elements of the facility and 

39 the site that will most affect risk to members of the public following decommissioning. While a

40 licensee must comply with all Commission regulations, a licensee whose site (or aspects of a 

41 site) have higher risk significance may need to provide a more rigorous demonstration to 

42 support compliance. Furthermore, the NRC staff generally will apply more scrutiny to reviews of 

43 such sites or situations with higher risk significance. This should result in a more effective and 

44 efficient regulatory process. The risk-informed regulatory approach to decommissioning is 

45 reflected in this volume, as shown by the following examples:

46

47 • The NRC has developed and is applying the concept of “decommissioning groups”

48 based on (1) the nature and the extent of the radioactive material present at a site and

49 (2) the complexity of the decommissioning process. The groups are generally related to

50 the potential risks associated with the site, in that the less complex sites with limited
2-3

1 distribution of radioactive material may pose lower risks (i.e., manageable risks) to 

2 individuals and populations during and following decommissioning (see Section 1.2).

3 • The NRC’s framework for decommissioning regulatory decision-making reflects the

4 iterative nature of the compliance demonstration process. The iterative approach to

5 decision-making for license termination provides a process for screening sites and for

6 directing additional data collection effort toward demonstrating compliance. The

7 framework is designed such that the level of complexity and rigor of analysis conducted

8 for a given site should be commensurate with the level of risk posed by the site (see

9 Section 1.4).

10 • This volume provides two different approaches for demonstrating compliance with the

11 dose-based decommissioning criteria, using either a dose modeling approach or a

12 DCGL approach. The dose modeling approach uses measurements of the actual

13 residual radioactivity at a site after cleanup to more realistically assess the potential

14 dose, and therefore the risk, associated with a decommissioned site. The DCGL

15 approach allows a licensee to calculate, a priori, a concentration limit (DCGL) for each

16 radionuclide based on the dose criteria of the LTR and to then demonstrate that the

17 residual radionuclide concentrations are below the DCGLs (see Section 2.5).

18 • This volume allows either a screening approach or a site-specific approach to

19 demonstrate compliance. The screening approach allows sites that pose lower potential

20 risks to demonstrate compliance through simpler, yet conservative, screening analysis

21 by adopting screening DCGLs developed by the NRC (see Sections 2.6 and 5.1 and

22 Appendix H).

23 • The NRC staff recommends using the data quality objectives (DQOs) process for

24 establishing criteria for data quality and developing survey designs. The process uses a

25 graded approach to data quality requirements, based on the type of survey being

26 designed and the risk of making a decision error based on the data collected. This

27 process aligns the resources expended to collect and analyze data with the

28 risk-significance of the data (see Section 3.2).

29 • The NRC provides for an approach to dose assessment that accounts for the

30 site-specific risk significance of radionuclides and exposure pathways. The NRC staff

31 allows a licensee to identify radionuclides and exposure pathways that may be

32 considered “insignificant,” based on their contribution to risk, and remove them from

33 further consideration (see Section 3.3). The NRC endorses the MARSSIM approach to

34 FSS design and execution, which results in a site-specific FSS design that is

35 commensurate with potential risks associated with a site, in terms of the likelihood of

36 exceeding the DCGLs at the site (see Section 4.4).

37 • The NRC staff supports a risk-informed approach to site-specific dose modeling for

38 compliance demonstration in several ways: (1) allowing for the site-specific selection of

39 risk-significant exposure scenarios, exposure pathways, and critical groups,

40 (2) expecting selection of conceptual models, numerical models, and computer codes

41 that incorporate the more risk-significant elements of a site, (3) expecting site-specific

42 data for the more risk-significant input parameters, and allowing for more generic data

43 for less risk-significant parameters, and (4) encouraging the use of probabilistic

44 techniques to evaluate and quantify the magnitude and effect of uncertainties in the risk
2-4

1 assessment, and the sensitivity of the calculated risks to individual parameters and 

2 modeling assumptions (see Appendix I and Q).

3 • The NRC allows for early partial release of a portion of a site before completion of

4 decommissioning for the entire site, based on the risks associated with the early partial

5 site release (PSR) (see Appendix K).

6 • The NRC staff supports a risk-informed graded approach for engineered barriers, and

7 this guidance includes an example of how the risk-informed approach is applied to

8 designing erosion protection barriers (see Appendix P). In addition, the staff supports a

9 risk-informed graded approach for selecting institutional controls and for long-term

10 monitoring and maintenance at restricted use sites, which allows licensees to tailor the

11 type of institutional controls and the specific restrictions on future site use based on a

12 risk framework and insights from dose assessments (see Section 17.7 and Appendix M

13 of Volume 1).

14 2.2 Flexibility in Submissions

15 The NRC staff expects that certain information will be included in licensees’ DPs, including the 

16 FSS design (if an FSS will be performed) and a description of the development of DCGLs or the 

17 dose assessment, as applicable. Volume 1 of this NUREG provides additional details on the 

18 expected content in these submittals. 

19

20 Some information is required by regulations (e.g., 10 CFR 30.36(g)(4)) and must be provided in 

21 the DP; the DP must include all of the following: 

22

23 • the conditions of the site, building, or area, sufficient to evaluate the acceptability of the

24 plan

25 • the planned decommissioning activities

26 • the methods used to ensure protection of workers and the environment against radiation

27 hazards during decommissioning

28 • the planned final radiation survey

29 • an updated cost estimate for decommissioning, comparison with decommissioning

30 funds, and a plan for ensuring the availability of adequate funds to complete

31 decommissioning

32 In addition, DCGLs are usually submitted in the DP. Therefore, the typical approach is for the 

33 licensee to obtain all the detailed information needed and to submit the information in the DP. 

34 Using the DP checklist (Appendix D of Volume 1 of this NUREG report) as a guide, licensees 

35 should coordinate with the NRC staff to determine what information should be included in the 

36 DP. For example, for a MARSSIM FSS, the licensee may perform sufficient characterization 

37 surveys to determine the appropriate number of samples to obtain for each survey unit. In this 

38 case, the NRC staff could approve both the survey design and the DP, and the FSS report 

39 (FSSR) may focus primarily on the results of the FSS.

40

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

42 For an FSS, the MARSSIM approach requires that certain information needed for the final 
2-5

1 radiological survey be developed as part of the remedial activities at the site; therefore, this 

2 information may not be available for the DP. Similarly, some aspects of the DCGL development 

3 or dose assessment may not be available before remediation and final surveys are complete. 

4

5 When some important information is not available at the time of the DP submission, licensees 

6 may either (1) make assumptions about the information or (2) commit to following a specific 

7 methodology to obtain the information. In the first case, assumptions will be considered by the 

8 NRC staff to be commitments to ensure and subsequently demonstrate that the assumption is 

9 true. The licensee would then submit the information at the completion of remediation, at the 

10 completion of FSS design, with the FSSR, or at some other appropriate time. For example, a 

11 facility uses the ratio of concentrations of thorium (Th)-232 to uranium (U)-238, along with 

12 measured concentrations of Th-232, in estimating the concentration of U-238. The licensee 

13 may have preliminary information about the ratio of concentrations and, if it is reasonable, may 

14 assume that that ratio would be valid for the conditions at the time of the FSS. The NRC staff 

15 could accept the use of the assumed value for the ratio. The licensee would demonstrate, at a 

16 later stage that the assumed value was valid, perhaps based on measurements made during 

17 the FSS.

18

19 In the second case, a licensee’s DP commits to following a specific methodology to obtain the 

20 information. For example, a facility may not have sufficient information at the time of the DP 

21 submission to determine the number of samples to be taken from each survey unit for a 

22 MARSSIM FSS. In this case, the licensee may commit to the procedure recommended in 

23 MARSSIM for determining the number of samples in a survey unit and document it in the DP. 

24

25 The licensee then may determine the number of samples for each survey unit as information is 

26 obtained. An FSSR could describe the FSS design, including the number of samples, which the 

27 NRC staff would evaluate, along with the FSS results.

28

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 licensee’s 

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. 

29

30 The licensee could take a similar approach to information needed to complete a dose 

31 assessment. One example is a facility for which the fraction of building-surface residual 

32 radioactivity that is removable has been determined during scoping surveys, but the licensee 

33 does not know whether remediation activities will change the fraction. In this case, the licensee 

34 might assume, for its dose assessment, that the measured fraction will remain unchanged. The 

35 NRC staff expects the licensee (1) to make measurements or calculations to demonstrate that 

36 the removable fraction was representative of the conditions when remediation is complete and 

37 (2) to demonstrate that the dose assessment is representative.

38

39 The NRC staff normally would not review DPs or FSSs that use assumptions in lieu of specific 

40 information that reasonably could be obtained before submission. In general, the NRC staff 

41 expects that assumptions used in developing DPs submitted for review would be limited to 

42 those parameters that could change as a result of the remediation or the FSS process itself or 
2-6

1 to those parameters for which information cannot reasonably be obtained at the time of DP 

2 submission. The NRC staff should consider other assumptions on a case-by-case basis.

3

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 facility’s 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.

4

5 2.3 Use of Characterization Data for Final Status Surveys 

6 Although the FSS is generally discussed as if it were an activity performed during a single stage 

7 of the radiation survey and site investigation (RSSI) process (see Chapter 4 and Table 4.1 for 

8 more about the RSSI process), this does not have to be the case. There is no requirement that 

9 an FSS be performed at the end of the decommissioning process. Data from other surveys 

10 conducted during the RSSI process—such as scoping, characterization, and remedial action 

11 support surveys—can provide valuable information for an FSS, provided the data are of 

12 sufficient quality.

13

14 In some cases, the data obtained from these other surveys may be sufficient to serve as an 

15 FSS. Licensees may plan the different phases of the RSSI to obtain data of sufficient quality to 

16 serve as or to supplement the FSS. The DQO process may be applied to all phases of the 

17 RSSI, with DQOs being as robust as those typically developed for the FSS. This approach may 

18 result in more costly characterization or remedial action support surveys (to support the more 

19 stringent DQOs), which may be balanced against the elimination of a separate FSS. 

20

21 2.4 Choice of Null Hypothesis for Final Status Survey Statistical Analysis

22 The default assumption used in the MARSSIM approach to FSSs and followed by the NRC staff 

23 is that the survey unit is considered contaminated above the limit, unless survey data show 

24 otherwise. Thus, the null hypothesis used for the MARSSIM FSS statistical tests is that the 

25 concentrations of residual radioactivity exceed the DCGLs. This assumption and null 

26 hypothesis are considered Scenario A. In most all cases, the NRC staff will consider 

27 Scenario A to be the appropriate choice. In some limited cases, a different assumption and null 

28 hypothesis, Scenario B, may be appropriate. Scenario B is typically used when the DCGL is 

29 within the range of background variability making it difficult to distinguish between residual 

30 radioactivity and background. This section (and Appendix A and G, Sections A.4.4 and G.6)

31 provide guidance on Scenario B. NUREG-1505, Revision 1, “A Proposed Nonparametric 

32 Statistical Methodology for the Design and Analysis of Final Status Decommissioning Surveys: 

33 Interim Draft Report for Comment and Use,” issued June 1998, forms the basis for most of the 

34 guidance on Scenario B and can be referred to for additional details. Table 2.1 summarizes the 

35 differences between Scenarios A and B.
2-7

1 Table 2.1 Comparison of FSS Statistical Test Scenarios

Characteristic Scenario A Scenario B

Assumption for statistical 

test

The survey unit is assumed to 

fail unless the data show it can 

be released.a

The survey unit is assumed to 

pass unless the data show 

that further remediation is 

necessary.

Null hypothesis The concentration of residual 

radioactivity in the survey unit 

exceeds the DCGLs.

The concentration of residual 

radioactivity in the survey unit 

is indistinguishable from 

background.

Scenario emphasis Compliance with a dose limit. Indistinguishable from 

background.

What is needed to reject 

the null hypothesis?

The concentration of residual 

radioactivity in the survey unit 

is less than the DCGL.

The concentration of residual 

radioactivity in the survey unit 

is distinguishable from 

background.

Rejecting the null

hypothesis means

The survey unit passes. The survey unit fails.

Increasing the number of 

measurements in a survey 

unit

Increases the probability that 

an adequately remediated 

survey unit will pass.

Increases the probability that 

an inadequately remediated 

survey unit will fail.

When should the scenario 

be used?

Should be used in most cases 

(i.e., default) when the DCGL 

is fairly large compared to 

measurement variability.

Should be used in special 

cases (i.e., exception) when 

the DCGL is 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.

2

3 Deciding which scenario to use and the process to make that decision are difficult issues. In 

4 most cases, when the area-wide DCGL (DCGLW) for the survey unit is large compared to the 

5 measurement variability, Scenario A should be chosen. This is because even residual 

6 radioactivity below the DCGLW should be measurable. In some cases, however, it may be more 

7 appropriate to demonstrate indistinguishability from background. When the DCGLW is small 

8 compared to measurement and/or background variability, Scenario B may be appropriate. This

9 is because residual radioactivity below the DCGLW may be difficult to measure. Background 

10 variability may be considered high when differences in estimated mean concentrations 

11 measured in potential reference areas are comparable to screening level DCGLs. Appendix G, 

12 Section G.6 of this volume provides an example of the use of Scenario B to demonstrate 

13 indistinguishability from background when the residual radioactivity consists of radionuclides 

14 that appear in background, and the variability of the background is relatively high (see also 

15 NUREG-1505 for additional details).

16

17 As mentioned above, the NRC staff’s default assumption is that the use of Scenario A is 

18 appropriate. The use of Scenario B is expected only for a small number of facilities, and the 
2-8

1 considerations for any given facility are expected to be site-specific. Therefore, the staff 

2 recommends that licensees contact the NRC early in the FSS design process to discuss 

3 considerations for their situation.

4

Cautions on the Use of Scenario B for FSS Statistical Tests

• Scenario B should typically only be used when residual radioactivity in the survey unit

is within the range of background variability, making it difficult to distinguish between

the residual radioactivity and background.

• 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.

5

6 2.5 Demonstrating Compliance Using Dose Assessment Methods versus 

7 Derived Concentration Guideline Levels and Final Status Surveys

8 There is flexibility in the general approach to demonstrating compliance with the dose criteria of 

9 10 CFR Part 20, Subpart E. Two major approaches are available, the dose modeling approach 

10 and the DCGL and FSS approach. The dose modeling approach involves characterizing the 

11 site post-remediation, if remediation is necessary, and performing a dose assessment. The 

12 DCGL and FSS approach requires the development of DCGLs and performing a FSS to 

13 demonstrate that the DCGLs have been met. Because the second option is more common, 

14 emphasis is placed on the use of DCGLs and FSSs as the compliance method in this NUREG. 

15 However, the two approaches are not mutually exclusive; a hybrid of both approaches can be 

16 used. In fact, the NRC recommends licensees assess the final estimated dose using data from 

17 the FSS even if the DCGL and FSS approach is used. Table 2.2 shows some advantages and 

18 disadvantages of the two approaches. 
2-9

1 Table 2.2 Comparison of Dose Modeling to DCGL and FSS Approaches to Compliance

Approach Advantages Disadvantages

Dose 

Modeling

• more realistic

• accounts for time of peak dose

for mixes of radionuclides

• can use additional data collected

during decommissioning for

site-specific analyses

• can guide remediation activities

and data collection

• may still need preliminary

cleanup goals or DCGLs to

design surveys or guide

remediation

• greater chance of additional

iterations of remediation and/or

site characterization

DCGLs and 

FSSs

• could be simpler to implement

• lower likelihood of not showing

compliance with dose criterion

after remediation

• by using sum of fractions,

provides level of conservatism

for radionuclide mix, if peak

dose from individual

radionuclides occurs at different

times

• additional modeling data (i.e., to

modify DCGLs) collected during

decommissioning cannot be

used without license

amendment

2

3 2.5.1 Dose Modeling Approach

4 Calculating the final dose is the most direct approach to show compliance with the dose criteria 

5 in Subpart E. Direct calculation of the total dose—from all radionuclides in a code that correctly 

6 accounts for the time of the peak dose for each radionuclide—is a more realistic measure of the 

7 potential dose from the site. Another advantage of this approach is that a licensee can use 

8 dose modeling information during the decommissioning process to guide additional site 

9 characterization, remediation, or other decommissioning options. Additional site 

10 characterization could be performed to reduce the level of conservatism in the dose model, 

11 parameters, or exposure scenario.

12

13 An advantage for sites that comply with the Subpart E criteria without any cleanup is that it may 

14 be unnecessary to create any DCGLs; however, the quality of the licensee’s site 

15 characterization data should be sufficient for use as an FSS.

16

17 A disadvantage of the dose modeling approach is that changes in the dose modeling, between 

18 the approval of the DP and the request for license termination, would result in requiring the NRC 

19 staff to review the new information before granting approval of license termination. This 

20 additional review step could result in further justification, modeling, remediation activities, or site 

21 characterization before approval is granted. This additional review step is similar to what can 

22 occur for a site that needs no remediation but uses site-specific dose modeling to show 

23 compliance as part of the DP.
2-10

1 Another disadvantage of using the dose modeling approach is that cleanup goals or final 

2 concentrations may need to be estimated (1) to provide assurance that the approach will result 

3 in compliance and (2) to design quality surveys, guide remediation activities, and perform 

4 additional site characterization.

5

6 2.5.2 DCGL and FSS Approach

7 For many sites, especially those that need remediation, the DCGL and FSS approach is a 

8 simpler system to show compliance with Subpart E. The DCGL and FSS approach is the one 

9 most commonly used for compliance with the LTR and is the one MARSSIM recommends. In 

10 the DCGL and FSS approach, the licensee commits to a single concentration value for each 

11 radionuclide (i.e., DCGLW) that results in a dose equal to the dose criterion. The DCGLW

12 derivation can use either generic screening criteria or site-specific analysis. The licensee then 

13 uses FSSs to demonstrate that the DCGLs have been met. For sites with multiple radionuclides 

14 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 15 The 

16 disadvantages of this approach include the following: 

17

18 • The sum of fractions approach (Section 2.7) has an underlying assumption that the peak

19 dose for every radionuclide occurs at the same time. This can result in an additional

20 level of conservatism, depending on the mix of radionuclides.

21 • Any changes in the DCGLs (e.g., because of new site information) may require a license

22 amendment and NRC staff review.

23 2.6 Merits of Screening versus Site-Specific Dose Assessments

24 The advantage of screening level analyses is that they require minimal justification, 

25 characterization, and NRC staff review. The disadvantages are that (i) only residual 

26 radioactivity associated with buildings surfaces, and surface soils are considered (may not be 

27 appropriate for subsurface residual radioactivity), and (ii) in most cases, screening values are 

28 expected to be more restrictive than DCGLs derived using site-specific dose modeling. While 

29 site-specific analyses allow more flexibility in estimating the risk for a particular site compared to 

30 screening-level analyses, site-specific analyses require site-specific information, and therefore, 

31 resources must be spent on obtaining data or providing support for especially risk-significant 

32 parameter values. Changes in pathways, exposure scenarios, and conceptual models may also 

33 require supporting information. Table 2.3 provides a brief summary of the attributes and merits 

34 of both approaches.

35

36 The models, exposure scenarios, and parameters used in screening are intended to be 

37 conservative, because the lack of information about a site warrants the use of conservative

38 models and default conditions to ensure that the derived dose is not underestimated. The 

39 screening analysis is intended to overestimate the dose, to ensure that, for 90 percent of the 

40 screening cases, the derived dose is not underestimated. In performing a screening analysis, 

41 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. 
2-11

1 calculating compliance, whereas in the site-specific analysis, the “peak of the mean” dose over 

2 time (e.g., 1,000 years) may be used. Deterministic analyses may also be used with sufficient 

3 support for those parameters that have a significant impact on dose as identified through 

4 sensitivity analysis. An analysis is considered to be site-specific when default parameters in the 

5 DandD code used for deriving screening values are changed, source term conditions are 

6 modified, or different models or codes are used.

7

8 2.7 Sum of Fractions

9 The sum of fractions is a simple, yet flexible, approach to deal with multiple radionuclides or 

10 sources. A source is any discrete material or medium that contains residual radioactivity. For 

11 example, a site with residual radioactivity in surface soil, groundwater, and buried piping has at

12 least three sources. The DCGLW is equivalent to the concentration of a single radionuclide from 

13 a single source that would provide 0.25 millisieverts per year (mSv/y) (25 millirem (mrem)/y) 

14 total effective dose equivalent (TEDE). The dose from each radionuclide and source should be 

15 calculated and then added together. If a licensee only complied with the DCGLW for each 

16 radionuclide in each source, the resulting total dose could be as high as 0.25 mSv/y 

17 (25 mrem/y) multiplied by the number of radionuclides multiplied by the number of sources. 

18 Unless there was only one source and one radionuclide, the resulting dose would not meet the 

19 limits detailed in Subpart E. The dose from all the radionuclides and sources must be equal to 

20 or less than the appropriate dose limit in Subpart E.

21

22 One simple way to calculate the dose from one radionuclide from one source is to calculate the 

23 relative ratio of the residual radioactivity concentration over the DCGLW. Then, the ratio is 

24 multiplied by 0.25 mSv/y (25 mrem/y). In fact, for multiple sources or radionuclides, the ratios 

25 can be added together, and the sum multiplied by the dose limit. Therefore, the sum of the 

26 ratios for all the radionuclides and sources may not exceed “1” (i.e., unity). For example, if 

27 radionuclides A and B are present at respective concentrations of Conc A and Conc B, and if 

28 the respective applicable DCGLs are Limit A and Limit B, then the concentration needs to be 

29 limited so that the following relationship exists to meet Subpart E:

30

Conc A

Limit A

Conc B

Limit B

31

+  1

(2-1)

32 Similarly, for multiple sources, the sum of the ratios resulting from the sum of the radionuclide 

33 contributions may not exceed unity. For example, if the site had a second source, also with 

34 radionuclides A and B, but in concentrations of Conc A0 and Conc B0, and DCGLs of Limit A0

35 and Limit B0, the following condition would need to be satisfied to meet Subpart E:

36

Conc A

Limit A

Conc B

Limit B

Conc A

Limit A

Conc B

Limit B

+ + + 

0

0

0

0

37

1

(2-2)
2-12

1 Table 2.3 Attributes of Screening and Site-Specific Analysis

Attribute Screening Site-Specific

Models/Codes DandD Version 2 or later 

versions

Any model/code compatible 

with site conditions and 

approved by the NRC staff.

Scope of Application Only for sites that meet the 

requirements for screening

analysis.

Any site

Parameters DandD default parameters Site-specific (physical 

parameters)2

Exposure 

Scenarios/Pathways

DandD default exposure 

scenarios/pathways

Exposure scenarios/pathways 

may be modified, based on site 

conditions.

Dose Metric and

Consideration of 

Uncertainty

The dose at the 90th percentile 

of the peak dose distribution 

within 1,000 years

Peak annual dose from a 

deterministic analysis or “peak 

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.

2

3 In the general form, the relationship of the ratios, commonly known as the “sum of the fractions” 

4 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. 
2-13

∑ ∑

Concs,r

Limits,r

N

r=1

M 1 s=1 ≤ 1 (2-3)

2

3 where Concs,r = the concentration of radionuclide r in source s, and

4 Limits,r = the DCGLW value for radionuclide r in source s.

5 For sites with a number of radionuclides and sources, it may be easier to partition the 

6 acceptable fraction between various sources or radionuclides. For example, a licensee could 

7 commit to keeping the ratio from the groundwater to less than 25 percent of the dose limit. 

8

9

10 One major, implicit assumption in using the sum-of-fractions approach is that peak doses for 

11 each radionuclide and source occur simultaneously. Because radionuclides can be transported 

12 through the environment at different rates, and a particular radionuclide may be dominated by a 

13 different pathway compared to another radionuclide, there are many radionuclides and 

14 contaminated media for which peak doses do not occur simultaneously. For example, 

15 radionuclides that result in predominantly external dose and are short-lived, such as cobalt 

16 (Co)-60, usually have a peak dose right after license termination. For radionuclides that result 

17 in peak dose through irrigation or drinking groundwater, the peak dose may not occur until years 

18 after license termination. When peak doses are from different radionuclides or sources occur at 

19 different times, the sum-of-fractions approach tends to overestimate that dose. In some 

20 situations, the overestimate may be significant and affect the compliance demonstration. The 

21 licensee could directly calculate the combined dose using final concentrations from the FSS to 

22 more accurately estimate the risk from the site (see Section 2.5).

23

24

25 2.8 Flexibility for Use of Institutional Controls and Engineered Barriers at 

26 Restricted Use Sites

27 The new guidance developed for restricted use sites includes risk-informed and 

28 performance-based approaches to institutional controls, engineered barriers, monitoring, and 

29 maintenance. These approaches not only enhance the attention to safety by being risk 

30 informed but also provide flexibility to licensees planning restricted use for a site. The 

31 approaches described allow licensees to select the most effective and efficient methods for 

32 restricting site use, designing engineered barriers to mitigate disruptive processes important to 

33 compliance, and planning monitoring and maintenance activities that are tailored to the specific 

34 site and indicators of potential disruptive processes and engineered barrier performance. 

35 Section 3.5, and Appendix P of this volume and Section 17.7 and Appendix M to Volume 1 

36 describe these approaches.

3-1

1 CROSS-CUTTING ISSUES

2 This chapter provides guidance on several cross-cutting issues that relate to multiple aspects of 

3 surveys, characterization, and dose modeling. The issues addressed in this chapter include the 

4 following:

5

6 • transparency and traceability of compliance demonstrations

7 • DQO process

8 • insignificant radionuclides and exposure pathways

9 • considerations for other constraints on allowable levels of residual radioactivity

10 • the use of engineered barriers

11 • 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 staff’s 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.

12

13 3.1 Transparency and Traceability of Compliance Demonstrations

14 Licensees submit information to justify their conclusions about compliance with 10 CFR Part 20, 

15 Subpart E. Because of insufficient justification, the NRC staff has found a number of licensee 

16 submittals to be inadequate for NRC to conclude it has reasonable assurance that the license 

17 termination rule criteria can be met. This section describes some considerations for improving 

18 the thoroughness of licensee submittals. Transparency refers to arguments or calculations with 

19 descriptions sufficient for an independent reviewer to replicate. Traceability refers to the

20 sources of information being relatable to the original source. The NRC staff encourages 
3-2

1 licensees to submit compliance demonstrations that are transparent and traceable. This should 

2 result in more efficient and effective staff reviews.

3

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

5 justification:

6 • describe the sources of data

7 • provide only summary data, if appropriate.

8 • to the extent that summary data are provided, include references to detailed data and

9 make them available to the NRC staff for review if requested (e.g., on an inspection)

10 • clearly describe the data, including units used, in tables and other presentations

11 • state the assumptions and ensure that the difference between assumptions and justified

12 data or parameters is clear

13 • provide justifications for parameters or arguments, especially when employing

14 nonstandard arguments or nondefault parameters

15 • describe uncertainties in data and parameters

16 3.2 Data Quality Objectives Process

17 Compliance demonstration is the process that leads to a decision as to whether or not a survey 

18 unit meets the release criteria. For most sites, this decision is supported by statistical tests 

19 based on the results of one or more surveys. In most cases, the initial assumption used by the 

20 NRC staff is that each survey unit is contaminated above the release criteria until proven 

21 otherwise (Scenario A). The surveys are designed to provide the information needed to reject 

22 this initial assumption. The NRC staff recommends using the Data Life Cycle as a framework 

23 for the planning, implementation, assessment, and decision-making phases of final surveys. 

24 Section 2.3 of MARSSIM discusses the major activities associated with each phase of the Data 

25 Life Cycle. 

26

27 One aspect of the planning phase of the Data Life Cycle is the DQO process, which is a series 

28 of planning steps for establishing criteria for data quality and developing survey designs. The 

29 DQO process consists of seven steps:

30

31 (1) statement of the problem

32 (2) identification of the decision

33 (3) identification of inputs to the decision

34 (4) definition of the study boundaries

35 (5) development of a decision rule

36 (6) specification of limits on decision errors

37 (7) optimization of the design for obtaining data
3-3

1 The output from each step influences steps later in the process. Even though the DQO process 

2 is depicted as a linear sequence of steps, it is iterative in practice; the outputs of one step may 

3 lead to reconsideration of prior steps.

4

5 The DQO process uses a graded approach to data quality requirements, defined according to 

6 (1) the type of survey being designed and (2) the risk of making a decision error based on the

7 data collected. This approach provides a more effective survey design, combined with a basis

8 for judging the usability of the data collected. Thus, the DQO process is a flexible planning tool

9 that licensees can use more or less intensively as the situation requires.

10

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

12 following:

13

14 • clarifying the study objective

15 • defining the most appropriate type of data to collect

16 • determining the most appropriate conditions for collecting the data

17 • specifying limits on decision errors that will be used as the basis for establishing the

18 quantity and quality of data needed to support the decision

19 Although the DQO process is generally used for surveys and the steps of an RSSI, the general 

20 concepts may also be applied to dose assessments. Chapter 5 contains additional information 

21 on data requirements related to dose modeling and DCGL development. Data of sufficient 

22 quality should be collected to ensure the technical defensibility of the modeling results focusing 

23 on risk-significant parameters. The DQO process should be used to guide data collection and 

24 analysis. Licensees are encouraged to apply the general concepts of the DQO process to all 

25 applicable parts of their compliance demonstration. The use of the DQO process can help 

26 ensure that the type, quantity, and quality of data and calculations used in decision-making will 

27 be appropriate for the intended application. Additional guidance on the use of the DQO process 

28 is in MARSSIM Section 2.3 and Appendix D and in EPA/600/R-96/055, “Guidance for the Data 

29 Quality Objectives Process,” issued August 2000. 

30

31 Experience has shown that it is helpful for the licensee to identify all appropriate DQOs in 

32 planning and designing the final status survey plan (FSSP). The process of identifying the 

33 applicable DQOs ensures that the survey plan requirements, survey results, and data evaluation 

34 are of sufficient quality, quantity, and robustness to support the decision on whether the cleanup 

35 criteria have been met. 

36

37 In purpose and scope, the DQO process can include a flexible approach for planning and 

38 conducting surveys and for assessing whether survey results support the conclusion that 

39 release criteria have been met. The DQO process can be an iterative process that continually 

40 reviews and integrates, as needed, new information in decision-making and the design of the 

41 final survey plan. Finally, the selection and optimization of DQOs should facilitate the later 

42 evaluation of survey results and decision-making processes during the data quality assessment 

43 (DQA) phase. Licensees have had difficulty developing DQOs, especially during the 

44 optimization step, and have not taken full advantage of the DQO process. Experience has 

45 shown that licensees often rigidly structure the process by relying too much on characterization 

46 data and are not readily open to the possibility of incorporating new information as it becomes 

47 available. This rigid approach makes implementing any changes difficult and is an inefficient 
3-4

1 use of resources that imposes time delays (e.g., the additional time required to determine how 

2 to implement any changes). 

3

4 3.3 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.

5

6 This section provides guidance on conditions under which radionuclides or exposure pathways 

7 may be considered insignificant and may be eliminated from further consideration. The dose 

8 criteria in 10 CFR Part 20, Subpart E, apply to the total dose from residual radioactivity. Thus, 

9 demonstrations of compliance should generally address the dose from all radionuclides and all 

10 exposure pathways. However, the NRC staff has determined that it is reasonable to eliminate 

11 radionuclides or pathways that are insignificant contributors to dose from further detailed 

12 consideration, although their dose contributions must be considered in demonstrating 

13 compliance with the radiological criteria for license termination. 

14

15 The NRC staff considers radionuclides and exposure pathways that contribute no more than 

10 percent of the dose criteria, considering uncertainty1 16 , to be insignificant contributors to dose. 

17 Because the dose criteria are performance criteria, this 10-percent limit is an aggregate 

18 limitation only. That is, the sum of the dose contributions from all radionuclides and pathways 

19 considered insignificant should be no more than 10 percent of the dose criteria (e.g., no more 

20 than 0.025 mSv/y (2.5 mrem/y) for the unrestricted release limit of 0.25 mSv/y (25 mrem/y)). No 

21 limitation on either single radionuclides or individual pathways is necessary. In cases of 

22 restricted release, where two dose criteria apply (one for institutional controls in place and one 

23 that considers the possibility of restrictions failing), the 10-percent limitation should be met for 

24 both dose criteria. In making a determination that radionuclides or pathways are insignificant, 

25 licensees should consider both reasonably foreseeable and less likely but plausible exposure 

26 scenarios (see Section 5 for more information). Licensees should also consider the presence of 

27 elevated areas, and potential in-growth of progeny from postulated insignificant radionuclides, 

28 when determining that the radionuclides or pathways contribute no more than 10 percent of the 

29 dose criteria and are, therefore, insignificant.

30

31 Typically, licensees would use characterization data to show that certain radionuclides or 

32 pathways are insignificant before remediation or the FSS. However, if remediation is planned, 

33 licensees may also be able to show through analysis that the dose contributions of certain 

34 radionuclides following remediation will contribute no more than 10 percent of the dose criteria. 

35

36 The approach discussed in the preceding paragraph (accounting for removal of radioactivity 

37 during remediation) is similar to the use of surrogate radionuclides that implicitly considers the 

38 dose contributions of hard-to-detect radionuclides (see Appendix A of this document and 

39 Section 4.3.2 of MARSSIM, Revision 1 for additional information on use of surrogate 

40 radionuclides). However, unlike surrogate radionuclides, the licensee has an additional burden 

41 of showing that the dose contributions of the radionuclides or pathways are insignificant. 

42 Characterization data can be used to define the relative activities of significant and 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 and 

Appendix I have more information on consideration of uncertainty.
3-5

1 radionuclides to determine the relative dose contributions of the radionuclides present at the 

2 site. Ratios should be conservatively selected so that they do not underestimate the potential 

3 dose contributions of the insignificant radionuclides (e.g., use of minimum detectable 

4 concentrations (MDCs) for undetected radionuclides, and use of the 95th percentile ratios of 

5 insignificant to significant radionuclides). The licensee should also show that the relative dose 

6 contributions of the insignificant radionuclides will not increase following remediation because of 

7 an increase in concentration or redistribution of residual radioactivity. In general, the NRC does 

8 not require post-remediation sampling of the insignificant radionuclides, due to their low risk 

9 significance. However, if there is a valid concern that the dose contributions of the postulated 

10 insignificant radionuclides could be significant following remediation, licensees may choose to 

11 manage this uncertainty as part of the DQO process (e.g., through post-remediation sampling of 

12 the insignificant radionuclides, similar to the approach used for surrogate radionuclides 

13 discussed in MARSSIM Section 4.3.2).

14

15 Once a licensee has demonstrated that radionuclides or exposure pathways are insignificant, 

16 then (1) the dose from the insignificant radionuclides and pathways must be accounted for in 

17 demonstrating compliance, but (2) the insignificant radionuclides and pathways may be 

18 eliminated from further detailed evaluations. For example, after sufficient site characterization, 

19 suppose a licensee shows that the dose from strontium (Sr)-90 at the facility is 0.02 mSv/y 

20 (2 mrem/y), which is less than 10 percent of the dose criterion for unrestricted use. In this case, 

21 Sr-90 can be considered insignificant and eliminated from the FSS and from detailed 

22 consideration in the dose modeling. However, the dose from Sr-90 has to be considered in 

23 demonstrating compliance with the dose criterion. In some cases, licensees may be able to 

24 show that the dose contributions of the insignificant radionuclides and pathways are much less 

25 than 10 percent of the dose criteria. However, it may be less burdensome on the licensee to 

26 show that the insignificant radionuclides and pathways contribute less than 0.025 mSv/y 

27 (2.5 mrem/y) than it is to show that the insignificant radionuclides and pathways contribute much 

28 less than 0.01 mSv/y (1 mrem/y), for example. Thus, it may be preferable to round the dose 

29 contributions of the insignificant radionuclides and pathways higher, and thereby provide a 

30 greater safety margin in meeting the dose criteria.

31

32 It is important for the licensee to document the radionuclides and pathways that it has 

33 considered insignificant and eliminated from further consideration and for the licensee to justify 

34 the decision to consider them insignificant. However, licensees and the NRC staff should be 

35 aware that remediation techniques (or other activities or processes) may result in an increased 

36 dose from the postulated insignificant radionuclides or pathways, such that the dose 

37 contributions are no longer insignificant. Thus, licensees should also demonstrate that the dose 

38 contributions of insignificant radionuclides and pathways deemed insignificant will not increase 

39 (or were not underestimated) as a result of other activities. 
3-6

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.

1

2 3.4 Considerations for Other Constraints on Allowable Residual Radioactivity

3 Situations or standards other than the dose criteria and ALARA requirements of Subpart E may 

4 constrain the final dose below 0.25 mSv/y (25 mrem/y). Two main causes for constraining the 

5 Subpart E dose limit are (1) a PSR and (2) other standards or regulations.

6

7 A PSR occurs when a licensee releases a portion of its site for unrestricted use before 

8 terminating the entire license. While the licensee should demonstrate that the residual 

9 radioactivity at the time of unrestricted release of the specific area meets the Subpart E dose 

10 limit, the residual radioactivity of the area should also be taken into account during final 

11 termination to demonstrate that the entire site has met the appropriate release criteria. 

12 Appendix K of this volume discusses dose modeling considerations for PSR. In general, the 

13 comments below are also applicable to PSRs.

14

15 Demonstrating compliance with the Subpart E dose limit does not eliminate the licensee’s 

16 requirement for meeting other applicable Federal, State, or local rules and regulations. These 

17 regulations from other governmental agencies may conflict with the requirements of Subpart E, 

18 as they may allow higher or lower levels of residual radioactivity on the site or may conflict in 

19 other ways, such as limiting decommissioning options or final status. Nevertheless, the staff 

20 should review a DP for compliance only with NRC requirements, including 10 CFR Part 20, 

21 Subpart D, which incorporates, where applicable, the requirements of 40 CFR Part 190, 

22 “Environmental Radiation Protection Standards for Nuclear Power Operations.” For example, in 

23 reviewing the appropriateness of proposed DCGLs or the number of samples per survey unit for 

24 an unrestricted site, the NRC staff would use the limit of 0.25 mSv/y (25 mrem/y), and not a 

25 State’s limit of 0.2 mSv/y (20 mrem/y). Thus, any requests for additional information would also 

26 be based on compliance with the limit of 0.25 mSv/y (25 mrem/y). Licensees should note that 

27 use of a lower dose standard does not necessarily lead to a lower clean-up level because of 

28 differences in assessment approaches including differences in exposure scenarios, models and 

29 parameters used by different agencies setting the dose standards.
3-7

1 3.5 Use of Engineered Barriers

2 This section, and the details provided in Appendix P, provide guidance to licensees considering 

3 the use of engineered barriers (e.g., engineered covers, including those designed for erosion 

4 protection, stabilizing cementitious materials, and reactive walls) to demonstrate compliance 

5 with radiological criteria for license termination. Section 3.5 and Appendix P also support 

6 Section 17.7.3 of Volume 1 by giving guidance on the information expected to be submitted in a 

7 DP for the engineered barrier analysis and the technical basis for engineered barrier 

8 performance.

9

10 In the Commission’s view, engineered barriers are distinct and separate from institutional 

11 controls (see Decommissioning Criteria for the West Valley Demonstration Project (M-32) at the 

12 West Valley Site: Final Policy Statement (67 FR 5003)). Generally, institutional controls are 

13 designed to restrict access, whereas engineered barriers are usually designed to inhibit water 

14 from contacting waste, limit releases, or mitigate doses to intruders.

15

16 Engineered barriers are passive, man-made structures or devices intended to enhance a 

facility’s ability to meet the dose criteria in the LTR.

2 17 Engineered barriers are usually designed 

18 to inhibit water from contacting waste and releasing radionuclides to groundwater, thereby 

19 reducing exposure from ingestion of contaminated water. Engineered surface barriers may also 

20 slow erosion or otherwise decrease the likelihood of the waste being exposed at the surface 

21 through human or biotic activity, thereby reducing dose from inadvertent intrusion and direct 

22 exposure to the waste. In some cases, engineered barriers may also be used to passively limit 

23 access of critical groups to residual radioactivity (e.g., a durable rock barrier that decreases the 

24 likelihood of excavation or drilling into residual radioactivity). Used in the general sense, an 

25 engineered barrier could be one of a broad range of barriers with varying degrees of durability, 

26 robustness, and isolation capability.

27

28 On the other hand, institutional controls are active controls that limit access to the site, and the 

29 use of it, to minimize disturbances to engineered barriers and ensure that the exposure from the 

30 residual radioactivity does not exceed the established criteria. Institutional controls include legal 

31 mechanisms (e.g., land use restrictions) and physical controls (e.g., signs, markers, 

32 landscaping, and fences) to prevent unauthorized access to the site and minimize disturbances 

33 to engineered barriers. Institutional controls may require financial assurance to ensure 

34 adequate control and maintenance of the site, legal enforceability, and an entity with the 

35 capability, authority, and willingness to enforce the controls. 

36

37 The licensee determines the functionality and robustness of barriers using the risk-informed 

38 graded approach described in Appendix P and evaluated on a site-specific basis for each 

39 license application. However, the general framework that a licensee should consider would not 

40 vary from licensee to licensee; only the depth and breadth of information supplied to 

41 demonstrate the performance of the engineered barriers may vary. Appendix P provides the 

42 general framework a licensee should consider for use of engineered barriers in the 

43 decommissioning process. 

44

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

46 However, there may be infrequent cases where engineered barriers are used as one 

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.
3-8

1 component of a decommissioning approach to achieve unrestricted use of a site. These cases 

2 should be infrequent because of the uncertainty associated with the long-term performance of 

3 engineered systems without monitoring and maintenance and because the goal should be to 

4 achieve unrestricted use without relying on engineered barriers. If an engineered barrier is used 

5 at an unrestricted use site, only the passive performance of the barrier to mitigate radiological 

6 impacts may be credited (i.e., performance of the barrier without monitoring, inspection, and 

7 maintenance) in the dose assessment to demonstrate compliance with the LTR dose criteria. 

8 The assessment of engineered barrier performance should consider reasonably foreseeable, as 

9 well as less likely but plausible, disruptive conditions from human activities and from natural 

10 events and processes. Results of reasonably foreseeable exposure scenarios should be strictly 

11 considered when demonstrating compliance with radiological criteria for license termination, 

12 while less likely but plausible exposure scenarios should also be considered to help risk-inform 

13 the decision-making process. 

14

15 Proposals to use engineered barriers for unrestricted use sites should be part of an overall plan 

16 for decontaminating and decommissioning a site, as presented in a licensee’s DP or LTP. The 

17 licensee should demonstrate that, at the time of license termination, the site meets LTR criteria, 

18 including the criterion to reduce residual radioactivity to ALARA levels. Therefore, licensees 

19 should consider whether the following would be consistent with ALARA requirements in the 

20 absence of engineered barriers: (1) removal and disposal of contaminated components and 

21 equipment, (2) decontamination (and demolition, if appropriate) of buildings, (3) removal and 

22 disposal of waste streams remaining on site from past operations, and (4) excavation and 

23 removal of large areas of soil contamination as waste. Therefore, while engineered barriers 

24 may be proposed to assist with meeting the LTR criteria, source removal must first be 

25 considered in demonstrating that residual radioactivity has been removed to ALARA levels. 

26 Chapter 6 and Appendix N contain additional information.

27

28 Similarly, for restricted use sites, under 10 CFR 20.1403(a) the licensee must show that further 

29 removal of residual radioactivity would result in net public or environmental harm or that leaving 

30 the residual radioactivity in place is ALARA. Licensees should also include other considerations 

31 (e.g., distance to disposal facility, efficient use of available disposal capacity at the offsite 

32 facility, unavailability of required treatment options, lack of disposal options other than leaving 

33 the contaminated materials on site, and the need to use funds to remediate nonradioactive 

34 hazards at the same site), if applicable and appropriate, in its determination of whether 

additional removal of residual radioactivity is reasonably achievable.

3 35 In their proposal to use 

36 engineered barriers, licensees should include all relevant information concerning the risks of 

37 using the proposed approach versus other remediation alternatives. 

38

39 Because of the wide range of licensed decommissioning sites, the LTR and the NRC’s 

40 decommissioning guidance are not prescriptive as to the criteria for, or acceptability of, 

41 site-specific engineered barriers. Therefore, the licensee has flexibility in the methods used to 

42 demonstrate compliance with the performance-based criteria of 10 CFR Part 20, Subpart E. 

43 Because of this flexibility and because engineered barrier designs are site-specific, it is very 

44 important for the licensee to clearly and completely document how it has considered site45 specific conditions (e.g., site-specific resources, climate, degradation mechanisms) in its 

46 engineered barrier designs and monitoring and maintenance program. 

47

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.
3-9

1 Appendix P provides the detailed framework for applying engineered barriers to achieve

2 decommissioning at a site, and an example of a graded approach to erosion covers. A 

3 summary of existing guidance and reference information on the application of engineered 

4 barriers at decommissioning sites is also listed in Appendix P. 

5

6 In summary, the following points should be considered when applying engineered barriers to 

7 achieve decommissioning at a site:

8

9 • Engineered barriers are distinct and separate from institutional controls.

10 • Only the passive performance of engineered barriers (i.e., no monitoring and

11 maintenance) can be relied on at unrestricted use sites.

12 • Passive performance and (when institutional controls are assumed to be in place) active

13 monitoring and maintenance of engineered barriers can be relied on at restricted use

14 sites. Active monitoring and maintenance are considered institutional controls and may

15 enhance the level of performance assumed for an engineered barrier beyond what is

16 assumed for a barrier’s passive performance. For most cover systems, long-term,

17 passive performance is not well understood.

18 • For licensees pursuing unrestricted use of their site, residual radioactivity has been

19 reduced to levels that are ALARA before reliance on engineered barriers to meet LTR

20 criteria. Furthermore, for licensees pursuing restricted use of their site, a demonstration

21 is required to show that additional removal of residual radioactivity to meet unrestricted

22 use criteria would result in net public or environmental harm or that further reductions

23 are not being made because levels that meet restricted use conditions are ALARA.

24 • Engineered barrier evaluation are reviewed on a case-by-case basis using a risk25 informed approach.

26 3.6 Surveying and Considering Risk Associated with Surface and Subsurface 

27 Soils

28 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 29

30 MARSSIM survey protocols, discussed in more detail in Chapter 4 and Appendices A–G, 

31 assume that residual radioactivity is present on the surfaces of buildings and soils. Likewise, 

32 certain dose modeling codes such as DandD, discussed in more detail in Chapter 5 and 

33 Appendices H and I, also make assumptions about the depth of residual radioactivity that could 

34 be important for assessing dose to potential receptors. For example, surface soils are important 

35 for assessing dose from certain exposure pathways such as direct radiation exposure, incidental 

36 soil ingestion, and inhalation. While surface soils could also support other important exposure 

37 pathways, such as plant growth and consumption, contaminated subsurface soils could also 

38 contribute significantly to radionuclide uptake and dose dependent on the plant type and 

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

40 underestimated.

4 For example, surface soil has been associated with the top 15 cm of soil that can typically be measured using scan 

instrumentation.
3-10

1 Human activities could cause subsurface soils to become surface soils from mixing or 

2 redistribution after license termination. For example, tilling soil to promote crop growth or

3 excavating soil for home construction or well drilling could lead to the redistribution of 

4 subsurface soils to the surface (see Appendix J). Before license termination, remediation of the 

5 site and sorting of relatively contaminated and uncontaminated soils could lead to the 

6 redistribution or reuse of contaminated subsurface soils on the surface (see Appendix G). While 

7 the definition of surface soil is an important consideration when it comes to surveying and 

8 assessing the dose impacts associated with residual radioactivity remaining at a site at the time 

9 of license termination, there is no clear line of demarcation of surface and subsurface soils; the 

10 final distribution of residual radioactivity should be considered in dose modeling and the final 

11 status survey design. Potential redistribution of radioactivity following license termination must 

12 also be considered when deriving subsurface soil DCGLs or when performing dose modeling to 

13 assess the impact of subsurface residual radioactivity (Appendix J). 

14

15 Despite the linkages to scan instrument capability and dose modeling discussed above, the 

16 depth of residual radioactivity in surface and subsurface soils assumed for the purpose of the 

17 FSS design should not be arbitrarily selected based on dose modeling assumptions or scan 

18 capability. Rather, the expected horizontal and vertical extent of residual radioactivity informed 

19 by characterization surveys should be considered in the FSS design and in assessing dose. 

20 Significant heterogeneity in residual radioactivity concentrations should also be considered 

21 when designing the FSS and assessing dose. Chapters 4 and Appendices A–G, provide 

22 guidance on for performing radiological surveys for surface and subsurface soils; building 

23 materials; and surface water and groundwater (Appendix F); Chapter 5, and Appendices I and J 

24 have guidance on evaluating the dose impacts associated with surface, subsurface, and 

25 heterogeneous residual radioactivity in soils and building surfaces. 

26

27 If subsurface residual radioactivity is present, soil sampling will likely be necessary to 

28 supplement scan surveys to adequately characterize the full vertical extent of residual 

29 radioactivity. Depth discrete sampling of soils may also be needed if there is significant vertical 

30 heterogeneity, and representation of the vertical heterogeneity is important in assessing dose. 

31 DCGLs derived from dose modeling should be consistent with the actual vertical extent of the 

32 residual radioactivity and significant vertical heterogeneity considered. For example, it may be 

33 important to differentiate soil layers based on vertical heterogeneity and derive more than one 

34 set of DCGLs (e.g., derive surface and subsurface soil DCGLs), if vertical heterogeneity is found 

35 to be important to dose. Alternatively, dose modeling could be used to confirm that radiological 

36 criteria for license termination are met based on the final configuration of residual radioactivity in 

37 soils, as measured in FSSs. Chapter 5 and Appendix I contain additional information on the

38 dose modeling approach. Appendix J contains information on scenarios that should be 

39 considered for buried residual radioactivity. Appendix I, Section I.2 specifically contains 

40 additional information on source term abstraction for heterogenous distributions. 
4-1

1 FACILITY RADIATION SURVEYS

2 4.1 Radiation Survey and Site Investigation Process

3

4

5

6

7

8

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. 

9 Table 4.1 Cross-References for Principle Steps in the Radiation Survey and Site 

10 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 

Survey

Section 4.2 of this volume

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

11 Note: As of the date of publication of this NUREG, MARSSIM, Revision 2, is in the process of being 

12 published. Because MARSSIM, Revision 2, has not yet been published, references to sections of 

13 MARSSIM in this volume are with respect to MARSSIM, Revision 1 (i.e., section numbers may be 

14 different in Revision 2). NRC staff plans to incorporate MARSSIM, Revision 2, into a future revision of

15 this volume.

16

17 4.1.1 Historical Site Assessment

18 The RSSI process uses a graded approach that starts with Site Identification and the HSA and 

19 is later followed by other surveys that lead to the FSS. In most cases, the radiological status of 

20 a site will already be known based on its prior use and the presence of radioactive material

21 although a records review could identify areas of use, disposal, or spills that have been 

22 overlooked. The HSA collects existing information describing a site’s complete history from the 

23 start of site activities to the present. The necessity for detailed information and the amount of 

24 effort to conduct an HSA depend on the type of site, associated historical events, regulatory 

25 framework, and availability of documented information. The main purpose of the HSA is to 

26 determine the current status of the site or facility, but the data collected may also be used to 

27 differentiate sites that need further action from those that pose little or no threat to human health 

28 and the environment (see Section 2.3). This screening process can provide a site disposition 

29 recommendation or propose additional surveys. Because much of the data collected during 
4-2

1 HSA activities are qualitative or are analytical data of unknown quality, many decisions on a site 

2 are the result of professional judgment.

3

4 The primary objectives of the HSA include the following:

5

6 • identify potential sources of residual radioactivity

7 • determine if sites pose a threat to human health and the environment

8 • differentiate impacted from nonimpacted areas

9 • provide input to scoping and characterization survey designs

10 • assess the likelihood of residual radioactivity migration

11 • identify additional potential radiation sites related to the site being investigated

12 The HSA typically consists of three phases: (1) preliminary investigation of the facility or site, 

13 (2) site visits or inspections, and (3) an evaluation of the site based on the information collected.

14 The HSA should identify special survey situations that may need to be addressed, such as

15 subsurface radioactivity; sewer systems, waste plumbing, and floor drains; ventilation ducts; and

16 embedded piping containing residual radioactivity. Appendix G of this volume includes

17 information on special survey situations. Section 2.4.2 and Chapter 3 of MARSSIM contain

18 additional guidance on the HSA.

19

20 4.1.2 Summary of Survey Types

21 The NRC’s regulations (10 CFR 20.1501(a)) require a licensee to make or cause to be made 

22 surveys that may be necessary for the licensee to comply with the regulations in Part 20, 

23 including the radiological criteria for license termination found in 10 CFR Part 20, Subpart E. 

24 The licensee would demonstrate compliance with this requirement by performing a FSS. The 

25 FSS will demonstrate that the licensee’s site or facility, or both, can meet the radiological criteria 

26 for license termination.

27

28 Other surveys (e.g., scoping surveys, characterization surveys, and remedial action support 

29 surveys) are used to identify areas with residual radioactivity but are typically not used to 

30 demonstrate compliance with the radiological criteria for license termination.

31

32 The NRC endorses the FSS methodology described in MARSSIM. The guidance in this chapter 

33 does not replace MARSSIM, and users of this chapter should be familiar with and use 

34 MARSSIM. Thus, it is intended that licensees will use this chapter and MARSSIM as guidance 

35 for acceptable approaches or methodologies to conduct surveys supporting decommissioning

36 and FSSs, in particular. The following text refers to specific sections of MARSSIM, Revision 1,

37 when applicable.

38

39 The measurement methods applied in assessing radiation and radioactivity levels can vary 

40 according to the objectives of the particular survey. It is expected that different types of surveys 

41 would be conducted during the course of decommissioning work, with each having a different 

42 emphasis while at the same time, sharing common elements. The sections below summarize 

43 six survey types. 
4-3

1 Background Survey—Although not specifically identified as a step in the RSSI process, this 

2 survey constitutes measurements of sites in areas surrounding the facility to establish the 

3 baseline; that is, the normal background levels of radiation and radioactivity. In some situations, 

4 historical measurements may be available from surveys performed before the construction and 

5 operation of a facility. The background survey takes on added importance if one may ultimately 

6 be comparing onsite cleanup units to offsite reference areas. Appendix A of this volume 

7 contains guidance on background surveys.

8

9 Scoping Survey—This survey, performed to augment the HSA, provides sufficient information 

10 (1) to determine if residual radioactivity is present that warrants further evaluation and (2) to

11 make initial estimates of the level of effort required for remediation and to prepare a plan for a

12 more detailed survey, such as a characterization survey. The scoping survey does not require

13 that all radiological parameters be assessed. Sections 2.4 and 5.2 of MARSSIM and

14 Section 4.2 of this volume contain additional guidance on the scoping survey.

15

16 Characterization Survey—This survey determines the type and extent of residual radioactivity 

17 on or in structures and environmental media. The survey should be sufficiently detailed to 

18 provide data for planning decommissioning actions, including remediation techniques, projected 

19 schedules, costs, waste volumes, and health and safety considerations during remediation. 

20 Section 4.2 of this volume contains additional guidance on characterization surveys.

21

22 Remedial Action Support Survey—This survey, which could be repetitive in nature, is conducted 

23 in what is effectively a real-time mode to guide cleanup efforts and ensure the health and safety 

24 of workers and the public. The effectiveness of the remediation efforts can be assessed as they 

25 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 26 Section 4.3 of this 

27 volume contains additional guidance on remedial action support surveys.

28

29 Final Status Survey—This survey demonstrates that residual radiological conditions satisfy the 

30 predetermined criteria for release for unrestricted use or, where appropriate, for use with 

31 designated restrictions. It is this survey that provides data to demonstrate that all radiological 

32 parameters (e.g., total surface activity, removable surface activity, exposure rate, and 

33 radionuclide concentrations in soil and other materials) satisfy the established guidelines and 

34 conditions. Section 4.4 of this volume contains additional guidance on FSSs. 

35

36 Confirmatory Survey—The regulator performs this survey to obtain data to substantiate the 

37 results of the licensee’s FSS. The objective of this type of survey is to verify that 

38 characterization, remediation, and final status actions and documentation, conducted as part of 

39 the RSSI process, are adequate to demonstrate that the site is radiologically acceptable, 

40 relative to applicable criteria. Section 15.4.5 of Volume 1 of this NUREG report contains 

41 additional information on confirmatory surveys.

42

43 These types of surveys are performed at various stages of the decommissioning process. Early 

44 on, where known residual radioactivity exists, the simplest of measurement approaches can be 

45 used to document the need to clean up a specific building surface or parcel of land. In practice, 

46 the simpler methods would generally be applicable to the scoping and remedial action support

47 surveys. The more complex methods, which produce data with higher precision and accuracy, 

48 will be required for background, characterization, final status, and confirmatory surveys. In 

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.
4-4

1 general, wherever measurements are to be performed at or close to background levels, greater 

2 sensitivity in the measurement is required.

3

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

5 It is possible that measurements collected in one survey can be used for another. For instance, 

6 if measurements sufficient in spatial coverage and with adequate detection limits were taken, 

7 the results of the scoping survey in an unaffected area could be used to support the FSS. The 

8 emphasis of the guidance in this volume is on the methods that can be applied to meet the 

9 requirements of the FSS, although they can be applied to other survey work as well. In 

10 performing decommissioning surveys, licensees should be cognizant of the survey 

11 methodologies and their limitations, especially where newer technologies are employed such as 

12 in situ gamma spectroscopy. Some information on the capabilities of this technology to detect 

13 discrete particles in soil can be found in the NRC sponsored study/report by the Oak Ridge 

14 Institute for Science and Education (ORISE) titled “Spatially-Dependent Measurements of 

15 Surface and Near-Surface Radioactive Material Using In situ Gamma Ray Spectrometry 

16 (ISGRS) For Final Status Surveys” (Chapman et al., 2006).

17

18 NRC released the final version of the Multi-Agency Radiological Laboratory Analytical Protocols 

19 manual (MARLAP) in 2004. The MARSSIM and the MARLAP manual are complementary 

20 guidance documents in support of cleanup and decommissioning activities. The MARSSIM

21 document contains guidance on how to plan and carry out a study to demonstrate that a site 

22 meets appropriate release criteria. It describes a methodology for planning, conducting, 

23 evaluating, and documenting environmental radiation surveys conducted to demonstrate 

24 compliance with cleanup criteria. The MARLAP manual provides guidance and a framework for 

25 both project planners and laboratory personnel to ensure that radioanalytical data will meet the 

26 needs and requirements of cleanup and decommissioning activities. 

27

28 The MARLAP manual recommends the use of a directed or systematic planning process. A 

29 directed planning process is an approach for setting well-defined, achievable objectives and 

30 developing a cost effective, technically sound sampling and analysis design that balances the 

31 data user’s tolerance for uncertainty in the decision process with the resources available for 

32 obtaining data to support a decision. For example, the NRC and licensees have determined 

33 that side-by-side surveys (with subsequent PSRs) are more efficient than waiting for a final 

34 sitewide confirmatory survey. The NRC and licensees should plan ahead and coordinate their 

35 schedules to implement efficient side-by-side confirmatory surveys. Appendix D contains more 

36 details on MARLAP and how it can enhance radiation monitoring.

37

38 Appendix D of this volume includes information on survey data quality and reporting, Chapter 5 

39 of MARSSIM provides survey checklists, Appendix E contains information on survey 

40 measurements, and Appendix G has information on special survey issues. 

41

42 4.1.3 Areas of Review

43 The NRC staff should review the results of the radiological characterization survey to determine 

44 whether it contains sufficient information to permit planning for site remediation that will be 

45 effective and will not endanger the remediation workers, to demonstrate that it is unlikely that 

46 significant quantities of residual radioactivity have gone undetected, and to provide information 

47 that will be used to design the FSS.

48

49 The purpose of the NRC staff review is to verify that the FSS design is adequate to demonstrate 

50 compliance with the radiological criteria for license termination.
4-5

1 The FSS review should determine whether the results demonstrate that the site, area, or 

2 building meets the radiological criteria for license termination.

3

4 The staff should note that NRC regulations require that DPs include a description of the planned 

5 final radiological survey. Recognizing the flexible approach discussed in Section 2.2 of this 

6 volume and that the MARSSIM approach allows certain information needed to develop the final 

7 radiological survey to be obtained as part of the remedial activities at the site, a licensee or 

8 responsible party should submit information on facility radiation surveys in one of two ways, as 

9 summarized below. Section 2.2 of this volume provides additional relevant guidance.

10

11 Method 1: 

12 The licensee or responsible party may submit the information contained in Sections 4.1–4.3 

13 of this volume as part of the DP, along with a commitment to use the MARSSIM approach in 

14 developing the final status survey. The licensee or responsible party would then submit the 

15 information discussed in Section 4.4 at the completion of remediation or design 

16 development for the final status survey for the site. The licensee or responsible party will 

17 submit the FSSR (Section 4.5) after performing the final status survey.

18

19 Method 2: 

20 The licensee or responsible party may submit the information contained in Sections 4.1–4.4 

21 of this volume, along with a commitment to calculate the number of sampling points that will 

22 be used in the final status survey, in accordance with the procedure described in MARSSIM. 

23 The licensee or responsible party would then submit the FSSR (Section 4.5) after 

24 performing the final status survey. If this method is used, the licensee or responsible party 

25 should include in the FSSR the information contained in the last three bullets under 

26 “Information to Be Submitted,” in Section 4.4 of this chapter.

27

28 4.1.3.1 Acceptance Review

29 The review should ensure that the licensee’s submittal contains the information summarized 

30 under the “Areas of Review,” as appropriate for the particular submittal. The NRC staff should 

31 ensure that the level of detail appears to be adequate for it to perform a detailed technical 

32 review but should not review the technical adequacy of the information, which it should 

33 determine during the detailed review.

34

35 4.1.3.2 Safety Evaluation

36 The material to be reviewed is both informational in nature and requires specific detailed 

37 technical analysis. The NRC staff should verify that the survey designs and results are 

38 adequate for demonstrating compliance with the radiological criteria for license termination.

39

40 4.1.4 Release Criteria

41 The NRC staff review is to verify that the licensee has provided appropriate release criteria, 

42 referred to as the DCGLs. Generally, the licensee should provide the DCGLW, for the survey 

43 unit average concentrations, and the applicable DCGLEMC (elevated measurement comparison)

44 for small areas of elevated concentrations, for all affected media. 
4-6

1 4.1.4.1 Acceptance Criteria

2 4.1.4.1.1 Regulatory Requirements

3 10 CFR 20.1402, 20.1403, and 20.1404

4

5 4.1.4.1.2 Regulatory Guidance

6 NUREG-1575, “Multi-Agency Radiological Survey and Site Investigation Manual 

7 (MARSSIM)”

8

9 4.1.4.1.3 Information to Be Submitted

10 The licensee should list the DCGL(s) that will be used to design the surveys and to demonstrate 

11 compliance with the radiological criteria for release, including the following:

12

13 • a summary table or list of the DCGLW for each radionuclide and affected medium of

14 concern

15 • the DCGLEMC (or areas factor derived from dose modeling) for each radionuclide and

16 media of concern if Class 1 survey units are present (Appendix A.2 of this volume

17 discusses classification of site areas)

18 • the appropriate DCGLW for the survey method to be used if multiple radionuclides are

19 present

20 The information to be submitted is also included as part of the master DP checklist provided in 

21 this NUREG report (see Section XIV.a from Appendix D of Volume 1).

22

23 4.1.4.2 Evaluation Criteria

24 The NRC staff should verify that, for each radionuclide and affected media of concern, the 

25 licensee has provided a DCGLW and, if Class 1 survey units are present, a table of DCGLEMCs 

26 (or area factors). The NRC staff should verify that the values presented are consistent with the 

27 values developed pursuant to dose modeling, as discussed in Chapter 5 of this volume. If 

28 multiple radionuclides are present, MARSSIM Sections 4.3.2, 4.3.3, and 4.3.4 describe 

29 acceptable methods to determine DCGLs appropriate for the survey technique.

30

31 4.2 Scoping and Characterization Surveys

32 4.2.1 Scoping Surveys

33 Early in the decommissioning process, the licensee identifies the potential residual radioactivity 

34 present at the site, the relative ratios of radionuclides, and the general extent of residual 

35 radioactivity—if any—both in activity levels and affected area or volume. Although the license 

36 and operational history documentation will assist to varying degrees in providing this 

37 information, it will often be necessary to supplement it with actual survey data. Therefore, the 

38 licensee shall perform a scoping survey typically consisting of limited direct measurements 

39 (exposure rates and surface activity levels) and samples (smears, soil, water, and material with 

40 induced activity) obtained (1) from site locations considered to be the most likely to contain 

41 residual activity and (2) from other site locations, including those immediately adjacent to the 
4-7

1 radioactive materials use areas. This survey provides a preliminary assessment of site 

2 conditions, relative to guideline values. The scoping survey forms the basis for initial estimates 

3 of the level of effort required for decommissioning and for planning the characterization survey.

4 Measurements and sampling in known areas of residual radioactivity do not need to be as 

5 comprehensive or be performed to the same sensitivity level as will be required for the 

6 characterization survey or FSS. However, when planning and conducting the scoping survey, 

7 the licensee should remember that some of the data, particularly from locations not affected by 

8 site operations, may be used as final survey results or to supplement the characterization or 

9 final survey results, or both. Similar measuring and sampling techniques as used for those 

10 categories of surveys, therefore, may be warranted.

11

12 Scoping surveys provide site-specific information based on limited measurements. The 

13 following are the primary objectives of a scoping survey:

14

15 • perform a preliminary hazard assessment

16 • support classification of all or part of the site as a Class 3 area

17 • evaluate whether the survey plan can be optimized for use in either the characterization

18 or final stage

19 • perform radiological status surveys

20 • provide data to address the requirements of other applicable regulations

21 • provide input to the characterization survey design, if necessary

22 Scoping surveys are conducted after the HSA is completed and consist of judgment 

23 measurements based on the HSA data. If the results of the HSA indicate that an area is 

24 Class 3 and no residual radioactivity is found, the licensee may classify the area as Class 3 and 

25 conduct a Class 3 FSS. If the scoping survey locates residual radioactivity, the licensee may 

26 consider the area to be Class 1 (or Class 2) for the FSS and typically conduct a characterization 

27 survey, collecting sufficient information to identify situations that require immediate radiological 

28 attention. Licensees should be aware that requirements of other applicable regulations 

29 (e.g., non-radiological constituents) may differ from NRC requirements. Appendix F to 

30 MARSSIM contains a comparison of MARSSIM guidance to other requirements.

31

32 4.2.2 Characterization Surveys

33 After identifying the affected locations, the licensee conducts a characterization survey to more 

34 precisely define the extent and magnitude of residual radioactivity. The survey should be 

35 sufficiently detailed to provide data for planning the remediation effort, including the remediation 

36 techniques, schedules, costs, and waste volumes, as well as necessary health and safety 

37 considerations during remediation. The type of information obtained from a characterization 

38 survey is often limited to that which is necessary to differentiate a surface or area as containing

39 or not containing residual radioactivity. A high degree of accuracy may not be required for such 

40 a decision when the data indicate levels well above the guidelines. On the other hand, when 

41 data are near the guideline values, a higher degree of accuracy is usually necessary to ensure 

42 the appropriate decision about the true radiological conditions. In addition, one category of 

43 radiological data (e.g., soil radionuclide concentration or total surface activity) may be sufficient 

44 to determine if there is residual radioactivity, and other measurements (e.g., exposure rates or 
4-8

1 removable residual radioactivity levels) may, therefore, not need to be performed during 

2 characterization. As the scoping survey example demonstrates, the choice of survey technique

3 should be commensurate with the intended use of the data, including considerations for 

4 possible future use of the results to supplement the FSS data.

5

6 Licensees typically submit site characterization summary information as part of their DP. If 

7 submitted site characterization information is insufficient to reasonably identify the extent and 

8 nature of residual radioactivity, the NRC may decline to accept and review the DP until such 

9 information is provided. The NRC may ask the licensee to submit site characterization plans or 

10 other site characterization information before submitting the DP, or the NRC may elect to meet 

11 with the licensee before or during site characterization work. However, licensees are not 

12 required to submit a separate site characterization or site characterization report unless required 

13 by a license condition. Rather, site characterization information is required as a component of 

14 the DP. Therefore, the NRC staff will only request site characterization and reports separate 

15 from the DP submittal when necessary to provide assurance the extent and nature of residual 

16 radioactivity are reasonably identified.

17

18 The characterization survey is generally the most comprehensive of all the survey types and 

19 generates the most data. This includes preparing a reference grid, systematic as well as 

20 judgment measurements, and surveys of different media to include surface soils and interior 

21 and exterior surfaces of buildings. Additionally, the characterization survey should identify all 

22 activated materials (typically Decommissioning Groups 4–7) and hard-to-detect radionuclides 

23 throughout the site. The decision as to which media will be surveyed is a site-specific decision 

24 addressed throughout the RSSI process (see MARSSIM).

25

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

27

28 Examples include the following:

29

30 • determining the nature and extent of residual radioactivity

31 • evaluating remediation alternatives (e.g., unrestricted use, restricted use, onsite

32 disposal, offsite disposal)

33 • developing input to pathway analysis and dose or risk assessment models for

34 determining site-specific DCGLs in becquerel/kilogram (Bq/kg), picocuries/gram (pCi/g),

becquerel/square meter (Bq/m2 35 ), or disintegrations per minute/100 square centimeters

(dpm/100 cm2 36 ), as applicable

37 • estimating the occupational and public health and safety impacts during

38 decommissioning

39 • evaluating remediation technologies

40 • developing input to the FSS design

41 • complying with requirements of other applicable regulations

42 This volume does not include detailed discussions of characterization survey design for each of 

43 these objectives; the user should consult other references for specific characterization survey 

44 objectives not covered. For example, DOE’s “Decommissioning Handbook,” issued 
4-9

1 March 1994, is a good reference for characterization objectives for evaluating remediation 

2 technologies or unrestricted or restricted use alternatives. Additionally, ANSI N13.59:2008, 

3 “Characterization in Support of Decommissioning Using the Data Quality Objectives Process” is 

4 a useful document which presents characterization strategies based on the DQO process. The 

5 licensee should consult other references (EPA, “Guidance for Conducting Remedial 

6 Investigations and Feasibility Studies Under CERCLA,” issued October 1988; EPA, “Superfund 

7 Removal Procedures,” 1988; “EPA, Federal Radiation Protection Guidance for Exposure of the 

8 General Public,” dated December 23, 1994; NUREG-1501, “Background as a Residual 

9 Radioactivity Criterion for Decommissioning—Draft Report,” issued August 1994) for planning 

10 decommissioning actions (e.g., remediation techniques, projected schedules, costs, and waste 

11 volumes) and health and safety considerations during remediation. Also, the specific modeling 

12 code documentation should determine the types of characterization data needed to support risk 

13 or dose modeling.

14

15 4.2.3 Areas of Review

16 The purpose of the NRC staff review is to verify that the licensee determined the radiological 

17 condition of the property well enough to permit planning for a remediation that will be effective 

18 and will not endanger the remediation workers, to demonstrate that it is unlikely that significant 

19 quantities of residual radioactivity have gone undetected, and to provide sufficient information 

20 for designing the FSS. Note that some licensees have used, or may request authorization to 

21 use, information developed during the characterization survey to support the final radiological 

22 survey.

23

24 Licensees may use characterization survey data to support the final radiological survey, as long 

25 as they can demonstrate that nonimpacted areas at the site have not been adversely affected 

26 by decommissioning operations and that the characterization survey data are of sufficient scope 

27 and detail to meet the “Information to Be Submitted” guidance for a final survey.

28

29 4.2.3.1 Acceptance Criteria

30 4.2.3.1.1 Regulatory Requirements

31 10 CFR 30.36(g)(4)(i), 40.42(g)(4)(i), 70.38(g)(4)(i), and 72.54(g)(1)

32

33 4.2.3.1.2 Regulatory Guidance

34 NUREG-1575, “Multi-Agency Radiological Survey and Site Investigation Manual

35 (MARSSIM)”

36

37 4.2.3.1.3 Information to Be Submitted

38 The information supplied by the licensee should be sufficient to allow the NRC staff to determine 

39 whether the characterization survey design is adequate to assess the radiological status of the 

40 facility. The licensee should describe the radiation characterization survey design and the 

41 results of the survey, including the following:

42

43 • a description and justification of the survey measurements for affected media (for

44 example, building surfaces, building materials (volumetric contamination) contamination,

45 surface soil, subsurface soil, surface water, groundwater, sediments, as appropriate)
4-10

1 • a description of the field instruments and methods that were used for measuring 

2 concentrations and the sensitivities of those instruments and methods

3 • a description of the laboratory instruments and methods that were used for measuring

4 concentrations and the sensitivities of those instruments and methods

5 • the survey results, including tables or charts of the concentrations of residual

6 radioactivity measured

7 • maps or drawings of the site, area, or building showing areas classified as nonimpacted

8 or impacted and visually summarizing residual radioactivity concentrations in impacted

9 areas

10 • a justification for classifying areas as nonimpacted

11 • a discussion of why the licensee considers the characterization survey to be adequate to

12 demonstrate that it is unlikely that significant quantities of residual radioactivity have

13 gone undetected

14 • a discussion of how areas or surfaces in a survey unit were surveyed or why they did not

15 need to be surveyed if considered to be inaccessible or not readily accessible

16 • for sites, areas, or buildings with multiple radionuclides, a discussion justifying the ratios

17 of radionuclides that will be assumed in the FSS or an indication that no fixed ratio

18 exists, and each radionuclide will be measured separately (note that this information

19 may be developed and refined during decommissioning, and licensees may elect to

20 include a plan to develop and justify final radionuclide ratios in the DP)

21 The information to be submitted is also included as part of the DP Checklist provided in this 

22 NUREG report (see Section XIV.b from Appendix D of Volume 1). 

23

24 Licensees should note that, if they elect to dispose of buildings and structures rather than leave 

25 them in place (for unrestricted release), the LTR does not apply to the material moved offsite 

26 from those buildings and structures. Rather, building and structure deconstruction and 

27 dismantlement materials can be released from the site in accordance with existing license 

28 conditions. The data from the characterization survey may be sufficient to demonstrate 

29 compliance with the conditions of the existing license for releasing material from the site. 

30 However, a characterization survey may not be required to demonstrate compliance with the 

31 license condition for releasing material from the site. Section G.2.1 of Appendix G of this 

32 volume provides additional guidance on the offsite disposition of materials.

33

34 4.2.3.2 Evaluation Criteria

35 The NRC staff should verify that the licensee has adequately characterized the site, area, or 

36 building relative to the location and extent of residual radioactivity. An adequate 

37 characterization is one that permits planning for a remediation that will be effective and will not 

38 endanger the remediation workers, demonstrates that it is unlikely that significant quantities of 

39 residual radioactivity have gone undetected, and provides information that will be used to design 

40 the FSS. The extent of detail in the information provided by the licensee should be appropriate 

41 for the specific site, area, or building.
4-11

1 The NRC staff should verify that the survey design and results demonstrate that the licensee or 

2 responsible party has adequately characterized the site. The characterization survey is 

3 adequate if it meets the criteria in the following guidance:

4

5 • Section 5.3 of MARSSIM for the characterization survey (the NRC staff may use the

6 “Example Characterization Survey Checklist” in Section 5.3 of MARSSIM for evaluating

7 the licensee’s submittal)

8 • MARSSIM Chapter 6 and Appendix E for instrument capabilities and sensitivities

9 • MARSSIM Section 4.8.4 for preparing areas for survey

10 4.3 Remedial Action Support Surveys

11 The effectiveness of remediation efforts in reducing residual radioactivity to acceptable levels is 

12 monitored by a remedial action support survey as the remediation effort is in progress. This 

13 type of survey activity guides the cleanup in a real-time mode; it also ensures that the 

14 remediation workers, the public, and the environment are adequately protected against 

15 exposures to radiation and radioactive materials arising from the remediation activities.

16

17 The remedial action support survey typically provides a simple radiological parameter, such as 

18 direct radiation near the surface being remediated. The level of radiation, below which there is 

19 reasonable assurance that the guideline values have been attained, is determined and used for 

20 immediate, infield decisions. Such a survey is intended for expediency and does not provide 

21 thorough or accurate data describing the final radiological status of the site.

22

23 The remedial action support survey is applicable to monitoring surfaces and soils or other bulk 

24 materials only if the radionuclides of concern are detectable by field survey techniques. For 

25 radionuclides and media that cannot be evaluated at guideline values by field procedures, 

26 samples are to be collected and analyzed to evaluate the effectiveness of remediation efforts. 

27 For large projects, the use of mobile field laboratories can provide more timely decisions on the 

28 effectiveness of remedial actions. Examples of situations for which remedial action support 

29 surveys would not be practicable are (1) when soil contains pure alpha or beta emitting 

30 radionuclides and (2) when very low energy beta emitters such as tritium are present on 

31 surfaces.

32

33 Licensees conduct remedial action support surveys to do the following:

34

35 • support remediation activities

36 • determine when a site or survey unit is ready for the FSS

37 • provide updated estimates of site-specific parameters used for planning the FSS

38 The determination that a survey unit is ready for an FSS following remediation is an important 

39 step in the RSSI process. Remedial activities may result in changes to the distribution of 

40 residual radioactivity within the survey unit. Thus, for many survey units, the site-specific 

41 parameters used during FSS planning (e.g., variability in the radionuclide concentration, 

42 probability of small areas of elevated activity) may need to be confirmed or reestablished 

43 following remediation. Obtaining updated values for these critical parameters should be 

44 considered when planning a remedial action support survey. In some cases, where 
4-12

1 concentrations of some radionuclides after remediation may be very low, it may be useful for 

2 licensees to show that certain radionuclides can be considered insignificant; in that case, further 

3 detailed evaluation as part of the FSS may not be necessary (see Section 3.3 of this volume). 

4 However, the dose from the insignificant radionuclides must be accounted for in demonstrating 

5 compliance with the applicable dose criteria. 

6

7 Note that this survey does not provide information that can be used to demonstrate compliance 

8 with the DCGLs and is an interim step in the compliance demonstration process. The FSS will 

9 then survey in detail areas that are likely to satisfy the DCGLs on the basis of the remedial 

10 action support survey. Alternatively, the remedial action support survey can be designed to 

meet the objectives of an FSS.

2 11 DCGLs may be recalculated, based on the results of the 

12 remediation process, although a license amendment may be needed to change (increase) 

13 previously approved DCGLs in a DP or LTP.

14

15 4.3.1 Areas of Review

16 Staff review of the description of the remedial action support surveys should verify that the 

17 licensee has designed these surveys appropriately and to assist in determining when remedial 

18 actions have been successful, so that it may begin the FSS. In addition, information from these

19 surveys may be used to provide the principal estimate of residual radioactivity variability that will 

20 be used to calculate the FSS sample size in a remediated survey unit. 

21

22 4.3.1.1 Acceptance Criteria

23 4.3.1.1.1 Regulatory Requirements

24 10 CFR 30.36(g)(4)(ii), 40.42(g)(4)(ii), and 70.38(g)(4)(ii), 

25

26 4.3.1.1.2 Regulatory Guidance

27 NUREG-1575, “Multi-Agency Radiological Survey and Site Investigation Manual 

28 (MARSSIM)”

29

30 4.3.1.1.3 Information to Be Submitted

31 The NRC staff should verify the licensee’s or responsible party’s description of the 

32 support survey includes the following information:

33 • a description of field screening methods and instrumentation

34 • a demonstration that field screening should be capable of detecting residual radioactivity

35 at the DCGLW

36 The information to be submitted is also included as part of the DP Checklist provided in this 

37 NUREG report (see Section XIV.c from Appendix D of Volume 1).

2 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. 
4-13

1 4.3.1.1.4 Evaluation Criteria

2 The NRC staff should verify that the description of the remedial action support surveys meets 

3 (1) the criteria in MARSSIM Section 5.4 for performing remedial action support surveys and

4 (2) the criteria in the applicable MARSSIM chapters listed in this volume for evaluating technical

5 issues, such as appropriate surveys instruments and survey instrument sensitivity.

6

7 4.4 Final Status Survey Design

8 Professional judgment and biased sampling are important for locating residual radioactivity and 

9 characterizing the extent of residual radioactivity at a site. However, the MARSSIM focus is on 

10 planning the FSS, which uses a systematic approach to sampling. Systematic sampling is 

11 based on rules that try to achieve the representativeness assumed by the statistical tests.

12

13 The licensee uses the FSS to demonstrate compliance with regulations. The primary objectives 

14 of the FSS are to do the following:

15

16 • verify survey unit classification

17 • demonstrate that the potential dose from residual radioactivity is below the release

18 criterion for each survey unit

19 • demonstrate that the potential dose from small areas of elevated activity is below the

20 release criterion for each survey unit

21 Data provided by the FSS can demonstrate that all radiological parameters satisfy the 

22 established guideline values and conditions.

23 4.4.1 Areas of Review

24 The purpose of the NRC staff’s review is to verify that the FSS design is adequate to 

25 demonstrate compliance with the radiological criteria for license termination.

26

27 4.4.1.1 Acceptance Criteria

28 4.4.1.1.1 Regulatory Requirements

29 10 CFR 20.1501(a), 30.36(g)(4)(iv), 40.42(g)(4)(iv), 70.38(g)(4)(iv), and 

30 72.54(g)(4)

31

32 4.4.1.1.2 Regulatory Guidance

33 • Draft NUREG-1505, Rev. 1, “A Nonparametric Statistical Methodology for the Design

34 and Analysis of Final Status Decommissioning Surveys – Interim Draft Report for

35 Comment and Use”

36 • NUREG-1575, “Multi-Agency Radiological Survey and Site Investigation Manual

37 (MARSSIM)”

38 • NUREG-1507, “Minimum Detectable Concentrations with Typical Survey Instruments for

39 Various Contaminants and Field Conditions”
4-14

1 4.4.1.1.3 Information to Be Submitted

2 The information supplied by the licensee should be sufficient to allow the NRC staff to determine 

3 whether the FSS design is adequate to demonstrate compliance with the radiological criteria for 

4 license termination. The information should include all of the following:

5

6 • a brief overview describing the FSS design

7 • a description and map or drawing of affected areas of the site, area, or building classified

8 by residual radioactivity levels (Class 1, 2, or 3) and divided into survey units, with an

9 explanation of the basis for division into survey units (maps should have compass

10 headings indicated)

11 • a description of the background reference areas and materials, if they will be used, and

12 a justification for their selection

13 • a summary of the statistical and other tests that will be used to evaluate the survey

14 results, including the elevated measurement comparison (EMC), if Class 1 survey units

15 are present; a justification for any test methods not included in MARSSIM; and the

16 values for the decision errors (α and β) with a justification for α and β values greater than

17 0.05 for Scenario A and B, respectively

18 • a description of scanning instruments, methods, calibration, operational checks,

19 coverage, and sensitivity for each media and radionuclide

20 • a description of the instruments, calibration, operational checks, sensitivity, and

21 sampling methods for in situ sample measurements, with a demonstration that the

22 instruments and methods have adequate sensitivity (noting that if a licensee uses an

23 advanced technology (e.g., in situ gamma spectroscopy), it must be shown to perform

24 with sensitivities that allow detection of residual radioactivity at an appropriate fraction of

25 the DCGL and corresponding investigation levels (ILs))

26 • a description of the analytical instruments for measuring samples in the laboratory,

27 including the calibration, sensitivity, and methodology for evaluation, with a

28 demonstration that the instruments and methods have adequate sensitivity

29 • a description of how the samples to be analyzed in the laboratory will be collected,

30 controlled, and handled

31 • a description of the FSS ILs and how they were determined

32 The information to be submitted is also included as part of the DP Checklist provided in this 

33 NUREG report (see Section XIV.d from Appendix D of Volume 1). Appendix A provides 

34 additional information about demonstrating the appropriate selection of survey instrumentation.

35

36 4.4.1.2 Evaluation Criteria

37 The NRC staff review should verify that the FSS design is adequate to demonstrate compliance 

38 with the radiological criteria for license termination. The FSS design is adequate if it meets the 

39 criteria in the following guidance:
4-15

1 • Appendix A to this volume, for general guidance on implementing the MARSSIM 

2 approach for conducting FSSs

3 • Appendix B to this volume, for guidance on alternative methods of FSS for simple

4 situations

5 • MARSSIM Sections 4.4 and 4.6 for classifying areas by residual radioactivity levels and

6 dividing areas into survey units of acceptable size

7 • MARSSIM Section 4.5 for methods to select background reference areas and materials

8 • NUREG-1505, Chapter 13, for a method to account for differences in background

9 concentrations between different reference areas

10 • MARSSIM Section 5.5.2 for statistical tests

11 • Appendix A to this volume, Section A.8.2, for decision errors

12 • MARSSIM Sections 6.5.3 and 6.5.4 for selection of acceptable survey instruments,

13 calibration, and operational checkout methods

14 • MARSSIM Section 6.7 for methods to determine measurement sensitivity; NUREG-1507

15 for instrument sensitivity information

16 • MARSSIM Sections 5.5.2.4, 5.5.2.5, 5.5.3, 7.5, and 7.6 for scanning and sampling

17 • MARSSIM Section 7.7 for sample analytical methods (Table 7.2 in Section 7.7 for

18 acceptable analytical procedural references)

19 • MARSSIM Sections 7.5 and 7.6 for methods for sample collection

20 • MARSSIM Section 5.5.2.6 for survey ILs

21 • Appendix G to this volume for surveys for special structural or land situations

22 4.5 Final Status Survey Report

23 To the extent possible, the FSSR should stand on its own with minimal information incorporated 

24 by reference. Although the FSS is discussed as if it were an activity performed at a single stage 

25 of the site investigation process, this does not have to be the case. Data from other surveys 

26 conducted during the RSSI process—such as scoping, characterization, and remedial action 

27 support surveys—can provide valuable information for an FSS, provided the data are of 

28 sufficient quality.

29

30 4.5.1 Areas of Review

31 The purpose of the NRC staff review is to verify that the results of the FSS demonstrate that the 

32 site, area, or building meets the radiological criteria for license termination. For licensees who 

33 have submitted a DP, the FSSR need only include the information described under

34 Section 4.5.1.1 (Acceptance Criteria). A licensee who has not submitted a DP should contact
4-16

1 the NRC staff to ensure its FSSR includes not only the information below but also any other 

2 relevant information the staff needs to carry out its review. 

3

4 4.5.1.1 Acceptance Criteria

5 4.5.1.1.1 Regulatory Requirements

6 10 CFR 20.1402, 20.1403, 20.1501, 30.36(j)(2), 40.42(j)(2), 70.38(j)(2), and 

7 72.54(l)(2)

8

9 4.5.1.1.2 Regulatory Guidance

10 NUREG-1575, “Multi-Agency Radiological Survey and Site Investigation Manual 

11 (MARSSIM)”

12

13 4.5.1.1.3 Information to Be Submitted

14 The information submitted by the licensee should be sufficient to allow the staff to determine 

15 whether the site, area, or building meets the radiological criteria for license termination. The 

16 information should include the following:

17

18 • an overview of the results of the FSS

19 • a summary of the DCGLs for the facility (if DCGLs are used)

20 • a discussion of any changes that were made in the FSS from what was proposed in the

21 DP or other prior submittals

22 • a description of the method by which the number of samples was determined for each

23 survey unit

24 • a summary of the values used to determine the number of samples and a justification for

25 these values

26 • the survey results for each survey unit, including the following:

27 o the number of samples taken for the survey unit

28 o a description of the survey unit, including (1) a map or drawing of the survey unit

29 showing the reference system and random start systematic sample locations for

30 Class 1 and 2 survey units, and random locations shown for Class 3 survey units and

31 reference areas, (2) a discussion of remedial actions and unique features, and

32 (3) areas scanned for Class 2 and 3 survey units

33 o the measured sample concentrations, in units that are comparable to the DCGLs

34 o the statistical evaluation of the measured concentrations

35 o judgmental and miscellaneous sample data sets reported separately from those

36 samples collected for performing the statistical evaluation
4-17

1 o a discussion of anomalous data, including any areas of elevated direct radiation

2 detected during scanning that exceeded the IL or any measurement locations in

3 excess of DCGLW

4 o a statement that a given survey unit satisfied the DCGLW and the EMC if any sample

5 points exceeded the DCGLW

6 • a description of any changes in initial survey unit assumptions relative to the extent of

7 residual radioactivity (e.g., material not accounted for during site characterization)

8 • a description of how ALARA practices were employed to achieve final activity levels

9 The information to be submitted is also included as part of the DP Checklist provided in this 

10 NUREG report (see Section XIV.e from Appendix D of Volume 1).

11

12 4.5.1.2 Review Procedures

13 After review of the FSSR, the NRC reviewer should have reasonable assurance that the FSSR 

14 demonstrates that residual radioactivity at the facility complies with the criteria of 

15 10 CFR Part 20, Subpart E. The following guidance discusses the minimal review that should 

16 be performed and how the reviewer should select survey units for more detailed reviews.

17

18 Section 4.5.1.1.3 describes the minimum information to be submitted in each FSSR. Additional 

19 information about the recommended level of documentation is in Appendix D of this volume. At 

20 individual facilities, the NRC reviewer many need additional information on site-specific issues 

21 and complex technical topics to evaluate the FSSR. In addition, the NRC reviewer may need to 

22 obtain previous NRC-generated reports on the FSS, including but not necessarily limited to 

23 inspections, confirmatory surveys, and any safety evaluation reports that may have addressed 

24 the FSS plan.

25

26 4.5.1.2.1 Minimal Technical Review

27 The NRC reviewer should review all of the following:

28

29 • the results of previously conducted in-process inspections and confirmatory surveys to

30 verify that the licensee has properly implemented the FSSP and associated procedures

31 • the licensee’s QA/QC program, if it has not been previously reviewed

32 • changes made to the DP or LTP, if not previously reviewed, to confirm that the changes

33 are not significant and are technically correct

34 • specific parts of the FSS and supporting data that affect the FSS that were not available

35 when the DP or LTP was approved (data that may include supplemental characterization

36 results, basis for final surrogate ratios for multiple radionuclides, or other data collected

37 to specifically support the FSS)

38 • issues (1) identified by interveners and stakeholders and (2) raised in allegations, to

39 ensure such issues have been satisfactorily resolved
4-18

1 • descriptions of the survey units, to determine if any special survey situations are present 

2 (see Appendix G of this volume for examples) 

3 • results of elevated measurement comparisons, to confirm that small areas of residual

4 radioactivity do not exceed the appropriate limits (e.g., DCGLEMC) and

5 • the results of the appropriate statistical tests (e.g., Wilcoxon Rank Sum (WRS) and sign

6 tests), to confirm that results indicate compliance

7

8 The purpose of the NRC staff review of in-process inspections, confirmatory surveys, and 

9 licensee procedures is to ensure all of the following: 

10

11 • the FSSs were implemented in accordance with the approved FSSP

12 • judgmental survey results are not used in the statistical tests and are evaluated

13 separately against the release criteria, and survey results obtained using random start

14 and systematic sampling are statistically treated separately for the purpose of

15 demonstrating compliance

16 • the QA/QC program was adequate and implemented for the FSS

17 • inadequacies in the FSS design or implementation were corrected (e.g., the licensee

18 improved the overall FSS design and implementation, using information from survey

19 units for which the release criteria were not initially met and resurvey or further

20 remediation was needed)

21 • results of confirmatory surveys, including split samples or independent measurements,

22 are consistent with results of licensee surveys

23 • appropriate instrumentation, with sufficient sensitivities, proper calibrations, and

24 adequately trained users, was used for surveys, scans, and measurements, as

25 described in the FSSP

26 4.5.1.2.2 Detailed Technical Review

27 Along with the minimal review described, the NRC reviewer may perform detailed reviews for a 

28 number of survey units. The number initially chosen for detailed review should use a 

29 risk-informed approach and the results of the minimal review. The reviewer should consider 

30 past inspection history, results of confirmatory surveys, the relative difference between residual 

31 radioactivity concentration and the associated DCGLs, the complexity of the FSSP, and the 

32 radionuclide mix. The detailed review could include confirming the selection process and 

33 location of measurements using survey unit maps or floor plans, checking measurement results 

34 using parameters that are specific to the survey methodology, and re-creating the appropriate 

35 MARSSIM statistical test results.

36

37 4.5.1.2.3 Selecting Survey Units for Detailed Reviews

38 Discriminating factors that may be used to select specific survey units for detailed review are 

39 listed below. A survey unit that is characterized by one or more of these factors should be 

40 considered for potential detailed review. However, the NRC reviewer should focus on survey 
4-19

1 units for which there are risk-significant issues, issues that are prevalent across a large number 

2 of survey units instead of isolated cases, and issues involving an inadequate basis for 

3 conclusions. 

4

5 These factors include any of the following:

6

7 • inconsistencies in defining survey units, including the following:

8 o size different from recommended size

9 o multiple areas now combined as one larger Class 1 survey unit

10 o Class 3 survey units that are bordered by Class 1 units

11 o survey units bordered by PSR areas

12 o gerrymandered survey unit boundaries

13 • application of nonstandard statistical tests (e.g., other than WRS test or Sign test)

14 • significant inconsistencies between the DP/LTP and implemented FSS, including the

15 following examples:

16 o use of surface and detector efficiencies that do not match survey methods, surface

17 features, and instrumentation used

18 o type of survey instrumentation

19 o sample collection method

20 o laboratory analytical methods

21 o any survey unit where the scan coverage is less than 100 percent for Class 1 areas

22 or less than the minimum commitment for Class 2 or 3 areas

23 o number of samples per survey unit

24 • survey units that were remediated

25 • survey units for which confirmatory surveys had results inconsistent with the licensee’s

26 FSS results

27 • any Class 2 survey unit with final measurement results near the DCGLW (e.g., greater

28 than 75 percent) or any Class 3 survey unit with significant residual radioactivity

29 (e.g., concentrations greater than 10–25 percent of the DCGLW)

30 • any survey unit that was downgraded in classification (i.e., from Class 1 to 2, Class 2

31 to 3, or Class 1 to 3, or from impacted to nonimpacted)

32 • units surveyed before resolution of QA/QC concerns
4-20

1 • significance of the variability in concentrations (i.e., heterogeneity) across survey units

2 • inconsistent approach or inadequate basis for determining surrogate radionuclide ratios

3 • significant changes to DP or LTP that affect the FSS or that were not previously

4 reviewed

5 • reclassification schemes not approved by the NRC staff

6 • use of MARSSIM survey methods and statistical tests when hot particles are present

7 • presence of systems and components, buried and embedded piping, or building

8 foundations slated to remain on the site after license termination

9 • survey units that combine, for demonstrating compliance, the results of random start or

10 systematic sampling patterns with biased or judgmental survey results

11 • a survey unit that involves surveying or sampling of media other than building surfaces

12 and surface soils (e.g., groundwater, surface water, sediments, or deep subsurface

13 soils)

14 • survey units with areas that are hard to access or have abnormal geometries

15 • any survey unit that combines survey results with a dose assessment or area factors to

16 demonstrate compliance (mixed approaches are used)

17 • the use of composite sampling to establish compliance with release criteria (i.e.,

18 DCGLs). Additional information on composite sampling can be found in Appendix O of

19 this volume.

20 4.5.1.2.4 Detailed Review Topics

21 The detailed review could include confirming the selection process and location of 

22 measurements using survey unit maps or floor plans, checking measurement results using 

23 parameters that are specific to the survey methodology, and re-creating the appropriate 

24 MARSSIM statistical test results. In performing detailed reviews, reviewers should consider, but 

25 not necessarily be limited to the following questions:

26

27 • Does the FSSR adequately address the issues previously discussed under the selection

28 criteria for detailed reviews, immediately above? For example, if a survey technique was

29 changed from the approved technique, did the FSSR adequately justify the new

30 technique?

31 • Are the probabilities of Type I and Type II errors acceptable?

32 • Does the licensee’s analysis rely on a large number of results expressed at minimum

33 detectable activity or MDC values?

34 • Are all of the static measurement or sampling locations for a survey unit taken from a

35 single random-start sampling set, without substitution (e.g., in cases where additional

36 remediation was performed)?
4-21

1 • Is there a discernible trend in results within and among survey units (e.g., when 

2 comparing survey methods, locations, or media matrices)?

3 • If there are discernible trends in the results, are the statistical tests appropriate?

4 • Are there any outliers in the data? How were they detected and was the disposition of

5 outliers appropriate?

6 • Are there any assumptions about the variability (variance) of the population?

7 • What analytical tools (statistical software packages) were used to analyze the data?

8 • What is the format of the presentation of results? Is it consistent for the survey units

9 reported? For example, are the measurement units consistent with the survey data, the

10 media measured, and the DCGLs?

11 The detailed review of the initially selected survey units may indicate issues that are prevalent 

12 across many units instead of isolated cases. In this case, the reviewer may decide to evaluate 

13 additional survey units in detail.

14

15 4.5.1.3 Evaluation Criteria

16 The NRC review should determine whether the FSSR is adequate to demonstrate compliance 

17 with the radiological criteria for license termination. The reviewer should verify that the 

18 licensee’s FSS results support the conclusion that each survey unit meets the radiological 

19 criteria for license termination. The FSS is adequate if it meets the following criteria: 

20

21 • MARSSIM Section 5.5.2 for the acceptable number of samples

22 • Appendix D of this volume for information on survey data quality and reporting

23 • Section A.10 from Appendix A of this volume for information on determining compliance

24 • MARSSIM Sections 8.3, 8.4, and 8.5 for interpretations of sample results

25 4.6 Issues not Covered in MARSSIM

26 MARSSIM’s main focus is on providing guidance for the design of the FSSs for residual 

27 radioactivity in surface soils and on building surfaces and evaluating the collected data. 

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

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

30 MARSSIM can be applied to surveys performed at vicinity properties—those not under licensee 

31 control—but the decision to apply MARSSIM at vicinity properties is outside the scope of 

32 MARSSIM. MARSSIM does not address other media (e.g., subsurface soil, volumetrically 

33 contaminated building materials, groundwater, surface water, sediments) containing residual 

34 radioactivity. Nor does it address the disposition of components and equipment that are not part 

35 of the survey unit. Some of the reasons for limiting the scope of the guidance to surface soils 

36 and building surfaces include (1) residual radioactivity is limited to these media for many sites 

37 following remediation, (2) since many sites have surface soil and building surfaces as the 

38 leading sources of residual radioactivity, existing computer models used for calculating the 

39 concentrations based on dose or risk generally consider only sources associated with surface 
4-22

1 soils or building surfaces, and (3) MARSSIM was written in support of cleanup rulemaking 

2 efforts for which supporting data are mostly limited to residual radioactivity in surface soils and 

3 on building surfaces. Table 4.2 summarizes the scope of MARSSIM. Although this table was 

4 taken from MARSSIM, it has been modified to be specific to the needs of NRC licensees.

5

6 This volume contains guidance for some topics beyond the scope of MARSSIM. Appendix F 

7 has guidance specific to the characterization of groundwater, surface water, and sediments. 

8 Chapter 5 and Appendices H, I, J, K, L, M, and Q contain other guidance pertaining to dose 

9 modeling. Guidance can be found in Appendix G for special characterization and survey issues 

10 such as subsurface residual radioactivity, embedded piping, sewer systems, and paved areas.
4-23

1 Table 4.2 Scope of MARSSIM 

Within Scope of MARSSIM Beyond Scope of MARSSIM

Guidance MARSSIM provides technical guidance on 

conducting radiation surveys and site investigations.

Regulation MARSSIM does not establish new 

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 with many components—some within the text of 

MARSSIM, others by reference.

Tool Box Many topics are beyond the scope of 

MARSSIM, including public participation programs, 

packaging and transportation of wastes for disposal, 

remediation and stabilization techniques, and training. 

Measurement The guidance given in MARSSIM is 

performance-based and directed toward acquiring sitespecific data.

Procedure The approaches suggested in MARSSIM 

vary depending on the various site data needs—there 

are no set procedures for sample collection, 

measurement techniques, storage, or disposal 

established in MARSSIM.

Modeling The interface between environmental pathway 

modeling and MARSSIM is an important survey design 

consideration addressed in MARSSIM.

Modeling Environmental pathway modeling and 

ecological endpoints in modeling are beyond the 

scope of MARSSIM. 

Soil and Buildings The two main media of interest in 

MARSSIM are surface soil and building surfaces with 

residual radioactivity.

Other Media MARSSIM does not cover other media, 

including subsurface soil, surface or subsurface water, 

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 

the FSS, as this is the deciding factor in judging if the site 

meets the release criterion.

Other Survey Types Although not the focus, 

MARSSIM provides less detailed information on 

scoping, characterization, and remedial action support 

surveys. 

Radiation MARSSIM only considers radiation-derived 

hazards.

Chemicals MARSSIM does not cover any hazards 

posed by chemical contamination.

Remediation Method MARSSIM assists in determining 

when sites are ready for an FSS and provides guidance 

on how to determine if remediation was successful.

Remediation Method MARSSIM does not discuss 

selection and evaluation of remedial alternatives, 

public involvement, legal considerations, or policy 

decisions related to planning.

DQO Process MARSSIM presents a systemized 

approach for designing surveys to collect data needed for 

making decisions such as whether to release a site.

DQO Process MARSSIM does not provide 

prescriptive or default values of DQOs.

DQA MARSSIM provides a set of statistical tests for 

evaluating data and lists alternative tests that may be 

applicable at specific sites.

DQA MARSSIM does not prescribe a statistical test 

for use at all sites.

2

5-1

1 DOSE MODELING EVALUATIONS

2 5.1 Introduction

3 Decommissioning plans typically include estimates of the potential future dose that could be 

4 caused by the residual radioactivity remaining on the site after decommissioning activities are 

5 completed. Calculating potential doses allows both the licensee and regulator to take site6 specific information into account in determining acceptable concentrations of residual 

7 radioactivity at the site using dose models and exposure scenarios that are as realistic as 

8 necessary for the given facility. This section has been written to maintain this flexibility. It 

9 includes the evaluation findings and supporting detailed technical guidance necessary to review 

10 the licensee’s dose and ALARA analyses. The discussion on decommissioning groups in 

11 Volume 1 of this NUREG series provides guidance on information to be submitted.

12

13 Dose modeling information is typically submitted as part of a DP or LTP, although in some 

14 cases it may be submitted separately or as part of a FSSR or other document. This chapter 

15 usually refers to DPs, although other types of reports are implied, if appropriate. The NRC staff 

16 should review all of the dose modeling information submitted by the licensee. For certain cases, 

17 such as screening analyses using default values or a lookup table, most of the review has 

18 already been completed in developing these tools and, therefore, the licensee need only submit 

19 minimal site information and justification in using these models, parameters, and exposure 

20 scenarios. In addition, the NRC staff should review the ALARA analyses, which are based, in 

21 part, on the dose modeling. Two general approaches exist to provide reasonable assurance 

22 that the final concentrations should meet the requirements of Subpart E:

23

24 • The licensee can commit to the exposure scenario(s), model(s), and parameters to be

25 used to evaluate compliance with the dose criterion using the final concentrations. The

26 licensee should project expected final concentrations in the DP to show that there is

27 reasonable assurance that the dose criterion will be met at the time of license

28 termination.

29 • The licensee can derive and commit to meeting nuclide-specific concentration limits

30 equivalent to the dose limit.

31 The “Decommissioning and License Termination Framework” (Figure 1.2), which generalizes 

32 the entire decommissioning process (e.g., Step 7 includes FSS and other requirements related 

33 to license termination), provides licensees with guidance on how to perform iterative dose 

34 analyses. The NRC staff review of dose modeling consists of evaluations in four general areas: 

35

36 • the source and source release assumptions

37 • an exposure scenario considering the site environment

38 • the mathematical model/computational method used

39 • the parameter values and a measure of their uncertainty

40 The actions taken as part of the loop suggested by Steps 8 through 12 of Figure 1.2 can result 

41 in the licensee modifying one or more of the above four parts. Licensees, generally, should not 
5-2

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

2 analyses.

3

4 In some cases, licensees may wish to include the iterative process as part of the DP. This is, 

5 generally, because site characterization is not initially complete enough to provide reasonable 

6 justification for assumptions used in modeling the site. Usually, such incorporation would be in 

7 the form of license conditions that need to be satisfied before license termination can occur. 

8

9 For example, a site may have initial data on groundwater contamination but does not currently 

10 have enough data on hydrological conditions to determine which survey units will be affected by 

11 the plume. Based on the limited data available, the licensee designates an area around the 

12 plume, and all survey units that involve that area will include the dose from the groundwater as 

13 part of the overall dose analyses. For the purposes of this example, the NRC could require the 

14 licensee, through a license condition (or other mechanism), to continue to characterize its 

15 groundwater. If the information confirms that the area affected by the groundwater 

16 contamination is the same or smaller than the assumed area, the licensee can proceed with the 

17 decommissioning process. If the licensee wishes to take advantage of the smaller area, or the 

18 data points to a larger affected area, the licensee may need to submit a license amendment 

19 request to modify the FSSP, the dose modeling, and any other area of the DP affected by the 

20 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 21 

22

23 As described by Figure 1.2 and the preceding example, the areas of dose modeling, site 

24 characterization, and FSSP are interdependent. This is an advantage as judicious use of dose 

25 modeling can help guide site characterization. In addition, both site characterization and FSSP

26 can guide development of reasonable exposure scenarios or modeling approaches. For 

27 example, the appropriate survey techniques may require more advanced modeling in some 

28 areas to make them cost effective to implement. 

29

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

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

32

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.
5-3

1 Table 5.1 Comparison and Description of Exposure Scenario Terms Used in this

2 Guidance 

Exposure Scenario Type Description

Plausible Exposure Scenarios

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 

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.

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

Implausible Exposure Scenarios (No Analysis is Required)

Implausible Land uses that, because of physical or other 

compelling limitations, could not occur

(e.g., residential land use for an underwater plot 

of land).

3

4

5

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 .
5-4

1 5.2 General Approach for Dose Modeling

2 The following section discusses the basic components that are involved in a dose modeling 

3 assessment. It is meant to provide an overview of how the pieces fit together. This section 

4 should give both licensees and reviewers a high-level understanding of the “big picture” review 

5 of the dose assessment evaluation. Section 5.3 provides additional details regarding the 

6 information that should be submitted with a DP or LTP request, Section 5.4 provides additional 

7 detail regarding the acceptance review performed by NRC staff upon receipt of the request,

8 Section 5.5 provides details related to the safety evaluation review, Section 5.6 summarizes 

9 review criteria, and Section 5.7 provides a list of additional guidance documents for use in the 

10 preparation of the request by the licensee as well as in the review by NRC staff’s review.

11

12 Chapter 4 of this volume addresses characterization of the residual radioactivity currently 

13 present at the site and radiological surveys. The information is based on measurements and 

14 knowledge of the site history. To perform dose modeling, the licensee should use the site 

15 information on residual radioactivity expected to be present at the completion of 

16 decommissioning to develop a generalized view of the site’s expected final source 

configuration3 17 . In developing the source term4 model, the licensee should consider the site 

18 measurements, the intended remedial actions, and the needs of both the conceptual model and 

19 the FSS.

20

21 For example, a site may have a large number of both historical and current measurements 

22 characterizing the residual radioactivity over a 10-hectare (25-acre) site. If the site information 

23 shows that residual radioactivity levels do not vary significantly, the licensee may assume that 

24 the source is a uniform layer of residual radioactivity over the site. If the site information shows 

25 significant variability in residual radioactivity concentration, then the licensee may conceptualize 

26 the site as two or more sources of residual radioactivity:

27

28 • one or more “hot spot” sources that represent the area(s) of elevated concentration

29 • a source that represents the larger area of residual radioactivity, which contains residual

30 radioactivity at a lower concentration compared to the elevated area

31 After the source configuration has been determined, the question becomes, “How could humans 

32 be exposed either directly or indirectly to residual radioactivity?” or “What is the appropriate 

33 exposure scenario?” Each exposure scenario should address the following scenario questions:

34

35 • How does the residual radioactivity move through the environment?

36 • 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 RESRADOFFSITE 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/basicref/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”.
5-5

1 • What is/are the likely land use(s) in the future for these areas?

2 • What are the exposure group’s habits that will determine exposure? (What do they eat

3 and where does it come from? How much? Where do they get water and how much?

4 How much time do they spend on various activities?)

5 In most situations, there are numerous possible exposure scenarios in which future human 

6 exposure groups could interact with residual radioactivity. The compliance criteria in 

7 10 CFR Part 20 for decommissioning does not require an investigation of all (or many) possible 

8 exposure scenarios; its focus is on the dose to members of the critical group for the compliance 

9 exposure scenario. The critical group is defined (at 10 CFR 20.1003, “Definitions”) as “the 

10 group of individuals reasonably expected to receive the greatest exposure to residual 

11 radioactivity for any applicable set of circumstances.” The compliance exposure scenario is the 

12 exposure scenario that leads to the largest peak dose to the average member of the critical 

13 group from the mixture of radionuclides. It may be based on a bounding exposure scenario, a 

14 screening exposure scenario, another exposure scenario using conservative assumptions about 

15 land uses or behaviors, or a scenario considering reasonably foreseeable future land uses for 

16 the area.

17

18 If the licensee bases its compliance exposure scenario on reasonably foreseeable land use

19 scenarios which are not clearly bounding, the licensee should also identify less likely but 

20 plausible land use scenarios. These are scenarios that could lead to higher doses compared to 

21 the reasonably foreseeable land use scenario used to demonstrate compliance with the LTR 

22 criteria. The evaluation of less likely but plausible exposure scenarios ensures that, if land uses 

23 other than the reasonably foreseeable land use were to occur in the future, unacceptably high 

24 risks would not result. 

25

26 By combining knowledge about the sources of residual radioactivity and the exposure scenario 

27 questions, the analyst can develop exposure pathways. Exposure pathways are the routes that 

28 residual radioactivity may take in traveling from the source, through the environment, to the 

29 receptor or human. Exposure routes can be fairly simple and direct (e.g., residual radioactivity 

30 in surface soil emits gamma radiation, which results in direct exposure of an individual standing 

31 on the soil), or exposure routes can be fairly complex and indirect (e.g., residual radioactivity in 

32 the surface soil leaches to unsaturated soil layers and is transported to the underlying aquifer 

33 and water from the aquifer is extracted for use as drinking water, which results in exposure to 

34 individuals ingesting groundwater). Exposure pathways typically fall into three principal 

35 categories identified by the manner in which the exposed individual interacts with residual

36 radioactivity present in environmental media: ingestion, inhalation, and external radiation 

37 exposure.

38

39 The exposure pathways for many of the exposure groups can be bounded by a smaller number 

40 of possible exposure groups. For example, at a rural site with surface soil residual radioactivity, 

41 two possible exposure groups are (1) a gardener who grows a small fraction of his or her fruits 

42 and vegetables in the soil and (2) a resident farmer who grows a larger fraction of his or her own 

43 food (i.e., the site supplies not only vegetables but also meat and milk). In this case, the 

44 resident farmer scenario could bound the gardener exposure scenario because the exposure 
5-6

1 pathways and specific parameter values associated with the gardener scenario are bounded by

the considerations for the resident farmer.5 2

3

4 As required by 10 CFR 20.1402, “Radiological criteria for unrestricted use,” expected doses are 

5 evaluated for the average member of the critical group, whose characteristics differ from those 

6 of the maximally exposed individual. This is not a reduction in the level of protection provided to 

7 the public but is an attempt to emphasize the uncertainty and assumptions needed in calculating 

8 potential future doses while limiting boundless speculation on possible future exposure 

9 scenarios. While it is possible to actually identify, with confidence, the most exposed member of 

10 the public in some operational situations (e.g., through monitoring, time-studies, distance from 

11 the facility), identifying the specific individual who should receive the highest dose some time 

12 (up to 1,000 years) in the future is impractical, if not impossible. Speculation on his or her 

13 habits, characteristics, age, or metabolism could be endless. The use of the “average member 

14 of the critical group” acknowledges that any hypothetical “individual” used in the performance 

15 assessment is based, in some manner, on the statistical results from data sets (i.e., the 

16 breathing rate is based on the range of possible breathing rates) gathered from groups of 

17 individuals. While bounding assumptions could be used to select values for each of the 

18 parameters (e.g., the maximum amount of meat, milk, vegetables, possible exposure time), the 

19 result could be an extremely conservative calculation of an unrealistic exposure scenario and 

20 could lead to excessively low allowable residual radioactivity levels. 

21

22 Calculating the dose to the critical group is intended to bound the individual dose to other 

23 possible exposure groups. The critical group is a relatively small group of individuals who, due 

24 to their habits, actions, and characteristics, could receive among the highest potential doses at 

25 some time in the future. By using the hypothetical critical group as the dose receptor, coupled 

26 with prudently conservative models, it is unlikely that any individual would actually receive doses 

27 in excess of those calculated for the average member of the critical group. The description of a 

28 critical group’s habits, actions, and characteristics should be based on credible assumptions, 

29 and the information or data ranges used to support the assumptions should be limited in scope 

30 to reduce the possibility of adding members of less exposed groups to the critical group. An 

31 analysis of the average member of the critical group’s potential exposure should also include, in 

32 most cases, some evaluation of the uncertainty in the parameter values used to represent

33 physical properties of the environment.

34

35 The definitions in 10 CFR Part 20 should be used when calculating dose to demonstrate 

36 compliance with the requirements of Subpart E. The intake-to-dose conversion factors from 

37 Federal Guidance Report No. 11, “Limiting Values of Radionuclide Intake and Air Concentration 

38 and Dose Conversion Factors for Inhalation, Submersion, and Ingestion,” issued September 

39 1988 (EPA 520/1-88-020) (EPA, 1988b), which are based primarily on adults, should be used 

40 when calculating internal exposures. As stated in EPA’s “Federal Radiation Protection 

41 Guidance for Exposure of the General Public,” dated December 23, 1994, implementing age 

42 and sex dependent limits for the general public would be difficult. Not only would the 

43 uncertainty associated with the impact of age and sex be significant, but the detailed 

44 consideration of age and sex is not generally necessary. 

45

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.
5-7

1 Since age-based dose conversion factors are not being used, the same dose conversion factors 

2 are applied to all individuals. Only in rare exposure scenarios will a non-adult individual receive 

3 a higher dose (i.e., take in more radioactive material) than an adult individual in a similar 

4 exposure scenario. One example is the milk pathway: children generally drink more milk 

5 annually than adults. If milk were the only pathway that would expose the individual to a dose, 

6 then the child would have a slightly higher dose than the adult. But in most situations, 

7 especially ones involving multiple pathways, the total intake of the adult is greater than that of a 

8 child. Therefore, for most multiple pathway exposure scenarios, such as screening analyses, 

9 the average member of the critical group should usually be assumed to be an adult with the 

10 proper habits and characteristics of an adult. As the licensee eliminates pathways or modifies 

11 the exposure scenario, the behavior and dietary habits of children may become important. In 

12 such cases, the licensees should contact the NRC staff for guidance.

13

14 By integrating the exposure scenario, source configuration, and knowledge about the applicable 

15 environmental transport routes involved in the exposure pathways, a conceptual model of the 

16 features and processes at the site can be created. The conceptual model is a qualitative 

17 description of the important environmental transport and exposure pathways and their 

18 interrelationships. Abstraction is commonly necessary to translate the concepts of a conceptual 

19 model into mathematical terms. Not only is model abstraction necessary, it can be useful to an 

20 analyst in explaining complex system behavior by reducing the system to its major components, 

21 thereby facilitating communication on the most important components of the system being 

22 simulated. 

23 The extent of the details provided in the conceptual model can impact the extent to which the 

24 features and processes described in the conceptual model can be translated into mathematical

25 terms. In particular, mathematical modeling of groundwater transport can be challenging for a 

26 variety of reasons. The characteristics and features that introduce challenges include, but are 

27 not limited to, the following:

28

29 • limited availability of site-specific data to describe important features, events, and

30 processes for a specific groundwater system

31 • limited availability of site-specific data on the source term

32 • natural heterogeneity of the site and changes in site characteristics over time

33 • complex interactions of the hydrologic, geologic, physical, and chemical processes

34 associated with the system

35 Model abstraction for hydrogeological systems may require the expertise of a qualified 

36 specialist. In some cases, the geology of an area may initially appear to be very complicated; 

37 however, after analysis, it may be determined that most of the geological detail need not be 

38 abstracted. For example, a bedrock aquifer layer may include smaller faults that could 

39 potentially serve as preferred pathways. If characterization studies indicate that the faults are 

40 filled with low permeable minerals and the intermittent nature of the faults prevents flow from 

41 being slowed or diverted, these faults may not need to be included in the conceptual model(s) 

42 and represented in the mathematical model. If, on the other hand, site characterization shows 

43 that the faults transport water more quickly than the rock matrix, flow through the faults should 

44 be represented in the mathematical model (e.g., the hydraulic conductivity of the matrix could be 

45 increased to represent the impact of small faults).
5-8

1 Going from a conceptual model to a mathematical model involves a number of assumptions and 

2 simplifications. For example, one part of a conceptual model of surface soil residual 

3 radioactivity involves the leaching of radionuclides through the soil and into the aquifer. In 

4 reality, the soil between the surface and the aquifer is usually formed by numerous layers of 

5 different types of soils with varying thickness across a site. For the purposes of dose modeling, 

6 the conceptual model is more focused on knowing how much activity is entering (and leaving) 

7 each major environmental compartment (such as the aquifer) than on precisely predicting the 

8 level of activity in the intervening material (e.g., any single soil layer between the surface and 

9 the aquifer). Therefore, the mathematical model may view the intervening soil layers as one 

10 layer or just a few layers, depending on the difficulty of justifying effective parameters that will 

11 mimic the real behavior. Users of off-the-shelf codes should be aware of and consider the 

12 appropriateness of the assumptions made in the computer model they are using.

13 The selection of parameter values (or ranges) for the features, events, and processes of a 

14 specific site depends not only on the site conditions and the exposure scenario(s) but also on 

15 the computer code (or mathematical model) being used. Nearly any data set will need to be 

16 transformed into one appropriate to the situation. This can be as straightforward as generating 

17 a sitewide effective soil density value or as complex as converting resuspension factor data into 

18 resuspension rates. The NRC has already factored these issues into the data used in the 

19 screening analyses, but licensees using site-specific information should justify their values.

20 The conversion of data into parameter values for use in these analyses requires consideration 

21 of uncertainty. In the past, the most common computer codes were deterministic and did not 

22 explicitly consider parameter uncertainty. Although it is not always necessary to use a 

23 probabilistic code to evaluate parameter uncertainty for site-specific analyses, licensees should 

24 provide some discussion of the level of uncertainty in the results and understand the most 

25 important factors influencing site-specific parameter values to ensure that there is sufficient 

26 information to support the results of the analysis. It should be noted that the type of uncertainty 

27 of prime interest to the NRC staff is uncertainty in the physical parameters (e.g., default 

28 behavioral and metabolic parameters found in NUREG/CR-5512, Volume 3, can be used with 

29 limited justification). Appendix Q includes additional guidance on considering uncertainty in 

30 dose modeling.

31 Licensees using probabilistic dose modeling should use the “peak of the mean” dose for 

32 demonstrating compliance with 10 CFR Part 20, Subpart E. Similar to all regulatory guidance, 

33 this NUREG report contains one approach for determining compliance with the regulations 

34 using probabilistic analyses. Use of “mean of the peaks” is also acceptable for demonstrating 

35 compliance. If the “mean of the peaks” dose is significantly higher than the “peak of the mean” 

36 dose, then “risk dilution” may be an issue in the probabilistic model. Appendix Q contains more 

37 information on risk dilution. If the licensee intends to use any probabilistic approach to calculate 

38 DCGLs, it should discuss its planned approach with the NRC staff.

39

40 5.3 Information to be Submitted

41 Dose modeling information is typically submitted as part of a DP or LTP, though in some cases 

42 it may be submitted separately or as part of an FSSR or other document. This information 

43 should include the licensee’s assessment of the potential doses resulting from the residual 

44 radioactivity remaining at the end of the decommissioning process, as well as a comparison of 

45 the potential doses against radiological criteria for license termination found in 10 CFR Part 20, 

46 Subpart E. Information needed for performing reviews of DPs varies depending on the 

47 decommissioning group. For certain cases, such as screening analyses using default values or 
5-9

1 a look-up table, most of the review has already been completed in developing these tools and, 

2 therefore, the licensee need only submit minimal site information and justification in using these 

3 models, parameters, and exposure scenarios. On the other hand, site-specific dose modeling 

4 associated with Decommissioning Groups 4 – 7 may require extensive details regarding the 

5 source term, exposure scenarios, analytical methods, and other details related to the proposed 

6 action. 

7

8 The information to be submitted is also included as part of the DP Checklist provided in this

9 NUREG report (see Checklist Section V.b from Appendix D of Volume 1).

10

11 5.3.1 Decommissioning Groups 1 – 3 (Unrestricted Release Using Screening Criteria)

12 For DPs for sites associated with Decommissioning Groups 1 – 3, the licensee should provide a 

13 description of the exposure scenarios proposed to show compliance with the regulations. The 

14 licensee’s dose modeling for building surfaces or surface soil using the default screening criteria 

15 should include the general conceptual model (for both the radiological contaminants of concern 

16 and the building or outside environment) of the site and a summary of the screening method 

used (i.e., running DandD6 17 or using the lookup tables [see Appendix H]).

18 The information to be submitted is also included as part of the DP Checklist provided in this 

19 NUREG report (see Checklist Section V.a.1. (Building Surface Residual Radioactivity) or V.a.2. 

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

21 5.3.2 Decommissioning Groups 4 – 5 (Unrestricted Release Using Site-Specific 

22 Information)

23 In addition to providing information on the exposure scenarios, the licensee should also provide 

24 information on the parameters used in the site-specific analysis. “Site-specific” is used in a 

25 general sense to describe all dose analyses except those based only on the default screening 

26 tools. This may be as simple as a few parameter changes from the default values in the DandD 

27 computer code to licensees using exposure scenarios, models, and parameter values that are 

28 only applicable at the licensee’s site. The information submitted should include the following:

29

30 • site-specific source term information, including nuclides of interest, configuration of the

31 source, areal variability of the source, release mechanisms, and so forth

32 • a description of the compliance exposure scenario, including a description of the critical

33 group

34 • a description of any other reasonably foreseeable or less likely but plausible exposure

35 scenarios considered

36 • a description of the conceptual model for the specific site, including the source term, the

37 physical features important to modeling the transport pathways, and the critical group

6 Decommissioning codes such as DandD can be obtained from NRC’s RAMP (Radiation Protection Computer 

Code Analysis and Maintenance Program).
5-10

1 • identification, description, and justification of the computer code(s) and/or mathematical 

2 model(s) used (e.g., hand calculations, DandD Version 2 (or later version), RESRAD

3 family of codes)

4 • a description of the parameters and the basis for the parameter values used in the

5 analysis

6 • a discussion about the effect of uncertainty on the results

7 • input and output files or printouts, if a computer program was used

8 The information to be submitted is also included as part of the DP Checklist provided in this 

9 NUREG report (see Checklist Section V.b from Appendix D of Volume 1).

10

11 5.3.3 Decommissioning Group 6 (Restricted Release)

12 The majority of the information needed for Decommissioning Group 6 is the same as what 

13 would be needed for Decommissioning Groups 4 – 5. In addition to the information listed above 

14 for Decommissioning Groups 4 – 5, details related to potential impacts associated with a loss of 

15 institutional controls and an ALARA analyses should be included.

16 The information to be submitted is also included as part of the master DP Checklist provided in 

17 this NUREG report (see Section V.c from Appendix D of Volume 1).

18 5.3.4 Decommissioning Group 7 (Alternate Release Criteria)

19 The same information provided for restricted release should also be provided for release using 

20 alternate criteria (see Section 5.3.3). Additionally, the licensee should provide information on 

21 the basis for alternative criteria, including an ALARA analysis. The types of information 

22 provided are included in NUREG-1757, Volume 1, Chapter 17.8 and Appendix M, as well as 

23 SECY-03-0069 and Regulatory Issue Summary 2004-08.

24

25 The information to be submitted is also included as part of the master DP Checklist provided in 

26 this NUREG report (see Section V.d from Appendix D of Volume 1).

27 5.4 Acceptance Review

28 Upon receipt of a DP, NRC staff should perform a high-level review of the contents of the 

29 submittal in an effort to organize the review and evaluation process. The NRC staff should 

30 organize the review by first looking at the overall scope of the dose modeling provided (possibly 

31 for several decommissioning options or critical groups, or both). This review should determine 

32 which decommissioning group and corresponding review criteria should be used and which 

33 specific dose modeling sections are applicable for the given DP. 

34

35 One acceptable way to organize this initial review is (1) to identify and confirm the principal 

36 sources of residual radioactivity (before and after remediation) and (2) to identify the 

37 decommissioning goal of the DP. Coupling these two sets of information, the NRC staff should 

38 have a good indication of what sections of the guidance apply. For decommissioning goals 

39 involving unrestricted release, the NRC staff should quickly evaluate the appropriate 

40 decommissioning group to which the licensee belongs (i.e., Decommissioning Groups 1-5). 

41 Sections 5.4 and 5.5 are organized by review step with subheadings in each section organized 
5-11

1 by decommissioning group. Therefore, it is important to determine the appropriate 

2 decommissioning group to effectively use the guidance in this chapter. 

3

4 The NRC staff should confirm that conditions at the site are consistent with the approach 

5 chosen by the licensee and the decommissioning group’s requirements (e.g., whether the 

6 conditions of the site are consistent with the modeling assumptions inherent in the screening 

7 analysis approach). A screening approach is generally inappropriate for sites exhibiting any of 

8 the following conditions (excluding those caused by sources of background radiation):

9

10 • subsurface soil residual radioactivity (screening analyses assume only surface soil

11 residual radioactivity)

12 • radionuclide residual radioactivity present in an aquifer

13 • buildings with volumetrically contaminated material

14 • radionuclide concentrations in surface water sediments

15 • sites that have an infiltration rate that is greater than the vertical saturated hydraulic

16 conductivity (i.e., resulting in the water running off the surface rather than only infiltrating

17 into the ground)

18 These conditions are inappropriate because they are inconsistent with the conceptual models 

19 used in developing the screening values. In other words, the conceptual model, parameters, 

20 and exposure scenarios used by the DandD computer code to conduct screening analyses and 

21 develop the screening tables presented in Appendix H, are generally incompatible with such 

22 conditions. Situations do exist where licensees may still use the screening analyses even if one 

23 or more of the conditions listed in the bulleted list above is applicable. For example, by 

24 conservatively assuming buried radioactive material is excavated and spread across the 

25 surface, the screening criteria may be applicable for use at the site. However, if exceptions are 

26 made, it is important to understand the underlying assumptions used to develop the screening 

27 values to ensure that the risk is not underestimated (e.g., if screening values are used for buried 

28 residual radioactivity it is important to consider the thickness of residual radioactivity and other 

29 factors to ensure that the use of surface soil screening values based on 15 cm of residual 

30 radioactivity adequately or conservatively represents the risk from the buried residual 

31 radioactivity). 

32 As part of the acceptance review, NRC staff should review the DP to ensure that the licensee or 

33 responsible party has included the necessary information and determine if the level of detail 

34 appears to be adequate for a detailed technical review. This acceptance review should include 

35 a general review of the DP table of contents and the individual sections of the submittal. NRC 

36 staff should also review the dose modeling portion of the DP but does not need to assess the 

37 technical accuracy or completeness of the information contained therein, which should be 

38 determined during the detailed technical review if the DP is accepted for review. The NRC staff 

39 should also verify that the licensee provided enough information to allow an independent 

40 evaluation of the potential dose resulting from the residual radioactivity after license termination 

41 and reasonable assurance that the decommissioning option will comply with regulations.

42

43 Specific considerations for acceptance reviews vary based on the decommissioning group 

44 associated with the site and the proposed path forward. The following subsections provide 

45 specific details regarding what should be considered based on whether NRC staff will be 
5-12

1 performing a review for a site using screening criteria, a review for sites using site-specific 

2 information and seeking unrestricted release, a review for sites using site-specific information 

3 and seeking restricted release or a review for sites using site-specific information and seeking to 

4 use alternate criteria.

5

6 5.4.1 Decommissioning Groups 1 – 3 (Unrestricted Release Using Screening Criteria)

7 Decommissioning sites included in Decommissioning Groups 1 – 3 can be evaluated using 

8 screening criteria. Decommissioning Group 1 includes sites in which licensed material was 

9 used in a way that precluded the release of radioactivity into the environment and would not be 

10 expected to have contaminated areas above the screening criteria. These sites generally 

11 include licensees that possessed and used sealed sources. These sites do not require a 

12 decommissioning plan and dose modeling is not performed. Decommissioning Groups 2 and 3 

13 include sites where residual radioactive contamination is present on building surfaces and in 

14 soils, and licensees are typically able to demonstrate that their facilities meet the provisions of 

15 10 CFR 20.1402 (“Radiological criteria for unrestricted use”) using a screening approach. 

16 Decommissioning Group 2 sites do not need to submit a DP since specific cleanup activities 

17 and procedures consistent with remediating the facility are already included in their current 

18 license. Decommissioning Group 3 sites require amendments to their license to incorporate 

19 necessary remediation procedures needed to decommission the site. Specific details related to 

20 these groups are included in NUREG-1757, Volume 1.

21

22 When evaluating a DP for a site in Decommissioning Groups 1 – 3, NRC staff should review the 

23 DP to ensure that, at a minimum, it contains the information needed to make conclusions 

24 regarding the DP’s compliance with 10 CFR 20.1402. For residual radioactivity on surface soils, 

25 the residential farmer scenario is considered, while for residual radioactivity on building 

26 surfaces, the building occupancy exposure scenario is considered. For screening analyses, the 

27 licensee should provide information for NRC staff to evaluate the appropriateness of use of a 

28 screening approach (e.g., information on the expected source configuration should be provided 

29 to allow staff to evaluate whether screening assumptions related to the source are met). The 

30 staff should also perform a high-level review of the dose modeling portion of the DP without 

31 assessing the technical accuracy and completeness of the information contained therein.

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

33

34 5.4.2 Decommissioning Groups 4 – 5 (Unrestricted Release Using Site-Specific 

35 Information)

36 Decommissioning Groups 4 – 5 include sites that are being considered for unrestricted release 

37 using site-specific dose analyses. Decommissioning Group 4 includes sites that are not found 

38 to include groundwater contamination while Decommissioning Group 5 sites include 

39 groundwater contamination. Additional details on these decommissioning groups are included 

40 in NUREG-1757, Volume 1.

41 Considerations when assessing the path forward for decommissioning actions with a goal of 

42 terminating the license under the unrestricted release requirements of 10 CFR 20.1402, include 

43 the primary exposure scenarios for the individuals exposed on the site. Exposure scenarios 

44 should be consistent with reasonably foreseeable land uses over the next decades, or use a 

45 bounding exposure scenario, such as a resident farmer. The residential farmer scenario is 

46 typically a bounding exposure scenario for residual radioactivity in the environment because this 

47 group includes a nearly comprehensive number of exposure pathways (footnote 13 explains 

48 how, in certain cases, a residential gardener scenario may lead to higher doses than a 
5-13

1 residential farmer scenario). In addition to pathways that may be limited by land use 

2 assumptions, site conditions, such as soil type or groundwater quality, may remove potential 

3 exposure pathways from consideration with the appropriate level of justification by the licensee. 

4 In rare instances, an exposure scenario involving offsite use of residual radioactivity may be the 

5 critical exposure scenario (e.g., buried contamination that is transported offsite via surface water 

6 where a critical group with unique exposure pathways or subsistence land use could be 

7 exposed). 

8

9 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 term7 10 , exposure 

11 scenario(s), conceptual model(s), numerical analyses (e.g., hand calculations or computer 

12 models), and uncertainty have been included to allow the NRC staff to make conclusions 

13 regarding the ability of the licensee to comply with 10 CFR 20.1402. The NRC staff may also 

14 perform a high-level review of the assumptions regarding the source term, the conceptual model 

15 of the site or building as appropriate, the exposure scenario(s), the mathematical method 

16 employed, and the parameters used in the analysis and their uncertainty.

17 5.4.3 Decommissioning Group 6 (Restricted Release)

18 Decommissioning Group 6 sites are considered for restricted release and therefore conclusions 

19 are made regarding the ability of the site to comply with 10 CFR 20.1403 at the time of license 

20 termination based on the information provided in the DP. In addition to the information listed 

21 above for Decommissioning Groups 4 – 5, the DP should consider, at a minimum, two different 

22 sets of exposure scenarios. One set of exposure scenarios should evaluate the dependence on

23 the proposed institutional controls or restrictions by assuming the institutional controls are 

24 effective. Depending on where the residual radioactivity is and what the proposed restrictions 

25 are, the exposure location(s) for the critical group could be either onsite or offsite. The second 

26 set of exposure scenarios should assume that institutional controls are “no longer in effect” in 

27 accordance with 10 CFR 20.1403(e) (i.e., institutional controls put in place by the licensee have 

28 failed to work properly, or effectively, and that the site will be used without knowledge of the 

29 presence of residual radioactivity). Although various times can be evaluated in sensitivity 

30 analyses, institutional controls should be assumed to be ineffective immediately after license 

31 termination. 

32 Chapter 6 of NUREG-1200, “Standard Review Plan for the Review of a License Application for a 

33 Low-Level Radioactive Waste Disposal Facility,” Revision 3, issued in April 1994, can be used 

34 as a guideline on the development of site-specific acceptance review criteria for restricted 

35 release as applicable (e.g., radioactive release and transport, intruder protection, erosion 

36 protection during the post-closure phase, which may be applicable for restricted release 

37 scenarios involving offsite receptors or use of engineered barriers). Additionally, Appendix J of 

38 this document contains information for use when considering intrusion scenarios for buried 

39 radioactivity. 

40 Similar to the acceptance review for Decommissioning Groups 4 – 5, NRC staff should ensure 

41 that, at a minimum, information on the source term, exposure scenario(s), conceptual model(s), 

42 numerical analyses (e.g., hand calculations or computer models), and uncertainty have been 

43 included. The NRC staff may also perform a high-level review of the assumptions regarding the 

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 RESRADOFFSITE offers several options to define the “source term.” Appendix I, Section 2, provides more information.
5-14

1 source term, the conceptual model of the site or building as appropriate, the exposure 

2 scenario(s), the mathematical method employed, and the parameters used in the analyses and 

3 their uncertainty.

4 5.4.4 Decommissioning Group 7 (Alternate Release Criteria)

5 Decommissioning Group 7 sites are evaluated using a licensee’s proposed alternate criteria. 

6 Section 5.4.3 for Decommissioning Group 6 and Chapter 6 of NUREG-1200, “Standard Review 

7 Plan for the Review of a License Application for a Low-Level Radioactive Waste Disposal 

8 Facility,” Revision 3, issued in April 1994, may be used as guidelines on the development of 

9 site-specific acceptance review criteria for alternate decommissioning criteria, as applicable.

10 When evaluating sites in Decommissioning Group 7 NRC staff should review the dose modeling 

11 information provided in the DP pertaining to the licensee’s proposed alternate criteria. It should 

12 use the findings and conclusions of the review under this section to evaluate the DP’s 

13 compliance with 10 CFR 20.1404, “Alternate Criteria for License Termination.” The staff should 

14 ensure that, at a minimum, information on the source term, exposure scenario(s), conceptual 

15 model(s), numerical analyses, and uncertainty have been included. The NRC staff may also 

16 perform a high-level review of the assumptions regarding the source term, the conceptual model 

17 of the site or building as appropriate, the exposure scenarios, the mathematical method 

18 employed, and the parameters used in the analyses and their uncertainty. The NRC staff 

19 should also review the public health and safety and protection of the environment as the basis 

20 for the alternate criteria.

21 5.5 Safety Evaluation Criteria and Review

22

23 When performing a safety evaluation, NRC staff should review the technical content of the 

24 information provided by the licensee to ensure that the licensee used defensible assumptions 

25 and models to calculate the potential dose to the average member of the critical group. The 

26 staff should also verify that the licensee provided enough information to allow an independent 

27 evaluation of the potential dose resulting from the residual radioactivity after license termination 

28 and to allow NRC staff to decide whether it has reasonable assurance that the decommissioning 

29 option will comply with license termination regulations in 10 CFR Part 20, Subpart E. A general 

30 outline and template for the development of a safety evaluation report is provided in NUREG31 1757, Volume 1, Appendix G.

32

33 Specific considerations for the safety evaluation vary based on the decommissioning group 

34 associated with the site and the proposed path forward. The following subsections provide 

35 specific details regarding what should be considered based on whether NRC staff will be 

36 performing a screening review, a site-specific review resulting in the release of a site for 

37 unrestricted use, a site-specific review resulting in the release of a site for restricted use, or the 

38 release of a site using alternate criteria.

39

40 5.5.1 Screening Safety Review Evaluation Criteria

41 Evaluation considerations will vary depending on whether the submittal uses only default 

42 screening methods and parameters, or if the licensee proposes the use of site-specific 

43 parameter values. When performing a safety evaluation of a licensee’s submittal proposing the 

44 use of a screening criteria NRC staff should determine if the screening criteria were used 

45 correctly by the licensee and whether the calculations provide reasonable assurance that 
5-15

1 potential doses would not exceed dose limits. If site-specific parameters are used, the analysis 

2 is not considered a screening analysis and Sections 5.5.2 through 5.5.4 should be consulted.

3 When licensees use the default screening methods and parameters inherent in the DandD code 

4 by either running the computer code or using lookup tables (see Appendix H), the NRC staff will 

5 have already reviewed and accepted nearly all areas of the analysis in developing the screening 

6 tool and should only need to review the source concentrations and distribution of residual 

7 radioactivity and the overall applicability of using the screening method with the associated 

8 residual radioactivity.

9 Specific considerations associated with the evaluation of surface soils and building surfaces, 

10 discussed below, also need to be reviewed. If the licensee did not directly calculate the dose 

11 from residual radioactivity but instead derived, or proposed to use, lookup tables to derive

12 DCGLs, the licensee should provide the basis or some discussion for their use. In cases where 

13 sufficient support is provided for a proposed approach, NRC staff would only need to review the 

14 information on the configuration of the residual radioactivity and the appropriate screening 

15 criteria section, below. For licensees who plan to use screening criteria, residual radioactivity 

16 should be reasonably represented by a homogeneous source (i.e., the source should not be 

17 overly heterogeneous). In general, there should be no elevated areas of residual radioactivity 

18 (e.g., “hot spots”) above the screening values. DandD is limited in its capability to consider the 

19 sensitivity of dose with respect to the area of residual radioactivity. However, if one or more 

20 elevated areas or “hot spots” of residual radioactivity are expected to be present at the site 

21 following decommissioning, these areas will also need to be considered in demonstrating 

22 compliance with radiological criteria for license termination. Appendix I, Sections I.2.3 and 

23 I.3.3.3, contain additional guidance on considering elevated areas of residual radioactivity. If

24 DCGLs are developed for the site, it is possible to consider elevated areas of residual

25 radioactivity through use of DCGLEMCs and address these areas during the FSS. Because it is

26 more difficult to develop DCGLEMCs when screening approaches are used, it is important for the

27 licensee to engage NRC staff as early as possible on acceptable approaches for deriving these

28 DCGLs or considering elevated areas.

29

30 When reviewing a screening analysis for building surfaces and surface soils NRC staff should 

31 use the following criteria:

32 • Source Configuration

33 The NRC staff should confirm that the actual measurements, facility history, and planned 

34 remedial action(s) support the source configuration used in the modeling by reviewing 

35 the portions of the DP on facility history, radiological status, and planned remedial 

36 action(s). 

37 When evaluating building surfaces the NRC reviewer should verify both the areal extent 

38 of residual radioactivity and the depth of penetration of the residual radioactivity into the 

39 building surfaces. The NRC reviewer should also determine if the physical configuration 

40 of the residual radioactivity can adequately be assumed to be a thin layer of residual 

41 radioactivity on the building surfaces. 

42 Similarly, when evaluating surface soils NRC staff should review both the areal extent of 

43 residual radioactivity and the depth of penetration of the residual radioactivity into the 

44 soil. The reviewer should determine if the physical configuration of the residual 
5-16

1 radioactivity can adequately be assumed to be a layer of surface soil containing residual 

2 radioactivity (i.e., no subsurface radioactivity). 

3 If, during the review, it is determined that the residual radioactivity is not limited to the 

4 building surfaces or the surface soil then use of the default screening criteria is not 

5 warranted without additional justification. The NRC reviewer should reclassify the 

6 licensee as Group 4 and evaluate the modeling using Section 5.5.2.

7 • Residual Radioactivity Spatial Variability

8 The NRC staff should review the licensee’s information on conditions before and those

9 projected after the decommissioning alternative is complete. Based on this information,

10 the NRC should determine whether it is appropriate to assume homogeneity (1) for the

11 whole facility or (2) for subsections of the facility when evaluating building surfaces.

12 Similarly, for surface soils, NRC staff should review the licensee’s information to

13 determine whether it is appropriate to assume homogeneity (1) for the entire affected

14 area or (2) for major subsections of the site. The NRC staff should then review the

15 adequacy of the licensee’s determination of a representative value (or range of values)

16 for the residual radioactivity concentration representing the source(s). For elevated

17 areas, the reviewer could use the general concepts related to DCGLEMCs and the more

18 detailed guidance on considering elevated areas of residual radioactivity discussed in

19 Sections I.2.3 and I.3.3.3 of Appendix I.

20 • Conceptual Models

21 A detailed review of the conceptual model is not necessary as the NRC staff addressed

22 these topics when it established the default screening methods. However, the reviewer

23 should verify that the site and DandD’s conceptual models are compatible. Situations

24 that would not allow the use of the DandD code as a screening tool would include those

25 where the source is not predominantly on building surfaces (i.e., volumetric source) or

26 use of the building could lead to higher predicted doses compared to the building

27 occupancy exposure scenario. A list of screening values for beta and gamma emitters

28 can be found in Appendix H, Table H.1.

29 In the case of surface soil evaluations, situations that would not allow use of the DandD

30 code as a screening tool would include those where the source is not predominantly

31 present in the surface soil, residual radioactivity is in the aquifer, or sites with infiltration

32 rates higher than the vertical saturated hydraulic conductivity (i.e., resulting in surface

33 runoff or a bathtub effect). Alternatively, additional information could be provided to

34 support a determination that the results tend to overestimate the dose. A list of

35 screening values can be found in Appendix H, Table H.2.

36 • Execution of the DandD Computer Code Dose Calculations
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1 If the licensee has used the DandD computer code to calculate the dose based on either 

2 current concentrations or projected final concentrations, the NRC staff should verify the 

3 following items, as applicable: 

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

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

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

7 modified.

8 o The licensee has used the 90th percentile (or higher percentile) of the dose

9 distribution to compare with the dose limit.

10 o If the appropriate annual peak dose is greater than 0.025 mSv (2.5 mrem), the

11 removable fraction of the residual radioactivity is 10 percent or less at the time of

12 license termination, or the removable fraction has been adjusted, as explained in

13 footnote a in Table H.1. Note: This item only applies to residual radioactivity on

14 building surfaces.

15

16 • DCGLs from the DandD Code or Lookup Tables

17

18 The licensee may use either the DandD computer code or the published lookup tables

19 for building surfaces in Appendix H, Table H.1, or for surface soils in Appendix H, Table 

20 H.2, to establish radionuclide-specific DCGLs equivalent to 0.25 mSv/y (25 mrem/y). If

21 the licensee proposes to use radionuclide-specific DCGLs, the NRC staff should verify

22 that the following conditions are true, as applicable:

23

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

25 (2) If more than one radionuclide is involved, there is reasonable assurance that the sum

26 of fractions (concentrations divided by DCGLs) (see Section 2.7) is no greater

27 than 1.

28 (3) For building surfaces, if the residual radioactivity is greater than 10 percent of the

29 respective screening DCGLs (Table H.1 from Appendix H of this volume), the

30 removable fraction is 10 percent or less at license termination, or the removable

fraction has been adjusted, as explained in footnote a in Table H.1.8 31 Note: This item

32 only applies to residual radioactivity on building surfaces.

33 If the licensee has used the DandD computer code to calculate the radionuclide-specific 

34 DCGLs, the NRC staff should also verify that the following two conditions are true:

35 (1) The output reports verify that no parameters (other than entering unit concentrations)

36 were modified.

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).
5-18

1 (2) The licensee has used the 90th percentile (or higher percentile) of the dose 

2 distribution to derive the DCGLs.

3 • Compliance with Regulatory Criteria

4 The licensee’s projections of compliance with regulatory criteria are acceptable provided

5 that the NRC staff has reasonable assurance that at least one of the following is true:

6 (1) The final concentrations result in a peak annual dose of less than 0.25 mSv

7 (25 mrem) and the licensee has committed to calculating the annual dose using a

8 screening analysis at license termination.

9 (2) The planned DCGLs are equal to or less than those provided by the screening

10 criteria, and the licensee has committed to ensuring the sum of fractions is no

11 greater than 1, if applicable.

12 5.5.2 Evaluation Criteria for Decommissioning Groups 4 – 5 (Unrestricted Release 

13 Using Site-Specific Information)

14 The NRC staff should determine the acceptability of the licensee’s projections of radiological 

15 impacts from residual radioactivity on the average member of the critical group during the 

16 compliance period. The information in the DP is acceptable if it is sufficient to ensure a 

17 defensible assessment of the possible future impacts from the residual radioactivity. The 

18 licensee’s assessment can be either realistic or prudently conservative. The information should 

19 allow an independent NRC staff evaluation of the assumptions used (e.g., source configuration, 

20 applicable transport pathways) and possible doses to the average member of the critical group.

21

22 The NRC staff should review the following information, as necessary, for each dose assessment 

23 of residual radioactivity that the licensee has submitted for the various decommissioning 

24 options. 

25 • Source Configuration and Release

26 The NRC staff should review the licensee’s dose modeling source term assumptions and

27 compare them with the current site information and planned remedial activities. The

28 model should be an appropriate abstraction (or simplification) of this information. Four

29 key areas of review for the source term assumptions are (1) source configuration,

30 (2) residual radioactivity spatial variability, (3) release mechanisms, and (4) chemical

31 form(s). Section I.2 from Appendix I of this volume provides additional guidance.

32 (1) Source Configuration

33 The NRC staff should confirm that the actual measurements, facility history, and

34 planned remedial action(s) support the source configuration used in the modeling by

35 reviewing the information in the facility history, radiological status, and planned

36 remedial action(s) portions of the DP. The review should include both the areal

37 extent of residual radioactivity and the depth (for soil or buildings) or volume (for

38 groundwater or buried material) of the residual radioactivity. The NRC reviewer

39 should also determine if the information provided supports the configuration

40 assumptions used in the exposure scenario(s) as well as the computer and

41 mathematical model(s) (e.g., a thin layer of residual radioactivity on the building

42 surfaces).
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1 (2) Residual Radioactivity Spatial Variability

2 The NRC staff should review the licensee’s residual radioactivity concentration

3 values for conditions both before and projected after the decommissioning alternative

4 is complete. For this subsection, NRC staff should review the spatial extent and the

5 degree of heterogeneity in the values. Based on this information, the NRC staff

6 should determine whether it is reasonable to assume homogeneity for each source

7 for either (1) the whole site or (2) specific subsections of the site. The staff should

8 then review the adequacy of the licensee’s determination of a representative value

9 (or range of values) for the residual radioactivity concentration in the dose model.

10 For elevated areas, the reviewer could use the general concepts related to

11 DCGLEMCs and more detailed guidance on considering the elevated areas of residual

12 radioactivity discussed in Sections I.2.3 and I.3.3.3 from Appendix I of this volume.

13 If the licensee used dose modeling to develop DCGLs instead of estimating final

14 concentrations and if spatial variability is a concern, then the licensee should develop

15 DCGLEMCs and provide this information in the DP or FSS. The NRC staff should

16 verify that spatial variability is adequately considered in DCGL development and

17 cleanup criteria are compatible with the assumptions made for dose modeling.

18 (3) Release Mechanisms

19 The NRC staff should review the licensee’s assumptions on mechanisms controlling

20 release of radioactivity from the source. Commonly used decommissioning dose

21 modeling codes such as DandD and the RESRAD family of codes contain built-in

22 release models. RESRAD-OFFSITE offers additional source release models that

23 are not available in DandD and RESRAD-ONSITE. For additional information, see

24 Section I.2.

25 (4) Chemical Form

26 The NRC staff should review the licensee’s assumptions about the adequacy of the

27 chemical form of the residual radioactivity and should determine whether the

28 licensee has considered possible chemical changes that may occur during the time

29 period of interest. Without any justification of possible chemical forms, the analysis

30 should use the bounding chemical form(s) (e.g., the chemical form(s) that give the

31 individual the highest dose per unit intake, as described in Federal Guidance Report

32 Number 11 (EPA, 1988b)). The licensee should provide an acceptable rationale for

33 other assumptions. Some acceptable rationales for using other chemical forms are

34 (1) chemical forms that would degrade quickly in the environment (e.g., uranium

35 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 36 ).

37 • Critical Groups, Exposure Scenarios, Pathways, Identification, and Selection

38 In its review, the NRC staff should confirm that the licensee has identified and quantified 

39 the most significant exposure scenarios based on available site- or facility-specific 

40 information, as well as the basis and justification for the licensee’s selected critical 

41 group. For exposure scenarios in which possible environmental pathways have been 

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.
5-20

1 modified or eliminated, the NRC staff should review the justifications provided by the 

2 licensee. Section I.3 has additional guidance on these subjects.

3 (1) Exposure Scenario Identification

4 The compliance exposure scenario is based on the location and type of source

5 (e.g., contaminated walls), the reasonably foreseeable land use, the general

6 characteristics and habits of the critical group (e.g., an adult light-industry worker),

7 and the possible pathways that describe how the residual radioactivity would incur

8 dose in humans. The licensee should provide justification for the exposure

9 scenario(s) evaluated.

10 The licensee should justify the possible land use(s) the site might experience in the

11 future and create exposure scenarios consistent with these uses. The licensee

12 should also justify its selection of a compliance exposure scenario from the possible

13 exposure scenarios derived from the current and projected land uses. The

14 compliance exposure scenario should result in the greatest exposure to the average

15 member of the critical group for all exposure scenarios given the mixture of

16 radionuclides. A licensee may choose to make a bounding assumption for land use

17 to derive the exposure scenario (e.g., assuming a rural land use for an urban

18 location) or base the exposure scenario on the reasonably foreseeable land use that

19 results in the highest dose. The other, less likely but plausible exposure scenarios

20 are considered to better risk inform the decision.

21 If the compliance exposure scenario is based on the reasonably foreseeable land

22 use, the licensee should justify the exposure scenario based on discussions with

23 land planners, meetings with local stakeholders, trending analyses of land use for the

24 region, or comparisons with land use in similar alternative locations. The time period

25 of interest for possible land use changes is within 100 years, depending on the rate

26 of change in the region and the peak exposure time. Note that the 100-year

27 timeframe described here is only for estimating future land uses; the licensee must

28 evaluate doses that could occur over the 1,000-year time period specified in the LTR.

29 The licensee should also identify what land uses are less likely but plausible and

30 evaluate exposure scenarios consistent with these less likely but plausible land uses.

31 If use of reasonably foreseeable land use exposure scenarios results in eliminating a

32 significant number of exposure pathways, the licensee may need to evaluate offsite

33 exposure scenarios to ensure they do not result in greater exposures, when

34 demonstrating compliance with the radiological criteria for license termination.

35 The licensee should provide a quantitative analysis or a qualitative argument

36 discounting the need to analyze all exposure scenarios generated from the

37 reasonably foreseeable land uses. The level of detail can vary between exposure

38 scenarios, and the licensee is expected to use simple analyses to limit the number of

39 detailed exposure scenarios. The licensee may use screening or generic analyses

40 to assist in determining the critical exposure scenario for compliance. With a mixture

41 of radionuclides, more than one compliance exposure scenario may be needed. The

42 peak dose from the exposure scenario(s) with the highest dose should be used to

43 demonstrate compliance.

44 Similarly, the licensee may provide either a quantitative analysis or a qualitative

45 argument discounting the need to analyze all exposure scenarios considering less

46 likely but plausible land uses. The staff will use the results of these analyses to

47 evaluate the degree of sensitivity of dose to overall exposure scenario assumptions

48 (and the associated parameter assumptions). The reviewer will consider both the
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1 magnitude and time of the peak dose from these exposure scenarios. If peak doses

2 from the less likely but plausible land use exposure scenarios are significantly above 

3 the dose standard, the licensee would need to provide greater assurance that the 

4 exposure scenario is less likely to occur, especially during the period of unacceptably

5 high dose.

6 The screening exposure scenarios for building surface residual radioactivity and soil 

7 residual radioactivity are described in NUREG-1549 and NUREG/CR-5512, 

8 Volumes 1, 2, and 3. Dose evaluations that use these exposure scenarios (i.e., the 

9 licensee changes parameter values or mathematical method but does not change 

10 the general exposure scenario) are acceptable if the exposure scenario is 

11 appropriate for the situation. In DPs where the licensee eliminates certain pathways, 

12 with justification, but still maintains the same general exposure scenario category, 

13 the NRC staff should find the exposure scenario identification to be acceptable. For 

14 example, a licensee may eliminate the use of groundwater because the near surface 

15 aquifer has total dissolved solids of 30,000 milligrams per liter (mg/L). The licensee 

16 still evaluates the impacts from crops grown in the residual radioactivity, but irrigation 

17 is provided by a noncontaminated source and therefore, the screening exposure 

18 scenario of a residential farmer is maintained.

19 (2) Critical Group Determination

20 In general, critical groups exposed to multiple exposure pathways result in higher

21 doses than groups that have more limited interaction with residual radioactivity.

22 NUREG-1549 and the NUREG/CR-5512 series detail the critical group assumptions

23 for the screening exposure scenarios. In DPs where the licensee has used the

24 screening exposure scenarios, the reviewer should verify that the proposed critical

25 group satisfies the assumptions listed in NUREG-1549 and the NUREG/CR-5512

26 series.

27 The licensee should provide either qualitative or quantitative justification that the

28 critical group is the highest exposed group for the assumed land use(s). The

29 selection of the critical group may be dependent on the assumption of the relative

30 mixture of radionuclides and sources of residual radioactivity present at the site. The

31 licensee should justify its compliance approach in these cases, as well as for the

32 critical exposure scenario for less likely but plausible exposure scenarios.

33 (3) Exposure Pathways

34 The DP should describe the exposure pathways to which the critical group is

35 exposed, except for cases where the licensee is using the screening exposure

36 scenarios and critical groups without modification. If the licensee has chosen to

37 modify the screening exposure scenario, the changes should be justified. In general,

38 the justification should be based on physical limitations or situations that would not

39 allow individuals to be exposed as described in the exposure scenario. The

40 exposure pathways should therefore be consistent with the land use assumptions,

41 exposure group behavior, and physical site conditions.

42 For example, acceptable justifications for removing the groundwater pathway include

43 the following: (1) the near surface groundwater is neither potable nor allowed to be

44 used for irrigation, (2) the aquifer volume is insufficient to provide the necessary

45 yields, and (3) there are current (and informed consideration of future) land use

46 patterns that would support elimination of the groundwater pathway (e.g., only short47 lived radionuclides are present at the site, which is currently located in an industrial
5-22

1 section of an urban area with restrictions on groundwater use). Justification of water 

2 quality and quantity of the saturated zone should be based on the classification 

3 systems used by EPA or the State, as appropriate. In cases where the aquifer is 

4 classified as not being a source of drinking water but is considered adequate for 

5 stock watering and irrigation but not a viable source of drinking water, the licensee 

6 can eliminate (i.e., does not need to consider) the drinking water pathway (and the 

7 fish pathway—depending on the exposure scenario). The licensee would still 

8 maintain the irrigation and meat and milk pathways, if appropriate, for the land use 

9 assumptions. 

10 Another example would be a rural site with a relatively small discrete outdoor area of 

11 residual radioactivity (compared to the area assumed in the default exposure 

12 scenarios). In this situation, it may be appropriate, based on the area of residual 

13 radioactivity, that gardening of some vegetables and fruits would still be an 

14 assumption, but the area would not be large enough to allow one to grow grain or 

15 raise animals for meat or milk.

16 • Conceptual Models

17 The NRC staff should review the adequacy of the conceptual model(s) used by the

18 licensee.

19 The conceptual model should qualitatively describe the following:

20 o the relative location and activities of the critical group

21 o both the hydrologic and environmental transport processes important at the site

22 o the dimensions, location, and spatial variability of the source represented in the

23 model

24 o the major assumptions made by the licensee in developing the conceptual model

25 The NRC reviewer should verify that the licensee adequately addressed the site 

26 conditions in the conceptual model and simplifying assumptions. Section I.4 has 

27 additional guidance on these subjects. 

28 • Calculations and Input Parameters

29 In its review, the NRC staff will confirm that the licensee has used a mathematical model

30 that is an adequate representation of the proposed conceptual model and the exposure

31 scenario. Section I.5 contains additional guidance on this subject.

32 (1) Execution of DandD Computer Code

33 If the licensee uses the DandD computer code in its analysis, the NRC staff should

34 verify the following points:

35 o The residual radioactivity is limited to the surface (either building or near surface

36 soil, as appropriate).

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

38 o The site conceptual model is adequately represented by DandD’s inherent

39 conceptual model.
5-23

1 o For residual radioactivity present on building surfaces, the licensee has modified 

2 the resuspension factor, as necessary, to account for the removable fraction 

3 expected to be present at the time of decommissioning. The default removable 

4 fraction assumed in DandD is 10 percent. If the removable fraction is expected 

5 to be greater than 10 percent, the licensee should account for higher removable 

fractions that might increase the resuspension factor.10 6 If site conditions are

7 consistent with assumptions in NUREG-1720, “Re-evaluation of the Indoor 

8 Resuspension Factor for the Screening Analysis of the Building Occupancy 

9 Scenario for NRC’s License Termination Rule,” with respect to activities and 

10 exposure scenarios, ventilation conditions, and low removable fractions at the 

11 time of decommissioning, the NUREG-1720 recommended resuspension factor 

12 value or parameter distribution may be used with minimal justification.

13 o For sites eliminating pathways, the licensee has used the appropriate parameters

14 in the DandD code as “switches” to turn off the pathways without unintentionally

15 removing others. For example, to remove the groundwater pathways, the

16 licensee should set the drinking water rate, irrigation rate, and pond volume

17 parameter values to zero.

18 o The licensee has provided adequate support for site-specific parameter values or

19 ranges, has not defaulted to default values, and has appropriately considered

20 parameter correlations.

21 o For modifications of behavioral parameters, the changes should be based on

22 acceptable changes in the critical group, and the mean values of the behavioral

23 parameters should be used, although use of the ranges is also acceptable.

24 o If the licensee has randomly sampled the parameter ranges in DandD, it has

25 used the “peak of the mean” or “mean of the peaks” dose to calculate the dose or

26 derive the DCGLs.

27 (2) Other Mathematical Methods

28 If the licensee uses other mathematical methods or codes, the NRC reviewer should

29 verify the following:

30 o The mathematical conceptual model is compatible with the site’s conceptual

31 model (e.g., RESRAD-ONSITE would not be an acceptable mathematical

32 method for sites with building surface residual radioactivity).

33 o For each parameter or parameter set, the licensee has adequately justified the

34 parameter value or range. For modifications of behavioral parameters, the

35 licensee should base the changes on acceptable changes in the critical group

36 and use the mean value (or full range) of the behavior.

37 o If the inhalation dose can be significant (e.g., due to the presence of

38 alpha-emitting radionuclides such as uranium or thorium), the NRC staff should

39 review the resuspension factor or rate and the assumptions about the degree of

40 removable residual radioactivity.

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.
5-24

1 o If the licensee is performing a probabilistic analysis, it has used the “peak of the 

2 mean” or “mean of the peaks” dose to either calculate the dose or derive the 

3 DCGLs.

4 • Uncertainty Analysis

5 The NRC staff should review the licensee’s discussion of the uncertainty resulting from 

6 the physical parameter values used in the analysis. The review should focus on the 

7 uncertainty analysis for the critical pathways or parameters. Reviewers should expect 

8 that the degree of uncertainty analysis depend on the level of complexity of the modeling 

9 (e.g., generally qualitative discussions for simple modeling to quantitative analyses for 

10 more complex sites). The overall acceptability of the uncertainty analysis should be 

11 evaluated on a case-by-case basis. Section I.7 of Appendix I and Appendix Q contain 

12 additional guidance on these subjects.

13 If the licensee evaluated exposure scenarios based on reasonably foreseeable land 

14 uses, it should provide either a quantitative analysis of or a qualitative argument 

15 discounting the need to analyze all exposure scenarios generated from the less likely but 

16 plausible land uses. The staff will use the results of these analyses to evaluate the 

17 degree of sensitivity of the dose to overall exposure scenario assumptions (and the 

18 associated parameter assumptions). The reviewer will consider both the magnitude and 

19 time of the peak dose from these exposure scenarios. If peak doses from the less likely 

20 but plausible land use exposure scenarios are significantly above the dose standards,

21 the licensee would need to provide greater assurance that the exposure scenarios are

22 unlikely to occur, especially during the period of unacceptably high dose.

23 • Compliance with Regulatory Criteria

24 The licensee’s projections of compliance with regulatory criteria are acceptable, provided 

25 that the NRC staff has reasonable assurance of the following:

26 (1) The licensee has adequately characterized the source and applied a technically

27 defensible source term.

28 (2) The licensee has analyzed the appropriate exposure scenario(s) and found that the

29 exposure group(s) adequately represents a critical group.

30 (3) The mathematical method and parameters used are appropriate for the exposure

31 scenario and parameter uncertainty has been adequately addressed.

32 (4) For deterministic analyses, the peak annual dose to the average member of the

33 critical group for the appropriate exposure scenario(s) for the option is less than (or

34 equal to) 0.25 mSv (25 mrem) or was used to calculate DCGLW.

35 (5) For probabilistic analyses, the “peak of the mean” or “mean of the peaks” dose to the

36 average member of the critical group for the appropriate exposure scenario(s) for the

37 option is less than (or equal to) 0.25 mSv (25 mrem) or was used to calculate

38 DCGLW.

39

40 (6) Either one of the following:

41 o The licensee has committed to using a specific exposure scenario, model, and

42 set of parameters with the final survey results to show final compliance with the

43 dose limit.
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1 o The licensee has committed to using radionuclide-specific DCGLs and will

2 ensure that the total dose from all radionuclides will meet the requirements of

3 Subpart E by using the sum of fractions.

4

5 5.5.3 Safety Evaluation Criteria for Decommissioning Group 6 (Restricted Release)

6 As discussed previously, sites in Decommissioning Group 6 are being considered for restricted 

7 release using site-specific dose analyses. Specific details describing the types of sites that can 

8 be considered for Decommissioning Group 6 are included in NUREG-1757, Volume 1.

9 The majority of the criteria used by the NRC staff to assess the acceptability of a site for 

10 restricted release are the same as criteria used for approving Decommissioning Groups 4 – 5. 

11 This includes specific areas of consideration, including the source configuration, release 

12 mechanisms, and chemical form of the waste. Similarly, likely exposure scenarios and 

13 corresponding exposure pathways as well as the modeling approach and possible uncertainty 

14 issues are also evaluated. Specific differences that need to be considered when assessing 

15 whether the DP is in compliance with 10 CFR 20.1403 include the following:

16 • When assessing whether the licensee has identified and quantified the most significant

17 exposure scenarios based on available site- and facility-specific information, dose

18 considerations will need to be made for a minimum of two sets of exposure scenarios.

19 One set of exposure scenarios addresses the situation when institutional controls are in

20 place and working properly. The other set of exposure scenarios addresses the possible

21 doses that may occur if institutional controls are assumed to no longer be in effect. For

22 purposes of the evaluation, the licensee should assume the institutional controls fail at

23 time = 0 years. The NRC staff should review the basis and justification for the licensee’s

24 selected critical group for each exposure scenario. Section I.3 of Appendix I and

25 Appendix M contain additional guidance on these specific areas.

26 • As discussed in the previous section for Decommissioning Groups 4 – 5, evaluations of

27 the compliance exposure scenarios proposed by the licensee includes consideration of

28 the location and type of source (e.g., contaminated walls), the reasonably foreseeable

29 land use, the general characteristics and habits of the critical group (e.g., an adult light30 industry worker), and the possible pathways that describe how the residual radioactivity

31 would incur dose in humans. When considering restricted releases associated with

32 Decommissioning Group 6, potential limitations for the use of a specific compliance

33 exposure scenario should also consider limitations based on the established institutional

34 controls associated with restricted release. This is to ensure that if land uses other than

35 the reasonably foreseeable land use was to occur in the future, significant exposures

36 would not result.

37 • When establishing exposure scenarios, the licensee may use proposed restrictions as a

38 basis for eliminating or changing specific exposure pathways. For example, for

39 Decommissioning Groups 4 – 5, acceptable justifications for removing the groundwater

40 pathway include the following: (1) the near surface groundwater is neither potable nor

41 allowed to be used for irrigation, (2) the aquifer volume is insufficient to provide the

42 necessary yields, and (3) there are current (and informed consideration of future) land

43 use patterns that would support elimination of the groundwater pathway. For

44 Decommissioning Group 6, specific site restrictions precluding groundwater use (e.g.,

45 permits, regulations, etc.) could also be justification for the removal of the groundwater
5-26

1 pathway from consideration. However, it is important to note that in the case that 

2 institutional controls are assumed to no longer be in effect, it may be necessary to 

3 evaluate the exposure scenarios and pathways eliminated for the case when institutional 

4 controls are assumed to be effective.

5 5.5.4 Safety Evaluation Criteria – Decommissioning Group 7 (Alternate Release Criteria)

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

7 An alternative release proposal in accordance with 10 CFR 20.1404 may allow a dose of up to 

8 1.0 mSv/y (100 mrem/y) for baseline conditions with restrictions in place. However, for 

9 restricted release sites and specifically for the case where institutional controls are assumed to 

10 no longer be in effect, the dose may not exceed the values in 10 CFR 20.1403(e). Furthermore, 

11 the other provisions of 10 CFR 20.1403 must also be met.

12

13 The material to be reviewed by NRC staff should ensure that the licensee used defensible 

14 assumptions and models to establish and demonstrate compliance with the proposed alternate 

15 criteria. The staff should also verify that the licensee provided (1) enough information to allow 

16 an independent evaluation of the assessment resulting from the residual radioactivity after 

17 license termination and (2) reasonable assurance that the proposed decommissioning option 

18 complies with regulations. Each evaluation should be performed on a case-by-case basis. 

19

20 5.6 Summary Review Criteria 

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

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

23 5.3 provides information on review criteria for residual radioactivity associated with buildings. 

24 Both Tables 5.2 and 5.3 provide information for screening (column 1) versus site-specific 

25 reviews (column 2); and unrestricted (top) versus restricted (bottom) release scenarios. Staff 

26 should review the information provided by the licensee to ensure an adequate basis is provided 

27 to support the modeling assumptions and parameters selected commensurate with their risk28 significance as determined through sensitivity and uncertainty analysis. Data of sufficient 

29 quality should be collected to ensure the technical defensibility of the modeling results and 

30 DCGLs. The DQO process should be used to guide data collection and analysis. Staff should 

31 review the DQOs, including QA/QC requirements, to ensure DQOs are met during the DCGL

32 development and the FSS used to demonstrate compliance with release criteria.
5-27

1 Table 5.2 Dose Modeling Review Criteria for Residual Radioactivity in Soils

Screening (Soils) Site-Specific (Soils)

Unrestricted Release

Review source assumptions

• Nuclide(s) of interest

• Only surface soil residual radioactivity is

present (residual radioactivity is present in

approximately the top 30 cm of soil)

• Extent of heterogeneity of residual

radioactivity (i.e., survey units should be

relatively homogeneous)

• Method of determining average

concentration

Review modeling assumptions related to the 

source term

• Nuclide(s) of interest

• Area of residual radioactivity

• Thickness of residual radioactivity

• Extent of heterogeneity of residual

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

• Evaluate whether default critical group

(resident farmer) is appropriate for the site

and whether there are any other critical

groups which could incur higher doses.

Review exposure scenario assumptions

• Assess whether the exposure scenario(s)

and critical group(s) used to demonstrate

compliance are appropriate for the site.

• 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

• Check to make sure site conditions are

consistent with the built-in conceptual

model in the DandD screening code (see

Appendix I, Table I.5)

o groundwater initially free of residual

radioactivity

o surface water sediments initially free of

residual radioactivity

Review conceptual site model assumptions

• Check to make sure site conditions are

consistent with the built-in conceptual

model in RESRAD (see Appendix I, Tables

I.6 and I.7)

• Evaluate the relative location and activities

of the critical group.

• Evaluate the distribution of residual

radioactivity in the environment (is residual
5-28

o infiltration rate is greater than the

vertical saturated hydraulic conductivity

radioactivity initially present in the 

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 

• Evaluate whether screening or look-up

values (see Appendix H) are being used.

• If screening values are being used and

multiple radionuclides are present, check to

make sure that the sum of fractions value

(concentrations divided by screening

values) is not greater than 1.

• The final concentration should result in a

peak annual dose of less than 0.25 mSv

(25 mrem/)

Review mathematical models 

• Evaluate the mathematical models used to

assess dose (guidance in this volume

focuses on use of DandD and probabilistic

RESRAD, which were sponsored by the

NRC for use in decommissioning) and

compatibility with the site conceptual model.

• If custom models or non-traditional codes

are used, evaluate whether the

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 

• Ensure that the default parameters

developed for DandD (see NUREG/CR5512, Volume 3) are used.

• The only parameters that require user input

are the list of radionuclides and their

associated concentrations.

Review Parameter Assumptions 

• Review appropriateness of physical

parameters developed for the site

(licensees may use DandD default

behavioral and metabolic parameters)

• If changes are made to DandD default

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 

• If modeling is performed to develop

screening values, ensure that DandD and

the 90th percentile of the dose distribution is

used.

Consideration of Uncertainty 

• Evaluate parameter ranges to ensure that

parameter uncertainty is appropriately

considered in the dose modeling. Default

parameter distributions are available in

RESRAD 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.

Table 5.2 Dose Modeling Review Criteria for Residual Radioactivity in Soils (cont.)
5-29

• 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 off-site 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. 1

Table 5.2 Dose Modeling Review Criteria for Residual Radioactivity in Soils (cont.)
5-30

1 Figure 5.3 Dose Modeling Review Criteria for Residual Radioactivity in Buildings

Screening (Building) Site-Specific (Building)

Unrestricted Release

Review source assumptions

• Residual radioactivity is present on building

surfaces (no volumetrically contaminated

building materials)

• Extent of heterogeneity of residual

radioactivity

• Method of determining average

concentration

Review source assumptions

• Nuclide(s) of interest

• Area of residual radioactivity

• Thickness of residual radioactivity (thin

surface or volumetric building

contamination)

• Extent of heterogeneity of residual

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

• Evaluate whether default critical group

(building occupancy) is appropriate for the

site and whether there any other critical

groups which could incur higher doses.

Review exposure scenario assumptions

• Assess whether the exposure

scenario(s) and critical group(s) used to

demonstrate compliance are appropriate

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

• Check to make sure site conditions are

consistent with the conceptual model builtin the DandD screening code

o only building surface residual

radioactivity is present and

o loose contamination is 10 percent or

less or the removable fraction has been

adjusted

Review conceptual site model assumptions

• Evaluate the relative location and

activities of the critical group.

• Evaluate the distribution of residual

radioactivity in the building (surface

versus volumetric; floors, walls, ceiling,

equipment, piping, or sewer lines).

• Evaluate representation of the building in

the model (geometry and building

ventilation of rooms, building materials).

Review mathematical models and screening 

method

• Evaluate whether screening or look-up

values (see Appendix H) are being used.

• If screening values are being used and

multiple radionuclides are present, check to

make sure that the sum of fractions

(concentrations divided by screening

values) is no greater than 1.

Review mathematical models

• Evaluate the mathematical models used

to assess dose (guidance in this volume

focuses on use of DandD and RESRADBUILD, which were sponsored by the

NRC for use in decommissioning) and

compatibility with the site conceptual

model.
5-31

• If modeling is performed to develop

screening values, ensure that DandD is

used.

• If custom models or non-traditional

codes are used, evaluate whether the

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 

• Ensure that the default parameters

developed for DandD (see NUREG/CR5512, Volume 3) are used.

• The only parameters that require user input

are the list of radionuclides and their

associated concentrations.

Review Parameter Assumptions 

• Review appropriateness of parameters

developed for the site (licensees may

use DandD default behavioral and

metabolic parameters).

• 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 

• If modeling is performed to develop

screening values, ensure that DandD and

the 90th percentile of the dose distribution is

used.

Consideration of Uncertainty 

• Evaluate parameter ranges to ensure

that parameter uncertainty is

appropriately considered in the dose

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 risksignificant 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)

1

Figure 5.3 Dose Modeling Review Criteria for Residual Radioactivity in Buildings (cont.)
5-32

1 5.7 Additional Regulatory Guidance

2 In addition to the information provided above, there are numerous other references that the 

3 licensee or the NRC reviewer can refer to when developing and reviewing the dose modeling 

4 portions of DPs. Some of these references are listed below.

5

6 5.7.1 Regulatory Guidance Related to Decommissioning Groups 1 – 3 (Unrestricted 

7 Release Using Screening Criteria)

8 • Appendix H to this NUREG report

9 • NUREG-1757, Vol. 1, “Consolidated NMSS Decommissioning Guidance:

10 Decommissioning Process for Materials Licensees”

11 5.7.2 Regulatory Guidance Related to Decommissioning Groups 4 – 5 (Unrestricted 

12 Release Using Site-Specific Information)

13 • Appendix I to this NUREG report

14 • NUREG-1549, “Decision Methods for Dose Assessment to Comply with Radiological

15 Criteria for License Termination”

16 • NUREG/CR-5512, Volume 1, “Residual Radioactive Contamination from

17 Decommissioning: Technical Basis for Translating Contamination Levels to Annual

18 Total Effective Dose Equivalent”

19 • NUREG/CR-5512, Volume 2, “Residual Radioactive Contamination from

20 Decommissioning: User’s Manual, DandD Version 2.1”

21 • Draft NUREG/CR-5512, Volume 3, “Residual Radioactive Contamination from

22 Decommissioning: Parameter Analysis”

23 • EPA, Federal Guidance Report Number 11, “Limiting Values of Radionuclide Intake and

24 Air Concentration and Dose Conversion Factors for Inhalation, Submersion, and

25 Ingestion,” September 1988

26 • EPA, Federal Guidance Report Number 12, “External Exposure to Radionuclides in Air,

27 Water, and Soil,” September 1993

28 5.7.3 Regulatory Guidance Related to Decommissioning Group 6 (Restricted Release)

29 • NUREG-1549, “Decision Methods for Dose Assessment to Comply with Radiological

30 Criteria for License Termination”

31 • NUREG-1573, “A Performance Assessment Method for Low-Level Waste Disposal

32 Facilities: Recommendations of NRC’s Performance Assessment Working Group”

33 • Draft NUREG-2175, “Guidance for Conducting Technical Analyses for 10 CFR Part 61”
6-1

1 ALARA ANALYSES

2 This chapter is applicable to Decommissioning Groups 2–7. Licensees in Decommissioning 

3 Groups 2 and 3 may only have to refer to the discussion of good housekeeping practices in 

4 Section 6.3. 

5

6 6.1 Safety Evaluation Review Procedures

7 6.1.1 Areas of Review

8 The NRC staff should review the information supplied by the licensee or responsible party to 

9 determine if the licensee has developed a DP that ensures that doses to the average member of 

10 the critical group are ALARA. Information submitted should include (1) a cost-benefit analysis 

11 (or qualitative arguments) demonstrating that the applicable ALARA requirement(s) for the 

12 licensee’s preferred decommissioning option will be met and (2) a description of the licensee’s 

13 preferred method for showing compliance with the ALARA requirement at the time of 

14 decommissioning. If the licensee proposes to use engineered barriers or intentional mixing to 

15 meet the LTR criteria for unrestricted use, it should complete an appropriate ALARA analysis, 

16 as described in Section 3.5 of this volume (for engineered barriers) or Section 15.13 of 

17 Volume 1 (for intentional mixing). Additionally, an ALARA evaluation for restricted use should 

18 follow guidance described in Appendix N of this NUREG-1757, Volume 2. The licensee should 

19 also follow the guidance in Appendix N of this volume when evaluating the eligibility tests for 

20 restricted use (10 CFR 20.1403(a), ALARA test, or test of net public or environmental harm) and 

21 use of a higher dose limit for restricted use with institutional controls not in place 

22 (10 CFR 20.1403(e)(2)(i), test for prohibitively expensive, or test for net public or environmental 

23 harm). 

24

25 6.1.2 Review Procedures

26 6.1.2.1 Acceptance Review

27 The NRC staff should review the DP to ensure that, at a minimum, it contains the information 

28 summarized under the above “Areas of Review.” The NRC staff should review the ALARA 

29 portion of the DP without assessing the technical accuracy or completeness of the information 

30 contained therein, which it should determine during the detailed technical review. The NRC 

31 staff should review the DP table of contents and the individual descriptions under “Areas of 

32 Review” (1) to ensure that the licensee or responsible party has included this information in the 

33 DP and (2) determine if the level of detail of the information appears to be adequate for a 

34 detailed technical review.

35

36 6.1.2.2 Safety Evaluation

37 The material supporting the ALARA portion of the DP to be reviewed is technical in nature and 

38 specific detailed technical analysis may be necessary. The NRC staff should evaluate the 

39 licensee’s dose estimates for various alternatives using the appropriate guidance in Chapter 5 

40 of this volume and should evaluate the licensee’s cost estimates using the guidance in 

41 Section 4.1 from NUREG-1757, Volume 3. 
6-2

1 6.2 Acceptance Criteria

2 6.2.1 Regulatory Requirements

3 10 CFR 20.1101(b), 20.1402, 20.1403(a), 20.1403(b), 20.1403(e), and 20.1404(a)(3)

4

5 6.2.2 Regulatory Guidance

6 Appendix N of Volume 2 of this NUREG report

7

8 6.2.3 Information to be Submitted

9 The information supplied by the licensee should be sufficient to allow the NRC staff to fully 

10 understand the basis for the licensee’s conclusion that projected dose or residual radioactivity 

11 concentrations and quantity (hereafter, the decommissioning goal) are ALARA. The 

12 decommissioning goal should be established at the point where the incremental benefits equal 

13 the incremental costs. The NRC staff review should verify that the following information is 

14 included in the description of the development of the decommissioning goal:

15

16 • a description of how the licensee will achieve a decommissioning goal that meets the

17 dose limit and ALARA requirement

18 • a quantitative cost-benefit analysis

19 • a description of how costs were estimated

20 • a demonstration that the doses to the average member of the critical group are ALARA

21 The information to be submitted is also included as part of the master DP Checklist provided in 

22 this NUREG report (see Section VII from Appendix D of Volume 1).

23

24 6.3 Evaluation Criteria

25 6.3.1 Evaluation of Good Practice Efforts

26 For ALARA during decommissioning, all licensees should use typical good practice, such as 

27 floor and wall washing, removal of readily removable radioactivity in buildings or in soil areas, 

28 and other good housekeeping practices. In addition, the FSSR should describe how the 

29 licensee employed these practices to achieve the final activity levels.

30

31 6.3.2 Evaluation of Cost-Benefit ALARA Analyses

32 The NRC staff review should verify that the qualitative descriptions provide reasonable 

33 assurance that the activities and decommissioning goal should result in doses to the average 

34 member of the critical group that are ALARA. For those situations in which a licensee prepares 

35 cost-benefit analyses, the NRC staff should ensure that the analyses are developed using the 

36 methodology described in Appendix N and are applied as described in the following text.
6-3

1 6.3.3 When Cost-Benefit Analyses are Unnecessary

2 In the following circumstances, the results of an ALARA analysis are known on a generic basis 

3 and the NRC staff considers that an analysis is not necessary (see Appendix N of this volume 

4 for more details):

5

6 • unrestricted use where excavated soil would be shipped to a LLW disposal facility for

7 disposal

8 • soil removal at a site to meet the unrestricted use dose criteria of 25 mrem/y

9 • remediation of building surfaces or surface soil to the NRC default screening levels (see

10 Appendix H for information about the screening levels)

11 • that no residual radioactivity distinguishable from background will remain at the site at

12 termination

13 • that loose residual radioactivity on building surfaces has been or will be removed

14 6.3.4 Calculation of Benefits

15 Appendix N of this volume discusses five different possible benefits: (1) collective dose averted, 

16 (2) regulatory costs avoided, (3) changes in land values, (4) esthetics, and (5) reduction in

17 public opposition. Numerical estimates will generally only be available for the first three

18 benefits, if they are appropriate. The licensee can make a qualitative analysis of the benefits,

19 especially if the costs are large (e.g., no matter what the change in land value is, the costs will

20 exceed the benefits). In most comparisons between alternatives in the same class (e.g., both

21 alternatives result in unrestricted release), the only important benefit should be the collective

22 dose averted. In comparisons between restricted and unrestricted release, the other benefits

23 can become important.

24

25 The collective dose averted is generalized as the incremental dose difference between the 

26 licensee’s approach (hereafter, preferred option) and the alternative under analysis. Therefore, 

27 the NRC staff should ensure that the licensee has calculated the benefits correctly by using the 

28 correct population density, area, and averted dose. This may require a technical analysis of the 

29 dose modeling, and the reviewer should use Chapter 5 for these cases. If the licensee has 

30 used discounting, the NRC staff should ensure that the proper rates were used. The licensee is 

31 not required to discount because the discount reduces the benefits of averting dose in later time 

32 periods.

33

34 One acceptable method of compliance with 10 CFR 20.1403(a) is to demonstrate that cleanup 

35 to the unrestricted release criteria is beyond ALARA considerations. In this case, a beneficial 

36 estimate should include costs that would be avoided if the site were to be released for 

37 unrestricted use, such as site control and maintenance costs, as well as an estimate of the 

38 additional regulatory costs associated with termination of a restricted site (e.g., development of 

39 an environmental impact statement, public meetings). Appendix N of this volume contains more 

40 details on compliance with ALARA requirements for restricted use. 

41

42 The NRC staff should ensure that the licensee has properly documented the basis for any 

43 estimates of changes in land values. Acceptable sources of such estimates include 
6-4

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

2 and the issues involved.

3

4 6.3.5 Calculation of Costs

5 The NRC staff should verify that the licensee has adequately estimated the effective monetary 

6 costs of the incremental remediation by using the equations in Appendix N of this volume. To 

7 review the calculated monetary costs of the incremental remediation, the NRC staff should use 

8 Section 4.1 of NUREG-1757, Volume 3, with the following changes (this may require calculating 

9 total cost estimates for the preferred option and each alternative):

10

11 • The cost estimate should be based on actual costs expected to be incurred by 

12 decommissioning the facility and should not assume that the work will be performed by 

13 an independent third-party contractor.

14 • The cost estimate does take credit for (1) any salvage value that might be realized from 

15 the sale of potential assets during or after decommissioning or (2) any tax reduction that 

16 might result from paying decommissioning costs or site control and maintenance costs.

17 • The decommissioning cost estimates should reflect the actual situation rather than 

18 maximized assumptions.

19 For each of the cost terms (e.g., disposal costs, worker fatalities), the incremental difference 

20 between the preferred and the alternative options may be negative (i.e., the alternative may cost 

21 less than the preferred option).

22

23 6.3.6 Compliance Methods at the Time of Decommissioning

24 There are two approaches to demonstrate compliance with the ALARA requirement at the end 

25 of decommissioning: (1) a predetermined acceptable dose limit or concentration guideline(s) or 

26 (2) an acceptable preferred option and decommissioning goal with organizational oversight and 

27 review during decommissioning. Both options have advantages and disadvantages. The 

28 licensee establishes the compliance method, with the staff reviewing the applicability, given the 

29 site-specific information.

30

31 6.3.6.1 Predetermined Compliance Measure

32 Under the predetermined compliance measure, the licensee would agree to meet the dose 

33 calculated for the preferred option or the radiological concentrations associated with that dose. 

34 This could be met by either establishing deterministic concentration limits for the site or 

35 agreeing to use a specified dose scenario with associated parameters and assumptions. If the 

36 licensee’s final survey results meet the self-imposed concentration limits (or dose limit), the 

37 licensee has met the ALARA requirement.

38

39 6.3.6.2 Performance-Based Compliance

40 Performance-based compliance allows a licensee to adjust its ALARA assessment during 

41 decommissioning to deal with actual site conditions experienced and actual costs incurred. The 

42 philosophy behind this compliance measure is very similar to how ALARA is handled during 

43 routine operations. The licensee’s DP needs to meet all of the following criteria to use this

44 approach:
6-5

1

2 • The preferred option, based on valid assumptions, would result in reducing residual

3 activity to ALARA levels, as described above.

4 • The licensee has established decommissioning guidelines (either dose or

5 concentrations) based on the DP’s analysis.

6 • The licensee has a documented method to review the effectiveness of the remediation

7 activities. This method should include all of the following:

8 o an ALARA committee or radiation safety officer for small licensees, similar to

9 operations requirements

10 o an appropriate review frequency established

11 o an acceptable set of criteria on the scope of activities or commitments that the

12 ALARA committee can change

13 o a commitment to prepare acceptable documentation of ALARA findings that result in

14 the licensee changing its remediation activities or decommissioning guidelines

15 o a commitment to provide the NRC annually with all necessary page changes to the

16 DP because of ALARA findings

17 At the end of remediation, a licensee using the performance-based approach should meet the 

18 following criteria:

19

20 • The final survey results satisfy the appropriate dose limit(s).

21 • Any substantial weaknesses in the ALARA program that were found during licensee

22 audits or NRC inspections have been resolved.

23 • Any deviation from the decommissioning goal presented in the DP is properly justified by

24 the ALARA committee findings.

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

7-1

1 BIBLIOGRAPHY AND SUPERSEDED DOCUMENTS

2 This section provides the reference list for the main body of this volume, categorized in the 

3 following subsections by type of reference document. Chapter 4 of Volume 1 of this NUREG 

4 series provided a more general list of decommissioning references, which included statutes, 

5 decommissioning regulations, decommissioning inspection manual chapters, and 

6 decommissioning inspection procedures.

7

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.

8

9 7.1 NRC Decommissioning Documents Referenced in the Main Body of Vol. 2

10 United States Nuclear Regulatory Commission (US NRC). “Manual for Conducting 

11 Radiological Surveys in Support of License Termination, Draft Report for Comment,” 

12 NUREG/CR-5849. NRC: Washington, DC. 1992a.

13 — — — — —. “Residual Radioactive Contamination from Decommissioning: Technical Basis 

14 for Translating Contamination Levels to Annual Total Effective Dose Equivalent,” 

15 NUREG/CR-5512, Vol. 1. NRC: Washington, DC. 1992b.

16 — — — — —. “Background as a Residual Radioactivity Criterion for Decommissioning—Draft 

17 Report,” NUREG-1501. NRC: Washington, DC. August 1994a.

18 — — — — —. “Draft Branch Technical Position on Site Characterization for Decommissioning.” 

19 NRC: Washington, DC. 1994b.

20 — — — — —. “Standard Review Plan for the Review of a License Application for a Low-Level 

21 Waste Disposal Facility,” NUREG-1200, Rev. 3. NRC: Washington, DC. 1994c.

22 — — — — —. “Working Draft Regulatory Guide on Release Criteria for Decommissioning: 

23 NRC Staff’s Draft for Comment,” NUREG-1500. NRC: Washington, DC. 1994d.

24 — — — — —. “Guidelines for Preparing and Reviewing Applications for Licensing of Non25 Power Reactors,” NUREG-1537. NRC: Washington, DC. 1996.

26 — — — — —. “NMSS Handbook for Decommissioning Fuel Cycle and Materials Licensees,” 

27 NUREG/BR-0241. NRC: Washington, DC. 1997.

28 — — — — —. “Decision Methods for Dose Assessment To Comply with Radiological Criteria 

29 for License Termination, Draft Report for Comment,” NUREG-1549. NRC: Washington, DC. 

30 1998a.
7-2

1 — — — — —. “Minimum Detectable Concentrations with Typical Radiation Survey Instruments 

2 for Various Contaminants and Field Conditions,” NUREG-1507. NRC: Washington, DC. 

3 1998b.

4 — — — — —. “A Proposed Nonparametric Statistical Methodology for the Design and Analysis 

5 of Final Status Decommissioning Surveys: Interim Draft Report for Comment and Use,” 

6 NUREG-1505, Rev. 1. NRC: Washington, DC. 1998c.

7 — — — — —. “Residual Radioactive Contamination from Decommissioning, Parameter 

8 Analysis, Draft Report for Comment,” NUREG/CR-5512, Vol. 3. NRC: Washington, DC. 1999.

9 — — — — —. “Multi-Agency Radiation Survey and Site Investigation Manual (MARSSIM),” 

10 NUREG-1575, Rev. 1, EPA 402-R-97-016, Rev. 1, DOE/EH-0624, Rev. 1. NRC: Washington, 

11 DC, U.S. Department of Defense, U.S. Department of Energy, U.S. Environmental Protection 

Agency. 2000a.

1 12 

13 — — — — —. “NMSS Decommissioning Standard Review Plan,” NUREG-1727. NRC: 

14 Washington, DC. 2000a.

15 — — — — —. “Draft Staff Guidance for Dose Modeling of Proposed Partial Site Releases.” 

16 Memorandum from John T. Greeves to John A. Zwolinski. NRC: Washington, DC. 2001a.

17 — — — — —. “Residual Radioactive Contamination from Decommissioning, User’s Manual, 

18 DandD Version 2.1,” NUREG/CR-5512, Vol. 2. NRC: Washington, DC. 2001a.

19 — — — — —. “Decommissioning Criteria for the West Valley Demonstration Project (M–32) at 

20 the West Valley Site: Final Policy Statement.” Federal Register, Vol. 67, No. 22, pp. 5003–

21 5012. 2002a.

22 — — — — —. “Re-Evaluation of the Indoor Resuspension Factor for the Screening Analysis of 

23 the Building Occupancy Scenario for NRC’s License Termiantion Rule—Draft Report for 

24 Comment,” NUREG-1720. NRC: Washington, DC. 2002b.

25 — — — — —. “Results of the License Termination Rule Analysis,” SECY-03-0069. NRC: 

26 Washington, DC. 2003a.

27 — — — — —. “Staff Requirements—SECY-03-0069—Results of the License Termination Rule 

28 Analysis,” SRM-SECY-03-0069. NRC: Washington, DC. 2003b.

29 — — — — —. “Results of the License Termination Rule Analysis of the Use of Intentional 

30 Mixing of Contaminated Soil,” SECY-04-0035. NRC: Washington, DC. 2004a.

31 — — — — —. “Staff Requirements—SECY-04-0035—Results of the License Termination Rule 

32 Analysis of the Use of Intentional Mixing of Contaminated Soil,” SRM-SECY-04-0035. NRC: 

33 Washington, DC. 2004b.

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/.
7-3

1 — — — — —. “Results of the License Termination Rule Analysis,” Regulatory Issue Summary 

2 2004-08. NRC: Washington, DC. 2004c.

3 — — — — —. “Consolidated Decommissioning Guidance – Decommissioning Process for 

4 Materials Licensees,” NUREG-1757, Vol. 1, Rev.2. NRC: Washington, DC. 2006a.

5 — — — — —. “Stakeholder Comments and Path Forward on Decommissioning Guidance To 

6 Address License Termination Rule Analysis Issues,” SECY-06-0143. NRC: Washington, DC. 

7 2006b.

8 — — — — —. “Staff Requirements—SECY-06-0143—Stakeholder Comments and Path 

9 Forward on Decommissioning Guidance to Address License Termination Rule Analysis Issues,”

10 SRM-SECY-06-0143. NRC: Washington, DC. 2006c.

11 — — — — —. “Decommissioning Oversight and Inspection Program for Fuel Cycle Facilities 

12 and Materials Licensees,” Inspection Manual Chapter 2602. NRC: Washington, DC. 2008.

13 — — — — —. “Guidance for Conducting Technical Anlayses for 10 CFR Part 612, Draft Report 

14 for Comment,” NUREG-2175. NRC: Washington, DC. 2015.

15 — — — — —. “Standard Review Plan for Evaluating Nuclear Power Reactor License 

16 Termination Plans,” NUREG-1700, Rev. 2. NRC: Washington, DC. 2018.

17 7.2 Other NRC Documents Referenced in this Volume 

18 Code of Federal Regulations (CFR). 10 CFR Part 20, “Standards for Protection against 

19 Radiation.” 

20 — — — — —. 10 CFR Part 30, “Rules of General Applicability to Domestic Licensing of 

21 Byproduct Material.” 

22 — — — — —. 10 CFR Part 40, “Domestic Licensing of Source Material.”

23 — — — — —. 10 CFR Part 50, “Domestic Licensing of Production and Utilization Facilities.” 

24 — — — — —. 10 CFR Part 70, “Domestic Licensing of Special Nuclear Material.” 

25 — — — — —. 10 CFR Part 72, “Licensing Requirements for the Independent Storage of Spent 

26 Nuclear Fuel and High-Level Radioactive Waste.” 

27 NRC. Staff Requirements Memorandum, SECY-98-144, “White Paper on Risk-Informed and 

28 Performance-Based Regulation.” NRC: Washington, DC. March 1999.

29

30 — — — — —. “A Performance Assessment Method for Low-Level Waste Disposal Facilities: 

31 Recommendations of NRC’s Performance Assessment Working Group,” NUREG-1573. NRC: 

32 Washington, DC. 2000.

33 — — — — —. “Environmental Review Guidance for Licensing Actions Associated with NMSS 

34 Programs,” NUREG-1748. NRC: Washington, DC. 2001.
7-4

1 — — — — —. “Standard Review Plan for the Review of a Reclamation Plan for Mill Tailings 

2 Sites Under Title II of the Uranium Mill Tailings Radiation Control Act,” NUREG-1620, Rev. 1.

3 NRC: Washington, DC. 2003.

4 — — — — —. “Jurisdiction Determinations,” SA-500, FSME Procedure Approval. ADAMS 

5 Accession No. ML110600350. NRC: Washington, DC. 2011.

6 — — — — —. “Clarification on the Determination of Regulatory Jurisdiction of Non-Federal 

7 Entities Conducting Cleanup Activities on Federal Property in Agreement States,” 

8 FSME-14-039. NRC: Washington, DC. 2014.

9

10 7.3 Other References in this Volume 

11 American National Standards Institute. “Characterization in Support of Decommissioning Using 

12 the Data Quality Objectives Process,” ANSI N13.59:2008. ANSI: New York, NY. 2008.

13 Atomic Energy Act of 1954, as amended. 42 U.S.C. §§2011 et seq.

14 Chapman, J.A., A.J. Boerner, E.W. Abelquist. “Spatially-Dependent Measurements of Surface 

15 and Near-Surface Radioactive Material Using In situ Gamma Ray Spectrometry (ISGRS) For 

16 Final Status Surveys.” 2006. 

17 Code of Federal Regulations. 40 CFR Part 190, “Environmental Radiation Protection Standards 

18 for Nuclear Power Operations”.

19 Comprehensive Environmental Response, Compensation, and Liability Act of 1980 (CERCLA). 

20 42 U.S.C. §§9601.

21 Department of Energy (US DOE). “Decommissioning Handbook,” DOE/EM-0142P. DOE: 

22 Washington, DC. 1994.

23 Environmental Protection Agency (US EPA). “Guidance for Conducting Remedial Investigations 

24 and Feasibility Studies Under CERCLA,” EPA/540/G-89/004. EPA: Washington, DC. 1988a.

25 — — — — —. “Limiting Values of Radionuclide Intake and Air Concentration and Dose 

26 Conversion Factors for Inhalation, Submersion, and Ingestion: Federal Guidance Report No. 

27 11,” EPA 520/1-88-020. EPA: Washington, DC. 1988b. 

28 — — — — —. “Superfund Removal Procedures,” OSWER Directive 9360.0-03B. EPA: 

29 Washington, DC. 1988c.

30 — — — — —. “External Exposure to Radionuclides in Air, Water, and Soil: Federal Guidance 

31 Report No. 12,” EPA 402-R-93-081. EPA: Washington, DC. 1993.

32 — — — — —. “Federal Radiation Protection Draft Guidance for Exposure of the General 

33 Public.” Federal Register, Vol. 59, p. 66414. US EPA: Washington DC. 1994.

34 — — — — —. “Guidance for the Data Quality Objectives Process, EPA QA/G-4,” EPA/600/R35 96/055. EPA: Washington, DC. 2000.
7-5

1 Resource Conservation and Recovery Act, 42 U.S.C. §§ 6901-6991 (1976).

2

3 Uranium Mill Tailings Radiation Control Act, 42 U.S.C. §§ 7901-7925 (1978).

4

5

6 7.4 Documents Superseded by this Volume

7 This volume supersedes the guidance documents listed in Table 7.4, and the superseded 

8 documents should no longer be used.

9 Table 7.1 Documents Superseded by this Report

Document Title Date

NRC Memorandum Draft Staff Guidance for Dose Modeling of Proposed 

Partial Site Releases

09/2001

NUREG-1727 NMSS Decommissioning Standard Review Plan 09/2000

NUREG/BR-0241 NMSS Handbook for Decommissioning Fuel Cycle and 

Materials Licensees

03/1997

Branch Technical 

Position

Draft Branch Technical Position: Screening 

Methodology for Assessing Prior Land Burials of 

Radioactive Waste Authorized Under Former 

10 CFR 20.304 and 20.302

10/1996

Branch Technical 

Position

Draft Branch Technical Position on Site 

Characterization for Decommissioning

11/1994

NUREG-1500 Working Draft Regulatory Guide on Release Criteria for 

Decommissioning: NRC Staff’s Draft for Comment 

08/1994

NUREG/CR-5849 Manual for Conducting Radiological Surveys in Support 

of License Termination 

06/1992

Branch Technical 

Position

Disposal of Onsite Storage of Thorium or Uranium 

Wastes from Past Operations

10/1981

10
7-6

1 This Volume 2 of this NUREG report also incorporates and updates numerous portions of 

2 NUREG-1727, “NMSS Decommissioning Standard Review Plan” (SRP), issued 

3 September 2000, specifically Chapters 5, 7, and 14; and Appendices C, D, and E of 

4 NUREG-1727 on dose modeling, ALARA, and facility radiation surveys. The NUREG-1727 

5 chapters and appendices that have been incorporated into this NUREG are superseded. This 

6 three-volume NUREG series (i.e., NUREG-1757, Volumes 1, 2, and 3) supersedes both 

7 NUREG/BR-0241, “NMSS Handbook for Decommissioning Fuel Cycle and Materials 

8 Licensees,” issued March 1997, and NUREG-1727 in their entirety.
IMPLEMENTING THE MARSSIM APPROACH FOR CONDUCTING 

FINAL RADIOLOGICAL SURVEYS

A-1

1 A.1 Introduction

2 This appendix is applicable to Decommissioning Groups 2–7.

3 Regulations of the NRC in Title 10 of the Code of Federal Regulations (10 CFR) 20.1501(a) 

4 require licensees to make or cause to be made surveys that may be necessary for the licensee 

5 to comply with the regulations in 10 CFR Part 20, “Standards for Protection against Radiation.”

6 The FSS is the radiation survey performed after an area has been fully characterized, 

7 remediation has been completed, and the licensee believes that the area is ready to be 

8 released. The purpose of the FSS is to demonstrate that the area meets the radiological criteria 

9 for license termination. The FSS is not conducted for the purpose of locating residual 

10 radioactivity; the HSA and the characterization survey perform that function.

11 The NRC endorses the FSS method described in NUREG-1575, “Multi-Agency Radiological 

12 Survey and Site Investigation Manual” (MARSSIM), Revision 1, issued August 2000, and 

13 references to MARSSIM sections within this appendix are specifically referring to MARSSIM 

Revision 1 (NRC, 2000).

1 14 This appendix (1) provides an overview of the MARSSIM approach 

15 for conducting a final radiological survey, (2) includes additional specific guidance on acceptable 

16 values for use in the MARSSIM method, (3) explains how to use the MARSSIM method in a way 

17 that is consistent with the dose modeling, (4) describes how to use the MARSSIM method to 

18 meet the NRC’s regulations, and (5) demonstrates how to extend or supplement the MARSSIM 

19 method to certain complex situations that may be encountered, such as how to address 

20 subsurface residual radioactivity. Note that the guidance in this appendix does not replace 

21 MARSSIM, and licensees and reviewers should refer to, and use, MARSSIM for designing final 

22 radiological surveys to support decommissioning. This guidance assumes a working knowledge 

23 of the MARSSIM approach and terminology and does not attempt to provide a comprehensive 

24 overview of the entire MARSSIM. In addition, for Decommissioning Groups 1–3, licensees may 

25 also use the alternative, simpler final survey methods described in Appendix B of this volume. 

26 Chapter 5 of MARSSIM contains survey checklists. These checklists are useful in implementing 

27 the steps of the RSSI process (Decommissioning Groups 3–7). These checklists are a useful 

28 tool for visualizing the sequential steps (i.e., design, performance, and evaluation) of the survey 

29 process. Furthermore, the use of these checklists should ensure that the necessary information 

30 is collected for each type of survey. Sites not using the RSSI process, such as 

31 Decommissioning Groups 1 and 2, should also find these checklists or parts of these checklists 

32 useful.

33 A.2 Classification of Areas by Residual Radioactivity Levels

34 The licensee should classify site areas based on levels of residual radioactivity from licensed 

35 activities. The area classification method contained in Section 4.4 of MARSSIM is acceptable to 

36 the NRC staff. Its essential features are described below. 

37

38 The licensee should first classify site areas as impacted or nonimpacted. Impacted areas are 

39 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, is being developed but is not reflected in this guidance. The decommissioning website should be 

consulted for issuance of technical reports providing guidance and a listing of any lessons learned between 

guidance revisions.
A-2

1 areas without residual radioactivity from licensed activities. Impacted areas should be identified 

2 by using knowledge of past site operations together with site characterization surveys. In the 

3 FSS, radiation surveys do not need to be conducted in nonimpacted areas. The licensee 

4 should classify impacted areas into one of the three classes, listed below, based on levels of 

5 residual radioactivity.

6 (1) Class 1 Areas: Class 1 areas are impacted areas that are expected to have

7 concentrations of residual radioactivity that exceed the derived concentration guideline

8 level or DCGLW (average concentrations over a wide area).

9 (2) Class 2 Areas: Class 2 areas are impacted areas that are not likely to have

10 concentrations of residual radioactivity that exceed the DCGLW.

11 (3) Class 3 Areas: Class 3 areas are impacted areas that have a low probability of

12 containing residual radioactivity.

13 Surveys conducted during operations or during characterization at the start of decommissioning 

14 are the basis for classifying areas. If the available information is not sufficient to designate an 

15 area as a particular class, the area either should be classified as Class 1 or should be further 

16 characterized. Areas that are considered to be on the borderline between classes should 

17 receive the more restrictive classification.

18 The NRC staff recognizes that a licensee may need to reclassify Class 1 areas to Class 2, when 

19 insufficient information is available for the initial classification. If more information becomes 

20 available to indicate that another classification is more appropriate, the guidance in MARSSIM 

21 allows for classifications to be changed at any time before the FSS. For more guidance on 

22 criteria for downgrading classifications (e.g., from Class 1 to Class 2), a licensee should refer to 

23 MARSSIM, Revision 1; in particular, Sections 2.2, 2.5.2, and 5.5.3. If a licensee plans to make 

24 use of reclassification during the RSSI process, it should include in the DP the criteria and 

25 methodology it plans to use for reclassification. In addition, a licensee contemplating the use of 

26 reclassification is encouraged to contact NRC staff.

27 For Class 1 and 2 soils, the licensee should determine whether a significant amount of

28 subsurface residual radioactivity is present, based on the HSA and site characterization. In this 

29 context “significant amounts of subsurface residual radioactivity” would be defined as an amount 

30 of radioactivity, or contaminated material (such as soil), that could contribute at least 10 percent 

31 of the dose criteria (see Section 3.3). 

32 The presence of subsurface soils is important, because subsurface soils cannot typically be 

33 measured by scan instrumentation and, therefore, scan surveys are not expected to be 

34 adequate for the purposes of an FSS. Although surface soils have been associated with the top 

35 15 centimeters (cm) of soil, which can typically be measured by scan instrumentation, the exact 

36 depth of residual radioactivity for which scan survey instrumentation can adequately detect 

37 residual radioactivity varies, based on several factors (e.g., survey instrument, radionuclide, and 

38 soil characteristics). 

39 Determining whether there is a significant amount of subsurface residual radioactivity should not 

40 require a complex set of characterization measurements. In most cases, there will be either 

41 significant amounts of residual radioactivity or only traces (such as in occasional small pockets 

42 or from leaching from surface layers by rainwater). When there is an insignificant amount of 

43 subsurface residual radioactivity, the MARSSIM survey methods for surface measurements are
A-3

1 acceptable. When there is a significant amount of subsurface residual radioactivity, the licensee 

2 should modify the dose modeling and survey methods to account for it.

3 The HSA usually determines the presence of subsurface residual radioactivity (see Chapter 3 of 

4 MARSSIM), applying knowledge of how the residual radioactivity was deposited. 

5 Characterization surveys to detect subsurface residual radioactivity in soil are not routinely 

6 conducted, unless there is reason to expect that subsurface residual radioactivity may be 

7 present. The need to survey or sample subsurface soil will depend, in large part, on the quality 

8 of the information used to develop the HSA, the environmental conditions at the site, the types 

9 and forms (chemical and radiological) of the radioactive material used at the site, the authorized 

10 activities and the manner in which licensed material was managed during operations.

The NRC staff’s 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.

11

12 A.3 Selection and Size of Survey Units

13 The licensee should divide the impacted area into survey units based on the classification 

14 described above. A survey unit is a portion of a building or site that is surveyed, evaluated, and 

15 released as a single unit. The licensee should give the entire survey unit the same area 

16 classification. Section 4.6 of MARSSIM contains a method acceptable to the NRC staff for 

17 dividing impacted areas into survey units. The important features of this method are 

18 summarized here. 

19 For buildings, it is normally appropriate to designate each separate room as either one or two 

20 survey units (e.g., floors with the lower half of walls and upper half of walls with ceiling), based 

21 on the pattern of potential of residual radioactivity. It is generally not appropriate to divide

rooms of normal size (100 square meters (m2 22 ) or less) into more than two survey units, because 

23 the dose modeling is based on the room being considered as a single unit. However, very large 

24 spaces such as warehouses may be divided into multiple survey units.
A-4

1

2

3

4

5

6

7

8

9

10

11

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. 

12 Table A.1 Suggested Survey Unit Areas (MARSSIM, Rev. 1, Roadmap, Table 1 

13 [NRC, 2000])

Suggested Survey Unit Area

Class Structures Land

1 up to 100 m2 up to 2,000 m2

2 100 to 1,000 m2 2,000 to 10,000 m2

3 no limit no limit

14

15

16

17

18

19

20

21

22

23

24

25

26

27

28

29

30

31

32

33

34

35

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. 

36
A-5

1 A.4.2 Characteristics of Soil Reference Areas

2 The objective is to select nonimpacted background reference areas where the distribution of 

3 measurements should be the same as would be expected in the survey unit if that survey unit 

4 had never been contaminated. Section 4.5 of MARSSIM contains an acceptable method for

5 selecting background areas, as briefly described below. 

6 Reference areas should have a soil type as similar to the soil type in the survey unit as possible. 

7 If there is a choice of possible reference areas with similar soil types, consideration should be

8 given to selecting reference areas that are most similar in terms of other physical, chemical, 

9 geological, and biological characteristics. Each reference area should have an area at least as 

10 large as the survey unit, if practical, to include the full potential spatial variability in background 

11 concentrations. Reference areas may be off site or on site, as long as they are nonimpacted. 

12 NUREG-1506, “Measurement Methods for Radiological Surveys in Support of New 

13 Decommissioning Criteria, Draft Report for Comment,” issued August 1995, provides additional 

14 information on reference area selection. Licensees should contact the NRC staff when they are 

15 unable to find a reference area that satisfies the above criteria.

16 A.4.3 Different Materials in a Survey Unit

17 Survey units may contain a variety of materials with markedly different backgrounds. An 

18 example might be a room with drywall walls, concrete floor, glass windows, metal doors, wood 

19 trim, and plastic fixtures. It is not appropriate to make each material a separate survey unit, 

20 because the dose modeling is based on the dose from the room as a whole and because a 

21 large number of survey units in a room would require an inappropriate number of samples. 

22 When there are different materials with substantially different backgrounds in a survey unit, the 

23 licensee may use a reference area that is a nonimpacted room with roughly the same mix of 

24 materials as the survey unit.

25 If a survey unit contains several different materials, but one material is predominant, or if there 

26 is not too great a variation in background among materials, a background from a reference area 

27 containing only a single material may still be appropriate. For example, a room may be mostly 

28 concrete but with some metal beams, and the residual radioactivity may be mostly on the 

29 concrete. In this situation, where the concrete predominates, it would be acceptable to use a 

30 reference area that contained only concrete. However, the licensee should demonstrate that 

31 the selected reference area will not result in underestimating the residual radioactivity on other 

32 materials.

33 The licensee may also use measured backgrounds for different materials or for groups of similar 

34 materials. When the licensee decides to use different measured backgrounds for different 

35 materials or for a group of materials with similar backgrounds, it is acceptable to perform a 

36 one-sample test on the difference between the paired measurements from the survey unit and 

37 from the appropriate reference material. Chapter 12 of NUREG-1505 describes, in detail, an 

38 acceptable method to do this.

39 For onsite materials, present either in buildings or as nonsoil materials present in outdoor 

40 survey units (e.g., concrete, brick, drywall, fly ash, petroleum product wastes), the licensee 

41 should attempt to find nonimpacted materials that are as similar as possible to the materials on 

42 the site. Sometimes such materials will not be available. In those situations, the licensee 
A-6

1 should make a good faith effort to find the most similar materials readily available or use

2 appropriate published estimates.

3

4

5

6

7

8

9

10

11

12

13

14

15

16

17

18

19

20

21

22

23

24

25

26

27

28

29

30

31

32

33

34

35

36

37

38

39

40

41

42

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 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) 

43 should be used at both the survey unit and the reference area. 
A-7

1 NUREG-1505 provides additional guidance on survey design, the methods of accounting for 

2 background radiation, and the nonparametric statistical methods for testing compliance with the 

3 decommissioning criteria in 10 CFR Part 20, Subpart E. Formulas contained in NUREG-1505

4 can be used to compute the required number of samples (measurement points) that will be 

5 needed in both the background reference and survey areas.

6 A.5 Methods to Evaluate Survey Results

7 All survey units should be evaluated to determine whether the average concentration in the

8 survey unit as a whole is below the DCGLW. If the radionuclide is not present in background, 

9 and the measurement technique is radionuclide-specific so that a comparison with a reference 

10 area is not necessary, the licensee should use a one-sample test, the Sign test. Section 8.3 of 

11 MARSSIM describes this test.

12

13 When the residual radioactivity contains a radionuclide present in the environment or when the 

14 measurements are not radionuclide specific, the survey unit should be compared to a reference 

15 area. When comparing the survey unit to a reference area, the licensee should use a two16 sample test, the WRS test. Section 8.4 of MARSSIM describes this test.

17 A.5.1 A Case for not Subtracting Background

18 An exception to using a two-sample test when a radionuclide is present in background is when 

19 the licensee plans to assume that all the radionuclide activity in the survey unit is caused by 

20 licensed operations and none is from background. This could be the case for cesium (Cs)-137, 

21 for example, because the levels in the environment are often so much less than the DCGLW that 

22 background concentrations may be ignored.

23

24 A.5.2 Elevated Measurements Comparison

25 Class 1 survey units that pass the Sign test or WRS test but have small areas with 

26 concentrations exceeding the DCGLW should also be tested to demonstrate that those small 

27 areas meet the dose criteria for license termination. This test is called the elevated 

28 measurement comparison (EMC). It is described in Section 8.5.1 of MARSSIM and 

29 summarized here.

30

31 To perform the EMC, the licensee first determines the size of the area in the survey unit with a 

32 concentration greater than the DCGLW, then determines the DCGLEMC for an area of that size. 

33 (The DCGLEMC is the concentration permitted in a limited area of a survey unit; see

34 Section A.8.6.) The average concentration in the area is calculated for comparison against the 

35 DCGLEMC. The EMC is acceptable if the following condition is met, as shown in Equation A-1 

36 (modified from MARSSIM Equation 8-2):

δ

DCGLW

+

average concentration in the elevated area− δ

DCGLEMC

37 < 1 (A-1)

38 Where δ = the average residual radioactivity concentration for all sample points in 

39 the survey unit only. 
A-8

1 If there is more than one elevated area, a separate term should be included for each one. As 

2 an alternative to the unity rule expressed in Equation A-1, the licensee can calculate the dose 

3 from the actual distribution of residual radioactivity, if an appropriate exposure pathway model is 

4 available.

5

6 A.6 Instrument Selection and Calibration

7 To demonstrate that the radiological criteria for license termination have been met, the 

8 measurement instruments should have an adequate sensitivity, be calibrated properly, and be 

9 checked periodically for proper response.

10 A.6.1 Calculation of Minimum Detectable Concentrations

11 The licensee should determine the MDC for the instruments and techniques to be used. The 

12 MDC is the concentration that a specific instrument and technique can be expected to detect 

13 95 percent of the time under actual conditions of use.

14 For scanning building surfaces for beta and gamma emitters, the MDCscan should be determined 

15 from the following equation (obtained by combining MARSSIM Equations 6-8, 6-9, and 6-10 and 

16 using a value recommended in this appendix for the index of sensitivity dʹ of 1.38, which is for 

17 95 percent detection of a concentration equal to MDCscan with a 60 percent false-positive rate).

MDC building surfaces ( )

B

p A t scan

i s

=

270 000 , .  138

 

18 (A-2)

19

20 where MDCscan = minimum detectable concentration for scanning building surfaces in 

picocuries per square meter (pCi/m2 21 )

270,000 = conversion factor to convert to pCi/m2 22

23 1.38 = index of sensitivity dʹ

24 B = number of background counts in time interval t

25 p = surveyor efficiency

26 ε i = instrument efficiency for the emitted radiation

27 ε s = source efficiency in emissions/disintegrations

A = probe’s sensitive area in cm2 28

29 t = time interval of the observation while the probe passes over the source, 

30 in seconds

31 Based on the measurements described in NUREG/CR–6364, “Human Performance in 

32 Radiological Survey Scanning,” issued August 2000, a surveyor efficiency p of 0.5 represents a 

33 mean value for normal field conditions and its use is generally acceptable. If the licensee wants 

34 to determine a value appropriate for particular measurement techniques, the information in 

35 NUREG/CR–6364 describes how to determine the value. NUREG-1507, “Minimum Detectable 

36 Concentrations with Typical Radiation Survey Instruments for Various Contaminants and Field 

37 Conditions,” issued June 1998, contains additional information on the interpretation of results 

38 reported in NUREG/CR–6364.

39 For scanning soil with a sodium iodide gamma detector, the MDCscan values given in Table 6.7 

40 of MARSSIM provide an acceptable estimate of MDCscan. 
A-9

1 For static measurements of surface concentrations, the MDCstatic may be calculated using the 

2 following equation (from Equation 3-10 in NUREG-1507).

MDCstatic =

3+4.65√B

K t

3 (A-3)

where MDCstatic = minimum detectable concentration in pCi/m2 4 or pCi/gram (g)

5 B = background counts during measurement time interval t

6 t = counting time in seconds

7 K = a calibration constant (best estimate) to convert counts/second to 

pCi/m2 8 or pCi/g, discussed further in NUREG-1507

9 Section 6.7.1 of MARSSIM shows an example using this equation.

10

11 The instruments used for sample measurements at the specific sample locations should have 

12 an MDCstatic less than 50 percent of the DCGLW, as recommended in Section 4.7.1 of 

13 MARSSIM. There is no specific recommendation for the MDCscan, but the MDCscan will 

14 determine the number of samples needed, as discussed in Section A.8.6 of this appendix.

15

16 The licensee should record all numerical values measured, even values below the “MDC” or 

17 “critical level,” including values that are negative (when the measured value is below the 

18 average background). Entries for measurement results should not be “nondetect,” “below 

19 MDC,” or similar entries, because the statistical tests can only tolerate a maximum of 40 percent 

20 nondetects.

21

22 A.6.2 Instrument Calibration and Response Checks

23 NRC regulations at 10 CFR 20.1501(c) require that the licensee periodically calibrate radiation 

24 measurement instruments used in surveys such as the FSS. 

25 For in situ gamma measurements, the detector efficiency (count rate per unit fluence rate) 

26 should be determined for the gamma energies of interest and the assumed representative depth 

27 distribution. The surface and volumetric distributions should be explicitly considered to evaluate 

28 potential elevated areas. To calibrate for the representative depth distribution, acceptable 

29 methods are to (1) use a test bed with radioactive sources distributed appropriately or (2) use 

30 primarily theoretical calculations that are normalized or verified experimentally using a source 

31 approximating a point source. The calibration of the source used for the verification source 

32 should be traceable to a recognized standards or calibration organization, for example, the 

33 National Institute of Standards and Technology. 

34

35 Some modern instruments are very stable in their response. Thus, as long as the licensee 

36 periodically performs instrument response checks to verify that the detector is operating 

37 properly, it may be acceptable to calibrate only initially without periodic recalibrations. The initial 

38 calibration may be performed by either the instrument supplier or the licensee, but in either 

39 case, 10 CFR 20.2103(a) requires that a record describing the calibration be available for NRC 

40 inspection.

41

42 A.6.3 Instrument Response Checks

43 Licensees should check the response of survey instruments with a check source each day 

44 before use to confirm constancy in instrument response and establish criteria for acceptable 
A-10

1 response. If the response is not acceptable, the licensee should consider the instrument as not 

2 responding properly and should not use it until the problem has been resolved. Measurements 

3 made after the last acceptable response check should be evaluated and discarded, if 

4 appropriate.

5

6 The check source should emit the same type of radiation (i.e., alpha, beta, gamma) as the 

7 radiation being measured and should give a similar instrument response, but the check source 

8 does not have to use the same radionuclide as the radionuclide being measured.

9

10 A.7 Scanning Coverage Fractions and Investigation Levels

11 Scanning is performed to locate small areas of elevated concentrations of residual radioactivity 

12 to determine whether they meet the radiological criteria for license termination. The licensee 

13 should perform scanning in each survey unit to detect areas of elevated concentrations. The 

14 licensee should establish ILs for investigating significantly elevated concentrations of residual 

15 radioactivity. Table A.2 shows acceptable scanning coverage fractions and scanning ILs for 

16 buildings and land areas. This table is based on MARSSIM Roadmap Tables 2 and 5.8.

17

18 Systematic scans are those conducted according to a preset pattern. Judgmental scans are 

19 those conducted to include areas with a greater potential for residual radioactivity. In Class 2 

20 areas, a 10 percent scanning coverage would be appropriate when there is high confidence that 

21 all locations would be below the DCGLW. A coverage of 25 percent to 50 percent would be 

22 appropriate when there may be locations with concentrations near the DCGLW. A coverage of 

23 100 percent would be appropriate, if there is any concern that the area should have had a 

24 Class 1 classification rather than a Class 2 classification. In Class 3 areas, scanning coverage 

25 is usually less than 10 percent. If any location exceeds the scanning IL, scanning coverage in 

26 the vicinity of that location should be increased to delineate the elevated area.
A-11

1 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 

lower walls of buildings, 10 to 50 % for 

upper walls and ceilings of buildings, 

systematic and judgmental 

>DCGLW or >MDCscan if MDCscan is

greater than DCGLW

3 Judgmental >DCGLW or >MDCscan if MDCscan is

greater than DCGLW

2 Sometimes the sensitivity of static measurements at designated sample points is high enough to 

3 detect significantly elevated areas between sample points. If the sensitivity is high enough, only 

4 this single set of measurements is necessary. For example, both scanning and sampling for 

5 cobalt-60, which emits an easily detectable gamma, can be done with a single set of in situ

6 measurements in some cases.

7

8 A.8 Determining the Number of Samples Needed

9 A minimum number of samples are needed to obtain sufficient statistical confidence that the

10 conclusions drawn from the samples are correct. The method described below from Chapter 5 

11 of MARSSIM is acceptable for determining the number of samples needed.

12

13 A.8.1 Determination of the Relative Shift

14 The required number of samples will depend on a ratio involving the concentration to be 

15 measured relative to the variability in the concentration. The ratio to be used is called the 

16 relative shift, Δ/σs. The relative shift, Δ/σs, is defined in Section 5.5.2.2 of MARSSIM as:





s 

W

s

DCGL LBGR

=

−

17 (A-4)

18

19 where DCGLW = derived concentration guideline

20 LBGR = concentration at the lower bound of the gray region. The LBGR is the 

21 concentration to which the survey unit must be cleaned to have an 

22 acceptable probability of passing the test (i.e., 1−β). 

23 σs = an estimate of the standard deviation of the concentration of residual 

24 radioactivity in the survey unit (which includes real spatial variability in 

25 the concentration as well as the precision of the measurement system)

26 The value of σs is determined either from existing measurements or by taking limited preliminary 

27 measurements of the concentration of the residual radioactivity in the survey unit at about 5 to

28 20 locations, as recommended in Section 5.5.2.2 of MARSSIM. If a reference area will be used

29 and the estimate of the standard deviation in the reference area, σr, is larger than the estimate 
A-12

1 of the standard deviation in the survey unit, σs, then the larger value should be used in the 

2 equation.

3 The LBGR should be set at the mean concentration of residual radioactivity that is estimated to 

4 be present in the survey unit. However, if no other information is available on the survey unit, 

5 the LBGR may be initially set equal to 0.5 DCGLW, as recommended by MARSSIM. If the 

6 relative shift, Δ/σs, exceeds 3, the LBGR should be increased until Δ/σs is equal to 3. The 

7 licensee may refer to Section 5.5.2.2 of MARSSIM for additional details and information.

8 A.8.2 Determination of Acceptable Decision Errors

9 A decision error is the probability of making an error in the decision on a survey unit by failing a 

10 survey unit that should pass or by passing a survey unit that should fail. When using the 

11 statistical tests, larger decision errors may be unavoidable when encountering difficult or 

12 adverse measuring conditions. This is particularly true when trying to measure residual 

13 radioactivity concentrations close to the variability in the concentration of those materials in 

14 natural background.

15 The α decision error is the probability of passing a survey unit where the actual concentration 

16 exceeds the release criterion. A decision error α of 0.05 is acceptable under the more favorable 

17 conditions when the relative shift, Δ/σs, is large (about 3 or greater). Larger values of α may be 

18 considered when the relative shift is small, to avoid an unreasonable number of samples. The 

19 β decision error is the probability of failing a survey unit where the actual concentration is equal 

20 to LBGR. Any value of β is acceptable to the NRC. 

21 A.8.3 Number of Samples Needed for the Wilcoxon Rank Sum Test

22 The minimum number of samples, N, needed in each survey unit for the WRS test may be 

23 determined from the following equation (adapted from MARSSIM Equation 5-1 with N redefined 

24 as the number of samples in the survey unit):

( )

( ) N

Z Z

Pr

= 

+

−

1 − −

2 3 05

1 1

2

2

 

.

25 (A-5)

26

27 where N = the number of samples in the survey unit

28 Z1−α = the percentile represented by the decision error α

29 Z1−β = the percentile represented by the decision error β

30 Pr = the probability that a random measurement from the survey unit exceeds a 

31 random measurement from the background reference area by less than the 

32 DCGLW when the survey unit median is equal to the LBGR concentration 

33 above background

34 1⁄2 = a factor added to MARSSIM Equation 5-1 because N is always defined in this 

35 guide as the number of samples in the survey unit

36 Tables 5.1 and 5.2 of MARSSIM contain values of Pr, Z1−α, and Z1−β. N is the minimum number 

37 of samples necessary in each survey unit. An additional N samples will also be needed in the 

38 reference area. If N is not an integer, the number of samples is determined by rounding up. In 

39 addition, the licensee should consider taking some additional samples (MARSSIM recommends 

40 20 percent) to protect against the possibility of lost or unusable data. Fewer samples increase 
A-13

1

2

3

4

5

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

6 the following equation (adapted from MARSSIM Equation 5-2): 

( )

( ) N

Z Z

Signp

=

+

−

1 1 − −

2

2

4 05

 

.

7 (A-6)

8

9 where N = number of samples needed in a survey unit

10 Z1−α = percentile represented by the decision error α

11 Z1−β = percentile represented by the decision error β

12 Sign p = estimated probability that a random measurement for the survey unit will be 

13 less than the DCGLW when the survey unit median concentration is actually 

14 at the LBGR.

15 Tables 5.2 and 5.4 of MARSSIM contain the values of Z1−α, Z1−β, and Sign p. In addition, the 

16 licensee should consider taking some additional samples (MARSSIM recommends 20 percent)

17 to protect against the possibility of lost or unusable data. Fewer samples increase the 

18 probability of an acceptable survey unit failing to demonstrate compliance with the radiological 

19 criteria for release. If a survey unit fails to demonstrate compliance because there were not 

20 enough samples taken, a totally new sampling effort may be needed unless resampling was 

21 anticipated.

22 A.8.5 Use of Two-Stage or Double Sampling

23 It may be desirable for a licensee to sample a survey unit a second time to determine 

24 compliance. “Two-stage sampling” and “double sampling” are two methods by which additional 

25 survey unit data can be acquired. Two-stage sampling refers to survey designs specifically 

26 intended to be conducted in two stages. Double sampling refers to the case when the survey 

27 unit design is a one-stage design, but allowance is made for a second set of samples to be 

28 taken if the retrospective power of the test using the first set of samples does not meet the 

29 design objectives. Use of either method should be considered as part of the DQO process 

30 when developing the design of the FSS. Appendix C of this volume contains information on the 

31 use of two-stage or double sampling.

32 A.8.6 Additional Samples for Elevated Measurement Comparison in Class 1 Areas

33 Additional samples may be needed when the concentration that can be detected by scanning, 

34 MDCscan, is larger than the DCGLW. The licensee should determine whether additional samples 

35 are needed in Class 1 survey units for the EMC when the concentration that can be detected by 

36 scanning, MDCscan, is larger than the DCGLW. The method in Section 5.5.2.4 of MARSSIM to 

37 determine whether additional samples are needed is acceptable to the NRC staff and is 

38 described here.

39

40 The area factor is the multiple of the DCGLW that is permitted in a limited portion of the survey 

41 unit. In Equation A-7, the ratio of the MDCscan to the DCGLW establishes the area factor (the 
A-14

1 multiple of the DCGLW) that can be detected by scanning (adapted from MARSSIM 

2 Equation 5-4):

area factor MDC

DCGL

scan

W

3

=

(A-7)

4

5 Using the methods in NUREG-1549, “Decision Methods for Dose Assessment To Comply with 

6 Radiological Criteria for License Termination, Draft Report for Comment,” issued July 1998, the 

7 licensee can determine the size of the area corresponding to the area factor, AEC. The number 

8 of sample points that may be needed to detect this area of elevated measurement 

9 concentration, NEMC, in a survey unit is:

10

N

A

A

EMC

EC

11

=

(A-8)

12

13 where A = the area of the survey unit

14 AEC = the area of concentration greater than DCGLW

15 If NEMC is larger than N, additional samples may be needed to demonstrate that areas of 

16 elevated concentrations meet the radiological criteria for license termination. However, the 

17 number of samples needed is not necessarily NEMC. The licensee can use the HSA and site 

18 characterization to determine how many additional samples may be needed. Based on what is 

19 known about the site, it may be possible to estimate a concentration that is unlikely to be 

20 exceeded. If there is a maximum concentration, the size of the area corresponding to this area 

21 factor for this concentration may be used for AEC in Equation A-8. Similarly, based on 

22 knowledge of how the radioactive material was handled or dispersed on the site, it may be 

23 possible to estimate the smallest area likely to have elevated concentrations. If this is so, that 

24 area can be used in Equation A-8. Likewise, the licensee could use the knowledge of how the 

25 residual radioactivity would be likely to spread or diffuse after deposition to determine an area 

26 AEC for Equation A-8. 

27

28 Figure D-7 of Appendix D to MARSSIM and Section 3.7.2 of NUREG-1505 show that a 

29 triangular grid is slightly more effective in locating areas of elevated concentrations. Therefore, 

30 a triangular grid generally should be used if NEMC is significantly larger than N and if areas 

31 similar in size or smaller than the grid spacing are expected to have concentrations at or above 

32 the area factor.

33

34 A.9 Determining Sample Locations

35 For the impacted areas, the licensee should establish a reference coordinate system, which is a 

36 set of intersecting lines referenced to a fixed site location or benchmark. Reference coordinate 

37 systems are established so that the locations of any point in the survey unit can be identified by 

38 coordinate numbers. A reference coordinate system does not establish the number of sample 

39 points or determine where samples are taken. A single reference coordinate system may be 

40 used for a site, or different coordinate systems may be used for each survey unit or for a group 

41 of survey units. Section 4.8.5 of MARSSIM describes an acceptable method to establish a 

42 reference coordinate system.
A-15

1 In Class 1 and Class 2 areas, the sampling locations are established in a regular pattern, either 

2 square or triangular. The method described below is from Section 5.5.2.5 of MARSSIM. 

3 After the number of samples needed in the survey unit has been determined, and the licensee 

4 has decided whether to use a square or triangular grid, sample spacings, L, are determined 

5 from Equations A-9 and A-10 (adapted from MARSSIM Equations 5-5, 5-6, 5-7, and 5-8).

L

A

N

= for a triangle grid

0.866

6 (A-9)

7

L

A

N

8

= for a square grid

(A-10)

9

10 where A = the survey unit area

11 N = the number of samples needed (in Class 1 areas, the larger of the number for 

12 the statistical test or the EMC).

13 The calculated value of L is then often rounded downward to a shorter distance that is easily 

14 measured in the field. 

15 A random starting point should be identified for the survey pattern. The coordinate location of 

16 the random starting point should be determined by a set of two random numbers, with one 

17 representing the x axis and the other, the y axis. The random numbers can be generated by 

18 calculator or computer or can be obtained from a table of random numbers. Each random 

19 number should be multiplied by the appropriate survey unit dimension to provide a coordinate 

20 relative to the origin of the survey unit reference coordinate system. 

21

22 Beginning at the random starting point, a row of points should be identified parallel to the x axis 

23 at intervals of L. For a square grid, the additional rows should be parallel to the first row at a 

24 distance of L from the first row. For a triangular grid, the distance between rows should be 

25 0.866 L, and the sample locations in the adjacent rows should be midway on the x axis between 

26 the sample locations in the first row. Sample locations selected in this manner that either do not 

27 fall within the survey unit area or cannot be surveyed because of site conditions should be 

28 replaced with other sample locations determined using the same random selection process that 

29 was used to select the starting point. MARSSIM Figure 5.5 contains an example illustrating the 

30 triangular grid pattern.

31

32 In Class 3 survey units and in reference areas, all samples should be taken at random locations. 

33 Each sample location should be determined by a set of two random numbers, one representing 

34 the x axis and the other the y axis. Each set of random numbers should be multiplied by the 

35 appropriate survey unit dimension to provide coordinates relative to the origin of the survey unit 

36 reference coordinate system. Coordinates identified in this manner that do not fall within the 

37 survey unit area or that cannot be surveyed because of site conditions should be replaced with 

38 other sample locations determined in the same manner. MARSSIM Figure 5.4 illustrates a 

39 random sample location pattern.

40
A-16

1

2

3

4

5

6

7

8

9

10

11

12

13

14

15

16

17

18

19

20

21

22

23

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 

24

Measurement Results Conclusion

Difference between maximum survey unit concentration and 

minimum reference area concentration is less than DCGLW.

Survey unit meets release 

criterion.

Difference between survey unit average concentration and 

reference area average concentration is greater than 

DCGLW.

Survey unit fails.

Difference between any survey unit concentration and any 

reference area concentration is greater than DCGLW and the 

difference of survey unit average concentration and 

reference area average concentration is less than DCGLW.

Conduct WRS test and EMC.

25
A-17

1 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 

concentration less than DCGLW.

Conduct Sign test and EMC.

2 Some facilities may have residual radioactivity composed of more than one radionuclide. When 

3 there are multiple radionuclides rather than a single radionuclide, the licensee should consider 

4 the dose contribution from each radionuclide. Section 2.7 of this volume contains information 

5 about using the sum of fractions approach for compliance when multiple radionuclides are 

6 present. 

7 When there is a fixed ratio among the concentrations of the nuclides, a DCGLW for each nuclide 

8 can be calculated. Compliance with the radiological criteria for license termination may be 

9 demonstrated by comparing the concentration of the single surrogate radionuclide that is 

10 easiest to measure with its DCGLW (which has been modified to account for the other 

11 radionuclides present). For example, if Cs-137 and Sr-90 are present, using measured 

12 concentrations of Cs-137 as a surrogate for the mix of Cs-137 and Sr-90 may be simpler than 

13 separately measuring Cs-137 and Sr-90, and may thus save labor and analytical expenses. 

14 When using a surrogate radionuclide to represent the presence of other radionuclides, a 

15 sufficient number of measurements, spatially distributed throughout the survey unit, should be 

16 used to establish a consistent ratio between the surrogate and the other radionuclides. 

17 Section 4.3.2 of MARSSIM provides additional information on the use of surrogate radionuclides 

18 for surveys.

19

20 When there is no fixed ratio among the concentrations of the nuclides, the licensee must 

21 evaluate the concentration of each nuclide via sampling. Compliance with the radiological 

22 criteria for license termination is then demonstrated by considering the sum of the concentration 

23 of each nuclide relative to its DCGLW, followed by an evaluation of all radionuclides of concern 

24 via the unity rule. Chapter 11 of NUREG-1505 describes an acceptable method for performing 

25 the evaluation. When there is no fixed ratio among the concentrations of the nuclides, and a 

26 large number of discrete samples are required, it may be possible to utilize a composite 

27 sampling strategy to increase the probability of elevated area detection and as a means to 

28 reduce analytical cost. However, this approach requires an evaluation of the exposure scenario 

29 related to the hard-to-detect radionuclide and should be performed on a case-by-case basis 

30 along with discussions with the regulator. Additional information on composite sampling can be 

31 found in Appendix O of this volume. 

32 In some cases in which multiple nuclides are present with no fixed ratio in their concentrations, 

33 the dose contribution from one or more of the nuclides in the mixture will dominate the total 

34 dose, and the dose from other radionuclides will be insignificant. For example, at a nuclear 

35 power plant, many different radionuclides could be present with no fixed ratio in their 
A-18

1 concentrations, but almost all of the dose would come from just one or two of the nuclides. 

2 Section 3.3 of this volume contains guidance on elimination of radionuclides or pathways from 

3 consideration. 

4 If a survey unit fails, the licensee should evaluate the measurement results and determine why it 

5 failed. MARSSIM, in Sections 8.2.2 and 8.5.3, and in Appendix D, provides acceptable 

6 methods for reviewing measurement results. If it appears that the failure was caused by the 

7 presence of residual radioactivity in excess of that permitted by the radiological release criteria, 

8 the survey unit should be re-remediated and resurveyed. However, some failures may not be 

9 caused by the presence of residual radioactivity. If it can be determined that this is the case, 

10 the survey unit may be released.

11 A.11 References

12 Nuclear Regulatory Commission (U.S.) (NRC). “Measurement Methods for Radiological 

13 Surveys in Support of New Decommissioning Criteria, Draft Report for Comment,”

14 NUREG-1506. NRC: Washington, DC. August 1995.

15

16 — — — — —. “Decision Methods for Dose Assessment to Comply with Radiological Criteria for 

17 License Termination, Draft Report for Comment,” NUREG-1549. NRC: Washington, DC. 

18 1998a.

19 — — — — —. “Human Performance in Radiological Survey Scanning,” NUREG/CR-6364. 

20 NRC: Washington, DC. 1998b.

21

22 — — — — —. “Minimum Detectable Concentrations with Typical Radiation Survey Instruments 

23 for Various Contaminants and Field Conditions,” NUREG-1507. NRC: Washington, DC. 

24 1998c.

25 — — — — —. “A Proposed Nonparametric Statistical Methodology for the Design and Analysis 

26 of Final Status Decommissioning Surveys–Interim Draft Report for Comment and Use,”

27 NUREG-1505, Rev. 1. NRC: Washington, DC. 1998d.

28

29 — — — —. “Multi-Agency Radiation Survey and Site Investigation Manual (MARSSIM),”

30 NUREG-1575, Rev. 1. EPA 402-R-97-016, Rev. 1; DOE/EH-0624, Rev. 1; U.S. Department of 

31 Defense; U.S. Department of Energy; U.S. Environmental Protection Agency; and NRC: 

32 Washington, DC. August 2000. 

33 — — — — —. Regulatory Issue Summary 2002–02, “Lessons Learned Related to Recently 

34 Submitted Decommissioning Plans and License Termination Plans.” NRC: Washington, DC. 

35 January 2002.
SIMPLE APPROACHES FOR CONDUCTING FINAL RADIOLOGICAL 

SURVEYS

B-1

1 B.1 Introduction

2 A large number of licensees may use a simplified method to demonstrate regulatory compliance 

3 for decommissioning, and thereby avoid complex FSSs. For Decommissioning Groups 1–3, 

4 licensees may use the simplified FSS method described in Appendix B to NUREG-1575, “Multi5 Agency Radiological Survey and Site Investigation Manual” (MARSSIM), Revision 1, issued 

6 August 2000, or the alternative protocol described in this appendix (Section B.3 below).

7

8 B.2 MARSSIM Simplified Method 

9 The simplified method in Appendix B of MARSSIM may be used by Decommissioning Group 1 

10 and some Decommissioning Group 2 licensees. These are sites where radioactive materials 

11 have been used or stored only in the form of (1) nonleaking, sealed sources, (2) short half-life 

12 radioactive materials (e.g., T1/2 ≤120 days) that have since decayed to insignificant quantities, 

13 (3) small quantities exempted or not requiring a specific license, or (4) a combination of the

14 above. MARSSIM, Revision 1, Appendix B gives the details of this simplified method.

15 B.3 Alternative Simplified Method 

16 This alternative method may be used by Decommissioning Groups 1–3 and is applicable only 

17 for surfaces of building structures and for surface soils. The following conditions are 

18 prerequisite to the use of this method:

19

20

21

22

23

24

25

26

27

28

29

30

31

• 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, “Minimum Detectable Concentrations with Typical Radiation Survey

Instruments for Various Contaminants and Field Conditions”)

1 32

33 If the above conditions are met, then the following simplified method may be used to design and 

34 conduct the FSS for each survey unit.

• Size is limited to 2000 m2for land areas and 100 m2 35 for structures.

1 Revision 1 of NUREG-1507 was issued in August 2020. 
B-2

1 • Scanning and sampling are to be performed as follows:

2 o 100-percent scan

3 o 30 samples

4 • The hot spot criterion is three times the DCGLW, applied to any sampling location.

5 • A quality control program ensures results are accurate and sources of uncertainty are

6 identified and controlled.

7 • The average concentration for the survey unit is compared to the DCGLW.

8 • Statistical tests that may be used include the parametric Student’s t-test, or non9 parametric WRS test when the radionuclide(s) of concern are in background, or the non10 parametric Sign test when the radionuclide(s) of concern are not in background

11 assuming an alpha (α) or false positive error of 5 percent. No statistical tests are

12 needed if all measurements are less than the DCGLW. MARSSIM (NUREG-1575) can

13 be consulted for additional information on the statistical tests.

14 The FSSR should provide a complete and unambiguous record of the radiological status of the 

15 site and should stand on its own with minimal information incorporated by reference (see 

16 Appendix D of this volume for additional information on reporting survey results).

17 B.4 References 

18

19

20

21

Nuclear Regulatory Commission (U.S.) (NRC). “Minimum Detectable Concentrations with 

Typical Radiation Survey Instruments for Various Contaminants and Field Conditions,”

NUREG-1507. NRC: Washington, DC. 1998.

— — — — —. “Multi-Agency Radiation Survey and Site Investigation Manual,” (MARSSIM), 

22 NUREG-1575. NRC: Washington, DC. 2000. 
USE OF TWO-STAGE OR DOUBLE SAMPLING FOR FINAL STATUS

SURVEYS

C-1

1 C.1 Introduction

2 This appendix contains information on the assessment of survey data when double or two-stage 

3 sampling in a survey unit is used to determine compliance. These sampling strategies utilize 

4 the initial sample data and a second round of supplemental sampling. This approach may be 

5 desirable when a survey unit fails the hypothesis test (i.e., the decision is made that the survey 

6 unit does not meet the release criterion) due to insufficient power as described in more detail 

below,

1 7 but the mean of the measured data is below the release criterion. Further information

8 on survey unit failures and possible remedies using the DQO process are discussed in Chapter 

9 8 of NUREG-1575, “Multi-Agency Radiological Survey and Site Investigation Manual” 

10 (MARSSIM), Revision 1, issued August 2000. 

11

12 In this appendix two-stage sampling denotes survey designs specifically intended to be 

13 conducted in two stages, whereas double sampling refers to one-stage survey designs that 

14 allow for the collection of a second set of samples if the retrospective power of the test using the 

15 first set of samples does not meet the design objectives. Double or two -stage sampling may be 

16 acceptable if it is incorporated into the final status survey design. Thus, before the initial round 

17 of sampling occurs, allowances for these sampling approaches should be discussed in the 

18 DQOs for the final status survey.

19

20 In general, adequate initial sampling to achieve the desired statistical power and error rates, or 

21 data collection in two stages, are both preferable to double sampling. As discussed in the next 

22 sections of this appendix, there are several factors to consider when additional random samples 

23 are collected. For example, double sampling is also generally not appropriate for Class 1 

24 survey units having confirmed areas of elevated activity, or Class 2 or Class 3 survey units 

25 because the need for a second set of samples raises the issue of survey unit misclassification. 

26

27 C.2 Double Sampling 

28 As noted in the preceding section C.1, situations can occur where a survey unit might have 

29 passed the final status statistical test, had the initial sampling design been powerful enough to 

30 reject the null hypotheses in Scenario A. That is, a retrospective examination of the power of 

31 the statistical tests reveals that the probability of detecting that the survey unit actually meets 

32 the release criterion was lower than that planned for during the DQO process. This could occur 

33 if the spatial variability in residual radioactivity concentrations was larger than anticipated. The 

34 power of the test specified during the DQO process depends on an estimate of the uncertainty. 

35 The power of the statistical test will be less than planned if the standard deviation is higher than 

36 expected. If samples were lost, did not pass analytical QA/QC, or are otherwise unavailable for 

37 inclusion in the analysis, the power will also be lower than was planned. In these situations, it 

38 might be desirable to take additional samples in the survey unit to improve the power of the 

39 statistical test. 

40 Draft NUREG/CR-5849, “Manual for Conducting Radiological Surveys in Support of License 

41 Termination,” issued June 1992, allowed the licensee to take additional samples in a survey unit 

42 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-β)). 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-2

1 confidence level on the mean was greater than the DCGLW. Because a 95-percent confidence 

2 interval is constructed using Student’s t-statistic rather than using a hypothesis test, the survey 

3 design does not consider Type II errors. The second set of samples was taken so that a 

4 Student’s t-test on the combined set of samples would have 90-percent power at the mean of 

5 the first set of samples, given the estimated standard deviation from the first set of samples. 

6 Such double sampling was to be performed only once.

7 Increasing the probability that a clean survey unit passes (power in Scenario A) using double 

8 sampling will also tend to increase the probability that a survey unit that is not clean will pass 

9 (Type I error). In addition, the two tests are not independent, because the data from the first set 

10 of samples is used in both. The increase in the Type I error rate is expected to be less than a 

11 factor of two based on the analysis in the following paragraphs of this appendix. The potential 

12 increase in the Type I error rate should be considered when designing the survey and 

13 determining acceptable error rates as part of the DQO process.

14 Two-stage or double sampling is not usually expected (nor is it encouraged) when the DQO 

15 process is used, as in MARSSIM, Revision 1. This is because the Type II error and the power 

16 desired are explicitly considered in the survey design process. If higher power in the test is 

17 desired, it should be specified as such. Sufficient samples should be taken to achieve the 

18 specified power. The value of this approach lies in the greater objectivity and defensibility of the 

19 decision made using the data. Nonetheless, it is recognized that there may be instances when 

20 some sort of double sampling is considered desirable. As discussed above, an example is

21 when it is difficult to estimate the standard deviation of the concentrations in a survey unit. A 

22 first set of data may be taken with an estimated standard deviation that is too low, and thus, the 

23 power specified in the DQO process may not be achieved. 

24 For planning purposes, it is important to understand the circumstances under which it is 

25 appropriate to combine this data with additional sample data to be used in the test of the final 

26 status survey. Consider the Sign test, as indicated in NUREG-1505, “A Proposed 

27 Nonparametric Statistical Methodology for the Design and Analysis of Final Status 

28 Decommissioning Surveys–Interim Draft Report for Comment and Use,” issued June 1998. 

29 Suppose N1 samples are taken. For the Sign test in Scenario A, the test statistic, S1, is equal to 

30 the number of survey unit measurements below the DCGLW. If S1 exceeds the critical value k1, 

31 then the null hypothesis that the median concentration in the survey unit exceeds the DCGLW is 

32 rejected (i.e., the survey unit passes this test). The probability that any single survey unit 

33 measurement falls below the DCGLW is found from 

p(C) = ∫ f(x)dx =

1

√2πσ

∫ e

−(x−C)

2

2σ2

DCGLW

−∞

dx =

DCGLW

−∞

Φ (

DCGLW−C

σ

34 ) (C-1)

35

36 C is the true, but unknown, mean concentration in the survey unit. Assuming the data are 

37 normally distributed, when C = DCGLW, then p = 0.5. More generally, if C is the true, but 

38 unknown, median concentration in the survey unit, p is also equal to 0.5.

39

40 The probability that more than k1 of the N1 survey unit measurements fall below the DCGLW is 

41 simply the following binomial probability:
C-3

 ( ) ( )

t k

N

t

N t

t

k

t

N N t

t

p p

N

t

p p

= +

−

=

 −









 − = −











  −

1

1

1

1

1

1

1 1

0

1

1 1 1

(C-2)

2

3 This is the probability that the null hypothesis will be rejected, and it will be concluded that the 

4 survey unit meets the release criterion. When the median concentration in the survey unit is at 

5 the DCGLW, this probability is just the Type I error rate, α. When C = DCGLW, p = (1− p) = 0.5, 

6 so

7

α = ∑ (

N1

t

)(0.5)

t

(0.5)

N1−t = (0.5)N1 ∑ (

N1

t

)

N1

t=k1+1

N1

t=k1+1

8 (C-3)

9

10

11 Suppose it is decided to allow the licensee to take a second set of samples of size N2. The test 

12 statistic, S, is equal to the number of the total of N = N1 + N2 survey unit measurements below 

13 the DCGLW. If S exceeds the critical value k, then the null hypothesis that the median 

14 concentration in the survey unit exceeds the DCGLW is rejected (i.e., the survey unit passes this 

15 test). In this case, the overall probability that the null hypothesis is rejected (i.e., the survey unit 

16 passes) is equal to the sum of the probabilities of the two events, labeled (Event 1) and (Event 

17 2) below, that are mutually exclusive:

18

19 (Event 1) the probability that more than k1 of the N1 survey unit measurements fall below 

20 the DCGLW, and

21 (Event 2) the probability that fewer than k1 of the first N1 survey unit measurements fall 

22 below the DCGLW but that more than k of the N total survey unit measurements fall below 

23 the DCGLW.

24 The test statistic, S, is then equal to the sum of S1 and S2, (S = S1 + S2), where S2 is the number 

25 of the second set of N2 survey unit measurements that fall below the DCGLW. S1 and S2 are 

26 independent, but S1 and S = S1 + S2 are not.

27 The covariance of S1 and S using E() to denote expected values, is 

Cov S S E S S E S E S

E S S S E S E S

E S E S S E S E S

N p p N p N N p N p N N p

N p p

( , ) ( , ) ( ) ( )

( ( )) ( ) ( )

( ) ( ) ( ) ( )

( ( ) ) ( )

( )

1 1 1

1 1 2 1

2

1 2 1

1

2

1

2 2

1 2

2

1 1 2

1

1

1

1

= −

= + −

= + −

= − + + − +

= −

28 (C-4)

29
C-4

1 Therefore, the correlation coefficient between S1 and S is

ρ(S1,S) =

N1p(1−p)

√N1p(1−p)(N1+N2

)p(1−p)

2

3

=

N1

√N1

(N1+N2)

4

5 (C-5)

= √

N1

(N1 + N2

6

7

= √

N1

N

8

9

10 To calculate the overall probability that the survey unit passes, one requires the joint probability 

11 of S1 and S,

12

13

( ) ( ) ( )

( ) ( ) ( )

( )

Pr S s S s S s S s s , Pr Pr

N

s

p p

N

s s

p p

N

s

N

s s

p p

s

N s s s N s s

s N s

1 1 1 1 2 1

1

1

2

1

1

1

2

1

1

1 1 12 1 1 1

1

= = = = = −

=











 −

−











 −

=













−











 −

− −− −

−

14 (C-6)

15 Therefore, the overall probability that the survey unit passes is

16

Pr( ) Pr()Pr( )

11 11 11

17

kS o r  =  +  kSandkSkSkS

)()(

2

1

1

1

1 1

1

21 2121

11 12

1 11

1

11

)1(

)1(

ss ssNN

ks sks

s sN

N

ks

pp

s

N

s

N

pp

s

N

+ +−+

  −

−

+=

− 































+

− 















=

 



18 (C-7)

19

20

21 The first term is equal to (or slightly less than) the Type I error rate α specified during the DQO 

22 process. The second term is the additional probability of a Type I error introduced by allowing 

23 double sampling.
C-5

1 Note that

( )

=   















−=

















−

















 −

















−

















 −=

−



 =

−

 =

−



 

 

)1( )Pr(

)1(

Pr )1(

1

2

0 1

1

1

2

0 1

1

11

1

1

1

kS

s

N

pp

sk

N

s

N

pp

sk

N

s

N

kS and ppkS

sN

N

ks

s

N

ks

k

s

s SN

N

ks

k

s

s sN

2 (C-8)

3

4 Thus, the Type I error rate would be at most doubled when double sampling is allowed.

5

6 For example, if a survey is designed so that N1 = 30, and α = 0.05, then the critical value for the 

7 Sign test is k1 = 19. Suppose the first survey results in 19 or fewer measurements that are less 

8 than the DCGLW. In addition, suppose the survey unit is sampled again, taking an additional 

9 N2 = 30 samples. Then the total number of samples is N = N1 + N2 = 60. The critical value for 

10 the Sign test with α = 0.05 and N = 60 is k = 36. When the survey unit concentration is equal to 

11 the DCGLW, p = 0.5, one has

12

( ) ( ) ( )

( ) ( )

( )

( ) ( )( ) ( )

Pr Pr Pr

. .

. .

. . .

S or S S S and S

s

s s

s

s s

s s s

s s s s

1 1 1

1

20

30 30

1

0

19

2

37

30 30 30

19 36 19 19 36

30

05 1 05

30 30

05 1 05

0 049 0 027 0 076

1

1 1

1 2 1

1 2 1 2

  =  +  

=











 −

+























 −

= + =

=

−

= = −

+ + − +



 

13 (C-9)

14

15 Thus, the total Type I error rate is about 50 percent greater than originally specified.

16

17 In conclusion, double sampling should not be used as a substitute for adequate planning. If it is 

18 to be used, it should be considered part of the DQO process. The procedure for double 

19 sampling (i.e., the size of the second set of samples, N2,) should be specified, recognizing that 

20 the Type I error rate could be up to twice that specified for the Sign test when only one set of 

21 samples is taken. 

22

23 Similar considerations apply for the WRS test; however, the calculation of the exact effect on 

24 the Type I error rate is considerably more complex and is not discussed in this appendix.

25

26 Finally, double sampling should never be necessary for Class 2 or Class 3 surveys, which are 

27 not expected to have concentrations above the DCGLW. These classes of survey unit should 

28 always pass after the first set of samples, because every measurement should be below the

29 DCGLW. The need for a second set of samples (i.e., failure to reject the null hypothesis) in 

30 Class 2 or Class 3 survey units would raise an issue of survey unit misclassification. In addition, 

31 double sampling is generally not appropriate for Class 1 survey units where elevated areas 

32 have been found.
C-6

1 In lieu of double sampling, a preferred approach is to plan for data collection in two stages and 

2 design the final status survey accordingly, as is discussed in the remainder of this appendix.

3

4 C.3 Two-Stage Sequential Sampling

5 In some cases, two-stage sampling may be used instead of a single-stage sample design. For 

6 example, if there are a large number of survey units of a similar type to be tested, a two-stage 

7 sampling procedure may result in substantial savings of time and money by reducing the 

8 average number of samples required to achieve a given level of statistical power.

9 An example of a two-stage sampling design using the Sign test is summarized here. In this 

10 example, N1 is the size of the first set of samples taken, and S1 is the number of these samples 

11 that are less than the DCGLW. Similarly, N2 is the size of the second set of samples taken, and 

12 S2 is the number of these samples that are less than the DCGLW. Let N = N1 + N2, and let 

13 S = S1 + S2. The procedure is as follows:

14 • If S1 > u1 then the survey unit passes (reject H0).

15 • If S1 < l1 then the survey unit fails.

16 • If l1 ≤ S1 ≤ u1 then the second set of samples is taken.

17 • If S = S1 + S2 > u2 after the second set of samples is analyzed, then the survey unit

18 passes.

19 An advantage of two-stage sampling is that it can reduce the total number of samples if there 

20 are many similar survey units of similar design. For given error rates α and β, the number of 

21 samples, N1, taken in the survey unit during the first stage of sampling will be less than the 

22 number, N0, required in the MARSSIM, Revision 1, tables. Unless the result is “too close to 

23 call,” this will be the only sampling needed. When the result is “too close to call,” l1 ≤ S1 ≤ u1, a 

24 second sample of size N2 is taken, and the test statistic S2 is computed using the combined data 

25 set, N1 + N2. While the size of the combined set, N = N1 + N2, will generally be larger than the 

26 number, N0, from the MARSSIM, Revision 1, tables, the expected sample size over many 

27 survey units will still be lower. Thus, a two-stage sampling scheme will be especially useful 

28 when there are many similar survey units for which the FSS design is essentially the same. 

29 Two-stage sampling may be used whether or not a reference area is needed (i.e., it may be 

30 used with either the Sign or the WRS test).

31 The remaining major issue is how to choose the critical values l1, u1, and u2. Hewett and 

32 Spurrier (1983) suggest three criteria:

33 (1) Match the power curve of the two-stage test to that of the one-stage test. The curves

34 are matched at three points. The points with power equal to α, 1−β, and 0.5 are

35 generally well enough separated to ensure a good match over the entire range of

36 potential survey unit concentrations.

37 (2) Maximize the power at the LBGR for given values of α and average sample size.

38 (3) Minimize the sample size for given values of α, and 1−β.
C-7

1 While any one of these criteria could be used, the first has received more attention in the 

2 literature. Thus, it may be more readily applied to the case of final status survey design. The 

3 other criteria would require further development.

4 Spurrier and Hewett (1975) initially developed a two-stage sampling methodology using 

5 criteria 1 assuming the data are normally distributed. They matched the two-stage power curve

6 to the one-stage power curve at three points: the first at α values of either 0.05 or 0.01; the 

7 second in the gray region where power is equal to 0.5; and the third at the lower bound of the 

8 gray region with a β value of 0.1 and power (1-β value) of 0.9. Table C.1 shows the values of l1, 

9 u1, and u2 obtained for six different sets of sample sizes, N1/N0, N2/N0, expressed as fractions of 

10 the sample size, N0, which would be required for the one-stage test with equivalent power. The 

11 term E(N)/N0, is the maximum expected combined sample size for the two-stage test relative to 

12 the sample size, N0, which would be required for the one-stage test with equivalent power. This 

13 number is almost always less than one, but it depends on how close the actual concentration in 

14 the survey unit is to the DCGLW. Clearly, if the concentration is over the DCGLW, the survey unit 

15 is likely to fail on the first set of samples. If the concentration is much lower than the DCGLW, 

16 the survey unit is likely to pass on the first set of samples. It is only when the true concentration 

17 in the survey unit falls within the gray region that there will be much need for the second set of 

18 samples. The fact that the maximum E(N)/N0 is almost always less than one indicates that the 

19 overall number of samples required for a two-stage FSS will almost never exceed the number 

20 required for a one-stage test, even if the true concentration of the survey unit falls in the gray 

21 region between the LBGR and the DCGLW.

22 The power to distinguish clean from dirty survey units is relatively low when the true 

23 concentration is in the gray region because the power falls from 1−β at the LBGR to α at the 

24 DCGLW. Thus, when the true concentration is in the gray region, a larger number of cases will 

25 require a second set of samples. The gray region is exactly where the results are “too close to 

26 call.” However, if the true concentration of the survey unit is below the LBGR or above the 

27 DCGLW, the actual average number of samples will be closer to N1, because the second set of 

28 samples will seldom be needed.

29 In 1976, Spurrier and Hewett dropped the assumption of normality and extended their 

30 methodology to two-stage Wilcoxon Signed Rank and WRS tests. The procedure depends on 

31 an extension of the Central Limit Theorem to the joint distribution of the test statistics S1 and S = 

32 S1 + S2. Spurrier and Hewett suggest that the approximation works reasonably well for sample 

33 sizes as small as nine.

34 In this appendix, their method is also applied to the Sign test.

35 For the Sign test, one computes

S

S N

N

1

1 1

1

2

4

=

−

+

36 (C-10)

where S1

+ 37 is the usual Sign test statistic (i.e., the number of measurements less than the

38 DCGLW).

39 Using Table C.1, 

40 • if S1 > u1 then reject the null hypothesis (the survey unit passes),
C-8

1 • if S1 < l1 then do not reject the null hypothesis (the survey unit fails),

2 • if l1 ≤ S1 ≤ u1 then take the second set of samples.

3 If a second set of samples is taken, then compute

4

5

S

S S N N

N N

S N

N

=

+ − +

+

=

−

+ + +

( ) ( )

( )

1 2 1 2

1 2

2

4

2

4

6 (C-11)

7

8 Using Table C.1,

9 • if S > u2 then reject the null hypothesis (the survey unit passes),

10 • if S ≤ u2 then do not reject the null hypothesis (the survey unit fails).

11 This test relies on “a large sample approximation.” That is, one is assuming that the sample 

12 size is large enough that the joint distribution of S1 and S is bivariate standard normal with 

13 correlation coefficient

ρ(S1, S) = √

N1

N

14 (C-12)

15 Some simulation studies may be done to determine quantitative bounds on the accuracy of this 

16 approximation.

17 The choice of which set of sample sizes should be used is dependent on how confident one is 

18 of passing.

19 For Class 2 and Class 3 survey units (discussed in Appendix A of this volume), case 3 with 

20 N1/N0 = 0.2 and N2/N0 = 1.0 might be reasonable. In these classes of survey units, no individual 

21 sample concentrations in excess of the DCGLW are expected. The probability of passing on the 

22 first set of samples should be close to one. Therefore, it makes sense to choose a design with 

23 the minimum number of samples required in the first set. 

24 For Class 1 survey units (discussed in Appendix A of this volume), case 2 with N1/N0 = 0.4 and 

25 N2/N0 = 0.8 might be more appropriate. There is some chance that the survey unit will not pass 

26 on the first set of samples, so it may be desirable to reduce Max E(N)/N0 from 0.999 to 0.907 by 

27 taking more samples in the first set.

28 If the gray region has been expanded to increase Δ/σ, case 1 or 4 would be a more 

29 conservative choice. In this situation, statistical power has been compromised somewhat, so it 

30 may be important to reduce the risk of having a larger average total number of samples (as 

31 indicated by the potential Max E(N)/N0 even further).

32 Scan sensitivity will also affect the ability to use two-stage designs in Class 1 survey units. It 

33 would have to be determined if the DCGLEMC can be detected when only N1 samples are taken. 

34 If not, the sample size would have to be increased until the MDCscan is lower than the DCGLEMC. 
C-9

1 In this situation, the choice of N1, and the average savings possible with two-stage sampling 

2 may be severely limited.

3 Table C.1 Critical Points for Two-Stage Test of Normal Mean for a One-Sided Alternative 

α = 0.05 α = 0.01

N1/N0 N2/N0 u1 l1 u2 Max

E(N)/N0

u1 l1 u2 Max

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

4 Source: Spurrier and Hewett (1975).

5 For the WRS test, at each stage, one sets the number of measurements required in the survey 

6 unit, n1 and n2, and in the reference area m1 and m2, relative to the number required for the 

7 one-stage test n0 = m0 = N0/2 specified in Table 5.3 of MARSSIM, Revision 1. There is an 

8 additional requirement that n1/n2 = m1/m2, which should be satisfied with sufficient accuracy for 

9 most MARSSIM, Revision 1 designs. Minor departures due to small differences in sample size 

10 caused by filling out systematic grids or the loss of a few samples should not severely affect the 

11 results. 

12 One now computes

( )

( )

S

W m n m

n m n m

R

1

1 1 1 1

1 1 1 1

1 2

1 12

=

− + +

+ +

/

/

13 (C-13)

14

where W1

R 15 is the usual WRS test statistic (i.e., the sum of the ranks of the adjusted reference

16 area measurements).
C-10

1 Using Table C.1,

2 • if S1 > u1 then reject the null hypothesis (the survey unit passes),

3 • if S1 < l1 then do not reject the null hypothesis (the survey unit fails),

4 • if l1 ≤ S1 ≤ u1 then take the second set of samples.

5 If a second set of samples is taken, then compute the following.

6

( ) ( )( )

( ) ( ) ( )

( )

( ) S

W W m m m m n n

m m n n m m n n

W m m n

mn m n

R

=

+ − + + + + +

+ + + + + +

=

− + +

+ +

+ +

1 2 1 2 1 2 1 2

1 2 1 2 1 2 1 2

1 2

1 12

1 2

1 12

/

/

/

/

7 (C-14)

8

9 Using Table C.1, 

10 • if S > u2 then reject the null hypothesis (the survey unit passes),

11 • if S ≤ u2 then do not reject the null hypothesis (the survey unit fails).

12 This test relies on “a large sample approximation.” That is, one is assuming that the sample 

13 size is large enough that the joint distribution of S1 and S is bivariate standard normal with 

14 correlation coefficient

( ) ( ) ( ) S S m n m n 1 1 1

15

, = + + /

(C-15)

16 Some simulation studies would be needed to determine some quantitative bounds on the 

17 accuracy of this approximation.

18

19 C.4 An Alternative Two-Stage, Two-Sample Median Test

20 A different approach to this testing problem has been suggested by Wolfe (1977). In this 

21 procedure, a specific number of sample measurements are made in a reference area, and the 

22 median, M, is calculated and the DCGLW added. Survey unit samples are then analyzed until r

23 of them are found to be below M. The test statistic, nr, is the number of survey unit samples that 

24 have been analyzed. Smaller values of nr indicate that the survey unit meets the release 

25 criterion. For Class 2 and Class 3 survey units, in particular, one would expect that nr = r. In 

26 that case, the number of reference area measurements, m, and the value of r are chosen to 

27 meet the DQOs for the Type I error rate. In each survey unit, r samples are taken. If all are less 

28 than M, one rejects the null hypothesis that the survey unit exceeds the release criterion. If any

29 one of them exceeds M, the null hypothesis will not be rejected. Thus, the total number of 

30 samples needed in each survey unit may be relatively small. In addition, as soon as one 

31 sample is measured above M, the result of the test is known. Thus, it may not be necessary to 

32 analyze every survey unit sample. Of course, the need to identify elevated areas may preclude 

33 the use of this method in some circumstances. However, the potential savings when the 

34 analytical costs are high may make this procedure attractive. As stated previously, if a licensee 
C-11

1 is considering the use of approaches discussed in this appendix, contact with the NRC staff is 

2 strongly encouraged early in the planning process.

3 C.5 References 

4 Gogolak, C.V. “Use of Two-Stage or Double Sampling in Final Status Decommissioning 

5 Surveys.” ADAMS Accession No. ML20035F387. Forty-Sixth Annual Meeting of the Health 

6 Physics Society: Cleveland, OH. June 13, 2001.

7 Hewett, J.E. and J.D. Spurrier. “A Survey of Two Stage Tests of Hypothesis: Theory and 

8 Application.” Communications on Statistics—Theory and Methods. Vol. 12, No. 20: 

9 pp. 2307–2425. 1983.

10 Nuclear Regulatory Commission. “A Proposed Nonparametric Statistical Methodology for the 

11 Design and Analysis of Final Status Decommissioning Surveys–Interim Draft Report for 

12 Comment and Use,” NUREG-1505, Rev. 1. NRC: Washington, DC. 1998.

13

14 Nuclear Regulatory Commission. “Manual for Conducting Radiological Surveys in Support of 

15 License Termination,” NUREG/CR-5849, Draft Report for Comment. NRC: Washington, DC. 

16 1992.

17

18 Spurrier, J.D., and J.E. Hewitt. “Double Sample Tests for the Mean of a Normal Distribution.” 

19 Journal of the American Statistical Association. Vol. 70, No. 350: pp. 448–450. 1975.

20 Spurrier, J.D., and J.E. Hewitt. “Two-Stage Wilcoxon Tests of Hypotheses.” Journal of the 

21 American Statistical Association. Vol. 71, No. 356: pp. 982–987. 1976.

22 Wolfe, D.A. “Two-Stage Two Sample Median Test.” Technometrics. Vol. 19, No. 4: 

23 pp. 459–501. 1977.

SURVEY DATA QUALITY AND REPORTING

D-1

1 D.1 INTRODUCTION

2 The MARSSIM and the MARLAP manual are complementary guidance documents in support of 

3 cleanup and decommissioning activities. The MARSSIM document provides guidance on how 

4 to plan and carry out a study to demonstrate that a site meets appropriate release criteria. It 

5 describes a methodology for planning, conducting, evaluating, and documenting environmental 

6 radiation surveys conducted to demonstrate compliance with cleanup criteria. Chapter 4 and 

7 Appendix A provide more details on MARSSIM. The MARLAP manual provides guidance and a 

8 framework for both project planners and laboratory personnel to ensure that radioanalytical data 

9 will meet the needs and requirements of cleanup and decommissioning activities.

10 Radioanalytical data are commonly generated to support activities such as characterization and 

11 survey of radiologically contaminated sites, effluent and environmental monitoring of nuclear 

12 facilities, emergency response to accidents involving radiological materials, cleanup and 

13 decommissioning of nuclear facilities, and radioactive waste management. Numerous 

14 significant decisions, affecting the health and safety of the public and the environment, are 

15 frequently based on the available radioanalytical data. Considering these activities, the 

16 decisions associated with the radioanalytical data may involve issues pertaining to the extent 

17 and depth of contamination and associated remedial actions; demonstration of compliance with 

18 the cleanup criteria; demonstration of compliance with the effluent release criteria; assessment 

19 of effluent radiological releases and corrective measures; assessment of actions in response to 

20 incidents or accidental releases of radiological materials; and issues involving waste storage, 

21 transport, and disposal. In addition, radioanalytical data commonly influence decisions related 

22 to the cost of remedial actions as well as decisions involving environmental monitoring 

23 strategies and designs.

24 The MARLAP manual was developed to provide guidance and a framework for project planners, 

25 managers, technical reviewers, and laboratory personnel to ensure that the radioanalytical data 

26 produced by surveys will meet the needs and requirements for cleanup and decommissioning 

27 activities. The MARLAP manual addresses the need for a nationally consistent approach to 

28 producing radioanalytical laboratory data that meet a project’s or program’s data requirements. 

29 The guidance provided by MARLAP is both scientifically rigorous and flexible enough to be 

30 applied to a diversity of projects and programs. The MARLAP manual (NRC document 

31 NUREG-1576 and U.S. Environmental Protection Agency (EPA) document 

32 EPA 402-B-04-001A-C) is issued in three volumes (printed version and CD-ROM) and is found 

33 on the Internet at: http://www.nrc.gov/reading-rm/doc-collections/nuregs/staff/sr1576.

34 The NRC staff encourages licensees to follow the recommendations in MARLAP.

35

36 D.2 An Overview of MARLAP 

37 The MARLAP manual is divided into two main parts. Part I provides guidance on using a 

38 performance-based approach for the three phases of radioanalytical projects, including (1) the 

39 planning phase, (2) the implementation phase, and (3) the assessment phase. These three

40 main phases and associated processes should result in analytical data of known quality 

41 appropriate for the intended use. Table D.1 provides an overview of the three main phases, the 

42 processes associated with each phase, and the anticipated outputs for each process. 

43 Figure D.1 illustrates an overview of MARLAP terms and processes and interactions of the 

44 radioanalytical project manager with the laboratory performing the analysis. The MARLAP

45 manual processes and terms described in Table D.1 and Figure D.1 are consistent with 

46 standard practices of the American Society for Testing of Materials (ASTM) for the generation of 
D-2

1 environmental data. Chapters 3 through 9 of the MARLAP manual provide a detailed 

2 description of MARLAP phases and specific processes. It should be noted that it is not a 

3 regulatory requirement to follow or use MARLAP processes as described in Figure D.1; 

4 however, these processes are believed to be flexible and scientifically rigorous to be applied for 

5 the generation of radioanalytical data of the desired quality for the intended use.

6

7 Part II of MARLAP provides technical information on the laboratory analysis of radionuclides. 

8 Specifically, Part II highlights common radioanalytical problems and how to correct them. It also 

9 provides options for analytical protocols and discusses the pros and cons of these options. It 

10 should be noted that Part II does not provide step-by-step instructions on how to perform certain 

11 laboratory procedures or tasks. However, Part II provides guidance to assist laboratory 

12 personnel in selecting the best approach for a particular laboratory task. For example, 

13 Chapter 13 does not contain a step-by-step instruction on how to dissolve a soil sample; 

14 however, it does provide information on acid digestion, fusion techniques, and microwave 

15 digestion, to help the analyst select the most appropriate technique or approach for particular 

16 sample characteristics and project needs. Part II presents detailed technical information on 

17 (1) field and sampling issues that affect laboratory measurements, (2) sample receipt,

18 inspection, and tracking, (3) laboratory sample preparation, (4) sample dissolution,

19 (5) separation techniques, (6) quantification of radionuclides, (7) data acquisition, reduction, and

20 reporting for nuclear counting instrumentation, (8) waste management in a radioanalytical

21 laboratory, (9) laboratory quality control, (10) measurement uncertainty, and (11) detection and

22 quantification capabilities. MARLAP adopted the International Organization for Standardization

23 (ISO) processes, terms, and expressions for analytical measurements, quantifications, and

24 estimation of uncertainty.

25 MARLAP also presents technical details on specific topics outlined in Parts I and II. 

26 Appendices A through E support Part I for the following specific topics: Appendix A, “Directed 

27 Planning Approaches”; Appendix B, “The Data Quality Objective Process”; Appendix C, 

28 “Measurement Quality Objectives for Method Uncertainty and Detection and Quantification 

29 Capability”; Appendix D, “Content of Project Plan Documents”; and Appendix E, “Contracting 

30 Laboratory Services”. Appendix F supports Part II for the specific topic on laboratory 

31 subsampling, and Appendix G provides a compilation of statistical tables.

32 D.3 Use of MARLAP in Decommissioning and Cleanup Projects 

33 MARLAP presents a useful approach and methodology applicable to radioanalytical projects for 

34 cleanup and decommissioning activities. The major processes of the Data Life Cycle are 

35 described briefly below for application in cleanup and decommissioning activities: 

36 D.3.1 The Planning Phase

37 As illustrated in Table D.1 and Figure D.2, planning documents could include Quality Assurance 

38 Project Plans (or QAPPs), Work Plans, Sampling and Analysis Plans, Data Validation Plans, 

39 and Data Quality Assessment Plans. Different organizations may use different terms for these 

40 documents but typically the set of documents include common elements. As provided in 

41 ANSI/ASQC E-4, QAPPs or other planning documents should detail the QA, QC, and other 

42 technical requirements that must be implemented to ensure that the results of the work will meet 

43 stated performance criteria. MARLAP selected EPA’s QAPP as a model for project plan 

44 documents because (i) it is closely associated with the DQO planning process, and (ii) widely 

45 accepted guidance on content (EPA, 2004; EPA, 2002). Chapter 4 and Appendix D of MARLAP

46 contain additional information on the scope and content of planning documents.
D-3

1 The directed planning process for cleanup and decommissioning typically involves the following 

2 radioanalytical aspects:

3 • Stating the cleanup problem: Identify the analytes of concern, matrix of concern,

4 regulatory requirements, sampling constraints, primary decisionmakers, available

5 resources, and existing data and their reliability.

6 • Identifying the cleanup decision: Assess different analytical protocols, identify items

7 of the analytical protocol specifications, and determine how sample collection will affect

8 the measurement quality objectives (MQOs).

9 • Identifying the inputs to the cleanup decisions: Define characteristics of the

10 analytes and matrix, assess the concentration range for the analyte of interest, and

11 define action levels.

12 • Defining the decision boundaries: Identify background and temporal and spatial

13 trends of data and determine limitations of current analytical protocols.

14 • Developing a decision rule and tolerable decision error rates: For example, the

15 decision rule may be defined as, “If the mean concentration of analyte x in the upper

16 15 cm of the soil is greater than z Bq/g, then an action would be taken to remove the soil

17 from the site.” Estimates should be made of uncertainties in the data considering action

18 levels and/or derived concentration guidelines.

19 • Specifying limits on decision error rates: Evaluate the range of possible parameter

20 values and the allowable difference between the action level and the actual value.

21 • Optimizing the strategy for obtaining data: This process may involve optimization of

22 the design for data collection through coordination with the different team members. The

23 process also involves developing analytical protocols specifications and establishing

24 performance measures of the MQOs.

25 D.3.2 The Implementation Phase

26 The radioanalytical process is a compilation of activities, starting from the time a sample is 

27 collected and ending with the data reduction and reporting. Figure 2 illustrates the typical 

28 components of an analytical process used for radiological characterization and survey of 

29 contaminated sites. Certain cleanup or decommissioning projects may not include all of the 

30 components listed in Figure 2. The analytical protocols usually comprise a compilation of 

31 specific procedures or methods and are performed in succession, depending on the particular 

32 analytical process. Using a performance-based approach, a number of alternative protocols 

33 might be appropriate for a particular analytical process. A major component of the analytical 

34 protocol is the analytical method. The radioanalytical process should also include analytical 

35 uncertainty, analytical error, precision, bias, and accuracy of the method used. 

36

37 D.3.3 The Assessment Phase

38 The assessment phase focuses on three major steps: 

39 (1) Data verification: This step ensures that the laboratory conditions and operations are

40 in compliance with the statement of work and the project’s QA project plan. The
D-4

1 verification process would examine the laboratory standard operating procedures. It 

2 would also check for consistency and comparability of the data, correctness of the data 

3 calculations, and completeness of the results and data documentation.

4 (2) Data validation: This step addresses the reliability of the radioanalytical data. It

5 addresses the analyte and matrix types, as well as the uncertainty of the measurement

6 to support the intended use. Validation flags (qualifiers) are typically applied to data that

7 do not meet the acceptance criteria established to meet the project DQOs and MQOs.

8 (3) Data quality assessment (DQA): This step represents the scientific and statistical data

9 evaluation aspects to determine if data are of the right type, quality, and quantity to

10 support the intended use. The DQA is more global in its purview, such that it considers

11 the combined impacts of all project activities on data quality and its usability.

12 D.4 Benefits of Using MARLAP in Decommissioning and Cleanup Projects

13 MARLAP is an extensive document that presents comprehensive guidance and information on 

14 the three phases of the radioanalytical Data Life Cycle. MARLAP emphasizes the importance of 

15 establishing the proper linkages among these phases. Use of MARLAP in decommissioning 

16 and cleanup projects can benefit the user in the following respects:

17 • MARLAP ensures the generation of radioanalytical data of acceptable quality for the

18 intended use.

19 • MARLAP minimizes time and effort expended in generating unacceptable data.

20 • MARLAP enhances public trust in the radioanalytical data generated by licensees and

21 regulators.

22 • MARLAP minimizes efforts applied to justifying data and may limit any litigation costs.

23 • Because MARLAP uses an early coordinated approach to develop the radioanalytical

24 data DQOs and MQOs, this approach would require early coordination and inputs from

25 the decisionmakers, the project manager, shareholders, concerned team members, and

26 the analyst (see Figure D.1). Therefore, this approach should resolve issues or

27 difficulties related to sampling, sample tracking, sample preservation, analysis, data

28 quality, time, and costs early in the process.

29 • MARLAP provides flexibility in selecting the appropriate analytical method, using a

30 performance-based approach that considers the DQOs, the MQOs, and the available

31 resources.

32 • MARLAP enhances regulatory reviews of radioanalytical data and saves time and effort

33 for site characterization, environmental monitoring, decommissioning, and remediation.

34
D-5

1 Table D.1 The Radioanalytical Data Life Cycle

PHASE PROCESS PROCESS OUTPUTS PLANNING

Directed Planning Process Development of DQOs and MQOs, including 

Optimized Sampling and Analytical Designs 

Plan Documents Project Plan Documents, including Quality 

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

IMPLEMENTATION

Sampling Laboratory Samples

Analysis Laboratory Analysis, including QC Samples 

and Complete Data Package

ASSESSMENT

Verification Verified Data and Data Verification Report

Validation Validated Data and Data Validation Report

DQA Assessment Report

Data of Known Quality Appropriate for the Intended Use 

2
D-6

1

2

3 Figure D.1 MARLAP Road Map—Key Terms and Processes 
D-7

1

2

3

4

5

6

7

8

9

10

11

12

13

14

15

16

17

18

19

20

21

22

23

24

25

26

27

28

29

30

31

32

33

34

35

36

37

38 Figure D.2 Typical Components of the Radioanalytical Process 

From MARLAP (NUREG-1576)

Sample Tracking

Data Reduction and 

Reporting

Quality Assurance 

Quality Control

Field Sample Preparation

Sample Receipt and 

Inspection

Laboratory Sample 

Preparation

Sample Dissolution

Chemical Separation of 

Radionuclides of Concern

Preparation of Samples 

for Instrument 

Measurements

Instrument Measurement
D-8

1 D.5 References

2 American Society for Quality Control (ANSI/ASQC) E-4. “Specifications and Guidelines for 

3 Quality Systems for Environmental Data Collection and Environmental Technology Programs.” 

4 American Society for Quality Control: Milwaukee, Wisconsin. 1995.

5 American Society for Testing and Materials (ASTM). “Standard Practice for Generation of 

6 Environmental Data Related to Waste Management Activities: Development of Data Quality 

7 Objectives,” ASTM D5792. 1995.

8 — — — — —. “Standard Guide for Data Assessment for Environmental Waste Management 

9 Activities,” ASTM D6233. 1998.

10 Environmental Protection Agency (EPA). “Guidance for Data Quality Assessment: Practical 

11 Methods for Data Analysis,” (EPA QA/G-9). EPA/600/R-96/084. EPA: Washington, DC.

12 2000a. 

13 — — — — —. “Guidance for the Data Quality Objective Process (EPA QA/G-4),” 

14 EPA/600/R-96/055. EPA: Washington, DC. 2000b. 

15 — — — — —. “Guidance for Quality Assurance Project Plans. EPA QA/G-5,”

16 EPA/240/R-02/009. EPA: Office of Environmental Information. Washington, DC. 2002.

17 — — — — —. U.S. Department of Defense, U.S. Department of Energy, U.S. Department of 

18 Homeland Security, U.S. Nuclear Regulatory Commission, U.S. Food and Drug Administration, 

19 U.S. Geological Survey, and National Institute of Standards and Technology. “Multi-Agency 

20 Radiological Laboratory Analytical Protocols (MARLAP),” NUREG-1576, EPA402-B-04-001A, 

21 NTIS PB2004-105421, Vol. I, II, and III and Supp. 1. Washington, DC. 2004.

22 International Organization for Standardization (ISO)/International Electrotechnical Commission 

23 (IEC). “International Vocabulary of Basic and General Terms in Metrology.” ISO/IEC Guide 

24 99:2007. ISO: Geneva, Switzerland. 2007.

25 — — — — —. “Guide to the Expression of Uncertainty in Measurement.” ISO: 

26 Geneva, Switzerland. 1995.

27 Nuclear Regulatory Commission (U.S.) (NRC). “Multi-Agency Radiation Survey and Site 

28 Investigation Manual,” (MARSSIM), NUREG-1575. NRC: Washington, DC. August 2000.
MEASUREMENTS FOR FACILITY RADIATION SURVEYS

E-1

1 E.1 Introduction

2 This appendix is applicable to all decommissioning groups. All surveys, whether simple or 

3 complex FSSs, require information on the reasons for instrument selection, the nature of the 

4 radionuclides, measurement techniques and procedures, MDCs of the instruments 

5 (measurement systems), and instrument calibration. Therefore, the information presented in 

6 this appendix would apply to a simple survey used to demonstrate compliance with regulatory 

7 decommissioning criteria, as well as a complex FSS.

8

9 This appendix contains limited, general information on survey techniques and survey 

10 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 11 , MARLAP, 

12 and various NUREGs for more detailed information. These are important areas for the conduct 

13 of surveys in the RSSI process and include the basic modes for determining levels of radiation 

14 and radioactivity at a site, instrument and scanning detection limits, instrument calibration, and 

15 laboratory measurements for samples. The data from the FSS are the deciding factor in judging 

16 if the site meets the release criteria.

17 Radiological conditions that should be determined for license termination purposes include any 

18 combination of total surface activities, removable surface activities, exposure rates, radionuclide 

19 concentrations in soil, or induced activity levels. To determine these conditions, field 

20 measurements and laboratory analyses may be necessary. For certain radionuclides or 

21 radionuclide mixtures, the licensee may have to measure both alpha and beta radiation. In 

22 addition to assessing the average radiological conditions, the licensee should identify small 

23 areas with elevated levels of residual radioactivity and determine their extent and activities. 

24 There are three basic modes in which one can operate in determining the levels of radiation and 

25 radioactivity at a site. They are scanning with hand-held survey instruments, direct 

26 measurements with these same or larger instruments, and sample collection at the site followed 

27 by analysis in the laboratory. In many cases, the licensee will use some combination of these 

28 modes to obtain data, although the exact mix would be expected to vary according to the 

29 application.

30 In practice, the licensee uses the DQO process to obtain a proper balance among the uses of 

31 various measurement techniques. In general, there is an inverse correlation between the cost 

32 of a specific measurement technique and the detection levels being sought. Depending on the 

33 survey objectives, important considerations include survey costs and choosing the optimum 

34 instrumentation and measurement mix.

35 The decision to use a measurement method as part of the survey design is determined by the 

36 survey objectives and the survey unit classification. Scanning is performed to identify areas of 

37 elevated activity that other measurement methods may not detect. Direct measurements are 

38 analogous to collecting and analyzing samples to determine the average activity in a survey 

39 unit. 

40

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-2

1 E.2 Direct Measurements (Fixed Measurements)

2 To conduct direct measurements of alpha, beta, and photon surface activity, instruments and 

3 techniques providing the required detection sensitivity are selected. The selection of the type of 

4 instrument and method of performing the direct measurement depends on the type of residual 

5 radioactivity present, the measurement sensitivity requirements, and the objectives of the 

6 radiological survey.

7

8 Direct measurements may be collected at random locations in the survey unit. Alternatively, 

9 direct measurements may be collected at systematic locations and supplement scanning 

10 surveys to identify small areas of elevated activity. Direct measurements may also be collected 

11 at locations identified by scanning surveys as part of an investigation to determine the source of 

12 the elevated instrument response. Professional judgment may also be used to identify locations 

13 for direct measurements to further define the areal extent of residual radioactivity and to 

14 determine maximum radiation levels within an area, although these types of direct 

15 measurements are usually associated with preliminary surveys (i.e., scoping, characterization, 

16 remedial action support). Licensees should document all direct measurement locations and 

17 results.

18 If the equipment and methodology used for scanning are capable of providing data of the same 

19 quality required for direct measurement (e.g., detection limit, location of measurements, ability 

20 to record and document results), then scanning may be used in place of direct measurements. 

21 Similarly, the usage of in situ measurement instrumentation may be possible if sufficient data 

22 quality can be achieved. In both cases, proposed approaches should be developed using the 

23 DQO process and should be communicated to NRC staff. Results should be documented for at 

24 least the number of locations required for the statistical tests. In addition, some direct 

25 measurement systems may be able to provide scanning data, provided they meet the objectives 

26 of the scanning survey.

27 Chapter 6 of MARSSIM includes information on radiation measurements. Specifically, 

28 Section 6.4.1 of MARSSIM contains information on direct measurements for alpha-, beta-, and 

29 gamma-emitting radionuclides.

30 E.3 Scanning Measurements 

31 Scanning is the process by which the operator uses portable radiation detection instruments to 

32 detect the presence of radionuclides on a specific surface (i.e., ground, wall, floor, equipment). 

33 The term scanning survey describes the process of moving portable radiation detectors across 

34 a suspect surface with the intent of locating residual radioactivity. Investigation levels for 

35 scanning surveys are determined during survey planning to identify areas of elevated activity. 

36 Scanning surveys are useful in locating radiation anomalies indicating residual gross activity 

37 that might require further investigation or action.

38

39 Areas of elevated activity typically represent a small portion of the site or survey unit. Thus, 

40 random or systematic direct measurements or sampling on the commonly used grid spacing 

41 may have a low probability of identifying these areas. Scanning surveys are often relatively 

42 quick and inexpensive to perform. For these reasons, the licensee typically performs them 

43 before direct measurements or sampling. This avoids spending time fully evaluating an area 

44 that may quickly prove to contain residual radioactivity above the IL during the scanning 

45 process. Based on the HSA, surfaces to be surveyed, and survey design objectives, licensees 

46 conduct scans that would indicate all radionuclides potentially present, using surrogate 
E-3

1 measurements where appropriate. Documenting scanning results and observations from the 

2 field is very important. For example, licensees should document a scan that identified relatively 

3 sharp increases in instrument response or identified the boundary of an area of increased 

4 instrument response. This information is useful when interpreting survey results.

5 Chapter 6 of MARSSIM includes information on radiation measurements. Specifically, 

6 Section 6.4.2 of MARSSIM contains information on scanning measurements for alpha-, beta-, 

7 and gamma-emitting radionuclides.

8 E.4 Sampling 

9 For certain radionuclides that cannot be effectively measured directly in the field, the licensee 

10 should collect samples of the medium under investigation (e.g., soil) and then analyze them with 

11 a laboratory-based procedure. On the simplest level, this would include the analysis of a smear 

12 sample using a gross alpha-beta counter. More involved analyses would include gamma 

13 spectrometry, beta analysis using liquid scintillation counting, or alpha spectrometry following 

14 separation chemistry.

15

16 Samples from a variety of locations may be required, depending upon the specific facility 

17 conditions and the results of scans and direct measurements. Inaccessible surfaces cannot be 

18 adequately evaluated by direct measurements on external surfaces alone; therefore, those 

19 locations that could contain residual radioactive material should be accessed for surveying. 

20 Residue can be collected from drains using a piece of wire or plumber’s “snake” with a strip of 

21 cloth attached to the end; deposits on the pipe interior can be loosened by scraping with a 

22 hard-tipped tool that can be inserted into the drain opening. Particular attention is paid to “low 

23 points” or “traps” where activity would likely accumulate. The need for further internal 

24 monitoring and sampling is determined on the basis of residue samples and direct 

25 measurements at the inlet, outlet, cleanouts, and other access points to the pipe interior.

26 Residual activity will often accumulate in cracks and joints in the floor. These are sampled by 

27 scraping the crack or joint with a pointed tool such as a screwdriver or chisel. Samples of the 

28 residue can then be analyzed; positive results of such an analysis may indicate possible 

29 subfloor residual radioactivity. Checking for activity below the floor will require accessing a 

30 crawl space (if one is present) or removing a section of the flooring. Coring, using a 

31 commercially available unit, is a common approach to accessing the subfloor soil. After 

32 removing the core (where the diameter may range from a few centimeters to up to 

33 20 centimeters), direct monitoring of the underlying surface can be performed and samples of 

34 soil collected.

35 Coring is also useful for collecting samples of construction material that may contain activity that 

36 has penetrated below the surface or activity induced by neutron activation. This type of 

37 sampling is also applicable to roofing material, which may contain embedded or entrapped 

38 contaminants. The profile of the distribution and the total radionuclide content can be 

39 determined by analyzing horizontal sections of the core.

40 If residual activity has been coated by paint or some other treatment, the underlying surface and 

41 the coating itself may contain residual radioactivity. If the activity is a pure alpha or low-energy 

42 beta emitter, measurements at the surface will probably not be representative of the actual 

43 residual activity level. In this case, the licensee can remove the surface layer from a known 

area, usually 100 cm2 44 , using a commercial stripping agent or by physically abrading the surface. 

45 The removed coating material is analyzed for activity content and the level converted to units of 
E-4

disintegrations per minute (dpm)/100 square centimeters (cm2 1 ) for comparison with guidelines 

2 for surface activity. The licensee takes direct measurements on the underlying surface, after 

3 removing the coating.

4 MARSSIM and MARLAP contain information on sampling and laboratory analysis for 

5 decommissioning. Chapter 10 of MARLAP discusses field and sampling issues that affect 

6 laboratory measurements.

7 E.5 Minimum Detectable Concentrations 

8 Detection limits for field survey instrumentation are important criteria in the selection of 

9 appropriate instrumentation and measurement procedures. For the most part, the licensee 

10 determines detection limits to evaluate whether a particular instrument and measurement 

11 procedure is capable of detecting residual activity at the regulatory release criteria. One may 

12 demonstrate compliance with decommissioning criteria by performing surface activity 

13 measurements and directly comparing the results to the surface activity DCGLs. However, 

14 before any measurements are performed, the survey instrument and measurement procedures 

15 to be used should be evaluated to ensure they possess sufficient detection capabilities relative 

16 to the surface activity DCGLs.

17

18 The measurement of residual radioactivity during surveys in support of decommissioning often 

19 involves measuring residual radioactivity at near-background levels. Thus, the licensee should

20 determine the minimum amount of radioactivity that may be detected by a given survey 

21 instrument and measurement procedure. In general, the MDC is the minimum activity 

22 concentration on a surface or within a material volume that an instrument is expected to detect 

23 (i.e., activity expected to be detected with 95 percent confidence). It is important that this 

24 activity concentration, the MDC, is determined a priori (i.e., before survey measurements are 

25 conducted).

26 As generally defined, the detection limit, which may be a count or count rate, is independent of 

27 field conditions such as scabbled, wet, or dusty surfaces. That is, the detection limit is based on 

28 the number of counts and does not necessarily equate to measured activity under field 

29 conditions. These field conditions do, however, affect the instrument’s “detection sensitivity” or 

30 MDC. Therefore, the licensee should not use the terms MDC and detection limit

31 interchangeably.

32 In MARSSIM, MARLAP, and other NRC NUREGs, the MDC corresponds to the smallest activity 

33 concentration measurement that is practically achievable with a given instrument and type of 

34 measurement procedure. That is, the MDC depends not only on the particular instrument 

35 characteristics (e.g., instrument efficiency, background, integration time) but also on the factors 

36 involved in the survey measurement process (U.S. Environmental Protection Agency, 

37 EPA 520/1-80-012, “Upgrading Environmental Radiation Data,” issued August 1980), which 

38 include surface type, source-to-detector geometry, and source efficiency (e.g., backscatter and 

39 self-absorption).
E-5

1 MARLAP Section 3.3.7, “Method Performance Characteristics and Measurement Quality 

2 Objectives,” and Chapter 20, “Detection and Quantification Capabilities,” discuss MDCs.

3 E.6 Survey Minimum Detectable Concentrations 

4 During radiological surveys in support of decommissioning, scanning is performed to identify the 

5 presence of any locations of elevated direct radiation. The probability of detecting residual

6 radioactivity in the field is affected not only by the sensitivity of the survey instrumentation when 

7 used in the scanning mode of operation but also by the surveyor’s ability. The surveyor will

8 decide whether the signals represent only the background activity, or whether they represent 

9 residual radioactivity in excess of background.

10 The MDC of a scan survey, referred to as scan MDC or MDCscan, depends on the intrinsic 

11 characteristics of the detector (e.g., efficiency, window area), the nature (e.g., type and energy 

12 of emissions) and the relative distribution of the residual radioactivity (e.g., point versus 

13 distributed source and depth of residual radioactivity), scan rate, and other characteristics of the 

14 surveyor. Some factors that may affect the surveyor’s performance include the costs 

15 associated with various outcomes—e.g., cost of missed residual radioactivity versus cost of 

16 incorrectly identifying areas as containing residual radioactivity—and the surveyor’s a priori

17 expectation of the likelihood of residual radioactivity being present. For example, if the surveyor 

18 believes that the potential for residual radioactivity is very low, as in an unaffected area, then a 

19 relatively large signal may be needed for the surveyor to conclude that residual radioactivity is 

20 present. NUREG/CR–6364, “Human Performance in Radiological Survey Scanning,” issued 

21 March 1998, contains a complete discussion of the human factors as they relate to the 

22 performance of scan surveys.

23 Signal detection theory provides a framework for the task of deciding whether the audible output 

24 of the survey meter during scanning was due to background or signal plus background levels. 

25 An index of sensitivity (dʹ) that represents the distance between the means of the background 

26 and background plus signal, in units of their common standard deviation, can be calculated for 

27 various decision errors—Type I error (α) and Type II error (β). As an example, for a correct

28 detection or true positive rate of 95 percent (1−β) and a false positive rate (α) of 5 percent, 

29 dʹ is 3.29 (similar to the static MDC for the same decision error rates). The index of sensitivity is 

30 independent of human factors, and therefore, the ability of an ideal observer (i.e., theoretical 

31 construct) may be used to determine the minimum dʹ that can be achieved for particular decision 

32 errors. The ideal observer makes optimal use of the available information to maximize the 

33 percent of correct responses and thus provides an effective upper bound against which to 

34 compare actual surveyors. Computer simulations and field experimentation can then be 

35 performed to evaluate the surveyor efficiency (p) relative to the ideal observer. The resulting 

36 expression for the ideal observer’s minimum detectable count rate (MDCR), in counts per 

37 minute (cpm), can be written:

MDCR d b i s i =  i ( ) ( ) =

i 

/

38

60 60 / /

(E-1)

39 where MDCR = minimum detectable (net) count rate in cpm,

40 bi = background counts in the observation interval, 

41 si = minimum detectable number of net source counts in the observation 

42 interval, and 

43 i = observational interval (in seconds), based on the scan speed and areal 

44 extent of the residual radioactivity.
E-6

1 Scan MDCs are determined from the MDCR by applying conversion factors to obtain results in 

2 terms of measurable surface activities and soil concentrations. As an example, the scan MDC

3 for a structure surface can be expressed as:

4

Scan MDC MDCR

p

probe area

cm

i s

=

 

100 2

5 (E-2)

6

7 Chapter 6 of NUREG-1507, “Minimum Detectable Concentrations with Typical Radiation Survey 

8 Instruments for Various Contaminants and Field Conditions,” issued June 1998, contains a 

9 discussion of survey MDCs. This discussion includes scan MDC equations for both 

10 building/structure surfaces and land areas.

11

12 E.7 Survey Instrument Calibration

13 Before determining the MDC for a particular instrument and survey procedure, it is necessary to 

14 introduce the expression for total alpha or beta surface activity per unit area. In the ISO Guide 

15 7503–1, “Evaluation of Surface Contamination,” 1988, the ISO recommends calculating the total 

16 surface activity, AS, as in the following expression:

17

( )( )( ) 

s

S B B

i S

R R

W

=

+ −

 

18 (E-3)

19

20 where RS+B = the gross count rate of the measurement in cpm,

21 RB = the background count rate in cpm, 

22 εi = the instrument or detector efficiency (unitless), 

23 εs = the efficiency of the residual radioactivity source (unitless), and 

W = the area of the detector window (cm2 24 ).

(For instances in which W does not equal 100 cm2 25 , probe area corrections are necessary to 

convert the detector response to units of dpm per 100 cm2 26 .)

27 This expression clearly distinguishes between instrument (detector) efficiency and source 

28 efficiency. The product of the instrument and source efficiency yields the total efficiency, ε tot. 

29 Currently, surface residual radioactivity is assessed by converting the instrument response to 

30 surface activity using one overall total efficiency. This is not a problem, provided that the 

31 calibration source exhibits characteristics similar to the surface residual radioactivity—including 

32 such characteristics as radiation energy, backscatter effects, source geometry, and 

33 self-absorption. In practice, this is hardly the case; more likely, total efficiencies are determined 

34 with a clean, stainless steel source, and then those efficiencies are used to measure residual

35 radioactivity on a dust-covered concrete surface. By separating the efficiency into two 

36 components, the surveyor has a greater ability to consider the actual characteristics of the 

37 surface residual radioactivity.

38 The instrument efficiency is defined as the ratio between the net count rate of the instrument 

39 and the surface emission rate of a source for a specified geometry. The surface emission rate, 

40 q2π, is defined as the “number of particles of a given type above a given energy emerging from 
E-7

1 the front face of the source per unit time” (ISO 7503–1). The surface emission rate is the 2π 

2 particle fluence that embodies both the absorption and scattering processes that affect the 

3 radiation emitted from the source. Thus, the instrument efficiency is determined by 

4

( )( )( ) 



S

S B

i S

R R

W

=

+ −

 

5 (E-4)

6

7 The instrument efficiency is determined during calibration by obtaining a static count with the 

8 detector over a calibration source that has a traceable activity or surface emission rate or both. 

9 In many cases, it is the source surface emission rate that is measured by the manufacturer and 

10 certified as National Institute of Standards and Technology traceable. The source activity is 

11 then calculated from the surface emission rate, based on the assumed backscatter and self12 absorption properties of the source. The theoretical maximum value of instrument efficiency is 

13 one.

14 The source efficiency, ε s, is defined as the ratio between the number of particles of a given type 

15 emerging from the front face of a source and the number of particles of the same type created 

16 or released within the source per unit time (ISO 7503–1). The source (or surface) efficiency 

17 takes into account the increased particle emission due to backscatter effects, as well as the 

18 decreased particle emission due to self-absorption losses. For an ideal source (no backscatter 

19 or self-absorption), the value of εs is 0.5. Many real sources will exhibit values of ε s less than 

20 0.5, although values greater than 0.5 are possible, depending on the relative importance of the 

21 absorption and backscatter processes. Source efficiencies may either be determined 

22 experimentally or simply selected from the guidance contained in ISO 7503–1.

23 Some of the factors that affect the instrument efficiency, ε I, include detector size (probe surface 

24 area), window density thickness, geotropism, instrument response time, and ambient conditions 

25 such as temperature, pressure, and humidity. The instrument efficiency also depends on the 

26 radionuclide source used for calibration and the solid angle effects, which include source-to27 detector distance and source geometry.

28 Some of the factors that affect the source efficiency, ε s, include the type of radiation and its 

29 energy, source uniformity, surface roughness and coverings, and surface composition 

30 (e.g., wood, metal, concrete).

31 The licensee assesses surface activity levels by converting detector response, through the use 

32 of a calibration factor, to radioactivity. Once the detector has been calibrated and an instrument 

33 efficiency (εi) established, several factors still need to be carefully considered when using that 

34 instrument in the field. These factors involve the background count rate for the particular 

35 surface and the surface efficiency (εs), which addresses the physical composition of the surface 

36 and any surface coatings. Ideally, the surveyor should use experimentally determined surface 

37 efficiencies for the anticipated field conditions. The surveyor needs to know how and to what 

38 degree these different field conditions can affect the sensitivity of the instrument. A particular 

39 field condition may significantly affect the usefulness of a particular instrument (e.g., wet 

40 surfaces for alpha measurements or scabbled surfaces for low-energy beta measurements).
E-8

1 One of the more significant implicit assumptions commonly made during instrument calibration 

2 and subsequent use of the instrument in the field is that the composition and geometry of 

3 residual radioactivity in the field is the same as that of the calibration source. This may not be 

4 the case, considering that many calibration sources are fabricated from materials different from 

5 those that comprise the surfaces of interest in the field (e.g., activity plated on a metallic disc 

6 (Walker, “Proper Selection and Application of Portable Survey Instruments for Unrestricted 

7 Release Surveys,” issued 1994)). This difference usually manifests itself in the varying 

8 backscatter characteristics of the calibration and field surface materials.

9 Generally, it will not be necessary to recalculate the instrument MDC to adjust for the field 

10 conditions. The instrument detection limit (in net counts or net count rate) remains the same, 

11 but the surface activity MDC may be different (due to the varying εs).

12 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 13 . The ISO 7503 series was updated 

14 in 2016, prior to Revision 2 of this NUREG report. Therefore, a comparison of the 1988 and 

15 2016 versions of ISO 7503 was performed to understand any differences in approach to surface 

16 or source efficiency, and to determine if the methodology/terminology in NUREG-1507 should 

17 be updated. A detailed discussion of this analysis will be provided in Revision 1 of NUREG15073 18 , while a brief overview of the comparison is provided below. 

19

20 Several of the basic concepts and methods were compared between the 1988 and 2016 ISO 

21 7503 series. The concepts of “source efficiency” and an “ideal source” were compared, and it 

22 was concluded that they were presented as essentially the same (noting that a different “P23 Factor” terminology was used for the 2016 series). The equations to evaluate contamination 

24 measurement data, and the associated usage of εs and the P-Factor, were compared between 

25 the two series, and these equations were essentially the same (while different terminology has 

26 been used). A comparison was performed between recommended default or “conservative” εs

27 values and P-Factors in the two series, and these values utilized the same assumptions and 

28 made the same recommendations for default source efficiencies (again noting that different

29 terminology was presented). Based upon the ISO 7503 series comparison, it has been 

30 concluded that the 2016 revision of ISO 7503-1 presented no compelling reason to update the 

31 usage of the “surface efficiency” concept and terminology for Revision 1 of NUREG-1507. 

32 Therefore, the equations for MDC and surface activity measurement, which were originally 

33 developed using ISO 7503-1:1988 remain valid and are not changed from the previous NUREG34 1757 version. 

35

36 However, the comparison between the 1988 and 2016 ISO 7503 series identified that there is a 

37 need to consider weighted detection efficiencies for use with multiple radionuclides or with 

38 complex decay series, as ISO 7503-3:2016, in particular, presented many new concepts in this 

39 area. To address this need, Revision 1 to NUREG-1507 will include “weighted efficiency” 

40 calculations that utilize the concepts of instrument efficiency, source efficiency, and emission 

41 intensity, while also considering the relative fraction of radionuclides and branching ratios.

42 Chapter 4 of NUREG-1507 covers survey instrument calibration and the effects of efficiency 

43 changes on MDC. Chapter 5 of NUREG-1507 discusses variables affecting efficiencies in the 

2 MARSSIM, Rev. 1, and NUREG-1507, Rev. 0, reference ISO 7503-1:1988. The NRC has found this standard 

acceptable for use by NRC licensees to calculate MDCs and surface activity measurements. 

3 As of the date of publication of this volume, NUREG-1507, Revision 1, is under development.
E-9

1 field. Chapter 20 of MARLAP discusses instrument efficiency and the minimum detectable net 

2 instrument signal.

3 E.8 Laboratory Measurements 

4 Frequently during surveys in support of decommissioning, it is not feasible, or even possible, to 

5 detect the residual radioactivity with portable field instrumentation; thus arises the need for a 

6 laboratory analysis of media samples. This is especially the case for such media samples as 

7 soil, which result in significant self-absorption of the radiation from the residual radioactivity. 

8 Another common situation that necessitates the use of laboratory analyses occurs when the 

9 residual radioactivity is difficult to detect even under ideal conditions. This includes residual 

10 radioactivity that emits only low-energy beta radiation (e.g., H-3 and Ni-63) or x-ray radiation 

11 (e.g., Fe-55). Laboratory analyses for radionuclide identification, using spectrometric 

12 techniques, are often performed during scoping or characterization surveys. Here, the principal 

13 objective is to simply determine the specific radionuclides present in the residual radioactivity, 

14 without necessarily having to assess the quantity of residual radioactivity. Once the licensee 

15 identifies the residual radioactivity, it selects sufficiently sensitive field survey instrumentation 

16 and techniques to demonstrate compliance with the DCGLs.

17 Samples collected during surveys for decommissioning purposes should be analyzed by trained 

18 individuals using the appropriate equipment and procedures at a well-established laboratory, 

19 which uses either in-house or contractor laboratory services. There should be written 

20 procedures that document both (1) the laboratory’s analytical capabilities for the radionuclides of 

21 interest and (2) the QA/QC program that ensures the validity of the analytical results. Many of 

22 the general types of radiation detection measuring equipment used for survey field applications 

23 are also used for laboratory analyses, usually under more controlled conditions that provide for 

24 lower detection limits and greater delineation between radionuclides. Laboratory methods often 

25 also involve a combination of both chemical and instrumental techniques to quantify the low 

26 levels expected to be present in samples from decommissioning facilities.

27 To reemphasize, a thorough knowledge of the radionuclides present, along with their chemical 

28 and physical forms and their relative abundance, is needed to select appropriate laboratory 

29 methods. With this information, it may be possible to substitute certain gross 

30 (i.e., nonradionuclide-specific) measurement techniques for the more costly and time31 consuming wet chemistry separation procedures and relate the gross data back to the relative 

32 quantities of specific contaminants. The individual responsible for the survey should be aware 

33 that radiochemical analyses require lead times that will vary according to the nature and 

34 complexity of the request. For example, a laboratory could provide a fairly quick turnaround on 

35 gamma spectrometry because sample preparation is likely limited to the media being dried and 

36 homogenized before being measured in a standard geometry. In comparison, alpha 

37 spectrometry usually involves sample preparation that also includes chemical separation and 

38 will typically require a longer lead time. Some factors influencing the analysis time include

39 (1) the nuclides of concern, (2) the type of samples to be analyzed, (3) the QA/QC

40 considerations required, (4) the availability of adequate equipment and personnel, and (5) the

41 required detection limits.

42 For relatively simple analyses, such as gross alpha and gross beta counting of smears and 

43 water samples, liquid scintillation spectrometry for low-energy beta emitters in smear and water 

44 samples, and gamma-spectrometry of soil, it is usually practical to establish in-house laboratory 

45 capabilities. The more complicated and labor-intensive procedures, such as alpha 
E-10

1 spectrometry, Sr-90 and low-energy beta emitters (e.g., H-3, Ni-63) in soil samples, should be 

2 considered candidates for contract laboratory analyses.

3 Analytical methods should be capable of measuring levels below the established release 

4 guidelines; detection sensitivities of 10 to 25 percent of the guideline should be the target. 

5 Although laboratories will state detection limits, these limits are usually based on ideal situations 

6 and may not be achievable under actual measurement conditions. Also, remember that 

7 detection limits are subject to variation from sample to sample, instrument to instrument, and 

8 procedure to procedure, depending upon sample size, geometry, background, instrument 

9 efficiency, chemical recovery, abundance of the radiations being measured, counting time, self10 absorption in the prepared sample, and interference from other radionuclides present.

11 MARSSIM and MARLAP contain information on sampling and laboratory analysis for 

12 decommissioning. MARLAP Sections 12, 13, 14, and 15 discuss laboratory sample 

13 preparation, sample dissolution, separation techniques, and quantification of radionuclides.

14 E.9 References

15 Nuclear Regulatory Commission (U.S.) (NRC). “Human Performance in Radiological Survey 

16 Scanning,” NUREG/CR-6364. NRC: Washington, DC. 1997.

17 — — — — —. “Minimum Detectable Concentrations with Typical Radiation Survey Instruments 

18 for Various Contaminants and Field Conditions,” NUREG-1507. NRC: Washington, DC. 1998. 

19 — — — — —. “Multi-Agency Radiation Survey and Site Investigation Manual,” (MARSSIM, 

20 Rev. 1), NUREG-1575. NRC: Washington, DC. 2000.

21 Environmental Protection Agency (U.S.) (EPA). “Upgrading Environmental Radiation Data,”

22 EPA 520/1–80–012. EPA: Washington, DC. 1980.

23 International Organization for Standardization. “Evaluation of Surface Contamination—Part 1: 

24 Beta Emitters and Alpha Emitters (first edition).” ISO 7503–1. ISO: Geneva, Switzerland. 

25 1988.

26 Walker, E. “Proper Selection and Application of Portable Survey Instruments for Unrestricted 

27 Release Surveys.” Bechtel Environmental, Inc. Presented at 1994 International Symposium on 

28 D&D. April 24–29, 1994.
SURFACE WATER AND GROUNDWATER CHARACTERIZATION

F-1

1 F.1 Introduction

2 This appendix includes guidance on surface and groundwater characterization to support 

3 development of conceptual site models, development of hydrologic inputs to dose assessment 

4 models, and radiological surveys to estimate dose associated with existing groundwater 

5 contamination. Surface water and groundwater characterization is important for

6 Decommissioning Group 4 sites with surface water contamination and Decommissioning 

Groups 5–7 that have existing groundwater1 7 contamination. Characterization is also important 

8 for sites that have a medium to high potential for groundwater contamination (see Section F.4) 

9 based on historical site activities even in the absence of evidence of groundwater 

10 contamination. 

11

12 If groundwater is contaminated, characterization of groundwater is an essential component of 

13 the dose modeling used in the estimation of doses to demonstrate compliance with the license 

14 termination requirements in 10 CFR Part 20, “Standards for Protection against Radiation,” 

15 Subpart E, “Radiological Criteria for License Termination.” In these cases, unmodified

16 screening derived concentration guideline levels (DCGLs) for soil are inappropriate to use, 

17 because the screening levels assume surface water and groundwater are initially 

18 uncontaminated. Likewise, if site-specific dose modeling is used to demonstrate compliance 

19 with radiological criteria for license termination, the dose contributions of contaminated surface 

20 water or groundwater should be taken into consideration in the dose assessment. 

21

22 This appendix references reports, methods and software that should assist in characterization of 

23 surface water and groundwater and associated dose modeling. However, it should be noted 

24 that other reports, methods, and software may also be appropriate. If other approaches are 

25 utilized, it may be prudent for the licensee to contact NRC to ensure that the approaches are 

26 appropriate prior to any significant resource input by the licensee.

27

28 F.2 Planning for Surface Water and Groundwater Characterization 

29 The licensee should plan surface water and groundwater characterization in a manner that 

30 maximizes the utility of the information to be collected during the various stages of radiological 

31 surveys. For example, for a particular site, a licensee may show that the surface water pathway 

32 is not likely to be significant in terms of existing and potential future exposure to the public. In 

33 such a case, the need for detailed characterization of the surface water system decreases. On 

34 the other hand, the identification of groundwater contamination during the preliminary scoping 

35 survey may warrant installing and sampling additional monitoring wells to define the nature and 

36 extent of groundwater contamination.

37 In some instances, groundwater may be unsuitable for specific uses, such as human and 

38 livestock consumption, but may be acceptable for crop irrigation. In addition, some aquifers 

39 may not have the yield to support crop irrigation but may produce enough water for human 

40 consumption. In some instances, the EPA or a State agency may have declared that the 

41 aquifer in question is unfit for human or livestock use. Accordingly, the licensee should address 

42 this type of information, because it will support site exposure scenario development and dose 

43 modeling. The State agency may also have rules that apply to groundwater resource 

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). 
F-2

1 classifications. Section I.3.3.3 from Appendix I and Appendix M to this volume provides 

2 guidance on modification of waterborne exposure pathways. 

3 F.3 Development of CSMs and Mathematical Models

4 A conceptual site model (CSM) provides a hypothetical framework for contaminant source, 

5 geologic, hydrologic (including water usage), chemical, biologic, and demographic 

6 characteristics for the site (NRC, 2007). The CSM provides the basis for understanding flow 

7 and transport at the site for abstraction into a dose assessment model and is the starting point 

8 for numerical models if contamination is present in the surface water and groundwater. In 

9 general, CSMs should be updated as new information becomes available. The complexity of 

10 CSMs should be commensurate with site risk and at an appropriate level to demonstrate 

11 radiological criteria for license termination can be met. ASTM E1689-95 (2014) provides 

12 information on the development of CSMs that may be useful to licensees.

13

14 The model abstraction process involves the representation of major components of a complex 

15 system in a conceptual model. Conceptual models should represent or be able to describe the

16 important features and processes of the system. A simplified representation of the conceptual 

17 model is developed so that the conceptual model can be more easily represented by a 

18 mathematical model. Mathematical models translate the assumptions of a conceptual model 

19 into the formalism of mathematics (IAEA, 2004). Mathematical models are represented by 

20 equations and, if these equations are solved, the output can include information such as 

21 radionuclide concentrations in media, doses to humans, and temporal evolution of the system. 

22 Numerical codes can solve such equations, and constructed numerical models are used to 

23 solve the equations represented in the mathematical models, thereby simulating the properties 

24 of features and processes represented in a conceptual model. Model simplification is the

25 process for reducing the complexity of a complex numerical model into a simpler numerical 

26 model while still maintaining the validity of the simulation results and is another form of model 

27 abstraction.

28

29 Compared to model simplification (i.e. reducing a complex numerical model to a less complex 

30 numerical model), abstracting the physical reality that is represented in a conceptual model to a 

31 mathematical model is typically more vulnerable to error. In addition, it can be difficult to 

32 validate how successfully the physical system is abstracted into a mathematical model whereas 

33 it is usually easier to make sure that a simplified numerical model is adequately representing a 

34 more complex numerical model. However, unlike model simplification, model abstraction is a 

35 required and necessary step in the performance assessment methodology. Not all features and 

36 processes can or should be included in a mathematical model. As an analogy, news reporters 

37 must abstract facts from a real set of complex circumstances or else the news story becomes 

38 too lengthy to be useful for most readers or viewers. Not only is model abstraction necessary, it 

39 can be useful to an analyst in explaining complex system behavior by reducing the system to its 

40 major components and thereby improve analyst communications with stakeholders regarding 

41 the system being simulated. Additional information on model abstraction and simplification can 

42 be found in NUREG-7026 and NUREG-6884.

43

44 Model abstraction builds on insights gained from characterizing the system and developing 

45 exposure scenarios. In a dose assessment, several model abstractions typically support an 

46 overall assessment of a facility’s ability to demonstrate compliance with release criteria. These 

47 abstractions usually include models of projected climate and infiltration, source term release, 

48 transport through environmental media including the groundwater, and potential exposures to a 

49 receptor in the biosphere. If surface water or groundwater are viable pathways, a CSM of the 
F-3

1 hydrologic system is needed to adequately model these systems. Thus, adequate 

2 characterization of the subsurface is necessary to construct defensible hydrogeological 

3 conceptual models to perform dose modeling to estimate future impacts to surface water and 

4 groundwater, as well as to estimate doses from existing groundwater contamination. Although 

5 monitoring data can be useful in determining whether groundwater is contaminated and to 

6 determine the potential risk associated with groundwater dependent pathways, deficiencies in 

7 the monitoring well network and monitoring approach may compromise the ability to use existing 

8 monitoring well data for this purpose. In these cases, the licensee can manage uncertainty in 

9 exposure point concentrations through collection of additional data, and through additional 

10 modeling, including uncertainty/sensitivity analysis. Licensees should discuss acceptable 

11 approaches to assessing the risk and dose contributions associated with surface water and 

12 groundwater dependent pathways.

13

14 Environmental impact assessments for decommissioning may also benefit from the 

15 development of a CSM. The information needs of environmental impact assessments may 

16 differ slightly from a dose assessment for a safety case as non-radiological and other impacts 

17 associated with surface water and groundwater use must also be considered. Although specific 

18 to information needs to support preparation of environmental impact assessments, Regulatory 

19 Guide 4.2, Section 2.2, contains useful information regarding the collection of hydrological data 

20 for the purpose of environmental assessment, but may also be useful for creation of CSMs and 

21 development of dose assessment models (NRC, 2018). 

22

23 It is also important to note that transients and spatial and temporal variability may need to be 

24 considered when simulating contaminant flow and transport if important to decision-making. For 

25 example, river stage may control groundwater flow direction under a site with changes in flow 

26 direction occurring over smaller timeframes, but net flow may be towards the river considering 

27 longer timeframes. A determination will need to be made if representation of short-term

28 behavior is important to demonstrating compliance or will help provide a better understanding of 

29 system behavior and potential uncertainties in assessment of risk. Additional information on 

30 consideration of scale affects is provided in Appendix Q. 

31

32 F.4 Indicators for Potential Groundwater Contamination 

33 The level of effort associated with subsurface characterizations is dependent in large part on the 

34 extent of residual radioactivity in the subsurface and the transport of residual radioactivity to 

35 groundwater. As described in Table 1.1 of Volume 1 of this NUREG report, Decommissioning 

36 Groups 5–7 are sites that have the potential for residual radioactivity in groundwater. Based on 

37 the experience gained from operational and decommissioning NRC licensed sites, the following 

38 is a list of potential indicators for groundwater contamination at decommissioning sites

(NUREG-1496, Appendix C, Attachment E, Table C.E.1).

2 39 They are illustrative only and are not 

40 intended to constitute a complete list:

41 • High Potential: if a site has a history of, or currently has the following:

42 o unlined lagoons, pits, canals, or surface-drainage ways that received

43 radioactively contaminated liquid effluent

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). 
F-4

1 o lined lagoons, pits, canals, or surface drainage ways that received radioactively 

2 contaminated liquid effluent, where the lining has leaked or ruptured, or where 

3 overflow has occurred

4 o septic systems, dry wells, or injection wells that received radioactively

5 contaminated liquid effluent

6 o storage tanks, waste tanks, and/or piping (above or below ground) that held or

7 transported radioactively contaminated fluids and are known to have leaked

8 o liquid or wet radioactive waste buried on site (i.e., burial under previous

9 regulations found in 10 CFR 20.302 “Method for Obtaining Approval for Proposed

10 Disposal Procedures”, or 10 CFR 20.304 “Disposal by Burial in Soil”, (or the

11 current 10 CFR 20.2002, “Method for Obtaining Approval of Proposed Disposal

12 Procedures”))

13 o an accident or spill on site where radioactive material was released exterior to a

14 building

15 o wet bulk waste (e.g., sludge or tailings) stored exterior to buildings or used as

16 backfill

17 o containerized-liquid waste, stored exterior to buildings, that has leaked

18 • Medium Potential: if a site has a history of or currently has the following:

19 o surface water or atmospheric discharge of radioactive effluents including authorized

20 releases and spills (e.g., releases in compliance with 10 CFR Part 20, Appendix B

21 effluent concentrations or spills)

22 o radioactive contamination detected on the roof of a building

23 o radioactive contamination detected in the floor cracks or sump of a building

24 o an accident or spill on site, where liquid radioactive material was released to the

25 interior of a building

26 o the presence of aging unmonitored underground storage tanks or underground

27 piping that held radioactively contaminated fluids, and are not known to have

28 leaked (e.g., unmonitored tanks or piping which could have released radioactivity

29 due to corrosive environmental or service conditions, use of material types

30 potentially susceptible to corrosion [carbon steel], or design flaws)

31 o a history of incineration of radioactive waste exterior to onsite buildings

32 o dry bulk waste (e.g., sludge or tailings) stored exterior to buildings or used as

33 backfill

34 o solid containerized waste, stored exterior to buildings, that has leaked

35 • Low Potential: if a site has a history of or currently has the following:
F-5

1 o underground storage tanks or underground piping early in their service life that held 

2 radioactively contaminated fluids and are not known to have leaked

3 o dry bulk waste stored inside the buildings

4 o a sealed-source-only license

5 The potential for groundwater contamination at any of these sites is conditioned by certain site 

6 characteristics, such as depth of groundwater, amount of yearly precipitation, and hydraulic 

7 conductivity, and by certain source characteristics such as half-life, solubility, and distribution 

8 coefficient.

9 F.5 Groundwater Characterization

10 If existing groundwater contamination is known to be present, or if it is determined that residual 

11 radioactivity has a medium to high likelihood of reaching groundwater, radiological surveys of 

12 groundwater are expected. 

13 Characterization of groundwater contamination, including all significant radiological constituents, 

14 along with inorganic and organic constituents and related parameters, should be adequate to 

15 determine the following:

16 • extent and concentration distribution of contaminants

17 • source (known or postulated) of radioactive contaminants to groundwater

18 • background groundwater quality

19 • rate(s) and direction(s) of contaminated groundwater migration

20 • location of groundwater plume and concentration profiles (i.e., maximum concentration

21 in the vertical and lateral extent)

22 • assessment of present and potential future effects of groundwater withdrawal on the

23 migration of groundwater contaminants

24 • potential safety and environmental issues associated with remediating the surface water

25 and groundwater

26 • chemical form/speciation of the radionuclides; effect of the non-radiological constituents

27 on the mobility of the radionuclides

28 • whether the remediation activities and radiation control measures proposed by the

29 licensee are appropriate for the type and amount of radioactive material present in the

30 surface water and groundwater

31 • whether the licensee’s waste management practices are appropriate

32 • whether the licensee’s cost estimates are plausible
F-6

1 Besides licensee process discharges, other mechanisms may affect groundwater. For example, 

2 sumps that capture infiltrating groundwater may affect the local groundwater elevation during 

3 pumping. In some situations, sumps collect groundwater at the lowest elevation of a building, 

4 with pumping going on continuously. Such pumping has been shown to affect the local 

5 groundwater elevation (i.e., cone of depression). 

6 Typical analytical parameters include gross alpha particle activity, gross beta particle activity, 

7 specific radionuclide concentrations, gamma spectrum analysis for all gamma-emitting 

8 radionuclides suspected to be present, sulfate, chloride, carbonate, alkalinity, nitrate, total 

9 dissolved solids, total organic carbon, Eh, pH, calcium, sodium, potassium, iron, and dissolved 

10 oxygen. Additional analytical parameters may be necessary to characterize any suspected 

11 contamination. Other regulatory agencies that have jurisdiction over the decommissioning effort 

12 may require characterization of the non-radiological constituents and related parameters

13 including or in addition to those listed above. Therefore, licensees should contact Federal, 

14 State, or local government bodies responsible for regulating groundwater. 

15 The licensee should determine the extent of contamination and background groundwater quality 

16 based on groundwater monitoring data from a suitable monitoring well network. The reference 

17 section lists the following guidance documents on acceptable groundwater monitoring 

18 techniques: Korte and Ealey (1983), Korte and Kearl (1984), NUREG-6948, NUREG/CR-7221,

19 and NUREG/CR-1388, USGS (1977, 1996 and 2018), EPA (1977, 1980, 1985, 1986, and 

20 2009a), and ASTM D5092/D5092M (2016a). The actual number, location, and design of 

21 monitoring wells depend on the size of the contaminated area, the type and extent of 

22 contaminants, the background groundwater quality, the hydrogeologic system, and the 

23 objectives of the monitoring program. For example, if the objective of monitoring is only to 

24 indicate the presence of groundwater contamination, the licensee will need relatively few 

25 downgradient and upgradient monitoring wells. In contrast, if the objective is to develop a 

26 detailed characterization of the distribution of constituents within a complex aquifer as the 

27 design basis for a corrective action program, a large number of suitably designed and installed 

28 monitoring wells and well points may be necessary. 

29 Planned site characterization activities should be flexible enough to allow for the installation of 

30 additional monitoring wells during the characterization effort if either (1) preliminary 

31 characterization indicates contamination where previously unanticipated, or (2) there is a need 

32 to delineate the vertical or lateral extent of contaminant plumes. Monitoring well locations, 

33 contaminant concentrations, and contaminant sources should be plotted on a map (or a series 

34 of maps for multiple contaminants) to show the relationship among contamination, sources, 

35 hydrogeologic features and boundary conditions, and property boundaries. At sites with 

36 significant vertical migration of contaminants, the Decommissioning Plan (DP) should also 

37 provide hydrogeologic cross sections that depict the vertical distribution of contaminants in 

38 groundwater. The vertical exaggeration of the cross-sections should not exceed 10 times.

39
F-7

1 generic completion design.

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47 Figure F.1 General Monitoring Well Cross Section (Adapted from Figure 3 [NRC, 1994]) 

SATURATED ZONE VADOSE ZONE FROST 

ZONE ZONE OF LESSER PERMEABILITY SUMP/SEDIMENT TRAP

 BOTTOM CAP

 SURVEYOR'S PIN (FLUSH MOUNT) GAS VENT TUBE STEEL PROTECTOR CAP WITH LOCKS

 SCREENED INTERVAL

 POTENTIOMETRIC SURFACE

 ANNULAR SEALANT

 BOREHOLE DIAMETER = 10" TO 12"

 (NOMINAL DIMENSION)

 WELL DIAMETER = 4"

 CEMENT AND SODIUM

 BENTONITE MIXTURE

 CONTINUOUS POUR CONCRETE CAP AND

 WELL APRON (EXPANDING CEMENT)

 CONCRETE WELL APRON

 (MINIMUM RADIUS OF 3 FEET AND 4 INCHES THICK)

 WELL CAP 1/4" GAS VENT 

--- --- --- --- --- --- --- --- --- --- --- --- --- --- --- --- --- --- --- --- --- --- --- --- --- --- --- --- --- --- --- --- --- --- --- --- --- --- --- --- --- --- --- --- --- --- --- --- --- --- --- --- --- --- --- --- --- --- --- --- --- --- --- --- --- --- --- ---

--- --- --- --- --- --- --- --- --- --- --- --- --- --- --- --- --- --- --- --- --- --- --- --- --- --- --- --- --- --- --- --- --- --- --- --- --- --- --- --- --- --- --- --- --- --- --- --- --- --- --- --- --- --- --- --- --- --- --- --- --- --- --- --- --- --- --- ---

--- --- --- --- --- --- --- --- --- --- --- --- --- --- --- --- --- --- --- --- --- --- --- --- --- --- --- --- --- --- --- --- --- --- --- --- --- --- --- --- --- --- --- --- --- --- --- --- --- --- --- --- --- --- --- --- --- --- --- --- --- --- --- --- --- --- --- ---

 FILTER PACK (2 FEET OR 

 LESS ABOVE SCREEN)

decom-016.ppt

052902
F-8

1 The DP should also describe the groundwater characterization program used to characterize 

2 the extent and distribution of contaminants in the groundwater. Depending on the complexity of 

3 the site, the DP can include the detailed information as described below or can summarize this 

4 information and then reference the documents containing the supporting details. The 

5 description should provide monitoring well completion diagrams explaining elevation, internal 

6 and external dimensions, type of casings, type of backfill and seal, type of the screen and its

7 location and size, borehole diameter and elevation and depth of hole, and type and dimension 

8 of riser pipe and other necessary information on the wells. Figure F.1 illustrates an acceptable

9 generic design for well completion. The reference section lists the following documents that 

10 may contain useful information on monitoring well installation for specific objectives and 

11 geohydrologic conditions: ASTM (2016a, 2016b, and 2018a), EPA (1991a, and 2013), and 

12 USGS (1996). ASTM (2018b) may provide useful information on how to remove monitoring or 

13 remediation wells from service without creating a conduit for contaminant migration from the 

14 surface or between subsurface geologic units.

15 The DP should document or reference the sampling techniques, methodology, and procedures. 

16 Site characterization procedures and methods should generally adhere to national practices and 

17 standards (e.g., American Society for Testing and Materials (ASTM), the USGS, EPA, 

18 U.S. Department of Energy Environmental Monitoring Laboratory (DOE/EML), and National 

19 Institute of Standards and Technology (NIST)). The DP should identify specific analytical 

20 methods that conform to generally accepted protocols and methods, such as those endorsed by 

21 NIST and DOE/EML, or other methods established through a comprehensive peer review and 

22 recommendation process (e.g., American National Standards Institute (ANSI)/American Society 

23 of Mechanical Engineers (ASME), “Quality Assurance Program Requirements for Nuclear 

24 Facilities,” 1986). Korte and Kearl, “Procedures for the Collection and Preservation of 

25 Groundwater and Surface Water Samples and for the Installation of Monitoring Wells,” 1984, 

26 provides forms for documenting well summary information, samples, chain of custody, QA

27 information for field chemical analyses, and sample location and identifier.

28

29 The site characterization program should include sufficient sampling and analysis of 

30 groundwater samples collected upgradient from the site to develop a representative 

31 characterization of background groundwater quality. Background groundwater quality should 

32 not exhibit any influence from contaminants released by the site and should be representative of 

33 the quality of groundwater that would exist if the site had not been contaminated. The site 

34 characterization should also assess any temporal or spatial variations in background 

35 groundwater quality. If sources of contamination other than the site are present, the licensee 

36 should evaluate the potential impact of such sources to determine the degree of their 

37 groundwater contamination.

38 F.6 Monitoring Practices and Procedures 

39 Depending on the complexity of the site, the DP can include the detailed information as 

40 described below, or the DP can summarize this information and then reference the documents 

41 containing the supporting details.

42 The site characterization should include a description of all surface water and groundwater

43 characterization activities, methods, and monitoring installations sufficient to demonstrate that 

44 the methods and devices provided data that are representative of site conditions. It should also 

45 describe the monitoring practices, procedures, and QA programs used to collect water quality 

46 data. Monitoring well descriptions, for example, should include location, elevation, screened 

47 interval(s), depth, construction and completion details, and the hydrologic units monitored. 
F-9

1 Aquifer test descriptions should include testing configuration, test results, and a discussion of 

2 the assumptions, analytical techniques, test procedures, pretesting baseline conditions, 

3 limitations, errors in measurements, and final results. The description of the water quality 

4 sampling and analysis program should include or reference the procedures for sampling, 

5 preserving, storing, and analyzing the samples, including QA/QC protocols implemented. All 

6 methods should be consistent with current standard methods and practices (e.g., ASTM, USGS, 

7 EPA, NIST, and ANSI/ASME). The reference section lists the following documents containing 

8 additional guidance on applicable methods for sampling and analyzing water quality samples: 

9 Korte and Ealey (1983); Korte and Kearl (1984); DOE (1988 and 1993); ANSI/ASME (1986); 

10 EPA (1977, 1985, 1986, 1987a, 1991b); and NUREG-1293, NUREG-1383, and Regulatory 

11 Guide 4.15 (NRC, 1979). The licensee should document and explain any deviations from the 

12 standard methods used.

13 F.7 Monitoring Network and Sampling Frequencies 

14 The monitoring wells are usually placed in the following critical locations (NRC, 2007):

15

16 (1) Source areas,

17 (2) Within and immediately down-gradient of the source area;

18 (3) Fringe portions and boundary of the plume;

19 (4) Regulatory compliance boundaries.

20

21 To address the effect of physical/chemical heterogeneity on contaminant fate and transport, 

22 certain monitoring wells should be placed in relatively high transmissive zones with highest 

23 contaminant concentrations in the targeted monitoring areas; monitoring wells may need to be 

24 placed in locations that will reduce uncertainties with respect to geochemical conditions, and 

25 aquifer hydraulic properties. NUREG/CR-6948, NUREG/CR-7221, EPA (2009a), and Barcelona 

26 et al. (1989) provide information on how to design and implement a more optimal groundwater 

27 monitoring well network to refine site conceptual and numerical models and improve the 

28 capabilities for detection of contamination. 

29

30 It is important to note that operational environmental monitoring of groundwater, although 

31 adequate for its intended purpose, may not be adequate for site characterization and to support 

32 dose assessments. To support site characterization and dose assessments, information 

33 supplied by licensees may need to address the types and movement of radioactive 

34 contamination in groundwater at the facility, as well as the magnitude and extent of this 

35 contamination. The actual number, location, and design of monitoring wells depend on the size 

36 of the contaminated area, the nature and extent of contamination, the background quality, 

37 hydrogeologic system, and the objectives of the monitoring program. For example, if the only 

38 objective of monitoring is to indicate the presence of groundwater contamination, relatively few

39 downgradient and upgradient monitoring wells are needed. In contrast, if the objective is to 

40 develop a detailed characterization of the distribution of constituents within a complex aquifer as 

41 the design basis for a corrective action program, a relatively large number of suitably designed 

42 and installed monitoring wells may be necessary. Power reactors normally have groundwater

43 monitoring programs as part of their radiological environmental monitoring programs (REMPs). 

44 Although data derived from a REMP may provide useful information, the data still tend to be 

45 insufficient to allow the staff to fully understand the fates and transport of radioactive materials
F-10

1 in the subsurface environment at the facility, as well as the magnitude and extent of this 

2 contamination (see for example, RG 1.185, Post-Shutdown Decommissioning Activities Report, 

3 lessons learned on groundwater monitoring). Therefore, a licensee may need to gather 

4 additional data to address this lack of understanding.

5 If remediation is necessary, remedial performance confirmation monitoring may be conducted to 

6 demonstrate that remediation is occurring according to expectations. Although remedial 

7 expectations and, consequently, appropriate performance monitoring analyses are site-specific 

8 in nature, reduction in contaminant concentrations to specified levels is generally expected as 

9 selected remedies are being performed (Pope et al., 2004). Data analysis is useful in assessing 

10 progress toward contaminant reduction objectives include evaluation of temporal trends in 

11 contaminant concentrations or mass, comparisons of observed contaminant distributions with 

12 predictions or predefined targets, among other approaches. Evaluations of adequate progress 

13 toward restoration objectives are sometimes difficult due, in large measure, to subsurface 

14 spatial variability and to a lesser extent, measurement variability. This will often necessitate 

15 optimization of the employed remedial remedy, with relatively dense monitoring networks to 

16 reduce uncertainty to acceptable levels. 

17

18 The remedial action objective of attaining permitted standards, such as groundwater protection 

19 standards, or DCGLs should be demonstrated before monitoring and the license are terminated 

20 to ensure that the required standards are actually achieved in the long-term (Pope et al., 2004). 

21 The demonstration of the attainment of cleanup objectives should include sufficient verification 

22 monitoring once the standards are achieved to evaluate the effects of variations caused by 

23 active remedy. The length of the verification period should be based on site-specific conditions 

24 and on objective statistical analyses of the data. Site-specific conditions to consider include the 

25 response times of the hydrogeological system to seasonal or annual variations. Statistical 

26 methods useful in these evaluations include analyses of temporal trends in contaminant 

27 concentrations and comparisons with the specified concentration standards (e.g., Cohen et al., 

28 1994; EPA, 1992; and ITRC, 2013).

29

30 The licensee should establish surface water and groundwater quality and water levels on a set 

31 frequency, based on site-specific considerations. For sites with extensive groundwater

32 contamination, the licensee should design and install a network of monitoring wells to provide a 

33 high probability of detecting and characterizing existing contamination and determining 

34 background groundwater quality. It should measure groundwater levels in piezometers and 

35 monitoring wells that provide a sufficiently accurate change in hydraulic head to determine the 

36 horizontal hydraulic gradient within the uppermost aquifer and vertical hydraulic gradient with 

37 adjacent units. It should measure water levels at least on a quarterly basis for a minimum of 

38 one (1) year to determine temporal variations in the hydraulic conditions. After this period, it 

39 should adjust the frequency of water level measurements to reflect the anticipated impact on

40 hydraulic heads by site-specific events and conditions (e.g., tides, rises in river stage and bank 

41 storage, increased precipitation, water year variations). The initial monitoring program should 

42 be designed so that it can be integrated into the programs for operational and post-operational 

43 periods. The reference section lists the following documents containing applicable methods for 

44 groundwater sampling and for measuring water levels: EPA (1977, 1985, 1986, and 1987a), 

45 USGS (1977, 2018), and Korte and Kearl (1984), and NUREG-1388. 

46 The licensee should determine the sampling frequency for evaluating spatial and temporal 

47 variations in groundwater quality, including radiological and associated non-radiological 

48 constituents, based on the site-specific temporal variations in flow directions and 

49 hydrogeochemical conditions. After an initial sampling period in which each monitoring well is 
F-11

1 sampled at a frequency to establish site-dependent temporal variations throughout a year, it 

2 should collect and analyze representative samples generally on a quarterly basis from key 

3 monitoring wells to obtain representative estimates of the temporal variation of water quality in 

4 the uppermost aquifer and adjacent units. After this initial period, it should adjust the sampling 

5 frequency to reflect variations in the hydraulic gradient and hydrochemistry. If concentrations of 

6 principal radiological constituents change by more than about 10–20 percent between sampling 

7 events, it should increase the frequency of sampling in an attempt to characterize the temporal 

8 variability of groundwater quality. For unconfined groundwater systems, less than biannual 

9 sampling schedules may not capture important temporal variations. For most sites, it should be 

10 sufficient to sample on a quarterly basis (i.e., one sample per well per calendar quarter) to 

11 characterize temporal changes in water quality. More frequent sampling, such as bimonthly

12 may be necessary, however, especially at sites involving offsite or potential offsite 

13 contamination of groundwater resources with more reactive, geochemical constituents in a 

14 dynamic hydrological system (Barcelona et al., 1989). The sampling frequency needs to be 

15 evaluated with respect to the monitoring objective and time-frame over which the monitoring 

16 network will be conducted. NUREG/CR-6948, NUREG/CR-7221, EPA (2009a), and Barcelona 

17 et al. (1989) include acceptable approaches for assessing frequency of groundwater sampling.

18 Quarterly sampling of surface water and sediments should be sufficient at most sites. This 

19 sampling should be supplemented by additional sampling to characterize the surface water

20 system at representative low- or high-stage flow conditions (e.g., minimum annual, 7-day 

21 average low flow or maximum annual, 7-day average high flow). The licensee should use this 

22 information to bound the existing and projected impacts of the release of contamination on 

23 adjacent surface water bodies.

24 Note that the incremental benefit of sampling decreases with increasing sampling frequency 

25 when there is an autocorrelation in the data. If this autocorrelation is large, a relative low 

26 sampling frequency is necessary to avoid sampling redundancy, and the total length of the 

27 sampling period must increase to achieve a sufficient return on sampling. The adequacy of the 

28 sampling frequency needs to be interpreted or viewed in terms of the time horizon of the 

29 sampling program and monitoring objectives (Barcelona et al., 1989).

30 F.8 Surface Water and Sediments 

31 Surface water can include ponds, creeks, streams, rivers, lakes, coastal tidal waters, oceans, 

32 and other bodies of water. Note that certain ditches and intermittently flowing streams qualify as 

33 surface water. The licensee should evaluate the need for surface water samples on a case-by34 case basis. It should base its surveys for water on appropriate environmental standards for 

35 water sampling. If the body of water is included in a larger survey unit, then it should take 

36 sediment samples at sample locations selected by the normal method, without taking the body 

37 of water into consideration. In addition to the discussion below, MARSSIM (Sections 3.6.3.3 

38 and 5.3.3.3) provides some guidance on evaluating the likelihood for the release of 

39 radionuclides into surface water and sediments and on concerns related to characterization and 

40 sampling.

41 Onsite and offsite normal effluent discharges to surface water may be permitted under the 

42 license or by wastewater and stormwater permits and registrations. Radioactive Effluent 

43 Reports track these releases and provide dose assessment results for the appropriate scenario 

44 pathway. In addition, Radiological Environmental Monitoring Reports record radiological results 

45 of offsite sampling of water and sediments at discharge locations. For effluent discharges to 

46 onsite lakes or ponds at power plants, leakage to the groundwater may occur, though it need 
F-12

1 not be reported again during operations (NRC, 2017b). However, at decommissioning, the 

2 residual radioactive in the lake or pond sediments and any leakage to the groundwater must be 

3 treated as residual radioactivity that needs to be dispositioned in accordance with the guidance 

4 in this NUREG. For offsite effluent discharges, offsite decommissioning activities are not 

5 required by NRC regulations. State and local entities may require different treatment of offsite 

6 areas that are contaminated by normal effluent discharges. For NRC, however, characterization 

7 may be needed to assess environmental impacts as part of the environmental assessment or 

8 impact statement. Similarly, groundwater seepage of onsite residual contamination to offsite 

9 surface waters must be incorporated into environment assessments or impact statements. Any 

10 onsite groundwater contamination during decommissioning must be incorporated into the dose 

11 assessment and site release criteria. 

12 For sites that are located near surface water streams and could reasonably affect surface water 

13 pathways, the site characterization program should establish background surface water quality 

14 by sampling upstream of the site being studied or areas unaffected by any known activity at the 

15 site. Water should be collected as grab samples from the stream bank in a well-mixed zone. 

16 Depending on the significance and the potential for surface water contamination, it may be 

17 necessary for certain sites to collect stratified samples from the surface water to determine the 

18 distribution of contaminants within the water column. Surface water quality sampling should be 

19 accompanied by at least one round of stream sediment quality sampling to assess the 

20 relationship between the composition of the dissolved solids, the suspended sediment, and the 

21 bedload sediment fractions. The licensee should determine water levels and discharge rates of 

22 the stream at the time samples are collected and should also consider the effects of variability of 

23 the surface water flow rate. Based on the results of the HSA and preliminary investigation 

24 surveys, it should conduct surface scans for gamma activity in areas likely to contain residual 

25 activity (e.g., along the banks). Korte and Kearl (1984) and USGS (1977, 2018) describe 

26 applicable methods for surface water and sediment sampling. In addition, Fleischhauer and 

27 Engelder (1984) present suggested procedures for stream sediment sampling. The EPA 

28 guidance documents mentioned above are also applicable. In some cases, the Radiological 

29 Environmental Monitoring Program (REMP) data from a facility’s operating period may provide 

30 useful information to support the characterization program, although there are limitations of the 

31 use of REMP data to support site characterization. 

32 The licensee should conduct surface water sampling in areas of runoff from active operations. 

33 In case of direct discharge into a stream, it should monitor and sample the outfall and the 

34 stream upstream and downstream from the outfall. It should conduct preliminary 

35 characterization of the contamination levels by measuring gross alpha and total beta particle 

36 activity (total and dissolved) and by obtaining a gamma spectrum for surface water samples. It 

37 should be noted that determination of gross alpha activity (and low-energy beta emitters, as

38 well) may be of limited value for samples containing elevated total or dissolved solids 

39 concentrations because of sample attenuation. In such instances, gamma spectroscopy might 

40 be the only recourse. Specific radionuclide analysis may be needed, depending on the level of 

41 activities and type of radionuclides. Non-radiological parameters, such as specific conductance, 

42 pH, and total organic carbon, may be used as surrogate indicators of potential contamination, 

43 provided a clear relationship is established between radionuclide concentration and the level of 

44 the surrogate. Additional analysis for other parameters like volatile and semivolatile 

45 compounds, chelating agents, pesticides, and polychlorinated biphenyls may also be necessary 

46 if they affect the mobility of radiological constituents and to evaluate the potential environmental 

47 effects of decommissioning. 
F-13

1 The licensee should carefully record each of the surface water and sediment sampling locations 

2 on the appropriate survey form. It can also use surface water flow models to assist in 

3 estimating contaminant concentrations or migration rates.

4 F.9 Geochemical Conditions 

5 The licensee should also describe geochemical conditions at the site and their association with 

6 groundwater and contaminants, giving consideration specifically to geochemical conditions that 

7 enhance or retard contaminant transport. Geochemical data should include information on solid 

8 composition, buffering capacity, redox potential, pH, sorption (represented as a range of 

9 distribution coefficients (Kd) for each radiological constituent), and other relevant geochemical 

10 data. Piper and Stiff diagrams may be useful for visualizing the geochemistry of the water. In 

11 some cases, it may be necessary for licensees (or responsible parties) to use appropriate 

12 geochemical codes to understand and quantify geochemical mechanisms that significantly 

13 affect transport of radiological and non-radiological contaminants and their potential fate 

14 (e.g., geochemical speciation codes such as PHREEQC (Parkhurst and Appelo, 2013) are 

15 typically used for this purpose)). 

16 In general, licensees or responsible parties should estimate the values of Kd through laboratory 

17 column or batch sorption measurements considering site-specific conditions, if found to be 

18 important to estimating dose. If not found to be important to dose, site-specific Kd

19 measurements may not be necessary and generic ranges may be used from published tables 

20 (e.g., Sheppard and Thibault, 1990; EPA, 1999, 2004, 2005; and Yu et al., 2015). Licensees (or 

21 responsible parties) may use appropriate geochemical codes to understand and quantify 

22 geochemical mechanisms that significantly affect transport of radiological and non-radiological 

23 contaminants and their potential fate (e.g., MINTEQ (EPA 1984); EQ3/6 (Daveler and 

24 Woolery, 1992); PHREEQC (Parkhurst and Appelo, 2013)). For complex sites with the 

25 possibility of transient variable geochemical conditions, the use of surface complexation models 

26 in geochemical speciation models may be needed to adequately represent the variability of Kd

27 (e.g., NRC, 2003a). 

28 Appendix I of this volume discusses additional information on groundwater parameters 

29 necessary for dose modeling.

30 F.10 Surface Water and Groundwater Models to Support Dose Modeling

31 Appendix I contains additional information on hydrogeological conceptual models found in 

commonly used decommissioning dose modeling codes such as DandD3 32 (NRC, 2001) and 

33 RESRAD-ONSITE (Kamboj et al., 2018). The inherent assumptions and limitations of these 

34 models is addressed. RESRAD-ONSITE does not consider existing groundwater contamination 

35 and only addresses the potential, future transport of residual radioactivity and contamination of 

ground and surface water and associated doses.4 36 

37 In cases where existing groundwater contamination is present, it is likely necessary to construct 

38 a more complicated groundwater model to support development of DCGLs. Additionally, in 

3 Decommissioning codes such as DandD can be obtained from NRC’s RAMP (Radiation Protection Computer 

Code Analysis and Maintenance Program).

4 RESRAD-OFFSITE 3.2 considers sources located below the water table, although guidance on potential use of 

RESRAD-OFFSITE to consider existing groundwater contamination 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. 
F-14

1 these cases some allowance for the existing groundwater contamination must be made (e.g., 

2 fractionation of the release criteria to account for multiple contaminated media) to ensure the 

3 potential dose is not underestimated. Although a licensee always has the option of assessing 

4 the peak dose within the 1000-year compliance period based on more realistic modeling, in 

5 many cases it may be more straightforward to perform more bounding type analyses to simplify 

6 the decommissioning process if compliance can be easily demonstrated.

7 As a joint effort, the NRC, EPA, and DOE have developed specific guidance on selecting and 

8 applying surface water and groundwater models (EPA 1994a, b, c). The reference section lists 

9 other documents containing supporting details (NUREG/CR-6805 and NUREG/CR-6948; ASTM 

10 D5447-17 (2018c), NCRP 1985, 1996; EPA 1987a, 1987b, 2009b; NAS 1999).

11

12 As previously discussed, model simplification is the process for reducing the complexity of a 

13 numerical model into a simpler numerical model while still maintaining the validity of the 

14 simulation results. Although specific computer codes may be discussed or referenced in this 

15 guidance, the NRC does not endorse the use of any particular code or modeling software 

16 package for analyzing the performance of a decommissioning or disposal site. However, it 

17 would be useful to discuss commonly used codes for developing groundwater models by NRC 

18 licensees and to know which codes are better suited for constructing complex models and which 

19 codes usually provide good results when used to construct simple models. A discussion of 

20 commonly used codes and a comparison of the model results obtained by codes used to 

21 construct complex models and the results from codes used for simple models should provide for 

22 a better understanding of the model simplification process and information on how model 

23 simplifications can affect the modeling results.

24

25 F.10.1 Codes for Development of More Complex Models

26 The MODFLOW and MT3DMS codes are commonly used to simulate multi-dimensional flow 

27 and transport in the subsurface. These codes can simulate three-dimensional flow and 

28 transport, and therefore can generally better project contaminant concentrations at various 

29 points of exposure compared to codes that are only able to simulate flow and transport in one30 dimension. Of course, sufficient data must be available to construct these more complex 

31 models, or the additional code capabilities are of limited value. In some cases, it may be 

32 preferable or necessary to use simpler codes to simulate groundwater flow and transport when 

33 adequate safety margin is available and more conservative assumptions can be made, or when 

34 data or resources are limited making it difficult for an analyst to construct a more complex model 

35 of the site. However, when additional complexity is warranted there are several off the shelf 

36 codes available to assist with constructing these more complex three-dimensional groundwater 

37 flow and transport models.

38

39 The MODFLOW code is a groundwater flow modeling code developed by the USGS that solves 

40 the groundwater flow equation in up to three dimensions using finite-difference approximations. 

41 The code is public domain free software written primarily in FORTRAN and can be compiled 

42 and run on Microsoft Windows or Unix-like operating systems. The first version of MODFLOW 

43 was published in 1984. It has a modular structure and can be modified to address different 

44 applications or problems. MODFLOW-2005 has many new capabilities compared to older 

45 versions (Harbaugh, 2005). MODFLOW 6 is the current core MODFLOW version distributed by 

46 the USGS (Langevin, et al., 2017; Hughes, et al., 2017). The previous core version, 

47 MODFLOW-2005, is actively maintained, supported, and able to handle supplemental 

48 MODFLOW features not yet updated for MODFLOW 6. 

49
F-15

1 The MODFLOW-2005 code simulates steady and non-steady flow in an irregularly shaped flow 

2 system in which aquifer layers can be confined, unconfined, or a combination of confined and 

3 unconfined. Flow from external stresses, such as flow to wells, areal recharge, 

4 evapotranspiration, flow to drains, and flow through river beds, can be simulated. Hydraulic 

5 conductivities or transmissivities for any layer may differ spatially and be anisotropic (restricted 

6 to having the principal directions aligned with the grid axes), and the storage coefficient may be 

7 heterogeneous. Specified head and specified flux boundaries can be simulated as can a head 

8 dependent flux across the model's outer boundary. MODFLOW 6 uses a new format of blocks 

9 and keywords for input of model data and is written using an object-oriented design. 

10 MODFLOW 6 uses a control-volume finite difference approach that can utilize structured or 

11 unstructured grids; MODLFOW-2005 requires a Cartesian grid system. MODFLOW 6 presently 

12 supports one type of process model — the Ground Water Flow Model (GWF). Multiple GWF 

13 models can be coupled and concurrently simulated, such as nesting of models at different 

14 scales. Other models may be added in the future that can seamlessly couple with the GWF, 

15 such as a groundwater transport model, a surface-water model, and a pipe network model. The 

16 new MODFLOW 6 framework will enable the ability to solve multiple, tightly coupled, numerical 

17 models in a single system of equations, or simplify the computational effort by loosely coupling 

18 the different models. For continuity, licensees may choose to continue using MODFLOW-2005; 

19 but if starting fresh or the site has hydrogeologically-significant features that exhibit as complex 

20 geometries, licensees should consider using MODFLOW 6.

21

22 Other options exist for sites with complex geometries. Codes developed using finite volume or 

23 finite element representations may be needed to better represent the geometries of the system. 

24 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 5 25 finite-element code has 

26 been used by DOE due to its ability to consider radioactive decay, can solve problems involving 

27 transient or steady state fluid flow, and mass transport in multi-phase, variably saturated, porous 

28 or fractured media. The porous/fractured media may be anisotropic and heterogeneous, 

29 arbitrary sources (injection or pumping wells) may be present and, chemical reactions or 

30 radioactive decay may take place. The geometry may be 2D or 3D, Cartesian or Cylindrical and 

31 the mesh may be structured or unstructured.

32

33 The MT3DMS code is a modular three-dimensional transport model for the simulation of 

34 advection, dispersion, and chemical reactions of dissolved constituents in groundwater systems 

35 (Bedekar, et al., 2016). The MT3DMS code uses a modular structure similar to the structure 

36 utilized by MODFLOW. The MT3DMS code is used in conjunction with MODFLOW in a two37 step flow and transport simulation. Heads and cell-by-cell flux terms are computed by 

38 MODFLOW during the flow simulation and are written to a specially formatted file. This file is 

39 then read by MT3DMS and utilized as the flow field for the transport portion of the simulation. 

40 The MT3DMS code differs from MT3D in that it allows for multi-species transport, supports 

41 additional solvers, and allows for cell-by-cell input of all model parameters.

42

43 F.10.2 Codes Typically Used in Decommissioning

44 The RESRAD-ONSITE (Kamboj et al., 2018) and RESRAD-OFFSITE (NRC, 2020) codes are 

45 commonly used for calculating groundwater concentrations and groundwater-dependent dose 

46 for decommissioning dose assessments. Details of the RESRAD-ONSITE groundwater models

47 are discussed in Appendix I. 

48

5 PORFLOW was developed by Analytic & Computational Research, Inc. (ACRi, 2008).
F-16

1 The RESRAD-ONSITE code includes a mass balance and non-dispersion model to compute 

2 groundwater concentrations at an onsite well. Neither the RESRAD-ONSITE mass balance, nor 

3 the non-dispersion model, considers dispersion, while RESRAD-OFFSITE can consider 

4 advection and dispersion in calculating groundwater concentrations at a receptor well. The 

5 groundwater transport model in RESRAD-OFFSITE considers 1-D advection (straight or curved 

6 flow path), along with 3-D dispersive transport in the saturated zone. Likewise, while only 1-D 

7 advection is considered in RESRAD-ONSITE, RESRAD-OFFSITE considers 1-D advection, and 

8 1-D dispersive transport in the unsaturated zone. Furthermore, the unsaturated zone, saturated

zone, and contaminated zone6 9 can be subdivided into smaller zones to increase the accuracy of

transport simulations.7 10

11

12 Both RESRAD-ONSITE and RESRAD-OFFSITE have the capability to consider variable 

13 transport rates of progeny created during transport in groundwater. The RESRAD-OFFSITE 

14 code has two groundwater transport algorithms: the first algorithm considers variable transport 

rates of parents and progeny, and the second algorithm models longitudinal dispersion8 15 . When 

16 either variable transport rates or dispersion is clearly dominant, the RESRAD-OFFSITE user 

17 should choose the transport algorithm that is most important to increase computational 

18 efficiency. When both the longitudinal dispersion and the variable transport rates are important, 

19 the user has the option of subdividing the transport pathway into a number of subzones to more 

20 accurately simulate the transport of progeny in transport, although this approach may 

21 significantly increase computation times. Only the zone where the progeny atoms are created 

22 would not consider both processes (longitudinal dispersion and variable transport rates of

23 parents and daughters).

24

25 Benchmarking of RESRAD-ONSITE with RESRAD-OFFSITE (onsite simulation)

26

27 Code developers benchmarked RESRAD-OFFSITE against RESRAD-ONSITE (Yu, 2006) using

28 an earlier version of the RESRAD-OFFSITE code (i.e., benchmarking was conducted prior to 

29 the 2007 release of RESRAD-OFFSITE 2.0). The results of the benchmarking exercises 

30 showed that RESRAD-OFFSITE could mimic the results of RESRAD-ONSITE when certain 

31 parameters were changed consistent with the RESRAD-ONSITE conceptual model. Notable 

32 differences between initial simulations run with RESRAD-ONSITE and RESRAD-OFFSITE 

33 included travel times to the point of compliance that were attributable to differences in use of 

34 porosity in the transport calculations (i.e., effective porosity is used in RESRAD-ONSITE, while 

35 total porosity is used in RESRAD-OFFSITE). Another noteworthy difference in results was 

36 observed for the water-dependent pathways due to accumulation of radioactivity in soil from 

37 application of contaminated irrigation water that is considered in RESRAD-OFFSITE but is not 

38 considered in RESRAD-ONSITE.

39

40

6 The capability to more accurately simulate transport through the contaminated zone through use of sub-zones was 

added in RESRAD-OFFSITE, Version 3.0

7 It is important to note that RESRAD-OFFSITE also has the capability of mimicking the RESRAD 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.

8 Both RESRAD-OFFSITE transport algorithms account for transverse dispersion.
F-17

1 Benchmarking of RESRAD-OFFSITE with Other Codes

2

The RESRAD-OFFSITE9 3 code was also benchmarked with several other codes as detailed in 

4 Gnanapragasam et al. (2000). The benchmarking study is instructive as it provides extensive 

5 discussion on the differences in modeling results for more complex and less complex models. 

6 Specifically, the model comparison study illustrates how differences between codes in their 

7 treatment of longitudinal and transverse dispersion and differences in ability to model variable 

8 transport rates of parents and daughters impact the results. The radionuclides considered in 

9 the benchmarking study included relatively short-lived Sr-90, and U-234 decay chain members 

10 (U-234, Th-230, Ra-226, Pb-210, and Po-210). Source loading spanned 200 years for Sr-90 

11 and 100 years for U-234 (at a constant rate of 0.88 Ci/y). The RESRAD-OFFSITE, PRESTO, 

12 MMSOILS, and MEPAS codes were the four primary codes tested in this study. They have 

13 different capabilities with respect to simulation of dispersion and variable transport rates of 

14 progeny. Increasing dispersivity with distance, and variable pore water velocity along the flow 

15 path were also considered in this study.

16

17 The results of the comparison showed markedly different results for PRESTO at increasing 

18 distance from the source, and at longer times after source loading ceases (when longitudinal 

19 dispersion is important). The PRESTO longitudinal profiles and breakthrough curve results are 

20 different from other codes because PRESTO is the only code of the four codes tested that does 

21 not consider longitudinal dispersion. However, at shorter distances and for constant sources, 

22 the differences between PRESTO and the other codes are less significant.

23

24 Another important difference between the codes tested is related to the calculation of 

25 retardation. Because RESRAD-OFFSITE is the only code to consider immobile pore water in 

26 the calculation of the retardation factor, the calculated retardation factors can be significantly 

27 higher for radionuclides with low retardation factors and if the effective and total porosity are 

28 significantly different (e.g., in the Gnanapragasam et al. (2000) study RESRAD-OFFSITE

29 calculated retardation factors are higher by a value of 0.56 (unitless) for all radionuclides due to 

30 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-lived10 31 radionuclides.

32

33 All of the codes are slightly different when it comes to important processes for progeny created 

34 in transport (i.e., longitudinal dispersion and consideration of variable transport rates of parents 

35 and progeny). The PRESTO code considers neither longitudinal dispersion, nor variable 

36 transport rates of progeny. The version of RESRAD-OFFSITE tested in the study only 

37 considered variable transport rates of daughters but not longitudinal dispersion of progeny 

38 (additional features were added to RESRAD-OFFSITE in subsequent versions as described 

39 above). When longitudinal dispersion is not considered, a sharp front is observed in the 

40 breakthrough curves of progeny; however, if variable transport rates of progeny are considered, 

41 as is the case for RESRAD-OFFSITE, higher retardation factors of progeny (e.g., Th-230 has a 

42 retardation factor of 198 compared to a retardation factor of 48 for U-234) can lead to long tails 

43 in the breakthrough curves. RESRAD-OFFSITE’s capability of modeling variable transport rates

44 of progeny also leads to lower peak concentrations of radionuclides such as Th-230, which has 

45 a higher retardation factor than its parent. On the other hand, higher peak concentrations of 

46 radionuclides such as Ra-226, Pb-210, and Po-210 result due to (1) the high retardation factor 

47 of Th-230, and (2) the lower retardation factors of the radionuclides compared to their 

9 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.

10 Short-lived with respect to the observation time or transport time at the point of observation.
F-18

1 immediate parents. Improvements to the accuracy of the breakthrough curves were observed 

2 when the aquifer was broken up into smaller segments and longitudinal dispersion and variable 

3 transport rates of progeny (compared to parents) was considered in all segments with the 

4 exception of the segment of transformation (where either longitudinal dispersion or variable 

5 transport rates of the progeny were considered). Presumably, this study led to improvements in 

6 the progeny transport models for some of the evaluated codes, including more recent versions 

7 of RESRAD-OFFSITE.

8

9 The various codes studied also differ with respect to the consideration of dilution in a pumping 

10 well (i.e., some codes consider just aquifer concentrations, while other codes such as RESRAD11 OFFSITE and PRESTO can consider dilution in a well). Depending on the location of the well 

12 and other factors, the impact of pumping may be more or less significant. For example, the 

13 study concludes that at exposure points far away from the source, transverse dispersion can be 

14 expected to distribute contamination nearly uniformly near the well given the size of the plume in 

15 relation to the capture zone. On the other hand, if the well is located close to the source and the 

16 plume is narrow or not well-developed, then pumping may lead to dilution of the narrow 

17 contaminant plume via mixing with clean water in the larger (than plume) volume of water pulled 

18 in by the well.

19

20 A prototype version of RESRAD-OFFSITE was also evaluated in a study conducted by the 

21 Biospheric Model Validation Study II Working Group on Uranium Mill Tailings (BIOMOVS II 

22 1996). The predictions of well water concentrations and offsite soil accumulation made by the 

23 prototype version of RESRAD-OFFSITE were stated by code developers to compare well with 

24 the predictions of other codes participating in the study (Yu, 2006).
F-19

1 F.11 References

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9

10 ASTM D5092/D5092M, “Standard Practice for Design and Installation of Groundwater 

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13 ASTM D6724/D6724M − 16, “Standard Guide for Installation of Direct Push Groundwater 

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16 ASTM D5521/D5521M − 18, “Standard Guide for Development of Groundwater Monitoring 

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19 ASTM D5299/D5299M – 18, “Standard Guide for Decommissioning of Groundwater Wells, 

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30 Bedekar, V., Morway, E.D., Langevin, C.D., and Tonkin, M. “MT3D-USGS Version 1: A U.S. 

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35 BIOMOVS II, “Long-Term Contaminant Migration and Impacts from Uranium Mill Tailings —

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25 Standards, Volume 2: Ground Water.” EPA, Office of Policy, Planning, and Evaluation, 

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27

28 — — — — —. EPA 402-R-94-012, “A Technical Guide to Ground-Water Model Selection at 

29 Sites Contaminated with Radioactive Substances.” EPA: Washington, DC. June 1994a.

30

31 — — — — —. EPA 500-B-003, “Environmental Protection Agency, Assessment Framework for 

32 Ground-Water Model Applications.” EPA: Washington, DC. 1994b.

33 — — — — —. EPA 500-B-004, “Environmental Protection Agency, Ground-Water Modeling 

34 Compendium: Fact Sheets, Descriptions, Applications, and Costs Guidelines.” EPA: 

35 Washington, DC. 1994c.

36 — — — — —. EPA 402-R-99-004B. “Understanding Variation in Partition Coefficient, Kd, 

37 Values, Volume II: Review of Geochemistry and Available Kd Values for Cadmium, Cesium, 

38 Chromium, Lead, Plutonium, Radon, Strontium, Thorium, Tritium (3H), and Uranium.” EPA,

39 Office of Air and Radiation: Washington, DC. 1999.
F-21

1 — — — — —. EPA 402-R-04-002C. “Understanding Variation in Partition Coefficient, Kd, 

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4 Washington, DC. 2004.

5 — — — — —. EPA/600/r-05/074. “Partition Coefficients for Metals in Surface Water, Soil, and 

6 Waste.” EPA, Office of Research and Development: Washington, DC. 2005.

7 — — — — —. EPA 530/R-09-007, “Statistical Analysis of Groundwater Monitoring Data at 

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9 Washington, DC. 2009a. 

10 — — — — —. EPA/100/K-09/003, “Guidance on the Development, Evaluation, and Application 

11 of Environmental Models.” EPA, Council for Regulatory Environmental Modeling: Washington, 

12 DC. 2009b.

13 — — — — —. SESDGUID-101-R1, “Design and Installation of Monitoring Wells.” EPA: 

14 Athens, Georgia. 2013. 

15 Fleischhauer, C. and T. Engelder. “Procedures for Reconnaissance Stream-Sediment 

16 Sampling.” Bendix Field Engineering Corporation. GJ/TMC-14. 1985.

17 Gnanapragasam, E.K., et al. “Comparison of Multimedia Model Predictions for a Contamination 

18 Plume Migration Scenario,” Journal of Contaminant Hydrology 46(1–2):17–38. 2000.

19

20 Harbaugh, A.W., “MODFLOW-2005, The U.S. Geological Survey Modular Ground-water 

21 Model—the Ground-Water Flow Process: U.S. Geological Survey Techniques and Methods 6-

22 A16.” USGS: Reston, Virginia. 2005.

23

24 Hughes, J.D., Langevin, C.D., and Banta, E.R. “Documentation for the MODFLOW 6 

25 Framework” U.S. Geological Survey Techniques and Methods, Book 6, Chapter A57, 40 pages.

26 USGS: Reston, VA. https://doi.org/10.3133/tm6A57. 2017.

27

28 International Atomic Energy Agency (IAEA). "Safety Assessment Methodologies for Near 

29 Surface Disposal Facilities (ISAM), Vol. I – Review and Enhancement of Safety Assessment 

30 Approaches and Tools; Vol. II – Tests Cases." IAEA: Vienna, Austria. 2004.

31 Interstate Technology & Regulatory Council (ITRC). “Groundwater Statistics and Monitoring 

32 Compliance, Statistical Tools for the Project Life Cycle.” GSMC-1. ITRC: Washington, DC. 

33 2013. 

34

35 Kamboj, S., E. Gnanapragasam, and C. Yu. “User’s Guide for RESRAD-ONSITE Code Version 

36 7.2. Argonne National Laboratory ANL/EVS/TM-18/1. ANL, Environmental Science Division: 

37 Argonne, IL. 2018.

38

39 Korte, N. and D. Ealey. “Procedures for Field Chemical Analyses of Water Samples.” Bendix 

40 Field Engineering Corporation. Technical Measurements Center Report GJ/TMC-07. DOE: 

41 Grand Junction, CO. 1983.

42 Korte, N. and P. Kearl. “Procedures for the Collection and Preservation of Groundwater and 

43 Surface Water Samples and for the Installation of Monitoring Wells.” Bendix Field Engineering 
F-22

1 Corporation. Technical Measurements Center Report GJ/TMC–08. DOE: Grand Junction, CO. 

2 1984.

3 Langevin, C.D., Hughes, J.D., Banta, E.R., Niswonger, R.G., Panday, Sorab, and Provost, A.M. 

4 “Documentation for the MODFLOW 6 Groundwater Flow Model.” U.S. Geological Survey 

5 Techniques and Methods, Book 6, Chapter A55, 197 pages. USGS: Reston, VA. 

6 https://doi.org/10.3133/tm6A55. 2017.

7 National Academy of Sciences (NAS). “Risk Assessment of Radon in Drinking Water.” National 

8 Academy Press: Washington, DC. 1999.

9 National Council on Radiation Protection and Measurements (NCRP). NCRP Report No. 76, 

10 “Radiological Assessment: Predicting the Transport, Bioaccumulation and Uptake by Man of 

11 Radionuclides Released to the Environment.” NCRP Publications: Bethesda, Maryland. 1985.

12 — — — — —. NCRP Report No. 123, Vol. 2, “Screening Models for Releases of Radionuclides 

13 to Atmosphere, Surfaces, Water, and Ground.” NCRP Publications: Bethesda, Maryland. 

14 1996.

15 Nuclear Regulatory Commission (U.S.) (NRC). Regulatory Guide 4.15, Rev. 1, “Quality 

16 Assurance for Radiological Monitoring Programs (Normal Operations)—Effluent Streams and 

17 Environment.” NRC: Washington, DC. February 1979.

18 — — — — —. NUREG-1293, “Quality Assurance Guidance for Low-Level Radioactive Waste 

19 Disposal Facility.” NRC: Washington, DC. January 1989a.

20 — — — — —. NUREG-1383, “Guidance on the Application of Quality Assurance for 

21 Characterizing a Low-Level Radioactive Disposal Site.” NRC: Washington, DC. 

22 November 1989b.

23 — — — — —. NUREG-1388, “Environmental Monitoring of Low-Level Radioactive Waste 

24 Disposal Facility.” NRC: Washington, DC. December 1989c.

25 — — — — —. “Draft Branch Technical Position on Site Characterization for Decommissioning.” 

26 NRC: Washington, DC. November 1994.

27 — — — — —. NUREG-1496, “Generic Environmental Impact Statement in Support of 

28 Rulemaking on Radiological Criteria for License Termination of NRC–Licensed Nuclear 

29 Facilities.” NRC: Washington, DC. July 1997.

30 — — — — —. NUREG-1575, Rev.1. “Multi-Agency Radiation Survey and Site Investigation 

31 Manual (MARSSIM)”. EPA 402-R-97-016, Rev.1. DOE/EH-0624, Rev.1. NRC: Washington, 

32 DC. 2000.

33 — — — — —. NUREG/CR-5512, Vol. 2, “ Residual Radioactive Contamination From 

34 Decommissioning: User's Manual DandD Version 2.1.” NRC: Washington DC. 2001.

35 — — — — —. NUREG/CR-6820. “Application of Surface Complexation Modeling to Describe 

36 Uranium (VI) Adsorption and Retardation at the Uranium Mill Tailings Site at Naturita, 

37 Colorado.” NRC: Washington DC. 2003a.

38
F-23

1 — — — — —. NUREG/CR-6805, “A Comprehensive Strategy of Hydrogeologic Modeling and

2 Uncertainty Analysis for Nuclear Facilities and Sites.” NRC: Washington, D.C. July 2003b.

3

4 — — — — —. NUREG/CR-6884, “Model Abstraction Techniques for Soil-Water Flow and 

5 Transport.” NRC: Washington, D.C. December 2006. 

6

7 — — — — —. NUREG/CR-6948, Volume 1, “Integrated Ground-Water Monitoring Strategy for 

8 NRC-Licensed Facilities and Sites: Logic, Strategic Approach and Discussion.” NRC: 

9 Washington, DC. 2007.

10 — — — — —. NUREG/CR-7026, “Application of Model Abstraction Techniques to Simulate 

11 Transport in Soils.” NRC: Washington, D.C. March 2011. 

12

13 — — — — —. NUREG/CR-7221, “Integrating Model Abstraction into Subsurface Monitoring 

14 Strategies.” NRC: Washington, D.C. March 2017a.

15

16 — — — — —. Regulatory Guide 4.25, Rev. 0, “Assessment of Abnormal Radionuclide

17 Discharges in Ground Water to the Unrestricted Area at Nuclear Power Plant Sites.” NRC: 

18 Washington, DC. 2017b.

19 — — — — —. Regulatory Guide 4.2, Rev. 3, “Preparation of Environmental Reports for 

20 Nuclear Power Stations.” NRC: Washington, DC. September 2018.

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31

32 Pruess, K., C. Oldenburg, and G. Moridis. “TOUGH2 User’s Guide, Version 2.” Lawrence 

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F-24

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4

5 Yu, C., E. Gnanpragasam, J.J. Cheng, and B. Biwer. “Benchmarking of RESRAD-OFFSITE: 

6 Transition from RESRAD (onsite) to RESRAD-OFFSITE and Comparison of the RESRAD7 OFFSITE Predictions with Peer Codes”, ANL/EVS/TM/06-3, DOE/EH-0708, Environmental 

8 Science Division, Argonne National Laboratory. 2006. 

9

10 Yu, C., S. Kamboj, C. Wang, and J-J. Cheng. “Data Collection Handbook to Support Modeling 

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SPECIAL ISSUES ASSOCIATED WITH DOSE MODELING, 

CHARACTERIZATION, AND SURVEY

G-1

1 G.1 Introduction

2

3

4

5

6

7

8

9

10

11

12

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, “MultiAgency Radiological Survey and Site Investigation Manual (MARSSIM), Revision 1, issued 

August 2000. In these situations, licensees may need to perform characterization and a FSS to 

demonstrate compliance with the license termination criteria in 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. 

13 This Appendix G applies, either in total or in part, to Decommissioning Groups 4–7.

14 G.2 Surveys for Special Situations in Buildings 

15 The survey method described thus far in this volume (e.g., Chapter 4 and Appendix A) applies 

16 to simple ideal geometries in a straightforward manner; however, there are likely to be some 

17 additional special situations at actual sites that will need further consideration. For each 

18 situation discussed below, it is assumed that the HSA and minimal site characterization have 

19 located and given a rough estimate of the concentration of residual radioactivity present.

20 G.2.1 Structures Versus Equipment

21 G.2.1.1 Background

22 The NRC staff acknowledges that the relationship between the LTR for unrestricted use of a site 

23 (dose criteria of 0.25 mSv/y (25 mrem/y) and ALARA found in 10 CFR 20.1402), and existing 

24 guidance for unrestricted releases of solid materials from a site on a case-by-case basis under 

25 10 CFR 20.2002 may have been unclear. In particular, the criteria for the LTR and for releases 

26 of solid materials with small or no amounts of residual contamination before license termination 

27 are different. Consistent with the LTR, once a site meets the radiological criteria for unrestricted 

28 use and the NRC terminates the license, solid material may be removed from a site. However, 

29 before license termination, material cannot be removed from the site for unrestricted use unless 

30 it meets either (1) criteria already approved for the licensed facility (e.g., in a license condition), 

31 for surficially contaminated materials, or (2) the few mrem/y criterion for the case-by-case 

32 approach for volumetrically contaminated materials (see Section 15.11 in Volume 1 of this 

33 NUREG). One rationale for the difference in criteria is that the technical basis for the LTR 

34 assumes that individuals are generally exposed to residual radioactivity at a single location (the 

35 site), while, for releases of solid material, an individual may be exposed to materials through 

36 several exposure scenarios at offsite locations. For more information about the relationship 

37 between the LTR and the case-by-case approaches to release of solid materials from a site, see

38 the LTR Analysis Commission Paper, SECY-03-0069, “Results of the License Termination Rule 

39 Analysis,” dated May 2, 2003, and the associated Regulatory Issue Summary 2004-08, “Results 

40 of the License Termination Rule Analysis,” issued May 2004. 

41 This section focuses on compliance with the LTR, in particular, the building structure-related 

42 materials that may be left on site at license termination and the criteria that should apply. 

43 Section 15.11 of Volume 1 of this NUREG provides more information about current approaches 

44 to releases of solid material before license termination.
G-2

1 G.2.1.2 Implementation

2 The LTR applies to building structures that remain in place after decommissioning and does not 

3 apply to releases of equipment from the facility before license termination. If licensees elect to 

4 dismantle building structures and dispose of the associated materials off site (in accordance 

5 with applicable regulatory requirements), rather than leave the building structures in place (for 

6 unrestricted use), the LTR does not apply to the associated materials moved off site before 

7 license termination. Materials licensees may release equipment and building structure 

8 deconstruction and dismantlement materials in accordance with existing license conditions. 

9 Reactor licensees (licensed under 10 CFR Part 50, “Domestic Licensing of Production and 

10 Utilization Facilities”) may release equipment and building structure deconstruction and 

11 dismantlement materials in accordance with the guidance in Inspection and Enforcement 

12 Circular 81-07, “Control of Radioactively Contaminated Material,” dated May 14, 1981; 

13 Information Notice 85-92, “Surveys of Wastes Before Disposal from Nuclear Reactor Facilities,” 

14 dated December 2, 1985; and Information Notice 88-22, “Disposal of Sludge from Onsite 

15 Sewage Treatment Facilities at Nuclear Power Stations,” dated May 12, 1988. Licensees 

16 should refer to Section 15.11 of Volume 1 of this NUREG report and should contact the NRC 

17 staff for further guidance on equipment and solid material releases.

18 When the LTR was developed, the NRC assumed that decommissioning generally would 

19 include the removal of systems and components from onsite buildings before license 

20 termination. However, with experience, it has become clear that each licensee uses a different 

21 approach for decommissioning, and these approaches are not necessarily consistent with the 

22 original assumptions of the LTR. Differences are the result of factors such as (1) the potential 

23 for reuse of systems and components, (2) cost of recycling and price of scrap metal and 

24 concrete, and (3) cost and availability of disposal options.

25 It is clear from the LTR technical basis in NUREG-1496, “Generic Environmental Impact 

26 Statement in Support of Rulemaking on Radiological Criteria for License Termination of NRC–

27 Licensed Nuclear Facilities,” issued July 1997, and NRC draft Regulatory Guide DG-4006, 

28 “Demonstrating Compliance with the Radiological Criteria for License Termination,” issued 

29 July 1998, that the LTR was not intended to apply to releases of “equipment” from the facility. 

30 “Equipment” includes anything not attached to, or not an integral part of, the building structure. 

31 On the other hand, previous guidance (the previous version of Section G.1.1 of NUREG-1757, 

32 Vol. 2, Rev. 0) was not prescriptive enough to provide a definitive answer about whether 

33 systems and components must be considered “building structures” or “equipment.” The 

34 previous guidance considered “doors, windows, sinks, lighting fixtures, utility lines, built-in 

35 laboratory hoods and benches, and other types of built-in furniture” to be part of the structure. 

36 Under that guidance, those items could be included in the FSS and left in place at license 

37 termination. It could be argued that, based on the examples provided, many plant systems and 

38 components also could be considered “building structures,” and, therefore, left in place at 

39 license termination. This previous guidance may have been inconsistent with the discussion in 

40 the LTR Analysis Commission Paper, SECY-03-0069, which described an expectation that 

41 removable materials and equipment would generally not be present at the time of license 

42 termination. 

43 The staff has identified a number of acceptable approaches to clarify what building structure44 related materials may be left on site at the time of license termination and what criteria should 

45 be applied to those materials.
G-3

1 For this discussion only, the NRC staff uses the following descriptions of building structures, 

2 systems and components, and equipment: 

3 • “Building structures” include floors, walls, and roofs; components embedded in floors,

4 walls, and roofs (e.g., embedded piping); and items that are attached to and are an

5 integral part of the buildings (e.g., doors and windows).

6 • “Systems and components” include items attached to a building structure that are not an

7 integral part of the building but provide important functions to the building (e.g., utility

8 lines, sinks, lighting fixtures, built-in laboratory hoods and benches, polar cranes (in

9 power reactors), and major process equipment).

10 • “Equipment” includes items not attached to the building structure that are generally

11 readily removable from the building. Examples of equipment include furniture or

12 appliances that are not built into or attached to the structure; stocks of chemicals,

13 reagents, metals, and other supplies; motor vehicles; and any other items that normally

14 would not be conveyed with a building when it is sold.

15 G.2.1.3 Building Structures, and Systems and Components that may be left in place at 

16 License Termination

17 The NRC staff finds the following approaches acceptable to determine what materials may be 

18 left in buildings at license termination. 

19

20

21

22

23

24

25

26

27

28

29

30

31

32

33

34

35

36

37

38

• Materials Left On Site Meet Previously Approved Release Criteria—Building

structures and systems and components may be left in place if residual radioactivity in all

materials is within the licensee’s 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 On Site Meet “Few Millirem per Year”—Building 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 On Site 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 Approaches—Licensees 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.

39

40 For all approaches, the residual radioactivity in building structures, systems and components, 

41 and all other media at the site (e.g., soils or groundwater) must be in compliance with the 
G-4

1 applicable criteria of the LTR (e.g., for unrestricted use, doses must not exceed 0.25 mSv/y 

2 (25 mrem/y) and must be ALARA). 

3 Licensees will perform dose assessments (or use NRC-approved screening dose assessments) 

4 to demonstrate compliance with the dose criteria. Typically, licensees may not need to evaluate 

5 potential offsite future use exposure scenarios, such as removal of soil for fill material or road 

6 base or reuse of concrete as road bed material, because such offsite use exposure scenarios 

7 are usually bounded by onsite use exposure scenarios. However, for some of the dose 

8 assessments needed for the above approaches, when less conservative and more realistic 

9 exposure scenarios are selected, the onsite exposure scenarios may no longer bound potential 

10 offsite use exposure scenarios. Thus, in these cases, the licensee should evaluate offsite use 

11 exposure scenarios. For additional guidance, see Section I.3.3.3.6 of Appendix I of this volume.

12 G.2.1.4 Equipment not covered by the LTR

13 The LTR does not apply to equipment, so equipment should not be left on the site at license 

14 termination. Equipment should be released under the current approaches for releases of solid 

15 materials, as discussed in Section 15.11 of Volume 1 of this NUREG report or could be 

16 disposed of as radioactive waste. 

17 G.2.2 Residual Radioactivity Beneath the Surface

18 The HSA and characterization surveys may indicate that residual radioactivity is present 

19 beneath the surface. In the dose modeling, direct exposure, inhalation, and ingestion pathways 

20 may be important for residual radioactivity on the surface. However, if the residual radioactivity 

21 is located in subsurface soils, additional pathways may become important to dose, and 

22 subsurface soil DCGLs may be derived. In these cases, surveys may be conducted, and results 

23 may be interpreted in a manner consistent with the dose modeling (see Section G.3 for 

24 additional details).

25 The FSS surveys cracks and crevices in the same manner as other building surfaces, except 

26 that these areas should receive judgmental scans when scanning coverage is less than 

27 100 percent.

28 For painted-over residual radioactivity, the licensee should use the HSA and characterization 

29 surveys to determine whether residual radioactivity was fixed in place by being painted over. If 

30 so, it may consider the process for its removal in developing the parameters for the dose 

31 modeling and may interpret the survey results in a manner consistent with the dose modeling.

32 G.2.3 Sewer Systems, Waste Plumbing Systems and Floor Drains

33 The HSA and characterization surveys determine whether there are unusual or unexpected 

34 levels of residual radioactivity in sewer systems and floor drains. Residual radioactivity in sewer 

35 systems and floor drains generally does not contribute to the dose pathways in the building 

36 occupancy exposure scenario or the residential exposure scenario. Thus, the licensee should 

37 calculate the dose from residual radioactivity in sewer pipes using a site-specific exposure 

38 scenario and conduct the FSS in a manner consistent with that scenario. If the sewer water is 

39 sent to an onsite drainage field or cesspool, the licensee should evaluate and survey any 

40 residual radioactivity as subsurface residual radioactivity. If unusual or unexpected results are 

41 found during the characterization survey, the licensee should handle the situation on a case-by42 case basis.
G-5

1 If sewage is sent to an onsite drainage field, any residual radioactivity is subsurface, and the 

2 survey methods discussed in Section G.3.1 are appropriate.

3 G.2.4 Ventilation Ducts

4 The HSA and characterization surveys indicate whether residual radioactivity may be present. 

5 External duct surfaces of ventilation ducts are surveyed as if they are a part of the building 

6 surface. The licensee should survey internal duct surfaces in a manner consistent with the dose 

7 modeling assumptions.

8 G.2.5 Piping and Embedded Piping

9 Embedded piping is piping embedded in a durable material, typically concrete, that cannot be 

10 easily removed without significant effort and tools. The HSA and characterization surveys 

11 indicate whether residual radioactivity is present in piping. The normal room surveys will 

12 adequately account for direct (external gamma) radiation from the pipes when the pipes are in 

13 place and undisturbed. The direct (external gamma) dose from the pipes will be in addition to 

14 the dose from the residual radioactivity on surfaces in the room. It may also be necessary 

15 consider building renovation that would disturb the piping, as described in NUREG/CR-5512, 

16 Volume 1, “Residual Radioactive Contamination from Decommissioning,” issued October 1992. 

17 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. 

18

19 G.3 Surveys for Special Situations on Land 

20 G.3.1 Subsurface Residual Radioactivity Surveys

21 MARSSIM addresses radiological surveys of surface soils only (i.e., subsurface radiological 

22 surveys are not within the scope of MARSSIM). Because the MARSSIM FSS method was 

23 designed specifically for residual radioactivity in surface soils, if significant amounts of residual 

24 radioactivity are located at depth (e.g., significant quantities of residual radioactivity in soils 

25 deeper than approximately 15 centimeters), the presence of subsurface residual radioactivity 

26 should be taken into consideration in designing the FSS. The licensee should first determine 

27 whether it needs surveys of subsurface residual radioactivity. The HSA and other surveys will 

28 play an important role in determining whether there is likely to be residual radioactivity in the 

29 subsurface. Modeling can also be used to supplement survey data to determine the potential 

30 for residual radioactivity to be present in significant quantities in subsurface soils or groundwater

31 due to environmental transport. If the survey data and supplemental modeling indicate that 

32 there is little likelihood of significant subsurface residual radioactivity, then subsurface surveys 

33 are likely unnecessary.

34 If the survey data indicates that there is substantial subsurface residual radioactivity, and the 

35 licensee plans to terminate the license with some subsurface residual radioactivity in place, the 
G-6

1 FSS should consider the subsurface residual radioactivity to demonstrate compliance with the 

2 radiological criteria for license termination. To prepare for the FSS, the characterization survey 

3 determines the depth of the residual radioactivity. In addition to conventional drilling, the 

4 licensee may consider the use of exploratory trenches and pits, where the patterns, locations, 

5 and depths are determined using prior survey results or HSA data. 

6 Performing radiological surveys at sites with significant quantities of subsurface residual 

7 radioactivity is more complex compared to surveys of surface soils given the relative 

8 inaccessibility of the subsurface regions (e.g., subsurface soils cannot be scanned for elevated 

9 areas without the extraction of subsurface materials). Additionally, heterogeneous materials are 

10 often encountered in the subsurface, and the presence of contaminated groundwater also 

11 presents challenges to subsurface radiological surveys (see Appendix F). Because the 

12 MARSSIM methodology relies heavily on scanning to identify elevated areas of concern, 

13 alternative or supplemental methods are needed when residual radioactivity is present in the 

14 subsurface. Modeling may help inform and supplement collection of radiological survey data 

15 and help alleviate the challenge of adequately characterizing the subsurface when scanning is 

16 not a viable option. NUREG/CR-7021 (A Subsurface Decision Model for Supporting

17 Environmental Compliance) presents a framework focused on development of a conceptual site 

18 model referred to as a contamination concern map (CCM). The CCM describes the extent, 

19 location, and significance of residual radioactivity relative to the decision criteria. The CCM can 

20 be developed with the aid of visualization, geographic information system (GIS), and 

21 geostatistical software. As additional data are collected, the CCM transitions from a more 

22 qualitative description to a more quantitative and detailed map. Subsurface concentration 

23 estimates and uncertainty measures serve as surrogates to scanning to facilitate more optimal 

24 sampling designs and decision-making. The approach laid out in NUREG/CR-7026 (Application 

25 of Model Abstraction Techniques to Simulate Transport in Soils) presents one potentially 

26 acceptable method that may be used in conjunction with radiological survey data to 

27 demonstrate compliance. For complex decommissioning cases where subsurface residual 

28 radioactivity and groundwater contamination are present, it is important to work with NRC early 

29 in the process to discuss acceptable approaches for demonstrating compliance with radiological 

30 criteria for license termination.

31

32 As discussed above, GIS and geostatistical software are available to assist with designing, 

33 performing, and evaluating the results of radiological investigations. GIS tools can be used to 

34 help with creation of conceptual models (e.g., provides spatial context and a better 

35 understanding of site features that may control or enhance radionuclide transport in the 

36 environment). Figures created with GIS software can also assist with identifying relatively 

37 homogeneous areas of residual radioactivity for delineation of survey units. Examples of 

38 features that can be captured on a figure using GIS tools include the following:

39

40 • Study area and property boundary

41 • Buildings where residual radioactivity may be present

42 • Roads

43 • Surface water features (streams, ponds, runoff basins, ditches, culverts)

44 • Underground features (underground storage tanks, piping)

45 • Topography, surface geology, and outcrop locations

46 • Hydrostratigraphic surfaces and isopach maps
G-7

1 • Water table and potentiometric surfaces

2 • Sampling locations

3 • Monitoring well locations

4 • Contaminant distributions

5 Geostatistical tools can be used to create figures showing contaminant distributions, predict 

6 radionuclide concentrations in areas where no data exists, and identify areas with a higher 

7 probability of residual radioactivity above levels of concern. This information can be beneficial 

8 in designing the scoping, characterization, and remediation surveys to define the nature and 

9 extent of residual radioactivity (e.g., optimizing the number and locations of samples). Figure 

10 G-1 presents an example use of geostatistical software.

11

12 If subsurface residual radioactivity is present, dose modeling may be conducted for both surface 

13 (if present) and subsurface soils and DCGLs developed for each. In these cases, the 

14 MARSSIM methodology will need to be supplemented or an alternative methodology will need 

15 to be developed to demonstrate compliance with radiological criteria for license termination 

16 (MARSSIM only addresses residual radioactivity at the surface). Because the depth and 

17 thickness of residual radioactivity are correlated to dose, the modeling should reflect the actual 

18 distribution of radioactivity in the survey unit. For example, for certain radionuclides (e.g., those 

19 whose risk is dominated by the plant ingestion pathway), the thickness of residual radioactivity 

20 is strongly correlated to dose. If the modeling assumes a thinner layer of residual radioactivity 

21 than is present, then the risk could be significantly underestimated. If the modeling assumes a 

22 thicker layer of residual radioactivity than is present, then the risk could be significantly 

23 overestimated. Additionally, for some radionuclides (e.g., those whose risk is dominated by the 

24 external dose pathway), the surface concentration may drive the risk as radiation emitted from 

25 residual radioactivity located at greater depth may be attenuated in the soil column and not 

26 contribute to dose. Therefore, if vertical heterogeneity is an issue, then it may be necessary to 

27 take discrete samples to ensure that higher concentration residual radioactivity at the surface is 

28 not diluted in cleaner materials at depth. Dose modeling can be used to determine the 

29 sensitivity of dose to these parameters, and the soil sampling design should ultimately be 

30 consistent with the modeling used to develop the DCGLs. Ideally, sufficient resolution in the 

31 sampling data would be available to evaluate vertical heterogeneity and calculate appropriate 

32 concentrations for comparison against DCGLs derived for specific depths and thicknesses 

33 and/or for the total thickness of residual radioactivity to ensure dose is not underestimated. 

34

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 extrapolate data in areas 

where no data is 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).

35
G-8

1

(a)

(b)

(c)

2 Figure G.1 Use of GIS and Geostatistical Tools (a) to Identify Potential Areas where

3

4

5

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 

6 Characterization (bottom) 

7 When the licensee establishes the appropriate DCGLs and mixing volumes, based on an 

8 acceptable site-specific dose assessment, the FSS takes core samples to the measured depth 

9 of the residual radioactivity. The number of cores to be taken is initially the number (N) required 
G-9

1 for the WRS or Sign test, as appropriate. The adjustment to the grid spacing for an EMC is 

2 more complicated than for surface soils, because scanning is not applicable. The core samples 

3 should be homogenized over a soil thickness that is consistent with assumptions made in the 

4 dose assessment, typically not exceeding 1 meter in depth. It is not acceptable to average 

5 radionuclide concentrations over an arbitrary soil thickness. The appropriate test (WRS or Sign) 

6 is then applied to the sample results. Triangular grids are recommended, because they are 

7 slightly more effective in locating areas of elevated concentrations. Site-specific EMCs may 

8 also need to be developed to demonstrate regulatory compliance and should take into 

9 consideration key radionuclides, pathways, and exposure scenarios important to dose. For

10 subsurface residual radioactivity at depth, the groundwater pathway and total inventory may 

11 drive the risk (i.e., small elevated areas of concentration may not be important to dose). Most 

12 intrusion scenarios assume some minimum degree of mixing of excavated soils; therefore, 

13 mixing arguments can be presented when determining the minimum volume of soil of interest in 

14 developing EMCs. The NRC has not yet developed generic guidance for performing an EMC 

15 for subsurface samples; therefore, licensees should discuss this matter with the NRC staff on a 

16 case-by-case basis. 

17

18 The sampling approach described above may not be necessary if sufficient data to characterize 

19 the subsurface residual radioactivity are available from other sources. For example, for some 

20 burials conducted under prior NRC regulations, the records on the material buried may be 

21 sufficient to demonstrate compliance with the radiological criteria for license termination.

22 G.3.2 Surveys of Excavations and Use of Backfill Soil for Excavated Land Areas

23 G.3.2.1 Surveys of Excavations

24 In cases where a licensee must remediate a site through excavation of subsurface residual 

25 radioactivity above clean-up levels, several options are available to the licensee to demonstrate 

26 compliance with radiological criteria for license termination. Although a backfilled excavation 

27 represents the final configuration of the site, it is a reasonable expectation that the licensee will 

28 perform the FSS on the open excavation prior to backfilling, if the survey can be performed 

29 safely. This is due to the potential cost and difficulty associated with adequately sampling a 

30 backfilled survey unit and the fact that scanning the entire depth of backfill would likely not be 

31 possible for most situations. Sampling and scanning of the open excavation are also beneficial 

32 by helping to ensure that residual radioactivity above levels that would lead to an exceedance of 

33 the dose criteria are removed and appropriately disposed. When a FSS is performed on an 

34 open excavation, it is important to document the locations and depth range below final grade 

35 represented by sampling as well as the general topographical layout of the excavation relative 

36 to final grade to understand the final distribution of residual radioactivity at the site and to 

37 facilitate comparison to release criteria. Additionally, it is important for the licensee to 

38 communicate with the NRC staff to allow the NRC to perform a confirmatory survey that 

39 provides independent evaluation of radiological conditions before the licensee backfills the 

40 excavation. 

41

42 Survey instruments should be used that are appropriate for evaluating the radioactive 

43 contamination of interest and all accessible surfaces of Class 1 survey units should be 

44 evaluated. Specialized equipment such as extended instrument cables/poles, trench boxes, or 

45 other tools and equipment may be needed to perform surface scans and to obtain samples from 

46 locations within the excavation and sidewalls suspected of contamination. Deep excavations 

47 may require significant preparatory work both to perform the excavation and to make an 

48 excavation safe to access, examples of this include pumping to lower the water table, use of 
G-10

1 steel pilings to shore up excavation side walls, and appropriate respiratory protection if toxic 

2 chemicals are present or could displace oxygen. In all cases, the health and safety of workers 

and the public should take priority.1 3

4

5 There are several approaches a licensee can take for sampling an excavation. Although 

6 MARSSIM does not apply to subsurface soils, a MARSSIM-based approach may be extended 

7 to subsurface problems. The survey classification of an excavation should consider whether the 

8 entire excavated area, including the floor and the sidewalls, has the same contamination 

9 potential. Also, if remediation of soil took place to meet the release criteria, then a MARSSIM 

10 Class 1 classification would be appropriate. In this case, if the floor and sidewalls of the 

11 excavation both possess the same potential for contamination, then the same level of survey 

12 effort should be applied to both areas. However, there may be cases where sloped sidewalls 

13 into an excavation are created to provide safe access for remediation activities, and the 

14 contamination potential of the sidewalls could be lower than the bottom of the excavation. In 

15 this example, the new sidewalls may be a Class 2 buffer area, whereas the floor of the 

16 excavation is a Class 1 area. As such, it is important to evaluate excavation practices and 

17 sampling strategies during the DQO process to ensure that classification and survey 

18 methodologies account for actual site conditions. It is also important to consider the potential 

19 for ongoing decommissioning activities to contaminate buffer areas in an excavation. Although 

20 the sidewalls or bottom of the excavation may not be planar surfaces, sampling locations could 

21 be established based on a flat, final graded area (see Figures G-2 (a) and (b)). The depth of 

22 sampling of the excavation should be considered during the DQO process. Sampling of sloped 

23 excavation sidewalls that are excavated to a depth shallower than the bottom of the excavation 

24 pose additional considerations due to their difference in elevation below grade. Different 

25 sample collection strategies such as sampling more horizontally or vertically into the sidewall,

26 may also be considered during the DQO process. In addition, a sampling strategy may need to 

27 take into account the slope of sidewalls, sample density, thickness of strata, and multiple 

28 DCGLs developed to represent different soil intervals. Licensees are encouraged to discuss 

29 with the NRC staff the appropriate level of detail and technical features of their sampling 

30 strategy. 

31

32 In many cases, licensees have used a layered approach in which multiple subsurface layers or 

33 strata are considered individually and then the cumulative risk from the multiple layers or strata 

34 are assessed. If the sidewalls of the excavation are considered the same Class 1 survey unit 

35 consistent with the bottom of the excavation, then the cumulative risk from all of the strata 

36 (below and above the bottom of the excavation) should be considered. Alternative approaches 

37 are available to assess the data. 

38

39 As noted above, in some cases multiple sets of DCGLs may have been approved for different 

40 strata or layers below final grade based on dose modeling. For example, surface soil is 

41 important to the external dose and inhalation pathways, intermediate depth soil corresponding 

42 to the till depth is potentially important to the plant ingestion pathway, and subsurface soil may 

43 be important to the groundwater pathway. Intrusion scenarios are considered to develop 

44 subsurface DCGLs for otherwise inaccessible subsurface soils at depth (see Appendix J). In 

45 cases where different sets of DCGLs are developed for different strata, it is important to ensure 

46 the average contaminant concentration in each designated stratum is lower than the applicable 

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. 
G-11

1 DCGL and that any elevated areas are appropriately investigated and addressed. A sum of 

2 fractions approach can be used to assess the cumulative risk associated with multiple strata. 

3 For example, a sum of fractions approach may entail calculating the sum of fractions for each 

4 sampling location considering the entire soil column, and then calculating the average sum of 

5 fractions for all the locations if the sum of fractions is greater than 1 for any one location, or 

6 evaluating the sum of fractions for multiple strata using the average concentration for each 

7 strata. In all cases, it is important to understand the basis for development of the DCGLs 

8 (thickness, depth, distribution, and area of residual radioactivity) for each stratum to ensure that 

9 risk is not underestimated. It is also important to understand the dose modeling assumptions 

10 with respect to the reuse of soils in the excavation (i.e., use of clean or slightly contaminated 

11 soils) to ensure the risk is not underestimated. In complex cases involving subsurface residual 

12 radioactivity, it is always prudent to calculate the final estimated dose for the compliance 

13 scenario(s) based on the final configuration and measured radionuclide concentrations of 

residual radioactivity at the site through dose modeling.2 14

15

16 As stated above, it is important to adequately assess the radiological characteristics of residual 

17 radioactivity in soils used for backfilling an excavation. For each of the strata for which DCGLs 

18 are developed, radioanalytical results of backfill soil sampling should be analyzed and 

19 appropriately included in the sum of fractions calculations discussed in the preceding 

20 paragraph. In some cases, DCGLs developed for strata below the excavation bottom could be 

21 significantly greater than DCGLs developed for more shallow strata (i.e., strata associated with 

22 backfill soil). Therefore, it is important to ensure that the backfill soil concentrations are

23 compared to the appropriate DCGL to ensure that the risk is not underestimated. In many 

24 cases, the backfill soil is sampled after the soil is excavated, in addition to the surveys 

25 conducted before the soil is emplaced in the excavation. 

26

27 In some cases, it may be appropriate and more straightforward to derive a single set of DCGLs 

28 based on the entire thickness of residual radioactivity, rather than develop multiple sets of 

29 DCGLs for different strata thicknesses. However, sensitivity analyses may need to be

30 conducted to ensure that the risk is not underestimated using this approach. For example, for 

31 radionuclides dominated by the external dose pathway, a similar DCGL may be developed for a 

32 small thickness of residual radioactivity (e.g., 15 cm) compared to a larger thickness of residual 

33 radioactivity, as residual radioactivity at depth may not contribute to the external dose pathway 

34 due to attenuation in the soil column. The concentration of the surface layer would then be 

35 appropriate to compare to the DCGL. If the plant pathway is important, soil concentrations 

36 down to 1 m may be important to dose (depending on the plant types and depth of roots). 

37 When a mix of radionuclides are present, certain radionuclides may be transported to greater 

38 depths than others. DCGLs derived for all of the radionuclides based on the depth of residual 

39 radioactivity for one radionuclide may lead to underestimate of risk for other radionuclides. In 

40 these cases, care should be taken to ensure that residual radioactivity in the subsurface is 

41 adequately characterized to determine the lateral and vertical extent of contamination and 

42 elevated concentrations are appropriately considered (i.e., depth discrete surveys or sampling 

43 may be needed to capture heterogeneity in soil concentrations and ensure that elevated areas 

44 above DCGLs are not diluted by averaging concentrations with clean soils). 

45

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.
G-12

1 Figure G.2 Sampling Strategies for Excavations (A) Use of MARSSIM Survey Design

2

3

4

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 

5 be Derived are Depicted) [Cross-Section Map on bottom]

6 Excavated land areas at sites undergoing decommissioning will often require backfill soils. 

7 Multiple options exist for backfill sources, such as offsite areas, nonimpacted areas on site, or 

8 impacted areas on site that have been appropriately surveyed. 

9

Class 1 Survey Unit—

Survey bottom of the

excavation

Class 1 or 2 Survey UnitSurvey sidewalls (see B)

A

B

Strata 1

Strata 2

Strata 3
G-13

1 G.3.2.2. Backfill from Nonimpacted Onsite and Offsite Areas

2 Licensees have typically proposed to use backfill from non-impacted areas onsite or from offsite 

3 locations. If the licensee is assuming there is no added residual radioactivity in the backfill, an 

4 analysis should be performed to support this assumption (i.e., that the backfill soils do not 

5 contain residual radioactivity). Residual radioactivity, as defined in this NUREG series, includes 

6 radioactivity from all licensed and unlicensed sources used by the licensee, but excludes 

7 background radiation. If there is uncertainty that backfill soils are non-impacted, one potential 

8 method to support this assumption of no added residual radioactivity would be to use a two 

9 sample statistical test such as a Scenario B type analysis to show indistinguishability from 

10 background, as described in Chapter 13 of NUREG–1505. See Section G.6 for more 

11 information about Scenario B. If Scenario B is used, the licensee should discuss appropriate 

12 values for the LBGR, and UBGR (or width of the gray region) with the NRC. Although some 

13 form of radiological survey is expected, other approaches proposed by the licensee may also be 

14 acceptable. 

15

16 G.3.2.3. Backfill from Impacted Onsite Areas 

17 Re-use of soils from radiologically impacted areas as backfill at a site undergoing 

18 decommissioning has been allowed by the NRC on a case-by-case basis. Licensees should 

19 continue to discuss proposed soil re-use plans with the NRC, as there are potentially complex 

20 issues associated with radiological measurement capabilities and site-specific dose 

21 assessments. The following guidance may assist licensees in developing re-use plans, though 

22 site-specific conditions may lead to additional issues for consideration. 

23

24 If licensees plan to re-use soil from impacted areas onsite, characterization and radiological 

25 surveys should be completed to the rigor of a FSS. The MARSSIM survey unit classification 

26 (i.e., Class 1, 2, or 3) should be considered to ensure that the number of measurements and 

27 scanning coverage are adequate. Licensees should also consider the depth at which surface 

28 soils can be measured via surface scanning and adjust excavation methodologies accordingly

29 (e.g., a Class 1 area should receive 100% surface scanning followed by systematic soil 

30 sampling). In this case, if the entire depth of re-use soil cannot be adequately surveyed it may 

31 be necessary to excavate and survey soil via lift depths consistent with surface soil dose 

32 modeling and instrument capabilities. Similarly, licensees should ensure that the sampling and 

33 scanning performed for Class 2 and 3 areas satisfy the survey objectives. Additionally, the 

34 effects of re-using soil onsite should be evaluated in the context of the final site dose 

35 assessment. The licensee should consider the final configuration of reused soil and may take 

36 credit for clean cover materials in estimating dose for residual radioactivity remaining at the site. 

37 The reuse of impacted soils, depending on the methods used, may include intentional mixing of 

38 soils containing residual radioactivity. Licensees should refer to Volume 1, Section 15.13, of 

39 this NUREG for additional guidance on intentional mixing of soils to meet license termination 

40 criteria. It is recommended that the licensee fully consider the implications of re-using impacted 

41 soils during the planning and DQO processes and discuss plans with the NRC as needed. 

42 Additional details regarding radiological surveys for reused soils are discussed in the following 

43 paragraphs.

44

45 Excavated soils may be segregated at a site, based on the level of residual radioactivity for 

46 potential reuse (e.g., impacted soil may be stockpiled for possible reuse as backfill or grading 

47 materials on a site). If the residual radioactivity includes radionuclides that can be measured in 

48 a scan survey (i.e., contaminants are not hard-to-detect radionuclides that cannot be detected in 
G-14

a scan survey3 1 ), then the licensee could use scanning as a first method to evaluate the soil. If

2 scanning indicates that the soil contains residual radioactivity above background when it is 

3 expected to be free of residual radioactivity, or above release criteria when DCGLs are 

4 developed for reused soils, then the licensee should determine whether it is acceptable to reuse 

the soil at the site. In some cases, the soils may not be suitable for stockpiling and reuse4 5 . The 

6 stockpiling of soil should be limited so that only scanned strata are excavated before the newly 

7 exposed soil is scanned. The thickness of the excavated lifts should be consistent with the 

8 scanning instrument capability and scanning coverage consistent with the survey unit

9 classification. Alternatively, soils can be excavated, transported to a suitable laydown area, and 

10 then scanned in the laydown area similar to a FSS prior to stockpiling. In these cases, the 

11 thickness of soil in the laydown area for scanning should be consistent with the scanning 

12 instrument capability and survey unit classification (e.g., in a Class 1 area, the areal coverage 

13 should be 100 percent and the survey instrument should be able to detect residual radioactivity 

14 at the bottom of a soil layer in the laydown area). Stockpiled soil and materials from impacted 

15 areas that are heterogeneous may result in local areas of increased radioactivity levels. In 

16 these situations, it may be useful to rescan the soil and materials after emplacement to 

17 determine whether additional evaluations or spot remediation is warranted, especially if the soils

18 are used for final grade material.

19

20 In some cases, licensees have used conveyor belts to scan excavated materials. Soil

21 characteristics such as soil type, density, and moisture content, and measurement sensitivity 

22 are factors to consider in the DQO process when using these systems. 

23

24 In summary, when stockpiled soils are planned for use as backfill or grading material and 

25 remain onsite at the time of license termination, it is important to develop a survey plan that 

26 takes into account the characteristics of the soil, measurement methods to be used before and 

27 after emplacement, and applicable radiological criteria, which may be based on site-specific 

28 dose modeling. 

29

30 Although DCGLs may have been developed to guide site remediation activities, it is also 

31 prudent to perform dose modeling using the final configuration and concentrations of residual 

32 radioactivity at the time of license termination. 

33

34 G.3.3 Rubble, Debris, and Rocks

35 Rubble, debris, and rocks can include naturally occurring rocks (either in place or in piles), 

36 pieces of concrete or rubble from buildings that have been razed, sheet metal disposed of as 

37 trash, asphalt, fly ash, and similar material. The HSA and characterization surveys determine 

38 the volumetric extent and residual radioactivity concentration. If the materials are contaminated, 

39 they would be disposed of as radioactive waste. If the radioactivity is not substantially elevated, 

40 the licensee may evaluate the rubble, debris, and rocks as part of a larger survey unit. When 

41 these materials will be evaluated as part of a larger survey unit and when they are found on a 

42 relatively small fraction of the area of a survey unit, the volumetric soil DCGL should be used 

43 uniformly throughout the survey unit. However, the licensee should justify the reasonableness 

44 of modeling rocks and rubble as soil.

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.
G-15

1 G.3.4 Paved Parking Lots, Roads, and Other Paved Areas

2 The HSA and characterization surveys determine whether the residual radioactivity is on or near 

3 the surface of the paving and whether there are significant concentrations of residual 

4 radioactivity beneath the paving. If the residual radioactivity is primarily on top of the paving, 

5 then the licensee should take the measurements as if the area were normal soil. Depending on 

6 how large the paved area is, the licensee may include it as part of a larger survey unit, or it may 

7 be its own survey unit. If the residual radioactivity is primarily beneath the paving, the licensee 

8 should survey it as subsurface residual radioactivity, as discussed above.

9 G.3.5 Recontamination

10 Finally, measures should be employed to avoid recontamination of previously remediated areas 

11 and the need for additional surveys or analyses to demonstrate compliance with release criteria. 

12 Recontamination of cleaned survey units or areas on site can occur particularly at 

13 decommissioning sites where (i) decommissioning is conducted in phases, (ii) partial site 

14 releases are conducted, and (iii) remediation occurs over long time-frames. Contamination of 

15 cleaned areas from residual radioactivity in unremediated areas can occur due to environmental 

16 and human-induced transport such as through surface water runoff, air transport; and soil 

17 moving or demolition activities. Should NRC staff suspect recontamination of previously 

18 surveyed areas, additional surveys, dose modeling, or remediation may be necessary to 

19 demonstrate compliance with release criteria.

20

21 G.4 Surveys Associated with Multiple Radionuclides and Media

22 In cases where residual radioactivity is present in multiple media and multiple radionuclides are 

23 present, it may be necessary to adjust the dose modeling approach used to derive DCGLs, as 

24 well as the survey design approach used to demonstrate compliance with radiological criteria for 

25 license termination to account for the cumulative dose from multiple radionuclides and media. 

26 For example, consider a site with residual radioactivity present on the surfaces of a building, 

27 surface and subsurface soils, and nearby stream bed sediments. Multiple key radionuclides 

28 contributing significantly to the dose are present in each of the media and different key 

29 radionuclides are important for each of the media. Given the previous industrial use of the site 

30 and presence of residential areas in close proximity to the site, reasonably foreseeable future 

31 land use scenarios include industrial and residential use of the site within 100 years of license 

32 termination. Exposure to stream bed sediments may also occur due to recreational use of the 

33 stream such as swimming, boating, and fishing. Therefore, a recreational scenario may also be 

34 considered for the streambed sediments. DCGLs are typically derived for each of the 

35 contaminated media based on the release criteria (e.g., 0.25 mSv/y [25 mrem/y] for unrestricted 

36 release). However, given limitations in the dose modeling codes used to derive the DCGLs, 

37 receptors are typically assumed to be exposed to just one of each of the media. Likewise, given 

38 use of different DCGLs for each of the different media, the survey design does not automatically 

39 consider the contributions of multiple media in demonstrating compliance with release criteria.

40

41 With respect to consideration of multiple radionuclides, the sum of fractions approach can be 

42 used to account for multiple radionuclides in the same survey unit. Chapter 11 in NUREG-1505 

43 (NRC, 1998) provides guidance and an example on how to use the sum of fractions approach to 

44 design the survey to determine the appropriate number of samples to demonstrate compliance 

45 with release criteria while ensuring decision criteria are met (e.g., acceptable Type I [false 

46 positive] and Type II [false negative] errors). However, this example is only applicable to a 

47 single survey unit type (e.g., soil, building, streambed sediments).
G-16

1

2 While surface and subsurface soils may have separate DCGLs, similar to the approach used for 

3 multiple radionuclides, surface and subsurface soils can be considered together when 

4 demonstrating compliance because they can be assumed to be located in the same survey 

5 unit(s). However, soil survey units are typically separate from building survey units and 

6 streambed sediment survey units. Therefore, if the demonstration of compliance is made for 

7 each media survey unit type based on clean-up levels derived at the compliance limit, if a 

8 member of the public could be exposed to multiple media and survey unit types in the same 

9 year, the annual release limit could be exceeded. 

10

11 Licensees can take a conservative approach to addressing cumulative dose associated with 

12 each of the media by re-calculating the DCGLs for each of the media assuming a portion of the 

13 compliance limit (e.g., 45 percent of the 0.25 mSv/y [25 mrem/y] unrestricted release limit for (i) 

14 the soils, and (ii) building; and 10 percent for (iii) streambed sediments). It is important to note 

15 that re-calculation of the DCGL is necessary prior to designing the survey to ensure that the 

16 survey design is based on the correct width of the gray region (e.g., additional samples may be 

17 needed if the width of the gray region is smaller due to the lower DCGL or UBGR used in the 

18 survey design). While assuming both a residential and industrial scenario occur at the same 

19 site may be overly conservative with occupancy factors approaching 100 percent or greater, this 

20 would be an acceptable approach to consideration of cumulative dose associated with all 

21 sources/media and radionuclides. 

22

23 Similar to methods used to consider elevated areas or “hot spots” (see Appendix I), the licensee 

24 may also develop DCGLs based on more reasonable exposure assumptions if a receptor is 

25 assumed to be exposed to multiple sources or media. For example, if the dose contributions 

26 from both the building and the soils are considered in demonstrating compliance, occupancy 

27 factors may be adjusted lower to account for the time spent in the building and time spent 

28 outside in an industrial scenario when developing DCGLs for the building and soils. Codes such 

29 as RESRAD-BUILD and RESRAD-ONSITE allow for occupancy factor parameter values to be 

30 adjusted such that the total time fraction is not greater than 1 (e.g., the residential farmer 

31 scenario assumes time fractions of 0.12 (outdoors) and 0.66 (indoors), and building occupancy 

32 scenario assumes a time fraction of 0.25 for a total time fraction greater than 1.0) . Other 

33 acceptable approaches for adding realism to the dose calculations may be available and should 

34 be discussed with the NRC staff early in the process.

35

36 G.5 Consideration of Elevated Areas in Survey Designs

37 Using the FSS approach, sites are typically broken up into a number of survey units without 

38 regard to the presence of elevated areas. In cases where multiple elevated areas exist in 

39 relatively close proximity to one another, the licensee should consider restructuring the survey 

40 units so that the elevated areas are evaluated in as few survey units as practical and which 

41 meet the guidance for survey unit size and have a shape similar to typical local property lots for 

housing or industry.5 42 This may be unnecessary if physical barriers exist such that it would be 

43 unlikely that future land use would result in a single occupant having potential exposure from all 

44 elevated areas. Alternatively, if the elevated areas do not contribute significantly to the potential 

45 exposure, it may be possible to simply show that the contributions of the elevated areas do not 

46 exceed the criteria for any of the adjoining survey units when considered as an additional 

47 source of exposure. 

5 Novel dose modeling approaches to addressing potential gross over-estimations in dose when considering 

multiple elevated areas is found in Appendix I.
G-17

1

2 With regard to delineation of elevated areas, sampling data and scanning data are typically 

3 used to define the boundaries of the elevated area. When defining the boundaries of the 

4 elevated area, the area should not be intentionally established to incorporate multiple sample 

5 results that are significantly less than the DCGLs and which would, therefore, intentionally lower 

6 the average activity level of the elevated area. In cases where the survey unit boundaries split 

7 the elevated area, the entirety of the elevated area should be considered as being in the 

8 confines of both survey units or else the survey units should be reconfigured so that it exists in 

9 its entirety in just one survey unit. Other approaches may be proposed by the licensee that may 

10 be acceptable and should be discussed with the NRC staff early in the process.

11

12 When elevated areas are encountered, good housekeeping practices consistent with ALARA 

13 criteria should be considered. For example, if it is practical or cost-beneficial to remove a small 

14 area of elevated activity (e.g., a couple buckets of contaminated soil or caulk in a floor seam), 

15 removal activities would likely be considered consistent with ALARA criteria. Minor removal

16 activities would also help alleviate complications associated with consideration of elevated area 

17 contributions to dose and interpretation of anomalous survey results.

18

19 G.6 Additional Information and Example on Conducting Scenario B

20 As noted in Section 2.4, the default assumption for FSSs is that the survey unit is considered 

21 contaminated above the limit (i.e., the null hypothesis is that the concentrations of residual 

22 radioactivity exceed the DCGLs). This assumption and null hypothesis are considered Scenario 

23 A. Typically, statistical tests are used to demonstrate that the average or median of the

24 measurements in the survey unit are not above the DCGL, with the burden of proof on the

25 Iicensee to show that the survey unit meets the release criteria. In special cases (i.e.,

26 exceptions) it may be appropriate to assume an alternative scenario in which the survey unit is

27 considered clean (i.e., the null hypothesis is that the concentrations of residual radioactivity

28 meet the release criterion). This alternative, Scenario B, should typically only be used when

29 residual radioactivity in the survey unit is within the range of measurement and/or background

30 variability, making it difficult to distinguish between the residual radioactivity and background.

31 For example, sites with highly variable background and comparatively small DCGLs may desire

32 to demonstrate that measurements in the survey unit are indistinguishable from measurements

33 in reference areas. Other Scenario B applications may include the use of gross alpha or gross

34 beta measurements, or nuclide-specific measurements when the nuclide is in background,

35 similar to Scenario A. Licensees considering the use of Scenario B for compliance with 10 CFR

36 20 Subpart E are strongly encouraged to contact with the NRC staff early in the planning

37 process. In all cases, the licensee should discuss its plans with the NRC to determine the

38 acceptability of using Scenario B, as well as determining appropriate values for the test

39 parameters.

40 As a Scenario B example, this section describes five major steps to demonstrate 

41 indistinguishability from background when the Wilcoxon Rank Sum and Quantile tests are used: 

42 (1) assess background variability using the Kruskal-Wallis test; (2) determine a concentration of

43 radioactivity that is indistinguishable from background; (3) perform the Wilcoxon Rank Sum test;

44 (4) perform the Quantile test if the survey unit “passes” the Wilcoxon Rank Sum test; and (5)

45 perform the elevated measurement comparison test. Information on these methods is

46 contained in Chapters 6, 7 and 13 of NUREG-1505, and summarized below.

47
G-18

1 Prior to performing statistical tests, measurements are made in several different background 

2 reference areas. The number of samples needed depends on the probability of Type I error (α) 

3 and Type II error (β) that are considered acceptable. The null hypothesis is that there is no 

4 significant variability between different background reference areas, so a Type I error would 

5 incorrectly conclude there is a significant difference between the background reference areas 

6 when there is no significant difference. Using the DQO process, a value of α is selected for an

7 acceptable frequency of a Type I error rate. Table 13.5 of NUREG-1505 provides results of 

8 calculations of the power of the F-test (parametric complement to the nonparametric Kruskal9 Wallis test) as a function of the number of reference areas and the number of measurements in 

10 each reference area for α values of 0.05, 0.10 and 0.20. Based on the results of these 

11 calculations, NUREG-1505, Chapter 13, indicates that 4 reference areas with between 10 and 

12 20 measurements in each reference area should generally be adequate, and that an α value of 

13 0.10 is a reasonable default value. Other α values, number of reference areas and number of 

14 measurements in each reference area could be found to be reasonable during the DQO 

15 process. For planning purposes, it may be convenient to select multiples of five measurements 

16 for performing the subsequent Quantile test because Table A.7a of NUREG-1505 contains 

17 values as multiples of five survey measurements. 

18

19 In this Scenario B example, measurement data from 10 measurements in 4 reference areas are 

20 provided in Table G-1. The Kruskal-Wallis test is performed on this data to determine if 

21 significant variability exists in the background reference areas. The measurements for the 

22 reference areas are pooled and ranked, and then the sum of the ranks for the individual 

23 reference areas (Ri) and the mean measurement are determined. A test statistic (K) is obtained 

24 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; ni 25 is the number of 

measurements in a given reference area; and Ri 26 is the sum of the ranks of the measurements in 

27 a given reference area: 

K =

12

N(N + 1)

(∑

Ri

2

ni

k

i=1

28 ) − 3(N + 1)

29 The value of K is then compared with the critical value (Kc) found in Table 13.1 of NUREG-1505. 

30 If K is greater than Kc, the null hypothesis is rejected, and it is concluded that there is significant 

31 variability between the reference areas. For the example data in Table G-1, the K value of 14.0 

32 is compared to Kc for three reference areas (four reference areas minus one reference area). 

33 The value of Kc is lower than K, which ranges from 11.3 (for an α value of 0.01) to 4.6 (for an α

34 value of 0.2). Therefore, the null hypothesis can be rejected with high confidence. It is 

35 concluded that there is significant variability between the reference areas, which helps justify 

36 use of Scenario B for determining indistinguishability from background.

37 As mentioned above, an alternative to the Kruskal-Wallis test is the F-test to determine if 

38 variability between the means of potential reference areas is statistically significant. Chapter 13 

39 of NUREG-1505 contains information on the use of the F-test for this purpose. Although the 

40 Kruskal-Wallis test (or F-test) is used to determine if it is appropriate to consider reference area 

41 variability in applying Scenario B, it may be acceptable to not conduct either test if it is decided 

42 that there are significant differences among the potential reference areas for a survey unit. 

43 Thus, if it is considered appropriate to give background variability the benefit of the doubt, the 

44 Kruskal-Wallis and F-test test need not be conducted. Not conducting these tests is essentially 

45 the same as setting α equal to 1.0. 
G-19

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

6

19 1.02 0.32 8.01 2.96

5 2.08 1.97 1.09 1.50 24 21 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

533

16

28

0.09 1.93 7.84 2.02

8 1.98 0.05 2.77 2.47 22 4 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

2 The next step is to determine the LBGR. The LBGR is the concentration level above

3 background that may be considered distinguishable from background and can be established 

4 using a measure of the variability among the background reference areas. The mean square 

between reference areas, sb

2

, and mean square within reference areas, sw

2 5 , are used to 

calculate the component of the variance (ω2 6 ). These test parameters can be calculated using 

7 methods described in Chapter 13 of NUREG-1505 or can be found in analysis of variance 

8 [ANOVA] output from commonly used statistical software (see Table G-2). The difference in 

9 concentration that is distinguishable from background may be expressed in terms of an 

10 appropriate multiple of ω. The equation below is based on equation 13-13 in NUREG-1505, 

11 where M is a multiplier selected using the DQO process and no is equal to the number of 

12 measurements per reference area when the number of measurements in each reference area is 

13 the same. 

LBGR = M√ω̂

2 = M√

(Sb

2 − Sw

2

)

n0

14

15 Equation 13-13 in NUREG-1505 can be used when the number of measurements in the 

16 reference areas are not the same. Using the DQO process, a value for M needs to be 

17 determined. Based on information contained in section 13.4 of NUREG-1505, a value of 3 ώ is 

a reasonable default. Using the ANOVA output in Table G-2, the LBGR = 3 x √ω̂

2 18 = 2.22. Note 

19 that the difference in means between reference areas 2 and 4 in Table G-1 is 1.82, which is 

20 consistent with the LBGR calculated based on 3 ώ. 

21

1 Table G-1 Calculation of ω2 for the Example Data
G-20

1 Table G-2 Analysis of Variance for Example Data

Source of 

Variation

Sum of 

Squares

Degrees of 

Freedom

Mean Square F Statistic

Between 

Groups 19.56 3 6.52 6.69

Within 

Groups 35.08 36 0.97

Total 54.65 39

2

3 Section 9.5 of NUREG-1505 contains information on calculating the sample size for the 

4 Wilcoxon Rank Sum test under Scenario B, and is essentially the same method used in 

5 Scenario A. Also, the Wilcoxon Rank Sum test is performed similarly as it is used in Scenario 

6 A. The data for an example Wilcoxon Rank Sum test using Scenario B are shown in column A

7 of Table G-3. In column B, the label “R” is inserted to denote a reference area measurement,

8 and the label “S” to denote a survey unit measurement.

9

10 In Scenario B, the survey unit measurements are adjusted by subtracting the LBGR (e.g., 3 ω̂)

11 from each survey unit measurement, whereas in Scenario A the reference area measurements

12 are adjusted. Column C of Table G-3 contains the adjusted data obtained by subtracting the 

13 LBGR (142) from the survey unit measurements. The ranks of the adjusted data in Column C 

14 are listed in Column D. Next, the adjusted survey unit measurements are ranked, which range 

15 from 1 to 24, since there is a total of 12 + 12 = 24 measurements. The sum of all the ranks is 

16 N(N + 1)/2 = (24)(25)/2 = 300. Column E contains only the ranks belonging to the adjusted 

17 survey unit measurements. The sum of the ranks of the adjusted survey unit data, Ws, is 194.5. 

18

19 When using the DQO process to determine an acceptable probability of a Type I error, it should 

20 be noted that the total probability of a Type I error in assessing the survey unit data is the sum 

21 of the probabilities of Type I errors for both the Wilcoxon Rank Sum and Quantile tests.

22 Therefore, the acceptable probabilities of Type I errors in the Wilcoxon Rank Sum and Quantile 

23 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 αWRS 24 is 0.025. For comparing the statistic Ws with the critical 

25 value, Table A.4 of NUREG-1505 is used. However, for Scenario B, the meaning of m and n 

26 are the reverse of those in Scenario A. For Scenario B, m is the number of survey unit 

27 measurements and n is the number of reference area measurements in this table. From Table 

28 A.4 of NUREG-1505, for values of α for the Wilcoxon Rank Sum test = 0.025 and n = m = 12,

29 the critical value is 184. Because the sum of the adjusted survey unit ranks, 194.5, is greater

30 than the critical value, 184, the null hypothesis that the survey unit concentrations do not exceed

31 the LBGR is rejected (i.e., the site is determined to be dirty). In this Scenario B example, the

32 true survey unit residual radioactivity is judged to be in excess of 142 above background.

33 The Quantile test is only used in Scenario B and if the Wilcoxon Rank Sum test fails to reject the 

34 null hypothesis. Whereas the Wilcoxon Rank Sum test is best at detecting excess radioactivity 

35 that is uniformly distributed, the Quantile test is intended to recognize excess radioactivity in 

36 only a fraction of the survey unit (i.e., higher concentrations of radioactivity in a fraction of the 

37 survey unit). Although the null hypothesis was rejected and the Quantile test need not be 

38 performed, for this Scenario B example the Quantile test is presented for illustrative purposes 
G-21

1 using the data in Table G-3. The data are adjusted and ranked in the same manner as in the 

2 Wilcoxon Rank Sum test, then the ranks are sorted, and the area associated with the rank is 

3 identified. The null hypothesis of the Quantile test is that there is no residual radioactivity above 

4 the LBGR in any part of the survey unit. 

5 Table A.7 of NUREG-1505 provides for the Quantile test the critical value, k, of the largest r 

6 measurements for different values of n, the number of measurements from the survey unit, and 

7 m, the number of measurements from the reference area. Different sub-tables are provided in 

8 Table A.7 of NUREG-1505 for different α values. The same rankings in Column D of Table G-3

9 for the Wilcoxon Rank Sum test can be used for the Quantile test. If k or more of the r largest 

10 measurements in the combined ranked data set are from the survey unit, the null hypothesis is 

11 rejected. 

12

13 Columns F and G of Table G-3 show the sorted ranks of the adjusted data, and the location 

14 associated with each rank (i.e., “R” for reference area and “S” for survey unit). From Table A.7b 

15 of NUREG-1505, the closest entry to n = m = 12 is for n = m = 10, with corresponding values of r 

16 = 7, k = 6 and α = 0.029. Thus, the null hypothesis is rejected if six of the seven largest adjusted 

17 measurements are from the survey unit. From Table G-3, we find that only five of the seven 

18 largest adjusted measurements are from the survey unit, so the null hypothesis is not rejected 

19 based on the Quantile test. The values of n and m that were used are close to, but not equal to, 

20 the actual values, so the α value will be different from that listed in the table. It is prudent to 

21 check a few other entries in Table A.7b of NUREG-1505 that are near the actual sample size. 

22 Additionally, Chapter 7 in NUREG-1505 provides equations to calculate exact and approximate 

23 values of α for the Quantile test as a function of n, m, k, and r. 

24 It is recommended that an elevated measurement comparison is conducted, regardless of the 

25 outcome of the WRS and Quantile test. This consists of determining if any measurement in the 

26 remediated survey unit exceed a specified investigation level. If so, then additional investigation 

27 of the data is required to determine if there are elevated measurements that were not identified 

28 by the statistical tests.

29

30
G-22

1 Table G-3 Scenario B WRS and Quantile Tests for Class 2 Interior Drywall Survey Unit

2

3 (Measurements from the reference area and the survey unit are denoted by R and S, 

4 respectively)

5

6

7

 A B C D E F G

1

Data Area Adjusted 

Data

Ranks Survey 

Unit 

Ranks

Sorted 

Ranks

Location 

Associated 

With Sorted 

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
G-23

1 G.7 Integration of Dose Modeling and Radiological Surveys

2 Pathway dose or risk modeling is oftentimes used to determine clean-up levels or DCGLs used 

3 as decision criteria in statistical tests discussed in Chapter 8 of MARSSIM and Appendix A. 

4 Because DCGLs are an integral part of the survey design, model integration with the survey 

5 design is an important topic discussed in various sections of this guidance. 

6 Assuming the pathway dose or risk modeling is representative of actual survey unit conditions, 

7 the survey design should be compatible with modeling used to derive clean-up levels or DCGLs. 

8 Because the distribution, thickness, depth, and area of residual radioactivity are directly related 

9 to dose (or risk) in many commonly used decommissioning modeling codes, the FSS should be 

10 designed consistent with the modeling assumptions related to these key parameters to the 

11 extent practical. For example, if vertical heterogeneity is an issue and DCGLs are sensitive to 

12 the distribution of residual radioactivity within the soil column, DCGLs could be developed for 

13 different soil intervals (e.g., 0 to 5 cm and 5 to 15 cm). Depth discrete sampling may then be 

14 necessary for comparison against the DCGLs developed for each soil interval. For certain 

15 radionuclides and pathways (e.g., Cs-137 and external dose pathway), depth discrete sampling 

16 may also be important to ensure that higher concentration residual radioactivity located near the 

17 surface is not diluted with clean or cleaner radioactivity located deeper in the soil column, which 

could lead to an underestimate of risk6 18 . Likewise, the thickness of contamination can also be 

19 important to risk. If residual radioactivity is located deeper in the soil column then assumed in 

20 the dose modeling, then the risk could be underestimated. Subsurface residual radioactivity 

21 should be adequately characterized and if present, appropriate methods should be developed to 

22 evaluate whether the site is clean (the MARSSIM methodology summarized in Appendix A was 

23 developed for surficial soils and building surfaces only; other methods may be necessary to 

24 make decisions regarding release for sites with subsurface or volumetric residual radioactivity).

25 Additionally, the dose and risk pathway modeling typically assume that contamination is 

26 relatively homogeneous across the model domain and most models only accept a single, 

27 average concentration for each radionuclide as input. If, in contrast, the survey unit is 

28 heterogeneous (e.g., spotty or elevated areas of radioactivity), the assumptions in the modeling 

29 may be violated, and the effectiveness of the statistical tests may be reduced. In these cases, 

30 the survey design team may need to consider other methods to mitigate or assess the impact of 

31 heterogeneous distributions such as more careful delineation of survey units or added emphasis 

32 and reliance on EMC tests. For Scenario B, the quantile test used in conjunction with the WRS 

33 test is also useful in evaluating whether a portion of the survey unit has unacceptable levels of 

34 contamination. 

35 Finally, it is important to note that dose or risk modeling typically uses mean concentrations of 

36 radionuclides from survey unit measurements, while the Sign and WRS tests are tests on the 

37 median. While the median concentration is a good approximation for the mean concentration if 

38 the radionuclide distributions are symmetric (or not highly skewed), in some cases the mean 

39 concentration could exceed the median concentration. The average of the survey unit 

40 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).
G-24

1 should always be compared to the DCGL before a site is released (Scenario A) to help ensure 

2 that a site with residual radioactivity above the DCGL is not released.

3 G.8 References 

4 NRC. Inspection and Enforcement Circular No. 81-07, “Control of Radioactively Contaminated 

5 Material.” U.S. NRC: Washington, DC. May 14, 1981.

6 — — — — —. Fuel Cycle Policy and Directive, FC 83-23, “Termination of Byproduct, Source 

7 and Special Nuclear Material Licenses.” U.S. NRC: Washington, DC. 1983.

8 — — — — —. Information Notice No. 85-92, “Surveys of Wastes Before Disposal from Nuclear 

9 Reactor Facilities.” U.S. NRC: Washington, DC. December 2, 1985.

10 — — — — —. Information Notice No. 88-22, “Disposal of Sludge from Onsite Sewage 

11 Treatment Facilities at Nuclear Power Stations.” U.S. NRC: Washington, DC. May 12, 1988.

12 — — — — —. “Generic Environmental Impact Statement in Support of Rulemaking on 

13 Radiological Criteria for License Termination of NRC-Licensed Nuclear Facilities,”

14 NUREG-1496. U.S. NRC: Washington, DC. November 1996. 

15 — — — — —. NUREG 1505, Revision 1. “A Nonparametric Statistical Methodology for the 

16 Design and Analysis of Final Status Decommissioning Surveys—Interim Draft Report for 

17 Comment and Use.” U.S. NRC: Washington, DC. June 1998.

18 — — — — —. NUREG-1575, Rev. 1, “Multi-Agency Radiation Survey and Site Investigation 

19 Manual (MARSSIM).” EPA 402-R-97-016, Rev. 1; DOE/EH–0624, Rev. 1; U.S. Department of 

20 Defense; U.S. Department of Energy; U.S. Environmental Protection Agency; and the U.S. 

21 NRC: Washington, DC. August 2000.

22 — — — — —. NUREG/CR-5512, Vol. 1, “Residual Radioactive Contamination from 

23 Decommissioning: Technical Basis for Translating Contamination Levels to Annual Total 

24 Effective Dose Equivalent.” U.S. NRC: Washington, DC. October 1992.

25 — — — — —. NUREG/CR-7021, “A Subsurface Decision Model for Supporting

26 Environmental Compliance,” NRC: Washington, D.C. January 2012. 

27

28 — — — — —. NUREG/CR-7026, “Application of Model Abstraction Techniques to Simulate 

29 Transport in Soils,” NRC: Washington, D.C. March 2011. 

30

31 — — — — —. NUREG-1496, “Generic Environmental Impact Statement in Support of 

32 Rulemaking on Radiological Criteria for License Termination of NRC–Licensed Nuclear 

33 Facilities.” Division of Regulatory Applications, Office of Nuclear Regulatory Research, U.S.

34 NRC: Washington, DC. July 1997.

35 — — — — —. Draft Regulatory Guide DG-4006, “Demonstrating Compliance with the 

36 Radiological Criteria for License Termination.” Office of Nuclear Regulatory Research, U.S. 

37 NRC: Washington, DC. August 1998.
G-25

1 — — — — —. Regulatory Issue Summary 2002-02, “Lessons Learned Related to Recently 

2 Submitted Decommissioning Plans and License Termination Plans.” U.S. NRC: 

3 Washington, DC. January 2002. 

4 — — — — —. Regulatory Issue Summary 2004-08, “Results of the License Termination Rule 

5 Analysis.” U.S. NRC: Washington, DC. May 2004.

6 — — — — —. SECY-03-0069, “Results of the License Termination Rule Analysis.” Policy 

7 Issue Memorandum from W.D. Travers to the Commissioners. U.S. NRC: Washington, DC. 

8 May 2, 2003. 

CRITERIA FOR CONDUCTING SCREENING DOSE MODELING 

EVALUATIONS

1 H.1 Introduction

2 This appendix consists of the technical guidance for the use of the screening criteria, applicable 

3 to Decommissioning Groups 1–3. Section I.8 of Appendix I contains details of the references 

4 cited below.

5 This section pertains to NRC staff’s review of a licensee’s demonstration of compliance with the 

6 dose criteria in 10 CFR Part 20, “Standards for Protection against Radiation,” Subpart E, 

7 “Radiological Criteria for License Termination,” using a screening approach dose analysis. The 

8 NRC staff should review the screening analysis using one or more of the currently available 

9 screening tools: 

10 • a lookup table for common beta- and gamma-emitting radionuclides for building surface

11 residual radioactivity (Volume 63 of the Federal Register (FR), page 64,132

12 (63 FR 64132); November 18, 1998)

13 • a lookup table for common radionuclides for soil surface residual radioactivity

14 (64 FR 68395; December 7, 1999)

15 • screening levels derived using DandD Version 2.1, or the most current version, for the

16 specific radionuclide(s) and using the DandD code’s default parameters

17 Other tools for performing a screening analysis might become available in the future, or the 

18 NRC staff may modify current lookup tables or develop additional lookup tables (e.g., the NRC 

19 staff may develop lookup tables for the common alpha-emitters for building surfaces using 

modified parameter values in current versions of DandD1 20 ). In addition, the NRC staff may 

21 consider the use of other screening tools (e.g., other lookup tables or other conservative 

22 codes/models) after evaluating and comparing these screening tools with the current screening 

23 codes.

24 A licensee usually conducts a screening analysis for simple sites with building surface 

25 (i.e., nonvolumetric) and/or with surficial soil (approximately 15 centimeters (cm) (6 inches (in.)) 

26 residual radioactivity. The analysis usually employs simple and conservative models/codes and 

27 parameters, under generic exposure scenarios and default site conditions, to define the 

28 screening derived concentration guideline levels (DCGLs) equivalent to the dose criteria. 

29 Because of the conservative nature of the screening analysis approach, the screening DCGLs 

30 are typically more restrictive than the site-specific DCGLs. A screening analysis may save 

31 licensees time and effort by reducing the amount of site characterization, modeling analysis, 

32 and reviews needed, versus those needed when using a site-specific analysis approach.

33 To review a screening analysis, the NRC staff first needs to make a generic assessment and 

34 evaluation of a licensee’s justification that the site is qualified for screening. In addition, the

35 NRC staff should be familiar with the tools (e.g., models, codes, and calculations) and 

36 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 NRC’s License Termination Rule,” issued June 2002, documents the NRC 

staff’s 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.

H-1
1 issues that the NRC staff may encounter in the generic screening analysis reviews and includes 

2 recommendations of approaches for addressing and resolving these issues. 

3 H.1.1 Issues in Performing Screening Analysis

4 The NRC staff may encounter issues with the screening analysis, including (1) the definition of 

5 screening and the transition from a screening to a site-specific analysis, (2) the qualification of 

6 the site for screening, in terms of site physical conditions and compatibility with the modeling 

7 code’s assumptions and default parameters, and (3) the acceptable screening tools (e.g., code, 

8 lookup tables), approaches, and parameters that the staff can use to translate the dose into 

9 equivalent screening concentration levels. Each one of these issues is the subject of discussion 

10 in the following subsections:

11 H.1.2 Screening Definition and Approaches for the Transition from Screening to Site12 Specific Analysis 

13 The NRC staff may encounter some inconsistencies about the definition of the term “screening” 

14 in dose analysis, which may cause confusion about the transition from a screening to a site15 specific analysis. These inconsistencies become more apparent when dividing screening 

16 approaches into multiple levels (NCRP Report No. 123, “Screening Models for Releases of 

17 Radionuclides to Atmosphere, Surface Water, and Ground,” dated January 22, 1996; NCRP 

18 Report No. 129, “Recommended Screening Limits for Contaminated Surface Soil and Review of 

19 Factors Relevant to Site-Specific Studies,” dated January 29, 1999). In some cases, screening 

20 and site-specific terms are mixed, and the term “site-specific screening” is used 

21 (NUREG/CR-5512, Volume 1, “Residual Radioactive Contamination from Decommissioning: 

22 Technical Basis for Translating Contamination Levels to Annual Total Effective Dose 

23 Equivalent,” issued October 1992). In certain cases, screening is categorized by the type of 

24 models used (e.g., simple and conservative models versus more advanced and complex 

25 models) and the extent of data and information needed to support the dose analysis.

26 Within the context of NUREG-1757, the NRC staff should consider the definition of screening as 

27 the process of developing DCGLs at a site using either (1) the NRC’s lookup tables in 

28 63 FR 64132 and 64 FR 68395, or (2) the latest version of the DandD code developed by the 

29 NRC to perform the generic screening analysis.

30 When licensees either (1) select other approaches or models for the dose analysis or (2) modify 

31 the DandD code default parameters, exposure scenarios, or pathways, the NRC staff considers 

32 licensees to be performing site-specific analyses. With regard to footnote “a” of Table H.1, the 

33 use of values of the fraction of removable surface contamination other than 0.1 or 1.0 (as 

34 described in the footnote) in the DandD code is considered a site-specific analysis and the staff 

35 should use Section 5.2 to review it.

36 While there is no requirement that licensees consider the use of screening criteria, they should 

37 recognize the advantages and disadvantages of selecting a screening approach for 

38 demonstrating compliance with the dose criteria. Section 2.6 of this volume discusses the 

39 merits of using screening versus using site-specific analysis.

H-2
1 H.1.3 Qualification of the Site for Screening

2 The NRC staff should be aware that a screening analysis, for demonstrating compliance with 

3 the dose criteria in 10 CFR Part 20, Subpart E, may not be applicable for certain sites because 

4 of the status of contaminants (e.g., location and distribution of radionuclides), or because of 

5 site-specific physical conditions. Therefore, the staff should assess the source characteristics 

6 (e.g., spatial distribution of residual radioactivity) to ensure consistency with the source 

7 configuration assumptions in the DandD code. Further, the NRC may determine that there 

8 could be conditions at the specific site that cannot be handled by the simple screening model, 

9 because of the complex nature of the site or because of the simple conceptual model in the 

10 DandD screening code.

11 When using the screening approach for demonstrating compliance with the dose criteria in 

12 10 CFR Part 20, Subpart E, licensees need to demonstrate that the particular site conditions 

13 (e.g., physical and source conditions) are compatible and consistent with the DandD model 

14 assumptions (NUREG/CR-5512, Volume 1). In addition, the default parameters, exposure 

15 scenarios, and pathways must also be used in the screening dose analysis. Therefore, 

16 reviewers should examine the site conceptual model, the generic source term characteristics, 

17 and other attributes of the site to ensure that it is qualified for screening.

18 The NRC staff should verify that the following site conditions exist for each of the residual 

19 radioactivity conditions:

20

21 • Building Surface Residual Radioactivity:

22 o The residual radioactivity on building surfaces (e.g., walls, floors, ceilings) should be

23 surficial and nonvolumetric (e.g., ≤10 millimeters (mm) (0.39 inches (in.)) of

24 penetration).

25 o Residual radioactivity on surfaces is mostly fixed (not loose), with the fraction of

26 loose (removable) residual radioactivity no greater than 10 percent of the total

27 surface activity. Note, for cases when the fraction of removable contamination is

28 undetermined or higher than 0.1, licensees may assume, for screening purposes,

29 that 100 percent of surface contamination is removable, and therefore the screening

30 values should be decreased by a factor of 10 (see footnote “a” to Table H.1).

31 o The screening criteria are not being applied to surfaces such as buried structures

32 (e.g., drainage or sewer pipes) or equipment within the building without adequate

33 justification; such structures, buried surfaces, and clearance of equipment should be

34 treated on a case-by-case basis.

35 • Surface Soil Residual Radioactivity:

36 o The initial residual radioactivity (after decommissioning) is contained in the top layer

37 of the surface soil (e.g., approximately 15 cm (6.0 in.)).

38 o Subsurface soil (e.g., approximately 15 cm (6.0 in.) or greater below the surface) in

39 the unsaturated zone and the groundwater are initially free of residual radioactivity.

40 o The vertical saturated hydraulic conductivity at the specific site is greater than the

41 infiltration rate (e.g., there is no ponding or surface runoff).

H-3
1 Questions have also been raised about the appropriateness of using a screening analysis at 

2 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 3 )). Initially, the NRC staff evaluated the 

4 effect of a large contaminated area on the derived screening dose and determined that this 

5 effect is trivial for sites with the dominant dose arising from direct exposure or inhalation. As 

6 modeled by DandD with its default parameter set, this effect could be appreciable for sites with 

7 a significant dose contribution associated with the ingestion pathway (specifically ingestion 

8 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 9 ), the dose may be 

10 underestimated under worst-case conditions by a factor of 2 to 3. However, further staff 

11 analysis showed that, because of the conservative assumptions of the DandD code, it is more 

12 likely that the derived dose (based on the use of other codes or the use of a site-specific 

13 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 14 ), the change in actual risk due to this effect is 

15 not appreciable. In summary, assuming that the site is qualified for screening based on the 

16 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 17 )).

18

19 It should be noted that the NRC staff should also evaluate complex site conditions that may 

20 disqualify the site for screening. Examples of such complex site conditions may include highly 

21 fractured formation, karst conditions, extensive surface water contamination, and highly 

22 nonhomogeneous distribution of residual radioactivity. Therefore, reviewers should ensure that 

23 the site meets the definition of a “simple site” to qualify for screening (see Section 1.2 of this 

24 document for additional details).

25

26 H.1.4 Acceptable Screening Tools

27 In the past, it may not have been clear what screening tools the NRC has determined to be 

28 acceptable. Some may believe that using simple, common codes (other than DandD), with their 

29 deterministic default parameters, may be acceptable to derive the desired screening derived 

30 concentration guideline levels (DCGLs). Others may believe that the use of any lookup tables 

31 published by certain scientific committees or authorities may be used to convert concentration 

32 levels directly into doses for purposes of complying with Subpart E. Questions have also been 

33 raised on the use of the DandD code for screening, particularly whether modifying input default 

34 parameters is acceptable for screening.

35

36 The NRC staff should accept, for screening analyses, the following currently available screening 

37 tools:

38

39 • a lookup table (Table H.1) for common beta- and gamma-emitting radionuclides for

40 building-surface residual radioactivity (63 FR 64132; November 18, 1998)

41 • a lookup table (Table H.2) for common radionuclides for soil surface residual

42 radioactivity (64 FR 68395; December 7, 1999)

43 • screening levels derived using the latest version of DandD for the specific radionuclide

44 and using code default parameters and parameter ranges

45 The screening values in Tables H.1 and H.2 are intended for single radionuclides. For 

46 radionuclides in mixtures, the “sum of fractions” rule should be used (see Section 2.7 of this 

47 volume). Table H.1 values that provide screening values for beta and gamma emitters 

G-25

H-4
1 associated with building surfaces were derived using DandD Version 1 and its deterministic, 

2 default parameter set, before the development of probabilistic DandD Version 2. Table H.2 

3 values were derived using a version of DandD that was similar to and the predecessor of 

4 DandD Version 2. The screening values in Table H.2 are based on the selection of the 90th 

5 percentile of the output dose distribution for each specific radionuclide or radionuclide with the 

6 specific decay chain. Behavior parameters were set at the mean of the distribution of the 

7 assumed critical group. The metabolic parameters were set either at the Reference Man or at 

8 the mean of the distribution for an average human.

9

10

11 For a radionuclide with its progeny present at equilibrium, the “+C” values of Table H.2 should 

12 be interpreted carefully. As described in footnote “c” to Table H.2, these “+C” values are 

13 concentrations of the parent radionuclide only but account for dose contributions from the 

14 complete chain of progeny in equilibrium with the parent radionuclide. For example, U-238+C 

15 lists the soil screening value as 18.5 Bq/kg (0.5 pCi/g). This means that it also assumed the 

16 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.

17

18

19 The current NRC staff position is to limit screening to its lookup tables or the execution of the 

20 latest version of DandD code with the default parameters and distributions. As indicated above, 

21 the NRC staff may develop additional lookup tables or modify the screening tables based on 

22 refining certain sensitive parameters in the future. For example, NUREG-1720 documents the 

23 NRC staff’s evaluation of the default resuspension factor used in DandD Version 1.0. 

24 NUREG-1720 provides specific recommendations related to the use of the resuspension factor 

25 in screening analyses for the building occupancy exposure scenario. If site conditions are 

26 consistent with assumptions in NUREG-1720 (with respect to activities and scenarios, 

27 ventilation conditions, and low removable fractions at the time of decommissioning), the NRC 

28 staff has determined that it is acceptable to use the NUREG-1720-recommended resuspension 

29 factor value or parameter distribution with minimal justification. However, until such time as the 

30 NUREG-1720-recommended parameter value or distribution is used in developing an updated 

31 screening table for the building occupancy exposure scenario, the NRC staff considers use of 

32 the NUREG-1720 values a site-specific analysis (because the underlying assumptions in 

33 NUREG-1720 must be verified by the licensee before use of the recommended resuspension 

34 factor values). The NUREG-1720-recommended resuspension factor is approximately 10 times 

35 less than the default value used in DandD Version 1 and would, thus, lead to significantly higher 

36 (or less restrictive) screening values for radionuclides dominated by the inhalation pathway 

37 (e.g., alpha-emitting radionuclides), had such screening values been published in 

63 FR 641322 38 . 

39

40 ORISE (2017) provides screening values for sites potentially contaminated with discrete 

41 sources of Ra-226 developed using the DandD code and three exposure scenarios including (i) 

42 industrial building occupancy, (ii) residential building occupancy, for residual radioactivity on 

43 building surfaces; and (iii) resident farmer for residual radioactivity associated with soils. Site44 specific dose modeling is required for sites with groundwater contamination or with site 

45 conditions that are otherwise inconsistent with the dose modeling assumptions. The screening 

46 values are stated to be inappropriate for use for other types of Ra-226 contaminated 

47 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. 

H-5
1 Table H.1 Acceptable License Termination Screening Values of Common Radionuclides 

2 for Building-Surface Contamination 

Radionuclide Symbol Acceptable Screening Levelsafor 

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 % of surface contamination is removable,

and therefore the screening levels should be decreased by a factor of 10. Users may calculate sitespecific 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

note 4, of footnote 3 to 10 CFR Part 20, Appendix B for a description of sum of fractions. An unofficial

copy of the note is available on the NRC website at https://www.nrc.gov/reading-rm/doccollections/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. 

H-6
Table H.2 Screening Valuesa 1 (pCi/g) of Common Radionuclides for Soil Surface 

2 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
1 Table H.2 Screening Values (pCi/g) of Common Radionuclides for Soil Surface

2 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.

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).

H-8
1 H.2 References

2 Code of Federal Regulations (CFR). 10 CFR Part 20, “Standards for Protection against 

3 Radiation.” 

4 NRC. “Residual Radioactive Contamination from Decommissioning: Technical Basis for 

5 Translating Contamination Levels to Annual Total Effective Dose Equivalent,”

6 NUREG/CR-5512, Volume 1. NRC: Washington, DC. 1992.

7 — — — — —. “Residual Radioactive Contamination from Decommissioning, User’s Manual, 

8 DandD Version 2.1,” NUREG/CR-5512, Vol. 2. NRC: Washington, DC. 2001.

9 — — — — —. “Residual Radioactive Contamination from Decommissioning, Parameter 

10 Analysis, Draft Report for Comment,” NUREG/CR-5512, Vol. 3. NRC: Washington, DC. 1999.

11

12 — — — — —. “Supplemental Information on the Implementation of the Final Rule on 

13 Radiological Criteria for License Termination,” Federal Register: Vol 63, p. 64132. November 

14 18, 1998.

15 — — — — —. “Supplemental Information on the Implementation of the Final Rule on 

16 Radiological Criteria for License Termination,” Federal Register: Vol 64, p. 68395. December 7, 

17 1999.

18 — — — — —. “Re-Evaluation of the Indoor Resuspension Factor for the Screening Analysis of 

19 the Building Occupancy Scenario for NRC’s License Termiantion Rule—Draft Report for 

20 Comment,” NUREG-1720. NRC: Washington, DC. 2002.

21 National Council on Radiation Protection and Measurements (NCRP). “Screening Models for 

22 Releases of Radionuclides to the Atmosphere, Surface Water, and Ground,” NCRP Report

23 No. 123. NCRP: Bethesda, MD. 1996.

24 — — — — —. “Recommended Screening Limits for Contaminated Surface Soil and Review of 

25 Factors Relevant to Site-Specific Studies,” NCRP Report No. 129. NCRP: Bethesda, MD. 

26 1999.

H-9

TECHNICAL BASIS FOR SITE-SPECIFIC DOSE MODELING 

EVALUATIONS

I-1

1 I.1 Introduction

2 This appendix consists of the technical guidance for the use of site-specific dose modeling, 

3 applicable to Decommissioning Groups 4–7.

4

5 I.1.1 Background

6 On July 21, 1997, the NRC published a final rule on “Radiological Criteria for License 

7 Termination,” in the Federal Register (62 FR 39058), which was incorporated as Subpart E, 

8 “Radiological Criteria for License Termination,” to 10 CFR Part 20, Standards for Protection 

9 against Radiation.” In 1998, the NRC staff developed a draft regulatory guide (DG), 

10 “Demonstrating Compliance with the Radiological Criteria for License Termination” (DG–4006) 

11 (NRC, 1998e), and a draft document, NUREG-1549, “Decision Methods for Dose Assessment 

12 to Comply with Radiological Criteria for License Termination” (NRC, 1998d), in support of the 

13 final rule. In addition, the staff developed a screening code “DandD” for demonstrating 

14 compliance with the dose criteria in 10 CFR Part 20, Subpart E.

15

16 On July 8, 1998, the Commission approved publication of the draft guidance, DG-4006, the draft 

17 NUREG-1549, and the DandD screening code for interim use for a 2-year period (i.e., from 

18 July 8, 1998, through July 7, 2000), in “Staff Requirements—SECY–98–051—‘Guidance in

19 Support of Final Rule on Radiological Criteria for License Termination,’” dated July 8, 1998 

20 (NRC, 1998c). In addition, the Commission directed the staff to develop a SRP for 

21 decommissioning and provide the Commission with a timeline for developing it, maintain a 

22 dialogue with the public during the interim period, address areas of excessive conservatism, 

23 particularly in the DandD screening code, develop a more user-friendly format for the guidance, 

24 and use a probabilistic approach to calculate the TEDE to the average member of the critical 

25 group (NRC, 1998c).

26

27 The NRC staff published NUREG-1727, “NMSS Decommissioning Standard Review Plan,” in 

28 September 2000. Chapter 5 of the SRP (which is incorporated into Chapter 5 of this volume) 

29 addresses the NRC staff review of the licensee’s dose modeling to demonstrate compliance 

30 with the criteria in 10 CFR Part 20, Subpart E. The NRC staff developed Appendix C of the 

31 SRP (Appendix I of this volume) as a technical information support document for performing 

32 staff evaluations of the licensee’s dose modeling. It presents detailed technical approaches, 

33 methodologies, criteria, and guidance to the staff reviewing dose modeling for demonstrating 

34 compliance with the dose criteria in 10 CFR Part 20, Subpart E. To develop Appendix C of the 

35 SRP, the NRC used an iterative process with the public, which included licensees, Federal 

36 agencies, States, and other interested individuals. To support this process, the staff conducted 

37 seven public workshops and gave several presentations to national and international 

38 professional groups, stakeholders, the Interagency Steering Committee on Radiation Standards, 

39 and the Conference of Radiation Control Program Directors, as well as to the NRC’s Advisory 

40 Committee on Nuclear Waste. In addition, the NRC posted the draft Appendix C (formerly the 

41 Technical Basis Document) on the NRC’s website and invited interested individuals to 

42 comment.

43

44 Since the publication of the license termination rule (LTR), the NRC staff has tested the DandD 

45 code for complex sites and addressed the issue of excessive conservatism in the DandD code. 

46 In addition, the NRC developed a new probabilistic DandD code (i.e., DandD Version 2.1) to 

47 reduce the excessively conservative approach in the initial version. Further, the staff developed 

48 the RESRAD-ONSITE and RESRAD–BUILD probabilistic codes for site-specific analysis, which 

49 also responds to the Commission’s direction to use a probabilistic approach to calculate the 
I-2

1 TEDE to the average member of the critical group. Later, RESRAD-OFFSITE was developed 

2 and also included probabilistic capabilities.

3

4 Licensees using probabilistic dose modeling should use the “peak of the mean” dose 

5 (see Section I.7.3.2 from Appendix I of this volume) for demonstrating compliance with 

6 10 CFR Part 20, Subpart E. Similar to all regulatory guidance, this NUREG contains one 

7 approach acceptable to the NRC staff for determining compliance with the regulations 

8 using probabilistic analyses. Use of “mean of the peaks” is also acceptable for 

9 demonstrating compliance. If the “mean of the peaks” dose is significantly higher than 

10 the “peak of the mean” dose, “risk dilution” may be an issue in the probabilistic model. 

11 Consult Appendix Q for more information on the potential for and impacts of “risk 

12 dilution.” To use any probabilistic approach to calculate DCGLs, the licensee should 

13 discuss its approach with the NRC staff.

14

15 I.1.2 Brief Description and Scope

16 This section is divided into the following different topic areas, as summarized below.

17

18 • Section I.2 presents NRC approaches for reviewing the conceptual representation of the

19 distribution and release of residual radioactivity from soils and building materials. This

20 section describes areas of review for conceptualization and representation of the source

21 and source term in dose modeling used to demonstrate compliance with radiological

22 criteria for license termination

23 • Section I.3 focuses on areas of review and criteria for modifying the two generic

24 exposure scenarios used in screening-level analyses: modifications to (1) the “resident

25 farmer” and (2) the “building occupancy” exposure scenarios. Section I.3, and

26 Appendices L and M, discuss the type of information a licensee should provide to

27 support the modification of default (screening) exposure scenarios. Section I.3 also

28 presents approaches for establishing site-specific exposure scenarios and pathways

29 reflecting the activities and behaviors of identified critical group(s), based on current and

30 reasonably foreseeable future land use, site restrictions, and other physical conditions

31 associated with a decommissioning site.

32 • Section I.4 provides approaches for developing conceptual and mathematical models in

33 dose modeling. This section presents approaches for collecting and using

34 characterization data to develop conceptual and mathematical models of the site and

35 issues associated with model simplification and abstraction. This section also includes

36 information on the underlying conceptual models in the DandD and RESRAD codes and

37 associated limitations.

38 • Section I.5 presents approaches and criteria for NRC staff acceptance of computer

39 codes/models. This section discusses review aspects pertaining to specifications,

40 testing, verification, documentation, and QA/QC of the licensee’s codes and models.

41 This section also addresses reviews applicable to embedded numerical models for the

42 source term, fate and transport, and biosphere (exposure) modeling.

43 • Section I.6 describes approaches for the selection and modification of input parameters

44 used in dose modeling.
I-3

1 • Section I.7 provides information on evaluating uncertainty and identifying parameters

2 most important to dose in analyses used to demonstrate compliance with LTR criteria.

3 • Section I.8 compiles the references used throughout the appendix.

4

I.2 Source and Source Term Abstraction1 5

6 I.2.1 Introduction

7 Source abstraction is the process of developing a conceptual representation of the residual 

8 radioactivity present at a site, focusing on the geometry and distribution of contamination in the 

9 environment. Typically, the radiological conditions at a site undergoing decommissioning are 

10 relatively complex. Source abstraction is necessary to allow the detailed radiological 

11 characterization of the site to be incorporated into the mathematical and computer models that 

12 are used to demonstrate compliance with the dose-based criteria in 10 CFR Part 20, Subpart E. 

13 The abstraction process involves generalizing the radiological characteristics across the site to 

14 produce a simplified representation, which should facilitate the modeling of radiological impacts. 

15 This guidance makes a distinction between parameters that characterize the source, including 

16 parameters related to the concentration of radionuclides, thickness, and area of residual 

17 radioactivity; and those parameters that characterize the release of radioactivity in the 

18 environment (i.e., that help define the source term). The source term considers parameters that 

19 define the source concentrations and configuration of residual radioactivity, as well as the 

20 release mechanism (e.g., solubility–controlled, desorption, or diffusion-limited release). 

21 Although source and source term abstraction are a necessary part of the dose modeling 

22 process, the licensee should take care to ensure that the conceptual representation of the 

23 source and source term developed in the abstraction process are not oversimplified in a manner 

24 that results in the underestimation of potential radiological impacts. 

25

26 As discussed in Chapter 5 of this volume, source configuration and source term abstraction 

27 serve as the starting point for the dose modeling process. The conceptual abstraction of the 

28 source and source term is combined with (1) the physical characteristics of the site, and

29 (2) characteristics of the average member of the critical group to develop a conceptual model for

30 the site being studied. Thus, the conceptual model also provides a representation of the natural

31 environment through which radioactivity can be transported, as well as applicable exposure

32 scenarios and pathways to members of the public who may be exposed to the radioactivity.

33 The conceptual model can be used to determine acceptability of use of computer codes that

34 contain their own built-in conceptual models with respect to how residual radioactivity is

35 transported in the environment and how humans can be exposed to radioactivity.

36

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., solubilitycontrolled, 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”.
I-4

1 Volume 1 of NUREG-1757 and Chapter 4 of this volume discuss the information the licensee is 

2 expected to provide about the existing radiological characterization of the site. The licensee 

3 should describe the types, levels, and extent of residual radioactivity in contaminated materials 

4 at the site. This should include residual radioactivity in all media (including buildings, systems, 

5 and equipment that will remain after license termination; surface and subsurface soil; and 

6 surface water and groundwater). The source configuration and source term abstraction should 

7 be based on the characterization of the radiological status (e.g., historical site assessment;

8 records of leakage or disposal). The licensee should explicitly relate the information provided in 

9 the discussion of radiological status of the site with its assumptions on the source configuration 

10 in dose modeling. The reviewer should be able to clearly interpret the relationship.

11

12 Generally, in the source abstraction process, the licensee may focus on several specific 

13 elements, which include the following:

14

15 • The licensee should identify the radionuclides of concern, taken directly from the

16 description of the site’s radiological status. The licensee should identify the

17 radionuclides based on the pre-remediation radiological status. It should include all

18 radionuclides potentially present at the site, so that their presence or absence may be

19 verified during the FSS, except as noted in Chapter 4, “Facility Radiation Surveys,” and

20 Section 3.3, “Insignificant Radionuclides and Exposure Pathways,” of this volume.

21 • The licensee should describe the physical/chemical form(s) of the contaminated media

22 anticipated at the time of FSS and site release. The licensee should indicate whether

23 the residual radioactivity will be limited to building surfaces or surface soil, or both, or

24 whether the residual radioactivity will involve other media, such as subsurface soil,

25 debris or waste materials (e.g., sludge, slag, tailings), or ground and surface water.

26 Information on the physical/chemical form(s) of the contaminated media will also help

27 determine whether source term assumptions are appropriate.

28 • The licensee should delineate the spatial extent of the residual radioactivity anticipated

29 at the time of FSS and site release. The delineation of the spatial extent should include

30 descriptions of (1) the areal extent of radionuclides throughout the site and (2) the

31 vertical extent of soil residual radioactivity of radionuclides below the ground surface.

32 The delineation of spatial extent and depth should establish the source areas and

33 volumes. Source areas and volumes may differ for individual radionuclides.

34 • The licensee should define the distribution of each radionuclide throughout the

35 delineated source areas and volumes anticipated at the time of FSS and site release.

36 The distribution of a radionuclide through the source should be defined in terms of

37 representative volumetric or areal concentrations. In addition, for volumetrically

38 contaminated soil, the licensee may provide an estimate of total activity of each

39 radionuclide.

40 • The licensee should define sources in surface water or groundwater, if any, based on

41 environmental monitoring and sampling of aquifers and surface water bodies. A site with

42 groundwater or surface water contamination may be categorized as “complex” and may

43 require more advanced dose modeling analysis (see Appendix F for additional

44 information on surface water and groundwater characterization).

45 In the source abstraction process, the licensee should identify the radionuclides of concern and 

46 have sufficient information to determine if residual radioactivity is surficial and relatively 
I-5

1 homogeneous. Depending on the dose modeling approach, the licensee may or may not need 

2 to address the other elements, as discussed in more detail later in this section.

3

4 I.2.2 Issues Associated with Source Term Abstraction

5 The level of effort that a licensee expends to develop a conceptualization of the source and 

6 source term should be commensurate with the complexity of the site and the licensee’s 

7 approach to demonstrating compliance with the release criterion (i.e., screening versus site8 specific). Also, the focus should be on the source and source term characteristics anticipated to 

9 exist at the site at the time of FSS and release, after any planned remediation.

10

11 If a licensee plans to use the screening DCGLs published by the NRC in the Federal Register 

12 (see Appendix H), a licensee should only have to identify the radionuclides present at the site 

13 and demonstrate that the conditions at the site meet the prerequisites for using the screening 

14 values (i.e., residual radioactivity is limited to building surfaces or approximately the uppermost 

15 15 centimeters (cm) (6 inches (in.)) of surface soil and show there is no contamination of 

groundwater or surface water2 16 ), as discussed further in Section I.2.3 of this appendix. 

17

18 If a licensee anticipates that residual radioactivity will be limited to building surfaces or surface 

19 soils at the time of the FSS but considers the published DCGLs overly restrictive, it may develop 

20 site-specific DCGLs. In this case, the licensee would most likely have to delineate the 

21 anticipated areal extent and depth of residual radioactivity. However, the licensee would not 

22 have to discuss the anticipated spatial variability of radionuclide concentrations within the 

23 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 24

25

26 The licensee should provide a site-specific dose assessment if the residual radioactivity is not 

27 limited to building surfaces or surface soil. In this case, the licensee would have to delineate the 

28 spatial extent (laterally and vertically) of the residual radioactivity and discuss the spatial 

29 variability of the physical, chemical, and radiological characteristics of the contaminated media.

30

31 Ideally, the source characteristics at a site would be relatively uniform, justifying simplified 

32 source abstraction. However, this is often not the case. Issues may arise when the residual 

33 radioactivity projected at a site at the time of release is inconsistent with the ideal case. These 

34 issues may include the following:

35

36 (1) Spatial extent

37 o limited areal extent of residual radioactivity

38 o irregular areal shape

2 If surface water or groundwater are contaminated, it may still be possible to use screening values if the dose 

contributions from the surface water or groundwater contamination 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. 
I-6

1 o varying depth of residual radioactivity in soil

2 (2) Spatial variability

3 o nonuniform distribution of radioactivity throughout a site

4 o limited areas of relatively elevated radionuclide concentrations

5 o multiple noncontiguous areas of residual radioactivity

6 o nonuniform physical and chemical characteristics

7 The sections below discuss approaches to addressing these technical issues.

8

9 I.2.3 Approach to Source Abstraction

10 Source abstraction data needs will depend on the approach used to demonstrate compliance 

11 with radiological criteria for license termination presented in the licensee’s DP. Generally, the 

12 licensee will use one of the two following approaches to dose modeling:

13

14 (1) Develop DCGLs for each radionuclide that would lead to a dose at the release criterion,

15 and then demonstrate through the FSS that median residual radioactivity concentrations

16 at the site are equal to or below the DCGLs with a certain specified level of confidence.

17 (2) Assess dose associated with the actual distribution of residual radioactivity at the site to

18 determine whether the residual radioactivity will result in a dose equal to or below the

19 dose-based release criterion.

20 In the first approach, the licensee intends to demonstrate at the time of the FSS that residual 

21 radionuclide concentrations across the site are below a prespecified concentration limit with 

22 some prespecified degree of confidence. The design of the FSS would be based on the 

23 proposed DCGLs, in accordance with NUREG-1575, “Multi-Agency Radiological Survey and 

24 Site Investigation Manual (MARSSIM)” (NUREG, 2000a). The MARSSIM process considers 

25 variability in determining the number of samples needed to demonstrate compliance with the 

26 radiological criteria for license termination. DCGLEMCs are used to account for the dose 

27 contributions of smaller elevated areas of residual radioactivity. Knowledge about the 

28 characteristics (e.g., area and thickness) of elevated areas may also assist with the 

29 development of DCGLEMCs. 

30 In the second approach, the licensee intends to assess potential radiation doses that may result 

31 from specified levels of radioactive material. The contaminated material may not be limited to 

32 building surfaces or surface soils but may include contaminated subsurface soil, debris, and 

33 waste. The licensee’s dose modeling should demonstrate that the residual radioactivity should 

34 not result in radiation doses in excess of applicable regulatory limits. Most likely, this modeling 

35 approach would require that the licensee incorporate information on both the spatial extent and 

36 spatial variability of radioactivity in the source abstraction.

37 Table I.1 summarizes source abstraction information needs, depending on the licensee’s dose 

38 modeling approach and whether the licensee is providing screening or site-specific analyses. 

39 This table can serve as an index for the reviewer of the licensee’s source abstraction.

40
I-7

1 Table I.1 Summary of Source Abstraction Information Needs for Two Types of Dose2 Modeling Approaches (Screening versus Site-Specific) 

Approach to 

Demonstrating 

Compliance

Screening Site-Specific

DCGLs No source term abstraction is 

necessary beyond radionuclide 

identification.

(assume unit radionuclide 

concentrations to calculate 

DCGLs (i.e., to calculate what 

concentration leads to a dose at 

the limit))

Delineate proposed lateral and 

vertical extent of residual 

contamination. Smaller areas of 

elevated activity can be 

considered through DCGLEMCs.

(assume unit radionuclide 

concentrations to calculate DCGLs 

(i.e., to calculate what 

concentration leads to a dose at 

the limit))

Dose Modeling Use actual concentrations with 

DandD Version 2 (or later 

version) and ensure that residual 

radioactivity is surficial and 

spatial variability is minimal.

Site–specific source abstraction 

incorporates spatial extent and 

variability.

3

4 I.2.3.1 Dose Modeling Approach One: Develop DCGLs

5 The MARSSIM approach, as documented in NUREG-1575 (NRC, 2000a) and discussed in 

6 Chapter 4 of this volume, requires that a licensee establish a set of DCGLs before conducting 

7 an FSS. In fact, the design of the FSS should be based on the identified DCGLs. DCGL is 

8 defined in MARSSIM as:

9

10 “...a derived, radionuclide-specific activity concentration within a survey unit 

11 corresponding to the release criterion....DCGLs are derived from activity/dose 

12 relationships through various exposure pathway scenarios.”

13

14 The DCGLW is the concentration of a radionuclide that, if distributed uniformly across a survey 

15 unit, would result in an estimated dose equal to the applicable dose limit. The DCGLEMC is the 

16 concentration of a radionuclide that, if distributed uniformly across a smaller limited area within a 

17 survey unit, would result in an estimated dose equal to the applicable dose limit. DCGLEMC is 

18 specific to the size of the area for which it is derived.

19

20 Two approaches are possible for developing DCGLs: screening and site-specific analysis.

21

22 SCREENING DCGLs

23

24 The NRC has published radionuclide-specific screening DCGLs in the Federal Register for 

25 residual building-surface radioactivity and residual surface-soil radioactivity (see Appendix H, 

26 Table H-1 and H-2; or Table 5.19 and 6.91 in NUREG/CR-5512, Volume 3 for additional 

27 radionuclides). The DCGLs in the Federal Register are intended to be concentrations that, if 

28 distributed uniformly across a building or soil surface, would individually result in a dose equal to 
I-8

1 the dose criterion. The licensee may adopt these screening DCGLs without additional dose 

2 modeling, if the site is suitable for screening analysis. Alternatively, the licensee may use the 

3 DandD computer code to develop screening DCGLs. The licensee would use the code to 

4 determine the dose attributable to a unit concentration of a radionuclide and scale the result to 

5 determine the DCGLW for the radionuclide. Either of these methods for identifying screening 

6 DCGLs requires the licensee (1) to identify the radionuclides of concern for the site and (2) to 

7 demonstrate that the source and model screening assumptions are satisfied. Thus, this 

8 approach requires essentially no source abstraction. 

9

10 Typically, before designing an FSS, the licensee identifies a DCGLEMC for each radionuclide for 

11 smaller limited areas of radioactivity (e.g., for areas between sampling locations). However, the 

12 underlying assumption for use of screening analyses is that the residual radioactivity is 

13 homogeneous. By default, DandD calculates dose based on the assumption that the size of the 

14 contaminated area is “unlimited.” Although DandD provides an option to allow the user to enter 

15 a “limited area” of contamination, due to the simplistic manner in which DandD corrects the dose 

16 for smaller, limited areas of contamination, DandD is not ideally suited for calculating 

17 DCGLEMCs. Therefore, it is recommended that licensees use other codes or approaches to 

18 develop DCGLEMC values, if there is a need for allowing higher concentrations above the 

19 DCGLW in smaller areas between sampling locations. These would be considered “site-specific” 

20 analyses in that they would not be using the DandD code with the default screening values. 

21 Section I.3.3.3.5 of this appendix contains additional information on this topic. If the licensee 

22 can show that residual radioactivity is relatively homogeneous and concentrations of 

23 radionuclides averaged over relatively small exposure areas (areas that are consistent with the 

24 exposure pathway assumptions in DandD) are less than the screening values or DandD-derived 

25 DCGLs, then development of DCGLEMC values should not be needed. Consult the DandD 

26 online help for more information on calculating the average concentrations for use in the DandD 

27 code.

28

29 SITE-SPECIFIC DCGLs

30

31 The licensee may choose to identify site-specific DCGLs if (1) the site conditions are not 

32 consistent with screening criteria or (2) the licensee believes the screening DCGLs are 

33 unnecessarily restrictive. As defined in MARSSIM, Rev. 1 (2000a) the licensee may derive site34 specific DCGLs from activity or dose relationships through various exposure pathway scenarios. 

35 “Site-specific” in this context may refer to the selection of conceptual models/computer models, 

36 physical (site) input parameter values, or behavioral or metabolic input parameter values. 

37 These aspects of site-specific analyses are discussed in other sections of this document. “Site38 specific” may also refer to the source or source term abstraction.

39

40 From the MARSSIM perspective, identifying a site-specific DCGLW still begins with assuming a 

41 uniform radionuclide concentration across some source area (building surface) or volume of 

surface soil4 42 . The site-specific DCGLW for a particular radionuclide may be identified by 

43 evaluating the dose resulting from a unit concentration and then scaling the result to the dose 

44 limit. Spatial variability of the radionuclide concentration within the area or volume is not 

45 evaluated in calculating the DCGLs but is taken into account in the statistical analysis of the 

46 data collected during the FSS. When developing the site-specific DCGLs, the licensee should, 

47 however, take the spatial extent into account, including the horizontal and vertical extent of 

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.”
I-9

1 contamination. While some examples in MARSSIM use the top 15 cm of soil as surface soil, for 

2 the purpose of compliance measurements, licensees should consider the actual depth of 

3 residual radioactivity which should be consistent with the dose model used to generate site 

4 DCGLs. Soil sampling should be representative of the chosen surface soil depth, and surface 

5 scanning methods should be capable of detecting contamination within the entire depth of 

6 surface soil. Accordingly, scan MDCs should be determined based upon the depth of 

7 contaminated soil associated with the dose model used to generate surface soil DCGLs. If 

8 subsurface residual radioactivity is present, dose modeling may be conducted for both surface 

9 and subsurface soils and DCGLs developed for each. In these cases, the MARSSIM 

10 methodology will need to be supplemented or an alternative methodology will need to be 

11 developed to demonstrate compliance with radiological criteria for license termination. 

12

13 Through the FSS, the licensee would have to demonstrate that the DCGLW is satisfied within the 

14 specified exposure area for each survey unit. The licensee should consider the area assumed 

15 in the dose modeling for consistency against the area of the survey unit and the area over which 

16 concentrations are averaged for comparison against DCGLs. The licensee should also develop 

17 DCGLEMC values for smaller areas within the larger area of residual radioactivity as part of the 

18 FSS design (e.g., develop DCGLEMC values for smaller areas in the survey unit between soil 

19 sampling locations that may be missed by sampling). If it is certain that the residual 

20 radionuclide concentration is limited to a specific lateral extent, the licensee may also 

21 incorporate the “area of residual radioactivity” into the identification of DCGLs. Computer 

22 modeling codes, such as RESRAD-ONSITE, allow the user to directly specify the area of 

23 residual radioactivity and adjust the dose as a function of the specified area in a unique manner 

24 for each exposure pathway. 

25

26 The NRC recommends that the licensee use dose modeling to calculate DCGLEMCs. For 

27 example, it can use RESRAD-ONSITE and RESRAD-BUILD to calculate dose for the larger 

28 area of concern (i.e., the entire contaminated area) and the dose from the smaller areas of 

29 potentially elevated concentration, separately. It can then use dose modeling results to 

30 calculate the cleanup levels for the larger area of concern and for the smaller elevated areas 

31 (DCGLW and DCGLEMCs, respectively). If only a single elevated area is present, the EMC is 

32 acceptable if Equation 1 is met (modified from MARSSIM, Equation 8.2)

33

34

δ

DCGLW

+

average concentration in the elevated area− δ

DCGLEMC

35 < 1 Equation 1

36

37 Where 



= the average residual radioactivity concentration for all sample points in the 

38 survey unit. 

39

40 In cases where there is more than one elevated area, a separate term should be included in the 

41 calculation for each area of elevated activity. The unity rule is satisfied when radionuclide 

42 mixtures yield a combined fractional concentration limit that is less than or equal to one. In 

43 situations where there is more than one radionuclide at a single source (elevated area or “hot 

44 spot”), the sum of the individual ratios also cannot be greater than or equal to 1. The sum of 

45 fractions rule applies in situations where there are multiple radionuclides and sources. For 

46 example, consider a site where residual radioactivity is present in two elevated areas or hot 

47 spots, areas A and B, as well as in a larger area of concern. The concentrations of 

48 radionuclides 1, 2, and 3, in elevated or hot spot areas A and B, and the larger area, W, must all 

49 be considered and the sum of fractions must be less than or equal to 1 (see Equation 2). Note 

50 that the general area concentrations CXW for each radionuclide, x=1, 2, or 3, can be subtracted 
I-10

1 from the elevated area concentrations, CXA and CXB for each radionuclide, in Equation 2 

2 consistent with Equation 1.

3

C1A

DCGL1A

+

C2A

DCGL2A

+

C3A

DCGL3A

+

C1B

DCGL1B

+

C2B

DCGL2B

+

C3B

DCGL3B

4 +

5 Equation 2

C1W

DCGL1W

+

C2W

DCGL2W

+

C3W

DCGL3W

6 ≤ 1

7

8 The following examples present different options for considering the contributions of multiple 

9 elevated areas within a single larger area. In general, two or more smaller areas modeled 

10 independently and combined will result in a higher dose then if contaminant concentrations are 

11 averaged across a single area. Although the higher dose may be considered conservative, it is 

12 often unrealistic, because it may assume, for example, that an individual spends all of his or her 

13 time in multiple locations simultaneously. The examples below demonstrate how the 

14 assumptions made about the contamination on a site can affect whether or not the site passes 

15 the cleanup criteria.

16

17 Example 1—Base Case

18

For this example, a survey unit of 10,000 m2 19 is uniformly contaminated with americium (Am)-241 

20 at a concentration of 9 pCi/g. The DCGL is 10 pCi/g. Because the concentration (9 pCi/g) is 

21 below the DCGL, the survey unit will pass.

22

• A 10,000 m2 survey unit with uniform Am-241

contamination (no elevated areas)

• Am-241 concentration = 9 pCi/g

• DCGLW = 10 pCi/g

23

9

pCi

g

10pCi

g

24 = 0.9 0.9 < 1.0 → Survey unit passes cleanup criteria

25

26 Example 2—Single Elevated Area in the Survey Unit

27

Example 2 considers a 10,000 m2 survey unit with a single 300 m2 28 elevated area contaminated 

29 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 30 ) is 12. As expected, the smaller contaminated area 

31 within the larger survey unit results in a smaller sum of fraction (0.75), which is less than 1. 

32 Therefore, the survey unit with these characteristics also passes the cleanup criteria.

33

• 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
I-11

1

9

pCi

g

12pCi

g

2 = 0.75 0.75 < 1.0 → Survey unit passes cleanup criteria

3

4 Example 3—Multiple Elevated Areas or Hot Spots in a Single Survey Unit

5

6 As shown in Example 3, evaluating multiple smaller elevated areas often results in a higher 

7 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 8 elevated areas

within the same 10,000 m2 9 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 10 elevated area was present, the dose 

11 criterion could be easily met, summing the dose contributions of the three individual elevated

12 areas together results in a value greater than 1 and therefore, exceeds the cleanup criteria.

13

14

• 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

15

A 100 m2 16 elevated area:

17

9

pCi

g

18.6

pCi

g

18 = 0.48

19

Combining the results of each of the three 100 m2 20 elevated areas:

21

9

pCi

g

18.6

pCi

g

+

9

pCi

g

18.6

pCi

g

+

9

pCi

g

18.6

pCi

g

22 = 1.45

23

24 In this case the higher dose is unrealistic, because it assumes that a receptor spends all of his 

25 or her time on each contaminated area simultaneously. The licensee can determine a more 

26 reasonable estimate of potential dose by combining the individual elevated areas into a single 

27 larger elevated area within the survey unit, as in Example 2. 

28

29 In addition to specifying a limited area of residual radioactivity in developing the site-specific 

30 DCGLs for soil, the licensee should appropriately represent the vertical extent of residual 

31 radioactivity within the area. The screening DCGLs and the DandD code assume that residual 

32 radioactivity is contained within the uppermost 15 cm of soil. If the licensee intends to leave 

33 residual radioactivity at depths below approximately 15 cm, the calculation of the DCGLW should 
I-12

1 reflect a greater thickness of residual radioactivity. Otherwise, leaving residual radioactivity 

2 below 15 cm may not be acceptable.

3

4 For subsurface residual radioactivity (i.e., residual radioactivity at depths greater than 

5 approximately 15 cm), the NRC staff should evaluate whether the licensee has reviewed 

6 existing historical site data (including previous processes or practices) and site characterization 

7 data to establish an adequate conceptual model of the subsurface source specifically about the 

8 horizontal and vertical extent of residual radioactivity. The licensee should evaluate lateral and 

9 vertical trends of variation in concentration for each specific radionuclide. Because certain 

10 radionuclides have higher mobility than others, radionuclide ratios may not be maintained as 

11 constant across subsurface soil. In other words, radionuclide concentrations within the 

12 unsaturated zone may vary, depending on the original source location and the time since the 

13 source existed. The NRC staff should evaluate whether the licensee has evaluated the physical 

14 and chemical properties of the source and the attenuating properties of the subsurface materials 

15 to assess the potential for radionuclide leaching and transport. In this context, the reviewer 

16 should evaluate the selected physical parameters and conceptual model of the site against the 

17 actual subsurface hydrostratigraphy to evaluate the acceptability of the parameter selections if 

18 found to be important to dose. The reviewer should also consider (1) heterogeneity in 

19 subsurface soils, and (2) the depth to the water table if found to be important to dose. 

20

21 If the thickness of residual radioactivity that will remain at a site is generally uniform across the 

22 site, the licensee may choose to use an upper bounding value for modeling the thickness. The 

23 NRC reviewer should evaluate the representative thickness value proposed by the licensee to 

24 ensure that the value selected does not lead to a significant underestimate of dose, particularly 

25 for sites where the vadose thickness is quite variable. For example, given the variable depth to 

26 the water table, a licensee may propose to use an area-weighted approach to assign the 

27 vadose zone thickness in dose modeling. However, if the timing of peak dose from groundwater 

28 dependent pathways is important to the compliance demonstration, then representation of the 

29 area with a thinner vadose zone may be important, for example.

30

31 If appropriate, the licensee should provide maps and cross sections detailing the proposed 

32 lateral and vertical extent of residual radioactivity left on the site.

33

34 I.2.3.2 Dose Modeling Approach Two: Assess Dose Based on Actual Concentrations

35 An alternative objective that a licensee may have for performing and submitting dose modeling 

36 may be to assess doses attributable to specific quantities of radioactive material. Although the 

37 development of DCGLs focuses on the determination of radionuclide concentrations 

38 corresponding to a specified dose, the dose assessment objective focuses on the determination 

39 of doses corresponding to specified radionuclide concentrations. 

40

41 In this situation, the licensee should give much more attention to the source abstraction and 

42 address all elements of the source term abstraction:

43 • identify the radionuclides of concern

44 • delineate the spatial extent of residual radioactivity

45 • represent the spatial variability of residual radioactivity

46 • incorporate the spatial variability of physical and chemical characteristics of the

47 contaminated media
I-13

1 The licensee should focus on the distribution of radioactive material expected to be present at 

2 the time of FSS and subsequent site release. The licensee may assess doses attributable to 

3 existing radiological conditions at the site, if it can demonstrate that the existing radiological 

4 conditions reasonably bound conditions expected at the FSS, from a dose perspective.

5

6 The first two elements of source abstraction—radionuclides of concern and spatial extent—were 

7 considered in the discussion of source abstraction for the development of DCGLs. Spatial 

8 variability was not considered, since it is statistically evaluated as part of the FSS. If the dose 

9 modeling approach using actual radionuclide concentrations is used, however, spatial variability 

10 should be factored into the source abstraction before modeling.

11

12 Assuming that the licensee has identified the radionuclides of concern and delineated the 

13 spatial extent of residual radioactivity, it should project the residual radionuclide concentration 

14 distribution and total residual radionuclide inventory across the site. The licensee should tie this 

15 projection directly to the characterization of existing radiological conditions at the site. The site 

16 may then be divided into relatively large areas that are radiologically distinct, based on 

17 radionuclide concentration or depth of residual radioactivity. The licensee should statistically 

18 demonstrate that the radionuclide concentrations or depth within an area may be relatively 

19 uniform, taking into account the spatial distribution of the data. Similarly, within the larger areas, 

20 the licensee should statistically delineate relatively small areas of projected elevated 

21 radionuclide concentrations or increased depth. (The licensee should discuss the reason for 

22 leaving the elevated concentrations in place as residual radioactivity.)

23

24 When complete, the licensee’s source term abstraction should define a site divided into 

25 relatively large areas of statistically uniform radionuclide concentrations and residual 

26 radioactivity depth. Within these areas may be relatively small areas of elevated concentration 

27 or increased depth. Assuming that the physical and chemical conditions across the site are 

28 relatively uniform, the licensee may use this source abstraction for modeling and proceed with 

29 the dose assessment. 

30

31 The following is a suggested approach:

32

33 • Consider each relatively large area independently and initially ignore the relatively small

34 elevated areas within each large area.

35 • Assess dose based on the properties of a large area, taking the areal extent into

36 account.

37 • Repeat the dose assessment but assume an essentially infinite areal extent. The

38 specific approach will depend on the computer modeling code used. This should

39 quantify the impact of dividing the site into artificial modeling areas.

40 • Assess dose attributable to each limited area of elevated concentration, assuming no

41 residual radioactivity exists outside the limited area. This may then be combined with

42 the dose attributable to the surrounding larger area, to assess the impact of leaving the

43 elevated concentrations.

44 In some cases, it may not be practical to separate a site into areas with relatively uniform 

45 radionuclide concentrations; sometimes areas to be evaluated will have nonuniform distributions 

46 of concentrations. In such cases, for performing the second step above, there may be a 

47 question about what statistical value best represents the radionuclide concentration for the large 
I-14

1 area. Log-normal distributions occur frequently in nature and are not unexpected when 

2 surveying contaminated sites. For log-normal distributions, the geometric mean is often used as 

3 a descriptor of the distribution. However, the geometric mean concentration should not be used 

4 as the average value for the source concentration for dose calculations, as use of the geometric 

5 mean could lead to a significant underestimate of the dose compared to use of the arithmetic 

6 mean. Because (1) the dose rate is proportional to radionuclide concentration, (2) it is 

7 reasonable to assume that the receptor spends an equal amount of time in each area of the 

8 site, and (3) each characterization data point represents an equal area, use of the arithmetic 

9 mean (and not the geometric mean) is more technically defensible for calculating average 

10 concentrations for use in dose modeling (second step above). If samples are not taken 

11 randomly or systematically (and thus data points represent unequal areas), weighted means 

12 may be appropriate, with application of weighting factors consistent with the assumptions of 

13 receptor exposures. 

14

15 The above discussion does not specifically address the determination of relatively significant 

16 large or small areas. This designation will depend on the areal assumptions underlying the 

17 computer modeling code used. For example, the DandD code considers the area of cultivation 

18 to be uniformly contaminated and irrigated. The area of cultivation depends on the cultivation 

19 requirements defined by the specific exposure scenario. While DandD allows the user to 

20 specify either (1) an “unlimited” area of contamination consistent with the underlying conceptual 

21 model, or (2) smaller areas of contamination, the relationship between area and dose is defined 

22 in a simplistic manner. Furthermore, the area of contamination is not treated as an uncertain 

23 parameter and, therefore, an automated sensitivity analysis on the area of contamination is not 

24 available. With respect to the RESRAD family of codes, ingestion pathway doses are largely 

25 dependent on the area of the source; however, the impact of area on dose varies depending on 

26 the specific pathway being considered. For example, soil and plant ingestion doses scale 

directly with areas up to 1,000 m2, 27 5 while animal pathway doses scale directly with areas up to 

20,000 m2 28 . The licensee should discuss and justify the designation of relatively large and 

29 relatively small areas, based on the computer code used. The licensee can provide additional 

30 information (e.g., results of alternative scenarios evaluating the sensitivity of the dose modeling 

31 results to area) to lend more support for the compliance demonstration.

32

33 As discussed above, the licensee may also have to consider the impact of multiple areas of 

34 elevated concentration within a single larger area. In general, modeling two small areas

35 independently and combining the results of the two dose assessments should result in a higher 

36 dose than if the two areas were combined and modeled as a single area (see Examples 1–3 in 

37 Section I.2.3.1). The higher dose may be unrealistic if it assumes that the receptor location 

38 relative to each contaminated area is such that the dose is maximized from each contaminated 

39 area independently. For a more reasonable estimate of potential dose, these smaller areas 

40 may be combined into a single larger area, if the concentrations within the smaller areas are 

41 comparable (e.g., see Example 3 in Section I.2.3.1). If this is not the case, then the licensee 

42 may model each smaller area individually and modify the exposure scenario and critical group 

43 assumptions for each area (e.g., time spent on each area) and combine the results. 

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 and RESRAD-OFFSITE assume 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 m2is 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.
I-15

1 The example illustrated in Figure I.1 presents an acceptable method for considering the 

2 contributions of multiple elevated areas or “hot spots” within a larger contaminated area, when 

3 the concentrations of the elevated areas are variable with each other, as well as with the larger 

4 area of residual radioactivity. Consider a site with two relatively small areas of elevated 

5 radioactivity in comparison to levels of radioactivity over a much larger area. Because most 

6 dose modeling codes, including DandD and RESRAD-ONSITE, assume that the receptor is

7 located in the center of the contaminated area, the “conceptual model” in Figure I.1 depicts 

8 overlapping contaminated areas. However, to avoid overestimating the dose when the 

9 contaminated areas are overlaid and summed, the larger contaminated area is run first with an 

10 average concentration of 3 pCi/g. Next, the second most elevated area or “hot spot,” which has 

11 an average concentration of 10 pCi/g, is simulated with a concentration of 3 pCi/g less than the 

12 average concentration of 10 pCi/g (or with a value of 7 pCi/g), so as not to double-count the 

13 activity in the larger area to which it is summed. The second most elevated area or hot spot is 

assumed to be 200 m2instead of 100 m2, to account for the fact that 100 m2 14 of the simulated 

contaminated area overlaps the most elevated 100 m2 15 area that is considered separately. 

16 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 17 area, (see yellow 

18 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 19 areas at the same 

20 time, and assuming that the receptor is located in the center of the hottest area on site for the 

21 entire exposure period is conservative. Finally, the most elevated area or hot spot, which has 

22 an average concentration of 16 pCi/g, is simulated with a concentration 10 pCi/g less than the 

23 average concentration of 16 pCi/g (or with a value of 6 pCi/g), so as not to double-count the 

24 activity assigned in the larger area simulation of 3 pCi/g, and not double-count the activity 

25 assigned in the second most elevated area simulation of 7 pCi/g. Although the geometry and 

26 locations of the elevated areas or “hot spots” differ in the “conceptual model” versus the “actual” 

27 configuration depicted in Figure I.1, the assumed geometry and elevated area location tends to 

28 overestimate the dose with the receptor standing directly on top of the hottest contaminated 

29 area on the site and in relatively close proximity to the second most contaminated area on site. 

30 Depending on the actual size and geometry of the elevated areas being simulated, this method 

31 may produce overly conservative results. If less conservative methods are needed to 

32 demonstrate compliance, the licensee may propose alternative methods that will require 

33 approval by NRC reviewers on a case-by-case basis.

34 Figure I.2 illustrates how vertical heterogeneity can be considered in dose modeling. In the 

35 example, variability in concentration for the first 5 cm of soil and the next 10 cm of soil is 

36 considered in the dose modeling. In the example, two simulations are run and the doses from 

37 the simulations added together. In the first simulation, the entire 15 cm of soil is considered but 

38 using the lower concentration of the deeper 5 to 15 cm soil interval. In the second simulation, 

39 the top 5 cm of soil is considered using the difference in concentration between the shallower 0 

40 to 5 cm and the deeper 5 to 15 cm soil intervals. The doses from the two simulations are 

41 summed to provide a dose estimate for the two soil intervals represented in the model.

42
I-16

1

2 Figure I.1 Approach to Considering Multiple Elevated Areas or Hot Spots

3
I-17

1

2 Figure I.2 Approach to Considering Vertical Heterogeneity

3

4
I-18

1 Case Study 1: Evaluation of the Impact of Variability of Exposure Area Concentrations

2 Using the dose modeling approach, analysts can evaluate the uncertainty in dose modeling 

3 predictions based on spatial variability in exposure area concentrations using a code such as 

4 SADA. A case study by the NRC staff used soil sampling data to calculate average 

5 concentrations for various exposure areas at a thorium-contaminated site. The default 

exposure area first considered was a 2,000 m2 6 area, which is of sufficient size to support the 

7 assumed external exposure pathway occupancy factors for a resident gardener, as well as the 

8 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 9

exposure areas. The NRC then used the distribution of 2,000 m2 10 exposure area concentrations 

11 for comparison against an area-wide DCGL derived from dose modeling. The staff also 

12 performed sensitivity analyses that considered a range of exposure area sizes to provide 

13 additional information with respect to other types of uncertainty, such as institutional, land use, 

14 or behavioral uncertainties (e.g., uncertainty in future land use, including the size and location of 

15 parcels of land a resident receptor may reside on following license termination and 

16 redevelopment of the released site). It is important to note that, as the size of the exposure 

17 area changes, so may the exposure pathways, occupancy factors, and other exposure 

18 scenario-related parameters. For example, if one were to consider relatively small exposure 

19 areas appropriate for an apartment or condominium resident, which will result in higher 

20 maximum exposure area concentrations, then it may be (1) unreasonable to assume certain 

21 pathways are viable, such as homegrown produce ingestion, and (2) shielding factors for the 

22 apartment or condominium residence may need to be adjusted to more accurately calculate 

23 external pathway dose. In these cases, dose modeling can estimate dose for the smaller 

24 exposure areas using more representative parameters and pathways appropriate for the 

25 exposure scenario being considered, with appropriate justification. Care should also be taken to 

26 ensure that the statistics are not skewed due to the presence of large areas of unimpacted soils 

27 (e.g., only Class 1–3 areas should be included in the analysis for particularly large sites with 

28 significant portions of the site that are not radiologically affected by facility operations).

29 Evaluation of uncertainty or variability in exposure area concentrations will lead to a more 

30 comprehensive evaluation of risk to the average member of the critical group. Analysis of a 

31 range of exposure areas, including smaller exposure areas that may be appropriate for 

32 alternative exposure scenarios, will better inform the decisionmaker with respect to less likely 

33 but plausible exposure scenarios that may be more limiting. As the size of the exposure areas 

34 decreases and approaches the size of true elevated areas or hot spots, SADA can also provide 

35 valuable information on the distribution of elevated area concentrations at a site. Consideration 

36 could then be given to the number and location of elevated areas on the site to ensure that 

37 potential risk is well understood. In some cases, it may be reasonable to assume the receptor 

38 spends his time at only one elevated area or to consider some form of time or spatial weighting 

39 if the elevated areas are located a considerable distance apart and the exposure area is rather 

40 large in relation to the elevated areas. 

41
I-19

1 I.3 Criteria for Selecting and Modifying Exposure Scenarios, Pathways, and 

2 Critical Groups

3 I.3.1 Introduction

4 After the source term has been evaluated, the question becomes, “How could humans be 

5 exposed either directly or indirectly to residual radioactivity?” or “What is the appropriate 

6 exposure scenario?” Each exposure scenario should address the following questions:

7

8 (1) How does the residual radioactivity move through the environment?

9 (2) Where can humans be exposed to the environmental concentrations?

10 (3) What is the likely land use(s) in the future for these areas?

11 (4) What are the exposure group’s habits that will determine exposure? (For example, what

12 do they eat and where does it come from? How much? Where do they get water and

13 how much? How much time do they spend on various activities?)

14 The ultimate goal of dose modeling is to estimate the dose to a specific receptor. Broad 

15 generalizations of the direct or indirect interaction of the affected receptors with the residual 

16 radioactivity can be identified for ease of discussion among the licensee, regulator, public, and 

17 other interested parties. Exposure scenarios are defined as reasonable sets of activities related 

18 to the future use of the site. Therefore, exposure scenarios describe future land uses, human 

19 activities, and the behavior of the natural system.

20

21 In most situations, there are numerous possible exposure scenarios of how future human 

22 exposure groups could interact with residual radioactivity. The compliance criteria in 

23 10 CFR Part 20 for decommissioning do not require an investigation of all (or many) possible 

24 exposure scenarios; their focus is on the dose to members of the critical group. The critical 

25 group is defined (at 10 CFR 20.1003) as “...the group of individuals reasonably expected to 

26 receive the greatest exposure to residual radioactivity for any applicable set of circumstances.”

27

28 By combining knowledge about the answers to Questions 1 and 2, the licensee can develop 

29 exposure pathways. These are the routes that residual radioactivity travels through the 

30 environment, from its source, until it interacts with a human. They can be fairly simple 

31 (e.g., surface-soil residual radioactivity emits gamma radiation, which results in direct exposure 

32 to the individual standing on the soil) or they can be fairly involved (e.g., the residual 

33 radioactivity in the surface soil leaches through the unsaturated soil layers into the underlying 

34 aquifer, and the water from the aquifer is pumped out by the exposed individual for use as 

35 drinking water, which results in the exposed individual ingesting the environmental 

36 concentrations). Exposure pathways typically fall into three principal categories, identified by 

37 the manner in which the exposed individual interacts with the environmental concentrations 

38 resulting from the residual radioactivity: ingestion, inhalation, or external (i.e., direct) exposure 

39 pathways.

40

41 As required under Subpart E, the licensee evaluates the dose from residual radioactivity for the 

42 average member of the critical group, which is not necessarily the same as the maximally 

43 exposed individual. This is not a reduction in the level of protection provided to the public but an 

44 attempt to emphasize the uncertainty and assumptions needed in calculating potential future 

45 doses, while limiting boundless speculation on possible future exposure scenarios. Although it 
I-20

1 is possible to actually identify with confidence the most exposed member of the public in some 

2 operational situations (e.g., through monitoring, time studies, distance from the facility), 

3 identification of the specific individual who may receive the highest dose some time (up to 

4 1,000 years) in the future is impractical, if not impossible. Speculation on his or her habits, 

5 characteristics, age, or metabolism could be endless. The use of the “average member of the 

6 critical group” acknowledges that any hypothetical “individual” used in the performance 

7 assessment is based, in some manner, on the statistical results from data sets (e.g., the 

8 breathing rate is based on the range of possible breathing rates) gathered from groups of 

9 individuals. Although bounding assumptions could be used to select values for each of the 

10 parameters (e.g., the maximum amount of meat, milk, vegetables, possible exposure time), the 

11 result could be an extremely conservative calculation of an unrealistic exposure scenario and 

12 may lead to excessively low allowable residual radioactivity levels, compared to the actual risk.

13

14 Calculating the dose to the critical group is intended to bound the individual dose to other 

15 possible exposure groups, because the critical group is a relatively small group of individuals, 

16 who, because of their habits, actions, and characteristics, could receive among the highest 

17 potential doses at some time in the future. By using the hypothetical critical group as the dose 

18 receptor, coupled with prudently conservative models, it is highly unlikely that any individual 

19 would actually receive doses in excess of that calculated for the average member of the critical 

20 group. The licensee should base the description of a critical group’s habits, actions, and

21 characteristics on credible assumptions, and the information or data ranges used to support the 

22 assumptions should be limited in scope to reduce the possibility of adding members of less 

23 exposed groups to the critical group.

24

25 As low as is reasonably achievable (ALARA) analyses should use the dose based on the 

26 reasonably foreseeable land use for any cost-benefit calculations performed.

27

28 I.3.2 Issues in Selecting and Modifying Exposure Scenarios, Pathways, and Critical 

29 Groups

30 The definition of exposure scenarios, identification of a critical group with its associated 

31 exposure pathways, and the dose assessment based on that definition can be generic or site32 specific. Licensees might do the following:

33

34 • Use screening exposure scenarios, screening groups, and pathway parameters as

35 described in NUREG-1549 (NRC, 1998d) and the NUREG/CR-5512 series, “Residual

36 Radioactive Contamination from Decommissioning,” Volumes 3 and 4, issued

37 October 1999 (e.g., NRC, 1999c, 1999d). This can be used for either screening or site38 specific analyses.

39 • Use the default screening exposure scenarios as a starting point to develop more

40 site-specific pathway analyses or critical group habits.

41 • Develop site-specific exposure scenarios and critical groups and identify associated

42 exposure pathways from scratch.

43 To establish site-specific exposure scenarios, critical groups, or sets of exposure pathways, the 

44 licensee may need to justify its selections. For some licensees, this may require minimum 

45 amounts of site-specific data to support the assumptions inherent in the existing default 

46 screening exposure scenarios or for removing specific exposure pathways. For others, the 

47 licensee may need to thoroughly investigate and justify the appropriateness of the selected 
I-21

1 exposure scenarios or critical groups, which may include evaluation of alternative exposure 

2 scenarios or critical groups. If a licensee creates the exposure scenario and associated critical 

3 group based on site-specific conditions (e.g., at a site that is grossly different than the 

4 assumptions inherent in the default exposure scenarios), it should include documentation that 

5 provides a transparent and traceable audit trail for each of the assumptions used in developing 

6 the exposure scenario and critical group (e.g., justify the inclusion (or exclusion) of a particular 

7 exposure pathway).

8

9 I.3.3 Recommended Approaches

10 I.3.3.1 Screening Analyses

11 In the case of screening, the decisions involved in identifying the appropriate exposure scenario 

12 and critical group, with their corresponding exposure pathways, have already been made. 

13 Exposure scenario descriptions acceptable to the NRC staff for use in generic screening are 

14 developed and contained in NUREG/CR–5512, Volume 1. NUREG/CR–5512, Volume 3, and 

15 NUREG-1549 provide the rationale for applicability of the generic exposure scenarios, critical 

16 groups, and pathways at a site; the rationale and assumptions for exposure scenarios and 

17 pathways included (and excluded); and the associated parameter values or ranges 

18 (NUREG/CR-5512, Volume 3, contains detailed information on data to support development of 

19 parameter distributions). A description of the screening exposure scenarios and associated 

20 pathways is provided in Table I.2. 

21

22 I.3.3.2 Site-Specific Analyses

23 Site-specific analyses give licensees greater flexibility in developing the compliance exposure 

24 scenario. The licensee should justify its selection of this scenario based on reasonably 

25 foreseeable land use at the site. This scenario should result in an exposure to the public, such 

26 that no other exposure scenario, using reasonably foreseeable land use assumptions, will result 

27 in higher doses to its exposure group(s). The level of justification and analysis provided by the 

28 licensee will depend on how much credit is taken by adding “realism” to the exposure scenario 

29 based on site-specific information (i.e., how much credit is being taken for elimination of 

30 pathways or reduction in pathway contributions relative to more conservative assumptions made 

31 in screening exposure scenarios). As the analysis becomes more realistic, greater degrees of 

32 justification and, potentially ancillary analyses, will be required. For example, a site is currently 

33 zoned as industrial, and the local area is a mix of suburban, commercial, and industrial uses. 

34 Rural uses of the property are less likely but plausible for the foreseeable future. If it chose to 

35 use the generic screening exposure scenario, the licensee would need to provide limited 

36 justification. If the licensee proposed to use a maintenance worker exposure scenario assuming 

37 industrial land use as the compliance exposure scenario, it would need to provide quantitative 

38 analyses of, or a qualitative argument discounting the need to analyze, other competing 

39 exposure scenarios (based on industrial land use and on suburban or commercial land use) to 

40 justify the selection of the compliance exposure scenario. In addition, the licensee would need 

41 to provide analyses of the rural use of the land to show what impacts would occur from the less 

42 likely but plausible exposure scenario.

43

44 Site-specific analyses can use the generic screening exposure scenario(s) with little justification. 

45 The licensee may need to justify that the site contains neither physical features nor locations of 

46 residual radioactivity (other than those assumed in the screening analyses), which would 

47 invalidate the assumptions made in developing the exposure scenarios. If site or source 

48 features are found to be incompatible with exposure scenario assumptions, the licensee should 
I-22

1 justify why the generic exposure scenario are nonetheless appropriate for use in the dose 

2 modeling. A site can fail to meet the requirements of the conceptual model (see Section I.4 of 

3 this appendix) without invalidating the generic exposure scenario, and situations can arise 

4 where the default exposure scenario is no longer the limiting case. For example, the site may 

5 have preexisting groundwater contamination, which is counter to the assumptions in the 

6 conceptual model inherent in the screening models. However, this may not require any change 

7 in the exposure scenario, because the residential farmer scenario may still be an appropriate 

8 scenario, as it contains all of the appropriate exposure pathways, including groundwater use for 

9 drinking, irrigation, and for animals. Alternately, if the residual radioactivity were a volumetric 

10 source in the walls of a building, rather than on the building surfaces, the generic exposure 

11 scenario of an office worker may not be the exposure scenario leading to the critical group. For 

12 certain sets of radionuclides, a building renovation exposure scenario may be more limiting 

13 because of the exposure to the airborne concentration of material as the walls are disturbed.

14 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

15
I-23

1 The licensee can develop site-specific exposure scenarios, critical groups, and pathways for 

2 any situation. Some cases where changes to the default exposure scenarios or modification to 

3 exposure pathways are or are likely to be appropriate include the following:

4

5 (1) Major pathways (e.g., the groundwater pathway or agricultural pathways) associated

6 with the default screening scenarios could be eliminated, for either physical or site-use

7 reasons.

8 (2) The location of the residual radioactivity and the physical features of the site are outside

9 the major assumptions used in defining the default critical group and exposure

10 scenarios.

11 (3) Restricted use is proposed for a site.

12 The second case listed above can be ambiguous, as a number of assumptions key to the 

13 development of the DandD screening tool do not affect the exposure scenario description, and 

14 the NRC reviewer may need to evaluate whether the initial generic exposure scenario would still 

15 be appropriate for the site.

16

17 Modifying exposure scenarios or developing a site-specific critical group requires information on 

18 plausible uses of the site and demographics. Such information might include considerations of 

19 the prevailing (and future) uses of the land and physical characteristics of the site that may 

20 constrain site use. The licensee should categorize potential land uses as reasonably 

21 foreseeable, less likely but plausible, or implausible. Any land uses that similar property in the 

22 region currently has, or may have in the near future (e.g., approximately 100 years), should be 

23 characterized as reasonably foreseeable. The licensee should consider trends and area land 

24 use plans in determining the likelihood of potential land use. Land uses that are plausible, 

25 generally because similar land was historically used for similar purposes, could be characterized 

26 as less likely but plausible if found to be counter to the current trends or regional experience, 

27 (e.g., rural use of property currently in an urban setting). Implausible land uses are those that, 

28 because of physical limitations, could not occur (e.g., residential land use for an underwater plot 

29 of land). It may be necessary to evaluate several potential critical groups, based on different 

30 combinations of site-specific exposure scenarios developed from expected land use, pathways, 

31 and demographics, to determine the group receiving the highest exposure.

32

33 Depending on the resulting exposure scenarios, considerations of offsite exposure by either 

34 transport (e.g., through groundwater) or material transfer may be necessary to identify the 

35 critical group. Thus, the licensee should consider if offsite uses are reasonably foreseeable. If 

36 offsite uses are found to be reasonably foreseeable, such offsite uses should be analyzed to 

37 determine if the offsite user receives a higher dose compared to an onsite user and if offsite 

38 users should be identified as the critical group.

39

40 Similar considerations apply for restricted release. Thus, when analyzing the dose under 

41 restricted conditions, the nature of the critical group is likely to change because of these 

42 restrictions and controls. Site restrictions and institutional controls can restrict certain kinds of 

43 activities and land or water uses associated with the physical features of the site. The detailed 

44 definition of the exposure scenarios considered for restricted release need to include the impact 

45 of the control provisions on the location and behavior of the average member of the appropriate 

46 critical group.

47
I-24

1 For restricted use, licensees must also evaluate doses assuming institutional controls are no 

2 longer in effect. This evaluation should address (1) the associated degradation of engineered 

3 barriers assuming there is no active maintenance, and (2) exposure scenarios assuming a loss 

4 of institutional controls immediately following license termination (i.e., time=0 years). 

5 Section 3.5 of this volume contains additional information on modeling the performance and 

6 degradation of engineered barriers in dose modeling analyses.

7

8 The NRC license reviewer should evaluate the licensee’s justifications for its exposure 

9 scenarios using the following appropriate guidance. The guidance is characterized by the 

10 general approach used in developing the exposure scenarios, either (1) modifying existing 

11 generic exposure scenarios or (2) developing site-specific scenarios from “scratch.”

12

13 MODIFICATION OF GENERIC EXPOSURE SCENARIOS

14

15 First, the NRC license reviewer should evaluate whether the generic exposure scenario was 

16 applicable to the site before the licensee started modifying the exposure scenario, based on 

17 physical features or restrictions, and should identify the modifications and evaluate the 

18 licensee’s justification for those changes. Table I.3 lists some common exposure scenarios but 

19 is by no means comprehensive. The Sandia Letter Report, “Process for Developing Alternate 

20 Scenarios at NRC Sites Involved in D&D and License Termination,” issued January 2000 

21 (Thomas, et al., 2000)), which is included, in part, as Appendix M in this volume, provides 

22 information to assist a licensee or reviewer with respect to the modification of default exposure 

23 scenarios using site-specific information. Specific guidance on acceptable justifications for 

24 modifying the default exposure scenarios is provided below, based on different types of site25 specific information. Additionally, if the licensee’s intent is restricted release, the NRC should 

26 review the final exposure scenario for the case where restrictions are in place. Based on either 

27 site restrictions or site-specific data, the licensee’s justifications should support the elimination 

28 from the analysis of exposure scenarios and pathways. The NRC should focus the review on 

29 the most risk-significant pathways and model components.

30

31

32
I-25

1 Table I.3 Potential Exposure Scenarios for Use in Dose Assessments

• building occupancy (generic screening—based on NUREG/CR-5512)

• residential farmer (generic screening—based on NUREG/CR-5512)

• 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)

2

3 The licensee may need to evaluate whether the final modified exposure scenario is still the 

4 limiting reasonable representation of the critical group at the site. This may involve investigating 

5 exposure pathways not covered in the default exposure scenarios.

6

7 DEVELOPMENT OF ALTERNATIVE EXPOSURE SCENARIOS

8

9 In some decommissioning cases, either the location of the residual radioactivity, the physical 

10 characteristics of the site, or planned institutional restrictions may make the default exposure 

11 scenarios inappropriate. In other cases, the licensee may wish to provide a transparent and 

12 traceable development of the compliance and other exposure scenarios, starting with the 

13 potential land use and the site conditions. Development (and review) of alternative exposure 

14 scenarios may involve iterative steps to create the conceptual model of the site. For example, 

15 the licensee may (1) develop a generic list of exposure pathways, (2) develop the site 

16 conceptual model to screen the generic list, (3) aggregate or reduce the remaining exposure 

17 pathways to the major exposure pathways, and (4) reevaluate the conceptual model to verify 

18 that all the necessary processes are included.

19

20 A brief summary of the NRC–recommended pathway analysis process follows. Appendix K 

21 contains an example of exposure scenarios developed for PSR.

22

23 • Compile a list of exposure pathways applicable to any contaminated site. A number of

24 existing sources of information can be used, for example, NUREG/CR-5512, Volume 1

25 (NRC, 1992). Another source, although the guidance is more focused on offsite

26 exposures, is NUREG/CR-5453, “Background Information for the Development of a Low-
I-26

1 Level Waste Performance Assessment Methodology,” Volumes 1 and 2, issued 

2 December 1989 (NRC, 1989). 

3 • Categorize the general types of residual radioactivity at the site (e.g., sediment or soil,

4 deposits in buildings, surface residual radioactivity, surface water, groundwater, or

5 industrial products such as slag).

6 • Screen out pathways, for each contaminant type, that do not apply to the site.

7 • Identify the physical processes pertinent to the remaining pathways for the site.

8 • Separate the list of exposure pathways into unique pairs of exposure media (e.g., source

9 to groundwater, groundwater to surface water). Determine the physical processes that

10 are relevant for each exposure media pair and combine the processes with the pathway

11 links.

12 • Reassemble exposure pathways for each source type, using the exposure media pairs

13 as building blocks, thus associating all the physical processes identified with the

14 individual pairs with the complete pathway.

15 The licensee’s documentation of the decisions made about inclusion (or exclusion) of the 

16 various pathways should be transparent and traceable. An international working group 

17 established a methodology for developing models to analyze radionuclide behavior in the 

18 biosphere and associated radiological exposure pathways (i.e., the Reference Biospheres 

19 Methodology). BIOMOVS II published the methodology in its Technical Report No. 6, 

20 “Development of a Reference Biospheres Methodology for Radioactive Waste Disposal,” issued 

21 September 1996 (SSI, 1996), and included a list of international biosphere features, events, and 

processes.6 22 The report may be useful as a guide for additional information on a logical method 

23 to complete the pathway analysis above and include proper justification. Generally, the 

24 Reference Biospheres Methodology is more useful for complex sites that may have numerous 

25 physical processes that interact in such a way that a number of different exposure groups may 

26 need to be investigated to identify the critical group. Additional work has been done on 

27 implementing the Reference Biospheres Methodology by a working group of the International 

28 Atomic Energy Agency’s (IAEA’s) Biosphere Modeling and Assessment (BIOMASS) program 

29 (IAEA, 1999a, 1999b, 2001). Specifically, IAEA Working Document BIOMASS/T1/WD03, 

30 “Guidance on the Definition of Critical and Other Hypothetical Exposed Groups for Solid 

31 Radioactive Waste Disposal,” may provide additional information on developing a site-specific 

32 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, although the licensee may eliminate a residential scenario, as construction of a residence in the area of 

active erosion would not be realistic, given its proximity to surface water and uneven topography). 
I-27

1

2 I.3.3.3 Guidance on Specific Issues

3 Land Use

4 A licensee’s assumptions for land use should focus on current practice in the region. The 

5 region of concern can be as large as an 80-kilometer (km) (50-mile) radius. To narrow the focus 

6 of current land practices, the licensees can use information on how land use has been changing 

7 in the region and should give more weight to land use practices either close to the site or in 

8 similar physical settings. This can be very important for semirural sites that are being 

9 encroached upon by suburban residential development. Reviewers may wish to involve State 

10 and local land use planning agencies in discussions, if the licensee has not already requested 

11 their involvement.

12

13 Potential land uses should be categorized as reasonably foreseeable, less likely but plausible, 

14 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 15 ), should be characterized as reasonably 

16 foreseeable. Consideration should be given to trends and area land use plans in determining 

17 the likelihood of potential land use. The time frame of interest for exposure scenario 

18 development could be less than 100 years in certain cases and would depend on such factors 

19 as the rate of change in land use patterns in the area, radionuclides of interest, and the time of 

20 peak dose. For example, a site with residual cobalt-60, which has approximately a 5-year half21 life, would not likely need to explore possible land uses that may exist at the site beyond a few 

22 decades, because of the natural decay of the residual material. 

23

24 Land uses that are plausible, generally because similar land historically was used for the 

25 purpose but are counter to the current trends or regional experience should be characterized as 

26 less likely but plausible (e.g., rural use of property currently in an urban setting). Implausible 

27 land uses are those that, because of generally physical limitations, could not occur 

28 (e.g., residential land use for an underwater plot of land).

29

30 Land use justifications by licensees often rely on State or local codes, in building or well 

31 development to constrain future use. In general, licensees requesting unrestricted release 

32 should not rely solely on these factors as reasons to remove pathways or justify the exposure 

33 scenario unless (1) the radionuclides have a relatively short half-life (approximately 10 years or 

34 less) or (2) the dose from long-lived radionuclides reaches its peak before 100 years. Similarly, 

35 licensees requesting unrestricted release should not limit land use exposure scenarios based on 

36 commitments or require the enforcement of limitations by the licensee or another party (e.g., a 

37 licensee reiterates that the land will remain industrial by stating that the land will not be sold by 

38 the licensee after the license is terminated).

39

40 Licensees should base justifications of land use on (1) the nature of the land and reasonable 

41 predictions based on its physical and geologic characteristics, and (2) societal uses of the land, 

42 based on past historical information, current uses of it and similar properties, and what is 

43 reasonably foreseeable in the near future. The societal uses of the site in the future should be 

44 based on advice from local land planners and other stakeholders on what possible land uses 

45 are likely within a time period of around 100 years. The level of justification for the final land 

46 uses is inversely proportional to the level of realism assumed by the licensee. Limited 

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.
I-28

1 justification may be required for bounding analyses, while much more detailed justification, 

2 including alternative reasonably foreseeable and less likely but plausible exposure scenario 

3 analyses, may be needed for a situation with a smaller degree of conservatism in the analyses. 

4

5 Additional guidance is available on potential sources of land use information in Appendix M.

6

7 WATERBORNE EXPOSURE PATHWAYS

8

9 Removal of waterborne exposure pathways can range from global (e.g., all groundwater 

10 pathways) to specific (e.g., no drinking water but agricultural/fish pond use remains). 

11 Acceptable justifications are generally based on physical conditions at the site rather than local 

12 codes. The licensee should base its justification of water quality and quantity of the saturated 

13 zone on the classification systems used by the U.S. Environmental Protection Agency (EPA) or 

14 the State, as appropriate. Arguments involving depth to water table, or well production capacity, 

15 should have supporting documentation from either the USGS, an appropriate State agency, or 

16 an independent consultant.

17

18 NRC license reviewers should evaluate the reasons for the classification (e.g., information on 

19 water quality used as a basis for eliminating pathways). Where the aquifer is classified as not 

20 being a source of drinking water but is adequate for stock watering and irrigation, the licensee 

21 can eliminate the drinking water pathway but should still maintain the irrigation and meat/milk 

22 pathways. Aquifers may exceed certain constituents and still be able to be used for various 

23 purposes, because those constituents may easily be treatable (e.g., turbidity). In cases where 

24 the water may be treatable or because the degree of connection between the aquifer and 

25 surface water may make the use of the aquifer questionable, the reviewer should involve the 

26 EPA or the State, or both, as appropriate, in discussions on reasonable assumptions for the 

27 aquifer use.

28

29 AGRICULTURAL PATHWAYS

30

31 Agricultural pathways may be removed or modified for reasons such as (1) land use patterns, 

32 (2) poor-quality soil, (3) topography, and (4) size of contaminated area. Many justifications may

33 result in modification of the pathways, rather than complete elimination. For example, the poor

34 quality of the soil may make intensive farming activities impractical, but residential gardening

35 may still be reasonable.

36

37 Licensees using poor-quality soil as a justification for modifying the agricultural pathways should 

38 provide the reviewer with supporting documentation from the Soil Conservation Service, 

39 appropriate State or local agency, or an independent consultant. Reviewers should carefully 

40 consider whether the state of the soil would reasonably preclude all activities (e.g., because of 

41 high salinity of soil) or only certain activities. In most cases, soil quality can reasonably preclude 

42 activities such as intensive farming but could allow grazing or small gardens.

43

44 When reviewing justifications involving topography, the NRC reviewer should limit speculation of 

45 future topographical changes from civil engineering and evaluate the reasonableness of the 

46 critical group performing its activities on the current topography, for example, a slope. The 

47 licensee should provide supporting documentation in the form of pictures, USGS or similar 

48 topographic maps, hand-drawn maps, or a detailed description of how the topography would 

49 limit farming. NRC reviewers may wish to visit a site to evaluate the topography firsthand. 

50
I-29

1 AGE-DEPENDENT CRITICAL GROUPS

2

3 The definitions in 10 CFR Part 20 should be used when demonstrating compliance with the 

4 requirements of Subpart E. EPA’s Federal Guidance Report No. 11, “Limiting Values of 

5 Radionuclide Intake and Air Concentration and Dose Conversion Factors for Inhalation, 

6 Submersion, and Ingestion,” issued September 1988, should be used when calculating internal 

7 exposures by using the intake-to-dose conversion factors, which are based primarily on adults. 

8 EPA’s “Federal Radiation Protection Draft Guidance for Exposure of the General Public,” 

9 Volume 59, dated December 23, 1994 indicates that implementing age and sex dependent 

10 limits for the general public is difficult and due to uncertainty in the impact of these factors on 

11 dose that detailed consideration of age and sex is generally unnecessary. 

12

13 Since age-based dose conversion factors are not being used, the same dose conversion factors 

14 are applied to all individuals. Only in rare exposure scenarios will a nonadult individual receive 

15 a higher dose (i.e., intake more radioactive material) than an adult individual in a similar 

16 exposure scenario. One example may be related to the milk pathway. Children generally drink 

17 more milk annually than adults. If milk were the only pathway that would expose the individual 

18 to a dose, then the child would have a slightly higher dose than the adult. But in most 

19 situations, especially ones involving multiple pathways, the total intake of the adult is greater

20 than that of a child. Therefore, for most multiple pathway exposure scenarios, such as 

21 screening analyses, the average member of the critical group should usually be assumed to be 

22 an adult, with the proper habits and characteristics of an adult. As the licensee eliminates 

23 pathways or modifies the exposure scenario, the behavior and dietary habits of children may 

24 become important. In such cases, the licensees should contact the NRC staff for guidance.

25

26 IMPACT OF AREA ON EXPOSURE PATHWAYS AND SCENARIOS

27

28 As discussed above, the default exposure scenario for surface soil assumes large areas of 

29 homogeneous surface residual radioactivity. If the area of residual radioactivity is smaller than 

30 that needed to support the exposure pathway (e.g., an insufficient area to support the 

31 production of the quantity of contaminated crops assumed to be consumed by the receptor), the 

32 licensee may propose modifying the exposure pathways to account for the effect area has on 

33 the critical group’s activities. The licensee can use different approaches to account for smaller 

34 areas of elevated activity, including the following two methods:

35

36 (1) Reduce the calculated dose by modifying the exposure time or usage parameters

37 accordingly.

38 (2) Modify the exposure scenario and pathways and/or modify the calculation method to

39 account for the size of the residual radioactivity.

40 These methods may be built into some dose assessment codes for surficial soil (e.g., RESRAD41 ONSITE), but the user should understand how the codes consider the area of contamination in 

42 adjusting pathway doses. For example, RESRAD-ONSITE and RESRAD-BUILD are commonly 

43 used to perform site-specific dose modeling and to calculate DCGLs. These codes uniquely 

44 consider the size of the area of the contaminated zone in calculating the dose for each pathway. 

45 It should be noted, however, that RESRAD-ONSITE and RESRAD BUILD do not adjust the 

46 occupancy factors for smaller areas of contamination. As the size of the elevated areas

47 decrease, full occupancy assumptions may become increasingly unrealistic and overly 

48 conservative (e.g., by default, RESRAD-ONSITE assumes that the receptor is located in the 
I-30

1 center of the contaminated zone). In these cases, the licensee may present arguments for 

2 modified occupancy assumptions or alternative exposure scenario assumptions for smaller 

3 areas of residual radioactivity, based on the expected habits and characteristics of members of 

4 critical groups, as well as the characteristics of the site being evaluated. Certain exposure 

5 pathways, such as meat and plant ingestion, are also affected by the size of the elevated area. 

6 Therefore, the licensee can present arguments for why the dose from certain pathways may be 

7 limited due to the area of the elevated concentration and adjust other parameters, as necessary, 

8 to avoid overly conservative dose calculations and associated overly conservative cleanup 

9 levels. For example, if the contaminated plant fraction is set to -1, RESRAD-ONSITE adjusts 

10 the total amount of contaminated plant products ingested based on the size of the contaminated 

11 area. If other dose modeling codes are used, the licensee may justify using lower consumption 

12 rates of contaminated homegrown produce or animal products, based on the size of the 

13 elevated area.

14 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 15 These factors are (1) provided in 

16 lookup tables included in various regulatory documents, (2) calculated using the formula 

17 provided for field analysis in the “User’s Manual for RESRAD Version 6,” and (3) calculated 

18 using RESRAD Version 6.5. Differences between area factor values listed for a specific 

19 radionuclide can be attributed to a variety of factors, including the use of different models or 

20 methods, modeling parameter values (e.g., length parallel to aquifer flow in the RESRAD 

21 nondispersion model), and modeling assumptions. The EPA Soil Screening Guidance (EPA, 

22 1996) and RESRAD Field Formula methods are generally more limiting for radionuclides where 

23 the dose is attributable primarily from ingestion pathways and comparable to other methods for 

24 radionuclides where the dose is attributable primarily to the external exposure pathway.

25

26 Some general observations related to cleanup levels for smaller areas and general 

27 characteristics of area factors include the following:

28

29 • The DCGLEMC value will always be greater than or equal to the DCGLW value (i.e., area

30 factors will always be larger than 1). In other words, higher concentrations or cleanup

31 levels are allowable for smaller elevated areas compared to larger areas of residual

32 radioactivity to meet the same radiological criteria for license termination.

33 • With respect to water-independent pathways:

34 o RESRAD-calculated area factors for gamma-emitting radionuclides where the risk is

35 dominated by the external exposure pathway generally have more restrictive or lower

36 area factors that vary nonlinearly with the area of contamination (see for example

37 cesium (Cs)-137 and Co-60 in Table I.4).

38 o RESRAD-calculated area factors for the inhalation exposure pathway also vary

39 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.
I-31

1 o RESRAD-calculated area factors for ingestion pathways such as soil ingestion,

2 animal product ingestion, and plant ingestion generally have area factors that scale

3 directly to the size of the contaminated area.

4 • With respect to water-dependent pathways:

5 o RESRAD-calculated area factors for the drinking water and fish ingestion pathways

6 generally have more limiting area factors compared to other water-dependent

7 pathways and can be more limiting than water-independent pathways.

8 o RESRAD-calculated area factors for ingestion pathway doses incurred from use of

9 contaminated irrigation water are generally lower and DCGLs are generally higher or

10 less restrictive compared to other water dependent and water-independent

11 pathways, and therefore, area factors for these pathways are generally not limiting

12 and are not as risk significant.

13 The NRC staff should review the following information provided by the licensee:

14

15 • summary table or list of the DCGLW(s) for each radionuclide and impacted medium of

16 concern

• the DCGLEMC for each radionuclide and medium of concern, if Class 1 survey units9 17 are

18 present

19 • the appropriate DCGLW for the survey method to be used if multiple radionuclides are

20 present

21 The NRC should review licensee-calculated DCGLEMCs to ensure that the values are developed 

22 based on dose modeling (i.e., default area factors found in the literature should not be used to 

23 assign DCGLEMCs). Consideration can be given to site-specific conditions, including the 

24 contributions of individual exposure pathways to the overall dose, since exposure scenarios and 

25 pathways can vary from one site to another. Additionally, the NRC should carefully review the 

26 licensee’s approach for considering multiple radionuclides and elevated areas.

27

28

9 Class 1 survey units are impacted areas that are expected to have concentrations of residual radioactivity that 

exceed the DCGLW.
I-32

1 Table I.4 Comparison of Area Factor Values from Different References

2

10000 m2 1000 m2 100 m2 10 m2 1 m2

Am-241

NUREG-1505 1 1.01 1.86 13.4 109

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

Cs-137

NUREG-1505 1 1.1 1.41 2.41 11

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

Co-60

NUREG-1505 1 1.1 1.23 2.12 9.81

NUREG-1575 1 1.1 1.2 2.1 9.8

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

Th-232

NUREG-1505 1 1.03 1.75 3.12 12.3

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

U-238

NUREG-1505 1 1.04 2.27 11.1 30.5

NUREG-1575 1 1.3 6.7 11.1 30.6

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

3 Notes:

4 • NUREG-1505 (NRC 1998b), Table 8.1, reports radionuclide-specific area factors 

5 calculated using RESRAD-ONSITE, Version 5.7.

6 • NUREG-1575 (NRC 2000a), Table 5.6, reports radionuclide-specific area factors 

7 calculated using RESRAD-ONSITE, Version 5.6.

8 • Area factors associated with the EPA Soil Screening Guidance (EPA, 1996) are the 

9 reciprocals of the area correction factors reported in Table 5.2 of the document. The 

10 area correction factors are based on the external exposure pathway only.

11 As indicated above, when the extent of residual radioactivity becomes smaller, some of the 

12 activities are no longer viable as reasonable assumptions for exposure. Generally, the first 

13 pathways affected are animal husbandry activities, because of the larger area needed for 

14 grazing and growing fodder. As a general rule, as the area gets smaller, the more the exposure 

15 scenario transforms into a residential gardener exposure scenario, so long as the initial residual 

16 radioactivity begins in the surface soil. For cases where the residual radioactivity is not in the 
I-33

1 surficial soil, the original area of residual radioactivity may not be as important in exposure 

2 scenario development, because some of the primary transport mechanisms result in 

3 redistribution of the radionuclides over larger areas (i.e., groundwater used as irrigation).

4 One common mistake in licensee submittals is that DCGLEMCs are typically not provided for 

5 residual radioactivity on building surfaces. When the screening DCGLW values were published 

6 in the Federal Register (see Appendix H), associated DCGLEMCs were not published. Although 

7 newer versions of DandD (Version 2) allow specification of a limited area of contamination, 

8 DandD adjusts dose based on area in a simplistic manner. Therefore, the licensee may wish to 

9 calculate DCGLEMCs for building surfaces using the RESRAD–BUILD computer code. 

10 OFFSITE EXPOSURE SCENARIOS

11 As discussed above, in rare situations, the exposure scenario resulting in the highest exposures 

12 from the residual radioactivity will be an offsite use exposure scenario. For these evaluations, 

13 the dose limits in 10 CFR Part 20, Subpart E, remain applicable, even though the situation may 

14 seem similar to the clearance of materials before license termination. In these scenarios, the 

15 exposure to the radioactive material will occur, because it has been removed from the current 

16 location, and this results in either new or enhanced exposure pathways. For example, a site 

17 has poor groundwater characteristics (thereby, allowing the licensee to remove the groundwater 

18 pathway from any applicable exposure scenarios), and the reasonably foreseeable land use is 

19 either commercial or industrial. The primary contaminant is technetium (Tc)-99, which primarily 

20 results in dose through either the groundwater or vegetable pathways, both of which are not 

21 applicable to the physical characteristics of the site or land use assumptions. The residual 

22 radioactivity is present in the site’s topsoil. A possible offsite exposure scenario is where, during 

23 construction of any commercial interest on the site after license termination, the removed topsoil 

24 is sold for use in a residential setting. In this case, it is likely that the topsoil with residual 

25 radioactivity will be unintentionally mixed with other topsoil at the offsite location. Licensees can 

26 use generic analyses to screen the importance of offsite uses with such sources as 

27 NUREG-1640, “Radiological Assessments for Clearance of Materials from Nuclear Facilities.” 

28 (NRC 2003b)

29 Even if offsite use is not considered reasonably foreseeable, offsite exposure scenarios may be 

30 less likely but plausible scenarios and should be analyzed as exposure scenarios, to understand 

31 the robustness of the analysis.

32 DETERMINING THE COMPLIANCE EXPOSURE SCENARIO

33 In many situations, a licensee will be faced with selecting a compliance exposure scenario from 

34 a potentially large suite of exposure scenarios and exposure groups. The licensee is expected 

35 to base its demonstration of compliance on the reasonably foreseeable exposure scenario 

36 resulting in the highest peak dose during the compliance period, consistent with the definition of 

37 the critical group. Licensees may find it advantageous to use an iterative approach to screen all 

38 the potential exposure scenarios. This will allow the licensees to focus their more detailed 

39 analyses on the important exposure scenarios. Licensees may be able to use information from 

40 NUREG/CR-5512 (NRC 1992), NUREG-1640 (NRC 2003b), and NUREG-1717 (NRC 2001), as 

41 well as other licensees’ analyses to screen their potential exposure scenarios with quantitative 

42 methods. Licensees also may be able to provide qualitative arguments to demonstrate that the 

43 dose from certain exposure scenarios is bounded by the dose of higher level exposure 

44 scenarios (e.g., a residential gardening exposure scenario will bound the dose for the residential 
I-34

1 nongardening exposure scenario). The licensees should provide justifications on the basis, 

2 method, and results of their exposure scenario screening in their DP.

3 Even after screening the exposure scenarios, a licensee will likely be left with a few exposure 

4 scenarios that may require detailed analyses to determine which will result in the critical group. 

5 For licensees with multiple radionuclides, determining the compliance exposure scenario

6 commonly depends on the final mixture of radionuclides. This can provide a dilemma for 

7 licensees creating DCGLs. The licensee must show that the final concentrations at the site 

8 meet the dose criteria of 10 CFR Part 20, Subpart E. Two possible approaches that the 

9 licensee may use to show compliance are, but are not limited to, the following:

10 (1) Use the most limiting DCGL for each radionuclide, regardless of the exposure scenario,

11 and use the sum of fractions, ignoring the exposure scenario basis for each DCGL. This

12 approach requires limited justification. It will always either estimate the same dose as

13 the individual exposure scenarios or overestimate the dose. Generally, it will greatly

14 overestimate the dose for the individual exposure scenarios.

15 (2) Commit to demonstrating the final dose for each of the important exposure scenarios in

16 the FSS reports. This approach will require the licensee to establish operational DCGLs

17 to fully use MARSSIM (see Section 2.5).

18 The licensee needs to provide either a quantitative analysis of, or a qualitative argument 

19 discounting the need to analyze, all the exposure scenarios generated from the less likely but 

20 plausible land uses. The results of these analyses will be used by the staff to evaluate the 

21 degree of sensitivity of dose to overall exposure scenario assumptions (and the associated 

22 parameter assumptions). Analyses of less likely but plausible exposure scenarios are not 

23 meant to be ‘worst-case’ analyses and should not use a set of ‘worst-case’ parameters. 

24 Selection of parameters for less likely but plausible exposure scenarios should be consistent 

25 with the guidance in this appendix. The reviewer will consider both the magnitude and time of 

26 the peak dose from these exposure scenarios. If the peak dose from the less likely but 

27 plausible land use exposure scenarios is significant, the licensee would need to provide greater 

28 assurance that the exposure scenario is unlikely to occur, especially during the period of peak 

29 dose. The licensee may be able to show that the compliance exposure scenario bounds the 

30 results of all or many of the exposure scenarios associated with the less likely but plausible land 

31 uses.

32 I.3.4 Generic Examples

33 The following examples describe situations where the default pathways may be removed or 

34 modified. Note that the examples assume that an adequate level of justification has been 

35 provided by the licensee.

36

37 I.3.4.1 Removal of Groundwater Pathways

38 A licensee has extensive contamination of the upper soil horizon and the upper aquifer, which is 

39 unconsolidated, and it wishes to remove the groundwater pathway because the upper aquifer 

40 would not be used as a water source. The aquifer shows relatively high levels of microbial 

41 activity, turbidity, and nitrates. In addition, adjacent to the site is a small patch of wetlands that 

42 shows a great deal of communication with the upper aquifer. The potential yield rate of the 

43 upper aquifer is sufficient for domestic use, but there is a better quality, confined aquifer, with a 

44 horizon at a depth of approximately 30 meters (100 feet). The licensee has also demonstrated 
I-35

1 that the deeper aquifer will not become contaminated from the upper aquifer. Considering all of 

2 these reasons in combination, it is questionable whether the upper aquifer would actually be 

3 used. Although it may be possible for someone to treat the contaminants and use the aquifer, 

4 there are better sources of water easily available. After consultation with the EPA and the 

5 State, it is agreed that it would be unreasonable to assume someone would use the upper 

6 aquifer as a water source. Therefore, the licensee is allowed to remove the groundwater 

7 pathway from the exposure scenario.

8

9 I.3.4.2 Exposure Scenario Development for Buried Residual Radioactivity

10 EXAMPLE 1: SUBSURFACE SOIL

11 A site has residual radioactivity buried at a few feet below the surface and the licensee is 

12 requesting unrestricted release. The residual radioactivity does not have enough highly 

13 energetic gamma emitters to result in an external dose in the current configuration. Two 

14 scenarios can be developed (without any other site-specific information): (1) leaching of the 

15 radionuclides to the groundwater, which is then used by a residential farmer, and (2) exposure 

16 to the buried residual radioactivity by house construction for a resident farmer with the displaced 

17 soil, which includes part of the residual radioactivity, spread across the surface. Exposure 

18 scenario 2 encompasses all the exposure pathways and, although not all of the source term is 

19 in the original position, leaching may occur both from the remaining buried residual radioactivity 

20 and the surface soil. In certain cases, alternative intrusion events could be more limiting and 

21 should be considered particularly if home construction into the buried residual radioactivity is 

22 precluded due to the depth of the residual radioactivity or presence of a cover. Appendix J of 

23 this volume describes in greater detail the consideration of residual radioactivity in subsurface 

24 soils.

25

26 EXAMPLE 2: EMBEDDED PIPING

27 At another site, the licensee is requesting unrestricted release of its site. It is removing the 

28 buildings but is evaluating the need to remove the concrete pads, which have embedded piping 

29 that contains the residual radioactivity. Two exposure scenarios can be reasonably envisioned. 

30 The first involves a resident farmer onsite, who builds a house on the concrete pad, without 

31 disturbing the embedded piping. Possible exposure pathways would be external dose from the 

32 piping and exposure to leached materials from the piping through groundwater use 

33 (e.g., drinking, irrigation). The second exposure scenario is similar to the building renovation 

34 exposure scenario discussed in Example 1, where the concrete pad and piping are removed 

35 from the site during a future construction project. The licensee should investigate both 

36 exposure scenarios to find the limiting scenario.

37

38 I.3.4.3 Exposure Scenario Development for Restricted Release

39 For this example, the site restrictions planned for an alternate site include a restriction on the 

40 deed that the property can be used only for parkland, and an engineered cover is placed over 

41 the residual radioactivity. The engineered cover is contoured for use as parkland and has a 

42 vegetative cover (i.e., not a mound covered in rip-rap). Three exposure scenarios are easily 

43 envisioned for the restricted release analysis. The first is recreational use of the property as a 

44 city park or golf course, which would limit exposure scenarios to possible external exposure. 

45 The second would involve offsite use of groundwater that contains radionuclides leached from 

46 the buried residual radioactivity. The default offsite user would be a resident farmer using the 
I-36

1 groundwater for all water needs. The third exposure scenario would be a worker maintaining 

2 the park.

3

4 The doses assuming the loss of institutional control (i.e., the deed restriction) immediately 

5 following license termination (or time=0 years) and degradation of the engineered cover also

6 must be evaluated. Scenarios should consider how critical groups could be exposed to the 

7 residual radioactivity through disruptive events.

8

9 Consider a residential farmer who uses groundwater from the aquifer located under the site. An 

10 engineered cover may become compromised by the placement of buildings. The cover may still 

11 perform in some degraded function for some period of time. Whether buried residual 

12 radioactivity is transported to the surface by the construction of a basement under the resident 

13 farmer’s house would depend on the thickness of the engineered cover. If typical basement 

14 depths are deeper than the engineered cover’s thickness, some portion of residual radioactivity 

15 would be transported to the surface, mixed with the “clean” cover material, and spread over the 

16 site. If the typical basement depth is shallower than the engineered cover thickness, other 

17 disruptive events such as well construction or large-scale excavation of material may need to be 

18 considered to evaluate a scenario where residual radioactivity is brought to the surface (see 

19 Appendix J).

20

21 In another scenario, the engineered cover may become degraded from erosion and residual 

22 radioactivity redistributed through hydrological processes. The reasonableness of this scenario 

23 would depend on the thickness and erosion-resistance of the engineered cover (see Section 3.5 

24 for additional information on performance assessments for engineered barriers).

25

26 I.4 Criteria to Establish Conceptual and Mathematical Models

27 I.4.1 Introduction

28 Analyzing the release and migration of radionuclides through the natural environment and

29 engineered systems, at a specific site, requires the licensee to interpret the nature and features 

30 of the site so that the site can be represented by mathematical equations (i.e., mathematical 

31 models). This simplified representation of the site is commonly referred to as the conceptual 

32 model of the site.

33

34 Figure I.3 depicts the process of conceptual model development. In dose assessments, 

35 developing a conceptual model involves making an abstraction of site data into a form that is 

36 capable of being modeled. This development should generally involve making simplifying 

37 assumptions, including simplification of the appropriate governing equations, to reflect the 

38 physical setting. These simplifying assumptions are usually made in describing the geometry of 

39 the system, the spatial and temporal variability of parameters, the isotropy of the system, and 

40 the influence of the surrounding environment. The conceptual model should provide an 

41 illustration or description of site conditions, to show, or explain, contaminant distributions, 

42 release mechanisms, exposure pathways and migration routes, and potential receptors. In 

43 other words, the conceptual model should explain or illustrate how radionuclides enter, move 

44 through and/or are retained in, and leave the environment.

45
I-37

1

2

3

4

5

6

7

8

9

10

11

12

13

14

15

16

17

18

19

20 Figure I.3 Conceptual Model Development

21 As shown in Figure I.4, developing a conceptual model at a site is Step 3 of the 

22 decommissioning decision framework (see Figure 1.2 of this volume). Conceptual model 

23 development follows after assimilation of site data (Step 1) and definition of scenarios (Step 2), 

24 because information from these two steps feeds into its development. In other words, the 

25 conceptual model should be based on what is known about the site from data and information 

26 gathered as part of Step 1, and how the site evolves during the period covered by the analysis, 

27 based on the assumed land use defined under Step 2.

28

29 Mathematical models are a quantitative representation of the conceptual model. Because the 

30 conceptual model provides the linkage between site conditions and features (Steps 1 and 2) 

31 and the computer code(s) (with its associated mathematical models) used in the dose analysis 

32 (Step 4 of the decommissioning framework), it is a key step in a dose assessment and should 

33 not be taken lightly.

34

Conceptual Model

Mental Construct

Site Data Scenario

Mathematical

Models

Engineering

Designs

Computer code
I-38

1

2 Figure I. 4 Decommissioning Decision Framework 

3 I.4.2 Technical Issues

4 Uncertainties in conceptual models can be large and possibly even larger than uncertainties in 

5 parameters used in the analysis (James and Oldenburg, “Linear and Monte Carlo Uncertainty 

6 Analysis for Subsurface Contaminant Transport Simulation,” issued November 1997). Thus, 

7 conceptual model uncertainties can be a significant source of uncertainty in the overall dose 

8 assessment. Uncertainties in the conceptual model(s) are generally caused by incomplete 

9 knowledge about the natural system being analyzed and differing views about how to interpret 

10 data representing the system.

11

12 Development of conceptual models is a subjective process based on the interpretation of limited 

13 (or in most cases, sparse) site data. From these limited data, the licensee should determine the 

14 key processes and features at the site and how they are likely to affect the movement of 

15 radionuclides through the environment. Because the conceptual model of the site is based on 

16 incomplete information, it is possible that multiple interpretations of the same data can be 

17 derived. A licensee should also determine the appropriate level of simplification acceptable for 

18 representing the site. An overly simplified conceptual model may leave out key site features or 

19 conditions that are important in estimating where radionuclides are likely to be transported (thus, 

20 where people might be exposed) and when they might get there (thus, the radionuclide 

21 concentration when it arrives). On the other hand, an overly complex conceptual model may 

22 introduce unnecessary uncertainty and costs into the analyses. As a broad example, simple 
I-39

1 models contained in screening codes may oversimplify features and processes at a specific site. 

2 The licensee also should ensure that the model provides the appropriate level of detail. It is 

3 important that the conceptual model have sufficient detail and scope for a license reviewer to be 

4 able to assess the appropriateness of the computer codes used in the analysis and the 

5 defensibility of the assumptions made. In summary, key issues in developing and presenting 

6 the conceptual model are (1) identifying the important site features and processes that need to 

7 be included in the conceptual model, (2) deciding among possible competing interpretations of 

8 the site data, and (3) determining the level of detail needed to describe those features and 

9 processes.

10

11 Some important staff insights gained from reviewing decommissioning dose assessments and 

12 performance assessments are summarized below. Insights are grouped into those related to 

13 (1) model abstraction and (2) model simplification. Issues associated with model abstraction

14 and simplification are generally important for more complex sites, while screening models and

15 codes discussed in the next section I.4.3 may be appropriate for simpler sites provided the

16 underlying assumptions of the conceptual models for those models and codes are consistent

17 with site conditions.

18

19 I.4.2.1 Model Abstraction

20 Some level of abstraction is required to translate the concepts of a conceptual model into 

21 mathematical terms. An abstracted model can be something as simple as a data value or 

22 lookup table. The methods to produce abstracted models are often not particularly rigorous and 

23 can introduce quantified uncertainties and biases. Important aspects can be lost in the 

24 abstraction process, and, for this reason, the reduction is often undertaken in such a way that it 

25 produces a conservative result. It is therefore important to clearly document the model 

26 abstraction and recognize the potential impacts due to the abstraction.

27

28 Insights gained on abstracting models of hydrogeological systems include the following:

29

30 • Very complicated sites may not need to have all geological and hydrogeological features

31 and processes represented in the model, so that model abstraction may be less

32 problematic than originally estimated.

33 • Many of the processes that govern transport of radionuclides in the unsaturated zone

34 are essentially the same as those that govern transport in the saturated zone. However,

35 effective hydraulic properties have a nonlinear dependence on soil moisture content, so

36 that flow in the unsaturated zone can be strongly influenced by extreme, but not

37 necessarily uncommon, conditions.

38 Additional important insights gained from the NRC review of examples of model abstraction 

39 include the following:

40

41 • Code selection for a particular modeling exercise should be judicious to ensure that code

42 limitations do not lead to nonconservative or unrealistic dose modeling predictions

43 (e.g., lack of consideration of complex source terms, important transport processes such

44 as diffusion and dispersion, or complex flow systems).

45 • Model abstractions can successfully represent the essential elements of the system

46 being simulated, increase computational efficiency, enable a more complete evaluation
I-40

1 of model and parameter uncertainty, and lead to a better understanding of the system 

2 being simulated.

3 • Additional complexity can and should be added to a model if certain processes or

4 parameters are found to be important (i.e., if the additional complexity significantly

5 influences the results).

6 • Important pathways of exposure and scenarios should be evaluated to ensure that the

7 most limiting pathways and scenarios are considered. In some cases, it is not intuitive to

8 determine the most risk-significant exposure scenario.

9 • Source term assumptions may have a significant impact on modeling results

10 (e.g., source orientation and geometry, source distribution, and elevation of release).

11 I.4.2.2 Model Simplification

12 Model simplification is the process for reducing the complexity of a numerical model into a 

13 simpler numerical model while still maintaining the validity of the simulation results. 

14 NUREG/CR-6884, “Model Abstraction Techniques for Soil-Water Flow and Transport,” issued 

15 December 2006 (NRC 2006), presented a systematic and objective approach to model 

16 simplification relevant to subsurface flow and transport modeling. This approach included 

17 (1) justifying the need for the model simplification, (2) reviewing the context of the modeling

18 problem, (3) selecting applicable model simplification techniques, (4) determining model

19 simplification directions, and (5) simplifying the complex model in each direction. When

20 performing a model simplification, various categories of techniques are relevant to the

21 subsurface flow and transport modeling. These categories include selecting from a predefined

22 hierarchy of models, changes in spatial dimensionality (e.g., three dimensional to two

23 dimensional), and scale change, including upscaling, aggregation, and metamodeling.

24

25 The model simplification process starts with an existing complex model that can be calibrated 

26 and used in simulations. Justifying the need for the model simplification is usually associated 

27 with unsatisfactory results from the complex model in some way. This can either be related to 

28 the complex model being too expensive, too large, and too difficult to run and calibrate; input is 

29 too hard to obtain; or output is difficult to understand. After the model simplification process, the 

30 resulting model output should provide information that is both necessary and sufficient to make

31 a decision on the issue(s) of interest.

32 Some important insights gained from the NRC review of previous examples of model 

33 simplifications include the following:

34 • While in many cases, simpler models and codes developed to demonstrate compliance

35 with regulatory criteria are purposefully constructed to err on the side of higher dose, the

36 use of simpler models and codes does not guarantee conservative results. Irrespective

37 of the level of complexity of the model used to facilitate decision-making, the analyst

38 should ensure that the model is adequately supported and that the impact of model

39 simplification is well understood.

40 • The necessary level of complexity of a model is dependent on many factors, including

41 the complexity of the site, uncertainty in site parameters and processes, the questions

42 that are being addressed by the models, and the safety margin.
I-41

1 • With respect to physical dispersion, models that consider dispersion can generally 

2 provide more realistic estimates of concentrations and dose compared to simpler models 

3 that do not consider dispersion. For example, more complex three-dimensional 

4 groundwater flow and transport models can explicitly consider longitudinal and 

5 transverse dispersion, while one-dimensional models may consider no dispersion at all 

6 or longitudinal dispersion, or only implicitly consider dispersion. Consideration of 

7 physical dispersion can lead to higher or lower concentrations (and the resultant dose), 

8 or to no significant impact at all, depending on such factors as the radionuclides driving 

9 the results, the presence of multiple sources, receptor locations, numerical model 

10 construction decisions, and site complexity (e.g., heterogeneity). Because, in many 

11 cases, dispersion assumptions can significantly affect the results, the treatment of 

12 dispersion in a model should be adequately studied to ensure that concentrations and 

13 dose are not underestimated.

14 • The essential elements of a complex, three-dimensional groundwater flow and transport 

15 model can be identified and considered in a simpler flow and transport model. However, 

16 data of sufficient quality needed to calibrate the complex model must be available and a 

17 sufficient number of intermediate outputs and observations extracted to ensure that the 

18 simpler model accounts for risk-significant processes. If this approach is used, the 

19 model abstraction can sometimes be helpful in identifying and communicating key 

20 parameters and processes that are difficult to identify or explain using the complex 

21 model alone.

22 For complex sites, when relatively simple models are used to demonstrate compliance with 

23 regulatory criteria, the analyst should provide sufficient information (e.g., relatively more and 

24 less complex model comparisons) to show that the model simplifications do not lead to a 

25 significant loss of fidelity in the results that could be important to decision-making. Appendix F 

26 contains additional information about conceptual site model development, model abstraction 

27 and simplification pertaining to hydrological models.

28

29 I.4.3 Recommended Approach

30 I.4.3.1 Screening

31 An acceptable dose assessment analysis need not incorporate all the physical, chemical, and 

32 biological processes at the site. The scope of the analysis, and accordingly, the level of 

33 sophistication of the conceptual model, should be based on the overall objective of the analysis. 

34 A performance assessment conceptual model can be simple, if it still provides satisfactory 

35 confidence in site performance. For an initial screening analysis, little may be known about the 

36 site from which to develop a conceptual model. Computer codes used for screening analyses 

37 are generally intended to provide a generic and conservative representation of processes and 

38 conditions expected for a wide array of sites. Accordingly, the generic conceptual model in such 

39 codes may not provide a close representation of conditions and processes at a specific site. 

40 Such a generic representation is still acceptable, as long as it provides a conservative 

41 assessment of the performance of the site. 

42 In general, the conceptual models within DandD are expected to provide a conservative 

43 representation of site features and conditions. Therefore, for screening analyses, the NRC staff 

44 should consider such generic conceptual models to be acceptable, provided it is acceptable to 

45 assume that the initial radioactivity is contained in the top layer (building surface or soil) and the 

46 remainder of the unsaturated zone and groundwater are initially free of residual radioactivity. In 
I-42

1 using DandD for site-specific analyses, it is important to ensure that a more realistic 

2 representation of the site that is consistent with what is known about the site would not lead to 

3 higher doses. Table I.5 lists some site features and conditions that may be incompatible with 

4 the generic conceptual models within DandD. The relative importance of the incompatibilities 

5 varies with the exposure scenario and radionuclides involved. More information on the 

6 assumptions of the model is available in the development documentation (e.g., the 

7 NUREG/CR-5512 series).

8 For any site where it is known that one or more of these conditions or features are present, the 

9 licensee should provide an appropriate rationale on why the use of the DandD should not result 

10 in an underestimation of potential doses at the specific site. 

11 As an example, DandD is inappropriate for analyzing sites that contain hydrogen H-3 and C-14

12 in soil, because DandD considers only the inhalation dose from particulates in the air and does 

13 not consider the loss of H-3 and C-14 from the soil to the air as a gas or vapor. To adjust 

14 results from the DandD resident farmer exposure scenario for analyzing sites that contain either 

15 H-3 or C-14 (Haaker, “Upper bound for inhalation dose from carbon-14 vapor and tritium vapor,”

16 published June 1999) (Haaker, 1999), (1) determine the area of the contaminated zone, (2) run

17 DandD for the site with only H-3 or C-14, (3) read the associated activity ratio factor for the

18 given area from Figure I.5, and (4) estimate the potential missed dose by multiplying the

19 inhalation dose calculated from DandD by the activity ratio factor to account for the dose

20 associated with the gas or vapor phase.

21
I-43

1 Table I.5 Site Features and Conditions that may be Incompatible with those Assumed in 

2 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 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

3

4
I-44

1

2

3

4

5

6

7

8

9

10

11

12

13

14

15 Figure I.5 Activity Ratio of Vapor to Particulate as a Function of Contaminated Area 

16 (Used to Account for Missed H-3 and C-14 Inhalation Dose from Vapor Phase 

17 in DandD Residential Exposure Scenario)

18 I.4.3.2 Site-Specific Analyses

19 For site-specific analyses, the intent is to provide a more realistic assessment of doses based 

20 on more site-specific information and data. Presumably for such analyses, more is known about 

21 the site from which to develop a conceptual model. For site-specific analyses, the licensee 

22 should provide a schematic or verbal description of the problem that it is attempting to analyze. 

23 Even when using a computer code that has a predefined conceptual model, it is important for 

24 the licensee to identify any site features or conditions that may differ from those assumed in the 

25 code. In developing a site-specific conceptual model or identifying potential limitations with a 

26 predefined conceptual model, the licensee should consider the issues listed in Table I.6.

27

28 Because conceptual models are developed based on limited data, in most cases, more than 

29 one possible interpretation of the site can be justified based on the existing data. The licensee 

30 should address this uncertainty by developing multiple alternative conceptual models and 

31 proceeding with the conceptual model(s) that provides the most conservative estimate of the 

32 dose and yet is consistent with the available data. Consideration of unrealistic and highly 

33 speculative conceptual models should be avoided. Consistent with the overall dose modeling 

34 framework of starting with simple analyses and progressing to more complex modeling, as 

1E+03 

1E+04 

1E+05 

1E+06 

Activity Ratio

10 100 1000 10000 100000 1000000 

Contaminated area (m2)

Activity Ratio of Vapor to Particulate
I-45

1 warranted, it may be advisable for the analyst to begin with a simple, conservative analysis that 

2 incorporates the key site features and processes and progress to more complexity only as 

3 merited by site data. It is important to stress that a simple representation of the site, in itself, 

4 does not mean that the analysis is conservative. It is incumbent on the licensee to demonstrate 

5 that its simplification is justified, based on what is known about the site and the likelihood that 

6 alternative representations of the site would not lead to higher calculated doses.

7 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

8

9 In general, there are two primary areas of the dose analysis where the conceptual model is 

10 expected to change from one site to another; these are related to the source term and 

11 environmental transport. Aspects of the analysis related to the exposure pathways in the 

12 biosphere and dosimetry are largely determined by the exposure scenario and the assumed 

13 behavior of the critical group. Accordingly, models related to the exposure pathways in the 

14 biosphere and dosimetry should not change from one site to another, unless there is a 

15 significant change in the exposure scenario and associated critical group. The principal 

16 environmental transport pathways that should have to be considered in a dose assessment are

17 groundwater (including transport through the unsaturated zone), surface water, and air.

18 The conceptual model of the source area should describe the contaminants and how they are 

19 likely to be released into the environment. Specifically, it should describe key features and 

20 processes such as the infiltration of water into the source area, the geometry of the source 

21 zone, the distribution of contaminants, release mechanisms, the physical form of the 

22 contaminants, near-field transport processes, and containment failure. If the contaminants are 

23 assumed to be uniformly distributed, this is an important assumption that needs to be justified 

24 because, in general, contaminants may not be uniformly distributed (see discussion under 

25 Section I.2 of this appendix). The source description should clearly identify how the 

26 contaminants are assumed to be released from the media. Common release mechanisms are 

27 diffusion, dissolution, surface release, and gas generation. The source description should also 

28 identify key processes and features that may retain or limit the release of contaminants from the 

29 source area (e.g., solubility and sorption). In addition, the description of near-field transport 

30 should state assumptions made about the dimensionality. In general, the assumption of one31 dimensional vertical flow should be appropriate, unless there is some type of barrier present that 

32 may hinder flow in the vertical direction. The description of the source term should also 
I-46

1 describe failure mechanisms for any containment (e.g., corrosion, concrete degradation, or 

2 cover degradation), if containers or other forms of containment are present.

3

4 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 5 

6 Release rates are typically expressed as a fractional release rate. Dissolution/precipitation is 

7 more likely to be the key process in locations such as at a point source, an area where high 

8 contaminant concentrations exist, or where steep pH or oxidation-reduction (redox) gradients 

9 exist. Adsorption/desorption will likely be the key process controlling inorganic contaminant 

10 migration in areas where the naturally present constituents are already in equilibrium and only 

11 the anthropogenic constituents (contaminants) are out of equilibrium, such as in areas far from 

12 the point source (EPA, 1999a; and EPA, 1999b).

13

14 Use of a Kd model results in the release of residual radioactivity decreasing over time as the 

15 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 16 -05 moles per liter (mol/L), and 

17 (2) desorption from a solid for six different assumed Kds (0.1, 1, 10, 50, 100, 200, 500, and 

18 1,000 L/kilogram (kg)). The total source is depleted more rapidly when solubility control is 

assumed (at a solubility of 3.5x10 19 -05 mol/L) compared to a leaching model using a Kd greater 

20 than around 50 L/kg, although the release rates may be higher initially for the Kd model. At 

21 lower Kds (i.e., less than about 50 L/kg), Figure I.6 shows the sorption model would deplete the 

22 total source more quickly than the solubility model. The Kd below which source depletion is 

23 more rapid than solubility limited release is a function of the bulk soil density, effective porosity, 

24 assumed solubility limit, and source inventory. Important characteristics of the curves shown in 

25 Figure I.6 are the following:

26

27 • Release rates (i.e., mass released per unit time) in a solubility-limited model are 

28 constant over time and the source decay curve is linear.

29 • The fractional release rate (i.e., fraction of source released per unit time) in a Kd model is 

30 constant over time (i.e., quantity released decreases over time) and the release curve is 

31 similar to an exponential, first order decay curve.

32

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. 
I-47

1

2 Figure I.6 Source Depletion for Solubility and Kd Release Models (for Six Values of Kd

3 Expressed in L/kg)

4 The conceptual model of the groundwater pathway should describe how contaminants could 

5 migrate through the unsaturated and saturated zones to potential receptors (e.g., a well, spring, 

6 or surface water). Essential features that should be included in the conceptual model include 

7 hydrostratigraphic units; boundary conditions; the physical and chemical form of the residual 

8 radioactivity (e.g., dissolved, suspended sediment, gas, speciation, complexation), the structural 

9 features of the geology (i.e., those that influence contaminant transport such as fractures, faults, 

10 and intrusions), and geochemical conditions and gradients important to contaminant transport. 

11 Important processes that should be characterized include the dimensions and state conditions 

12 (e.g., steady state) of flow, dimensions and state conditions of transport (e.g., dispersion), 

13 chemical and mass transfer processes (e.g., sorption, precipitation, complexation), and 

14 transformation processes (e.g., radioactive ingrowth and decay). Although contaminant 

15 migration through both the unsaturated and saturated zones is best represented in three 

16 dimensions, it may be appropriate to assume only one or two dimensions, if this provides a 

17 more conservative representation of contaminant migration, or if it can be demonstrated that 

18 migration in one or more other directions is not expected to result in exposure to potential 

19 receptors.

20 The conceptual model of the surface water pathway should describe potential contaminant 

21 migration to potential receptors through surface water bodies, such as lakes, streams, channels, 

22 or ponds. Essential features that should be included in the conceptual model include: the 

23 geometry of the surface water body (i.e., boundaries and boundary conditions), the physical 
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1 form of the contaminants (e.g., dissolved or solid), and physical and chemical properties. Key 

2 processes that should be described include: the dimensions and state conditions of flow and 

3 transport, chemical and mass transfer processes (e.g., sorption, precipitation, volatilization), and 

4 transformation. One key boundary condition that should be described is how the contaminants 

5 are expected to initially mix or interact with the surface water.

6

7 The conceptual model of the air pathway should describe potential contaminant migration 

8 through the air to potential receptors. Essential features that should be included in the 

9 conceptual model are similar to those for the other environmental pathways—namely, the 

10 geometry (i.e., boundaries and boundary conditions), form of contaminants (e.g., particulates or 

11 gases), and physical and chemical properties. Key processes that should be described include 

12 the dimensions and state conditions of flow and transport and the transformation processes.

13 SITE-SPECIFIC COMPUTER CODES

14 Three common computer codes used for site-specific analyses are RESRAD-ONSITE,

RESRAD-OFFSITE, and RESRAD–BUILD11 15 . All have predefined conceptual models. 

16 Therefore, in using these codes, it is important for the licensee to demonstrate that key site 

17 features and conditions are consistent with the modeling assumptions within the codes or, 

18 where they are not consistent, the analysis may not result in an underestimation of potential 

19 doses. Additional information is provided in Section I.5.3 regarding the built-in conceptual

20 models in these codes.

21

22 LIMITATIONS OF SITE-SPECIFIC COMPUTER CODES 

23

24 In general, the conceptual models within the RESRAD family of codes are expected to provide 

25 an acceptable generic representation of site features and conditions. Table I.7 lists some 

26 specific site features and conditions that may be incompatible with this generic representation. 

27 At any site where it is known that one or more of these conditions or features are present, the 

28 licensee should provide appropriate justification for use of the computer code.

29

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.
I-49

1 Table I.7 Site Features and Conditions that may be Incompatible with the Assumptions 

2 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

3

4 I.4.4 Generic Examples

5 I.4.4.1 Screening

6 A hypothetical research and development facility is authorized to use radiological chemicals 

7 through an NRC license. Because the research and development facility plans to discontinue 

8 its use of radioactive material, it wants to decommission the facility and terminate its license. An

9 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 10 ). The wall area is 

430 m2(4,600 ft2). In addition, an outside area of roughly 930 m2(10,000 ft2 11 ) was used for dry 

12 storage of chemicals. A preliminary characterization program has determined that 

13 approximately 10 percent of the building floor area and 5 percent of the wall area are 

14 contaminated with Cs-137 and Co-60. Surficial soils covering an area of approximately 

2,500 m2(27,000 ft2 15 ) are contaminated from windblown dust and runoff from spills in the storage 

16 area. The soils are also contaminated with Cs-137 and Co-60.
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1 The licensee proposes to use a screening analysis, using DandD, to demonstrate compliance 

2 with the LTR. A building occupancy exposure scenario is assumed for the building and a 

3 residential farmer exposure scenario is assumed for the contaminated soils. Based on what is 

4 known about the site, the licensee certifies that the use of the generic conceptual models within 

5 DandD is appropriate for the analysis.

6

7 I.4.4.2 Site-Specific

8 A hypothetical manufacturing facility has a former radioactive waste burial area that may be 

9 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 10

(22,000 ft2 11 ). The trenches, which are roughly 0.9 m (3 ft) deep, are covered with 1.2 m (4 ft) of 

12 native soil. A review of site operating records shows that the radionuclides of concern are 

13 natural uranium, enriched uranium, and natural thorium.

14 Based on information from the local county agricultural extension office and published reports, 

15 the geology and hydrogeology at the site are described as follows. This description shows that 

16 no site features or conditions in Table I.7 are present at this site.

17 The surface geology at the site contains 14 to 27 m (46 to 89 ft) of till consisting 

18 primarily of fine, silty sand to sandy silt with narrow, discontinuous sand lenses. 

19 Sandstone bedrock underlies the unconsolidated till. A shallow unconfined 

20 aquifer occurs in the unconsolidated till. The average depth to the water table 

21 ranges between 3 and 4 m below the land surface. The mean horizontal 

22 hydraulic conductivity is roughly 60 m/y (197 ft/y). The average vertical hydraulic 

23 conductivity of the till is estimated to be an order of magnitude less. The 

24 hydraulic gradient is estimated to range between 0.006 and 0.021. The mean 

25 precipitation at the site is roughly 0.8 m/y (30 in./y). The site is located in the 

26 reach of a surface water drainage basin that has a drainage area of 

approximately 500,000 m2(5.4 million ft2 27 ).

28 The licensee assumes a residential farmer exposure scenario as a reasonable future land use 

29 and proposes to use the RESRAD-ONSITE computer code for the dose analysis. Because the 

30 contaminated media is trash, an assumption is made that the trash degrades and becomes 

31 indistinguishable from soil. In addition, the metal drums are assumed to have degraded away. 

32 Given the relative short lifespan for metal drums and the long half-life of the radionuclides, this 

33 should be a reasonable assumption. The cover is also assumed to be breached through the 

34 construction of a basement for the house. The contaminated soil is assumed to be uniformly 

35 mixed with the excavated cover. Because the trash is assumed to be indistinguishable from 

36 soil, it is also assumed that, once the cover is breached, the future hypothetical farmer may not 

37 recognize the contaminated material as contaminated. The licensee also assumes that the 

38 hypothetical future well is located at the center of the residual radioactivity because of limited 

39 bases for assuming otherwise.

40 The licensee determines that the other aspects of conceptual models within RESRAD-ONSITE

41 are acceptable for analyzing the problem.

42
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1 I.5 Criteria for Selecting Computer Codes/Models

2 I.5.1 Introduction

3 Dose assessment commonly involves the execution of numerical model(s) that mathematically 

4 represent the conceptual model of the contaminated site. The numerical models used to 

5 implement the mathematical equations are usually linked via the conceptual model and codified 

6 in a software package known as “the code.” The words “code” and “model” are frequently used 

7 to express the software package, including the embedded numerical models or the specific 

8 models contained in the code. For example, “DandD code” may refer to the software package, 

9 including the associated exposure models (e.g., the water use model, food-ingestion pathway 

10 model, inhalation-exposure model) embedded in the code. The “DandD model” may also refer 

11 to DandD software, the DandD conceptual model, or any of the numerical models, or the group 

12 of models used in the code (e.g., DandD groundwater model). Within the context of this 

13 volume, the word “code” will refer to the software package and the associated numerical 

14 models. However, the word “model” will refer to the mathematical representation of the 

15 conceptual model, including representation of the specific exposure scenario and pathways. 

16 This section describes the process and criteria used in the selection of codes and models for 

17 dose assessments.

18 The codes and models used in the dose assessment can be either generic screening 

19 codes/models or site-specific codes/models. Regardless of the intent of the use of the 

20 code/model (e.g., for screening or site-specific analysis), the reviewer should ensure that the 

21 licensee properly documents and verifies the dose assessment codes/models and the 

22 associated databases in accordance with rigorous QA/QC criteria acceptable to the NRC. 

23 Currently, the only acceptable generic screening code is DandD. If the licensee uses site24 specific models/codes, it should justify the conceptual model used (see Section I.4.3.2 of this 

25 appendix). The NRC staff should also review the source term model(s), the transport models, 

26 the exposure models, and the overall dose models and assess the QA/QC documentation and 

27 level of conservatism of any alternative code/model.

28 This section describes the generic issues associated with the selection of the screening and 

29 site-specific codes/models that the NRC staff may encounter and recommends approaches and 

30 criteria for its acceptance of the codes/models. In addition, this section presents a generic 

31 description of common dose assessment codes, DandD, RESRAD-ONSITE and RESRAD–

32 BUILD. The NRC developed or funded (in part) the development of these codes. In addition, 

33 the NRC staff and licensees have used these codes to demonstrate compliance with the dose 

34 criteria in Subpart E.

35

36 I.5.2 Issues in Selection of Computer Codes/Models

37 The major issues associated with the selection of computer codes/models include the following: 

38 • Generic criteria for the selection of computer codes/models: This issue pertains to

39 the NRC staff’s review criteria of code aspects related to QA/QC requirements,

40 specifications, testing, verification, documentation, interfacing, and other features related

41 to uncertainty treatment approaches.

42 • Acceptance criteria for the selection of site-specific codes/models: This issue

43 pertains to the NRC staff’s review of additional specific requirements for justifying the
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1 use of the conceptual model, the numerical mathematical models, the source term 

2 model and its abstraction, and the transport and exposure pathway models.

3 • Options for the selection of deterministic or probabilistic site-specific codes: This

4 issue pertains to the NRC staff’s review of the justification to support the decision to use

5 either of these two approaches.

6 A generic description of the DandD Version 2 is presented below to familiarize users with this 

7 code. Further, the rationale for the development of DandD Version 2 and the issue of excessive 

8 conservatism in DandD Version 1 are addressed. A description of the approaches to minimize

9 such excessive conservatism, using DandD Version 2, site-specific input data, or use of other 

10 models/codes is included.

11 For a site-specific analysis, the NRC staff should accept any model or code that meets the 

12 criteria described below in “Generic Criteria for the Selection of Codes/Models.” However, the 

13 staff is expected to conduct a more detailed and thorough review of less common codes/models 

14 (e.g., codes other than DandD and the RESRAD family of codes), specifically those developed 

15 by licensees. The NRC sponsored development of the probabilistic RESRAD-ONSITE

(Version 6) and RESRAD–BUILD (Version 3) codes for site-specific analysis12 16 . These have 

17 already been reviewed for QA/QC and are acceptable.

18 The selection of appropriate models/codes for complex sites may also present challenges. For 

19 example, sites with multiple source terms, with significant groundwater or surface water 

20 contamination, or sites with existing offsite releases, may require more advanced codes/models 

than commonly used codes such as DandD or RESRAD-ONSITE13 21 . Complex sites may also 

22 include sites with engineered barrier(s), or with complex hydrogeological conditions such as 

23 highly fractured geologic formations. Because of site complexity and variability, there are no 

24 standard dose analysis review criteria for these sites.

25

26 I.5.3 Recommended Approach 

27 I.5.3.1 Generic Criteria for Selection of Codes/Models

28 The generic criteria under this subsection pertain to the NRC staff review of codes/models other 

29 than commonly used codes; specifically, those developed or modified by the NRC staff 

30 (i.e., other than DandD, RESRAD-ONSITE, and RESRAD–BUILD). The reviewer should use 

31 the generic criteria when the codes/models have no readily available documentation of testing, 

32 verification, and QA/QC review. In this context, the reviewer should use the following generic 

33 criteria in reviewing the codes/models selected for the dose assessment:

34 • The NRC staff should review the adequacy and completeness of the database available

35 on QA/QC aspects of the code/model. The QA/QC database should be comparable to

36 the NRC’s QA/QC requirements (NUREG/BR-0167, “Software Quality Assurance

37 Program and Guideline,” issued February 1993 (NRC, 1993b) and NUREG-0856, “Final

38 Technical Position on Documentation of Computer Codes for High Level Waste

39 Management,” issued June 1983 (NRC, 1983)). The QA/QC should include information

40 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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1 with the assumed conceptual model(s) used in the code, and accuracy of the software to 

2 reflect the model’s mathematical formulation and correct representation of the process or 

3 system for which it is intended.

4 • The NRC staff should ensure that the software used for the code is in conformance with

5 the recommendations of the Institute of Electrical and Electronics Engineers

6 (IEEE) Standard 830-1984, “IEEE Guide for Software Requirement Specifications.”

7 • The NRC staff should review the adequacy and appropriateness of the code/model

8 documentation with regard to (1) software requirements and intended use, (2) software

9 design and development, (3) software design verification, (4) software installation and

10 testing, (5) configuration control, (6) software problems and resolution, and (7) software

11 validation.

12 • For uncommon codes/models, the NRC staff should review code data, including (1) a

13 software summary form, (2) a software problem/change form, (3) a software release

14 notice form, and (4) a code/model user’s manual, which covers code technical

15 description, software source code, functional requirements, and external interface

16 requirements (e.g., user interface, hardware interface, software interface, and

17 communication interface), if necessary.

18 • The NRC staff should review the conceptual model of the selected code to ensure

19 compatibility with the specific site conceptual model, including the pathways and the

20 exposure scenario. The source term assumptions of the selected code should also be

21 compatible with the site-specific source term. The staff may accommodate minor

22 modifications in the source term conceptual model, as long as the basic model

23 assumptions are not violated.

24 • The NRC staff should verify that the exposure scenario of the selected code is

25 compatible with the intended scenario for the site. For example, models/codes designed

26 for the onsite exposure scenario may not be appropriate for assessment of an offsite

27 exposure scenario.

28 • The NRC staff should review the selected model/code formulation to account for

29 radionuclide decay and progenies. The code should have proper and timely formulation,

30 as well as linkages of decay products with the receptor location and the transport

31 pathways, via corresponding environmental media.

32 • The NRC staff should examine the documentation of the selected code/model

33 performance; specifically, test and evaluation, as well as a code comparison with

34 commonly used (accepted) codes and models (e.g., DandD and the RESRAD family of

35 codes). The staff should also review documentation on code/model verification, if

36 available, to support decisions for code acceptance.

37 • The NRC staff should review code/model features of the sensitivity/uncertainty analysis

38 to account for variability in the selection of input parameters and uncertainty in the

39 conceptual model and multiple options for the interpretation of the system.
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1 I.5.3.2 Acceptance Criteria for Selection of Site-Specific Codes/Models

2 This issue involves the NRC staff’s review of additional requirements supporting the justification 

3 for using the conceptual model, the numerical mathematical models, the source term model and 

4 its abstraction, and the transport and exposure pathway models.

5 CONCEPTUAL MODELS

6 The NRC staff review should compare the conceptual model for the site with the conceptual 

7 model(s) in the selected code, to ensure compatibility with site-specific physical conditions and 

8 pathway assumptions for the critical group receptor.

9 NUMERICAL MATHEMATICAL MODELS

10 The staff should review the equations used in the code to implement the conceptual model and 

11 the numerical links between mathematical models to ensure correctness and consistency. For 

12 codes developed or modified by the NRC staff (e.g., DandD, RESRAD-ONSITE, and RESRAD–

13 BUILD), minimal review is needed, because the NRC revised these codes and examined them 

14 early for consistency with its QA/QC requirements. For less commonly used codes, or codes 

15 developed locally by user(s), the NRC staff should verify the numerical mathematical models, 

16 including the numerical links between these models. In this context, the reviewer may examine, 

17 if necessary, each mathematical model used for the specific transport-exposure pathway, to 

18 ensure that the code is designed for its intended use.

19

20 SOURCE TERM MODELS

21

22 The NRC staff should review the source term model(s) used for the specific site. In this context, 

23 the review should include the following source term aspects: 

24 • Building Occupancy Exposure Scenario Source Term: The NRC staff should review 

25 the HSA and other relevant data on the extent of the source and its depth (e.g., within 1 

26 to 10 millimeters (mm) (0.04 to 0.39 in.) deep into the building surface or more). Based 

27 on this review, the reviewer should identify the source as surficial or volumetric and 

28 examine the assumptions made for the loose/fixed fractions of the source. The review 

29 should address the sources of residual radioactivity on surfaces that are not integral 

30 parts of the building (e.g., equipment, pipes, and sewer lines) separately, because the

31 applicable model and exposure scenario could be different. Therefore, source term 

32 model assumptions for such surfaces should be reviewed on a case-by-case basis. 

33 The NRC staff should also review the radionuclide mixture comprising the source and 

34 whether a constant ratio is assumed in the dose analysis, as well as determine if 

35 surrogate radionuclides are being used. The latter two situations may require additional 

36 NRC staff verification of the source definition and review of consistency with the 

37 intended final survey methodology. 

38 The review should also include the use of multiple sources (e.g., multiple rooms). 

39 Certain codes may provide an option to define multiple sources in various configurations,

40 such as two to three rooms, with multiple-story buildings. The source term under these 

41 conditions allows for source depletion due to open air circulation and common 

42 ventilation. For example, the RESRAD–BUILD code model uses two- or three-room 

43 models with two- or three-story buildings, allowing for air exchange within the rooms, 
I-55

1 and source depletion. The review should include the indoor air-quality model 

2 (e.g., building ventilation and infiltration), and the indoor air-concentration model, as well 

3 as the adaptation of the air-quality model in the RESRAD–BUILD code, to ensure 

4 consistency with the site-specific conditions. The review should verify the input 

5 parameters associated with these models. The NRC staff may accept such site-specific 

6 source term models after an assessment of the compatibility of the source term model 

7 with the conceptual model of the site. The review should also include the physical 

8 parameters defining the source term, to ensure consistency with site-specific conditions, 

9 and the occupancy parameters, to ensure consistency with the exposure scenario.

10 • Resident Farmer Exposure Scenario Source Term: The NRC staff should examine

11 the licensee’s source information to determine if the nature and extent of residual

12 radioactivity is consistent with the model/assumptions in the selected code. The review

13 should include evaluation of the vertical and horizontal extent of residual radioactivity, to

14 verify model assumptions on the area and thickness of the contaminated zone, and to

15 determine if there is subsurface and/or groundwater contamination at the site. The

16 DandD model assumes residual radioactivity is located in surface soils only

17 (approximately 15 cm or less). For contaminated zone thicknesses significantly greater

18 than 15 cm, DandD may underestimate the dose and justification would be needed to

19 use DandD to assess the dose from subsurface residual radioactivity. The contaminated

20 area and its shape should also be assessed to check for possible correction of the area

21 and/or for geometry of the source. Additionally, the NRC staff should determine if credit

22 is taken for a clean cover or a barrier over the contaminated zone and the basis for the

23 modeling assumption. Support for the assumed level of performance of the engineered

24 cover or barrier should be evaluated within the context of institutional control

25 assumptions (e.g., active maintenance for the case where institutional controls are in

26 effect; and only passive performance in the case when institutional controls are assumed

27 to no longer be in effect). Although sensitivity analysis on the timing of failure of

28 institutional controls can be performed, institutional controls should be assumed to fail at

29 time=0 years. The evaluation should consider assumptions regarding the performance

30 and degradation of the engineered covers and barriers during the compliance period

31 (e.g., 1,000 years).

32 The NRC staff should also review the physical and chemical form of the source to 

33 evaluate the adequacy of the underlying soil leaching model(s) available in the selected 

34 code. This review should help assess the source mass-balance model and the transport 

35 model within the concerned environmental media. In addition, a review of these source 

36 term aspects would help establish consistencies for the selection of relevant parameters. 

37 The review should include the horizontal distribution and homogeneity of the source and 

38 the variation of source concentration with depth. The NRC staff should use either an 

39 upper-bounding value for modeling the thickness or an area-weighted approach to 

40 calculate the representative thickness, if this approach does not significantly 

41 underestimate the potential dose (e.g., it may not be appropriate to represent an area 

42 with a thin vadose zone with a thicker vadose zone based on source thickness 

43 averaging, if the thicker vadose zone assumption could lead to longer transport times 

44 and lower exposure concentrations due to radiological decay). In certain cases, multiple 

45 sources or more advanced subsurface source term modeling may be needed to 

46 adequately assess dose; these would be evaluated on a case-by-case basis.

47
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1 TRANSPORT MODELS

2

3 The transport models simulate transport mechanisms of contaminants from the source to the 

4 receptor. The NRC staff should review transport models for consistency and compatibility with 

5 respect to (1) the source term, (2) the exposure scenario defined for the critical group receptor, 

6 and (3) the simplified conceptual model, which describes site-specific physical conditions. The 

7 transport models may include the diffusive and advective transport of contaminants via air, 

8 surface water, and groundwater. The transport models can be overly simplified, using simple 

9 conservative assumptions such that minimal characterization data would be required to execute 

10 the model(s). Transport models can also be very complex, requiring advanced mathematical 

11 derivation and extensive site-specific, or surrogate data about the site.

12 For the building occupancy exposure scenario, the associated transport models (e.g., transport 

13 models for ingestion, inhalation, and direct exposure pathways) of the DandD code are simple 

14 and conservative. For example, the ingestion pathway depends on the effective transfer rate of 

15 the removable surface residual radioactivity from surfaces to hands and from hands to mouth. 

16 The inhalation transport model depends largely on mechanical disturbance of the contaminated 

17 surface, resuspension of residual radioactivity in the air, and subsequent breathing of 

18 contaminated air. The external dose formulation assumes exposure from a nonuniform source 

19 of residual radioactivity on the walls, ceiling, and floor of a room. This model was found to be 

20 comparable to the infinite plane source for the building occupancy exposure scenario (NRC, 

21 1992).

22 For the resident farmer exposure scenario, the associated DandD transport models include 

23 models of contaminants transport to groundwater, to surface water (e.g., three-box model that 

24 relies on transfer of contaminate through leaching), and to air (e.g., through dust mass loading 

25 and indoor resuspension). Transport models of contaminants via the air include dust loading, 

26 resuspension of contaminated soil, and use of mass loading factor for deposition. Transfer of 

27 contaminants from the soil/water to plants, fish, animals, and animal products are calculated 

28 using a water use model, along with transfer factors, translocation factors, and bio-accumulation 

29 factors. Separate models were used for C-14, and H-3, as described in NUREG/CR-5512, 

30 Volumes 1, 2, and 3. The RESRAD-ONSITE model can consider residual radioactivity in 

31 surface soils (approximately 15 cm) or thicker contaminated zones, with an idealized cylindrical 

shape of the contaminated zone14, 32 and allows for a cover at the top of the contaminated zone, if 

33 appropriate.

34 In general, the NRC’s review of the selected code should include transport models and the 

35 appropriateness of such models with respect to the site-specific conditions (e.g., area, source, 

36 unsaturated zone, and aquifer conditions). In addition, the staff should review, for compatibility 

37 and consistency, the transport model assumptions and the generic formulation pertaining to the 

38 applicable pathways of the critical group exposure scenario. The extent of the transport model 

39 review depends on the familiarity of the NRC staff with these models. Because they developed 

40 or modified certain commonly used codes/models (e.g., DandD, RESRAD-ONSITE, and 

41 RESRAD-BUILD), the NRC staff is more familiar with them, and they would require less of a 

42 review than for less common codes/models developed by users or other parties. The NRC 

43 review should also include updated new models or code versions and studies on code/models 

44 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.
I-57

1 The RESRAD–BUILD code is more advanced than the DandD code because it employs 

2 multiple sources and more advanced particulate air transport models. In other words, each 

3 contaminated location may be considered a distinct source. Depending on its geometric 

4 appearance, the source can be defined either as a volume, an area, or a point source. The 

5 RESRAD–BUILD code depends on erosion of the source and transport of part of its mass into 

6 the indoor air environment, resulting in airborne residual radioactivity. The RESRAD–BUILD 

7 model differs from DandD because it assumes air exchange among all compartments of the 

8 building. In other words, the model assumes that the airborne particulates are being loaded into 

9 the indoor air of the compartment and then transported to the indoor air of all compartments of 

10 the building. In addition to air exchange between compartments, the indoor air model simulates 

11 air exchange between compartments and the outdoor air. Descriptions of models pertaining to

12 indoor air quality, air particulate deposition, inhalation of airborne dust, and ingestion of 

13 removable materials and deposited dust, were documented in an Argonne National Laboratory 

14 report “ANL/EAD/LD–3” (ANL 1994). The exposure pathways in the RESRAD–BUILD code 

15 include (1) the external exposure to radiation emitted directly from the source and from 

16 radioactive particulates deposited on the floors and exposure caused by submersion from 

17 radioactive particulates, (2) inhalation of airborne radioactive particulates, and (3) ingestion of 

18 contaminated material directly from the source, as well as airborne particulates deposited onto 

19 the surface of the building.

20

21 EXPOSURE PATHWAY MODELS

22

23 The exposure pathway models pertain to the formulation of the links between the radiological 

24 source, the transport of contaminants within environmental media, the critical group receptor 

25 location, and behaviors of the receptor that lead to its exposure to residual radioactivity through 

26 direct exposure, inhalation, and ingestion of contaminated water, soil, plants, crops, fish, meat, 

27 milk, and other dairy products. The NRC staff should review the conceptual model(s) that 

28 describe the human behaviors that lead, or control, the amount of receptor exposure. 

29 Therefore, the occupational, behavioral, and metabolic parameters describing these models 

30 should be reviewed and compared with the default model exposure scenarios and associated 

31 parameters. The NRC staff should review exposure model(s) and associated parameters to 

32 ensure conservatism, consistency, and comparability with site-specific conditions and exposure 

33 scenario assumptions. NUREG/CR–5512, Volumes 1, 2, and 3, provide detailed information on 

34 default parameters and approaches for changing parameters in dose modeling analysis.

35 I.5.3.3 Option for Selection of Deterministic or Probabilistic Site-Specific Codes

36 Licensees may select either a deterministic analysis or a probabilistic approach for 

37 demonstrating compliance with the dose criteria in 10 CFR Part 20, Subpart E. A deterministic 

38 analysis uses single parameter values for each variable in the code. In contrast, parameter 

39 distributions are specified for uncertain variables in a probabilistic analysis, sets of parameter 

40 values are selected via sampling from the specified parameter distributions, many sets of 

41 parameter values called realizations are run through the model, and a distribution of results is 

42 generated and evaluated. Although deterministic sensitivity analysis can be conducted, a 

43 deterministic analysis gives more limited information on the uncertainty in the results, based on 

44 uncertainty in the parameter values, and is generally less efficient in identifying important model 

45 sensitivities. Therefore, the deterministic approach may require a more elaborate justification of

46 code input parameter values and may require further analysis of doses using upper or lower 

47 bounding conditions.

48
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1 NRC-approved data sets for both DandD and the RESRAD family of codes are for the 

2 probabilistic calculation and not the deterministic mode.

3 Section I.7.3.2 of this appendix contains a detailed description of an NRC staff review for both 

4 deterministic and probabilistic analyses. 

5 I.5.3.4 Modeling of Subsurface Residual Radioactivity

6 For subsurface residual radioactivity (residual radioactivity deeper than approximately 15 cm 

7 (6 in.)), the NRC staff should review existing historical site data (including previous processes or 

8 practices) and site characterization data, to establish an adequate conceptual model of the 

9 subsurface source; specifically, the horizontal and vertical extent of residual radioactivity. 

10 Section I.2.3.1 describes approaches for subsurface source term abstraction for dose modeling 

11 analysis. In some cases, the licensee may wish to develop multiple DCGLs for surface and 

12 subsurface DCGLs. Additional information on integration of dose modeling with radiological 

13 surveys for license termination is found in Appendix G.

14 I.5.3.5 Generic Description and Development of DandD 

15 The DandD code has two default land use exposure scenarios: a building occupancy and a 

16 resident farmer exposure scenario. The building occupancy exposure scenario is intended to 

17 account for exposure to both fixed and removable residual radioactivity within a building. 

18 Exposure pathways included in the building occupancy exposure scenario include external 

19 exposure to penetrating radiation, inhalation of resuspended surface residual radioactivity, and 

20 inadvertent ingestion of surface residual radioactivity. The resident farmer exposure scenario is 

21 intended to account for exposure to residual radioactivity in soil. Resident farmer exposure 

22 scenario pathways include the following: external exposure to penetrating radiation; inhalation 

23 exposure to resuspended soil; ingestion of soil; and ingestion of contaminated drinking water, 

24 plant products, animal products, and fish. The predefined conceptual models within DandD are 

25 geared to assessing releases of radioactivity, transport to, and exposure along, these pathways.

26 For the building occupancy exposure scenario, DandD models external exposure to penetrating 

27 radiation as an infinite area source, using surface source dose rate factors from “Federal 

28 Guidance Report No. 12: External Exposure to Radionuclides in Air, Water and Soil,” issued 

29 September 1993 (EPA 1993). Exposure to inhalation of resuspended surface residual 

30 radioactivity is modeled as a linear static relationship between surface residual radioactivity and 

31 airborne concentrations. The model accounts for ingrowth and decay. Exposure to incidental 

32 ingestion of surface residual radioactivity is modeled with a constant transfer rate.

33 The generic conceptual models for the resident farmer exposure scenario are more complicated 

34 because of the large number of exposure pathways and considerations of release of 

35 radioactivity from the source area and transport of radionuclides in the environment. DandD 

36 models external exposure from volume soil sources when the person is outside as an infinite 

37 slab of residual radioactivity 15 cm (6 in.) thick, using dose rate factors from Federal Guidance 

38 Report No. 12 for volume residual radioactivity. When the person is indoors, exposure from 

39 external radiation is modeled in a similar manner, except the exposure is assumed to be 

40 attenuated through the use of a shielding factor (note: the higher the shielding factor, the lower 

41 the assumed attenuation). Exposure through ingestion of contaminated animal and plant 

42 products is modeled simply through the use of transfer factors. Instantaneous equilibrium is 

43 assumed to occur between radionuclide concentration in the soil and the concentration in 

44 plants, and between animal feed and animal products.
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1 The generic source term conceptual model in DandD assumes a constant release rate of 

2 radionuclides into the water and air pathways. Release of radionuclides by water is assumed to 

3 be downward and a function of a constant infiltration rate, constant contaminant zone thickness, 

4 constant moisture content, and equilibrium adsorption. DandD assumes that there are no 

5 radioactive gas or vapor releases. Release of radioactive particulates is assumed to be upward, 

6 instantaneous, uniform, and a function of a constant particulate concentration in the air and the 

7 radioactivity within the soil. Radionuclides in the contaminant zone are assumed to be uniformly 

8 distributed in a single soil layer, 15 cm (6 in.) thick. No transport is assumed to occur within the 

9 source zone, but radioactive decay is taken into account. In terms of containment, DandD 

10 assumes that there are no containers (or that they have failed), and that there is no cover over 

11 the contaminated zone.

12 The DandD generic conceptual model for the groundwater pathway assumes a single 

13 hydrostratigraphic layer for each of the unsaturated and saturated zones. The unsaturated zone 

14 (vadose zone) can be broken into multiple layers within DandD; however, each layer is 

15 assumed to have the same properties. For radionuclides entering the vadose zone, DandD 

16 accounts for adsorption-limited leaching by considering the vadose zone to behave as a well17 mixed chemical reactor with a constant water inlet and outlet rate set at the infiltration rate. 

18 Accordingly, it is assumed that the vertical saturated hydraulic conductivity of the unsaturated 

19 zone is greater than or equal to the infiltration rate (i.e., there is neither ponding nor runoff on 

20 the surface). The outlet concentration from one unsaturated zone layer to another is assumed 

21 to be a function of the constant infiltration rate, equilibrium partitioning, the thickness of the 

22 layer, a constant moisture content, and radioactive decay. Radionuclides entering the saturated 

23 zone are assumed to be instantaneously and uniformly distributed over a constant volume of 

24 water equivalent to the larger of either the volume of infiltrating water (i.e., the infiltration rate 

25 times the contaminated area) or the sum of the water assumed to be removed for domestic use 

26 and irrigation. Based on the default parameters in DandD, dilution in the groundwater pathway 

27 is based on the water use. No retardation is assumed to occur in the aquifer; however, 

28 radioactive decay is taken into account. A volume of contaminated water equivalent to the 

29 irrigation volume is assumed to be returned annually to the source zone. The concentration of 

30 radionuclides in the irrigation water is assumed to remain constant during the year. 

31 Radionuclides deposited on the vegetation are assumed to be removed at a constant rate. The 

32 DandD groundwater model should generally provide a conservative representation of the 

33 groundwater system, because it allows very little dilution and nominal attenuation.

34 The generic surface water conceptual model in DandD assumes that radionuclides are 

35 uniformly mixed within a finite volume of water representing a pond. Radionuclides are 

36 assumed to enter the pond at the same time and concentration as they enter the groundwater. 

37 Accordingly, there is assumed to be no transport of radionuclides through the groundwater to 

38 the pond and thus no additional attenuation (besides the initial groundwater dilution) is assumed 

39 for transport in the groundwater. The surface water model within DandD should provide a 

40 conservative dose estimate as long as a small volume is assumed for the surface water pond. 

41 Because the parameters in DandD are selected to provide a conservative dose estimate, the 

42 generic conceptualization of the surface water pathway should generally provide a conservative 

43 representation of transport of radionuclides through it. Figure I.7 shows the generic 

44 groundwater and surface water conceptual model within DandD.

45 The generic conceptual model of the air pathway in DandD assumes an equilibrium distribution 

46 between radionuclides in the air and soil. The concentration in air is assumed to be a function 

47 of the soil concentration and a constant dust loading in the air. Accordingly, all radionuclides in 

48 the air are assumed to be in a particulate form. The air pathway model within DandD is very 
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1 simple and should generally allow a conservative dose estimate, as long as a conservative 

2 particulate concentration is assumed. Because the default parameters in DandD are geared to 

3 be conservative, the air pathway in DandD should generally allow a conservative dose estimate.

4

5

6

7

8

9

10

11

12

13

14

15

16

17

18

19

20

21

22

23

24

25

26 Figure I.7 DandD Conceptual Model of the Groundwater and Surface Water Systems 

27 (from NUREG/CR-5621)

28 PROBABILISTIC DANDD VERSION 2

29 The NRC staff developed a probabilistic DandD Version 2, which updates, improves, replaces 

30 and significantly enhances the capabilities of Version 1.0. In particular, Version 2 allows full 

31 probabilistic treatment of dose assessments, whereas Version 1.0 embodied constant default 

32 parameter values and only allowed deterministic analyses. DandD implements the 

33 methodology and information contained in NUREG/CR–5512, Volume 1 (NRC, 1992), as well 

34 as the parameter analysis in Volume 3 (NRC, 1999c), which establishes the PDFs for all of the 

35 parameters associated with the exposure scenarios, exposure pathways, and models embodied 

36 in DandD.

Box 1

Box 2

Unsaturated-Soil

Layer

Surface-Soil

Layer

Thickness = H2

Thickness = H1

Ground-Water Aquifer Surface-Water Pond

Box 3

Ground-Water Well

I = Infiltration

Precipitation/Evaporation

Irrigation Domestic Use
I-61

1 Finally, DandD Version 2 includes a sensitivity analysis module that assists licensees and the 

2 NRC staff to identify those parameters in the screening analysis that have the greatest impact 

3 on the results of the dose assessment. Identification of risk-significant parameters helps 

4 licensees make informed decisions on the allocation of resources needed to gather site-specific 

5 information important to the compliance demonstration. 

6 I.5.3.6 Generic Description of the RESRAD Family of Codes 

7 Argonne National Laboratory developed the RESRAD family of computer codes under the 

8 sponsorship of the U.S. Department of Energy, and other agencies, such as the NRC. These 

9 codes are pathway analysis models designed to evaluate potential radiological doses 

10 associated with exposure of members of the public to residual radioactivity in soils and building 

11 materials, respectively. The RESRAD-ONSITE code uses a residential farmer exposure 

12 scenario (ANL/EAD/LD–2, “Manual for Implementing Residual Radioactive Material Guidelines 

13 Using RESRAD 5.0,” issued September 1993 (ANL, 1993a)) with nearly identical exposure 

14 pathways as the DandD residential exposure scenario described in NUREG/CR–5512, 

15 Volume 1 (NRC, 1992). The RESRAD–BUILD code uses a scenario that covers all exposure 

16 pathways in the DandD building occupancy exposure scenario, plus pathways corresponding to 

17 external exposures from air submersion and deposited material and to ingestion of deposited 

18 material. Previous sections in this appendix briefly described the RESRAD and RESRAD–

19 BUILD codes and conceptual models (see Section I.4.3.2). The RESRAD-OFFSITE code 

20 extends RESRAD-ONSITE capabilities to include offsite exposure locations, as well as including 

21 more complex source terms, groundwater, air and surface water models. Detailed descriptions 

22 of these codes are available in various technical documents, including:

23

24 • ANL/EAD/LD-2, “Manual for Implementing Residual Radioactive Material Guidelines

25 Using RESRAD, Version 5.0,” issued September 1993 (ANL, 1993a);

26 • ANL/EAD/LD-3, “RESRAD-BUILD: A Computer Model for Analyzing the Radiological

27 Doses Resulting from the Remediation and Occupancy of Buildings Contaminated with

28 Radioactive Material,” issued November 1994 (ANL, 1994);

29 • NUREG/CR-6697, “Development of Probabilistic RESRAD 6.0 and RESRAD-BUILD 3.0

30 Computer Codes,” issued December 2000 (NRC, 2000c);

31 • NUREG/CR-7268, “User’s Manual for RESRAD-OFFSITE Code Version 4, Vol. 1—

32 Methodology and Models Used in RESRAD-OFFSITE Code,” also ANL/EVS/TM-19/2,

33 Volume1, issued February 2020 (NRC, 2020b)

34 • NUREG/CR-7268, “User’s Manual for RESRAD-OFFSITE Code Version 4, Vol. 2—

35 User’s Guide for RESRAD-OFFSITE Version 4,” also ANL/EVS/TM-19/2, Volume 2,

36 issued February 2020 (NRC, 2020c).

37

38 The NRC staff and licensees widely used the deterministic versions of these codes before the 

39 LTR to estimate doses from radioactively contaminated sites and structures. The NRC 

40 sponsored development of the probabilistic versions (RESRAD-ONSITE, Version 6, and 
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RESRAD-BUILD, Version 3), and their default probabilistic data sets. These two codes 1

15 were 

2 selected because they possess all three of the following attributes:

3

4 (1) The software has been widely accepted, and there is already a large user base among

5 the NRC staff and licensees.

6 (2) The models in the software were designed, and have been applied successfully, to more

7 complex physical and residual radioactivity conditions than the DandD code.

8 (3) Verification and validation of these two codes are well documented (C. Yu, “RESRAD

9 Family of Codes and Comparison with Other Codes for Decontamination and

10 Restoration of Nuclear Facilities,” issued 1999 (Yu, 1999)).

11 It should be noted that the RESRAD-ONSITE code has been widely used and tested by national 

12 and international agencies and has gone through verification (HNUS-ARPD-94-174, 

13 “Verification of RESRAD—Case for Implementing Residual Radioactive Material Guidelines, 

14 Version 5.03,” issued June 1994 (HNUS, 1994)), dose model comparison (NRC, 1999d; Electric 

15 Power Research Institute (EPRI) TR–112874, “Comparison of Decommissioning Dose Modeling 

16 Codes for Nuclear Power Plant Use: RESRAD and DandD,” issued November 1999 

17 (EPRI, 1999), and benchmarking (DOE/ORO–2033, “Benchmarking Analysis of Three 

18 Multimedia Models: RESRAD, MMSOILS and MEPAS,” issued November 1995 (DOE, 1995)). 

19 RESRAD-ONSITE, RESRAD-OFFSITE, and RESRAD–BUILD codes are continuously being 

20 developed and updated with new code versions. Licensees may use updated versions of the 

21 RESRAD family of codes that have gone through adequate verification, validation, and QA 

22 testing and should document the version of the codes used in their DP.

23

24 RESRAD–BUILD

25 The RESRAD–BUILD code can be used to evaluate doses for the building occupancy exposure 

26 scenario. It considers exposure from external radiation at the source and air submersion, 

27 inhalation of airborne material, and inadvertent ingestion of radioactive material. Exposure to 

28 direct radiation at the source is calculated using surface source dose rate factors from Federal 

29 Guidance Report No. 12. The RESRAD–BUILD code incorporates correction factors to account 

30 for a finite area source, for any offset of the receptor from the axis of the disk of residual 

31 radioactivity, and for shielding by building materials. Exposure to external radiation from air 

32 submersion is calculated as an infinite cloud of material using dose rate conversion factors for 

33 an infinite cloud. The RESRAD–BUILD code models the airborne concentration of 

34 radionuclides using a dynamic model that accounts for the kinetic introduction and removal of 

35 radioactive material to and from indoor air. Exposure to incidental ingestion of radioactive 

36 material is modeled using a constant transfer rate.

37 RESRAD-ONSITE AND RESRAD-OFFSITE

38

39 The RESRAD-ONSITE code can be used for analyzing the resident farmer exposure scenario. 

40 As with the generic conceptual models used by DandD for analyzing the resident farmer and 

41 building occupancy exposure scenarios, the conceptual model in RESRAD-ONSITE (see 

42 Figure I.8) is more complex than the exposure scenario in RESRAD–BUILD. The RESRAD43 ONSITE code considers external exposure from volume soil sources when the receptor is 

15 RESRAD-OFFSITE was developed after the development of the probabilistic versions of RESRAD-ONSITE and 

RESRAD-BUILD.
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1 outside, using volume dose rate factors from Federal Guidance Report No. 12. Correction 

2 factors are used to account for soil density, areal extent of residual radioactivity, thickness of 

3 residual radioactivity, and cover attenuation. When the person is indoors, exposure from 

4 external radiation is modeled in a similar manner, except that additional attenuation is 

5 considered to account for protection afforded by the building. Exposure through ingestion of 

6 contaminated animal and plant products is modeled simply through use of transfer factors. 

7 RESRAD-OFFSITE extends RESRAD-ONSITE to offsite locations and features more complex 

8 source term; and atmospheric, groundwater, and surface water transport models. Additional 

9 details regarding each of the models is found below.

10 Source Term Models

11

12 The generic source term conceptual model in RESRAD-ONSITE assumes a time-varying 

release rate of radionuclides into the water and air pathways.

16 13 Radionuclides in the 

14 contaminated zone are assumed to be uniformly distributed. No transport is assumed to occur 

15 within the source zone, but radioactive decay is accounted for. In terms of containment, the 

16 radioactive material is not assumed to be contained (or containers are assumed to have 

failed).

17 17 RESRAD-ONSITE allows inclusion of a cover over the contaminated area. However, 

18 the cover is not assumed to limit infiltration of water. The cover provides shielding from gamma 

19 radiation and can also decrease the amount of radioactivity taken up by plants. Release of 

20 radionuclides by water is assumed to be a function of a constant infiltration rate, time-varying 

21 contaminant zone thickness, constant moisture content, and equilibrium adsorption. The 

22 contaminant zone is assumed to decrease over time from a constant erosion rate. RESRAD23 ONSITE assumes a uniform release of H-3 and C-14 gases, based on a constant evasion loss 

24 rate. Particulates are assumed to be instantaneously and uniformly released into the air as a 

25 function of the concentration of particulates in the air, based on a constant mass loading rate.

26 Figure I.9 provides conceptual information on the impact of the Kd on the magnitude of source 

27 release from the source zone and the mobility of contaminants in the natural environment, as 

28 conceptualized in RESRAD-ONSITE. Zones 1 and 2 depicted in Figure I.9 are the 

29 contaminated and unsaturated zones, respectively. For relatively low contaminated zone Kds, 

30 RESRAD-ONSITE predicts a higher concentration release into the environment, signified by the 

31 dark blue contaminated water leaching into the unsaturated zone (Cases A and B). For 

32 relatively high contaminated zone Kds, RESRAD-ONSITE predicts a lower concentration release 

33 into the environment, signified by the light blue contaminated water leaching into the 

34 unsaturated zone (Cases C and D). The Kd assigned to the unsaturated zone controls transport 

35 times to the aquifer with lower unsaturated zone Kds leading to faster transport rates or plumes 

36 located lower in the unsaturated zone (Cases A and C) and higher unsaturated zone Kds 

37 leading to slower transport rates or plumes located higher in the unsaturated zone (Cases B 

38 and D) after the same amount of time has elapsed. The Kd is used to calculate the leach rate 

39 using Equations 3 and 4. The RESRAD-ONSITE code uses the unsaturated zone distribution 

40 coefficient to calculate travel times from the contaminated zone to the saturated zone.

41

16 Additional source term options are available in RESRAD-OFFSITE 4.0 (NRC, 2020b). 

17 RESRAD-OFFSITE allows specification of the fraction of residual radioactivity that is releasable over time.
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1

2 Figure I.8 Conceptualization Modeled by RESRAD-ONSITE (ANL, 1993)

3

4 Figure I.9 RESRAD Near-Field Leaching and Transport Model (from RESRAD Training 

5 Workshop)
I-65

1 The RESRAD-ONSITE code calculates a leach rate based on a user-specified Kd for the 

contaminated zone.18 2 Alternatively, the user may specify (1) time since material placement and 

3 groundwater concentrations, (2) solubility limits, or (3) a leach rate from which contaminated 

4 zone Kds can be back-calculated for individual radionuclides, or allows for the correlation of 

plant to soil coefficient to Kd.

19 5 Only one of the methods can be used and, if more than one 

6 user-specified approach is indicated, the code selects the approach first specified in the order 

7 listed above. Ultimately, all options use the same approach to characterizing the source term 

8 from the contaminated zone, based on Kd and a calculated leach rate. 

9

Li =

I

θ

(cz) x To x Rdi

(cz)

10 Equation 3

11 Where

12 I = infiltration rate (m/y),

θ

(cz)

13 = volumetric water content of the contaminated zone,

14 To = initial thickness of the contaminated zone (m), and

Rdi

(cz)

15 = retardation factor in the contaminated zone for radionuclide i (dimensionless).

16

17 The retardation factor is calculated as follows:

Rdi= 1 +

ρb

θ

Kdi18 Equation 4

19 Where

ρb= bulk soil density (g/cm3 20 ),

Kdi= distribution coefficient for the ith principal radionuclide (cm3 21 /g), and

22 θ = volumetric water content (dimensionless).

23

24 The Kd is often one of the single most important parameter values in dose modeling, as it 

25 dictates the release rate of radionuclides out of the source or contaminated zone. Therefore, 

26 discussion on the importance of this parameter used in source term modeling is warranted. To 

27 estimate release rates and concentrations of radionuclides in groundwater, an analyst needs 

28 some means of quantifying the relationship between dissolved contaminants and the solid 

29 phase through which the contaminants move. In many cases, dose analysts resort to the use of 

Kds,

20 30 because the Kd model is readily available in commonly used contaminant transport 

31 models, and data are typically lacking to perform more complex modeling.

32

33 Simply put, Kds characterize the partitioning of contaminants between the solid and aqueous 

34 phases. The solid phase concentration is linearly related to the aqueous phase concentration 

35 by the Kd. Higher Kds indicate a greater degree of partitioning or “sorption” of a contaminant to 

36 solid materials. The term “sorption” is used when the actual mechanism of chemical removal 

37 from the aqueous phase is unknown. For example, sorption can refer to ion exchange, in which 

38 ions are loosely associated with a surface through differences in electrical charge, typically on 

39 porous materials with fixed charges, such as clays, zeolites, and organic material. Sorption also 

40 refers to surface complexation (more properly termed adsorption) in which the ion in solution 

41 reacts with ions on a solid’s surface in a way that is analogous to complex forming reactions that

18 RESRAD-ONSITE allow for separate specification of Kds for the unsaturated and saturated zones. 

19 The user has the option of toggling this option on or off.

20 Distribution coefficient and Kd are used interchangeably in this desk guide.
I-66

1 occur in solution. In both of these cases, sorption is subject to influences by the solution 

2 chemistry, notably pH and ionic strength. Changes in solution chemistry can, therefore, lead to 

3 a reversal in sorption and desorption. This is a significant consideration in long-term 

4 assessments, where water chemistry cannot necessarily be expected to remain constant, 

5 potentially resulting in sorption of a contaminant for a while, and then its release as water 

6 chemistry changes. Therefore, the potential for changes in geochemical conditions is an 

7 important consideration in decommissioning dose modeling.

8

9 Owing to its simplicity, the Kd model is widely applied in contaminant transport codes. Any 

10 computer code used to simulate the transport of a conservative, nonreactive constituent can be 

11 easily modified to solve the transport equation for a reactive (sorbing) constituent, if the 

12 partitioning coefficient is assumed to be constant (Glynn 2002)). It is expected that, for the vast 

13 majority of decommissioning sites, use of a Kd model is adequate. However, the selection of an 

14 appropriate value or set of values to use in performance assessment modeling can be a 

15 challenging task. Furthermore, the Kd model inherently assumes certain conditions that may not 

16 be valid: linearity of the sorption isotherm, equilibrium conditions, and the existence of only a 

17 single aqueous and solid phase species (simple and invariant sorption). If saturation of sorption 

18 sites, rapid transport, or variable chemical conditions is a concern, an analyst should carefully 

19 evaluate the appropriateness of using a Kd model to evaluate site risk. 

20 If the Kd appears to be inadequate to describe the release rate of radionuclides from the 

21 contaminated zone, RESRAD-ONSITE also allow editing of a release rate table that provides 

22 the user with additional flexibility in altering the source term model used by the code. In addition 

23 to the source term model offered in RESRAD-ONSITE, RESRAD-OFFSITE has the ability to 

24 delay release and to consider the fraction of the residual radioactivity that is releasable over 

25 time (e.g., to simulate the use of containers or engineered barriers). The RESRAD-OFFSITE 

26 code also provides options to account for first order rate controlled release with transport 

27 through the contaminated zone, as well as an instantaneous equilibrium desorption release 

28 option (see NUREG/CR-7127, “New Source Term Model for the RESRAD-OFFSITE Code 

29 Version 3,” issued December 2013 (NRC, 2013) and NUREG/CR-7189, “User’s Guide for 

30 RESRAD-OFFSITE,” issued April 2015 (NRC, 2015b) for additional details). RESRAD31 OFFSITE 4.0 was released in early 2020 and offers additional source term models including 

32 solubility-limited and diffusive release models (NRC, 2020a; and NRC, 2020b). If a more 

33 complex source term is needed than the first order leach rate models used in DandD and 

34 RESRAD-ONSITE, the licensee may consider use of a more complex model offered in 

35 RESRAD-OFFSITE, or other codes such as DUST-MS (Sullivan, 1996).

36 As a result of the NRC staff review of dose modeling to support decommissioning and 

37 performance assessment modeling to support low-level waste disposal, a number of 

38 observations and recommendations are noted with respect to the use of Kds in the abstraction of 

39 the source term and groundwater models. Table I.8 lists key considerations.

40
I-67

1 Table I.8 Distribution Coefficient Model and Model Simplification Considerations

Category Consideration or Recommendation

Model Extrapolation of experimental data to fit a linear Kd model could lead to 

significant inaccuracies in concentration and dose projections. For 

example, if experiments are conducted at relatively low 

concentrations, the calculated Kd may not be appropriate for higher 

concentrations. Saturation of sorption sites at higher concentrations 

may lead to nonlinearity in the sorption isotherm (i.e., less sorption 

may occur at higher concentrations). In another example, if 

experiments are conducted at relatively high concentrations, solubility 

limits may be exceeded and calculated Kds may be overestimated.

A particular Kd value or the Kd model, in general, may be inappropriate 

if nonequilibrium conditions are expected in the field. For example, 

under certain conditions (e.g., transport rates are rapid relative to 

sorption reaction rates, as may be the case under conditions of rapid 

fracture flow), sorption may be overestimated and dose/risk 

underestimated. For batch experiments, Kds may increase with 

equilibration time. However, if nonequilibrium conditions occur in the 

field, doses and risk could be significantly underestimated. 

Experiments used to derive site-specific Kds should be representative 

of field conditions. Column experiments could be run to study the 

impact of pore water velocity on sorption rates.

Use of a single Kd will likely be inappropriate under variable 

geochemical conditions. Methods to overcome this problem include 

use of spatially and temporally variable Kds, Kd lookup tables, or the 

establishment of functional relationships between Kd and other key 

parameters.

Model 

Simplifications

Averaging Kds of different species of the same radioisotope is 

generally not appropriate and can lead to significant errors in modeling 

results. In limited cases, the benefits of Kd averaging may outweigh 

the costs (costs such as calculation or decision error). However, 

adequate justification should be provided to support the approach, and 

reviewers should carefully evaluate the justification.

In some cases, errors associated with the averaging of Kds of different 

materials for a single species may be acceptably low. However, if 

differences in material Kds are significant, peak contaminant 

breakthrough times could be markedly different and lead to large 

inaccuracies in modeling predictions, particularly for relatively shortlived radionuclides. Adequate justification should be provided for any 

material Kd averaging approach and should include comparisons of 

results for the full range of parameter space being simulated. 

Inaccuracies related to timing of peak dose should be quantified and

clearly communicated.

2

3

4
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1 Groundwater Models

2

3 The RESRAD-ONSITE generic conceptual groundwater model assumes one or more horizontal 

4 homogeneous strata for the unsaturated zone. Transport in the unsaturated zone is assumed to 

5 result from steady-state, constant vertical flow, with equilibrium adsorption, and decay, but no 

6 dispersion. The RESRAD-ONSITE code has two different ways of modeling radionuclides once 

7 they reach the saturated zone. In the mass-balance approach, radionuclides entering the 

8 saturated zone are assumed to be instantaneously and uniformly distributed over a constant 

9 volume equivalent to the volume of water removed by the hypothetical well (as long as the 

10 pumping rate is larger than the rate of leachate entering the groundwater—if not, no dilution is 

11 assumed to occur in the groundwater). For the mass-balance approach, radionuclides are 

12 assumed to enter a well pumping immediately beneath the contamination zone. The mass13 balance approach is very similar to the groundwater modeling approach in DandD. In the 

14 nondispersion approach, transport in the saturated zone is assumed to occur in a single 

15 homogeneous stratum, under steady-state, unidirectional flow, with a constant velocity, 

16 equilibrium adsorption, and decay. It assumes no dispersion; however, radionuclides are 

17 assumed to be diluted by clean water as a function of the assumed capture zone of the 

18 hypothetical well, in relation to the width of residual radioactivity and the depth of residual 

19 radioactivity, in relation to the depth of the hypothetical well (see Table I.9 below). Radioactive 

20 decay and equilibrium adsorption are assumed to occur for the nondispersion approach. 

21 Radionuclides are assumed to enter a well located at the immediate downgradient edge of the 

22 contamination zone. For the nondispersion model, the calculated width of the effective pumping 

23 zone could be a factor of 2 larger than what one would predict from a steady-state capture zone 

24 analysis; this could lead to a slight overestimation in the amount of dilution (NRC, 1999d). 

25 Table I.9 Four Cases Evaluated in RESRAD’s Nondispersion Model to Calculate Dilution

Case Dilution Factor Condition Condition Parameters

1 1 dr ≤ A / l ζ ≥ dw dr=effective pumping width (m)

A=contaminated zone area

l= length of the contaminated zone 

parallel to the hydraulic gradient

ζ=distance from the water table to the 

lower boundary of the contaminant 

plume

dw=distance of well intake below the 

water table

Uw = well pumping rate

2 A * I / Uw dr > A / l ζ < dw

3 A * I * dw / Uw * 

ζ

dr > A / l ζ ≥ dw

4 ζ / dw dr ≤ A / l ζ < dw

26

27 In determining which of these two conceptual models to use, the licensee should consider 

28 where the hypothetical well may be located (i.e., either at the center of the residual radioactivity 

29 or at the edge of the residual radioactivity), the relative half-life of the radioactivity, and the 

30 potential capture zone of the hypothetical well. Use of the nondispersion model will generally 

31 result in lower estimated doses. Both models assume that radionuclides enter the well as soon 

32 as they reach the water table. However, the nondispersion model, unlike the mass-balance 

33 model, calculates the time it takes for the peak concentration to occur after the initial 

34 breakthrough. Accordingly, the nondispersion model accounts for radioactive decay during the 

35 interval between the initial breakthrough and the arrival of the peak concentration. Generally, 

36 the amount of decay should be small unless the radionuclides have short half-lives and are 

37 retarded. In addition, unlike with the mass-balance model, no assumption is made for the 
I-69

1 nondispersion model that all radionuclides released from the contaminated zone are withdrawn 

2 through the well. Therefore, the nondispersion model may include additional dilution. The only 

3 way that dilution is not considered is if the expected capture zone of the hypothetical well is 

4 small in relation to the width and thickness of the residual radioactivity and the plume is deeper 

5 than the well intake depth. Because the nondispersion model will generally give a lower 

6 estimated dose than the mass-balance model, it is important for the licensee to justify the use of 

7 this model for the specific analysis. Use of the mass-balance approach should always be 

8 acceptable. In Equations 5 and 6, use of the nondispersion model should be acceptable, 

9 without additional justification, for modeling long-lived radionuclides (i.e., where radioactive 

10 decay is not important) when either one of the following conditions is met:

11

U

w

v d

w

A

len

I

len d

v w

•

















 and

12 Equation 5

13

14

U

w A

len

len d

w

v d

I

v

w



















 and

15 Equation 6

16

where Uw = pumpage rate from the well (m3 17 /y);

18 ν = groundwater darcy velocity (m/y);

A = area of residual radioactivity (m2 19 );

20 dW = depth of well intake below water table (m);

21 len = length of residual radioactivity parallel to groundwater flow (m); and

22 I = infiltration rate (m/y).

23 As a general rule, use of the nondispersion approach should be acceptable when the area of 

24 residual radioactivity is known to be larger than the assumed capture area of the hypothetical 

25 well. Assuming an essentially flat water table and steady-state conditions, the capture area of 

26 the hypothetical well can be calculated in Equation 7 as follows:
I-70

A

w

U

w

I

=

















1 Equation 7

2

where AW = area of well capture (m2 3 );

UW = pumpage rate from the well (m3 4 /y); and

5 I = infiltration rate (m/y).

6 The RESRAD-OFFSITE code provides a more complex model for simulating groundwater 

7 transport to offsite locations. The RESRAD-OFFSITE code can consider advection and 

8 dispersion in calculating groundwater concentrations at a receptor well. The groundwater 

9 transport model in RESRAD-OFFSITE considers 1-D advection (straight or curved flow path), 

10 along with 3-D dispersive transport in the saturated zone. Likewise, while only 1-D advection is 

11 considered in RESRAD-ONSITE, RESRAD-OFFSITE considers 1-D advection and 1-D 

12 dispersive transport in the unsaturated zone. Furthermore, the unsaturated zone, saturated 

zone, and contaminated zone21 13 can be subdivided into smaller zones to increase the accuracy 

of transport simulations.22 14

15

16 Both RESRAD-ONSITE and RESRAD-OFFSITE have the capability to consider variable 

17 transport rates of progeny created during transport in groundwater. The RESRAD-OFFSITE 

18 code has two groundwater transport algorithms: the first algorithm considers variable transport 

rates of parents and progeny, and the second algorithm models longitudinal dispersion23 19 . When 

20 either variable transport rates or dispersion is clearly dominant, the RESRAD-OFFSITE user 

21 should choose the transport algorithm that is most important. Specification of the most 

22 important transport mechanism is important to increasing computational efficiency. When both 

23 the longitudinal dispersion and the variable transport rates are important, the user has the 

24 option of subdividing the transport pathway into a number of subzones to more accurately 

25 simulate the transport of progeny in transport, although this approach may significantly increase 

26 computation times. Only the zone where the progeny atoms are created would not consider 

27 both processes (longitudinal dispersion and variable transport rates of parents and daughters).

28

29 Code developers benchmarked RESRAD-OFFSITE against RESRAD-ONSITE (C.E. Yu, 

30 “Benchmarking of RESRAD-OFFSITE: Transition from RESRAD (onsite) to RESRAD-OFFSITE 

31 and Comparison of the RESRAD-OFFSITE Predictions with Peer Codes,” issued May 2006) 

32 (Yu, 2006). Benchmarking was performed with an earlier version of the RESRAD-OFFSITE 

33 code (i.e., benchmarking was conducted before the 2007 release of RESRAD-OFFSITE 

34 Version 2.0). The results of the benchmarking exercises showed that RESRAD-OFFSITE could 

35 mimic the results of RESRAD-ONSITE when certain parameters were changed consistent with 

36 the RESRAD-ONSITE conceptual model (onsite dose). Notable differences between initial 

37 simulations run with RESRAD-ONSITE and RESRAD-OFFSITE included travel times to the 

38 point of compliance that were attributable to differences in the use of porosity in the transport 

39 calculations (i.e., effective porosity is used in RESRAD-ONSITE while total porosity is used in 

21 The capability to more accurately simulate transport through the contaminated zone through use of subzones was 

added in RESRAD-OFFSITE Version 3.0.

22 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 pertains to just the offsite capabilities and not the onsite dose calculations, unless otherwise stated.

23 Both RESRAD-OFFSITE transport algorithms account for transverse dispersion.
I-71

1 RESRAD-OFFSITE). Another noteworthy difference in results was observed for the water2 dependent pathways due to accumulation of radioactivity in soil from applying contaminated 

3 irrigation water that is considered in RESRAD-OFFSITE but is not considered in RESRAD4 ONSITE.

5

6 Surface Water Models

7 The generic conceptual model of the surface water pathway in RESRAD-ONSITE assumes that 

8 radionuclides are uniformly distributed in a finite volume of water within a watershed. 

9 Radionuclides in surface water are assumed to be diluted as a function of the size of the 

10 contaminated area in relation to the size of the watershed. Radionuclides are assumed to enter 

11 the watershed at the same time as the groundwater. Accordingly, no additional attenuation is 

12 considered as radionuclides are transported to the watershed. The RESRAD surface water 

13 conceptual model assumes that all radionuclides reaching the surface water are derived from 

14 the groundwater pathway. Thus, neither the transport of radionuclides overland from runoff nor 

15 additional dilution from overland runoff is considered. 

16 A more complex surface water model was added to RESRAD-OFFSITE 4.0 released in early 

17 2020 (NRC, 2020a; and NRC, 2020b). In addition to contributions of residual radioactivity from 

18 groundwater to surface water, this model considers contributions of residual radioactivity 

19 primary contamination and catchment areas to surface water from surface water runoff and 

20 erosion, and deposition of particulates from atmospheric transport. Various sources and sinks 

21 of residual radioactivity are also considered in this model. 

22 For example, the water balance considers the interception of groundwater by the surface water 

23 body, inflow of runoff and stream flow from the catchment into the surface water body, 

24 precipitation on the surface water body, evaporation from the surface water body, stream flow 

25 and extraction out of the surface water body, and infiltration from the surface water body into the 

26 groundwater aquifer. The sediment balance considers the influx of eroded material, removal of 

27 sediments out of the surface water body due to flow, settling of suspended sediment, and other 

28 processes. Sources of radioactivity to the surface water include 

29 • discharge from surface water, 

30 • eroded particulates from the primary contamination, 

31 • surface water runoff from the catchment (also includes particulates deposited in the 

32 catchment through the air pathway), and 

33 • particulates deposited directly on the surface water body through the air pathway

34 Sinks include mass removal due to stream flow, infiltration from the surface water body to the 

35 aquifer, and burial of radioactivity in deeper sediments, radiological decay, and other processes. 

36 Equilibrium partitioning of radionuclides between the water, the suspended sediments and the 

37 bottom sediments is also assumed and represents a removal mechanism from the aqueous 

38 phase (Yu et al., 2019).

39 Air Models

40 The generic conceptual model of the air pathway in RESRAD-ONSITE use a constant mass 

41 loading factor and area factor to model radionuclide transport. The area factor, which is used to 
I-72

1 estimate the amount of dilution, relates the concentration of radionuclides from a finite area 

2 source to the concentration of radionuclides from an infinite area source. It is calculated as a 

3 function of particle diameter, wind speed, and the side length of a square-area source. The 

4 conceptual model assumes a fixed particle density, constant annual rainfall rate, and constant 

5 atmospheric stability. No radioactive decay is considered. Chang, et al. (1998) provides more 

6 detail. Tritium and C-14 gases are assumed to be uniformly mixed in a constant volume of air 

7 above the contaminated zone. RESRAD-ONSITE does not model the transport of H-3 or C-14 

8 as particulates in the air. 

9

10 A more complex atmospheric dispersion model is utilized in RESRAD-OFFSITE compared to 

11 RESRAD-ONSITE. The air dispersion model in RESRAD-OFFSITE uses a polar grid with 16 

12 sectors to specify wind direction. Meteorological inputs include joint frequency of wind speed 

13 and stability class for each of the 16 sectors. Pasquill-Gifford, Briggs rural and Briggs urban 

14 dispersion coefficients can be selected for dispersion model coefficients. Sector-averaged, 

15 ground-level air concentrations are calculated by the code for various distances down-wind of 

16 the source. Removal processes include dry and wet deposition, as well as radiological decay. 

17

18 Table I-10 provides summary information on the attributes and information requirements of the 

19 DandD versus RESRAD-ONSITE and RESRAD-BUILD computer codes. Additional information 

20 on the differences between the codes can be found in NUREG/CR-5512, Volume 4 (NRC, 

21 1999d).

22

23 I.5.4 Use of Codes and Models Other than DandD and the RESRAD Family of Codes

24 The NRC staff should provide flexibility for the possible use of other codes and models selected 

25 by licensees. However, less common codes, specifically those developed by users, may 

26 require more extensive NRC staff review and verification. In this context, the NRC may review 

27 the following pertinent aspects when using other less common codes:

28 • scope of code application and applicability to the concerned site

29 • extensive review of the generic code selection criteria listed previously

30 • review of the mathematical formulation of the associated models and the selected dose 

31 conversion factors

32 • review of the conceptual model, including the source term model, used in the code, and 

33 compatibility with site conditions

34 • review of code performance and comparison with commonly used and verified codes

35 • review of code capability on the handling of default pathways and consistency in 

36 selecting default parameters (e.g., occupancy, behavioral, and metabolic parameters)

37 • detailed review of code/model documentation and updates for code/model modifications, 

38 including QA/QC reviews
Table I.10 Summary Attributes and Information Needs of DandD; and RESRAD-ONSITE and RESRAD-B

DandD

RESRAD-ONSITE and RESRADOFFSITE

RE

General • Can be used for screening or sitespecific analyses for buildings and soils.

• Code QA/QC has already been

performed.

• Used for site-specific analyses for soils.

• Code QA/QC has already been performed.

• Used for si

buildings.

• Code QA/Q

performed.

Conceptual 

and 

Mathematical 

Models

• Conceptual site model must be

compatible with the assumptions in the

code. For soils, the residual radioactivity

is assumed to be associated with

surface soils, and no radioactivity is

assumed to be present in the subsurface

or in groundwater. A list of site

conditions that may be incompatible with

DandD is listed in Table I.5.

• For buildings, residual radioactivity is

assumed to be associated with building

surfaces.

• DandD cannot easily handle multiple

sources or sources with odd geometry

(e.g., piping, equipment, and sewer

lines).

• The conceptual models built-into RESRADONSITE is more robust compared to DandD.

RESRAD-ONSITE can handle various

exposure scenarios, subsurface residual

radioactivity in soils, buried radioactivity, and

better assess the dose associated with

smaller areas of elevated residual

radioactivity. Site conditions that are

incompatible with the conceptual model in

RESRAD-ONSITE are listed in Table I.7.

RESRAD-OFFSITE has all the capabilities of

RESRAD-ONSITE (i.e., can simulate a

RESRAD-ONSITE run) and can handle more

complex source terms, groundwater and

surface water transport.

• The concep

RESRAD-B

compared t

can handle 

source/rece

volumetric 

afforded by 

more comp

indoor air q

assess the 

smaller are

radioactivit

Parameter

Requirements

• Screening Assessments—minimal

parameter information is needed to use

screening code (list of contaminants and

concentrations).

• Site-specific physical information is needed

to develop the model. Default parameters

should not be used without justification

(DandD default behavioral and metabolic

parameters can be used with minimal

justification). Default parameter distributions

may be used to perform probabilistic

sensitivity analyses. However, site-specific

information should be used to develop or

provide support for parameter distributions

for risk-significant parameters if a

probabilistic analysis will be used to

demonstrate compliance, particularly if risk

dilution is an issue (see Appendix Q for

additional information on risk dilution).

• Site-specifi

construct th

source info

parameters 

justification 

and metabo

with minima

• Default par

used to per

analyses. 

information 

or provide s

distribution

parameters 

will be used 

I-7

3
I-74

1 I.6 Criteria for Selecting or Modifying Input Parameter Values 

2 I.6.1 Introduction

3 Any analytical approach to dose assessment should involve the selection of appropriate values 

4 for input parameters. Each computer modeling code or other analytical method that a licensee 

5 may use should have its own suite of input parameters. Also, unless the licensee is performing 

6 a screening analysis, each site should likely have its own defining characteristics that should be 

7 incorporated into the dose assessment through the selection of input parameter values.

8 This section provides general guidelines for the NRC staff to consider in evaluating a licensee’s 

9 selection of values for input parameters. This section addresses three aspects of parameter 

10 value selection:

11 (1) selection of parameter values and distributions

12 (2) technical justification to support value selection

13 (3) evaluation of the impact of parameter selection on dose assessment results

14 The deterministic parameter set described in NUREG/CR–5512, Volume 1 (NRC, 1992), and 

15 implemented in DandD Version 1 has been superseded by the parameter set described in

16 NUREG/CR–5512, Volume 3 (NRC, 1999c) and implemented in DandD Version 2. DandD 

17 Version 1 did not support probabilistic analyses and used a default deterministic input 

18 parameter set. 

19

20 I.6.2 Issues in Modifying Parameters

21 In addressing the three aspects of parameter value selection identified above, several issues 

22 should be discussed. First is the distinction between screening analysis and site-specific 

23 analysis, with respect to parameter value modification. Second is the appropriateness of 

24 accepting default input parameter values in site-specific analyses. Third is the level of 

25 justification expected to support the selection of site-specific input parameter values. The NRC 

26 staff should consider these issues in evaluating a licensee’s dose assessment.

27 I.6.2.1 Screening Analyses versus Site-Specific Analyses

28 A licensee may perform a screening analysis to demonstrate compliance with the radiological 

29 criteria for license termination specified in 10 CFR Part 20, Subpart E. The screening analysis 

30 described in Chapter 5 of this volume requires that the licensee either (1) refer to 

31 radionuclide-specific screening values listed in the Federal Register (63 FR 64132 and 

32 64 FR 68395), or (2) use the latest DandD computer code. A licensee pursuing the screening 

33 option may find that implementation of the DandD code is necessary if radionuclides not 

34 included in the Federal Register listings should be considered.

35 The NRC staff should ensure that a licensee performing a screening analysis using the DandD 

36 code limits its parameter modification to identifying radionuclides of interest and specifying the 

37 radionuclide concentrations and verify that it has not modified any other input parameter values. 

38 The output file generated by DandD identifies all parameter values that have been modified. 

39 Modifying any input parameter value from a default value will constitute a site-specific analysis.

40
I-75

1 I.6.2.2 Default Values versus Site-Specific Values

2 DandD and many other computer codes used for dose assessment provide the user with default 

3 values for the input parameters. Often, the user need only select radionuclides to execute the 

4 code. This allows the user to quickly obtain results with very little time expended in developing 

5 input parameter sets. This is basically how DandD Version 2 was to be used for screening 

6 analyses.

7 While little user input or thought may be required for licensees to run common decommissioning 

8 dose modeling codes, which contain built-in default parameters, care should be taken to ensure 

9 that selected parameter values are adequately supported. Default parameter values may be 

10 inappropriate for site-specific conditions and the selected exposure scenario. When using an 

11 off-the-shelf computer code and its default parameters, the user agrees with (1) the conceptual 

12 model used by the computer code, (2) the exposure scenario, and (3) the process used to 

13 select the default parameters so that they are appropriate for the site being modeled. 

14 Users of computer codes should have an adequate understanding of the conceptual and 

15 numerical models built into the code and how parameter inputs affect the results. Analysts

16 should provide support for the selection of input parameters by (1) determining what 

17 characteristics of the modeled system the parameter represents (how the parameter is used in

18 the code), and (2) developing a value for the input parameter that is appropriate for both the 

19 system being modeled and for the conceptual and numerical models implemented by the code. 

20 In fact, many default parameter values in the computer code are simply “placeholders” for site21 specific data. Experience has shown that the availability of default values for input parameters 

22 can result in the user performing a “site-specific” analysis characterized by the modification of 

23 parameter values for which site data are readily available and the acceptance of default 

24 parameter values for the remaining parameters, without an adequate understanding of the 

25 implications of accepting the default parameters on the results. For site-specific analyses, the 

26 NRC staff suggests that the licensee justify using both the model and the default parameters. 

27 The level of justification needed for each parameter value is not necessarily the same. 

28 Section I.7 discusses the need to conduct uncertainty and sensitivity analyses as a method of 

29 focusing licensee and NRC staff resources on the parameters and pathways most important to 

30 dose. 

31 The NRC staff has reviewed the DandD Version 2 and RESRAD-ONSITE, Version 6, computer 

32 codes and associated parameter distributions and considered them appropriate for dose 

33 assessments conducted to demonstrate compliance with dose-based standards in 

34 10 CFR Part 20, Subpart E. Later versions of the code may also be appropriate for use with 

35 appropriate verification and validation, as well as QA documentation. Use of approved 

36 parameter distributions (1) promotes consistency among analyses (where appropriate), 

37 (2) helps to focus licensee and NRC staff resources on parameters considered risk-significant

38 based on dose assessment results, and (3) facilitates the NRC’s review of the licensee’s dose

39 assessments. Therefore, most licensees can use the code and its default parameter

40 distributions with little justification. While default parameter distributions are approved for use in

41 DandD and RESRAD-ONSITE, this does not mean that the default parameter distributions can

42 be used in another computer code with no justification. Additionally, for especially risk43 significant parameter values, which may be determined through probabilistic sensitivity analysis,

44 the licenses should provide additional justification for its selection of a deterministic value for the

45 risk-significant parameter or use of the default parameter distribution to demonstrate compliance

46 with the dose standards in 10 CFR Part 20, Subpart E.
I-76

1 I.6.2.3 Justifying Site-Specific Parameter Values

2 The NRC reviewer should evaluate whether a licensee submitting a site-specific dose 

3 assessment has demonstrated that all parameter input values are appropriate for the site being 

4 modeled. However, this does not require the licensee to submit a detailed analysis to support 

5 the values selected for each and every input parameter. Instead, the level of justification 

6 required should be based on the parameter classification and should be commensurate with the 

7 significance of the parameter relative to the dose assessment results, as evaluated through 

8 sensitivity analyses. The sensitivity analyses should reflect the relative significance of exposure 

9 pathways. Note that the relative significance of exposure pathways may change as parameters 

10 are modified. 

11 Dose assessment input parameters may be generally classified as behavioral, metabolic, or 

12 physical. Behavioral parameters (B) collectively describe the receptor—the exposed individual 

13 for whom the dose received is being assessed. The values selected for these input parameters 

14 should depend on the behavior hypothesized for the exposed individual. Metabolic parameters 

15 (M) also describe the exposed individual but generally address involuntary characteristics of the

16 individual. Physical parameters (P) collectively describe the physical characteristics of the site

17 being modeled. These would include the geohydrological, geochemical, and meteorological

18 characteristics of the site. The characteristics of atmospheric and biospheric transport up to, but

19 not including, uptake by, or exposure of, the dose receptor would also be considered physical

20 input parameters.

21 There is always uncertainty associated with the behavior of a hypothetical receptor. For this 

22 reason, the licensee may accept a generically defined receptor for its analysis. The generically 

23 defined receptor is the “average member of the critical group.” The characteristics of this 

24 exposed individual and the criteria for modifying the characteristics for a site-specific analysis 

25 are discussed in Section I.3 of this appendix. The licensee may use DandD default values for 

26 the behavioral and metabolic parameters, with limited justification, if the values are consistent 

27 with the generic definition of the average member of the critical group and if the screening group 

28 is reflective of the exposure scenario. Table I.11 lists key default behavioral and metabolic 

29 parameters used in DandD. NUREG/CR-7267 also lists key default behavioral and metabolic 

30 parameters in Tables 6-1 and 6-3; and Appendix B, Tables B-1 and B-26, for the resident farmer 

31 and building occupancy scenarios, respectively.

32 In site-specific analyses, all efforts should be made by the licensee to use site-specific 

33 information for important physical parameters. “Site-specific” in this context includes 

34 (1) information directly related to the site, (2) information characterizing the region that is

35 consistent with site conditions, and (3) generic information that is consistent with the specific

36 geohydrologic conditions at the site (e.g., consistent with the surface-soil unsaturated-zone soil

37 classifications). The justification for site-specific physical parameter values should demonstrate

38 that the site-specific values selected are consistent with the known or expected characteristics

39 of the physical site being modeled. The level of justification should be based on the significance

40 of the parameter to the results of the dose assessment. The licensee should evaluate the

41 significance through sensitivity analyses (see Section I.7 of this appendix). Because of the

42 importance of groundwater, the NRC staff should verify that the licensee uses site-specific

43 values for all physical parameters (or parameter ranges) related to geohydrologic conditions. If

44 a licensee relies on the DandD default parameter distributions for especially risk-significant

45 parameters such as partition coefficients and soil-to-plant transfer factors, the reviewer should

46 evaluate whether the default parameter distributions are consistent with known or expected

47 conditions at the site to ensure that the doses are not significantly underestimated.
I-77

1 Table I.11 Key DandD Deterministic Behavioral and Metabolic Parameters Mapped to 

2 RESRAD-ONSITE and -BUILD Parameters

DandD Parameter^ DandD Value

RESRADONSITE/BUILD 

Parameter

RESRADONSITE/BUILD 

Value

Notes

Residential Exposure Scenario (RESRAD-ONSITE)

Indoor exposure 

period 0.66 Indoor time fraction 0.5 DandD lists as 240 days out of the year

Outdoor exposure 

period 0.11 **Outdoor time 

fraction 0.25

DandD lists as 40.2 days 

out of the year.

Gardening 

exposure period 0.01 DandD lists as 2.92 days out of the year

Indoor breathing 

rate 0.9 m3

/hr

&Inhalation rate 8400 m3

/y

DandD 7900 m3/y

Outdoor breathing 

rate 1.4 m3

/hr DandD 12,000 m3/y

Gardening 

breathing rate

1.7 m3/hr DandD 15,000 m3/y

Leafy vegetable 

consumption 21.4 kg/y Leafy vegetable consumption 14 kg/y

RESRAD reduces the 

consumption rate based on 

contaminated fraction.

Root vegetable 

consumption 44.6 kg/y

Fruit, vegetable, 

and grain 

consumption

160 kg/y

RESRAD reduces the 

consumption rate based on 

contaminated fraction

Fruit consumption 52.8 kg/y

RESRAD reduces the 

consumption rate based on 

contaminated fraction

Grain 

consumption 14.4 kg/y

RESRAD reduces the 

consumption rate based on 

contaminated fraction

Beef consumption 39.8 kg/y Meat and poultry 

consumption 63 kg/y

RESRAD reduces the 

consumption rate based on 

contaminated fraction

Poultry 

consumption 25.3 kg/y

Milk consumption 233 L/y Milk 

consumption 92 L/y

RESRAD reduces the 

consumption rate based on 

contaminated fraction

Egg consumption 19.1 kg/y NA NA

Fish consumption 20.6 kg/y Fish 

consumption 5.4 kg/y

RESRAD also includes a 

seafood consumption rate 

of 0.9 kg/y.

Water 

consumption 1.26 L/d Drinking water intake 510 L/y DandD 460 L/y

Indoor shielding 

factor 0.55 Shielding factor 0.7

This parameter may also 

be considered a physical 

parameter

3

4
I-78

1 Table I.11 Key DandD Deterministic Behavioral and Metabolic Parameters Mapped to 

2 RESRAD-ONSITE and -BUILD Parameters (cont.)

3

DandD Parameter DandD Value

RESRADONSITE/BUILD 

Parameter

RESRADONSITE/BUILD 

Value

Notes

Building Occupancy Exposure Scenario (RESRAD BUILD)

Time in building 45 hr/wk Indoor fraction 

(time) 0.5 DandD 0.27 time fraction

Breathing rate 1.4 m3/hr Breathing rate 18 m3/d DandD 33.6 m3/d

Ingestion rate 1.1E-05 

m2/hr Ingestion rate† 0.0001 m2

/hr Loose ingestion 

rate

1.1E-04 

m2/hr

4 _________________________

5 ^ NUREG/CR-5512, Volume 3 (NRC, 1999c), Table 6.87, presents a more exhaustive list of DandD parameters for the residential 

6 scenario. Average values of behavioral parameters (marked with a “B”) and metabolic parameters (marked with an “M”) are 

7 provided in Table 6.87.

8 **The outdoor and garden exposure periods must be summed to compare to the single RESRAD-ONSITE outdoor time fraction 

9 (0.12).

& The individual DandD breathing rates must be time weighted and summed to calculate an effective breathing rate of 8,600 m3 10 /hr

11 for use in RESRAD-ONSITE.

†There is only one ingestion rate in RESRAD-BUILD. The default loose fraction in DandD is 0.1 and 0.1 x 1.1x10-04 m2/hr=1.1x10 12 -05

m2 13 /hr.

14

15 I.6.3 Input Parameter Data Sets

16 I.6.3.1 DandD Default Probabilistic Parameter Set

17 Probabilistic analyses using the DandD computer code were performed to establish the 

18 screening values for surface-soil residual radioactivity that were published in the Federal 

19 Register in December 1999 (64 FR 68395). In performing these screening analyses, data were 

20 compiled for over 600 input parameters and reviewed by the NRC staff. These data are 

21 discussed in great detail in NUREG/CR–5512, Volume 3 (NRC, 1999c), and are directly 

22 incorporated into DandD Version 2. These data form the reference input parameter set for 

23 probabilistic analysis using DandD. The user is referred to NUREG/CR–5512, Volume 3, and 

24 the current version of the DandD computer code for information on the basis for the current 

25 default parameter distributions.

26 The DandD computer code may be used to evaluate radiological doses for two exposure 

27 scenarios: (1) the building occupancy exposure scenario and (2) the residential exposure 

28 scenario. A detailed discussion of these scenarios and the associated exposure pathways 

29 appears in NUREG-1549 (NRC, 1998d) and NUREG-CR/5512, Volume 1 (NRC, 1992).

30 As stated above, the licensee may use the default deterministic behavioral and metabolic 

31 parameters from NUREG/CR-5512, Volume 3 (NRC, 1999c), or the current version of the 

32 DandD computer code, with limited justification (see Table I.11). The justification should 

33 examine how the licensee’s exposure scenario is consistent with the generic exposure scenario 

34 used in DandD. Similarly, a licensee may use the parameter distribution for a physical 

35 parameter, provided it justifies why the parameter distribution is consistent with the site 

36 conditions. 
I-79

1 Note that licenses may not use deterministic physical parameter values without substantial 

2 justification (including sensitivity and uncertainty analyses).

3 I.6.3.2 DandD Default Deterministic Parameter Set

4 Several default parameter sets support deterministic analyses using the DandD code. 

5 NUREG/CR-5512, Volume 1 (NRC, 1992), initially presented both the conceptual and 

6 mathematical foundation of the DandD code and the deterministic values for many input 

7 parameters. Volume 3 of NUREG/CR-5512 (NRC, 1999c) incorporated much of the parameter 

8 information from Volume 1 in developing the default probabilistic input parameter set, making 

9 corrections and updating values as necessary. Therefore, a licensee should not refer to 

10 NUREG/CR-5512, Volume 1, as a primary source for a default deterministic parameter set. 

11 Additionally, it should not use the DandD Version 1 default parameter set as a reference data 

12 set for any parameters.

13 Licensees may perform deterministic analyses using DandD (Version 2 or later). This would 

14 require them to change all parameter distribution types to “constant” and specify a single value 

15 for each parameter. However, the NRC staff does not intend to provide a default deterministic 

16 input parameter set to be used in conjunction with DandD. Also, a licensee intending to support 

17 decommissioning activities with deterministic dose assessments should ensure that there is

18 sufficient information to support its demonstration of compliance (e.g., data to support 

19 parameter selection for important parameters identified through sensitivity analysis, as 

20 described in Section I.7 of this appendix).

21 I.6.3.3 RESRAD Default Probabilistic Parameter Set

22 The most recent versions of the RESRAD-ONSITE, RESRAD-OFFSITE, and RESRAD–BUILD 

23 computer codes include the option to perform probabilistic dose assessments. The RESRAD 

24 team at Argonne National Laboratory worked with the NRC staff to develop default input 

25 parameter distributions that may be used to perform probabilistic dose assessments with the 

26 RESRAD-ONSITE and RESRAD–BUILD codes. These default probabilistic input parameter 

27 distributions are documented in NUREG/CR-6697 (NRC, 2000c). Guidance for performing 

28 probabilistic dose assessments with RESRAD-OFFSITE can be found in NUREG/CR-7189 

29 (NRC, 2015b). Recently updated and newly developed RESRAD-ONSITE, RESRAD-OFFSITE, 

30 and RESRAD-BUILD parameter distributions that may be of use to licensees are available in 

31 NUREG/CR-7267.

32 I.6.3.4 RESRAD Default Deterministic Parameter Set

33 As a set, the RESRAD default deterministic parameters are not considered to be acceptable for 

34 performing dose assessments in support of decommissioning. A licensee may use the default 

35 probabilistic parameter distributions described in the preceding section as a starting point for its 

36 analyses. However, the licensee should identify and justify those parameters most important to 

37 dose and provide information to support the selected values. Care should be taken to avoid use 

38 of overly broad distributions that lead to risk dilution (see Appendix Q). ANL/EAIS-8, “Data 

39 Collection Handbook to Support Modeling the Impacts of Radioactive Material in Soil” (1993), 

40 which was used to support development of parameter values and distributions in the RESRAD 

41 family of codes was updated in 2015 and re-designated as ANL/EVS/TM-14/4. ANL/EVS/TM42 14/4 provides updated information on parameter definitions, typical ranges, variations, and 

43 measurement methodology. A summary of the development of the models in the RESRAD 

44 family of codes and associated parameters is found in NUREG/CR-7267 (2020a). NUREG/CR-
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1 7267 performs probabilistic analysis using parameter values and distributions found in 

2 ANL/EVS/TM-14/4, as well as other more recent information (e.g., updated dose conversion 

3 factors), for several radionuclides. Exposure scenarios evaluated include the following: the 

4 resident farmer using RESRAD-ONSITE, building occupancy using RESRAD-BUILD, and an 

5 offsite resident scenario with water transport and offsite resident scenario with air transport 

6 using RESRAD-OFFSITE 4.0. Default behavioral and metabolic parameter values from DandD 

7 are used and key parameters and pathways are identified. The methodology and information 

8 provided in NUREG/CR-7267 may be useful to licensees performing probabilistic sensitivity 

9 analysis to identify key parameters and pathways and providing support for parameters and 

10 distributions important to dose.

11 I.6.3.5 Input Data Sets for Other Computer Codes

12 A licensee may choose to use a computer code or analytical approach other than DandD or the 

13 RESRAD family of codes to perform the dose assessment in support of decommissioning. 

14 Each code or analytical approach should have a unique set of input parameters. However, 

15 there will likely be some input parameters that are also included in the DandD input parameter 

16 set.

17 The NRC staff should verify that a licensee lists all input parameters required in its analysis. For 

18 each parameter, the licensee should provide a discussion similar to the one in 

19 NUREG/CR-5512, Volume 3 (NRC 1999c), Chapters 5 and 6. The discussion should include 

20 the parameter name, a description of the parameter, a discussion of how the parameter is used 

21 in the dose assessment model, and the licensee’s classification of the input parameter 

22 (i.e., behavioral, metabolic, or physical). For the parameters being represented by constant 

23 values, the licensee should include the range of appropriate values for the parameter, the single 

24 value selected for the parameter, and the basis for the range and selected value, including 

25 references. The level of justification to be provided in the basis should be based on the 

26 classification of the parameter (i.e., behavioral, metabolic, or physical) and the relative 

27 significance of the parameter in the dose assessment.

28

29 For input parameters classified as “behavioral” or “metabolic,” the NRC staff should verify that

30 the licensee specifies values that are consistent with the default screening values for the DandD 

31 behavioral and metabolic parameters, as long as the definition of the critical group has not been 

32 modified (see Table I.11). Consistency may depend on the conceptual and numerical models 

33 underlying the code being used and the manner in which the parameters are used in the 

34 models. Using consistent behavioral and metabolic parameter values for the default critical 

35 group may support a relatively standardized definition of the average member of the critical 

36 group among analyses. The basis the licensee provides for these parameters should identify 

37 the comparable DandD parameters and discuss any adjustments necessary to accommodate 

38 differences between DandD and the code or analytical method being used. 

39 For the input parameters the licensee classifies as physical, with exceptions noted below, the 

40 NRC staff should verify that the licensee uses site-specific values when available. The licensee 

41 should include the soil classification for all soil units and specify consistent values for all 

42 geohydrologic parameters. For geochemical parameters, such as partition coefficients, the 

43 licensee may rely on DandD default probabilistic distributions for sensitivity analysis, as long as 

44 justification is provided to demonstrate that the ranges are consistent with geochemical 

45 conditions at the site. However, overly broad distributions for parameters that primarily affect 

46 the timing of peak dose, such as Kds, can lead to risk dilution (lowering of the peak of the mean 

47 dose) as discussed in Appendix Q. Licensees and reviewers should evaluate the potential for 
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1 risk dilution if the peak of the mean dose is used as the metric to demonstrate compliance with 

2 radiological criteria for license termination in probabilistic analyses. The licensee may need to 

3 modify the default parameter distributions to ensure consistency with site conditions. 

4 Additionally, it is important to note that the distributions may not be applicable to codes other

5 than DandD. For meteorological parameters, the licensee should use values that are based on 

6 applicable site or regional data. For physical parameters related to atmospheric and biospheric 

7 transport, the licensee may accept DandD default parameter ranges with minimal justification for 

8 sensitivity analysis, using NUREG/CR-5512, Volume 3 (NRC, 1999c), as a starting reference 

9 point. Physical parameters related to biosphere transport would include parameters such as 

10 crop yields, animal ingestion rates, transfer factors, and crop growing times. The NRC staff 

11 should evaluate whether the justification provided by the licensee demonstrates that the default 

12 values are consistent with conditions at the site. Generally, Kds and transfer factors are 

13 particularly risk significant. If the sensitivity analysis reveals the importance of these parameter 

14 values, care should be taken to select a deterministic value that is supported by site-specific 

15 data or to ensure that parameter distributions used in probabilistic analyses to demonstrate 

16 compliance with the performance objective are not expected to underestimate the potential 

17 dose.

18 I.6.3.6 Internal and Direct Exposure Dose Conversion Factors 

19 The NRC staff should review the dose conversion factors for inhalation and ingestion, to ensure 

20 that the factors used are those developed by EPA, published in Federal Guidance Report 

21 No. 11 (EPA, 1988). Similarly, the review should ensure that EPA’s external dose factors, 

22 although they may correct for actual area, published in Federal Guidance Report No. 12 

23 (EPA, 1993), are used or another appropriate code such as Microshield is used. These dose 

24 factors were selected to ensure that the dosimetry models used in deriving these factors are 

25 consistent with NRC regulations in 10 CFR Part 20.

26 More recent dosimetry methods may be acceptable for use and will be evaluated on a case-by27 case basis. For example, licensees may request an exemption from 10 CFR Part 20 to use 

28 more recent dose conversion factors (e.g., ICRP 72). ICRP 72 provides age dependent 

29 committed effective dose conversion factors for ingestion and inhalation of radioactivity. If a 

30 licensee proposes to use ICRP 72, the selection of exposure scenarios and critical groups 

31 should take into account differences in potential exposure due to age. Licensees should avoid 

32 “picking and choosing” dosimetry methods for different radionuclides (e.g., Federal Guidance 

33 Report No. 11 for six radionuclides and current international dose conversion factors for three 

34 radionuclides).

35 I.6.4 Recommended Approach to Parameter Modification 

36 Any analysis that does not meet the conditions of a screening analysis may be considered a 

37 site-specific analysis. This will include all analyses using the DandD computer code, where one 

38 or more input parameter values have been modified from default ranges (or values for 

39 behavioral and metabolic parameters), as well as analyses using analytical methods or 

40 computer codes other than DandD. 

41 I.6.4.1 Modifying the DandD Default Probabilistic Parameter Set

42 A reviewer should expect that a licensee who is modifying parameter values for a site-specific 

43 analysis using DandD is cognizant of the following:
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1 (1) what the parameter represents

2 (2) how the parameter is used in the DandD code

3 (3) the basis for the default parameter value

4 (4) which parameters are physically or numerically correlated

5 NUREG/CR-5512, Volumes 1–3, describes in detail what each parameter is intended to 

6 represent. Volume 1 (NRC, 1992) provides the original parameter definitions but has been 

7 superseded by Volume 3 (NRC, 1999c) for parameter values. Volume 1 also provides the 

8 mathematical formulations underlying the DandD code that should allow the user to 

9 (1) understand how each parameter is used and the implication of parameter modification on

10 the resulting calculated dose, and (2) identify numerical correlations among parameters.

11 Volume 2 (the DandD user’s manual) redefines several of the input parameters and

12 mathematical formulations based on implementing the Volume 1 methodology in the DandD

13 computer code. Finally, Volume 3 contains a detailed discussion of most input parameters,

14 allowing the user to fully understand the basis for the default ranges. Volume 3 includes a

15 parameter description and a discussion of how parameters are used in the code, a review of the

16 information sources on which the default values are based, a discussion of uncertainty in the

17 default parameter values, and insight into the selection of alternative parameter values. The

18 DandD user performing site-specific analyses should be cognizant of the information provided in

19 the three volumes of NUREG/CR-5512.

20 A licensee may modify DandD behavioral (B) and metabolic (M) input parameter values for the 

21 building occupancy and residential exposure scenarios to reflect the characteristics of the 

22 average member of a site-specific critical group. NUREG/CR-5512, Volume 3, provides the 

23 basis for the default value for each behavioral and metabolic parameter (see Table I.11). If the 

24 licensee modifies the values for these parameters, the NRC staff should verify that the licensee 

25 has defined a site-specific critical group. The licensee may provide site-specific parameter 

26 distributions that reflect the variability of the behavior of the average member of the site-specific 

27 critical group, or the licensee may use the mean of the site-specific information as a 

28 constant-value input for these parameters, consistent with the concept of the “average member” 

29 of the critical group. The level of justification required to support modification of behavioral and 

30 metabolic parameter values should be consistent with the sensitivity of results to changes in the 

31 parameter value.

32

33 For the DandD building occupancy exposure scenario, there are only three physical parameters: 

34 the resuspension factor (Rfo*), which is derived from the loose fraction (Fl) and the loose 

35 resuspension factor (Rfo). If the inhalation pathway dose can be significant, these parameters 

36 should be carefully selected. The default removable fraction assumed in DandD is 10 percent. 

37 If the removable fraction is expected to be greater than 10 percent, the licensee should account 

38 for higher removable fractions that might increase the resuspension factor. Because the default 

39 removable fraction is multiplied by the loose resuspension factor to derive the resuspension 

40 factor in DandD, either the removable fraction or the resuspension factor can be adjusted to 

41 account for removable fractions greater than the default value of 10 percent. If site conditions 

42 are consistent with assumptions in NUREG-1720 (with respect to activities and exposure 

43 scenarios, ventilation conditions, and low removable fractions at the time of decommissioning), 

44 NUREG-1720 (NRC, 2002) recommended the use of resuspension factor value or parameter 

45 distribution with minimal justification.
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1 There are many more physical parameters for the DandD residential exposure scenario. The 

2 physical parameters may be considered in several groups. The following physical parameters 

3 address the geohydrologic conditions: 

4 • Unsaturated Zone Thickness (H2)

5 • Soil Classification (SCSST)

6 • Porosity Probability (NDEV)

7 • Permeability Probability (KSDEV)

8 • Parameter “b” Probability (BDEV)

9 • Water Application Rate (AP)

10 • Surface Soil Porosity (N1)

11 • Unsaturated Zone Porosity (N2)

12 • Surface Soil Saturation (F1)

13 • Unsaturated Zone Saturation (F2)

14 • Infiltration Rate (INFIL)

15 • Surface Soil Density (RHO1)

16 • Unsaturated Zone Density (RHO2)

17 • Surface Soil Permeability (Ksat1)

18 • Soil Moisture Content (sh)

19 For these physical parameters, the licensee should use site-specific distributions and values. 

20 (As stated previously, “site-specific” in this context includes (1) information directly related to the 

21 site, (2) information characterizing the region that is consistent with site conditions, and 

22 (3) generic information that is consistent with the specific geohydrologic conditions at the site

23 (e.g., consistent with the unsaturated zone soil classification)).

24 The NRC staff should verify that the licensee has provided site-specific information for the 

25 thickness of the unsaturated zone and the soil classification. In addition, the licensee should 

26 ensure that the water application rate is consistent with the irrigation rate (behavioral 

27 parameter), if the licensee modifies the irrigation rate. Alternatively, the licensee may 

28 demonstrate, through sensitivity analyses, that the dose assessment results are insensitive to 

29 these parameters and use the default ranges.

30 Values for the derived parameters will be generated internally according to the soil classification 

31 indicated and the uniform distributions defined for the porosity probability (NDEV), the 

32 permeability probability (KSDEV), and the parameter “b” probability (BDEV). The NRC staff 

33 should verify that the licensee has not modified the uniform distributions for these three 
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1 parameters. If site-specific data are available, the licensee may proceed to modify the derived 

2 geohydrologic parameters, consistent with the information presented in NUREG/CR-5512, 

3 Volume 3 (NRC, 1999c).

4 The only geochemical parameter used in DandD is the element-specific partition coefficient. As 

5 documented in NUREG/CR-5512, Volume 3 (NRC, 1999c), the partition coefficients at a site are 

6 generally both dependent on geochemical conditions and independent of soil classification. If 

7 the licensee has used the default distributions, the reviewer should evaluate whether the 

8 defaults are consistent with known or expected conditions at the site.

9 The following physical parameters address radionuclide transport through the atmosphere and 

10 exposure to direct radiation:

11 • outdoor shielding factor (SFO)

12 • flood dust loading (PD)

13 • indoor resuspension factor (RFR)

14 • outdoor dust loading (CDO)

15 • indoor dust loading (CDI)

16 • indoor/outdoor penetration factor (PF)

17 • gardening dust loading (CDG)

18 The remaining physical parameters address characteristics of transport through the biosphere:

19 • growing periods (produce, forage, grain, hay) (TG_(#))

20 • animal product specific activity (SATac)

21 • livestock feeding periods (TF_(#))

22 • animal product yields (YA(#))

23 • interception fractions (R_(#))

24 • translocation factors (T_(#))

25 • contaminated fractions (x_(#))

26 • crop yields (Y_(#))

27 • wet-to-dry conversion factors (W_(#))

28 • animal ingestion rates (Q_(#))

29 • mass-loading factors (ML_(#))

30 • carbon fractions (fc_(#))
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1 • hydrogen fractions (fh_(#))

2 • hydrogen fraction: soil (fhd016)

3 • tritium equivalence: plant/soil (sasvh)

4 • tritium equivalence: plant/water (sawvh)

5 • tritium equivalence: animal products (satah)

6 These two groups of physical parameters describe characteristics of the transport of 

7 radionuclides through the atmosphere or biosphere up to the point of ingestion or inhalation by, 

8 or external exposure to, the receptor. The licensee may accept the default distributions for 

9 these parameters as long as they are consistent with conditions that may exist at the site in the 

10 future. The licensee should review the basis given in NUREG/CR-5512, Volume 3 

11 (NRC, 1999c), for the default distributions, to determine whether the basis is consistent with 

12 conditions hypothesized for the site. If not, the licensee should modify the input values 

13 accordingly. The NRC staff should ensure that the licensee documents this assessment for 

14 each of the physical parameters. Note that modifying several of these parameters (e.g., crop 

15 yields, animal product yields) should affect the derived behavioral parameters (e.g., area of land 

16 cultivated).

17 For the physical parameters, the licensee may use representative distributions or values. A 

18 representative distribution should take into account spatial and temporal variation of the 

19 parameter at the site. A representative distribution, for example, would be a precipitation rate 

20 based on the historical precipitation data for the site, if available, or from surrounding defensibly 

21 relevant monitoring locations. The arithmetic or geometric mean value is often used in defining 

22 a representative value. However, the calculation of a mean value should be weighted to 

23 account for nonuniform sampling or other nonuniform parameters (e.g., material volume) and 

24 parameter sensitivity and uncertainty. The licensee is not required to routinely adopt worst25 case, bounding, upper- or lower-percentile, or other overly conservative values in defining 

26 distributions.

27 The review of this information should be facilitated if the licensee presents the information in a 

28 tabular or list format. The NRC staff should verify that the licensee has listed every DandD input 

29 parameter with the default screening distributions or value (for behavioral or metabolic 

30 parameters). For those parameters for which the licensee is using site-specific values (e.g., the 

31 physical parameters), the licensee should provide the range of plausible values for the site, the 

32 selected distribution or value, and supporting justification, including references. 

33 I.6.4.2 Modifying the RESRAD Default Probabilistic Parameter Set

34 A licensee using the RESRAD-ONSITE, RESRAD-OFFSITE, or RESRAD–BUILD codes may 

35 change parameters from the default values to reflect a site-specific critical group or site-specific 

36 conditions, or to incorporate site-specific data. As discussed in the preceding section, the NRC 

37 staff should expect that a licensee who is modifying parameter values for a site-specific analysis 

38 is cognizant of the following:

39 • what the parameter represents

40 • how the parameter is used in the code
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1 • the basis for the default parameter value

2 • which parameters are physically or numerically correlated

3 The licensee should refer to the current code documentation to determine the basis for the 

4 parameter distributions and how they are used in the code and provide references to the 

5 documentation. With respect to the basis for the default parameter distributions and values, the 

6 licensee should refer to NUREG/CR-6697 (NRC, 2000c) and NUREG/CR-7267 (NRC, 2020a). 

7 When modifying parameter distributions and values, the licensee should consider whether the 

8 parameters are classified as behavioral, metabolic, or physical. For behavioral and metabolic 

9 parameters for which probability distributions have been developed, the licensee may adopt the 

10 DandD default distribution, or the mean of the DandD default distribution, as long as the 

11 licensee has not modified the definition of the critical group (see Table I.11). For behavioral and 

12 metabolic parameters for which distributions have not been developed, the licensee should use 

13 values or distributions that are consistent with the DandD default distributions, as applicable. 

14 A licensee may modify behavioral and metabolic default input parameter values to reflect the 

15 characteristics of the average member of a site-specific critical group. The licensee may modify 

16 the values for these parameters if the licensee has defined a site-specific critical group. The 

17 licensee may provide site-specific parameter distributions that reflect the variability of the 

18 behavior of the average member of the site-specific critical group use the mean of the 

19 site-specific information as a constant-value input for these parameters, consistent with the 

20 concept of the “average member” of the critical group. The level of justification required to 

21 support modification of behavioral and metabolic parameter values should be consistent with 

22 the sensitivity of the results to changes in the parameter.

23 For the physical parameters, the licensee should use site-specific information addressing 

24 geohydrologic and meteorologic conditions. The level of justification for the parameter values 

25 should be based on sensitivity analyses. Alternatively, sensitivity analyses may be used to 

26 support the use of default distributions or representative values. Care should be taken to 

27 ensure that use of overly broad distributions for parameters that effect the timing of peak dose 

28 do not lead to risk dilution (see Appendix Q for additional details).

29 For the physical parameters describing geochemical conditions (i.e., Kds), the licensee should 

30 use values that are consistent with known or expected site conditions. The RESRAD default 

31 distributions may be acceptable for sensitivity analyses, as long as the values are consistent 

32 with site-specific conditions. If the peak of the mean from a probabilistic analysis is used to 

33 demonstrate compliance with radiological criteria for license termination, as stated in the 

34 previous paragraph, the range of Kd values should not be overly broad (i.e., significantly outside 

35 of expected site-specific conditions) due to the potential for risk dilution, or the spreading out of 

36 peak dose over time leading to an artificially low peak of the mean dose that is not reflective of 

37 site risk. A sensitivity analysis can be used to determine the relative importance of the 

38 parameter values on dose. Parameter distributions for parameters having a significant impact 

39 on the results should be carefully selected based on site-specific conditions.

40 For the remaining physical parameters (atmospheric and biospheric transport), the licensee may 

41 use distributions or representative values that are consistent with the RESRAD default 

42 distributions, as applicable, as long as the default distributions are consistent with known or 

43 expected site conditions.
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1 I.6.4.3 Sensitivity Analyses

2 The level of justification required to support site-specific parameter values should be 

3 commensurate with the sensitivity of the results of the dose assessment to the selected values. 

4 Section I.7 of this appendix discusses sensitivity analyses in detail.

5 I.6.4.4 Site-Specific Distribution Coefficients for Soil or Concrete

6 The following describes an acceptable approach for the developing input Kd values for soil or 

7 concrete for use in site-specific dose modeling codes.

8 It is noted that Kd values commonly reported in the literature may vary by as much as six orders 

9 of magnitude for a specific radionuclide. Generally, no single set of ancillary parameters, such 

10 as pH and soil texture, is universally appropriate in all cases for determining appropriate Kd

11 values. Although Kd values are intended to represent adsorption, they are in most cases a 

12 lumped parameter representing a myriad of processes. Given the above, the proper selection 

13 of a range of Kd values, for either soils or concrete, from the literature will require judicious 

14 selection.

15 The licensee is encouraged to use sensitivity analyses to identify the importance of the Kd

16 parameter on the resulting dose. The sensitivity analysis should encompass an appropriate 

17 range of Kd values. The input range for the sensitivity analysis may be obtained from literature, 

18 DandD default distributions, RESRAD probabilistic default distributions, and ideally site-specific 

19 information. 

20 A commonly used reference for Kd data, Sheppard and Thibault, “Default Soil Solid/Liquid 

21 Partition Coefficients, Kds, for Four Major Soil Types: A Compendium” (Sheppard and 

22 Thibault, 1990), provides lognormal distribution parameters for a long list of radionuclides by soil 

23 type. Sheppard et al. “Soil Nuclide Distribution Coefficients and Their Statistical Distribution” 

24 (Sheppard et al., 1984), and Sheppard and Evenden, “Comparison of Partition Coefficients for 

25 54Mn and Soil-Extractable Mn, Including Relationship to Plant Uptake” (Sheppard and 

26 Evenden, 1989) are cited in Sheppard and Thibault (1990) as providing support for the use of 

27 lognormal distributions to characterize Kds. Although lognormal distributions are commonly 

28 used, it is important to note that the selection of an appropriate parameter distribution to 

29 describe Kd uncertainty or variability should be based on the goodness of fit of the model to the 

30 data available to define the distribution. Undoubtedly, Sheppard and Thibault’s (1990) 

31 conclusion that Kd data follow a lognormal distribution is, based in large part, on the inclusion of 

32 a wide range of values from the literature representative of a number of sites. Depending on the 

33 objective of the modeling exercise (e.g., to understand sensitivity of the results to potential 

34 range of Kds representative of a number of conditions and sites, or to understand uncertainty in 

35 the results given the potential variability of Kds for a particular site when uncertainty has already 

36 been reduced with respect to the site characteristics), the most appropriate parameter 

37 distribution to describe the Kd or coefficients for a particular radionuclide and site may vary. 

38 Therefore, the most appropriate distribution for any particular site, or project, cannot be 

39 determined a priori and should be based on a careful review of the data. The licensee can use 

codes such as MATLAB® 40 statistics toolbox to develop parameter distributions and evaluate 

41 parameter correlations, if sufficient data are available.

42

43 Using the results of the sensitivity analysis, the licensee can obtain support for its selection of Kd

44 value. For example, if higher Kd values result in a larger dose, an input Kd value should be 

45 selected from the upper range of values that represent site-specific conditions. If lower Kd
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1 values result in the larger dose, an input Kd value should be selected from the lower range of 

2 values that represent sites-specific conditions. If dose and compliance risk are sensitive to the 

3 selection of Kd, it may be necessary to conduct experiments using site materials to provide 

4 support for Kd values used in dose modeling. For those isotopes where the Kd does not have a 

5 significant impact on the dose assessment based on a sensitivity analysis (i.e., the dose results 

6 are not sensitive to Kd), limited justification will be needed to support selection of the parameter 

7 value.

8 As a result of the NRC staff review of dose modeling to support decommissioning and 

9 performance assessment modeling for low-level waste disposal, a number of observations and 

10 recommendations are noted with respect to the use of Kds in dose modeling to demonstrate 

11 compliance with radiological criteria for license termination or in performance assessment 

12 analyses. Table I.12 lists key considerations related to selection of parameter values and 

13 consideration of uncertainty.

14

15 DANDD

16 Data in Table 6.86 of NUREG/CR-5512, Volume 3, “Residual Radioactive Contamination from 

17 Decommissioning—Parameter Analysis,” issued 1999 (NRC, 1999c), may be useful in the 

18 development of Kd parameter distributions in a sensitivity analysis but should be representative 

19 of or expected to encompass site-specific conditions.

20

21 RESRAD FAMILY OF CODES

22 Licensees should not use RESRAD default parameter values (including Kd values). The code23 included defaults serve primarily as place holders that enable it to be run; it was assumed that 

24 site-specific values would be developed. However, the default Kd parameter distributions, 

25 developed for RESRAD-ONSITE, Version 6.0, may be useful in the development of Kd

26 parameter distributions for use in sensitivity analysis but should be representative or expected 

27 to encompass site-specific conditions. Updated information on parameter distributions is 

28 provided in NUREG/CR-7267.

29 Following sensitivity analysis with the appropriate Kd ranges, support can be provided for a 

30 value in the upper or lower range of the distribution, whichever endpoint results in the highest 

31 derived dose. If the use of Kd values near the endpoints of the distribution leads to doses above 

32 the release criteria, a site-specific Kd value may be needed to demonstrate compliance with the 

33 radiological criteria for license termination. 

34 METHODS FOR DEVELOPMENT OF SITE-SPECIFIC Kd

35

36 There are five general methods used to measure Kd values: laboratory batch method, in situ

37 batch method, laboratory flow-through (or column) method, field modeling method, and Koc

38 method (EPA, 1999, page 62). Each approach has its own built-in assumptions and 

39 consequently, its own distinct advantages and disadvantages compared to other methods. 

40 Ultimately, differences in assumptions can lead to significant differences in calculated Kds when 

41 using these various methods as discussed in more detail below.

42

43 Experimental methods used to calculate Kds have several built-in assumptions and limitations. 

44 For example, intrinsic to the batch method is the assumption of achieving equilibrium (or at least 

45 constant concentration of the contaminant) and that this equilibrium is entirely reversible. This 
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1 can be addressed by taking measurements to ascertain that concentrations have become 

2 constant. Additionally, results of an experiment alone cannot distinguish between adsorbed and 

3 precipitated species with precipitation sometimes being attributed to sorption.

4 Table I.12 Distribution Coefficient Parameter and Uncertainty Evaluation Considerations

Category Consideration or Recommendation

Parameters Literature values based on soil type are not considered site-specific 

values. Furthermore, Kds may not be strongly correlated to soil type.

Development of site-specific Kds (versus use of literature values) may 

be necessary to inform the decision-making process, if the decision 

being made is sensitive to selection of the parameter value over a 

plausible range of parameter space.

Experiments used to develop site-specific parameters should be 

representative of field conditions. Complicating mechanisms that may 

be present in the field may be difficult to reproduce, study, or interpret 

in the laboratory. Experimental artifacts may lead to erroneous 

conclusions on sorption potential in the field. Complicating 

mechanisms that may affect sorption include variable saturation, 

presence of a colloidal or anion exclusion mechanism, particle size 

heterogeneity, variable solution chemistry, and differences in sorption 

and desorption reactions and rates. Experimental artifacts include 

issues with material storage and preparation, sorption on filters or 

containers, nonrepresentative soils or aqueous chemistry, introduction 

of incompatible components that may chemically alter the system, and 

flow channeling or variability.

Treatment of 

Uncertainty

Overly broad Kd parameter distributions in probabilistic analyses could 

lead to risk dilution and poor decision-making. A significant difference 

between peak of the mean dose versus mean of the peak dose is an 

indication of risk dilution. If there is evidence of risk dilution, effort 

should be made to constrain the Kd parameter range. A metric 

different than the peak of the mean dose could also be used to 

evaluate compliance. Broad distributions may be acceptable for the 

purpose of conducting a sensitivity analysis.

In some cases, when ground-water-dependent pathways dominate the 

dose, lower distribution coefficients may lead to higher doses. In other 

cases, when non-ground-water-dependent pathways dominate the 

dose, higher distribution coefficients may lead to higher doses, 

particularly if in-growth of daughter products can increase the dose/risk 

over time.

5

6 Typically, only the total aqueous concentration of a contaminant is measured, with no analysis 

7 of speciation; implicit in this measurement is the assumption that only one aqueous phase 

8 species is present. A potential problem with assuming that only one aqueous species is present 

9 when, in reality, there are multiple species of varying mobility, is that the measured Kd

10 represents an average Kd that may not provide sufficient information to accurately model 

11 contaminant transport. For example, if two species are present, one with significantly lower Kd

12 than the other, faster travel times and potentially higher peak concentrations associated with the 
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1 more mobile fraction will not be simulated using an average Kd, potentially leading to an 

2 underestimate of contaminant concentrations. Column experiments have distinct advantages 

3 over batch experiments in their ability to identify fractions of significantly higher mobility, as 

4 evidenced by significantly earlier breakthrough times for a fraction of the total source activity. 

5 Although column experiments may be able to identify relatively mobile fractions, accurately 

6 identifying mechanisms and selecting appropriate transport parameters for predictive modeling 

7 may be more difficult.

8

9 Although various experimental methods are routinely used to measure or calculate distribution 

10 coefficients, experimental conditions can compromise their effectiveness in developing 

11 representative distribution coefficients. Experimental conditions that can affect Kd

12 measurements include pH; redox speciation; dissolved gas and solute concentrations; 

13 temperature and pressure; physical, chemical and mineralogical characteristics of available 

14 sorbing surfaces; stirring rates and/or flow rates (P.G. Glynn, “Modeling Np and Pu Transport 

15 with a Surface Complexation Model and Spatially Variant Sorption Capacities: Implications for 

16 Reactive Transport Modeling and Performance Assessments of Nuclear Waste Disposal Sites” 

17 (Glynn, 2002)), among others. As a result, thought and care should be applied to developing 

18 the experimental plan for sorption studies. The selection of materials, including appropriate site19 specific sediment and groundwater, is of primary importance. American Society for Testing and 

20 Materials (ASTM) C-1733, “Standard Test Method for Distribution Coefficients of Inorganic 

21 Species by the Batch Method” (ASTM, 2010) discusses the necessary precautions that have 

22 been found to be important.

23

24 There are a variety of published tests for Kd measurements. Experimenters should be judicious 

25 in their selection of methods to develop sorption parameters. Some tests are appropriate to 

26 measure sorption of metals and radionuclides, while some are designed for organics. Some 

27 tests use batch, equilibrium methods, while others are flow through. Some tests address 

28 sorption while others look at desorption (through the use of leaching tests). The following 

29 section briefly discusses some of these tests and their advantages and disadvantages. EPA 

30 (1999) provides a good evaluation of the different methods used to derive distribution 

31 coefficients and the advantages, disadvantages, and assumptions for each of the methods.

32

33 Batch Methods to Determine Kds

34

35 Batch adsorption tests for the distribution coefficient (or partition coefficient, or Kd) are 

36 essentially simple tests in which a known quantity of a solid (e.g., soil or mineral sample) is 

37 contacted with a volume of water containing a known (often spiked) concentration of some 

38 species of interest. After allowing the system to come to steady state (actually, it is constant 

39 concentration), the aqueous concentration of the species of interest is measured and compared 

40 to the starting concentration. Typically, just the solution is measured and the mass of sorbed 

41 species is determined by subtraction of the before and after concentrations. While in principle 

42 this is a simple test, details are important and can make big differences in the Kd value obtained. 

43 Among the important factors, probably the greatest is the pH of the contact solution. For the 

44 NRC’s purposes, it is important that site-specific materials, both soil and groundwater, be used 

45 and that these materials be handled in ways that do not disturb their chemical properties.

46

47 Currently, ASTM has two approved batch Kd methods. There were several earlier ASTM tests 

48 that have either been withdrawn or have not been renewed (ASTM tests need to be renewed 

49 every 5 years or they are dropped from the published methods). Both are batch tests and both 

50 are specifically designed to quantify the sorption of inorganic species. This is an important 

51 point; tests for organics can be more complex than those for inorganics because, for organics, 
I-91

1 one needs to consider degradation and volatilization of the contaminant. A few tests for organic 

2 contaminants are listed below but are not described.

3

4 ASTM D4646-3 (2016), Standard Test Method for 24-h Batch Type Measurement of 

5 Contaminant Sorption by Soils and Sediment, is “applicable in screening and providing relative 

6 rankings of a large number of samples for their sorption affinity in aqueous leachate/geomedia 

7 suspensions.” While this method may be useful for screening, it is generally considered to be 

8 too short to establish steady-state conditions, which are needed for models. This time 

9 dependence is very much controlled by the element or species being tested; for example, 

10 anionic species typically need a number of days to reach constant concentration. The method 

11 suggests using ASTM D-4319 (2002) for longer tests. However, D-4319 has not been renewed

12 (the standard was withdrawn in 2007). Because of the short duration of the test, D-4646 is not 

13 recommended for generating Kds for use in performance assessment modeling.

14

15 ASTM C-1733-10 is a new method, the scope of which is to “quantify uptake onto solid 

16 materials by a batch method.” It is intended “to assess sorption of dissolved ionic species 

17 subject to migration through pores and interstices of site-specific geomedia.” This method 

18 provides a substantial discussion of factors that can influence Kd values, with emphasis on being 

19 aware of these effects or controlling them in the laboratory. The method discusses its use with 

20 radioactive tracers and recommends a preliminary Kd test to determine the time needed to reach 

21 constant solution concentrations of the species of interest. A correction method is given for 

22 materials containing grains larger than 2 mm. The “Significance and Use” section of the method 

23 states, “Because of the sensitivity of Kd to site-specific conditions and materials, the use of 

24 literature derived Kd values is strongly discouraged.” This method is suggested in Regulatory 

25 Guide 4.7, Revision 3, “General Site Suitability Criteria for Nuclear Power Stations,” issued 

26 March 2014 (NRC, 2014), that was issued as DG-4021 in December 2011. Ebert and Petri, 

27 “Uptake of Cs and Sr on San Joaquin Soil Measured Following ASTM Method C-1733,” (Ebert 

28 and Petri, 2012) evaluated the repeatability and sensitivity to test parameters for this test 

29 method; for example, an average Kd of 373 mL/g for Cs was determined with a standard 

30 deviation of 21 mL/g or 5.6 percent.

31

32 Other batch methods are published but not as official consensus standard methods. One is the 

33 “Standard Method Used at Pacific Northwest National Laboratory for Measuring Laboratory 

34 Batch Kd Values.” It is published as Appendix C in Volume 1 of EPA 402-R-99-004A, 

35 “Understanding Variation in Partition Coefficient, Kd, Values Volume 1: The Kd Model of 

36 Measurement, and Application of Chemical Reaction Codes” (EPA, 1999). This method is 

37 oriented toward radioactive tracers and stipulates a 7-day contact period unless time is a 

38 parameter to be investigated. 

39

40 EPA published another batch method in 1992 in a Technical Resource Document entitled 

41 “Batch-Type Procedures for Estimating Soil Adsorption of Chemicals,” EPA/530/SW-87/006-F 

42 (Roy et al., 1992). Chapter 17 of this informative report, Laboratory Procedures for Generating 

43 Adsorption Data, contains a detailed procedure with provisions for both organic and inorganic 

44 contaminants. In this method, emphasis is placed on assessing the impacts of various test 

45 parameters such as contact time and solid/solution ratios. The results are used to generate

46 isotherms. 

47

48 Serne and Relyea, in “Waste/Rock Interactions Technology Program: The Status of 

49 Radionuclide Sorption-Desorption Studies Performed by the WRIT Program,” issued April 1992

50 (PNL-3997), discuss laboratory methods development for batch tests and report on a round51 robin test that was conducted among nine participants using Cs, Sr, and plutonium (Pu) tracers. 
I-92

1 Results were very scattered and Pu adsorption data were regarded as not satisfactory, with two 

2 to three orders of magnitude differences. On examining the results of the tests, the problem 

3 was ascribed to Pu retention on containers and formation of Pu particles. Overall, it was

4 concluded that several uncontrolled parameters may have affected results: (1) method of tracer 

5 addition to solution, (2) solution-to-rock ratio, (3) initial tracer concentration in influent solution, 

6 (4) particle size distribution, (5) solid-solution separation method, (6) sample containers, and

7 (7) temperature.

8

9 Although they are not the subject of this report, a few standard test methods for adsorption of 

10 organic chemicals are listed below. While they are oriented toward organics, these methods 

11 present useful information for all sorption tests.

12

13 • ASTM E1195-01(2008), “Standard Test Method for Determining a Sorption Constant

14 (Koc) for an Organic Chemical in Soil and Sediments”

15 • EPA 712-C-08-009, (2008), “Fate, Transport and Transformation Test Guidelines,

16 OPPTS 835.1230, Adsorption/desorption (Batch Equilibrium)”

17 • OECD 2000, “OECD Guideline for Testing of Chemicals, Adsorption-Desorption Using a

18 Batch Equilibrium Method, Method 106”

19 In Situ Batch Kd

20

21 In situ batch methods may also be used to develop site-specific Kds. Using this method, cored 

22 samples are extracted directly from the aquifer. Phase separation is accomplished through 

23 centrifugation or filtration and contaminant concentrations in both phases are measured. The 

24 advantage of this approach, compared to the laboratory Kd method, is that actual samples are 

25 used to calculate the Kds. There is also a greater likelihood that the sample is closer to 

26 equilibrium. A disadvantage of this approach is that surface concentrations of most metal 

27 contaminants are typically low and difficult to measure. Energy dispersive x-ray analysis and 

28 x-ray fluorescence, with typical detection limits around 10,000 and 100 ppm, respectively, are

29 commonly used to measure contaminant concentrations on the solid phase. Lower detection

30 limits may be obtained by dissolving the solid phase sample with acids and through use of

31 inductively coupled plasma spectroscopy, inductively coupled plasma/mass spectroscopy, or

32 atomic adsorption spectroscopy techniques. In addition to the detection limit problem and

33 similar to problems with the laboratory batch method, no differentiation can be made between

34 sorption and precipitation. Additionally, some trace contaminants may be associated with

35 crystalline lattice sites of minerals present in soils rather than being reversibly sorbed to mineral

36 surfaces. For anthropogenic radionuclides present at trace levels, it may be acceptable to

37 assume that the total mass measured on the solid is reversibly sorbed and that precipitation and

38 lattice site associations are negligible (EPA, 1999).

39

40 Column Experiments to Determine Kds

41

42 Column tests are used much less often than batch tests and, in fact, searching the literature has 

43 resulted in no standard methods for soil adsorption tests using columns. Of course, all sorts of 

44 adsorption columns are used, from laboratory scale to immense industrial columns, so the 

45 concept is readily and often applied to soils. There are papers and reports that describe the 

46 apparatus and methods specifically for soil column tests and a few that compare results of batch 

47 and column tests. Most column tests are used to assess the behavior of organic chemicals, but 

48 there are many that investigate inorganic chemicals. There are standard column leach test 
I-93

1 methods that have many similarities to adsorption experiments, especially the apparatus used;

2 see ASTM D-4874; Leaching Solid Material in a Column Apparatus.

3

4 Column adsorption tests are substantially more complex, both physically and analytically, than 

5 batch tests. Relyea et al. (“Methods for Determining Radionuclide Retardation Factors: Status 

6 Report,” 1980 (Relyea et al., 1980)) provide a good comparison and discussion of the two types 

7 of tests. The experimental apparatus for column tests includes a column that will contain the

8 material to be tested and not leak, a low-flow pump, a reservoir to hold the influent solution, and 

9 a sample collection system. Compared to a batch test, more information is needed to interpret 

10 a column test, including column dimensions, mass of added material, porosity, bulk density, flow 

11 rate, and sample collection times. Typically, column tests require many samples over time, as 

12 opposed to a limited number in batch tests (to establish constant concentration). Depending on 

13 the Kd of the system being tested, column tests may need to run for weeks or months (or more) 

14 to establish breakthrough. Alternatively, the column material can be extruded and sectioned 

15 and the solids analyzed for profiles of the species of interest. Interpretation of results generally 

16 requires computer models, and it is good practice to perform a measurement of flow rate and 

17 dispersion using a nonreactive tracer such as tritiated water.

18

19 There is general agreement on some experimental procedures for column tests. Columns 

20 should be run with water entering from the bottom of the column. This “upflow” mode helps 

21 eliminate preferential pathways. Columns should be more than 20 times the width of the largest 

22 grains of the tested material. Column fabrication materials should be chosen to minimize 

23 sorption on walls, for example. The design should include a means of holding the tested 

24 material in place. Flow rates should be considered carefully, because they control residence 

25 time. Serne and Relyea (1982) discuss some advantages and disadvantages of column 

26 methods. Relyea et al (1980) provide, in Appendix C, a method for a column procedure and 

27 data reporting. Little information is given for the apparatus; the emphasis is on data reporting 

28 and interpretation. 

29

30 Fjeld et al., “Final Report Column Tests to Study the Transport of Plutonium and Other 

31 Radionuclides in Sedimentary Interbed at INEEL” (Fjeld et al., 2000), provide some 

32 methodology for column tests used to investigate transport of radionuclides such as Sr, Am, Pu, 

33 neptunium (Np), and U. The analysis of column data, the interpretation of breakthrough curves, 

34 and the geochemistry controlling them are useful. Although they are focused on organics, 

35 column packing, conservative tracer tests, and retention time concerns are universal.

36

37 Field Transport Rate Approach

38

39 This field approach to deriving Kd values uses groundwater monitoring data, source term data, 

40 and other flow and transport parameters to back-calculate Kds based on travel times. A dual 

41 tracer test, using a nonsorbing conservative tracer and a tracer of the element of interest may 

42 be useful. Analytical models can be used to solve for the retardation factor from which a 

43 distribution coefficient can be calculated. Alternatively, numerical methods can be used to 

44 determine Kd through calibration with measured plume concentrations.

45

46 Potential Problems for Sorption Experiments

47

48 The following sections describe several experimental artifacts and methods of overcoming them 

49 to make them more useful for predicting contaminant mobility in the environment.

50
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1 Hysteresis—One complicating factor in sorption experiments is that many contaminants have 

2 been observed to sorb more readily than desorb from mineral or organic surfaces, a 

3 phenomenon referred to as hysteresis. Most batch experiments measure sorption, not 

4 desorption, although desorption may be much slower than sorption. If sorption reactions are not 

5 completely reversible or desorption is kinetically limited, the contaminant may be present for 

6 longer periods of time than assumed and remedial options (e.g., pump and treat) may be 

7 hampered.

8

9 Gravel Issue—The “gravel issue” is the problem that transport modelers face when converting 

10 laboratory-derived Kd values based on experiments using the less-than-2-mm fraction into 

11 values that can be used in systems containing particles greater than 2 mm in size. For 

12 example, experiments using less than a 2-mm fraction may overestimate sorption due to the 

13 higher surface area of smaller particles. EPA (1999) proposed two methods: the first considers 

14 the fraction of larger and smaller size particles present at the site. The experimentally derived 

15 Kd using 2-mm-size particles is adjusted to account for larger materials. To adjust the Kd, site 

16 soils are characterized to identify the fraction of particles with diameters greater than 2 mm and 

17 a Kd of 0 L/kg is applied to the larger particle size fraction. This approach has been adopted by 

18 ASTM C-1733. The second approach adjusts the experimentally derived Kd using the ratio of 

19 surface area in the experiment to the surface area of the material in the field. This method 

20 assumes a relatively small specific surface area of gravel-size material. 

21

22 Colloid Formation—While mathematical models are available to consider a separate colloidal 

23 phase, evaluating the impact of colloids on contaminant transport predictions is hampered by 

24 the difficulty in characterizing the colloidal phase in the field and through experimentation. 

25 Additionally, a lack of mechanistic information on the factors that lead to the generation and 

26 removal of colloids in site-specific applications adds to the uncertainty in contaminant transport 

27 projections. Contaminant mass sorbed to colloidal particles that remain in suspension after 

28 separation of the solid and aqueous phases is sometimes included with the total aqueous phase 

29 concentration, resulting in a significant underestimation of the distribution coefficient. For 

30 example, use of a centrifuge for liquid/solid phase separation may result in formation of a very 

31 thin zone of suspended particles at the liquid surface where surface tension holds finer particles 

32 in solution. When pipets are used to remove supernatant solution, the pipet should be inserted 

33 sufficiently below the liquid surface to limit extraction of suspended particles. Likewise, the pipet 

34 should be significantly above the bottom of the sample to avoid extraction of settled particles. 

35 Filtering following centrifuging is desirable; however, care should be taken to avoid filters that 

36 might sorb contaminants. Pretreatment of filters may also help minimize container sorption 

37 biases.

38

39 Materials—The selection of inappropriate materials for the experiments can seriously degrade 

40 the results of sorption tests. The soil or other solid phase must be representative of the material 

41 that the contaminant will flow through. Similarly, the aqueous phase must be sampled from the 

42 appropriate unit or must be carefully matched to the subject water, if simulated groundwater is 

43 used. Most especially, speciation altering parameters (such as pH and complexing ions) must 

44 be controlled. Care must be given to storage of these materials before the experiments start. 

45 Experimental materials should be judiciously chosen, making sure that no sorption takes place 

46 on containers or filters.

47
I-95

1 I.7 Uncertainty/Sensitivity Analyses 

2 I.7.1 Introduction

3 Uncertainty is inherent in all dose assessment calculations and should be considered in 

4 regulatory decision-making. In general, a dose assessment has three primary sources of 

5 uncertainty: (1) uncertainty in the models, (2) uncertainty in the scenarios, and (3) uncertainty in 

6 the parameters (NUREG/CR-4604, “Statistical Methods for Nuclear Material Management,” 

7 issued December 1988 (NRC, 1988a); DOE Conference, “Treatment of Uncertainty in 

8 Low-Level Waste Performance Assessment,” held November 19–21, 1991 (DOE, 1991)). As 

9 stated in Section J.4 of this appendix, models are simplifications of reality and, in general, 

10 several alternative models may be consistent with available data. Uncertainty in scenarios is 

11 the result of a lack of knowledge about the future of the site. Parameter uncertainty results from 

12 incomplete knowledge of the model coefficients.

13 The NRC’s risk-informed approach to regulatory decision-making suggests that an assessment 

14 of uncertainty be included in estimating doses. Specifically, the Probabilistic Risk Assessment 

15 (PRA) Policy Statement (60 FR 42622, August 16, 1995) states, in part, “The use of PRA 

16 technology should be increased in all regulatory matters to the extent supported by the state of 

17 the art in PRA methods and data, and in a manner that complements NRC’s deterministic 

18 approach and supports the NRC’s traditional defense-in-depth philosophy.” In the past, dose 

19 assessments in support of NRC decommissioning requirements have primarily included the use 

20 of deterministic analyses. The deterministic approach has the advantage of being simple to 

21 implement and easy to communicate to a nonspecialist audience. However, it has a significant 

22 drawback in not allowing consideration of the effects of unusual combinations of input 

23 parameters and by not providing information on uncertainty in the results, which would be 

24 helpful to the decisionmaker. Furthermore, a deterministic analysis that had a high assurance 

25 of not being exceeded would have to rely on the use of pessimistic estimates of each parameter 

26 of the model, often leading to overly conservative evaluations. Even with the use of probabilistic 

27 analyses, it is generally recognized that not all sources of uncertainty can be considered in a 

28 dose assessment nor need to be considered. The primary emphasis in an uncertainty analysis 

29 should be to identify the important assumptions and parameter values that, when altered, could 

30 change the decision.

31 A sensitivity analysis performed in conjunction with the uncertainty analysis can be used to 

32 identify parameters and assumptions that have the largest effect on the result. Sensitivity 

33 analysis provides a tool for understanding and explaining the influence of these key 

34 assumptions and parameter values on the variability of the estimated dose. Once identified, 

35 additional resources can be spent on refining risk-significant parameters. Sensitivity analyses 

36 also are useful in focusing NRC staff reviews on those issues most important to the decision37 making process.

38 I.7.2 Issues in Uncertainty/Sensitivity Analyses 

39 Uncertainty analysis imparts more information to the decisionmaker than deterministic analysis. 

40 It characterizes a range of potential doses and the likelihood that a particular dose may be 

41 exceeded.

42 An important issue in uncertainty and sensitivity analysis is that not all sources of uncertainty 

43 can be easily quantified. Of the three primary sources of uncertainty in dose assessment 

44 analyses, parameter uncertainty analysis is most mature. However, approaches for quantifying 
I-96

1 conceptual model and scenario uncertainty are less well developed. Difficulties in predicting the 

2 characteristics of future society, especially those influencing exposure, can lead to large 

3 uncertainties. At most, one is able to assert that the assessment has considered an acceptably 

4 complete suite of scenarios (Flavelle, “A Quantitative Measure of Model Validation and Its 

5 Potential Use for Regulatory Purposes” (Flavelle 1992)). For these reasons, we make no 

6 attempt to quantify formally model or scenario uncertainty, although to a certain extent, these 

7 are captured in parameter uncertainty analyses. Choices of the exposure scenarios and 

8 conceptual model(s) to be used for the site are discussed in Sections I.3 and I.4, of this 

9 appendix, respectively.

10 Uncertainty analyses frequently use the Monte Carlo method. Input variables for the models are 

11 selected randomly from probability distribution functions, which may be either independent or 

12 correlated to other input variable distributions. Critics of formal uncertainty analysis have often 

13 pointed out that limitations of knowledge about the nature and extent of correlation among 

14 variables fundamentally limit our ability to make meaningful statements about the degree of 

15 uncertainty in dose assessments (Smith et al., “The Effect of Neglecting Correlations When 

16 Propagating Uncertainty and Estimating the Population Distribution of Risk” 

17 (Smith et al., 1992)).

18

19 Because the results of an uncertainty analysis provide a distribution of doses, it should be 

20 recognized that some percentage of the calculated doses may exceed the regulatory limit. A 

21 key issue that should be addressed in the treatment of uncertainty is specifying how to interpret 

22 the results from an uncertainty analysis in the context of a deterministic regulatory limit. Agency 

23 practice has not been to require absolute assurance that the regulatory limit will be met, so 

24 regulatory compliance could be stated in terms of a metric of the distribution such as the mean, 

25 or a percentage of calculated doses allowed to exceed the limit. Even for a deterministic 

26 analysis, it is recognized that the reported dose is simply one of a range of possible doses that 

27 could be calculated for the site; therefore, there is still an issue of where this calculated dose 

28 should lie in terms of the unquantified spectrum of possible doses.

29

30 In summary, the key issues in addressing uncertainty are (1) incorporating alternative 

31 conceptual models and scenarios to identify a complete suite of possibilities, (2) determining 

32 how to select appropriate parameter distribution and ranges, along with the associated 

33 correlation between parameters for the analysis, and (3) specifying the metric of the dose 

34 distribution to use in determining compliance with the dose limit.

35

36 I.7.3 Recommended Approach 

37 I.7.3.1 Screening Analyses 

38 Often the first step in evaluating site compliance should be a screening analysis. At preliminary 

39 stages of the evaluation, there may be little information available about the site. Therefore, the 

40 NRC’s screening approach is designed to ensure that there is high confidence that the dose has 

41 not been underestimated. As discussed in Sections I.3 and I.4 of this appendix, the models and 

42 exposure scenarios used in screening were selected to represent generic conditions and are 

43 intended to be “prudently conservative.” The screening analysis assumes that all that is known 

44 about a site is the source term. Accordingly, the default parameters were selected to make it 

45 unlikely for the dose that would be calculated using site-specific information to exceed the 

46 screening dose.
I-97

1 To develop screening values, the NRC staff performed a Monte Carlo analysis, using the 

2 DandD Version 1 code, with values of the input parameters sampled from wide ranges selected 

3 to represent the variability in those parameters across the United States. The default values of 

4 input parameters for the DandD code (i.e., the values that the code would use without 

5 specification by the user) were then chosen from distributions of those parameters based on the 

6 90th percentile of the output dose distribution from the Monte Carlo analysis, as illustrated in 

7 Figure I.10 (NRC, 1999a).

8

9

10

11

12

13

14

15

16

17

18 Figure I.10 Treatment of Parameter Uncertainty in DandD Version 1

19 The intent of the specification of default parameter values, exposure scenarios, and conceptual 

20 models in the DandD code was to ensure that there should be less than a 10 percent probability 

21 that the calculated dose using site-specific information would exceed the dose limit. Because 

22 the default parameters, exposure scenarios, and conceptual models in DandD Version 1.0 were 

23 designed to provide high confidence that the dose would not be underestimated, a licensee 

24 using the screening criteria did not need to quantify the uncertainty in the dose analysis. The 

25 calculated results may be considered to represent a “prudently conservative” estimate of the 

26 dose (i.e., the calculated dose is likely an overestimation of the true dose). In many cases, 

27 however, the default parameter values chosen were highly conservative, making the outcome of 

28 the deterministic analysis overly stringent.

29 DandD Version 2 was designed to allow Monte Carlo analysis, which gives a distribution of 

30 doses, as illustrated in Figure I.11. The code automatically performs the probabilistic analyses 

31 and aggregates the results for the user. To maintain consistency in approaches used for 

32 Versions 1 and 2, and previously published screening tables, the 90th percentile of the dose 

33 distribution was used to determine screening values that demonstrate compliance with the 

34 unrestricted release criteria in 10 CFR Part 20, Subpart E. Default parameter PDFs were 

35 incorporated into the code for screening analyses; therefore, for those, the license reviewer may 

36 only need to ensure that these aforementioned default parameters were used. See Appendix H 

Parameter 1

Parameter 2

!

!

!

Parameter N

Dose Distribution

90th percentile

Default Parameter Set

Parameter 1

Parameter 2

!

!

!

Parameter N

pdf

pdf

pdf

Dose

DandD Version 1
I-98

1 for additional details on the screening values and the approaches used to develop the screening 

2 values.

3

4

5 Figure I.11 Treatment of Parameter Uncertainty in DandD Version 2

6

7 I.7.3.2 Site-Specific Analyses 

8 DETERMINISTIC ANALYSIS

9 For site-specific analyses, the treatment of uncertainty in deterministic and probabilistic 

10 analyses should be handled differently. The NRC’s risk-informed approach to regulatory 

11 decision-making suggests that an assessment of uncertainty should be included in dose 

12 analyses. However, in some cases, such analyses may not be needed (e.g., bounding type 

13 analyses). Because no information is provided on the uncertainty in bounding analyses, it is 

14 important for the licensee to demonstrate that the single reported estimate of the peak dose is 

15 likely to be an overestimation of the actual peak dose. Use of conservatism in only some 

16 aspects of the analysis may not necessarily result in a conservative estimate of the dose. 

17 Uncertainties in the conceptual model may be larger than uncertainties in the parameters used 

18 in the analysis; therefore, use of conservative parameter values does not necessarily ensure a 

19 conservative estimate of the dose. To ensure that the results from a deterministic analysis are 

20 unlikely to underestimate the dose, the NRC recommends that the licensee use the approaches 

21 discussed in Sections I.3 and I.4 of this appendix for developing land use exposure scenarios 

22 and conceptual models. In addition, the licensee should use conservative values for key 

23 parameters and should use the approaches discussed below on performing sensitivity analyses 

24 when identifying key parameters in the analysis.

Parameter 1

Parameter 2

!

!

!

Parameter N

Dose Distribution

90th percentile

pdf

pdf

pdf

Dose
I-99

1 PROBABILISTIC ANALYSIS

2 Although bounding analyses are a good starting point for determining regulatory compliance, 

3 the demonstration that a single, deterministic result is bounding may be too difficult to support. 

4 For site-specific probabilistic analysis, the licensee may use the peak of the mean dose to 

5 demonstrate regulatory criteria have been met.

6

7 A single deterministic calculation using the mean values of parameters is unlikely to result in the 

8 mean dose.

9

10 Parameter uncertainty analysis provides a quantitative method for estimating the uncertainty in 

11 calculated doses, assuming that the structure of the model is an adequate representation of the 

12 real world and that the exposure scenario is an appropriate reflection of potential future land use 

13 at the site. Several methods have been developed for quantifying parameter uncertainty, 

14 including (1) analytical methods, (2) Monte Carlo methods, (3) response surface methods, and 

15 (4) differential methods (DOE/LLW-100, “Guidelines for Sensitivity and Uncertainty Analysis of

16 Performance Assessment Computer Codes, National Low-Level Waste Management Program,”

17 issued September 1990 (DOE, 1990)). In addition, alternative approaches, such as the first18 order reliability method, have recently been applied on a wide variety of environmental problems

19 (DOE, 1998). Of these methods, the Monte Carlo methods are recommended, because they

20 are easy to implement and provide significant versatility.

21

22 Monte Carlo methods can be applied to either linear or nonlinear models and analytical or 

23 numerical models. Input parameter uncertainties are represented as PDFs. Parameter values 

24 randomly sampled from PDFs are used as inputs to multiple runs or “realizations” of the model.

25 For probabilistic analyses, the peak of the plot of mean dose over time should be compared with 

26 the regulatory standard to determine compliance. Equation 8 shows how the mean dose as a 

27 function of time can be derived. For Monte Carlo Runs:

28

29

Mean t

Dose t

N

i

k i

k

N

( )

( )

=

=



1

30 Equation 8

31

32 where Mean(ti) = mean dose at time ti

33 Dosek(ti) = dose at time ti for run k

34 ti = time in years

35 i = time steps (1 to 1000)

36 Essentially, a mean dose is determined at each discrete time in the analysis. A plot is then 

37 made of these means over time. The mean dose provides the “best estimate” of dose at each 

38 discrete time. The overall peak of these best estimates is then used to determine compliance 

39 with the rule. Figure I.12 shows how such a plot would be used to determine compliance with 

40 the regulations.
I-100

1

2

3 Figure I.12 Application of “Peak of the Mean” Dose.

4 Licensees using probabilistic dose modeling should use the “peak of the mean” dose for 

5 demonstrating compliance with the 10 CFR Part 20, Subpart E. Similar to all regulatory 

6 guidance, this NUREG report contains one approach for determining compliance with the 

7 regulations using probabilistic analyses. Use of “mean of the peaks” is also acceptable for 

8 demonstrating compliance. If the “mean of the peaks” dose is significantly higher than the “peak 

9 of the mean” dose, then “risk dilution” may be an issue in the probabilistic model. Consult 

10 Appendix Q for more information on the potential for and impacts of “risk dilution.” If the 

11 licensee intends to use any probabilistic approach to calculate DCGLs, it should discuss its 

12 planned approach with the NRC staff.

13

14 I.7.4 Input Parameter Distributions for Monte Carlo Analysis

15 A key aspect of any Monte Carlo analysis is defining the ranges and statistical distribution of 

16 parameters treated as uncertain in the analysis. It is important for the licensee to avoid 

17 assigning overly restrictive ranges that suggest an unwarranted precision in the state of 

18 knowledge. On the other hand, the specification of unreasonably large ranges may not account 

19 for what is known about a parameter and also may lead to “risk dilution.” The distributions used 

20 in the analysis should characterize the degree of belief that the true but unknown value of a 

21 parameter lies within a specified range of values for that parameter. 

22 Sensitivity results are generally less dependent on the actual distributions assigned to the input 

23 parameters than they are on the ranges chosen for the parameters. However, distributional 

24 assumptions can have a large impact on the dose distribution (SNL, SAND90-7103, “Sensitivity 

25 Analysis Techniques and Results for Performance Assessment at the Waste Isolation Pilot 

26 Plant” [SNL, 1991]). Resources can often be used most effectively by performing a Monte Carlo 

27 analysis in an iterative manner. Initially, rather crude ranges and distribution assumptions can 

28 be used to determine which input variables dominate the behavior of the calculated dose. 

29 Often, most of the variation in the calculated dose is caused by a relatively small subset of input 

30 parameters. Once the licensee identifies the most important input parameters, it can 

31 concentrate resources on characterizing their uncertainty. This avoids spending a large effort 

32 characterizing the uncertainty in parameters that have little impact on the dose (SNL, 1991).

Time (years)

Regulatory Limit

Peak
I-101

1 A reasonable strategy for assigning distributions for parameters used in Monte Carlo analyses is 

2 summarized below (NUREG-1573, “A Performance Assessment Method for Low-level Waste 

3 Disposal Facilities: Recommendations of NRC’s Performance Assessment Working Group,” 

4 issued October 2000 (NRC, 2000b):

5 • Select parameters to be assigned distributions—Not all parameters of the system

6 under study require specification of a distribution. Those parameters that may well be

7 distributed, but ultimately have little impact on the results, can be assigned constant

8 values. Even if a parameter is known to have a significant effect on the results, its value

9 may be specified at a constant value if it can be demonstrated that the choice leads to a

10 conservative result.

11 • Assign distributions for important parameters—The assignment of parameter

12 distributions is usually a matter of the quantity of available data.

13 • Ample data available—Where there are ample data, empirical distributions of a

14 parameter can be generated directly.

15 • Sufficient data available—Data plotted as histograms or in probability coordinates can

16 be used to identify standard distributional forms (e.g., normal, lognormal, and uniform).

17 • Parameters with some data—Where there are insufficient data to estimate the shape

18 of an empirical distribution, data may be supplemented by other soft information. For

19 example, if there were a mechanistic basis for assigning a given distribution, or if a

20 distribution were well known for the parameter, on a regional basis, this information

21 could be used to estimate the likely shape of the distribution. Alternatively, the new data

22 can be used to supplement a priori, non-site-specific parameter distribution

23 (e.g., Bayesian updating).

24 • Parameters with insufficient information—If sufficient data are not available but there

25 were other kinds of data that imply the likely behavior of a parameter, then it may be

26 possible to supplement the desired data indirectly. An example of such a procedure is

27 the use of root uptake factors to infer distribution coefficients in soil (Oak Ridge National

28 Laboratory (ORNL)-5786, “A Review and Analysis of Parameters for Assessing

29 Transport of Environmentally Released Radionuclides Through Agriculture,” issued

30 September 1984 (ORNL, 1984). Although use of root uptake factors to derive

31 distribution coefficients for soil is provided as an example, inferred correlations between

32 parameters should be adequately supported. If only incomplete information is known

33 about the parameter (e.g., its mean, or its range), and no correlations to other types of

34 data are available, then the choice of the parameter distribution should reflect the

35 uncertainty. The distribution should have the least-biased value, which is generally a

36 wide distribution encompassing all the possible values. One procedure to ensure that

37 the distribution has the least bias is known as the “maximum entropy formalism,” based

38 on Shannon’s informational entropy (M.E. H “Reliability Based Design in Civil

39 Engineering” (Harr, 1987)). This formalism allows the investigator to pick the distribution

40 based on the kinds of information available on the parameter to ensure that the result is

41 the least biased; for example, if only the range of the data is known, a uniform

42 distribution over the range is least biased. Table I.13 describes the maximum entropy

43 solutions for several classes of data (Harr, 1987). Other, empirical sources of guidance

44 for choosing parameter distributions can be found in several other references

45 (International Atomic Energy Agency (IAEA)-SS-100, “Evaluating the Reliability of
I-102

1 Predictions Made Using Environmental Transfer Models” (IAEA 1989); National Council 

2 on Radiation Protection and Measurements, “A Guide for Uncertainty Analysis in Dose 

3 and Risk Assessments Related to Environmental Contamination,” dated May 10, 1996 

4 (NCRP 1996a)).

5 • Parameter correlations—Many of the parameters used in the probabilistic analyses

6 may be correlated to other parameters. Some parameter distributions may, in fact, be

7 used to derive other distributions (e.g., root uptake factors may be used to derive soil

8 distribution coefficients). Also, correlations are expected on physical grounds, such as

9 the relationship between hydraulic gradient and permeability. Where available, these

10 correlation coefficients can then be used to generate correlated values of distributed

11 parameters. This may help to avoid the situation where two correlated quantities are

12 treated as uncorrelated, leading to unlikely combinations of parameters (e.g., high

13 gradient and high-hydraulic conductivity). The effects of assumed minimum versus

14 assumed maximum levels of correlation can be investigated to evaluate the importance

15 of including an explicit estimate of dependency between model parameters. In some

16 cases, explicit modeling of the dependency between model parameters is possible,

17 based on knowledge about the explicit mechanistic reasons for the dependencies. In

18 general, it is more important to consider the effect of dependency when correlations are

19 strong among the model’s most sensitive parameters (see discussion below on

20 identifying sensitive parameters); weak correlations between sensitive parameters and

21 strong correlations among insensitive parameters will generally have very little impact on

22 the overall calculated dose (NCRP, 1996a).

23 Table I. 13 Maximum Entropy Probability Distributions (Adapted from Harr, 1987)

Given Constraints on Data Assigned Probability Density

Minimum and maximum only Uniform

Expected value only Exponential

Expected value and standard deviation Normal

Expected value, standard deviation, 

minimum and maximum

Beta

Mean occurrence rate between

arrival of independent events

Poisson

24

25 I.7.5 Sensitivity Analysis

26 Uncertainty and sensitivity analyses are closely linked, and ideally, they should be considered 

27 together. The primary aim of a sensitivity analysis is to identify the input parameters that are the 

28 major contributors to the variation or uncertainty in the calculated dose. Identifying these key 

29 parameters is essential for building a defensible case in support of the assessment. It is very 

30 important for the licensee to justify the value or range of values used in the assessment to 

31 represent these key parameters. Several of the more popular sensitivity methods used in other 

32 performance assessments conducted at the NRC are presented, very briefly, below 

33 (NRC, 1999). It may be necessary for the licensee to use more than one approach in identifying 

34 the key parameters.
I-103

1 The licensee should focus on the pathways and radionuclides that are providing the greatest 

2 dose. If these pathways are modified or eliminated, the sensitivity analysis should be 

3 reevaluated to verify the important parameters for the analysis, consistent with the iterative 

4 nature of the “Decommissioning and License Termination Framework” (see Section 1.5 of this 

5 volume). For sites with a suite of radionuclides, the licensee may use expected concentrations 

6 or relative ratios of radionuclides to focus resources on the overall critical pathways and 

7 parameters. In addition, the licensee should evaluate the effects of uncertainty on the relative 

8 ratios. 

9 I.7.5.1 Deterministic Sensitivity Analysis 

10 Two types of sensitivity analysis techniques are widely used: deterministic and Monte Carlo. 

11 The first, deterministic sensitivity analysis, calculates the change in the output result (i.e., peak 

12 dose) with respect to a small change in the independent variables, one at a time. The following 

13 formula illustrates the normalized sensitivity coefficient calculated from a deterministic analysis.

S

X

d X

d

X

i

i

i i

=

























( )





14 Equation 9

15

where = sensitivity coefficient 

 = baseline value of the parameter 

 ( = peak dose for the baseline case

 = change in peak dose 

 = change in parameter 

th

th

S

X i

d X

X i

i 

i 

i

i

)





d

16

17

18 Variable transformations, such as normalization, used in this example, are described further 

19 below.

20

21 The advantage of the deterministic technique is that it is unambiguous in terms of demonstrating 

22 a cause and effect for the given conceptual model. The disadvantages are that at least one 

23 evaluation of the model should be performed for every independent variable, and the sensitivity 

24 result applies only locally (i.e., for one location in the space of all of the independent variables).

25 I.7.5.2 Statistical Sensitivity Analysis Techniques 

26 The techniques used herein (except deterministic analysis) rely on the use of the Monte Carlo 

27 method for probabilistically determining system performance. Statistical analysis of Monte Carlo 

28 results starts with a large pool of realizations (hundreds to thousands). These techniques 

29 determine sensitivities of the dependent variable (dose) to changes in the independent 

30 variables. The main advantage of these techniques is that they allow sensitivity to be 

31 determined over wide ranges of the independent variables, as opposed to the deterministic 

32 techniques that apply to only one point within the ranges. The disadvantage of statistical 

33 techniques is that it is often difficult to extract useful information on sensitivity except for a small 

34 set of the most important variables, because smaller sensitivities are obscured. A compilation 

35 of some of the more popular techniques for analyzing sensitivity from Monte Carlo results is 

36 presented below.
I-104

1 Usually, statistical sensitivity techniques have been applied to the set of peak doses drawn from 

2 the realizations. Sensitivity information from the ensemble of the peak doses is useful and is 

3 consistent with the “mean of the peaks” dose. For the “peak of the mean” dose, the set of 

4 doses drawn from the Monte Carlo runs at the time of the “peak of the mean” dose could be 

5 used.

6 SCATTER PLOT AND LINEAR REGRESSION ON ONE VARIABLE

7 In the scatter plot/single linear regression technique, peak TEDE is plotted versus each of the 

8 sampled input variables. This is often a good starting point for examining Monte Carlo results, 

9 because strong relationships between peak dose and the independent variables are often 

10 obvious. Single linear regression of Monte Carlo results may fail to show an unambiguous 

11 correlation, since other sampled parameters that affect the output are varying at the same time.

12 USE OF THE T-STATISTIC TO DETERMINE SIGNIFICANCE OF SINGLE LINEAR

13 REGRESSION PARAMETERS

14 The t-test estimates the confidence that an estimated parameter value differs from another 

15 value. In this case, it is used to determine if there is a specified (e.g., 95-percent) confidence 

16 that the slope (mi) of a single linear regression is different from zero (Benjamin and Cornell, 

17 “Probability, Statistics, and Decision for Civil Engineers” (Benjamin and Cornell, 1970).

18 The t statistic of the slope of the regression line is defined:

19

20

t m n

S x

S

i i

i =

2

2

,

21 Equation 10

22

23 where ti = t-statistic for regression coefficient i

24 mi = estimated value of regression coefficient (i.e., slope of the best-fit line for 

25 dose versus the independent variable i)

26 S = estimated standard deviation of dose

27 Si, x = estimated standard deviation of independent variable xi

28 n = number of samples

29 When the number of realizations is large, the t distribution may be represented by the normal 

30 distribution. The critical value to ensure 95-percent confidence that mi differs from zero under 

31 these conditions is 1.96. Equation 10 is used, therefore, to determine whether the absolute 

32 value of the t statistic for each independent variable is greater than 1.96. If not, then the 

33 hypothesis that the independent variable is significant is rejected.

34

35
I-105

1 PARTIAL RANK CORRELATION

2

3 The partial rank correlation coefficient measures the strength of the relationship between 

4 variables after any confounding influences of other variables have been removed. The partial 

5 rank correlation coefficient between X1 and Y, with the influence of X2 removed, is given by:

6

 

( )( )

( )( )



  

 

( ) X YX X Y X X YX

X X YX

1 2 2 2

1 1 2 2

1 2 2

1

2

1 1

=

−

− −

7 Equation 11

8

9 where ρ(X1YX2) = partial rank correlation coefficient between X1 and Y, with the influence 

10 of X2 removed

11 ρX1Y = rank correlation coefficient between X1 and Y

12 ρX1X2 = rank correlation coefficient between X1 and X2

13 ρYX2 = rank correlation coefficient between Y and X2

14 Stepwise Multiple Linear Regression

15 Stepwise multiple linear regression (stepwise regression) determines the most influential 

16 independent variables on output uncertainty according to how much each reduces the residual 

17 sum of squares (RSS) (SNL, 1991). The form of the regression equation is:

 y m x m x m x b = + + + + 1 1 1 1 n n18

... Equation 12

19

20 where y = dependent variable (i.e., peak dose)

21 xi = independent variables

22 mi = regression coefficients

23 b = intercept

24 The variables may be the raw variables, transformed variables (e.g., logarithms), or ranks (see 

25 Section I.7.5.3.2 of this appendix). The stepwise algorithm calculates the reduction in RSS for 

26 the independent variables in the order that gives the greatest reduction first. The regression 

27 coefficients (mi) are the partial derivatives of the dependent variable with respect to each of the 

28 independent variables; therefore, mi provides a measure of the relative change in output with 

29 respect to a change in the input variable, given that the other input variables are held constant.

30

31 NONPARAMETRIC TESTS

32

33 Nonparametric tests differ from regression and differential analyses in that they do not require 

34 fitting the data to prespecified functional form. The Kolmogorov-Smirnov test is one such test 

35 that determines whether a set of samples has been drawn from a specific distribution 

36 (NRC, 1988a). It is used to determine whether an independent variable is important by 

37 comparing a subset of the independent variable composed of the values from the highest 

38 category (e.g., 10 percent) of the peak TEDE realizations to the theoretical distribution of that 

39 independent variable. If the distributions are equivalent, then peak TEDE is not sensitive to the 

40 variable in question. Conversely, if the distributions are different, then the variable in question 

41 does have an effect on peak TEDE.
I-106

1 I.7.5.3 Variable Transformations and Their Attributes

2 Demonstrating the relationship among input and output variables can be enhanced by 

3 transforming the variables. This section describes some common variable transformations used 

4 in sensitivity analysis.

5 NORMALIZATION

6 In normalization, the input variable xi is transformed by dividing by its mean value (or another 

7 baseline, such as the median or the 90th percentile):

8

9

 x

x

x

i

i

i

*

10

=

Equation 13

11

12 Normalized variables are dimensionless and are scalar multiples of their baseline values. 

13 Dimensionless variables allow the comparison of sensitivities to other independent variables 

14 with different dimensions. Normalized variables are a natural outcome of sensitivity derived 

15 from regression of log-transformed variables. Such sensitivity measures describe only the 

16 relative change in the dependent variable (peak TEDE) to changes in the independent 

17 variables. Sensitivities calculated from normalized variables do not take into account the 

18 uncertainty in the independent variables.

19

20 RANK TRANSFORMATION

21 Rank transformation, a dimensionless transform, replaces the value of a variable by its rank 

22 (i.e., the position in a list that has been sorted from largest to smallest values) (Iman and 

23 Conover, “The Use of Rank Transform in Regression” (Iman and Conover 1979)). Analyses 

24 with ranks tend to show a greater sensitivity than results with untransformed variables and 

25 diminish the influence of the tails in highly skewed distributions.

26 LOGARITHMIC TRANSFORMATION

27 For situations in which input and output variables range over many orders of magnitude, it may 

28 be advantageous or even necessary to perform analyses on the logarithm of the variables 

29 instead of the variable values themselves. The log transformation is also valuable for creating 

30 regression equations, where the subprocesses of the model multiply each other to form the 

31 output variable. For the present situation, in which the dose calculation results from 

32 radionuclide releases from the waste form, transport through the geosphere, and uptake by 

33 humans, the processes are indeed largely multiplicative rather than additive. Log transforms, 

34 therefore, tend to give better fits to the Monte Carlo results than untransformed variables. The 

35 log transformation is generally used in conjunction with normalization.

36 STANDARDIZATION

37 The independent and dependent variables can be standardized by subtracting the mean and

38 dividing by the standard deviation; that is, 

39
I-107

x

x x

i

i

x

*

=

−



1 Equation 14

2 The advantage of standardization over normalization is that it inserts the approximate range of 

3 the variables into the sensitivity analyses. Therefore, a variable that has a large per-unit 

4 sensitivity, but is well known and has a narrow range, will have an increased sensitivity when 

5 standardized. Conversely, independent variables with wide ranges may show a reduced 

6 sensitivity when standardized.

7

8 Sensitivity measures based on standardized variables (standardized sensitivities) have the 

9 advantage of taking into account the uncertainty (in terms of the standard deviation) of the 

10 independent variable. This technique decreases the sensitivity if the range of the independent 

11 variable is large. Furthermore, the standardized sensitivities preserve the absolute values of 

12 peak TEDE, since the derivatives are divided by the standard deviation for the entire set of 

13 calculations rather than the mean peak TEDE at the evaluation point. 

14 I.7.6 Conclusions

15 Sensitivity analyses identify parameters of the models and assumptions that have the largest 

16 effect on the results. The licensee should use sensitivity results to determine if more 

17 information on key parameters is warranted to make a convincing case for the acceptability of 

18 the site for release. The sensitivity techniques discussed here evaluate sensitivity in different 

19 ways, and all have their strengths and weaknesses. A useful way to use sensitivity results is to 

20 employ several different techniques and then determine if a common set of parameters regularly 

21 turns out to be important.

22 Appendix Q contains additional information related to the NRC staff’s observations and 

23 recommendations related to consideration of uncertainty in dose and performance assessment 

24 analyses. Included are specific issues related to data representativeness (temporal and 

25 spatial), data correlations, use of generic data, consideration of alternative conceptual models, 

26 model integration, model abstraction, and preparation of sensitivity and uncertainty analyses.

27
I-108

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I-109

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I-110

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I-111

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I-112

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3 Washington, DC. July 1998(d).

4 — — — — —. Regulatory Guide DG-4006, “Demonstrating Compliance with the Radiological 

5 Criteria for License Termination, Draft.” NRC: Washington, DC. August 1998(e).

6 — — — — —. NUREG/CR-5621, “Groundwater Models in Support of NUREG/CR–5512.” 

7 NRC: Washington, DC. December 1998(f).

8 — — — — —. NUREG-1668, Volume 2, “NRC Sensitivity and Uncertainty Analyses for a 

9 Proposed HLW Repository at Yucca Mountain, Nevada, Using TPA 3.1.” NRC: 

10 Washington, DC. March 1999(a).

11 — — — — —. NUREG/CR-5512, Volume 2, “Residual Radioactive Contamination from 

12 Decommissioning, User’s Manual, Draft Report.” NRC: Washington, D.C. May 1999(b).

13 — — — — —. NUREG/CR-5512, Volume 3, “Residual Radioactive Contamination from 

14 Decommissioning, Parameter Analysis, Draft Report for Comment.” NRC: Washington, DC. 

15 October 1999(c).

16 — — — — —. NUREG/CR-5512, Volume 4, “Comparison of the Models and Assumptions 

17 used in the DandD 1.0, RESRAD 5.61, and RESRAD-Build Computer Codes with Respect to 

18 the Residential Farmer and Industrial Occupant Scenarios Provided in NUREG/CR-5512, Draft 

19 Report for Comment.” NRC: Washington, DC. October 1999(d).

20 — — — — —. NUREG-1575, Revision 1. “Multi-Agency Radiation Survey and Site 

21 Investigation Manual (MARSSIM).” NRC: Washington, DC. August 2000(a).

22 — — — — —. NUREG-1573, “A Performance Assessment Method for Low-level Waste 

23 Disposal Facilities: Recommendations of NRC’s Performance Assessment Working Group.” 

24 NRC: Washington, DC. October 2000(b).

25 — — — — —. NUREG/CR-6697, “Development of Probabilistic RESRAD 6.0 and 

26 RESRAD-BUILD 3.0 Computer Codes.” NRC: Washington, DC. December 2000(c).

27 — — — — —. NUREG-1727, “NMSS Decommissioning Standard Review Plan.” NRC: 

28 Washington, DC. September 2000(d).

29 — — — — —. NUREG-1717, “Systematic Radiological Assessment of Exemptions for Source 

30 and Byproduct Materials.” NRC: Washington, DC. June 2001.

31 — — — — —. NUREG-1720, “Re-evaluation of the Indoor Resuspension Factor for the 

32 Screening Analysis of the Building Occupancy Scenario for NRC’s License Termination Rule.” 

33 NRC: Washington, DC. June 2002.

34 — — — — —. NUREG-1620, Rev. 1, “Standard Review Plan for the Review of a Reclamation 

35 Plan for Mill Tailings Sites Under Title II of the Uranium Mill Tailings Radiation Control Act of 

36 1978.” NRC: Washington, DC. June 2003a.
1 — — — — —. NUREG-1640, “Radiological Assessments for Clearance of Materials from 

2

3

4

5

6

7

8

9

10

11

12

13

14

15

16

17

18

19

20

21

22

23

24

25

26

27

28

29

30

31

32

33

34 

35

36

Nuclear Facilities: Main Report (NUREG-1640, Volume 1).” NRC: Washington, DC. 

June 2003b.

— — — — —. NUREG/CR-6884, “Model Abstraction Techniques for Soil-Water Flow and 

Transport.” NRC: Washington, DC. December 2006. 

— — — — —. NUREG/CR-7127, “New Source Term Model for the RESRAD-OFFSITE Code 

Version 3.” NRC: Washington, DC. December 2013.

— — — — —. Regulatory Guide 4.7, Revision 3 (draft was issued as DG-4021 on 

December 30, 2011). “General Site Suitability Criteria for Nuclear Power Stations.” NRC: 

Washington, DC. March 2014.

— — — — —. NUREG-2175. “Guidance for Conducting Technical Analyses for 

10 CFR Part 61, Draft Report for Comment.” NRC: Washington, DC. March 2015(a). 

— — — — —. NUREG/CR-7189, “User’s Guide for RESRAD-OFFSITE,” Washington, DC. 

April 2015(b). 

— — — — —. NUREG/CR-7267, “Default Parameter Values and Distributions in RESRADONSITE V7.2, RESRAD-BUILD V3.5, and RESRAD-OFFSITE V4.0 Computer Codes,” also 

ANL/EVS/TM-20/1. NRC: Washington, DC. 2020a. 

— — — — —. NUREG/CR-7268, “User’s Manual for RESRAD-OFFSITE Code Version 4, 

Volume 1—Methodology and Models Used in RESRAD-OFFSITE Code,” also ANL/EVS/

TM-19/2, Volume 1. NRC: Washington, DC. 2020b.

— — — — —. NUREG/CR-7268, “User’s Manual for RESRAD-OFFSITE Code Version 4, 

Volume 2—User’s Guide for RESRAD-OFFSITE Version 4,” also ANL/EVS/TM-19/2, Volume 2. 

NRC: Washington, DC. 2020c.

OECD. Method 106. “OECD Guideline for Testing of Chemicals, Adsorption-Desorption Using a 

Batch Equilibrium Method,” OECD: Paris, France. 

OECD, 2012, “Thermodynamic Sorption Modeling in Support of Radioactive Waste Disposal 

Safety Cases NEA Sorption Project Phase III,” NEA No. 6914, Nuclear Energy Agency, 

Organisation for Economic Co-operation and Development, Radioactive Waste Management, 

ISBN 978-92-64-17781-9, Paris, France. 

Oak Ridge National Laboratory (ORNL). ORNL-5786, “A Review and Analysis of Parameters 

for Assessing Transport of Environmentally Released Radionuclides Through Agriculture.” 

ORNL: Oak Ridge, TN. September 1984. 

Office of Management and Budget (U.S.) (OMB), Washington, DC. “Economic Analysis of 

Federal Regulations under Executive Order 12866.” OMB: Washington, DC. 

January 11, 1996. 

Relyea, J. F., R. J. Serne and D. Rai, “Methods for Determining Radionuclide Retardation 

Factors: Status Report,” PNL-3349, Pacific Northwest Laboratory, Richland, WA, 1980. 

I-113

37
I-114

1 Roy, W.R., I.G. Krapac, S.F.J. Chou, and R.A. Griffin, “Batch Type Procedures for Estimating 

2 Soil Adsorption for Chemicals,” U.S. EPA Technical Resource Document, 

3 EPA/530/SW 87/006 F, U.S. EPA, Washington, DC, 1992.

4 Sandia National Laboratories. SAND90-7103, “Sensitivity Analysis Techniques and Results for 

5 Performance Assessment at the Waste Isolation Pilot Plant.” SNL: 1991.

6 Serne, R.J., J.F., Relyea, “Waste/Rock Interactions Technology Program: The Status of 

7 Radionuclide Sorption-Desorption Studies Performed by the WRIT Program,” PNL-3997, Pacific 

8 Northwest Laboratory, Richland, WA. April 1982. 

9 http://www.osti.gov/bridge/servlets/purl/5361396-3DRJ74/5361396.pdf

10 Sheppard, M.I. and D.H. Thibault, 1990. “Default Soil Solid/Liquid Partition Coefficients, Kds, 

11 for Four Major Soil Types: A Compendium.” Health Physics. 59(4): 471–482.

12 Sheppard, M.I., D.I. Beals, D.H. Thibault, P. O’Connor, “Soil Nuclide Distribution Coefficients 

13 and Their Statistical Distribution,” Limited Report AECL-8364, Chalk River Nuclear Labs, Atomic 

14 Energy of Canada, Chalk River, Ontario, 1984.

Sheppard, S.C., W.G., Evenden, “Comparison of Partition Coefficients for 15 54Mn and 

16 Soil-Extractable Mn, Including Relationship to Plant Uptake,” Canadian Journal of Soil Science, 

17 69: 351-365, 1989.

18

19 Sheppard, S.C., M.I., Sheppard, “Impact of Correlations on Stochastic Estimates of Soil 

20 Contamination and Plant Uptake,” Health Physics, 57: 653–657, 1989.

21 Smith, A.E., P.B. Ryan and J.S. Evans, 1992. “The Effect of Neglecting Correlations When 

22 Propagating Uncertainty and Estimating the Population Distribution of Risk.” Risk Analysis, 

23 Vol. 12, No. 4, 1992.

24 Sullivan, T., 1996. “DUST-MS Model Equations for Waste Form Leaching and Transport with 

25 Ingrowth Due to Radioactive Decay.” Brookhaven National Laboratory, Environmental and 

26 Waste Technology Center: Upton, New York. May 1996.

27 Swedish Radiation Protection Institute (SSI). Technical Report No. 6, “Development of a 

28 Reference Biospheres Methodology for Radioactive Waste Disposal, Final Report of the 

29 Reference Biospheres Working Group of the BIOMOVS II Study.” SSI: Stockholm, Sweden. 

30 September 1996.

31 Thomas, D., W. Hareland, W. Beyeler, D. Brosseau, J. Cochran, T. Feeney, T. Brown, P. Pohl, 

32 R. Haaker and B. Fogleman, 2000. Process for Developing Alternate Scenarios at NRC Sites

33 Involved in D&D and License Termination, Sandia Letter Report, Sandia National Laboratories,

34 January 2000. (ADAMS Accession No. ML003698994.)

35 Yu, C., 1999. ANL/EAD-4. “RESRAD Family of Codes and Comparison with Other Codes for 

36 Decontamination and Restoration of Nuclear Facilities.” 

37 Decommissioning and Restoration of Nuclear Facilities, M.J. Slobodien (ed.). Medical Physics 

38 Publishing: Madison, WI. 1999.
I-115

1 Yu, C., et. al. “User’s Manual for RESRAD Version 6,”, Environmental Science Division, 

2 Argonne National Laboratory, July 2001.

3 Yu, C., E. Gnanpragasam, J.J. Cheng, and B. Biwer, “Benchmarking of RESRAD-OFFSITE: 

4 Transition from RESRAD (onsite) to RESRAD-OFFSITE and Comparison of the 

5 RESRAD-OFFSITE Predictions with Peer Codes,” ANL/EVS/TM/06-3, DOE/EH-0708, 

6 Environmental Science Division, Argonne National Laboratory, May 2006.

Institute of Electrical and Electronics Engineers (IEEE) Standard 830-1984, “IEEE Guide for 

Software Requirement Specifications.” IEEE: Piscataway, N.J. 1984.

ASSESSMENT STRATEGY FOR BURIED MATERIAL

J-1

1 J.1 Generic Description of Situations

2 This appendix describes examples of situations warranting alternative conceptual models which 

3 may occur at sites undergoing decommissioning. The first situation discusses buried 

4 radioactive material or subsurface contamination at a site. The second situation discusses 

5 subsurface structures or basements which are backfilled and left as part of the end state of the 

6 site. 

7

8 J.1.1 Buried Radioactive Material or Subsurface Soil Contamination

9 A licensed site has the following characteristics for purposes of the conceptual model for buried 

10 radioactivity: 

11 • It has buried radioactive material or it has subsurface soil contaminated with residual

radioactivity beneath clean cover soils or fill material1 12 . (Sites prior to the 1980’s may

13 have 10 CFR 20.304 burial units that were in use before the January 28, 1981, effective

14 date of the final rule, “Standards for Protection Against Radiation; Burial of Small

15 Quantities of Radionuclides” (45 FR 71761), when Section 20.304 was removed from

16 the regulations).

17 • The site has no other sources of residual radioactivity, or the total dose associated with

18 all sources of residual radioactivity must be addressed and be below the dose limit (see

19 Section 2.7 of this NUREG report).

20 • Information on the inventory (e.g., radionuclide concentrations, disposal dates, waste

21 form) may be limited. However, the licensee has enough information to estimate or to

22 bound the total activity or concentrations of radioactive material present (see Section 4.0

23 for information on the HSA, Section 4.2 for information on characterization surveys, and

24 Section I.2 of Appendix I for information about source term abstraction).

25 • It is known that the material is deep enough that an external dose at the surface is not

26 possible in the current configuration.

27 • The site does not have any of the physical limitations that would make the use of a

28 RESRAD code to represent the conceptual model inappropriate (see Appendix I,

29 Section 4, Table I-9).

30 • The site is underlain by an unsaturated layer and aquifer. If the aquifer is potable, and

31 has sufficient yield, the groundwater pathway should be considered for drinking water

32 and/or irrigation. If the aquifer is not potable, or has insufficient yield, then the licensee

33 may present arguments for why the groundwater (drinking water) pathway does not

34 need to be considered, although the licensee may need to consider other uses of

35 groundwater (e.g., irrigation).

1 Licensees should provide support that the cover or fill materials contain no radioactive contamination. If potentially 

contaminated fill materials are used (e.g., reuse of rubblized concrete from the site as fill), the licensee must 

adequately survey the materials before reusing them for debris and consider the dose contributions of residual 

radioactivity present in fill materials. See Section G.3.2. for additional details on radiological surveys associated 

with fill materials.
J-2

1 • If the soil at the site is assumed to be capable of growing crops without significant soil 

2 engineering, then plant ingestion should be considered.

3 • Arguments may be presented for why certain pathways do not need to be considered.

4 Figure J.1 shows a simple conceptual figure of the site with buried radioactive material.

5

6 Figure J.1 Conceptual Illustration of Buried Material in Subsurface Soil at a Site

7

TCover

TWaste

Unsaturated Zone

Cap or Uncontaminated Soil Cover

Residual Radioactivity or Contaminated Zone

Saturated Zone

TUZ

TSZ
J-3

1 J.1.2 Backfilled Basements

2 In some cases, the site may have below-grade basements that will be backfilled and remain 

3 after decommissioning. For these sites, there may be some residual radioactivity left on the 

4 concrete surfaces prior to backfill, and a different conceptual model is warranted to derive site5 specific DCGLs for those concrete surfaces (see Figure J.2).

6

7

8 Figure J.2 Conceptual Model for Remaining Basements which have Residual 

9 Radioactivity and are Backfilled with Clean Fill 

10

Uncontaminated soil cap T=3 m

Residual radioactivity 

on concrete surfaces

Remaining basement backfilled with clean fill
J-4

1 J.2 Conceptual Models, Exposure Scenarios, Exposure Pathways, and Critical 

2 Group

3 To develop the exposure scenario(s) for the critical group, the analyst should address the 

4 following questions:

5

6 • How does the residual radioactivity move through the environment?

7 • Where can humans be exposed to the environmental radiological concentrations?

8 • What are the exposure group’s habits that will determine exposure? (What do they eat

9 and where does it come from? How much do they eat? Where do they get water and

10 how much water do they drink? How much time do they spend on various activities?)

11 In the absence of site-specific information, residential scenarios are generally assumed in the 

12 following examples for illustrative purposes, although arguments can be presented for why other 

13 scenarios are reasonably foreseeable and residential scenarios are less likely but plausible (or 

14 implausible). For more information on developing exposure scenarios, see Appendix I and 

15 Appendix M of this volume.

16

17 J.2.1 Buried Material Conceptual Models, Exposure Scenario, Exposure Pathways, and

18 Critical Group

19 For sites with buried material, one conceptual model could assume that all the buried 

20 radioactive material at depth was instead at the surface (Figure J.3). That is, this scenario takes 

21 no credit for the clean cover (the clean cover is removed in the conceptual model) and models 

22 leaching of the radionuclides into the groundwater which is then used by a resident, as well as 

23 considering direct radiation from exposure to the soil at the surface. Care should be taken to 

24 ensure the modeled vadose (unsaturated) zone thickness is equal to or less than the actual 

25 vadose zone thickness (i.e., the contamination zone should not be moved to the top of the cover 

26 but should remain at the actual elevation).

27

28
J-5

1 Figure J.3 Simplified Conceptual Model of Human Disturbance into Buried Residual 

2

3

4

Radioactivity (the top Panel Shows the Original Configuration of Residual 

Radioactivity and Human Disturbance Event [Construction of a Home with 

Basement]; the bottom Panel Shows a Conceptual Model with the Cover 

5 Assumed to have been Removed for Simplification)

6

House

TWaste

Unsaturated Zone

Residual Radioactivity or Contaminated Zone

Saturated Zone

TUZ

TSZ

House

TCover

TWaste

Unsaturated Zone

Soil Cover

Residual Radioactivity or Contaminated Zone

Saturated Zone

TUZ

TSZ

Actual location 

of residual 

radioactivity

The cover is 

assumed to be 

removed, 

exposing the 

waste to the 

surface but the 

thickness of the 

unsaturated 

and saturated 

zones remains 

the same. 
J-6

1 Alternatively, credit for the cover may be taken and two exposure scenarios can be developed 

2 (1) leaching of the radionuclides from their buried position to groundwater, and (2) intrusion into

3 the buried residual radioactivity by basement construction (or other large-scale excavation

4 scenario) or well drilling. In the second scenario the displaced soil, which includes part of the

5 residual radioactivity, is spread across the surface (Figure J.4 and Figure J.5). All the exposure

6 pathways are considered for the second scenario involving intrusion into the buried residual

7 radioactivity and, although not all the source term is in the original position, leaching will occur

8 both from the remaining buried residual radioactivity (if there is any) and the surface soil.

9

10 Figure J.4 Conceptual Model of Human Disturbance into Buried Residual Radioactivity 

11 (Showing some of the Waste Mixed with Clean Soil on the Surface [Conc1] and 

12 some Waste Remaining at Depth [Conc2]) 

13

14

Conc1

Conc2

TB=3 m

House

Unsaturated Zone

Soil Cover

Residual Radioactivity or Contaminated Zone

Saturated Zone

Original 

location of 

residual 

radioactivity

Residual 

radioactivity and 

clean soil cap 

material are brought 

to the surface (and

assumed to be well 

mixed)
J-7

Conc2

1 Figure J.5 Conceptual Model of Buried Radioactivity with a Well Drilled through the

2 Radioactive Material and the Drilling Spoils Spread out on the Surface

3 J.2.2 Additional Considerations for Backfilled Basement Conceptual Models, Exposure 

4 Scenarios and Pathways

5 For backfilled basements a similar approach can be considered as that for buried material. An 

6 intrusion event (well-driller scenario) that includes leaching into the groundwater can be 

7 developed. In the well-driller scenario, the backfilled basement is assumed to remain 

undisturbed except for a well drilled through a portion2 8 of the backfilled substructure for drinking 

9 water. 

10

11 The well-drilling scenario considers (i) pathways associated with use of contaminated 

12 groundwater from an onsite well and (ii) direct (and indirect) exposure to drilling spoils that are 

13 brought to the surface during the installation of the onsite well. These two exposure scenarios 

14 have different conceptual models. For the groundwater scenario (Figure J.6), licensees may be 

15 able to make simplifying assumptions if the simplifications do not underestimate the potential 

2 Because it is more pessimistic to assume that the driller stops drilling after encountering reinforced concrete 

(would lead to higher concentrations and dose with less clean material located below the concrete being brought to 

the surface and mixed with the residual radioactivity in the reinforced concrete), this assumption is acceptable. 

Arguments can be presented on why the driller might be able to drill through the reinforced concrete floor and bring 

up additional relatively clean material between the concrete floor and the drinking water aquifer. This would 

potentially allow cleaner soil below the concrete basement to be brought to the surface and mixed with the residual 

radioactivity associated with the concrete surface.

Well drilling spoils with 

residual radioactivity and 

clean soil cap are brought to 

the surface (and assumed to 

be well mixed).

Conc1

Original location of residual 

radioactivity

Uncontaminated soil cap or 

fill materials

Unsaturated Zone

Saturated Zone
J-8

1 dose (e.g., licensees may be able to assume all the residual radioactivity exists as a 

2 contaminated zone layer of soil at the elevation of the floor of the underground structure closer 

3 to the water table aquifer). To simplify the conceptual model further, the concrete floor and 

4 walls can be assumed to not affect flow, allowing more rapid transport of residual radioactivity to 

groundwater3 5 . If the simplifications lead to excessive pessimism, licensees have the option to 

6 model the actual configuration of residual radioactivity, degradation of cementitious materials, 

7 and flow and transport of residual radioactivity to groundwater. However, given the complexity 

8 and uncertainty in more realistically modeling these processes, it is expected that uncertainty 

9 will likely need to be managed with conservative assumptions. The licensee should 

10 communicate with the NRC early in the process to ensure that a technically acceptable 

11 approach is developed for these more complex problems. 

12

13 If the concrete floors and walls are assumed to be intact, then water could fill the basement and 

14 cause contamination of near-surface soils. Depending on the depth of the buried residual 

15 radioactivity below ground surface and mechanisms for upward transport, residual radioactivity 

16 transported near the surface could present additional exposure pathways and scenarios that 

17 need to be considered (e.g., exposure to residual radioactivity transported to the surface and/or 

18 exposure to residual radioactivity from basement excavation scenario if radioactivity below 3 m 

19 (10 ft) of the surface can be transported within 3 m (10 ft) of the surface). Licensees could take 

20 credit for measures to ensure that water does not fill up in the basement to eliminate this 

21 scenario from consideration. Additionally, licensees can take credit for cover materials and 

22 depth of the residual radioactivity in eliminating surface exposure pathways. The adequate 

23 depth will depend on site–specific information (e.g., water table level and variations in the water 

24 table) and will need to be evaluated on a case-by-case basis. 

25

26 For the well-driller direct exposure pathway (Figure J.7), a well is assumed to be installed 

27 through the fill material, and drilling spoils along with a portion of the concrete floor are brought

28 to the surface. The well installation may occur before any leaching from concrete occurs. The 

29 residual radioactivity in the concrete of the basement floor that is contacted by the borehole 

30 during installation of the well is assumed to be mixed with the column of drilling spoil fill material 

31 above the floor surface, brought to the ground surface, and spread. 

32

33

34

3 If residual radioactivity associated with the basement is below the water table, the licensee should ensure that the 

conceptual model for the site is compatible with the conceptual model of the selected code. Commonly used 

decommissioning dose modeling codes assume the residual radioactivity is in the vadose zone. More complex 

site-specific modeling may be warranted, or adjustments can be made to existing models/codes, if the limitations 

of the selected codes are well understood and the adjustments can be shown to not result in an underestimation of 

dose. See Appendix I for additional information.
J-9

1 Figure J.6 Conceptual Model for Groundwater Exposure Pathway from Remaining 

2

3

Basements which have Residual Radioactivity and are Backfilled with Clean 

Fill (where Residual Radioactivity is Assumed to be Released from the Concrete 

4 and Leach into the Groundwater) 

5

6

Uncontaminated soil cap T=3 m

Residual 

radioactivity 

on concrete 

surfaces 

Radioactivity leaches 

unobstructed from concrete 

walls and floor (residual 

radioactivity on walls can be 

assumed to be at the 

elevation of the floor) 
J-10

1 Figure J.7 Conceptual Model for Well-Driller Scenario 

2 Additional scenarios that may need to be considered for large backfilled subgrade structures 

3 (e.g., containment basements, auxiliary building basements, and/or turbine basements at a 

reactor site) may be described as large-scale excavation4 4 . A large-scale excavation scenario, 

5 which would bring up more residual radioactivity compared to a well-driller scenario, may need 

6 to be considered given the depth, geometry, and materials associated with the source. In this 

7 scenario, a portion or all the remaining backfilled basement walls and/or floor are removed in 

8 the future after license termination. In this scenario, all the residual radioactivity is assumed to 

9 remain on the concrete to maximize the concentrations in the concrete brought to the surface. 

10 An example conceptual model is shown in Figure J.8. Note that this figure is for illustrative 

11 purposes and does not represent a specific suggested geometry or scale of an excavation.

12 The dose to the construction worker, disposal facility worker, or other member of the public who 

13 may be exposed to the residual radioactivity on the excavated concrete and/or fill material would 

14 need to be considered. The radionuclide concentrations (Bq/kg or pCi/g) in the excavated 

15 concrete is directly related to the ratio of concrete surface area to concrete volume excavated. 

16 For substructures that have thick walls and for where there the contamination is limited to the 

17 surface, assuming a partial excavation that includes only the walls with the minimum thickness 

18 would result in a higher concentration. Arguments could be presented by the licensee why a 

19 large excavation scenario is not reasonably foreseeable or why certain pathways are less likely 

20 but plausible, or implausible and are simply used to inform decision-making, or do not need to 

21 be considered (see Chapter 5 for additional information on compliance, informative, or 

22 eliminated exposure scenarios).

4 For the case of large backfilled subgrade structures, the depth of the basement is typically greater than 3 m (10 ft), 

and the home basement excavation scenario can be eliminated from consideration for multiple reasons (depth of 

waste and potential difficulty constructing a residential basement through reinforced concrete structures). 

Uncontaminated soil cap T=3 m

Residual radioactivity is 

assumed to remain on 

concrete surfaces or within 

concrete if volumetric 

contamination is present. 

Well installation intersects 

with contaminated concrete 

surface. 

Contaminated drilling 

spoils from mixing clean 

fill with contaminated 

concrete
J-11

1 Figure J.8 Large-Scale Excavation Conceptual Model for Remaining Basement which 

2 has Residual Radioactivity and is Backfilled

3 J.3 Examples of Analyses Using the RESRAD Family of Codes

4 The RESRAD computer code was developed by Argonne National Laboratory (ANL) for the 

5 U.S. Department of Energy (DOE) to calculate site-specific residual radiation guidelines and 

6 radiation dose to future hypothetical onsite individuals at sites contaminated with residual 

7 radioactive material. DOE adopted the RESRAD-ONSITE code in Order 5400.5 for derivation 

8 of soil cleanup criteria and dose calculations, and it is widely used by DOE, other Federal 

agencies, and industry.5 9 

10

11 The RESRAD family of codes have an assumed conceptual model; therefore, the analyst only 

12 must determine if the assumed conceptual model is appropriate for the problem. However, 

13 unlike DandD, the RESRAD family of codes do not have prescribed exposure scenarios. The 

14 analyst should develop the exposure scenario by switching on or off various exposure pathways 

15 or adjusting parameters. For the standard resident farmer exposure scenario commonly used 

16 by the NRC staff, all the exposure pathways should be switched on for unrestricted use cases, 

17 with the exception of the radon pathway. The analyst should justify excluding any of the other 

18 pathways. For example, if it can be shown that the groundwater at the site cannot be used

19 because of either poor ambient water quality (e.g., salinity) or low yields, the licensee may 

20 justify elimination of the groundwater pathway. A finding that the groundwater is unsuitable is 

5 RESRAD-ONSITE is referred to as RESRAD in DOE Order 5400.5

Uncontaminated soil cap T=3 m

Residual radioactivity 

on concrete surfaces 

Large scale 

excavation of 

concrete

Residual radioactivity 

on excavated concrete 

and/or fill material 
J-12

1 typically made in coordination with State agencies. For more information on eliminating

2 pathways, see Appendix I, and Appendix M.

3 The RESRAD family of codes require that the radioactive inventory be input as a source 

4 concentration. Because the codes are designed for conducting site-specific analyses, it is 

5 expected that, for most analyses, the analyst will have data on radionuclide concentrations at 

the site.6 6 It should be appropriate to use the arithmetic average of the radionuclide 

7 concentration in the analysis (note this also includes any interspersing clean soil). The 

8 RESRAD-ONSITE and RESRAD-OFFSITE codes allow the user to input information on the 

9 area and thickness of the residual radioactivity (i.e., these are not fixed, although defaults are 

10 provided). For surface residual radioactivity (≤0.9 m (3 ft), which is the default root depth), the 

11 site-specific mean concentration, area of residual radioactivity, and thickness of the residual 

12 radioactivity can be input directly in the code. For deeper residual radioactivity, or if the residual 

13 radioactivity is capped (such as with burials), assumptions should be made about how much 

14 waste may be brought to the surface and how it may be mixed with uncontaminated soil. 

15 J.3.1 Modeling Buried Material Using RESRAD

16 Analyzing the conceptual model in Figure J.4 using RESRAD requires two simulations. During 

17 the first simulation, it is assumed that a small volume of waste due to construction of a house 

basement (600 m3(21,200 ft3 18 )) is brought to the surface and spread out over an area to a depth 

19 of 0.9 m (3 ft). The source concentration for this first simulation could be derived by multiplying 

20 the average concentration of the buried residual radioactivity by the fraction of the total 3 m

21 home (basement) construction excavation thickness that is assumed to contain residual 

22 radioactivity (i.e., credit is taken for mixing and dilution of the residual radioactivity in a larger 

23 volume of excavated soil, a portion of which is assumed to be free of residual radioactivity). The 

24 concentrations could take credit to account for radioactive decay. An area that could be used in 

the first simulation is 700 m2(7,550 ft2) (i.e., 600 m3(21,200 ft3 25 ) divided by 0.9 m (3 ft)). The 

26 assumed contaminated thickness would then be 0.9 m (3 ft). However, the licensee should 

27 perform a sensitivity analysis to understand the impact of assumptions on the distribution of 

28 residual radioactivity on the surface (area and thickness of the contaminated zone) to dose. If 

29 the dose results are significantly different, the licensee should consider changing the 

30 assumptions on the distribution of contamination to ensure that the doses are not significantly 

31 underestimated. The second simulation looks at effects from exposure from the remaining 

32 residual radioactivity. If the residual radioactivity is presently capped, it can be assumed to be 

33 covered for the second simulation, unless there are reasons to model the situation with the 

34 cover removed (e.g., high soil erosion rate). The source concentration for the second simulation 

35 should be the mean concentration for the waste area, including interspersing clean soil. The 

36 area and thickness of the residual radioactivity used in the second simulation would be based 

37 upon the true existing source configuration. Accordingly, to use this approach, the analyst will 

38 have to know something about the waste zone configuration. Additionally, the analyst may 

39 need to consider the geometry of the buried residual radioactivity in relation to the receptor 

40 located in the basement and other floors of the home. The RESRAD-ONSITE and RESRAD41 BUILD codes generically assume that residual radioactivity is located directly underneath the 

42 residence (residual radioactivity surrounding a basement of a residence is not explicitly 

6 RESRAD-ONSITE and RESRAD-OFFSITE are primarily designed to look at radioactively contaminated soils; 

therefore, for analyses involving other types of wastes, the analyst will have to make some assumptions about the 

waste form and how the radionuclides will be released from this waste form. These assumptions should be clearly 

laid out.
J-13

1 considered). It may be possible to use a shielding code to calculate an effective shielding factor 

2 to implicitly consider such an exposure scenario.

3 Analyzing the conceptual model in Figure J.5 would also require a dual simulation in RESRAD 

4 like what is described for the conceptual model in Figure J.4, except the concentration and 

5 volume of the material brought to the surface is based on drilling spoils as opposed to house 

6 construction. For example, one could assume a well that has a diameter typical for residual 

7 drinking water use in the area and is sufficiently deep to reach the aquifer is drilled through the 

8 buried radioactive material. The drilling spoils are then spread across the surface to a certain 

9 thickness (e.g., 0.15 m). The licensee may take credit for mixing the buried radioactive material 

10 with the clean material when it is brought to the surface in determining the concentration of the 

11 drilling spoils. Again, the licensee should perform a sensitivity analysis to understand the 

12 impact of assumptions on the distribution of residual radioactivity on the surface (area and 

13 thickness) to dose. The second simulation analyzes the effects from leaching of the remaining 

14 residual radioactivity into the groundwater.

15 An alternative to using the dual simulation approach is to simply assume that the waste is 

16 uniformly distributed over the source volume, taking no credit for the cover as shown in the 

17 lower portion of Figure J.3 (i.e., by assuming that the cap is not present). This should provide a 

18 conservative estimate of the dose. The thickness of the vadose zone should reflect the actual 

19 distance to the aquifer to ensure that the travel time to the aquifer is not significantly 

20 overestimated, particularly if short-lived radionuclides are present. Using this simpler approach, 

21 the analyst would use the mean concentration as the source concentration. In using either of 

22 these approaches, it may be important to assess the appropriateness of assuming that the 

23 activity is uniformly distributed over the waste volume (e.g., if elevated areas are present, it may 

24 be appropriate to evaluate the dose contributions of the elevated areas [see Sections A.5.2 and 

25 I.2.3 for additional information on evaluation of elevated areas]).

26 If all that is known is the source inventory (activities), such as at some old burial sites, the 

27 source concentration of the material brought to the surface can be calculated by dividing the 

28 total inventory by the total mass of the waste (i.e., volume of the waste multiplied by density of 

29 the waste) and then multiplying this average concentration of buried residual radioactivity by the 

30 fraction of the total 3 m home (basement) construction excavation thickness that is assumed to 

31 contain residual radioactivity (i.e., credit is taken for mixing and dilution of the residual 

32 radioactivity in a larger volume of excavated soil a portion of which is assumed to be free of 

33 residual radioactivity). The concentrations could take credit to account for radioactive decay. 

34 The density used to calculate the concentration should be the same density assumed in the 

35 analysis.

36 J.3.2 Modeling Backfilled Basements Using RESRAD

37 Analyzing the conceptual models for backfilled basements also require dual simulations in 

38 RESRAD. For example, in Figure J.7, the drilling spoils can be modeled as a contaminated 

39 zone at the surface. The well drilling may be assumed to stop once it hits the concrete floor, but 

40 the depth should be sufficient enough to bring up all surface or volumetric contamination in the 

41 concrete floor that it intersects. The concentration of the drilling spoils would be the result of 

42 mixing of the radioactivity on the concrete with the fill material above it which the drill displaces. 

43 The volume of material brought to the surface is calculated based on the borehole area and the 

44 depth of the basement floor. The drilling spoils are then spread at the surface. The licensee 

45 should perform a sensitivity analysis to understand the impact of assumptions on the distribution 

46 of residual radioactivity on the surface (area and thickness) to dose. 
J-14

1 The second simulation would analyze the effect of leaching into the groundwater from the 

2 radioactivity on the concrete walls and floors. The concrete walls and floor can be assumed to 

3 not affect flow of radioactivity and the radioactivity on the floor can be modeled as a layer of 

4 contaminated soil at the depth of the floor. To simplify the simulation, the contamination on the 

5 walls may also be assumed to be on the floor. A reasonable contaminated zone thickness 

6 should be selected that will not underestimate the potential dose associated with the buried 

7 residual radioactivity and should consider whether there is surface or volumetric contamination 

8 of the concrete. The RESRAD model will produce a Dose-to-Source Ratio (DSR) in units of 

9 mrem/y per pCi/g for each radionuclide. The licensee should confirm whether surficial or 

10 volumetric residual radioactivity is present on building surfaces and should be able to measure 

11 and appropriately account for any residual radioactivity at depth to ensure that potential dose is 

12 not underestimated. If only surficial residual radioactivity is confirmed by radiological survey to 

13 be present and concrete surface DCGLs are needed to perform the final status survey, the 

14 licensee will need to convert the volumetric DCGLs to surface DCGLs. This conversion factor 

15 will require an assumed thickness to the contaminated zone, along with assumed densities of 

16 the fill and concrete material. 
DOSE MODELING FOR PARTIAL SITE RELEASE

K-1

1 K.1 Dose Modeling Considerations for Partial Site Release

2 This appendix consists of the technical guidance for reviewing the release of a portion of a site 

3 before final termination of the entire site, a process called partial site release (PSR). The review 

4 is carried out under 10 CFR Part 20, “Standards for Protection against Radiation,” Subpart E, 

5 “Radiological Criteria for License Termination.” This guidance is generally applicable to 

6 Decommissioning Groups 2–5.

7 The NRC has developed the guidance in this appendix to encompass the needs of the most 

8 complex situations, but the licensee should tailor the specific informational needs for a PSR 

9 request to the complexity and safety significance of the proposed action. This appendix is split 

10 into three sections. The first section, which complements Chapter 5 of this volume, details the 

11 review criteria to be used in assessing compliance with Subpart E. The second section 

12 provides technical information, which supplements the guidance in Appendices H and I. The 

13 third section contains two hypothetical simple examples of PSR considerations.

14 The guidance is focused on PSR requests that occur before the NRC approves the DP, but it is 

15 also applicable for requests for phased releases after DP approval (see Section K.1.8).

16

17 K.1.1 Partial Site Release Reviews

18 For a PSR, dose modeling is not necessarily limited to the partial site but can also include 

19 residual radioactivity outside the partial site. Other areas on the site may contribute direct 

20 radiation or have natural processes that may move residual radioactivity to the area requested 

21 for release. For purposes of this section, “offsite sources” are potential sources of exposure that 

22 are not on the partial site but still on affected areas under (or previously under) the control of the 

23 licensee. For example, a licensee may have impacted groundwater under a portion of the site. 

24 At the time of the partial site request, the groundwater under the partial site may not be 

25 impacted. However, possible movement of this impacted groundwater (an “offsite source” of 

26 residual radioactivity) from the remaining site to the area requested to be released must be 

27 considered.

28 In addition to compliance analyses for the PSR, there should be evaluations of potential 

29 prospective analyses. These analyses should evaluate how the PSR could affect the license 

30 termination of the licensed site, including any additional PSRs. For example, releasing an area 

31 of the site at higher DCGLs than is likely for the rest of the site could constrain future 

32 decommissioning, forcing the licensee to use DCGLs for the rest of the site that are below what 

33 they could have been if the PSR had never occurred.

34

35 K.1.2 Incorporation into Review Process

36 The licensee may use either a screening or a site-specific dose assessment to show 

37 compliance with Subpart E. Although it may use generic screening analyses to create the 

38 DCGLs of PSRs, the overall review should be a site-specific review. The NRC reviewer should 

39 use the appropriate section of Chapter 5 to review the assessments and the additional 

40 considerations for source terms, exposure scenarios, and pathway identification detailed below.

41

42
K-2

1 K.1.3 Evaluation Information

2 The difference between a dose assessment for license termination of the entire site license and 

3 a dose assessment for a PSR is that other sources under control of the licensee may affect the 

4 potential dose on the PSR. In license termination of the entire site, when the site is released for 

5 unrestricted use, there are no offsite sources remaining under the control of the licensee to 

6 affect the projected dose for residents or workers using the site. In contrast, after a PSR, the 

7 remaining licensed site may still be operating and thus have dose contributions to the critical 

8 group receiving doses from the PSR, such as from surface water runoff or groundwater

9 migration. In addition, sources on the remaining licensed site may result in doses to the public 

10 after unrestricted release of the remaining site in the future, such that members of the PSR 

11 critical group receive doses from sources on both the PSR and on the remaining (now 

12 terminated) site.

13

14 The NRC staff should review the licensee’s assessment of offsite sources that may influence 

15 the dose analysis, and the evaluation of these sources would be like that of a source on the 

16 partial site. The development of the appropriate exposure scenarios for compliance evaluation 

17 should identify which sources NRC staff should focus on. The primary areas of additional 

18 consideration given to PSR cases in developing reasonable exposure scenarios are the 

19 following:

20

21 • How does the licensed site or a previous PSR influence the dose on the PSR

22 (e.g., effluent releases, groundwater plumes, future combined use)?

23 • How could the PSR influence dose estimates for the licensed site during its

24 decommissioning?

25 • How does the PSR influence previous PSRs (e.g., possible effects on the PSR’s final

26 DCGLs to limit the impacts on the previous PSR, so that the potential dose on the

27 previous PSR does not exceed Subpart E criteria)?

28 K.1.4 Development and Identification of Partial Site Release Exposure Scenarios

29 Based on the questions above, exposure scenarios can be divided into two categories: 

30 compliance and prospective. Analysis of both categories of exposure scenarios should assist in 

31 establishing the finality of the decision on the PSR.

32

33 Compliance exposure scenarios involve assessing the compliance of the proposed PSR, or the 

34 continued compliance of a previous PSR affected by the proposed PSR, with the Subpart E 

35 dose limit. Compliance exposure scenarios involve current or future exposure routes between 

36 the partial site and the previously released area or the licensed site (e.g., see the gamma 

37 radiation from the low-level waste storage area example in Section K.3.1). Compliance 

38 exposure scenarios that calculate exposures more than the regulatory limit or a licensee’s self39 imposed limit (e.g., from a previous PSR approval) should then entail remedial actions on the 

40 proposed PSR (not the previous PSR(s)) or more realistic dose assessments.

41 Prospective exposure scenarios involve assessing possible interactions between the partial site 

42 and any future decommissioning actions on the licensed site, including other PSR areas. The 

43 purpose of prospective analyses is to scope out the potential interactions in the future and 

44 address them, either by additional remediation of the partial site or by placing or acknowledging 

45 possible constraints on future decommissioning of the other sources.
K-3

1

2 K.1.4.1 Screening of Features, Events, and Processes

3 The NRC staff should review the licensee’s analysis using the worksheet in Appendix L of this 

4 volume to guide reviews of potential sources of interaction between the partial site and offsite 

5 sources. The purpose of this screening is to answer the questions from above, by identifying 

6 any potential interaction and evaluating the impact(s) on the dose calculations.

7 The licensee should have adequate justification for excluding each of the potential sources, 

8 transport processes, or exposure pathways not evaluated in the dose assessment analyses. 

9 Justification can be quantitative or qualitative.

10 Three methods are acceptable for handling the impact of offsite sources related to interactions 

11 that have not been screened out: (1) incorporate the source, transport mechanisms, and 

12 pathways into the conceptual model and the dose analyses, (2) remediate those sources, or 

13 (3) apply constraints on the PSR’s DCGLs, to accommodate potential exposures from offsite

14 sources or from previous PSRs.

15 Therefore, the NRC staff should verify the following:

16 • The licensee screened potential interactions with the licensed site and previous PSRs.

17 • The screening arguments are justified.

18 • The licensee properly addressed the remaining potential exposure pathways.

19 Section K.3.1 illustrates some of these considerations.

20 K.1.4.2 Screening the Use of the Partial Site and Other Areas by the Critical Group

21 A member of the critical group could potentially be exposed to higher doses than those resulting 

22 from the PSR alone. This would be through the use of other impacted areas after they have 

23 been released (including previous PSR areas), in addition to continuing the use of the partial 

24 site.

25 Three general situations can result in doses to individuals that are higher than for the PSR area

26 alone:

27 • One of the land area’s DCGLs took into account the small size of the area.

28 • Use of more than one exposure area would result in the dose receptor receiving

29 exposure from radionuclides or sources not present on the released area.

30 • Use of more than one exposure area would result in the dose receptor receiving

31 exposure from new exposure pathways or would increase the degree of exposure to a

32 current exposure pathway.

33 Section K.3.2 illustrates a hypothetical review of a situation involving multiple land uses.

34 If the licensee has used the same DCGLs for a previous PSR or commits to use the same 

35 DCGLs for areas surrounding the partial site, multiple uses of the areas is not likely to result in a 
K-4

1 higher dose, as long as none of the above situations is present and the exposure scenarios and 

2 assumptions used in the calculations are appropriate for all areas.

3 If the licensee has: (1) used different DCGLs; (2) has at least one of the above situations 

4 present; (3) found that the exposure scenarios and assumptions on the proposed PSR used for 

5 a previous PSR are no longer appropriate; or (4) has not committed to using the same analyses 

6 for surrounding areas (as long as it would be valid for the other areas, as well), then the NRC 

7 staff should evaluate the licensee’s analyses of potential multiple use exposure scenarios. For 

8 example, for interactions with a previous PSR, the staff should review: (1) any prospective 

9 analyses and associated constraints, if established, on previous PSR(s), (2) the estimated dose 

10 from the residual radioactivity related to both the previous PSR and the proposed PSR, and 

11 (3) any new or updated analyses performed by the licensee.

12 K.1.5 Partial Site Release Evaluation Criteria

13 The NRC staff should verify the following PSR considerations:

14 • For PSR and previous PSR interactions:

15 o The exposure scenarios used in the prospective analyses for the previous PSR, that

16 analyzed the interactions between the previous PSR and the area encompassed by

17 the proposed PSR, continue to be appropriate, or have been updated appropriately.

18 o The licensee incorporated any constraints imposed by the previous PSR that remain

19 appropriate in determining the DCGLs for the proposed PSR.

20 o The licensee appropriately identified those sources that may affect the dose to the

21 average member of the critical group, from either the previous PSR or the proposed

22 PSR.

23 o The licensee adequately justified each excluded potential source, transport

24 mechanism, and pathway.

25 o The licensee incorporated, or addressed by other appropriate means, any sources,

26 transport mechanisms, or pathways that could not be eliminated.

27 o The licensee evaluated (either quantitatively or qualitatively) reasonable exposure

28 scenarios to account for interactions between the previous PSR and the proposed

29 PSR. This includes the prospective analyses for the previous PSR, as well as any

30 new exposure scenarios that required evaluation based on new information.

31 o The DCGLs for the proposed PSR should not result in exposures exceeding the

32 dose limit for either the previous PSR or the proposed PSR. The dose assessment

33 for the proposed PSR should also include any appropriate contributions from the

34 licensed site.

35 • For the PSR and interactions with the licensed site, considering both current and future

36 sources (e.g., potential impacts from other decommissioning activities, or potential future

37 parallel use of impacted areas on the licensed site and the PSR):
K-5

1 o The licensee appropriately identified those current and potential future offsite 

2 sources that may affect the dose calculated for the partial site.

3 o The licensee adequately justified each excluded potential source, transport

4 mechanism, and exposure pathway.

5 o The licensee incorporated, or addressed by other appropriate means, any sources,

6 transport mechanisms, and exposure pathways that could not be eliminated.

7 o The licensee evaluated reasonable exposure scenarios to account for interactions

8 between the proposed PSR and the licensed site, including any prospective analyses

9 that estimate exposures after the licensed site is decommissioned.

10 o The licensee ensured that the DCGLs would not result in exposures exceeding the

11 dose limit at the proposed PSR. The dose assessment for the proposed PSR should

12 also include any appropriate contributions from previous PSRs.

13 o The licensee has clearly documented any constraints placed on current and potential

14 future sources of exposure on the licensed site.

15 K.1.6 Dose Modeling Approaches

16 Licensees proposing PSRs may still be able to use either dose modeling option: (i) screening 

17 values or (ii) site-specific analyses.

18 • If a licensee proposes to use the screening criteria, the NRC should ensure the

19 following:

20 o Interactions with the licensed site or previous PSRs have been appropriately

21 evaluated.

22 o Any sources of potential exposure from the licensed site have been either

23 constrained or remediated.

24 o Any sources of potential exposure increasing either the dose to residents or workers

25 from the proposed PSR or a previous PSR have been either constrained or

26 remediated.

27 o The screening criteria have been appropriately scaled by all the considerations

28 associated with the PSR. For example, in Section K.3.1, the licensee limited the

29 groundwater dose to 0.05 mSv (5 mrem). Therefore, the screening values for the

30 PSR’s surface soil would need to be scaled to 80 percent (0.2 mSv

31 (20 mrem)/0.25 mSv (25 mrem)) of the published values or those received by using

32 the current version of the DandD computer code.

33 o The PSR and its analysis meet the other requirements of Sections 5.1.1 and 5.1.2,

34 as appropriate.

35 • If a licensee uses site-specific modeling, the NRC staff must verify the following:
K-6

1 o The licensee has incorporated all sources from the licensed site or previous PSRs, 

2 as necessary, into the analyses.

3 o The licensee has properly reflected any constraints in calculating the DCGLs.

4 o The modeling meets all the other review criteria of Section 5.2, as appropriate.

5 K.1.7 License Termination: The Effect of Previous Partial Site Releases

6 At the time of final license termination, the NRC staff should take into consideration any 

7 previous PSRs. The entire site (including the previous PSRs) should meet the Subpart E dose 

8 limit. Reviewing the impact on the license termination is the same as that discussed under the 

9 first item in K.1.5, “For PSR and previous PSR interactions.” In this case, it is necessary to 

10 consider the rest of the licensed site as the PSR.

11 K.1.8 Use of Partial Site Release During Decommissioning

12 Reviewers can use this guidance when licensees request the release of portions of their site(s), 

13 either as part of a DP submittal or after the DP has been approved. After the DP has been 

14 approved, some of the issues are not as relevant. If the licensee has prepared a DP for the 

15 entire site, more information will likely be available at the time of the PSR. Importantly, the NRC 

16 reviewer may be able to review the PSR’s DCGLs, the DCGLs for other areas of the site, and 

17 any plans on continued remediation of other areas of the site. The licensee may still need to 

18 complete the prospective analyses of critical group behavior after the entire site is released, but 

19 these exposure scenarios are likely to be easier to define and evaluate.

20 K.2 Partial Site Release Technical Basis for Dose Modeling

21 K.2.1 Considerations for Partial Site Release Dose Assessments

22 Although the license termination requirements in Subpart E provide options for unrestricted and 

23 restricted release, PSRs would normally be used for unrestricted release. A PSR has many 

24 aspects in common with the existing approach for unrestricted release, and the available 

25 guidance is generally applicable. One key difference is that the PSR does not occur concurrent 

26 with license termination. As a result, continuation of licensed activities outside of the released 

27 area represents a potential source of exposure. In turn, the residual radioactivity from the PSR 

28 may affect dose analyses for other areas of the facility during subsequent PSR requests or 

29 eventual license termination.

30 Because this volume’s guidance requires that the dose assessment include all significant 

31 exposure pathways, the need to consider the potential for accumulation processes resulting in 

32 increased radionuclide concentrations over time is not a new concept for PSR. Nonetheless,

33 the importance of accumulation is increased under a PSR because the license will not be 

34 terminated. The existing site areas outside the partial site are not required to be remediated at 

35 the time of the PSR, thereby increasing the potential for accumulation on the partial site, which

36 could affect the dose assessment.

37 One of the most important concepts of the guidance for a PSR is the finality of the decision. 

38 The purpose of the guidance is to establish the scope of the review and focus the reviewer’s 

39 attention and resources on early identification of aspects important to compliance. The primary 

40 objective of the PSR guidance is to ensure that any PSR meets Subpart E requirements, even if 
K-7

1 potentially affected by later PSRs or license termination. The secondary objective of the PSR 

2 guidance is designed to ensure, at the time of license termination, that all prior released areas 

3 are considered and included, as necessary, in dose assessments to provide assurance that the 

4 entire site meets Subpart E requirements. To meet these two objectives, the licensee is asked 

5 to perform both compliance calculations for current conditions at the PSR (or the effect on a 

6 previous PSR) and prospective calculations to estimate the impact on other decommissioning 

7 activities by a licensee. This set of analyses should help ensure that the DCGLs chosen for the 

8 PSR do not result in the need for future remediation of the PSR or unduly constrain the 

9 decommissioning of the entire site.

10 The existence of a PSR may place constraints on future activities that occur nearby, including 

11 currently licensed activities to limit the potential for exposures to a critical group residing or 

12 working (nonradiation workers) on the released area from exceeding other public dose limits 

13 (see Section 3.4 of this volume). For example, 10 CFR 20.1301(d) requires that a fuel cycle 

14 facility comply with the requirements of 40 CFR Part 190, “Environmental Radiation Protection 

15 Standards for Nuclear Power Operations.” This standard (40 CFR Part 190) limits the total dose 

16 that a member of the public may receive from all fuel cycle activities. For the remaining site to 

17 show compliance with 40 CFR Part 190, the dose from the partial site and any effluents from the 

18 remaining site’s operations will need to be combined. The constraint, then, is not on the 

19 compliance of the PSR with Subpart E limits but on maintaining compliance with 

20 40 CFR Part 190 for the remaining activities (e.g., an independent spent fuel storage 

21 installation). Existing NRC effluent control and operational dose limits and their associated 

22 guidance should generally limit operational releases to acceptable levels. Adjustments may 

23 need to be made to effluent compliance calculations or environmental sampling areas to 

24 account for the removal of the PSR from licensee’s control (e.g., because of changing the site 

25 boundary). The NRC may develop additional guidance addressing these constraints (other than 

26 10 CFR Part 20, Subpart E) in the future.

27 The dose criterion of 40 CFR Part 190 is based on actual annual doses. Thus, it should be 

28 noted that a licensee does not have to use its Subpart E compliance calculation (generally a 

29 prospective calculation) as the dose contribution from the PSR for the compliance determination 

30 under 40 CFR Part 190. Information will be available to the licensee to estimate the actual 

31 annual dose based on the actual activities that occurred at the released site. For example, a 

32 partial site is released using an exposure scenario similar to a residence. The next year the 

33 licensee is performing calculations to show compliance with 40 CFR Part 190. In the year since 

34 the PSR was approved, the land was used as a public parking lot. For compliance with 

35 40 CFR Part 190, the licensee can evaluate the PSR dose to a member of the public using the 

36 parking lot instead of the residence exposure scenario used in approving the PSR for 

37 10 CFR Part 20, Subpart E. The licensee may also account for decay since the FSS.

38 For nonimpacted areas, the NRC technical review should normally be limited to the sufficiency 

39 of bases in the site characterization. A PSR, effectively, narrows the definition of nonimpacted, 

40 because of the possibility of future licensee actions resulting in an effect on the PSR. For 

41 example, in some cases, close proximity to existing operations, contaminated areas, future 

42 remediation sites, or potential storage areas may not allow a licensee to designate an area as 

43 nonimpacted and release it without a dose assessment. For impacted areas, PSR requests can 

44 involve a more complicated compliance demonstration and review effort. The site 

45 characterization should include areas of the site outside of the PSR to the extent, necessary to 

46 provide assurance that residual radioactivity from the licensed site (or a previous PSR) is 

47 unlikely to transport material to the partial site that would result in potential future exposures to 

48 users of the property (including subsurface water and groundwater).
K-8

1 K.2.2 Partial Site Release and Decommissioning Guidance

2 This NUREG report was written, in large part, to address decommissioning of sites as part of 

3 the license termination process. Consequently, it often discusses termination of the license as 

4 the end result of decommissioning. When applying this guidance to a PSR, most of the 

5 references to license termination should be regarded as implying the completion of the partial 

6 site decommissioning effort. Despite the frequent use of the term “license termination,” 

7 licensees should be aware that a PSR will not result in license termination, as the entire 

8 licensed site (including any PSRs) should meet the Subpart E requirements at the time of 

9 license termination. Therefore, true license termination issues only need to be considered in 

10 PSR reviews when assessing prospective analyses that may raise issues that need to be 

11 considered or analyzed at the time of license termination (e.g., creation of new license 

12 conditions that identify pathways that should be included in DCGL calculations). Licensees 

13 should also be aware that the existence of a PSR adjacent to impacted areas could place 

14 limitations on future decommissioning methods and actions related to the license termination 

15 (e.g., to minimize the potential for decommissioning to re-contaminate previously released 

16 areas).

17 This NUREG report uses the terms “site” and “facility” interchangeably. Most of the references 

18 will apply to the boundaries of the area proposed for a PSR (i.e., the area to be 

19 decommissioned). Exceptions to this are when the consolidated guidance discusses the need 

20 to collect site characterization information, in which case the terms “site” or “facility” can include 

21 areas beyond the boundary of the PSR and potentially encompass the entire site and any 

22 previous PSR(s), as necessary to establish contaminant source, transport, and exposure 

23 pathways for DCGL calculations. To resolve such safety concerns, the NRC staff is expected to 

24 use pre-submittal meetings with the licensee to develop the information needed on specific 

25 areas of the licensed site.

26 K.2.3 MARSSIM and Partial Site Release

27 The MARSSIM approach involves demonstrating compliance on a survey unit-by-survey unit 

28 basis. Survey units are determined based on the expected level of residual radioactivity in 

29 areas across the site, as well as spatial and topographical considerations. By allowing 

30 compliance demonstration by survey unit, the current approach is congruent with a PSR 

31 concept. As a result, in general, the licensee can apply the MARSSIM approach directly to a 

32 PSR without significant problems.

33 To limit the potential for interactive dose effects, the licensee should fully include in the 

34 proposed PSR final surveys any impacted areas that continue across the proposed PSR 

35 boundary. If buildings are intended for PSR, the licensee should include them unless it can 

36 demonstrate a low potential for future exposure of individuals in the PSR portions of the building 

37 from other impacted areas of the building and should include any significant dose contributions 

38 from outside the PSR area in determining the DCGLs.

39 K.2.4 Dose Modeling Specific Issues

40 The compliance methodology in this NUREG report emphasizes dose modeling to derive 

41 DCGLs to be used as input to the MARSSIM process (see Section 2.5 of this volume). Simple 

42 sites that only involve surface contamination and a low potential for migration of residual 

43 radioactivity should require straightforward dose calculations to derive DCGLs. Sites with both 

44 subsurface and groundwater residual radioactivity or migration of radioactive material from one 
K-9

1 area to another generally require more complex modeling and compliance demonstration 

2 methods.

3 A primary concern is that the dose calculation of DCGLs include all applicable transport and 

4 exposure pathways, because other parts of the site may not be remediated at the same time as 

5 the PSR area. The licensee should give special consideration to any potential for significant 

6 transport of material into the released area from outside the boundary, or from the released area 

7 to other areas of the licensed site or to another previously released area. DCGLs should 

8 account for the movement of radioactive material where accumulation processes could lead to 

9 significantly increasing media concentrations if the transport were included or would add new 

10 radionuclides or exposure pathways to the PSR dose assessment.

11 For example, an area designated for PSR may not have impacted groundwater, but an 

12 impacted area on the licensed site up-gradient has impacted groundwater that is expected to 

13 migrate into the PSR in the future. The licensee should include future groundwater residual 

14 radioactivity in the dose calculation for the PSR (unless its contribution would not be significant) 

15 or address it by other methods. The licensee may need to limit surface DCGLs for the PSR to 

16 ensure that the total PSR area, including any future groundwater dose, complies with the 

17 0.25 mSv/y (25 mrem/y) dose limit for the 1,000-year compliance period. Similar situations 

18 could exist with up-gradient surface or subsurface contamination (e.g., leaching and transport 

19 from the sources to the PSR area).

20

21 Similarly, residual radioactivity sources within the PSR area should be evaluated for potential 

22 transport to the licensed site or other previous PSR areas. Most licensees should be able to 

23 assess the potential for transport pathway communication between site areas using available 

24 site characterization information. Complex sites may require the collection of additional site 

25 characterization information (inside and outside of the PSR area) to support the evaluation of 

26 transport pathways. The scope of site characterization work should be consistent with the 

27 expected level of residual radioactivity and potential dose consequences.

28

29 The PSR process requires records of the PSR residual radioactivity to ensure inclusion in 

30 subsequent PSR analyses and in the overall site license termination process. Residual 

31 radioactivity in PSR areas may be a concern for site license termination when the potential 

32 exists for the migration and accumulation of radioactive materials to the licensed site. This 

33 circumstance may only be significant when several PSR areas exist, in close proximity, and

34 share common transport pathways with the licensed site, allowing the potential accumulation of 

35 transported material. Another situation arises when the critical group may use multiple PSR 

36 areas. In addition, the NRC’s approval of the PSR may have required the licensee to agree to 

37 limits on the dose contributions from decommissioning activities or residual radioactivity sources 

38 remaining on the licensed site.

39

40 For each subsequent PSR request and at license termination, the licensee should consider all 

41 prior PSR areas for potential contributions to dose when calculating partial or site DCGLs. For 

42 some sites, this could mean that prior PSRs could constrain the amount of residual radioactivity 

43 allowed in a future PSR request or at the remainder of a site at license termination. If the review 

44 of a PSR identifies important features, events, or processes (FEPs) to be considered at license 

45 termination, the NRC staff may impose a license condition to ensure the licensee addresses the 

46 matter.

47

48 The review of impacts on previous PSR areas is very important because the NRC approved the 

49 previous PSR request based on the calculations and evaluations that showed compliance with 
K-10

1 the Subpart E limit. If another portion of the site is decommissioned and released later, the 

2 NRC should review the possible impacts on the dose estimates at the previously released site. 

3 The NRC should first review the prospective analyses from its previous approval. If they remain 

4 valid and bound any impacts that the proposed PSR could cause, then it can consider the 

5 impact of the proposed PSR to be acceptable. The NRC should ensure that the DCGLs for the 

6 proposed PSR prevent the dose from previous PSR areas from exceeding the Subpart E limit.

7

8 Although exposure scenario development, especially for prospective analyses, does involve 

9 some speculation, both human behavior and FEPs should be based on present knowledge. To 

10 limit the degree of speculation, the NRC staff should focus on reasonable exposure scenarios 

11 and avoid speculation on activities that are not present in the region, are not reasonably likely to 

12 occur, or that would change human behavior or the FEPs. For example, a scenario that 

13 involves modifying the local topography to allow surface water to transport radioactive material 

14 from an impacted area to the PSR area is generally too speculative and not a reasonable 

15 scenario unless it is part of a remediation option.

16

17 An important part of the detailed technical review may be determining if a licensee has included 

18 all applicable exposure pathways in the DCGL calculations and provided sufficient bases for 

19 exclusion of exposure pathways. Applicable exposure pathways are determined by considering 

20 all three of the following:

21 (1) how the critical group can be exposed to localized residual radioactivity

22 (2) the potential for sources and transport of radioactive materials (from the PSR area, the

23 licensed site, or previous PSR areas) to the location of the applicable critical group

24 (3) concurrent use, if appropriate, of the PSR area and previous PSR areas by the critical

25 group

26 This NUREG report generally addresses exposure pathways for localized residual radioactivity, 

27 and the methods are relatively straightforward. This section focuses on analyzing sources and 

28 transport pathways, because the potential risk of additional sources of exposure affecting a PSR 

29 (or a previous PSR) increases when the entire site is not decommissioned at the same time. 

30 Exposure scenario definition and pathway identification are therefore key aspects of DCGL dose 

31 modeling that are affected by the unique circumstances of a PSR.

32 K.2.5 Features, Events, and Processes

33 Applicable source, transport, and exposure pathways comprise the exposure scenario for DCGL 

34 calculations. The licensee can calculate DCGLs using the all-pathway models, or pathways can 

35 be decoupled from the modeling and their results allocated to pathway-specific DCGLs that can 

36 be combined to generate a survey unit or PSR DCGLW that equates to the Subpart E dose limit. 

37 The licensee may use the site conditions, complexity, and level of risk to choose a specific 

38 method among several options for calculating DCGLs.

39 K.2.5.1 Screening Methods

40 The purpose of screening various sources, transport mechanisms, and exposure pathways is to 

41 evaluate whether the PSR may have processes that could result in radioactive material being 

42 transferred between the PSR area and either the licensed site or previous PSR areas. The first 

43 goal of the screening criteria is to eliminate various FEPs from consideration while minimizing 
K-11

1 the amount of information the NRC staff needs to make a decision. A second goal is for the 

2 screening criteria to factor in the availability and cost of obtaining the information (i.e., the first 

3 criterion should not require the development of a complex site-specific, three-dimensional 

4 groundwater model). The screening of these criteria should not only focus on the effect of the 

5 licensed site, or a previous PSR, on the PSR, but, also, the potential contribution of the PSR on 

6 the dose assessment for the entire site at the time of decommissioning, or the current 

7 compliance of a previous PSR with Subpart E. The following are the general categories of 

8 screening criteria:

9 • the presence of residual radioactivity in various media, including effluent releases from

10 the operating site (e.g., soil, groundwater, air)

11 • the availability of mechanisms to either move material from one location to another,

12 (e.g., groundwater movement) or project exposure from one area to another (e.g., direct

13 radiation)

14 • the availability of exposure pathways to cause a dose in humans after material is moved

15 or projected to the area

16 The licensee may screen out a medium, such as groundwater, that is found to contain residual 

17 radioactivity if it has minimal levels of residual radioactivity (compared with the residual 

18 radioactivity currently present in the media at the critical group location). If the source remains, 

19 then the licensee screens the transport mechanism(s) to evaluate the capability of the process 

20 to move material to the area of interest. This result can then be compared to the residual 

21 radioactivity levels for each radionuclide currently present in the area of interest or other 

22 processes for moving material. Finally, the licensee can screen the potential exposure 

23 pathways to remove those that would result in insignificant doses or are not present at the 

24 location where the material is being deposited.

25

26 In formulating a complete exposure scenario for a proposed PSR, the licensee should give initial 

27 consideration to available information (from the PSR area, the site, or any prior PSR) that can 

28 rule out further consideration of specific sources or transport pathways. Although investigation 

29 of potential sources and transport pathways can become complicated, the licensee can rule out 

30 potential sources and transport pathways with relatively simple and available information. 

31 Appendix L provides a worksheet of source, transport, and exposure pathways with questions 

32 that can be used for screening. Use of a “top-down” approach to screening can avoid an 

33 unnecessary and costly investigation into details that may not have a significant impact on 

34 DCGL calculations. It is expected that, once a potential release or transport pathway has been 

35 identified, licensees may provide simple, yet reasonably conservative, screening-type 

36 calculations to assess importance. Licensees may exclude pathways that make only a small 

37 dose contribution, if the pathway results in less than 10 percent of the dose limit, and the sum 

38 total of all pathway exclusions does not exceed 10 percent of the dose limit (see Section 3.3 of 

39 this volume). The licensee should clearly identify all screened pathways and should show 

40 sufficient bases for exclusion.

41

42 Example of Screening Process for Features, Events, and Processes

43

44 In Appendix L, a worksheet has been provided as one method of screening FEPs for a PSR. 

45 The purpose of the worksheet is to provide some general topics that can, in most cases, be 

46 considered, with generally available information, to minimize unnecessary site characterization, 
K-12

1 modeling, and review. The licensee can use the worksheet to develop both compliance and 

2 prospective analyses. Ultimately, if radionuclides cannot be released or transported to the 

3 critical group location, there is no point in including the FEP(s) in the dose assessment.

4

5 Site-specific conditions and available information may make it desirable for a licensee to initially 

6 focus on source, transport, or both, when trying to screen FEPs. In some cases, it may be 

7 necessary to conduct limited dose calculations to justify the exclusion of a source, transport 

8 mechanism, or exposure pathway. If the licensee cannot screen out a source, then it should 

9 consider it for transport screening. If pathways cannot be excluded using this worksheet, the

10 licensee should consider them in initial dose calculations, either by including them in the 

11 analysis or by modifying the dose limit using an agreed-upon limit. Results of the initial dose 

12 calculations can provide additional insights to the significance of pathways with respect to dose 

13 and may offer additional means for further refinement of the calculations to address only the 

14 important features and processes. The licensee should identify all source, transport, and 

15 exposure pathway exclusions from modeling and appropriately justify their exclusion.

16

17 The worksheet is split into three parts: (1) sources, (2) transport processes, and (3) exposure 

18 pathways. The method is to start with the source questions and follow the directions under 

19 each item, as necessary. The user should follow the path down until the item is screened out or 

20 needs to be considered in the analyses. After reaching the end of a path, the user should go 

21 back to where the branching occurred and continue with the questions, if applicable. For 

22 example, a site has some residual radioactivity in the soil, and the licensee reviews the 

23 questions under L.2.2.2 (“Soil Transport: Leaching”). The questions lead the user to L.2.4 

24 (“Groundwater”) and the user follows that path to its conclusion. The user then continues to 

25 evaluate that residual radioactivity in soil by returning to questions L.2.2.3–L.2.2.6.

26

27 In general, for each “yes” answer, the analysis continues to more detailed questions on that 

28 source and media type. Each “no” answer on a black bulleted question means that area (and its 

29 related questions) need no further evaluation for that specific source or media combination. For 

30 a black-bulleted question with a list of more detailed questions (i.e., with the dash bullets) to be 

31 excluded, all of the detailed questions need to have “no” answers. Some instructions may 

32 provide exceptions to this general rule.

33

34 For example, the last question of L.2.3.1 (“Deep Soil Transport: Leaching”) includes three 

35 specific transport mechanisms from deep soil. If the answers to all three were “no,” the leaching 

36 of the source would be screened out of the dose assessment. If the answer were “no” for 

37 surface water and other, but “yes” for groundwater, the user should address potential leaching 

38 of the deep soil source unless the groundwater transport or related exposure pathways were 

39 subsequently screened out.

40

41 K.2.5.2 Human-Induced Exposure Scenarios

42 Another source of exposure that may lead to interactive dose effects between the PSR and 

43 another impacted area under (or previously under) control of the licensees occurs when 

44 individuals using both the PSR area and the other impacted area(s) after the licensee no longer 

45 controls those areas. The concern is that a critical group could use the PSR area, such that it 

46 still receives a large fraction of the Subpart E dose limit, and reasonably use another impacted 

47 area that would lead to the critical group receiving, in total, doses in excess of the Subpart E 

48 limit.

49
K-13

1 Most prospective exposure scenarios evaluate the human-induced exposure scenarios after the 

2 other impacted area is released for unrestricted use. In cases where the human-induced 

3 exposure scenario involves a previous PSR, the analysis is one of compliance for the PSR, 

4 which could also verify that the human-induced exposure scenario may not result in exposures 

5 to the previous PSR, above the limit. The licensee can use self-identified and agreed-upon 

6 limits to address the exposures from the future use of areas that are on the licensed site (see 

7 the example in Section K.3.2).

8

9 Three situations can result in dose assessments higher than those for the PSR alone:

10

11 (1) One of the DCGLs for the land areas considered the small size of the area.

12 (2) Use of more than one impacted area would result in the dose receptor receiving

13 exposure from radionuclides not present in the PSR area.

14 (3) Use of more than one impacted area would result in the dose receptor receiving

15 exposure from new exposure pathways or would increase the degree of exposure to a

16 current exposure pathway.

17 Taking an area’s size into account when developing the DCGLs is a special case of the third 

18 bullet above. This is because size-related modifications for dose modeling usually result in 

19 reducing the number of pathways or amount of exposure, but these changes may not be 

20 obvious, especially if the code itself (like RESRAD-ONSITE) modifies the dose calculations.

21

22 K.2.6 Subsurface Residual Radioactivity

23 Subsurface residual radioactivity can exist in soils and deeper geologic strata. Common 

24 sources of subsurface residual radioactivity include material leached from surface soils, buried 

25 waste, and impacted groundwater. Impacted areas can be either saturated with groundwater or 

26 unsaturated (where water may percolate through but does not fill all pore spaces). Currently, 

27 this volume suggests applying the MARSSIM (surface-based) methodology to subsurface 

28 radioactivity, with a few modifications to address volume sources (see Appendix G, 

29 Section G.2.1). The NRC expects to update its guidance in the future to improve statistical 

30 methods for subsurface residual radioactivity. This section discusses special considerations for 

31 addressing subsurface residual radioactivity under the PSR exposure scenario(s), with an 

32 emphasis on pathway identification for DCGL calculations. Because addressing subsurface 

33 residual radioactivity is merely a component of the same dose modeling discussed in the 

34 previous sections, the same framework for DCGL calculations applies. For the purpose of 

35 discussion, surface water is included in some examples because of the interconnection between 

36 the surface water and groundwater systems.

37

38 If the MARSSIM methodology classifies a site as impacted, this volume suggests that surface 

39 water and groundwater surveys be designed on a site-specific basis. If the licensee does not 

40 have important information necessary to understand subsurface characteristics (including extent 

41 and amount/type of residual radioactivity) immediately available when requesting a PSR, some 

42 characterization of surface water flow, sediment movement, and groundwater flow for both the 

43 PSR and adjacent areas, as necessary, on the licensed site may be required to support the 

44 amendment request. The source locations, in conjunction with the site complexity, determine 

45 the surface and groundwater characterization needed at the time of a PSR request. The level of 

46 surveys for surface and groundwater residual radioactivity should factor in all three of the 

47 following:
K-14

1 (1) the extent of existing residual radioactivity of soil on the PSR

2 (2) the proximity of the PSR to existing and potential impacted areas on the licensed site

3 (3) the complexity of the surface and groundwater hydrology

4 As noted previously, a PSR that has been classified as impacted requires dose modeling. The 

5 licensee should assess subsurface residual radioactivity, once identified, for inclusion or 

6 exclusion in dose modeling to derive DCGLs. Residual subsurface radioactivity that contributes 

7 less than 10 percent of the dose limit does not need to be included in the DCGL calculations, as 

8 long as all exclusions do not consist of more than 10 percent of the dose limit, but its exclusion 

9 should be documented for future consideration at license termination.

10

11 Simple situations that need to include subsurface residual radioactivity in dose modeling may 

12 involve only radioactive material originating from the PSR area or only one offsite source of 

13 impacted groundwater in a relatively simple hydrology system. More complex PSR scenarios 

14 can involve numerous additional sources of residual radioactivity migrating from outside the 

15 PSR area or migrating from the PSR area onto a previous PSR area or the licensed site. An 

16 important aspect is the possibility of multiple sources coalescing in the surface or groundwater

17 systems (i.e., the additive effect of multiple sources from the licensed site, the PSR area, or 

18 previous PSR areas).

19

20 Licensees should consider all potential processes for the migration of material; however, they 

21 can easily exclude some pathways with available information (see Section 3 of Appendix I and 

22 Appendix L). There is a large dilution effect when radionuclides migrate into bodies of water, 

23 such as streams, rivers, lakes, and ponds. Sediment movement and groundwater flow are 

24 commonly slow processes, relative to surface water flow. Reduction of residual concentrations 

25 in groundwater (caused by mechanical mixing and sorption) and radioactive decay effects are 

26 associated with the longer time factor in the transport legs for sediment movement and 

27 groundwater transport. A clear example of pathway exclusion would be a PSR area located in a 

28 watershed that is isolated from the licensed site operations and impacted areas and is located 

29 upstream of all other offsite sources. It is reasonable, in this case, to neglect the residual 

30 radioactivity at the PSR area in dose modeling at the time of license termination if it can be 

31 demonstrated that there is no significant dose contribution to the site DCGLs. Another simple 

32 example is the exclusion of drinking water pathways given the absence of a drinking water 

33 aquifer accessible to the critical group.

34

35 Types of surface and groundwater features that could lead to focusing residual radioactivity 

36 from multiple, spatially separated source areas are separated into two categories—common 

37 features and site-specific features. To determine if focusing occurs, site characterization data 

38 identify spatially convergent groundwater flow directions or convergent surface water flow and 

39 sediment movement. The license should determine the level of site characterization by the 

40 potential for these features to occur at the site. Examples of each are described below.

41

42 The most common feature leading to convergent mass movement is a river, stream, or pond in 

43 a watershed. Multiple radionuclide sources at various locations around a watershed could all 

44 potentially migrate in the surface and subsurface towards the main stream channel or pond. All 

45 surface water in the watershed could be routed into the main channel or pond. Whereas most 

46 watersheds have an outlet, some lakes, ponds, or bogs may be the terminal point in a transport 

47 pathway where residual radioactivity may accumulate. Changing chemistry of the transport path 
K-15

1 (e.g., the reducing environment of a swamp) can also affect the deposition or dissolution or 

2 mobilization of specific contaminants.

3

4 In that same watershed, the uppermost aquifer may also focus groundwater flow into the 

5 stream, because gradients in the unconfined aquifers typically follow the topography and 

6 commonly seep into stream and river channels. The exception applies to uppermost aquifers 

7 with water tables that lie below the stream elevation; these aquifers would not necessarily seep 

8 into the stream channels or ponds and would not lead to a convergence of groundwater flow 

9 directions unless dictated by another feature.

10

11 Site-specific features, such as faults, karst terrains, and alluvial channel deposits, can focus 

12 water from diverse locations into single transport pathways. These features may lead to a 

13 channelization of flow in the subsurface. Licensees should first determine if such subsurface 

14 features exist at a site. If so, they can analyze the candidates for the potential to focus transport 

15 pathways from impacted areas of the licensed site or a previous PSR area.

16

17 Facilities and PSRs with the potential for multiple sources of residual radioactivity that could 

18 migrate to surface or groundwater should use or obtain sufficient site characterization data to 

19 show the possibility for convergent features to exist on the site. The licensee may have to get 

20 the site characterization data at the time of a PSR request if they are not already available. The 

21 licensee should assess the potential for overlapping transport pathways from multiple source 

22 areas, where those sources could be at the PSR area, the licensed site, or previous PSR areas.

23

24 K.2.7 Records and Documentation

25 Maintaining complete records of PSRs is important, because the information is likely to be 

26 needed for any subsequent PSR requests and at the time of license termination. The NRC staff 

27 should consider all prior licensing actions in the reviews for a license termination, of which a 

28 PSR is only one example. Similarly, the framework for PSR involves considering all prior PSRs 

29 and whether the residual radioactivity needs to be included in DCGL calculations for license 

30 termination. Because considerable time may elapse between a PSR and the eventual license 

31 termination, maintenance of complete records is an important aid to the licensee, as well as the 

32 NRC staff. Incomplete records may result in the need for additional site characterization at the 

33 time of license termination. Records should include identification of impacted areas and 

34 information describing the MARSSIM RSSI methods used and results obtained, including all 

35 PSR site characterization information that supports DCGL calculations. Any information 

36 supporting source, transport, or exposure pathway exclusions in the PSR area in common with 

37 the licensed site is important, as are any licensee agreed-upon limits used to simplify the 

38 previous dose assessments. This information may support a determination of whether PSRs 

39 must be included in DCGL calculations at the time of license termination. Volume 3 of this 

40 NUREG has more information on recordkeeping requirements.

41

42 K.3 Hypothetical Examples

43 K.3.1 Contributions from the Remaining Licensed Site

44 A licensee requests the release a portion of a site 10 years before the date it estimates it will 

45 provide the NRC with the DP. The PSR has surface soil residual radioactivity of cobalt-60 

46 (Co-60) and cesium-137 (Cs-137). Adjacent to the PSR, on the licensed site, is the low-level 

47 waste storage area, which is a potential source of gamma exposure to individuals within the 

48 released area. The only other potential offsite source is a groundwater plume from the licensed 
K-16

1 site. The licensee evaluates the two offsite sources and eliminates all other offsite sources 

2 because of the absence of valid transport mechanisms to allow significant impact on dose 

3 analyses. The licensee then takes the two following actions to address the remaining potential 

4 exposure sources:

5 • A berm is going to be built between the low-level waste storage area and the PSR area,

6 on the licensed site, to reduce the external gamma exposure. The berm is estimated to

7 reduce the potential dose from 0.05 mSv (5 mrem) to less than 0.001 mSv (0.1 mrem).

8 The contribution is now insignificant, and the source can be eliminated from further

9 consideration in estimating the dose for the compliance calculations for the PSR’s

10 DCGLs. The presence of the berm would then likely become a license condition.

11 Removal of the berm in the future may require reanalyzing the total dose to the critical

12 group to verify that exposures from the released area will not exceed the Subpart E limit

13 after unrestricted release of the low-level waste storage area.

Note that, if a previous PSR is impacted by a proposed PSR, or the 

decommissioning of the site, such that doses from the PSR area may exceed 

the Subpart E dose limit, constraints are to be placed on the current action(s) 

and do not require that the previous PSR be remediated (except as noted by 

10 CFR 20.1401(c)).

14

15 • The groundwater plume is estimated to reach the PSR area in approximately 15 years.

16 The licensee currently has no final plans to remediate the plume. Current conservative

17 dose modeling estimates the annual peak exposure to be approximately 0.05 mSv

18 (5 mrem) from an all-pathway analysis, using the groundwater concentration at its

19 current location. The licensee proposes partitioning the unrestricted dose limit for the

20 PSR and constraining the annual peak dose from the surficial soil residual radioactivity

21 in the PSR area to 0.2 mSv (20 mrem) and the groundwater dose to 0.05 mSv (5 mrem).

22 To ensure the groundwater concentrations do not exceed specific concentrations

23 associated with the agreed-upon limit of 0.05-mSv (5-mrem), the licensee may install

24 monitoring wells between the plume and the new licensed site boundary and develop a

25 corrective action plan to use in case the concentrations rise above some specified

26 fraction of the licensee’s agreed-upon limit.

27 Although the PSR should meet the Subpart E dose limits at the time of approval, these actions 

28 may further affect the final decommissioning of the licensed site. For example, assume, at the 

29 time of license termination, residual amounts of Co-60 and niobium-94 are in the surface soil 

30 around the low-level waste storage area and a building code requires the berm to be removed 

31 before the entire site is released for unrestricted use. When the berm is removed, external 

32 exposure would result in 0.04 mSv/y (4 mrem/y) to the average member of the critical group in 

33 the PSR area. Based on the concentrations from the PSR’s FSSR and the groundwater

34 agreed-upon limit, the total dose estimate for the PSR is now 0.29 mSv/y (29 mrem/y), which 

35 does not meet the Subpart E dose limit for unrestricted release. The effect of the PSR may limit 

36 the final DCGLs of the low-level waste storage area. Other options available to the licensee 

37 would be to reevaluate the PSR dose assessment to account for decay and new information on 

38 the dose from the groundwater plume or provide additional remediation of the groundwater

39 plume.

40
K-17

1 K.3.2 Use of Multiple Areas

2 A second licensee requests the release of an impacted portion of a licensed site. The area has 

3 residual radioactivity of uranium and thorium. The PSR area is rocky, with poor soil, and the 

4 licensee can provide adequate justification that the critical group would not plant extensive 

5 gardens nor use the groundwater under the released site. No offsite sources or transport 

6 mechanisms could affect the dose if the critical group used only the PSR area.

7 The licensed site has the closest other radioactive source, under the control of the licensee—

8 some groundwater concentrations of hydrogen-3 and chlorine-36 from old tracer tests. The land 

9 over the groundwater residual radioactivity is suitable for extensive gardening or farming, and 

10 the aquifer is potable. As part of the dose assessment, the licensee evaluates a prospective 

11 exposure scenario where the critical group may use both the PSR area and portions of the 

12 licensed site after it is decommissioned. After review of the sources, impacted areas, and 

13 routes of exposure, the licensee decides that a reasonable exposure scenario would involve the 

14 person living on the previously released site and using the offsite area and its impacted aquifer 

15 for drinking water and growing an extensive garden. The licensee, believing it should be easy 

16 to remediate the groundwater, addresses the offsite pathways in this exposure scenario by 

17 proposing an aggressive limit (i.e., a small fraction of the current dose estimate) on the dose 

18 from the waterborne pathways (which are all from the offsite area). Accounting for its agreed19 upon groundwater limit, the licensee calculates DCGLs for the PSR based only on the 

20 radioactivity in the PSR area, performs the FSS, and gains NRC approval of the PSR request 

21 for unrestricted use.

22

23 At the time of site decommissioning, years later, the licensee, having better characterized the 

24 groundwater plume and having run some well pumping tests, finds that it may be difficult to 

25 meet the limit it established on the groundwater dose without extensive remediation of the 

26 groundwater. Therefore, the NRC staff’s perspective is that at license termination, the licensee 

27 is effectively left with three options:

28

29 (1) Remediate the groundwater down to the licensee’s agreed-upon limit.

30 (2) Revise the licensee’s agreed-upon limit based on additional modeling (e.g., considering

31 actual FSS results for the PSR, decay of the sources, new information known about the

32 groundwater system, and associated residual radioactivity, or more realistic models of

33 groundwater dispersion and transport).

34 (3) Combine the two options.

35 The NRC views remediating a previous PSR as the option of last resort, consistent with its 

36 policy on intervention of terminated licenses. Obviously, if the licensee desires to remediate the 

37 previous PSR area, the NRC would not necessarily stop the licensee. However, the situation 

38 could involve several issues related to regulatory authority and require the current site owner’s 

39 approval.

40

Note that if the groundwater with residual radioactivity had been under a previous PSR 

instead of the licensed site, the options would have been different. The options would 

be to remediate the uranium and thorium residual radioactivity in the proposed PSR 

area, do more complex modeling, or combine the two.

41

WORKSHEET FOR IDENTIFYING POTENTIAL PATHWAYS FOR 

PARTIAL SITE RELEASE

L-1

1 L.1 Introduction

2 This volume provides a worksheet to assist the staff of the NRC and its licensee to screen 

3 potential sources and transport pathways considered in dose modeling for derived concentration 

4 guideline levels for a PSR. It is intended that the results of this worksheet summarize the 

5 exclusion or inclusion of each item and the screening argument, as well as reference the more 

6 complete screening argument, if necessary. The questions should be considered for all sources 

7 of residual radioactivity and potential critical group locations. Although this worksheet has been 

8 designed for use in identifying features, events, or processes that could result in additional 

9 sources of exposure for the critical group in the partial release, it can also be used for general 

10 exposure scenarios and pathway development.

11

The worksheet focuses on physical features, events, and processes that may transport 

radioactive material to the partial site. Additionally, it covers situations where offsite 

radioactive material may directly expose critical group members using the partial site 

release. It does not explicitly address sources or routes of exposure that result from 

the critical group using more than the partial release.

12

13 L.2 Instructions

14 The worksheet is split into three parts: (1) screening sources, (2) screening transport processes, 

15 and (3) exposure pathways. The method is to start with the source questions and follow the 

16 directions under each item, as necessary. The licensee should follow the path down until it 

17 screens out the item or decides to consider it in the analyses. After reaching the end of a path, 

18 the licensee should go back to where the branching occurred and continue with the questions, if 

19 applicable. For example, a site has some residual radioactivity in the soil and the licensee 

20 reviews the questions under Section L.4.2.2 (“Soil Transport: Leaching”). The questions lead to 

21 Section L.4.4, “Groundwater,” and the licensee follows that path to its conclusion. The licensee 

22 then returns to Sections L.4.2.3–L.4.2.5 to continue evaluating the source of residual 

23 radioactivity in the soil.

24

25 L.3 Screening Sources (Yes/No)

26 Answer the following questions that are appropriate for each possible source of residual

27 radioactivity.

28

29 L.3.1 Existing/Historical Residual Radioactivity (Yes/No)

30 • Is residual radioactivity present in the media? (yes/no)

31 o surface soil (less than 30 centimeters (cm) (1 foot (ft))?

32 o deep soil (greater than 30 cm (1 ft))?

33 o groundwater?

34 o surface water?

35 o structures?

36 o others?
L-2

1 For each media type, determine if there is enough residual radioactivity to include in 

2 dose calculations? (yes/no)

3 o surface soil (less than 30 cm (1 ft))? If “yes,” go to Section L.4.2.

4 o deep soil (greater than 30 cm (1 ft))? If “yes,” go to Section L.4.3.

5 o groundwater? If “yes,” go to Section L.4.4.

6 o surface water? If “yes,” go to Section L.4.5.

7 o structures? If “yes,” go to Section L.4.6.

8 o other? If “yes,” follow the process for the most similar media.

9

10 L.3.2 Current Operational Releases (Yes/No)

11 • Are there current effluents or ongoing disposals from the operating facility in the media

12 (e.g., onsite disposals may be authorized under Title 10 of the Code of Federal

13 Regulations (10 CFR) 20.2002, “Method for Obtaining Approval of Proposed Disposal

14 Procedures”)? (yes/no)

15 o gaseous or particulate release? If “yes,” go to Section L.4.1.

16 o surface soil (less than 30 cm (1 ft))? If “yes,” go to Section L.4.2.

17 o deep soil (greater than 30 cm (1 ft))? If “yes,” go to Section L.4.3.

18 o groundwater? If “yes,” go to Section L.4.4.

19 o surface water? If “yes,” go to Section L.4.5.

20 o other? If “yes,” follow the process for the most similar media.

21 • Are there ongoing or planned decommissioning activities involved with the media

22 containing residual radioactivity? (yes/no)

23 o gaseous or particulate release? If “yes,” go to Section L.4.1.

24 o surface soil (less than 30 cm (1 ft))? If “yes,” go to Section L.4.2.

25 o groundwater? If “yes,” go to Section L.4.4.

26 o surface water? If “yes,” go to Section L.4.5.

27 o structures? If “yes,” go to Section L.4.6.

28 o other? If “yes,” follow the process for the most similar media.
L-3

1 Screening Transport Processes (Yes/No)

2

3

4 L.4.1 Air Transport (Yes/No)

5 • Does the wind travel a significant portion of the year from the source to the critical group

6 location? (yes/no)

7 • Is the source location near enough to the critical group location to prevent significant

8 dilution of suspended or gaseous residual radioactivity? (yes/no)

9 • Do the structures, topography, and vegetation between the source and critical group

10 locations provide only small amounts of dispersion? (yes/no)

11 If the answer to any one of the above questions in this section is “no,” answer the following 

12 question. If all are “yes,” go to Section L.5.1.

13

14 • Can air-transported residual radioactivity from this source accumulate with other

15 source/air transport combinations that have been screened out, so that the combined

16 effect of all sources would result in a significant source of exposure? (yes/no)

17 If the answer is “yes,” air transport for this source and for any other sources identified by this 

18 question is not screened out. Go to Section L.5.1. If “no,” air transport for the source is 

19 screened out.

20

21 L.4.2 Surface Soil Transport (Yes/No)

22 Each source should go through all subsections. Screening out one subsection does not 

23 necessarily mean all subsections are screened out. To screen out the entire surface soil 

24 transport mechanism for a source, screen out Sections L.4.2.1–L.4.2.5 individually.

25

26 L.4.2.1 Erosion (Yes/No)

27 • Can the chemical or structural form of the residual radioactivity erode within the time

28 frame of the analysis? (yes/no)

29 • Is the rainfall, runoff, or wind speed sufficient to erode source contaminants? (yes/no)

30 • Is the source location close enough to the critical group location for erosion to transport

31 contaminants to the critical group location? (yes/no)

32 • Do the structures, topography, and vegetation between the source location and the

33 critical group favor transport of material to the critical group location? (yes/no)

34 If the answer to any one of the above questions in this section is “no,” answer the following 

35 question. If all are “yes,” skip the next question, and then answer the last question of this 

36 section.

37

L.4

Answer the following questions for the media type/source combination.
L-4

1 • Can the residual radioactivity from this source erode and accumulate with other

2 source/erosion transport combinations that have been screened out so that the

3 combined effect of all sources would result in a significant source of exposure? (yes/no)

4 If the answer is “yes,” the erosion subsection for this source and any other sources identified by 

5 this question is not screened out. Answer the following question. If “no,” the erosion subsection 

6 is screened out. Go to Section L.4.2.2.

7

8 • If erosion were to occur, where would the material arrive so that it can be transported to

9 the critical group location?

10 o direct overland flow? (yes/no) If “yes,” go to Section L.4.5 and answer surface water

11 questions for potential overland flow.

12 o surface water body? (yes/no) If “yes,” go to Section L.4.5.

13 o other? (yes/no) If “yes,” go to the appropriate similar transport mechanism.

14 If the answer to any one of these questions is “yes,” the erosion subsection for this source is not 

15 screened out. Proceed as directed by the specific question. After completing that pathway, 

16 return to Section L.4.2.2. If “no,” the erosion subsection is screened out. Go to Section L.4.2.2.

17

18 L.4.2.2 Leaching (Yes/No)

19 • Is the rainfall or infiltration amount sufficient for residual radioactivity to leach to a

20 significant degree? (yes/no)

21 • Will the residual radioactivity leach within the time frame of the analysis? (yes/no)

22 • Does the geochemistry of the soil and radionuclides (e.g., distribution coefficients (Kd))

23 allow leached residual radioactivity to reach the ultimate transport mechanism within the

24 time frame of the analysis (e.g., will the residual radioactivity be able to move through

25 the unsaturated zone and enter into the groundwater aquifer)? (yes/no)

26 If the answer to any one of the above questions in this section is “no,” answer the following 

27 question. If all are “yes,” skip the next question, and then answer the last question of this 

28 subsection.

29

30 • Can the leached residual radioactivity from this source accumulate with other

31 source/leach transport combinations that have been screened out so that the combined

32 effect of all sources would result in a significant source of exposure? (yes/no)

33 If the answer is “yes,” the leaching subsection for this source and for any other sources 

34 identified by this question is not screened out. Answer the following question. If “no,” the 

35 leaching subsection is screened out. Go to Section L.4.2.3.

36

37 • If leaching were to occur, where would the material arrive so that it can be transported to

38 the critical group location?

39 o groundwater aquifer? (yes/no) If “yes,” go to Section L.4.4.
L-5

1 o surface water body? (yes/no) If “yes,” go to Section L.4.5.

2 o other? (yes/no) If “yes,” go to the appropriate similar transport mechanism.

3 If the answer to any one of these above questions is “yes,” the leaching subsection for this 

4 source is not screened out. Proceed as directed by the specific question. After completing that 

5 pathway, return to Section L.4.2.3. If the answer to each of the bullets is “no,” the leaching 

6 subsection is screened out. Go to Section L.4.2.3.

7

8 L.4.2.3 Resuspension (Yes/No)

9 • Does the wind travel a significant portion of the year from the source to the critical group

10 location? (yes/no)

11 • Is the source location near enough to the critical group location to prevent significant

12 dilution of suspended or gaseous residual radioactivity? (yes/no)

13 • Do the structures, topography, and vegetation between the source location and the

14 critical group favor transport of the material to the critical group location? (yes/no)

15 • Can enough of the residual radioactivity be resuspended to affect the dose to the critical

16 group? (yes/no)

17 If the answer to any one of the above questions in this section is “no,” answer the following 

18 question. If all are “yes,” skip the next question and go to Section L.5.1. After completing that 

19 pathway, return to Section L.4.2.4.

20

21 • Can the resuspended residual radioactivity from this source accumulate with other

22 source/resuspension or air-transport combinations that have been screened out so that

23 the combined effect of all sources would result in a significant source of exposure?

24 (yes/no)

25 If the answer is “yes,” the resuspension subsection for this source and for any other sources 

26 identified by this question is not screened out. Go to Section L.5.1. After completing that 

27 pathway, return to Section L.4.2.4. If “no,” the resuspension subsection is screened out. Go to 

28 Section L.4.2.4.

29

30 L.4.2.4 Manual Redistribution: Excavation and Fill (Yes/No)

31 • Do source area characteristics allow future excavation and reuse? (yes/no)

32 • Would reuse be reasonable on or near the partial site? (yes/no) A “no” on this question

33 does not screen this subsection out.

34 • Would the source be able to become airborne as part of fugitive dust emissions?

35 (yes/no) If “yes,” go to Section L.4.2.3. A “no” on this question does not screen this

36 subsection out.

37 If the answer to the first bullet is “no,” or the second and third bullets are “no,” the manual 

38 redistribution subsection is screened out. Go to Section L.4.2.5. If manual redistribution is not 

39 screened out, go to Section L.5.2. After completing that pathway, return to Section L.4.2.5.
L-6

1 L.4.2.5 Direct Radiation (Yes/No)

2 • Are the radionuclides significant external hazards? (yes/no)

3 • Is the source location close enough to the critical group location to prevent a significant

4 reduction in the dose rate? (yes/no)

5 • Do the structures, topography, and vegetation between the source and critical group

6 locations provide inadequate shielding to minimize the external exposure? (yes/no)

7 If the answer to any one of the above questions in this section is “no,” the direct radiation 

8 subsection is screened out. If all are “yes,” go to Section L.5.2.

9

10 L.4.3 Deep Soil Transport (Yes/No)

11 Each source should go through both subsections below. Screening out one subsection does 

12 not necessarily mean that both subsections are screened out. To screen out the entire deep 

13 soil transport mechanism for a source, screen out both Section L.4.3.1 and Section L.4.3.2 

14 individually.

15

16 L.4.3.1 Leaching (Yes/No)

17 • Is the rainfall or infiltration amount sufficient for residual radioactivity to leach to a

18 significant degree? (yes/no)

19 • Will the residual radioactivity leach within the time frame of the analysis? (yes/no)

20 • Does the geochemistry of the soil and radionuclides (e.g., Kd) allow leached residual

21 radioactivity to reach the ultimate transport mechanism within the time frame of the

22 analysis (e.g., will the residual radioactivity be able to move through the unsaturated

23 zone and enter the groundwater aquifer)? (yes/no)

24 If the answer to any one of the above questions in this section is “no,” answer the following 

25 question. If all are “yes,” skip the next question, and then answer the last question of this 

26 subsection.

27

28 • Can leached residual radioactivity from this source accumulate with other source/leach

29 transport combinations that have been screened out so that the combined effect of all

30 sources would result in a significant source of exposure? (yes/no)

31 If the answer is “yes,” the leaching subsection for this source and for any other sources 

32 identified by this question is not screened out. Answer the following question. If “no,” the 

33 leaching subsection is screened out. Go to Section L.4.3.2.

34

35 • If leaching were to occur, where would the material arrive so that it can be transported to

36 the critical group location?

37 o groundwater aquifer? (yes/no) If “yes,” go to Section L.4.4.

38 o surface water body? (yes/no) If “yes,” go to Section L.4.5.
L-7

1 o other? (yes/no) If “yes,” go to the appropriate similar transport mechanism.

2 If the answer to any one of these is “yes,” the leaching subsection for this source is not 

3 screened out. Proceed as directed by the specific question. After completing that pathway, 

4 return to Section L.4.3.2. If the answers to all of the bullets are “no,” the leaching subsection is 

5 screened out. Go to Section L.4.3.2.

6

7 L.4.3.2 Manual Redistribution: Excavation and Fill (Yes/No)

8 • Do source area characteristics allow future excavation and reuse? (yes/no)

9 • Would reuse be reasonable on or near the partial site? (yes/no) A “no” on this question

10 does not screen this subsection out.

11 • Could the source become airborne as part of fugitive dust emissions? (yes/no) If “yes,”

12 go to L.4.2.3. A “no” on this question does not screen this subsection out.

13 If the answer to the first bullet is “no,” or all bullets are “no,” the manual redistribution subsection 

14 is screened out. If manual redistribution is not screened out, go to Section L.5.2.

15

16 L.4.4 Groundwater Transport (Yes/No)

17 • Is saturated groundwater in hydraulic connection with the radioactive source? (yes/no)

18 • Does the groundwater (including unconfined or confined aquifers, as necessary) flow

19 from the source to the location of the critical group? (yes/no)

20 • Is the aquifer fit for use? (yes/no)

21 • Is the aquifer potable? (yes/no)

22 • Is groundwater used for irrigation? (yes/no)

23 • Can the residual radioactivity enter the groundwater aquifer in significant amounts

24 (e.g., is the aquifer not protected from all potential migrating contaminants by

25 low-permeability geologic strata (e.g., clay layer))? (yes/no)

26 • Is the yield rate of the aquifer sufficient? (yes/no)

27 o household and drinking water? (yes/no)

28 o irrigation? (yes/no)

29 • Is the distance traveled from the source to the critical group location close enough to

30 prevent significant dilution and sorption of migrating radionuclides? (yes/no)

31 If the answer to any of the above questions (not related to irrigation) in this section is “no,” the 

32 drinking water/groundwater transport mechanism is screened out. If the questions related to 

33 irrigation are no, the irrigation/groundwater transport mechanism is screened out. If all are 

34 “yes,” go to Section L.5.3.

35
L-8

1 L.4.5 Surface Water Transport (Yes/No)

2 • Does surface water flow from the source of residual radioactivity (or from zones of

3 mobilized radionuclides) to the critical group location? (yes/no)

4 • Does the volume of surface water allow transport of significant concentrations of either

5 dissolved or suspended radioactive solids? (yes/no)

6 If the answer to either of the above questions in this section is “no,” the surface water transport 

7 mechanism is screened out. If both are “yes,” answer the following question.

8

9 • Is significant sediment buildup possible at the critical group location? (yes/no)

10 If the answer is “yes,” go to Section L.5.5. After completing that pathway, return to 

11 Section L.5.4. If the answer is “no,” go to Section L.5.4.

12

13 L.4.6 Structures (Yes/No)

14 L.4.6.1 Direct Radiation (Yes/No)

15 • Are the radionuclides significant external hazards? (yes/no)

16 • Is the source location close enough to the critical group location so that a significant

17 dose rate cannot be avoided? (yes/no)

18 • Do the structures, topography, and vegetation between the source and critical group

19 locations provide inadequate shielding to minimize the external exposure? (yes/no)

20 If the answer to any one of the above questions in this section is “no,” the direct radiation 

21 subsection is screened out. Go to Section L.4.6.2. If all are “yes,” go to Section L.5.2. After 

22 completing that pathway, return to Section L.4.6.2.

23

24 L.4.6.2 Leaching (Yes/No)

25 • Is the rainfall or infiltration amount sufficient for residual radioactivity to leach to a

26 significant degree? (yes/no)

27 • Will the residual radioactivity leach from the structure within the time frame of the

28 analysis? (yes/no)

29 • Does the geochemistry of the soil and radionuclides (e.g., Kd) allow leached residual

30 radioactivity to reach the ultimate transport mechanism within the time frame of the

31 analysis (e.g., will the residual radioactivity be able to move through the unsaturated

32 zone and enter the groundwater aquifer)? (yes/no)

33 If the answer to any one of the above questions in this section is “no,” answer the following 

34 question. If all are “yes,” skip the next question, and then answer the last question of this 

35 subsection.

36
L-9

1 • Can the leached residual radioactivity from this source accumulate with other

2 source/leach transport combinations that have been screened out so that the combined

3 effect of all sources would result in a significant source of exposure? (yes/no)

4 If the answer is “yes,” the leaching subsection for this source and for any other sources 

5 identified by this question is not screened out. Answer the next question. If “no,” the leaching 

6 subsection is screened out.

7

8 • If leaching were to occur, where would the material arrive so that it can be transported to

9 the critical group location?

10 o groundwater aquifer? (yes/no) If “yes,” go to Section L.4.4.

11 o surface water body? (yes/no) If “yes,” go to Section L.4.5.

12 o other? (yes/no) If “yes,” go to the appropriate similar transport mechanism.

13 If the answer to any one of these questions is “yes,” the leaching subsection for this source is 

14 not screened out. Proceed as directed by the specific question. If the answers to all these dash 

15 bullets are “no,” the leaching subsection is screened out.

16

17 L.5 Exposure Pathways

18 “No” on the black bullets will not eliminate the entire section.

19

20 L.5.1 Air Pathways (Yes/No)

21 • Based on the critical group habits and activities, are the following viable? (yes/no)

22 o inhalation? (yes/no)

23 o air submersion external dose? (yes/no)

24 • Is significant deposition viable? (yes/no) If “yes,” go to Section L.5.2 and consider the

25 potential soil pathways at the deposition area.

26 L.5.2 TSoil Pathways (Yes/No)

27 • Is external exposure viable? (yes/no)

28 • Is exposure through ingestion viable? (yes/no)

29 o direct soil ingestion? (yes/no)

30 o garden or crops? (yes/no)

31 o leafy vegetables? (yes/no)

32 o nonleafy vegetables? (yes/no)

33 o fruits? (yes/no)
L-10

1 o grain? (yes/no)

2 o animal husbandry? (yes/no)

3 o meat? (yes/no)

4 o milk? (yes/no)

5 o eggs? (yes/no)

6 • Is exposure through inhalation viable? (yes/no)

7 o indoors? (yes/no)

8 o outdoors? (yes/no)

9 L.5.3 Groundwater Pathways (Yes/No)

10 • Is exposure via drinking water viable? (yes/no)

11 • Is exposure via irrigation viable? (yes/no)

12 o garden or crops? (yes/no)

13 o animal husbandry? (yes/no)

14 o fish farming? (yes/no)

15 If irrigation is viable, go to Section L.5.2. Consider the soil pathways appropriate for the soil 

16 affected by the irrigation.

17 • Is water used for purposes other than household uses (including drinking water) or

18 irrigation (e.g., evaporative coolers, dust suppression)? (yes/no)

19 If “yes,” go to Section L.5.2. Consider the soil pathways appropriate for the impacts of the 

20 activity.

21

22 L.5.4 Surface Water Pathways (Yes/No)

23 • Is internal exposure viable? (yes/no)

24 o fish? (yes/no)

25 o drinking water? (yes/no)

26 o inadvertent intakes? (yes/no)

27 o via irrigation? (yes/no)

28 o garden or crops? (yes/no)

29 o animal husbandry? (yes/no)
L-11

1 If irrigation is viable, go to Section L.5.2. Consider the soil pathways appropriate for the soil 

2 affected by the irrigation.

3

4 • Is water used for purposes other than household uses (including drinking water) or

5 irrigation (e.g., evaporative coolers, dust suppression)? (yes/no)

6 If “yes,” go to Section L.5.2. Consider the soil pathways appropriate for the impacts of the 

7 activity.

8

9 • Are recreational activities viable? (yes/no)

10 If recreational activities are viable, go to Section L.5.2. Consider the exposure pathways 

11 appropriate for recreational activities in the water (e.g., incidental ingestion during swimming).

12

13 L.5.5 Sediments (Yes/No)

14 • Are recreational activities viable? (yes/no)

15 If recreational activities are viable, go to Section L.5.2. Consider the exposure pathways 

16 appropriate for recreational activities on the shore, or involving sediments (e.g., incidental 

17 ingestion from making sand castles).

18

19 • Is use of sediments for land-based activities viable (e.g., fill or crops)?

20 If use of sediments is viable, go to Section L.5.2. Consider the soil pathways appropriate for the 

21 impacts of the activity.

22

23 L.5.6 Documentation

24 The information from the worksheet should be summarized in tables. The tables should include 

25 (1) the source, (2) whether it is included or excluded, (3) the FEPs screened, (4) the screening

26 argument, and (5) the reference for the screening argument. For example, one format shown

27 below uses the example in Section 3.1 of Appendix K as a basis. The level of detail is only

28 needed for the question being used to screen out the source, transport mechanism, or pathway.

29 Common pathways using the same or similar screening arguments can be grouped (e.g., fourth

30 row of the example below).

31
L-12

1 Table L.1 Example of Summary Format 

Source Status Screening Pathwaya Screening Argument Reference

Groundwater 

(GW) Plume–

Area 4-10

Incl GW (3.1)–GW (4.4)–

GW (5.3)–Soil (5.2)

N/A N/A

GW Plume–

Area 4-12

Excl GW (3.1)–GW 

(4.4/YIELD)

Yield of Pico Aquifer 

<30 L/day.

DP Chapter 

3.7.3

Low-Level 

Waste (LLW) 

Storage Area

Incl Other (3.2)–Soil (4.6)–

Soil (5.2/direct)

N/A N/A

Excl Other (3.2)–Soil (4.1–

4.2)

No significant erosion or 

leaching of LLW area 

within 1,000 years.

DP Chapter 

4.1.5

Note:

a. Numbers in this column indicate the appropriate sections in Appendix L.

2
PROCESS FOR DEVELOPING EXPOSURE SCENARIOS USING SITESPECIFIC INFORMATION

M-1

1 M.1 Introduction

2 This appendix discusses modification of the two default or screening-level exposure scenarios, 

3 the residential farmer and the building occupancy exposure scenarios, using site-specific 

4 information. The residential farmer exposure scenario is applied to sites with land and water 

5 residual radioactivity, while the building occupancy exposure scenario is applied to sites with 

6 contaminated structures. A generic critical group, with associated default parameter 

7 representing the average member of each group, is associated with each exposure scenario. 

8 The default pathways, models, and parameter values for the critical group combine to form 

9 exposure scenarios.

10

11 The default exposure scenarios may not always be appropriate or may lead to overly 

12 conservative results. For example, there is significant variability among decommissioning sites 

13 with respect to physical features and characteristics of residual radioactivity remaining on site. 

14 The original purpose of the site, historical development, and the resulting processes that 

15 generated the site’s residual radioactivity vary widely. Residual radioactivity may occur in 

16 buildings, process equipment and other site structures, soils (surface and subsurface), ponds, 

17 lagoons, surface water, and groundwater. Sites are in urban and suburban; residential, 

18 commercial and industrial, rural, and agricultural areas; and many are located on or directly 

19 adjacent to rivers, lakes, oceans, estuaries, wetlands, flood plains, or wildlife areas. The 

20 physical and chemical form of residual radioactivity is highly variable. Residual radioactivity 

21 may be associated with slag, soils, sediments, sludge, debris, dust or sand piles, packaged 

22 waste (e.g., drums, crates), and dispersed in liquid media. These factors should be considered 

23 in developing reasonably foreseeable exposure scenarios, as well as those that are considered 

24 less likely but plausible.

25

26 This appendix steps through the process of developing alternative exposure scenarios and for 

27 eliminating various default pathways from the residential farmer exposure scenario and 

28 describes the information needed to support these modifications. Contact with state and local 

29 agencies is recommended early in the process to obtain information on state and local laws and 

30 regulations, and sources of information to support development of alternative scenarios and 

31 elimination of pathways. The process for developing alternative exposure scenarios 

32 complements the Decision-making Framework and is meant to be used in conjunction with the 

33 methodology discussed in Chapter 1 of this volume and in the guidance on exposure scenarios, 

34 exposure pathways, and critical groups discussed in Appendix I, Section I.3. 

35

36 While development of site-specific exposure scenarios is always an option, if screening 

37 assumptions and values are met, there may be no need to collect additional data to develop 

38 alternative exposure scenarios. If the dose criteria of 10 CFR 20.1402 are met through use of 

39 the screening values and the residual radioactivity has been reduced to levels that are ALARA, 

40 the site would be considered a candidate for unrestricted use. However, if screening level 

41 analyses do not support compliance with the release standard, then one option is to use site42 specific information to modify the resident farmer exposure scenario, including use of site43 specific information to support elimination of pathways that are inappropriate for the site in 

44 question. Sensitivity analysis would be helpful in determining what pathways and parameters 

45 have a significant effect on the dose. While other options are available to meet release criteria

46 (e.g., additional source removal), the focus of this appendix is on the development of alternative 

47 exposure scenarios. 

48
M-2

1 M.2 Introducing Site-Specific Information to Modify or Develop Alternate 

2 Exposure Scenarios

3 Site-specific information can be divided into two broad categories: cultural information and 

4 physical information. Physical information includes such characteristics as the location, climate, 

5 topography, geology, soil types, and water availably of the site. Cultural information is basically 

6 how the land is used by the human population. Physical properties of land are essentially 

7 unchanging, while cultural properties are constantly changing. In reality, physical properties 

8 change (sometimes as a result of cultural activities), but the change is slow compared to the 

9 cultural use of the land.

10

11 Since the initial dose assessment for this process used the resident farmer exposure scenario 

12 with NRC–approved default pathways and parameters, the introduction of either cultural or 

13 physical information about a decontamination and decommissioning site is likely to reduce the 

14 TEDE. 

15

16 While the two screening scenarios are expected to be conservative, it should be noted that they

17 might be less conservative than expected. Therefore, care should be taken when eliminating 

18 exposure pathways and exposure scenarios should be consistent with site conditions.

19

20 M.2.1 Cultural Information

21 For developing alternative exposure scenarios, the most important element of cultural 

22 information about any site is the future land use over the period residual radioactivity at a site 

23 will persist at risk-significant levels. The future is assessed based on the past and the present. 

24 Experience has shown that, while this is an inexact science, the near future can be estimated 

25 with some degree of accuracy. Although the near future depends on the location, the culture, 

26 and what is being estimated, for the purposes of this section the near future is considered 

27 100 years; future land use over longer periods of time are uncertain and may not need to be 

28 considered. The key to the assessment of future land use is the current and past use of the 

29 land at both the site and in the region. If future land use is reasonably expected to be either 

30 urban or industrial, the resident farmer scenario may potentially be omitted from evaluation as a 

31 compliance scenario, although it may be evaluated to provide additional information for the 

32 decision-making process (i.e., as a less likely but plausible exposure scenario). The licensee 

33 will need to address both the dose from reasonably foreseeable land use (for the compliance 

34 calculation) and less likely but plausible land uses (to risk-inform the decision on whether 

35 release criteria are met). With adequate justification, pathways may be eliminated from detailed 

36 consideration in dose assessments. However, care should be taken to ensure that spatial and 

37 temporal variability is considered when justifying elimination of pathways (e.g., if poor water 

38 quality is used to justify elimination of the groundwater pathway, then the analyst should 

39 consider whether poor water quality conditions are expected to persist into the future).

40

41 M.2.1.1 Current Land Use

42 The determination of current land use is the initial step in the process of estimating future land 

43 use. Land use should be determined not only for the site but also for the land within an 80-km 

44 (50-mile) radius surrounding the site. This assessment of land use does not need to be 

45 complicated or detailed; it should be simple, dividing the land into only three categories: urban, 

46 rural, or industrial.

47

48 Current land use can be determined through one or more of the following information sources:
M-3

1

2 • site description

3 • land cover maps

4 • topographic maps

5 • planning agencies

6 • zoning laws and maps

7 • land use plans

8 • aerial photographs or satellite imagery

9 • demographic data

10 • cultural and natural resource data

11 • site visits

12 Most of the United States has codified land use/zoning, and many administrative areas have 

13 developed land use master plans. For this reason, the primary source of information on current 

14 land use should be the planning agencies of the State, county, and municipality in which the site 

15 resides. 

16

17 A large amount of land use information is available on the Internet at websites maintained by 

18 government agencies or universities. Land use data is often collected, stored, and analyzed 

19 using a GIS. An Internet search that includes the name of the government entity and the phrase 

20 “GIS data” or “land use data” will usually find the website where GIS data, including land use 

21 data, can be downloaded or ordered. For locations where State or local land use data are 

22 unavailable, data collected by one or more Federal agencies, such as the USGS, 

23 U.S. Department of Agriculture (USDA), or the U.S. Census Bureau, can usually be found. In 

24 particular, the Multi-Resolution Land Characteristics Consortium maintains nationwide land 

25 cover data, from which land use can be inferred, via the National Land Cover Database. This 

26 program is a joint effort of numerous Federal agencies, including USGS, USDA, EPA, Bureau of 

27 Land Management, National Oceanic and Atmospheric Administration, National Park Service, 

28 U.S. Forest Service, U.S. Fish and Wildlife Service, National Aeronautics and Space 

29 Administration, and the U.S. Army Corps of Engineers. The National GAP Analysis Program is 

30 also a potential source of land cover data for much of the United States.

31

32 Assumptions and predictions on future land uses are important considerations in the 

33 development of exposure scenario definitions and descriptions for analysis. If the site currently 

34 exists in a highly populated urban area, a residential farmer exposure scenario is very unlikely. 

35 Exposure scenarios for certain sites may exclude exposures via agricultural pathways if 

36 agricultural land uses are clearly incompatible with existing and anticipated future conditions at 

37 the sites. Exposures via ingestion of contaminated groundwater may be discounted if the 

38 affected groundwater is of such poor quality as to preclude human consumption.

39
M-4

1 M.2.1.2 Use of Ponds as Fisheries

2 In addition to physical limitations on the likelihood of a farmer using a pond as a fishery, the 

3 licensee should use local cultural information to determine if local residents currently engage in 

4 this practice. This question might be answered by the USDA county extension agent nearest to 

5 the decommissioning site. 

6

7 In addition to exposure of the average member of the critical group through use of a pond as a 

8 fishery, contaminated groundwater could also discharge to surface water as conceptualized in 

9 RESRAD-ONSITE more generally (see Appendix I). In contrast to the DandD conceptual 

10 model, RESRAD-ONSITE allows dilution in surface water, based on the ratio of the watershed 

area to the area of the contaminated zone.

1 11 If the licensee eliminates the use of the pond as a 

12 fishery, based on site-specific information, it should consider if the surface water pathway is 

13 otherwise viable (e.g., discharge of groundwater to a site stream used for fishing).

14

15 M.2.1.3 Future Land Use

16 Licensees should consider specific local conditions when deciding how far into the future it can 

17 estimate land use. The general range for estimation is within 100 years. In areas where rapid 

18 change has occurred in the past, this cutoff might be considerably less than 100 years, whereas 

19 in other areas, such as the heart of New York City, it may be reasonable to argue that urban 

20 conditions should prevail for more than 100 years.

21

22 The first step in estimating future land use is to determine the current land use at the site. The 

23 licensee should also learn the past use of the land, because it is the combination of past and 

24 present uses that should indicate what changes have occurred and the rate of those changes. 

25 This information should be used in a documented process that a reviewer would be able to 

26 follow. This documentation should include the types and sources of material used and how the 

27 licensee determined the final projected use. 

28

29 Land use and changes in land use within the 80-km (50-mile) radius of the site are included in 

30 this process. For example, a site that is currently located in a rural area within 16–32 km (10–

31 20 miles) of a growing metropolitan area would likely be in the suburbs of the metropolitan area 

32 within a decade or two, depending on population growth.

33

34 The 80-km (50-mile) radius is only a suggestion for determining the size of the area to consider. 

35 There may be valid reasons for increasing or decreasing the area of consideration, depending 

36 on local conditions and the length of time the residual radioactivity presents a potential risk. 

37 Other factors that may influence this decision are critical pathways and the estimated 

38 distribution of residual radioactivity.

39

40 M.2.1.4 Sources of Information for Determining Future Land Use

41 EPA’s Office of Solid Waste and Emergency Response (OSWER) publishes useful guidance 

42 related to land use assessments including OSWER Directive No. 9355.7-04, “Land Use in the 

43 CERCLA Remedy Selection Process,” dated May 25, 1995, and OSWER 9355.7-06P, “Reuse 

1 RESRAD-OFFSITE 4.0 contains a more sophisticated surface water model that considers contamination of 

surface water via (i) sediment erosion and runoff, (ii) groundwater discharge, and (iii) airborne deposition. The 

code considers the water balance, sediment balance, and radionuclide balance of the surface water body in order 

to compute the aqueous concentration of radionuclides in the surface water body.
M-5

1 Assessments: A Tool to Implement the Superfund Land Use Directive,” dated June 4, 2001. 

2 Published in 2010, OSWER 9355.7-19 reaffirms OSWER 9355.7-04 indicating that the 1995 

3 directive continues to provide useful guidance on consideration of reasonably anticipated future 

4 land use in the Superfund remedy selection process. OSWER 9335.7-04 and -19 are intended 

5 to facilitate future remedial decisions at Superfund sites by outlining a public process and

6 sources of information which should be considered in developing reasonable assumptions 

7 regarding future land use. Land use is integral to determining the efficacy of the remedy with 

8 respect to long-term effectiveness, protectiveness and compliance with all applicable or relevant 

9 and appropriate requirements.

10

11 Because so much of the information used to describe current land uses and to determine 

12 possible future land uses is geographic in nature, the sources include GIS providers at both the 

13 national and State levels. State GIS organizations should be able to direct the licensee to local 

14 sources for much of this information and, in many cases, may have links to that information 

15 directly from their data sites.

16

17 Land use information types include the following:

18

19 • zoning laws and maps

20 • state and community comprehensive master plan

21 • demographics

22 • historical population growth patterns

23 • current site location relative to other land uses

24 • federal/state land designations for surrounding lands

25 • threatened and endangered species

26 • natural resource inventory information

27 • floodplain/wetlands designations

28 • local/regional geologic information

29 • wellhead protection areas/aquifer recharge areas

30 • state comprehensive groundwater protection program

31 • historical aerial photography

32 • environmental justice issues

33 Federal agencies managing data that may be relevant to land use include the following:

34

35 • United States Geological Survey
M-6

1 o Multi Resolution Land Characteristics Consortium (MRLC) National Land Cover 

2 Dataset (NLCD)

3 o Water Resources Division (WRD) National Spatial Data Infrastructure (NSDI) Node

4 o EarthExplorer

5 o GAP Analysis Project

6 • United States Department of Agriculture

7 o National Resources Conservation Service (NRCS)

8 o National Agricultural Statistics Service (NASS)

9 o Economic Research Service (ERS)

10 o Forest Service (USFS)

11 • EPA

12 • National Oceanic and Atmospheric Administration

13 • United States Department of Interior

14 o Fish and Wildlife Service National Wetlands Inventory (NWI)

15 o Bureau of Land Management (BLM)

16 • Census Bureau

17 o TIGER data

18 • U.S. Department of Transportation

19 M.2.1.5 Urban Gardens

20 The subsistence farm associated with the resident farmer is unlikely to exist in an urban 

21 situation, but gardens are very likely in urban and suburban settings. The “Victory Gardens” of 

22 World War II demonstrate this possibility. Exceptions would be places like the concrete and 

23 steel core of large cities like New York, where gardens using locally obtained soil and water 

24 would be highly unlikely.

25
M-7

1 Documentation to Be Submitted to the NRC

2

3 Current Land Use should be documented by maps, descriptions, or other information

4

5 Estimates of Future Land Use should be supported by the documented process described in 

6 Sections M.2.1.3 and M.2.1.4.

7

8 State and local agencies should be contacted to get information on local land use and 

9 restrictions.

10

11 M.2.2 Physical Information about the Site

12 Physical information about the site includes climate, topography, vegetation, and, most 

13 importantly, water. Since water is a key factor in many of the pathways, its availability and 

14 proximity are very important.

15

16 M.2.2.1 Groundwater and Surface Water

17 Groundwater is present at some depth at almost every site. If groundwater is only found at 

18 great depths, surface water may be ephemeral and may exist only in response to rainfall or 

19 snowmelt. Surface water for the resident farmer may be a fish pond that is connected to the 

20 groundwater.

21

22 Several key questions about groundwater should be answered using site-specific information. 

23 The most important question is on the availability of water. Subsequent questions concern its 

24 quality and suitability for use.

25

26 M.2.2.2 Is Groundwater Available?

27 The first question that should be answered is “Is groundwater available as a resource for the 

28 exposure scenario resident?” More specific questions include the following:

29

30 (1) Is it shallow enough and does it have sufficient yield to be reasonably pumped by the

31 resident to irrigate a small farm and provide domestic drinking water?

32 (2) Is it shallow enough to intercept and connect to a fish pond, and does it have sufficient

33 yield to sustain the pond?

34 For the first question, the resident would need to drill a well into a permanent aquifer that has 

35 water sufficient for his needs and then be able pump that water into his house and onto his 

36 crops. Under the assumption that the well drilling and pumping technology available to the 

37 resident is similar to what exists today, it would not be unreasonable for the farmer to drill a well 

38 and pump from a depth of 120 m (400 ft), but this depth should be considered somewhat 

39 subjective. Specific local conditions should be considered when deciding how deep an aquifer a 

40 subsistence farmer would be able to use. A commercial farmer would be likely to drill much 

41 deeper than a subsistence farmer would.

42

43 Local trends in groundwater decline should be considered. In areas where groundwater is 

44 being withdrawn at an unsustainable rate, water levels may be dropping. If it can be reasonably 

45 assumed that this trend may continue, the licensee should consider it when assessing the 

46 availability of groundwater for the resident farmer.
M-8

1

2 If groundwater is not available at a reasonable depth for drinking water or irrigation, it may also 

3 not be available for a pond. Under these circumstances, the resident farmer exposure scenario 

4 can be devolved to exclude all three of the major pathways based on groundwater usage: 

5 irrigation, drinking water, and aquatic (pond). If groundwater is unavailable, it is also reasonable 

6 to exclude the use of surface water, since the aquatic exposure scenario considers the 

7 concentration of radionuclides in the surface water to be related to the concentration in the 

8 groundwater aquifer (NUREG/CR-5512, Volume 1, “Residual Radioactive Contamination from 

9 Decommissioning: Technical Basis for Translating Contamination Levels to Annual Total 

10 Effective Dose Equivalent,” 1992 (Kennedy, 1992).

11

12 Even if groundwater were available at a reasonable depth, the yield may be insufficient for all 

13 uses. Under these circumstances, the resident farmer exposure scenario can be modified to be 

14 more realistic considering the yield (e.g., pathways can be eliminated, or parameters adjusted to 

15 account for the lower yield). State and local agencies should be contacted to get input on 

16 groundwater use restrictions, groundwater protection regulations, and for any objections to 

17 elimination of groundwater related pathways.

18

19 Documentation to Be Submitted to the NRC

20

21 Groundwater Unavailable: A USGS or independent consultant report showing that either 

22 groundwater does not exist, or that it is too deep (e.g., more than 120 m (400 ft)) to reasonably 

23 be used by a subsistence farmer.

24

25 State and local agencies should be contacted to get input on groundwater use restrictions, 

26 groundwater protection regulations, and support for elimination of groundwater related 

27 pathways.

28

29 If groundwater is available for drinking or irrigation, it may not be available for a fish pond. It 

30 would not be reasonable to expect that the farmer would continually pump water into a pond to 

31 maintain it as a fishery. The groundwater would have to be shallow enough that a sufficient 

32 pond level would be maintained through its connection to the pond. This would mean the 

33 groundwater would have to be no deeper than about 5 m (15 feet). Information about local 

34 topography and specific conditions at each site could be used to adjust this number up and 

35 down. If groundwater is not available for the pond, the aquatic pathway should be removed 

36 from the resident farmer exposure scenario.

37

38 Documentation to Be Submitted to the NRC

39

40 Groundwater Unavailable: A USGS or independent consultant report showing that either 

41 groundwater does not exist, or that it is too deep (e.g., more than 5 m (15 ft)) to connect to a 

42 surface water pond. 

43

44 State and local agencies should be contacted to get information on local land use practices, and 

45 to support elimination of the surface water pathway from consideration.

46

47 M.2.2.3 Is Groundwater Quality Suitable for Aquatic Life?

48 The quality of surface water is critical to support aquatic life and is affected by (1) the chemical 

49 and physical conditions that exist in the pond, (2) runoff from exposed soil, and (3) condensation 
M-9

1 or entrapment of contaminants from the air (e.g., pollutants, acid rain). EPA provides 

2 recommended surface water standards to support aquatic life, including minimum dissolved 

3 oxygen values and maximum contaminant levels (MCLs).

4

5 The concentration of dissolved oxygen in surface water is affected by the biochemical oxygen 

6 demand of the ecosystem. Sedimentation of suspended solids can cause a buildup of organic 

7 matter in sediments. These materials undergo metabolic degradation by aerobic soil 

8 microorganisms with the concomitant depletion of dissolved oxygen. Other contaminants, such 

9 as dissolved ammonia, can contribute to oxygen depletion by nitrification. Ammonia is toxic to 

10 fish and other aquatic animals. The presence of coliform bacteria is sometimes indicative of 

11 other, more virulent pathogens in surface water and should be considered when fish or other 

12 aquatic animals are produced for human consumption. If the quality of the groundwater (and 

13 hence the pond) lies outside of the acceptable standards for aquatic life, the licensee can 

14 remove the aquatic pathway from the resident farmer exposure scenario.

15

16 Documentation to Be Submitted to the NRC

17

18 Groundwater Unsuitable for Aquatic Life: A USGS or independent consultant report showing 

19 that groundwater quality is poorer than the standards listed for this use. 

20

21 State and local agencies should be contacted to get supporting information on elimination of the 

22 aquatic pathway based on groundwater quality.

23

24 M.2.2.4 Is Groundwater Quality Suitable for Agriculture?

25 The quality of groundwater for agricultural uses varies depending on the type of agribusiness or 

26 agricultural enterprises conducted at the site. For example, groundwater with infiltrated 

27 fertilizers and herbicides can be very beneficial to crop land through irrigation but can have an 

28 adverse effect on the health and productivity of livestock and poultry. Based on extensive 

29 USDA studies, recommended limits for chemicals in drinking water for livestock and poultry 

30 have been published and are available in an Internet search.

31

32 In addition to acute and chronic toxicity from chemicals, high concentrations of dissolved solids 

33 in drinking water can lead to various degrees of mineral toxicity in animals. Most minerals and 

34 dissolved solids found in water provide nutritional benefits when present within limited 

35 concentration ranges (e.g., selenium). At high concentrations, however, common minerals can 

36 lead to acute or chronic changes that affect the quality of animal products and overall 

37 productivity. 

38

39 The licensee should consider salinity, or total dissolved solids, when evaluating groundwater for 

40 animal consumption. Although 10,000 mg/L is acceptable under some conditions, the health, 

41 and ultimately the productivity, of animals are affected to various degrees by the salinity. 

42 Table M.1 provides a breakdown of conditions that have been observed and documented in 

43 livestock and poultry for various concentrations of dissolved solids in drinking water.
M-10

1 Table M.1 Effects of Salinity of Drinking Water on Livestock 

2

Salinity Level Limits for 

Drinking Water Conditions

Less than 1,000 mg/L Excellent for all classes of livestock and poultry

1,000–3,000 mg/L Satisfactory for all classes of livestock. May cause temporary and 

mild diarrhea in livestock and poultry not accustomed to such levels 

but should not affect their health or performance. 

3,000–5,000 mg/L Satisfactory for livestock, although they might very possibly cause 

mild diarrhea or be refused by animals not accustomed to such 

levels. Increased morbidity and decreased growth in poultry.

5,000–7,000 mg/L Marginal quality for livestock. Not suitable for poultry and pregnant 

and lactating animals.

7,000–10,000 mg/L Not suitable for pigs, and considerable risk for pregnant and lactating 

animals. In general, their use should be avoided.

Above 10,000 mg/L Unacceptable 

3

4 If the quality of the groundwater is less than what is considered acceptable for irrigation, the 

5 licensee should remove the irrigation pathway from the resident farmer exposure scenario.

6 If the quality of the groundwater is less than what is considered acceptable as a drinking source 

7 for farm animals, the licensee should remove that pathway from the resident farmer exposure 

8 scenario.

Documentation to Be Submitted to the NRC

Groundwater Unsuitable for Agriculture: A USGS or independent consultant report 

showing that groundwater quality is poorer than the standards listed for this use.

State and local agencies should be contacted to get supporting information on elimination of

use of groundwater for irrigation or livestock consumption based on quality.

9

10 M.2.2.5 Is Groundwater Suitable for Drinking Water?

11 The licensee can address this question by comparing the quality of the groundwater with EPA 

12 drinking water standards. Regulations for public water systems in the United State are in 

13 Title 40 of the Code of Federal Regulations (40 CFR) Part 141, “National Primary Drinking 

14 Water Regulations.” Primary drinking water standards specify approval limits for 

15 microorganisms, including bacteria and viruses, specific inorganic and organic chemicals, 
M-11

1 radionuclides, and turbidity, while secondary standards identified in 40 CFR Part 143, “National 

2 Secondary Drinking Water Regulations,” recommend limits on benign contaminants and define 

3 physical characteristics that address the aesthetics of drinking water (e.g., color and odor).

4

5 Drinking water standards are available for (1) inorganic chemicals, (2) organic chemicals, 

6 (3) radionuclides, and (4) microorganisms. Although turbidity is a measured physical

7 parameter, it is included with microorganisms, because turbid water is generally associated with

8 microorganisms or provides a medium for microbial growth. Although the secondary standards

9 are not regulated, they serve as a guide for water quality and may, in some instances, be

10 regulated at the State or local level.

Documentation to Be Submitted to the NRC

Groundwater Not Potable: A USGS or independent consultant report that shows that 

groundwater quality is poorer than the standards listed for this use.

EPA regulations should be used to support elimination of the groundwater pathway based on 

groundwater quality. State and local agencies should be contacted to get information to 

support elimination of the groundwater pathway.

11

12

13 M.2.3 Topography and Soil

14 M.2.3.1 Is Soil Suitable for Agriculture?

15 Soil performs several functions related to plant growth. It forms a media in which roots 

16 penetrate, thereby providing a source of stability and nourishment. Nourishment can be 

17 provided by the nutrients available in the soil, by fertilizers, or by soil amendments.

18

19 With suitable fertilizers or soil amendments, plants can readily be grown in “soil free” materials, 

20 such as mineral sand, gravel, perlite, pumice, crushed bricks, or glass wool. Consequently, the 

21 absence of soil in the traditional sense at a site does not eliminate plant ingestion as a pathway. 

22 Because soilless gardening requires more management than traditional gardening methods, it is 

23 more likely to be used for growing vegetables and herbs than for the production of commodity 

24 items such as grains or livestock fodder (R.E. Nicolls, “Beginning Hydroponics Soilless 

25 Gardening,” 1997 (Nicholls, 1997)). Agriculture could be excluded from an exposure scenario if 

26 the site is an outcropping of bedrock without appreciable soil or debris that could serve to 

27 anchor plants. 

28

29 Areas consisting of made land, where there is abundant debris and cobbles with little or no soil, 

30 would not lend themselves to mechanized agriculture in short-term exposure scenarios. In the 

31 absence of mechanized agriculture, commodity food items and fodder are not likely crops. 

32 However, it would be difficult to exclude vegetable gardens from exposure scenarios at such 

33 sites. In addition, it would be difficult to justify exclusion of livestock forage from exposure 

34 scenarios for such sites.

35 Agriculture pathways could be eliminated in short-term exposure scenarios if the soil is outright 

36 toxic or inhospitable to plants. As examples, (1) no agriculture is apt to occur on the bed of a 

37 dry salt lake, and (2) crops are not apt to be grown in made land that contains such a high 
M-12

1 percentage of concrete materials that extraordinary efforts would be required to maintain the soil 

2 pH in a range that is tolerated by plants.

3

4 If it can be documented that the soil at this site would not support the resident farmer’s 

5 agricultural efforts, the licensee could eliminate or modify this pathway.

6

Documentation to Be Submitted to the NRC

Soil Unsuitable for Agriculture: A Natural Resources Conservation Service (Soil Conservation 

Service) or independent consultant report that shows the quality of soil is poorer than the 

standards listed for this use.

State and local agencies should be contacted to get information to support elimination of 

agricultural pathways based on soil suitability.

7

8 M.2.3.2 Is Topography Suitable for Agriculture?

9 In the past few hundred years, dikes have been used to convert submerged land into productive 

10 farmlands. Today, explosives and earth-moving equipment can easily change features of the 

11 landscape, making them suitable for agricultural or residential use. Consequently, locality or 

12 accessibility may form a basis for eliminating certain agricultural pathways from exposure 

13 scenarios in the next century but not for a period of 1,000 years.

14

15 Ignoring the fact that topography may change with time as a result of civil engineering projects, 

16 there are probable limits to the types of terrain where mechanized agriculture can be used. 

17 Tractors are likely to always be unstable on slopes, so there is probably a practical limit on 

18 slopes that can be put under mechanized agriculture. While there is no predictable maximum 

19 safe slope that tractors may traverse without the danger of rollover, operating a tractor on a 

20 30-degree (2 to 1) slope is so hazardous that the average member of the critical group is not

21 likely to attempt it.

22 In the absence of mechanized agriculture, persons are more likely to practice gardening than to 

23 grow commodity food items. In fact, gardening is commonly practiced on hillsides using contour 

24 rows, terraces, or raised beds to minimize erosion. They may also be used to allow livestock to 

25 forage.

26 If the topography at the site is too steep or too erratic to support the type of farming expected 

27 within the resident farmer exposure scenario, the licensee could justify modification of the 

28 agricultural pathway in accordance with this finding. There may also be aspects of the 

29 topography that would limit farming or other specific activities at the site.

30
M-13

Documentation to Be Submitted to the NRC

Topography Unsuitable for Agriculture: A USGS or similar topographic map, hand-drawn 

map, or description with enough detail to illustrate the topography that limits farming at this 

site.

State and local agencies should be consulted regarding local land use practices and to obtain 

information to support modification of the agriculture pathway based on topography.

1

2 M.3 Summary

3 The process presented in this appendix is compatible with the Decommissioning Framework

4 discussed in Chapter 1. It discusses use of site-specific information to develop alternative 

5 exposure scenarios by eliminating pathways from the default resident farmer exposure scenario. 

6 Once the TEDE to an average member of the critical group drops below 0.25 mSv/y 

7 (25 mrem/y), the process is complete, and the licensee may proceed to license termination. 

8 Following the initial dose assessment and each of the iterative dose assessments, sensitivity 

9 analyses help the licensee focus on the introduction of evidence that can rule out those 

10 pathways that are responsible for the high dose.

11 The licensee uses physical and cultural information to answer a series of questions about the 

12 site. The future use of the land may be important in deciding what assumptions the licensee 

13 can make about the starting exposure scenario. Information on current land use, past land use, 

14 and a history of land use changes can determine the probable future use of the land. If the 

15 future land use can reasonably be predicted to be either urban or industrial, the resident farmer 

16 exposure scenario can be bypassed, allowing the licensee to concentrate on these other two 

17 land uses. 

18 The residential farmer exposure scenario is meant to be applied to sites with land and water 

19 residual radioactivity, and the building occupancy exposure scenario is to be applied to sites 

20 with contaminated structures. If a resident farmer exposure scenario is assumed, the most 

21 important aspect of the physical nature of the site is the nature and availability of water. The 

22 answers to each of four critical questions about water at the site can determine if major 

23 pathways can be removed from the exposure scenario. If groundwater is not available, the 

24 licensee can remove all of the pathways that rely on groundwater as a key component 

25 (irrigation, aquatic, and drinking). If groundwater is not suitable for aquatic life, the aquatic 

26 pathway can be removed. If groundwater is not suitable for agriculture, irrigation and drinking 

27 water pathways can be removed. If the water is not potable, the drinking water pathway can be 

28 removed. Detailed discussion in this appendix helps the licensee answer these questions, 

29 understand the standards that it would have to meet to rule out this pathway, and identify the 

30 documentation that it would have to present to the NRC.
M-14

1 M.4 Bibliography

2 Bonano, E.J., Hora, S.C., Keeney, R.L., and D. Winterfeldt, 1990, “Elicitation and Use of Expert 

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4 NUREG/CR-5411 (SAND89-1821), U.S. Nuclear Regulatory Commission, Washington, DC; 

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6 Cochran, J.R., 1996, “A Proposed Alternative Approach for Protection of Inadvertent Human 

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8 Waste Management ’96, Session 59, Waste Management Symposia, Inc., Tucson, AZ.

9 Cranwell, R.M., Campbell, J.E., Helton, J.C., Iman, R.L., Longsine, D.E., Ortiz, N.R., Runkle, 

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11 Waste: Final Report,” NUREG/CR-2452 (SAND81-2573), U.S. Nuclear Regulatory

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13 Cranwell, R.M., Guzowski, R.V., Campbell, J.E., and Ortiz, N.R., 1990. “Risk Methodology for 

14 Geologic Disposal of Radioactive Waste: Scenario Selection Procedure,” NUREG/CR-1667 

15 (SAND80-1429), U.S. Nuclear Regulatory Commission, Washington, DC; Sandia National 

16 Laboratories, Albuquerque, NM.

17 U.S. Department of Energy, 1994. “Decommissioning Handbook,” DOE/EM–0142P, U.S. DOE, 

18 Office of Environmental Restoration.

19 U.S. Department of Energy, 1998. “Effects of Future Land Use Assumptions on Environmental 

20 Restoration Decision Making.” DOE/EH–0413/980, U.S. DOE, Office of Environmental Policy 

21 and Assistance.

22 U.S. EPA, 2011. “Exposure Factors Handbook: 2011 Edition.” EPA/600/R-090/052F. Office of 

23 Research and Development, U.S. Environmental Protection Agency, Washington, DC. 

24 U.S. EPA, 1995. OSWER Directive No. 9355.7-04, “Land Use in the CERCLA Remedy 

25 Selection Process,” U.S. Environmental Protection Agency, Office of Solid Waste and 

26 Emergency Response, Washington, DC.

27 U.S. EPA, 2001. OSWER Directive No. 9355.7-06P, “Reuse Assessments: A Tool to 

28 Implement the Superfund Land Use Directive,” U.S. Environmental Protection Agency, Office of 

29 Solid Waste and Emergency Response, Washington, DC.

30 U.S. EPA, 2010. OSWER Directive No. 9355.7-19, “Considering Reasonably Anticipated 

31 Future Land Use and Reducing Barriers to Reuse at EPA-lead Superfund Remedial Sites,” 

32 U.S. Environmental Protection Agency, Office of Solid Waste and Emergency Response, 

33 Washington, DC.

34 Guzowski, R.V. 1991. “Evaluation of Applicability of Probability Techniques to Determining the 

35 Probability of Occurrence of Potentially Disruptive Intrusive Events at the Waste Isolation Pilot 

36 Plant,” SAND90-7100, Sandia National Laboratories, Albuquerque, NM.

37 Haaker, R., T. Brown, and D. Updegraff. “Comparison of the Models and Assumptions used in 

38 the DandD 1.0, RESRAD 5.61, and RESRAD–Build 1.50 Computer Codes with Respect to the 
M-15

1 Residential Farmer and Industrial Occupant Scenarios Provided in NUREG/CR–5512.” Draft 

2 Report for Comment, October 1999. (Note: This report is also SAND99-2147.)

3 Hora, S.C., D. von Winterfeldt and K.M. Trauth, 1991. “Expert Judgement on Inadvertent 

4 Human Intrusion into the Waste Isolation Pilot Plant.” SAND90-3063, Sandia National 

5 Laboratories, Albuquerque, NM.

6 IAEA, 1982. “General Models and Parameters for Assessing the Environmental Transfer of 

7 Radionuclides from Routine Releases,” International Atomic Energy Agency; Safety Series 

8 No. 57, Vienna, Austria:

9 IAEA, 1995. “Criteria for Clean-Up of Contaminated Areas,” Consultants Meeting Draft Report. 

10 International Atomic Energy Agency, Vienna, Austria, December 4–8, 1995.

11 Kennedy, Jr., W.E., and D.L. Strenge. 1992. “Residual Radioactive Contamination from 

12 Decommissioning: Technical Basis for Translating Contamination Levels to Annual Total 

13 Effective Dose Equivalent,” NUREG/CR-5512, Volume 1, U.S. Nuclear Regulatory Commission, 

14 Washington, DC. 

15 Kozak, M.W., N.E. Olague, R.R. Rao and J.T. McCord, 1993. “Evaluation of a Performance 

16 Assessment Methodology for Low-Level Radioactive Waste Disposal Facilities: Evaluation of 

17 Modeling Approaches, NUREG/CR-5927, Volume 1, U.S. Nuclear Regulatory Commission, 

18 Washington, DC. 

19 Leigh, C.D., B.M. Thompson, J.E. Campbell, D.E. Longsine and R.A Kennedy, 1993. “User’s 

20 Guide for GENII–S,” SAND91-056, Sandia National Laboratories, Albuquerque, NM.

21 Locke, P.A., M.C. Carney, Tran, N.L., T.A. Burke, 1998. Chemical and Radiation Environmental 

22 Risk Management: Foundations, Common Themes, Similarities and Differences, Workshop 

23 Proceedings, Annapolis, MD.

24 National Research Council. 1995. Technical Basis for Yucca Mountain Standards, National 

25 Academy Press, Washington, DC. 

26 NEA, 1992. “Systematic Approaches to Scenario Development,” Nuclear Energy Agency, 

27 Organisation for Economic Co-Operation and Development, Paris, France.

28 Nicholls, R.E., 1997. “Beginning Hydroponics Soilless Gardening,” Running, Running Press, 

29 Philadelphia, PA.

30 NRC, 1977. Regulatory Guide 1.109, “Calculation of Annual Doses to Man from Routine 

31 Releases of Reactor Effluents for the Purpose of Evaluating Compliance with 10 CFR Part 50, 

32 Appendix I.” U.S. Nuclear Regulatory Commission, Washington, DC.

33 NRC, 1981a, “Proposed Rules for Licensing Requirements for Land Disposal of Radioactive 

34 Waste,” Federal Register, Vol. 46, No. 142, pp. 38081–38015.

35 NRC, 1981b, NUREG-0782, “Draft Environment Impact Statement on 10 CFR Part 61,

36 ‘Licensing Requirements for Land Disposal of Radioactive Waste,’” U.S. Nuclear Regulatory 

37 Commission, Washington, DC.
M-16

1 NRC, 1982, NUREG-0945, “Final Environment Impact Statement on 10 CFR Part 61, ‘Licensing 

2 Requirements for Land Disposal of Radioactive Waste,’”, U.S. Nuclear Regulatory Commission, 

3 Office of Nuclear Material Safety and Safeguards, Washington, DC.

4 NRC, 1990, NUREG/CR-5256, “Components of an Overall Performance Assessment 

5 Methodology,”, U.S. Nuclear Regulatory Commission, Office of Nuclear Material Safety and 

6 Safeguards, Washington, DC.

7 NRC, 1993. Title 10 of the Code of Federal Regulations, Part 20, “Standards for Protection 

8 against Radiation.” U.S. Nuclear Regulatory Commission, Washington, DC.

9 NRC. 1998. NUREG-1549, “Decision Methods for Dose Assessment to Comply with 

10 Radiological Criteria for License Termination.” U.S. Nuclear Regulatory Commission, Office of 

11 Nuclear Regulatory Research, Washington, DC.

12 Sara, M.N., Standard Handbook for Solid and Hazardous Waste Facility Assessment, Lewis 

13 Publishers, Boca Raton, FL.

14 Trauth, K.M., S.C. Hora and R.V. Guzowski, 1993. “Expert Judgement on Markers to Deter 

15 Inadvertent Human Intrusion into the Waste Isolation Pilot Plant.” SAND92-1382, Sandia 

16 National Laboratories, Albuquerque, NM.

17 Viessman, Jr., Warren; Hammer, Mark J. “Water Supply and Pollution Control.” Fourth Edition. 

18 New York: Harper and Row; 1985.

19 Wernig, M.A., A.M. Tomasi, F.A. Duran, and C.D. Updegraff. “Residual Radioactive 

20 Contamination from Decommissioning, User’s Manual, Draft Report,” May 1999.

21 Wood, D.E., Curl, R.U., Buhl, T.E., Cook, J.R., Dolenc, M.R., Kocher, DC, Napier, B.A., Owens, 

22 K.W., Regnier, E.P., Roles, G.W., Seitz, R.R., Thorne, D.J., and Wood, M.I., 1992,

23 “Performance Assessment Task Team Draft Progress Report,” DOE/LLW-157, Revision 0,

24 Idaho National Engineering Laboratory, Idaho Falls, ID. Prepared for the U.S. Department of

25 Energy.

26 Wood, D.E., Curl, R.U., Armstrong, D.R., Cook, J.R., Dolenc, M.R., Kocher, DC, Owens, K.W., 

27 Regnier, E.P., Roles, G.W., Seitz, R.R., and Wood, M.I., 1994, “Performance Assessment Task 

28 Team Progress Report,” DOE/LLW-157, Revision 1, Idaho National Engineering Laboratory, 

29 Idaho Falls, ID. Prepared for the U.S. Department of Energy.
ALARA ANALYSES

N-1

1 N.1 Introduction

2 To terminate a license, a licensee must demonstrate that it has met the dose criteria in 

3 Subpart E and the requirements for reducing exposures ALARA. This section describes 

4 methods acceptable to the NRC staff for determining when it is reasonable to further reduce the 

5 (future) exposures of members of the public below the dose criteria. This section does not 

6 apply to, nor replace guidance for, operational ALARA programs. This guidance does involve 

7 the same principle as the operational ALARA guidance, as described in NRC Regulatory 

8 Guide 8.8, Revision 3, “Information Relevant to Ensuring that Occupational Radiation 

9 Exposures at Nuclear Power Stations Will Be as Low as Is Reasonably Achievable, issued 

10 June 1978:

11

12 “Reasonably achievable” is judged by considering the state of technology and the 

13 economics of improvements in relation to all the benefits from these improvements. 

14 (However, a comprehensive consideration of risks and benefits will include risks from non15 radiological hazards. An action taken to reduce radiation risks should not result in a 

16 significantly larger risk from other hazards.) 

17

18 For ALARA, as it relates to the license termination criteria, all licensees should use typical good 

19 practices, such as floor and wall washing, removal of readily removable radioactivity in buildings 

20 or in soil areas, and other good housekeeping practices. In addition, in the FSSR, licensees 

21 should describe how they employed these practices to achieve the final activity levels.

22

23 In light of the conservatism in the building surface and surface soil generic screening levels 

24 developed by the NRC, the staff presumes, absent information to the contrary, that licensees 

25 who remediate building surfaces or soil to the generic screening levels (see Appendix H) do not 

26 need to provide analyses to demonstrate that these screening levels are ALARA. In addition, if 

27 residual radioactivity cannot be detected, it may be assumed that it has been reduced to levels 

28 that are ALARA. An ALARA analysis is also unnecessary in cases where soil removal is

performed.

1 29 Therefore, the licensee may not need to conduct an explicit analysis to meet the 

30 ALARA requirement. 

31

32 After a site has been decommissioned and the license terminated in accordance with the criteria 

33 in 10 CFR Part 20, “Standards for Protection against Radiation, Subpart E, “Radiological Criteria 

34 for License Termination,” or after part of a facility or site has been released for unrestricted use 

35 in accordance with 10 CFR 50.83 and in accordance with the criteria in Subpart E, the 

36 Commission will require additional cleanup only if based on new information, it determines that 

37 the criteria of this subpart were not met and residual radioactivity remaining at the site could 

38 result in significant threat to public health and safety. Because ALARA represents an 

39 optimization technique to reduce doses below the dose criteria, it is not considered reasonable 

40 to reopen consideration of a previously released radiological area that meets the appropriate 

41 dose criterion.

42

43 In general, a method for determining whether exposures that would result from a proposed 

44 license termination approach are ALARA would have the following characteristics.

45

46 • The method is simple. The method for most licensee applications should be simple,

47 because the effort needed for very sophisticated models cannot generally be justified. In

1 See preamble to the license termination rule found at 62 FR 39058; July 21, 1997.
N-2

1 an ALARA analysis of a remediation action, the primary benefit (i.e., the collective 

2 radiation dose that may actually be averted in the future) is uncertain because future 

3 land uses, the number of people that may actually occupy a site, and the types of 

4 exposure scenarios are all uncertain. These uncertainties mean that the future collective 

5 dose cannot be known with precision. Because of the inherent limitation on the ability to 

6 precisely determine the future collective dose at a particular site, it is not useful to 

7 perform a complex analysis when a simple analysis can be appropriate. Licensees may 

8 use more complex or site-specific analyses if more appropriate for their specific 

9 situations (e.g., restricted release analyses, situations that include a number of 

10 unquantifiable benefits and costs).

11

12 • The method is not biased and uses appropriate dose modeling to relate

13 concentrations to dose. The determination of ALARA should not be biased. This is

14 different from demonstrating compliance with a dose limit. The analyses for dose

15 assessments and surveys for compliance with the dose criteria described in this volume

16 include a reasonably conservative bias for demonstrating compliance. Unlike a

17 demonstration of compliance, an ALARA analysis is an optimization technique that

18 seeks the proper balance between costs and benefits below the dose limit. To achieve a

19 proper balance, each factor in the ALARA analysis should be determined with as little

20 bias as possible. If the ALARA analysis were intentionally biased, it would likely cause a

21 misallocation of resources and could deprive society of the benefits from other uses of

22 the resources. Thus, the ALARA analysis should provide an unbiased analysis of the

23 remediation action, which can both avert future dose (a benefit to society) and incur

24 costs (a potential detriment, because it can deprive future generations of the return on

25 the investment of this money). Sections N.2 through N.6 discuss the methods that

26 licensees should use in estimating benefits and detriments, or costs, including scenarios,

27 models, and parameters for relating concentration to dose at a site.

28

29 • The method is usable as a planning tool for remediation. Before starting a

30 remediation action, the licensee should be able to determine generally what

31 concentration of residual radioactivity would require a remediation action to meet the

32 ALARA requirement. It would be inefficient if the licensee could not tell whether the area

33 would pass the ALARA test until after the remediation. Establishing ALARA

34 postremediation could also result in it being less likely for a licensee to remediate below

35 the dose limit(s) because of the additional manpower startup costs associated with

36 performing additional remediation.

37

38 • As much as possible, the method uses the results of surveys conducted for other

39 purposes. The demonstration that the ALARA requirement has been met should not

40 require surveys beyond those already performed for other purposes, such as the

41 characterization survey and the FSS. It would be inefficient (and unnecessary) to collect

42 additional sets of measurements to demonstrate that remediation actions were taken

43 wherever appropriate to meet the ALARA requirement, if the licensee could use

44 measurements undertaken for other purposes.

45 Issues raised in 72 FR 46102 that pertained to guidance in the previous revision to this volume

46 (Appendix N in NUREG-1757, Volume 2, Revision 1) are addressed in Revision 2 to this 

47 appendix. Issues included guidance on (i) discounting rates and (ii) monetary value of collective 

48 dose averted among others.

49
N-3

1

2 N.2 ALARA as it Applies to NRC Decommissioning Regulations

3 ALARA, as defined in 10 CFR 20.1003, means:

4

5 making every reasonable effort to maintain exposures to radiation as far 

6 below the dose limits in this part as is practical consistent with the purpose 

7 for which the licensed activity is undertaken, taking into account the state of 

8 technology, the economics of improvements in relation to state of 

9 technology, the economics of improvements in relation to benefits to the 

10 public health and safety, and other societal and socioeconomic 

11 considerations, and in relation to utilization of nuclear energy and licensed 

12 materials in the public interest.

13

14 The general requirement for ALARA is stated in 10 CFR 20.1101(b)

15

16 The licensee shall use, to the extent practical, procedures and engineering controls based 

17 upon sound radiation protection principles to achieve occupational doses and doses to 

18 members of the public that are ALARA.

19

20 This is a general form of ALARA that conceptually applies to all radiation protection dose limits 

21 under 10 CFR Part 20, as confirmed in Shieldalloy Metallurgical Corp. (Decommissioning of the 

22 Newfield, New Jersey Site), CLI-11-12, 74 NRC 460, 480 (Oct. 12, 2011), which states that 

23 “ALARA is a general requirement for all ‘doses to members of the public’ established in the 

24 ‘Radiation Protection Programs’ in 10 C.F.R. Part 20, including the license termination dose 

criteria.2 25 ” As such, all licensees should establish programs and controls with ALARA concepts 

26 in mind to reduce occupational doses and doses to members of the public to ALARA levels.

27

28 For license termination with either restricted or unrestricted release, doses to a member of the 

29 public must meet the release criteria (e.g., 0.25 mSv [25 mrem] per year for unrestricted release

30 or restricted release with institutional controls in place) but also ALARA. Therefore, ALARA 

31 should be considered in conjunction with the applicable release criteria including unrestricted 

32 release, restricted release with institutional controls in effect, and restricted release with 

33 institutional controls no longer in effect. 

34

35 The licensee’s implementation of the general ALARA principle from 10 CFR 20.1101(b) should 

36 be appropriate for the specific regulatory basis being utilized, which could include

37 10 CFR 20.1402 (“Radiological Criteria for Unrestricted Use”), 10 CFR 20.1403 (“Criteria for 

38 License Termination Under Restricted Conditions”), or 10 CFR 20.1404 (Alternate Criteria for 

39 License Termination) for unrestricted release, restricted release, or alternative release criteria, 

40 respectively. The following sections discuss general ALARA considerations for each of these 

41 regulatory bases. 

42

43 N.2.1 ALARA in Unrestricted Use

44 As described in the previous section, ALARA is a requirement of 10 CFR 20.1402. Thus, for 

45 most instances of unrestricted release, the ALARA requirement is to consider reasonably 

2 The applicability of 10 CFR 20.1101(b) to the 10 CFR Part 20, Subpart E, license termination dose criteria is 

discussed in Commission Order Shieldalloy, CLI-11-12, 74 NRC 460, (480-481). This order also provides 

additional details on the specific application of the ALARA regulation to the license termination rule.
N-4

1 achievable means of reducing dose below the 0.25 mSv (25 mrem) per year dose criterion of 

2 10 CFR 20.1402. 10 CFR 20.1402 indicates that a site will be considered acceptable for 

3 unrestricted use if residual radioactivity meets the 0.25 mSv per year dose standard “and that 

4 the residual radioactivity has been reduced to levels that are as low as reasonably achievable 

5 (ALARA)”. 10 CFR 20.1402 goes on to state that determination of the levels which are ALARA 

6 must take into account consideration of any detriments, such as deaths from transportation 

7 accidents, expected to potentially result from decontamination and waste disposal. 

8

9 N.2.2 ALARA in Restricted Release

10 Under 10 CFR 20.1403, a licensee must meet certain eligibility criteria before the NRC can 

11 terminate a license under restricted use conditions. Among other things, under 10 CFR 

12 20.1403(a) a site will be considered acceptable for license termination under restricted 

conditions if the licensee can demonstrate that further reductions in residual radioactivity3 13

14 necessary to comply with the provisions of 10 CFR 20.1402 would result in net public or 

15 environmental harm or were not being made because the residual levels associated with 

16 restricted conditions are ALARA. 

17

18 This eligibility determination implements the NRC’s preference for licensees to decommission to 

19 the unrestricted use criteria. The ALARA principle incorporated into § 20.1403(a) serves as a 

20 regulatory tool to limit the use of restricted release—effectively, to screen out sites that should 

21 be removing contamination to achieve unrestricted use. The NRC expects licensees to make 

22 every reasonable effort to achieve unrestricted release. Specifically, the requirement calls for a 

23 licensee seeking to use restricted release to analyze whether it would be cost-beneficial to 

24 remove enough radioactive contamination from the site so that doses to the public are no higher 

25 than 0.25 mSv/y (25 mrem/y) without reliance on restricted release controls. Such reduction of 

26 radioactivity is accomplished through removal of radioactive material or site decontamination. 

27 There are generally two alternative analyses that may be used to weigh the costs and benefits 

28 of removing radioactive contamination: (1) compare the potential benefits to the potential costs 

29 that are typically evaluated in an ALARA analysis, or (2) consider the net public and 

30 environmental harm as a cost and compare those costs against the health and environment31 related benefits of removing radioactive contamination. Sections N.3–N.6 contain details on the 

32 implementation of these two analyses. 

33

34 N.2.2.1 Restricted Release—ALARA with Institutional Controls and without

35 Once the eligibility for restricted release has been established under 10 CFR 20.1403(a), the

36 general ALARA requirement still applies to the other criteria of 10 CFR 20.1403 for restricted 

37 release. Specifically, 10 CFR 20.1403(b) and 20.1403(e) enumerate dose limits for restricted 

38 release with institutional controls in place and without, respectively. Despite the availability of 

39 institutional controls and engineered barriers to reduce dose under restricted release, licensees 

40 must reduce residual radioactivity to levels that provide reasonable assurance that doses will 

41 not exceed a maximum value or “cap” specified in paragraph (e) if institutional controls are no 

42 longer in effect. As stated above, reduction of radioactivity is accomplished through removal of 

43 radioactive material or site decontamination.

3 As noted in Shieldalloy Metallurgical Corp. (Decommissioning of the Newfield, New Jersey Site), CLI-13-6, 78 NRC 

155, 168-170 (August 5, 2013), such reduction of radioactivity is only accomplished through removal of radioactive 

material, and the eligibility for restricted release must consider whether further reductions, without considering the 

impacts of institutional controls and engineered barriers, are cost beneficial.
N-5

1

2 Paragraph (b) requires provisions for legally enforceable institutional controls that provide 

3 reasonable assurance that the TEDE from residual radioactivity distinguishable from 

4 background to the average member of the critical group will not exceed 0.25 mSv (25 mrem) per 

5 year. Although not specifically included in paragraph (b), the general ALARA requirement 

6 applies and the dose from residual radioactivity must be as low as reasonably achievable even 

7 with institutional controls in place.

8

9 Paragraph (e) requires, in part, that 

10

11 Residual radioactivity at the site has been reduced so that if the institutional controls 

12 were no longer in effect, there is reasonable assurance that the TEDE from residual 

13 radioactivity distinguishable from background to the average member of the critical 

14 group is as low as reasonably achievable and would not exceed either—

15

16 (1) 1.0 mSv (100 mrem) per year; or

17

18 (2) 500 mrem (5 mSv) per year provided the licensee—

19

20 (i) Demonstrates that further reductions in residual radioactivity necessary to

21 comply with the 1.0 mSv/y (100 mrem/y) value of paragraph (e)(1) of this section

22 are not technically achievable, would be prohibitively expensive, or would result

23 in net public or environmental harm.

24

25 As the underlined portions in the preceding text indicate, paragraph (e) includes an explicit 

26 reference to ALARA and a requirement for an additional analysis similar to one for ALARA in 10 

27 CFR 20.1403(a). First, the licensee must provide reasonable assurance that through reduction 

28 of residual radioactivity that the TEDE if institutional controls fail is ALARA. Second, to use the 

29 larger 5 mSv (500 mrem) per year cap, licensees must make an additional demonstration using 

30 criteria similar to those for 20.1403(a). For subparagraph (e)(2)(i), however, the requirement is 

31 not simply to reduce to as low as reasonably achievable, but to show that meeting the dose limit 

32 in paragraph (e)(1) is not “technically achievable, would be prohibitively expensive, or would 

33 result in net public or environmental harm.” Further, 10 CFR 20.1403(e)(2) is a regulatory tool 

34 to limit the use of the 500 mrem (5 mSv) per year cap, similar to the way that the ALARA 

35 requirement in 10 CFR 20.1403(a) limits the use of restricted release.

36

37 N.2.3 ALARA in Alternate Release Criteria

38 Under 10 CFR 20.1404, the Commission may terminate a license using alternate criteria greater 

39 than the dose criterion of 10 CFR 20.1402, 20.1403(b), and 20.1403(d)(1)(i)(A), if the licensee 

40 meets certain conditions. That is, the NRC might authorize release of a site with a TEDE to the 

41 average member of the critical group from residual radioactivity distinguishable from 

42 background in excess of 25 mrem (0.25 mSv) per year under the conditions specified in 

43 10 CFR 20.1404. The general ALARA requirement of 10 CFR 20.1101(b) continues to apply 

44 even if an alternate dose criterion is authorized under 10 CFR 20.1404. Additionally, the criteria 

45 for approval of alternate release criteria includes specific ALARA requirements: the licensee

46 must employ, to the extent practical, restrictions on site use according to the provisions of 

47 § 20.1403 in minimizing exposures at the site; and must reduce doses to ALARA levels, taking

48 into consideration any detriments such as traffic accidents expected to potentially result from

49 decontamination and waste disposal.
N-6

1

2 Based on these criteria, the NRC expects the licensee to consider requesting alternate release 

3 criteria only after first considering unrestricted and restricted release, and after implementing the 

4 provisions of 10 CFR 20.1403 on restrictions of site use to the extent practical to minimize dose. 

5

6 N.2.4 ALARA Measures and Radon

7 One area where the implementation of ALARA principles notably differs between unrestricted, 

8 restricted, and restricted release/alternate release criteria is that of radon mitigation: 

9

10 Because of [the] variations and the limitation[s] of measurement techniques, the 

11 Commission believes that it is not practical for licensees to distinguish between radon from

12 licensed activities at a dose comparable to a 0.25 mSv/y (25 mrem/y) dose criterion and 

13 radon which occurs naturally. Therefore, in implementing the final rule, licensees will not be 

14 expected to demonstrate that radon from licensed activities is indistinguishable from 

15 background on a site-specific basis. Instead this may be considered to have been 

16 demonstrated on a generic basis when radium, the principal precursor to radon, meets the 

17 requirements for unrestricted release, without including doses from the radon pathway.

18 (62 FR 39083)

19

20 Therefore, for unrestricted release, the NRC does not require licensees to include doses from 

21 the radon pathway as long as residual radium contamination is within the unrestricted use 

22 criteria. However, for restricted release:

23

24 ...it may not be reasonable to achieve levels of residual concentrations of radon precursors 

25 within the limit for unrestricted use...[F]or cases such as these, restricting site use by use of 

26 institutional controls could be considered by a licensee as a means to limit the doses from 

27 precursors by limiting access to the site. Under the restricted use provisions of the rule, 

28 these doses are required to be further reduced based on ALARA principles. (62 FR 39083)

29

30 For these reasons, licensees may need to consider institutional controls to reduce doses from 

31 radon, and its precursors, to comply with the ALARA principles of 10 CFR 20.1403. ALARA

32 considerations should address the practicality of radon mitigation techniques in structures as 

33 part of institutional controls. ALARA considerations for compliance with the general ALARA 

34 principle as applied to 10 CFR 20.1403(e) for controls not in place do not have to address the 

35 radon pathway.

36

37 N.3 Evaluation of Cost-Benefit ALARA Analyses

38 As discussed in Section N.1, compliance with the license termination criteria must include a 

39 demonstration that the dose to the average member of the critical group and the amount of 

40 residual radioactivity remaining on the site at license termination is ALARA. A simplified method 

41 for demonstrating compliance with the ALARA requirement is described below. Licensees may 

42 use more complex or site-specific analyses if more appropriate for their specific situation. In 

43 general, more complex analyses should follow the general concepts presented here. Evaluation 

44 of more complex analyses should be handled on a case-by-case basis and early involvement of 

45 the appropriate regulatory agencies and members of the public is suggested.

46

47 Sometimes it is very difficult or impossible to place a monetary value on an impact. A best effort 

48 should be made to assign a monetary value to the impact, because there may be no other way 

49 to compare benefits to costs. However, there may be situations for which a credible monetary 
N-7

1 value cannot be developed. In these situations, a qualitative treatment may be the most 

2 appropriate. Qualitative analyses should be evaluated on their merits on a case-by-case basis. 

3

4 N.3.1 Simplified Method for Cost-Benefit ALARA Analyses

5 The simplified method presented here is to estimate when a proposed remediation action is 

6 ALARA using generalized estimates for the remedial action. Evaluating whether a proposed 

7 action is ALARA involves considering potential further, alternative remedial actions that would 

8 result in further reduction in the dose to the average member of the critical group or further 

9 reduction in the amount of residual radioactivity that would remain on site at license termination, 

10 relative to the proposed action. The evaluation then involves comparing the proposed action 

11 and the potential alternative remedial actions in terms of desired beneficial effects (benefits) and 

12 undesirable effects (costs). If the proposed action provides the greatest benefits: cost ratio 

13 when compared to a further alternative action, then that is demonstration that the proposed 

14 action is ALARA in comparison to the alternative action and the alternative action is 

15 unnecessary. Conversely, if the alternative action provides the greatest benefits: cost ratio, it 

16 would be ALARA to select the alternative option. 

17

18 The various ALARA evaluations for license termination may all use this same simplified method, 

19 but some of the evaluations should differ in the nature of the baseline and alternative 

20 approaches being compared. The NRC staff should consider, in particular, whether the 

21 approaches being compared appropriately address the reduction of dose to the average 

22 member of the critical group or reduction of residual radioactivity remaining on site at license 

23 termination. For requests to use alternative release criteria under 10 CFR 20.1404, the NRC 

24 staff should evaluate the ALARA demonstrations on a case-by-case basis.

25

26 To compare the benefits and costs of a remediation action, it is necessary to use a comparable 

27 unit of measure. The unit of measure used here is the dollar; if possible, all benefits and costs 

28 are given a monetary value. The licensee can calculate benefits and costs as described in 

29 Sections N.2 through N.6. 

30

31 The licensee should apply the method during remediation planning, before the start of 

32 remediation but after completing some or all of the characterization work. The licensee should 

33 only use the method to determine whether and where it should take particular remediation 

34 actions to meet the ALARA requirement.

35

36 If the licensee has already decided to perform a remediation action, there is no need to analyze 

37 whether the action was necessary to meet the ALARA requirement. The analysis described in 

38 this section is needed only to justify not taking a further remediation action. For example, if a 

39 licensee plans to wash room surfaces (either to meet the dose limit or as a good practice), there 

40 is no need to analyze whether the remediation action of washing is necessary to meet the 

41 ALARA requirement.

42

43 Table N.1 gives an example of various benefits and costs. Other than Collective Dose Averted, 

44 the additional benefits listed are generally only important in comparisons between alternatives 

45 that address whether the licensee can pursue restricted release. The value of any benefit or 

46 cost can be negative in some cases.

47
N-8

1 Table N.1 Possible Benefits and Costs Related to Decommissioning 

Possible Benefits Possible Costs

Collective Dose Averted

Regulatory Costs Avoided

Changes in Land Values

Esthetics

Reduction in Public Opposition

Remediation Costs (including waste disposal costs)

Additional Occupational/Public Dose

Occupational Non-radiological Risks

Transportation Direct Costs and Implied Risks

Environmental Impacts

Loss of Economic Use of Site/Facility

2

3

4 N.3.2 Calculation of Benefits

5 N.3.2.1 Collective Dose Averted

6 In the simplest form of the analysis, the only benefit estimated from a reduction in the level of 

7 residual radioactivity is the monetary value of the collective averted dose to future occupants of 

8 the site. The licensee should base the collective averted dose on the same exposure 

9 scenario(s) used for the compliance calculations. Section N.2.7 discusses additional 

10 considerations related to groundwater residual radioactivity.

11

12 To calculate the benefit from collective averted dose, BAD, determine the present worth of the 

13 future collective averted dose and multiply it by a factor to convert the dose to monetary value, 

14 as shown in Equation N-1 below. 

15

16 NUREG/BR-0058, “Regulatory Analysis Guidelines of the U.S. Nuclear Regulatory 

17 Commission,” Revision 4, issued September 2004 describes present-worth calculations using 

18 both 3-percent and 7-percent real discount rates. Special considerations arise when comparing 

19 benefits and costs across generations, as is often the case with decommissioning. Thus, ethical 

20 and technical arguments can also support the use of lower discount rates. Therefore, if 

21 licensees anticipate important intergenerational consequences, such as for cases with 

22 radionuclides with half-lives of decades or longer, licensees should consider supplementing the 

23 analysis with an explicit discussion of the intergenerational concerns, such as how future 

24 generations will be affected by the regulatory decisions. Supplemental information could 

25 include a presentation of the values and impacts at the time they are incurred, with no present 

26 worth conversion. Also, licensees should consider a sensitivity analysis using a lower, but 

27 positive, discount rate. As a general principle, licensees should perform a sensitivity or 

28 uncertainty analysis, or both, whenever the values of key attributes can range widely. 

29 NUREG/BR-0058 contains additional discussion on the consideration of intergenerational 

30 consequences and sensitivity or uncertainty analyses.

31

32

BAD = VAD × PW(ADCollective 33 ) (N-1)

34

35 where BAD = benefit from an averted dose for a remediation action, in current 

36 U.S. dollars

37 VAD = value of averted dose, which is a conversion factor for the 

38 monetary value of radiation dose (dollars ($) per person-rem, see 

39 NUREG/BR-0058). This conversion factor is applied to averted 
N-9

1 dose (as in this case) and to dose incurred (as in the case of dose 

2 to workers or the public discussed in Section N.2.3).

3 PW(ADcollective) = present worth of a future collective averted dose in person-rem

4

5 The value of averted dose is periodically reviewed based on changes to the underlying 

6 assumptions. It is important to verify the current value from NRC regulatory guidance when 

7 performing a collective dose averted analysis. In addition to NUREG/BR–0058, licensees 

8 should consult NUREG-1530, “Reassessment of NRC’s Dollar per Person-Rem Conversion 

9 Factor Policy”. 

10

11 The present worth of the future collective averted dose can be estimated from Equation N–2, for 

12 relatively simple situations:

13

PW(ADCollective) = PD × A × 0.025 × F ×

Conc

DCGLw

×

1 − e

−(r + λ)N

r + λ

14 (N-2)

15

16

where PD = population density for the critical group scenario in people/m2 17 ;

A = area being evaluated in square meters (m2 18 );

19 0.025 = annual dose to an average member of the critical group from residual 

20 radioactivity at the derived concentration guideline level (DCGLW) 

21 concentration in rem/y;

22 F = effectiveness, or fraction of the residual radioactivity removed by the 

23 remediation action;

24 Conc = average concentration of residual radioactivity in the area being evaluated 

25 in units of activity per unit area for buildings or activity per unit volume for 

26 soil;

27 DCGLW= derived concentration guideline equivalent to the average concentration 

28 of residual radioactivity that would give a dose of 0.25 mSv/y (25 mrem/y) 

29 to the average member of the critical group, in the same units as “Conc”;

30 r = monetary discount rate in units per year;

31 λ = radiological decay constant for the radionuclide in units per year; and

32 N = number of years over which the collective dose will be calculated.

33

34 NOTE: When the discount rate, r, is zero and the radiological decay rate, λ, is very small so 

35 that r + λ → 0, this equation must be adjusted (see Section N.6).

36

37 The present worth of the benefit calculated by Equation N–2, above, assumes that the peak 

38 dose occurs in the first year. This is almost always true for the building occupancy scenario but 

39 not always true for the residential scenario, where the peak dose can occur in later years. In 

40 that case, Equation N–2 would overestimate the benefit. The licensee may perform a more 

41 exact calculation that avoids this overestimation of the benefit of remediation by calculating the 

42 dose during each year of the evaluation period and then calculating the present worth of each 

43 year’s dose. Section N.6 contains a detailed derivation of Equation N-2 and some of the other 

44 equations.

45

46 The DCGLW (derived concentration guideline level, average concentrations over a wide area) is 

47 based on compliance with the 0.25 mSv/y (25 mrem/y) dose limit. Base the population density, 

48 PD, on the dose scenario used to demonstrate compliance with the dose limit. Thus, for 

49 buildings, the licensee should estimate PD for the building occupancy scenario. For soil, it 
N-10

1 should base the PD on the residential scenario. The factor at the far right of the equation, which 

2 includes the exponential terms, accounts for both the present worth of the monetary value and 

3 radiological decay. 

4

5 If more than one radionuclide is present, the total benefit from a collective averted dose, BAD, is 

6 the sum of the collective averted dose for each radionuclide. When multiple radionuclides have 

7 a fixed concentration (i.e., secular equilibrium), residual radioactivity below the dose criteria is 

8 normally demonstrated by measuring one radionuclide and comparing its concentration to a 

9 DCGLW that has been calculated to account for the dose from the other radionuclides. In this 

10 case, the licensee may use the adjusted DCGLW with the concentration of the radionuclide 

11 being measured. The other case is where the ratio of the radionuclide concentrations is not 

12 fixed and varies from location to location within a survey unit; this benefit is the sum of the 

13 collective averted dose from each.

14

15 N.3.2.2 Regulatory Costs Avoided

16 This benefit usually occurs in ALARA analyses of restricted release versus unrestricted release 

17 decommissioning goals. By releasing the site with no restrictions, the licensee may avoid the 

18 various costs associated with restricted release. These costs can include 

19

20 • additional licensing fees for safety reviews and for developing an Environmental Impact

21 Statement,

22 • financial assurance for necessary control and maintenance of a site

23 (10 CFR 20.1403(c)),

24 • costs (including NRC-related) associated with public meetings or the community review

25 committee (10 CFR 20.1403(d)(2)), and

26 • future liability.

27

28 When evaluating the eligibility of a licensee’s proposal for restricted release according to 

29 10 CFR 20.1403(a), the NRC staff recommends that the regulatory costs avoided be included in 

30 the benefits of the unrestricted release decommissioning alternative, rather than included as 

31 costs related to the restricted release (see Section N.2.2).

32

33 N.3.2.3 Changes in Land Values

34 The licensee should account for any expected change in the value of the site or facility or 

35 surrounding land caused by the different decommissioning options. This may be difficult to 

36 quantify.

37

38 N.3.2.4 Esthetics/Reduction in Public Opposition

39 These can be very difficult to quantify. The licensee may wish to evaluate the effect of the 

40 available decommissioning options with respect to the overall esthetics (including the 

41 decommissioning activities themselves) of the site and surrounding area. Another factor the 

42 licensee may wish to consider is the potential reduction in opposition, if there is any, to the 

43 decommissioning activities or goal it is attempting to propose.

44
N-11

1 N.3.3 Calculation of Costs

2 The licensee should evaluate the costs of the selected alternative remediation actions being 

3 evaluated. When performing a fairly simple evaluation, the costs generally include the monetary 

4 costs of 

5

6 • the remediation action being evaluated,

7 • transportation and disposal of the waste generated by the action,

8 • workplace accidents that occur because of the remediation action,

9 • traffic fatalities resulting from transporting the waste generated by the action,

10 • doses received by workers performing the remediation action, and

11 • doses to the public from excavation, transport, and disposal of the waste. The licensee

12 may also include other costs that are appropriate for the specific case.

13

14 The total cost, CostT, which is balanced against the benefits, has several components.

15

CostT = CostR + CostWD + CostACC + CostTF + CostWDose + CostPDose + Costother 16 (N-3)

17

18 where CostR = monetary cost of the remediation action (may include “mobilization” 

19 costs);

20 CostWD = monetary cost for transport and disposal of the waste generated by the 

21 action;

22 CostACC = monetary cost of worker accidents during the remediation action;

23 CostTF = monetary cost of traffic fatalities during transporting of the waste;

24 CostWDose = monetary cost of dose received by workers performing the remediation 

25 action and transporting waste to the disposal facility;

26 CostPDose = monetary cost of the dose to the public from excavation, transport, 

27 and disposal of the waste; and

28 Costother = other costs as appropriate for the particular situation.

29

30 The licensee does not necessarily have to calculate all the costs described below. For 

31 example, if one or two of the costs can be shown to be in excess of the benefit and if none 

32 of the other costs are negative, the remediation action will be unnecessary. However, some 

33 of these costs may, in fact, be negative (i.e., the alternative may cost less than the preferred 

34 option) in some comparisons between alternative decommissioning options, and thus, the 

35 licensee may need to evaluate all costs.

36

37 N.3.3.1 Remedial Action Costs

38 Calculation of the incremental remedial action costs includes the standard manpower and 

39 mechanical costs. The licensee can account for any additional licensing fees from the NRC 

40 (e.g., if the option to meet the ALARA goal requires another year of remediation). Lower

41 concentrations may change sampling or survey requirements. The remedial action can 

42 consider increased survey costs but note that this is the incremental cost of surveying below the 
N-12

1 dose limit. Survey costs related to evaluating compliance at the dose limit are not part of the 

2 ALARA analysis.

3

4 N.3.3.2 Transport and Disposal of the Waste

5 The cost of waste transport and disposal, CostWD, may be evaluated according to Equation N-4.

6

7 CostWD = VA × CostV (N-4)

8

where: VA = volume of waste produced, remediated in units of m3 9 ; and

10 CostV = cost of waste disposal per unit volume, including transportation cost, in 

units of $/m3 11 .

12

13 N.3.3.3 Non-radiological Risks

14 The cost of non-radiological workplace accidents, CostACC, may be evaluated using 

15 Equation N-5.

16

17 CostACC = VSL × FW × TA (N-5)

18

19 where: VSL = monetary value of a statistical life (or fatality) (see NUREG-1530, 

20 “Reassessment of NRC’s Dollar per Person-Rem Conversion 

21 Factor Policy”)—this value is subject to periodic revision, so it is 

22 important to verify the current value from NUREG-1530 when 

23 performing an analysis on non-radiological workplace accidents;

24 FW = workplace fatality rate in fatalities/hour worked; and

25 TA = worker time required for remediation in units of worker-hours.

26

27 N.3.3.4 Transportation Risks

28 Calculate the cost of traffic fatalities incurred during the transportation of waste, CostTF, as in 

29 Equation N-6.

30

CostTF = VSL × (

VA

VSHIP

31 ) × FT × DT (N-6)

32

33 where: VSL = monetary value of a statistical life (or fatality) (see NUREG-1530, 

34 “Reassessment of NRC’s Dollar per Person-Rem Conversion Factor 

35 Policy”)—this value is subject to periodic revision, so it is important to verify 

36 the current value from NUREG-1530 when performing an analysis on non37 radiological workplace accidents;

VA = volume of waste produced in units of m3 38 ;

39 FT = fatality rate per truck-kilometer traveled in units of fatalities/truck-km;

40 DT = distance traveled in km; and

VSHIP = volume of a truck shipment in m3 41 .

42

43

44 The actual parameters should depend on the site’s planned method of waste transport. Some 

45 facilities may consider a mix of trucking and rail transport to take the waste to the disposal site. 
N-13

1 In these cases, the cost would be equivalent to the total fatalities likely from the rail transport

2 and the limited trucking, not just the trucking alone.

3

4 N.3.3.5 Worker Dose Estimates

5 Calculate the cost of the remediation worker dose, CostWDose, as shown in Equation N-7.

6

7 CostWDose = VAD × DR × T (N-7)

8

9 where: VAD = value of incurred dose, which is a conversion factor for the monetary 

10 value of radiation dose (dollars ($) per person-rem, see 

11 NUREG/BR-0058);

12 DR = TEDE rate to remediation workers in units of rems/hr; and

13 T = time worked (site labor) to remediate the area in units of person-hour.

14

15 The cost of worker dose usually should not be discounted, because the dose is all incurred 

16 close to the time of license termination.

17

18 N.3.3.6 Loss of Economic Use of Property

19 A cost in the “other” category could include the fair market rental value or economic use for the 

20 site during the time the additional remediation work is being performed. These costs are usually 

21 associated with locations such as laboratories, hospital rooms, and industrial sites. This cost 

22 may be added to the costs in Equation N-3.

23

24 N.3.3.7 Environmental Impacts

25 Another cost that could fall into the other category would be a remediation action that could 

26 damage an ecologically valuable area or cause some other adverse environmental impact. 

27 Include these impacts as costs of the remediation action.
N-14

1 N.3.3.8 Default Parameters

2 Table N.2 Acceptable Parameter Values for Use in ALARA Analyses (Shows Acceptable

3 Values for some of the Parameters used in Performing These Calculations)

Parameter Value Reference and Comments

Workplace accident 

fatality rate, FW

4.2 x 10-8/hr NUREG-1496, “Final Generic 

Environmental Impact Statement in 

Support of Rulemaking on Radiological 

Criteria for License Termination of NRCLicensed Nuclear Facilities,” and 

NUREG-1496, Volume 2, Appendix B, 

Table A.1, July 1997

Transportation fatal 

accident rate, FT

Trucks: 3.8 x 10-8/km NUREG-1496, Volume 2, Appendix B, 

Table A.1

Value of averted 

dose, VAD, and 

value of statistical 

life, VSL

values are updated 

periodically

NUREG/BR-0058, NUREG-1530. It is 

important to verify the current values (see 

Section N.2.2)

Monetary discount 

rate, r

0.03/y and 0.07/y 

discount rates, with 

special considerations 

for intergenerational 

consequences (as 

discussed in 

Section N.2.2)

NUREG/BR-0058

Number of years of 

exposure, N

Buildings: 70 years

Soil: 1,000 years

NUREG-1496, Volume 2, Appendix B, 

Table A.1

Population density, 

PD

Building: 0.09 person/m2

Land: 0.0004 person/m2

NUREG-1496, Volume 2, Appendix B, 

Table A.1

Excavation, 

monitoring, 

packaging, and 

handling soil

Soil: 1.62 personhours/m3 of soilNUREG-1496, Volume 2, Appendix B, 

Table A.1

Waste shipment 

volume, VSHIP

Truck: 13.6 m3/shipment NUREG-1496, Volume 2, Appendix B, 

Table A.1

4

5 N.3.4 Residual Radioactivity Levels that are ALARA for Unrestricted Use

6 The residual radioactivity level that is ALARA for license termination for unrestricted use is the 

7 concentration, ConcALARA at which the benefit from removal equals the cost of removal. If the 

8 total cost, CostT, is set equal to the present worth of the collective dose averted in Equation N-2, 

9 the ratio of the concentration, ConcALARA, to the DCGLW can be determined (derivation shown in 

10 Section N.7).
N-15

1

ConcALARA

DCGLw

=

CostT

VAD × PD × 0.025 × F × A

×

r + λ

1 − e−(r + λ)N

2 (N-8)

3

4 All the terms in Equation N-8 are as defined previously.

5

6 Since PD, N, and r are constants that have generic values for all locations on the site, the 

7 licensee only needs to determine the total cost, CostT, and the effectiveness, F, for a specific 

8 remediation action. If the concentration at a location exceeds ConcALARA, it may be cost 

9 effective to remediate the location by a method where the total cost is CostT. Note that the 

10 concentration, ConcALARA, that is ALARA can be higher or lower (more or less stringent) than the 

11 DCGLW, although licensees should meet the DCGLW. 

12

13 For cases of license termination for restricted use, the NRC staff and licensees should modify 

14 the equations derived in Section N.6 to include the additional benefits that are only applicable to 

15 comparisons of unrestricted and restricted use (see Section N.1). 

16

17 N.3.5 Examples of Calculations

18 Example 1: Washing Building Surfaces

19 This example considers a building with cesium-137 residual radioactivity (λ = 0.023/y). The 

remediation action to be considered is washing a floor of 100 m2 20 area. The licensee estimates 

21 that this may cost $400 and may remove 20 percent (F = 0.2) of the residual radioactivity. The 

22 most recent value of VAD, at the time of this revision, is $5100 per person-rem. For buildings, 

generic parameters are: PD = 0.09 person/m2 23 , r = 0.07/y, and N = 70 years. Using these values 

24 in Equation N-8:

25

ConcALARA

DCGLw

=

$400

$5100 × 0.2 × 0.025 × 0.09 × 100

×

0.07 + 0.023

1 − e−(0.07 + 0.023)70 26 (N-9)

27

ConcALARA

DCGLw

28 = 0.16 (N-10)

29

30 To meet the ALARA requirement, the floor should be washed if the average concentration 

31 exceeds about 16 percent of the DCGLW. This is more stringent than the dose limit. This 

32 calculation shows that washing building surfaces is often necessary to meet the ALARA 

33 requirement. If the surfaces are planned to be washed, there is no need for the licensee to 

34 perform the ALARA evaluation or to submit the evaluation to the NRC. If it is decided not to 

35 wash the building surfaces, the licensee could submit this evaluation and demonstrate in the 

36 FSS that all surfaces have a concentration below 16 percent of the DCGLW.

37

38 Example 2: Scabbling Concrete in a Building

39 This example is the same as above, except that it evaluates the use of a scabbling tool that 

40 removes the top one-eighth of an inch of concrete. The licensee estimates that the total cost of 

the scabbling may be $5,000 for the 100 m2 41 floor and that it may remove all the residual 

42 radioactivity so that F = 1. Using these values in Equation N-8 gives:

43

ConcALARA

DCGLw

=

$5000

$5100 × 1 × 0.025 × 0.09 × 100

×

0.07 + 0.023

1 − e−(0.07 + 0.023)70 44 (N-11)
N-16

1

ConcALARA

DCGLw

2 = 0.41 (N-12)

3

4 The licensee could decide to scabble depending on the concentrations present. In lieu of 

5 scabbling, the licensee could provide this analysis and demonstrate that the floor concentration 

6 is less than 0.41 DCGLW.

7

8 Example 3: Removing Surface Soil

In this example, soil with an area of 1,000 m2 9 is found to contain cobalt-60 (λ = 0.1315/y) 

10 residual radioactivity to a depth of 15 centimeters (cm) (6 inches (in.)). The licensee estimates 

11 that the cost of removing the soil (F = 1) may be $100,000. For soil, the generic parameters are 

PD = 0.0004 person/m2 12 , r = 0.03/y, and N = 1000 y. The most recent value of VAD, at the time of 

13 this review, is $5100 per person-rem. Using these values in Equation N-8 gives:

14

ConcALARA

DCGLw

=

$100,000

$5100 × 1 × 0.025 × 0.0004 × 1000

×

0.03 + 0.1315

1 − e−(0.03 + 0.1315)1000 15 (N-13)

16

ConcALARA

DCGLw

17 = 317 (N-14)

18

19 Thus, meeting the dose limit would be limiting by a considerable margin. Based on these 

20 results, it would rarely be necessary to ship soil to a waste disposal facility to meet the ALARA 

21 requirement. The licensee could use this evaluation to justify not removing soil. 

22

23 The advantage of the approach shown in these examples is that it allows the licensee to 

24 estimate a concentration at which a remediation action may be cost effective before starting 

25 remediation and before planning the FSS. Thus, it is a useful planning tool that lets the licensee 

26 determine which remediation actions may be needed to meet the ALARA requirement.

27

28 N.3.6 When Mathematical Analyses are not Necessary

29 In certain circumstances, the results of an ALARA analysis are known on a generic basis and an 

30 analysis is not necessary. For residual radioactivity in soil at sites that may have unrestricted 

31 release, generic analyses (see NUREG-1496 and the examples in Section N.2.5) show that 

32 shipping soil to a licensed low-level waste disposal facility is unlikely to be cost effective for 

33 unrestricted release, largely because of the high costs of waste disposal. An ALARA analysis is 

34 not needed for soil removal to meet unrestricted release at or below a dose criterion of 

0.25 mSv (25 mrem) per year.

4 35 Therefore, the licensee generally does not have to evaluate 

36 shipping soil to a low-level waste disposal facility to achieve exposure levels at or below the 

37 criterion for unrestricted release. However, this conclusion may not hold for waste disposal at 

38 other than licensed low-level waste disposal facilities. 

39

40 In light of the conservatism in the building surface and surface soil default screening levels 

41 developed by the NRC staff (see Appendix H of this volume), the reviewer presumes, absent 

42 information to the contrary, that licensees who remediate building surfaces or surface soil to the 

43 NRC default screening levels do not need to provide analyses to demonstrate that these 

44 screening levels are ALARA. 

4 See preamble to the license termination rule found at 62 FR 39058; July 21, 1997.
N-17

1

2 In addition, the NRC staff concludes that licensees would not be required to perform a 

3 quantitative cost-benefit ALARA analysis for cases where no residual radioactivity 

4 distinguishable from background remains or will remain on the site at termination. 

5

6 Removal of loose residual radioactivity from building surfaces is almost always cost effective, 

7 except when very small quantities of radioactivity are involved. Therefore, the NRC staff 

8 concludes that loose residual radioactivity normally should be removed from building surfaces 

9 and if it is removed, the cost-benefit analysis would not be needed.

10

11 N.3.7 Additional Considerations for Residual Radioactivity in Groundwater

12 The method described above is adequate for most situations and has minimal cost for analyses. 

13 However, if the site has residual radioactivity from site operations in groundwater, the licensee 

14 should include other factors, as described below, if it intends to release the site. 

15

16 If there is residual radioactivity from site operations in groundwater, it may be necessary to 

17 calculate the collective dose from consumption of the groundwater. Default or generic 

18 groundwater models typically assume that potable aquifers have small volumes and cannot 

19 supply large populations. When this is the case, dose calculations for the site critical group may 

20 adequately represent the collective dose from groundwater. However, when site-specific 

21 groundwater modeling is used, and the residual radioactivity is diluted in an aquifer of large 

22 volume and there is also an “existing population deriving its drinking water from a downstream 

23 supply using a downstream plume” (62 FR 39058; July 21, 1997), the collective dose for that 

24 population should be included in the ALARA calculation. The possibility of reducing the 

25 collective dose by remediation should be one of the items evaluated as one of the benefits, 

26 even if remediation would not affect the critical group’s doses significantly. Another 

27 consideration for groundwater residual radioactivity would be the reduction of any potential 

28 costs incurred by other entities, such as a public water supply utility, to meet the requirements of 

29 the Safe Water Drinking Act, if the licensee’s residual radioactivity levels would potentially lead 

30 to concentrations at the wellhead that would exceed the U.S. Environmental Protection 

31 Agency’s Maximum Contaminant Levels.

32

33 N.4 Determination of “Net Public or Environmental Harm” 

34 Subpart E, 10 CFR 20.1403(a) and 10 CFR 20.1403(e)(2)(i) address circumstances in which a 

35 licensee may demonstrate that further remediation would cause net public or environmental

36 harm. The calculation to demonstrate net public or environmental harm is a special case of the 

37 general ALARA calculation described above that compares the benefits in dose reduction to the 

38 cost of doses, injuries, and fatalities incurred or to the cost of environmental degradation. The 

39 calculation does not consider the monetary costs for performing further remediation, CostR, or 

40 the costs of waste disposal, CostWD. Thus, if the benefit from averted dose BAD is less than the 

41 sum of the costs of workplace accidents, CostACC, the costs of transportation fatalities, CostTF, 

42 the costs of remediation worker dose, CostWDose, and the costs of public dose, CostPDose, or less 

43 than the costs of any environmental degradation, CostED, then there is net public or 

44 environmental harm. Thus, there is net public or environmental harm if:

45

Net harm if: BAD < CostACC + CostTF + CostWDose + CostPDose 46 (N-15)

47

48 or
N-18

1

2 Net harm if: BAD < CostED (N-16)

3

4 In some cases, it may be very difficult to assign a credible monetary value to environmental 

5 degradation. For example, environmental harm could be caused by an action such as 

6 remediation of a wetlands area. There may be no way to assign a monetary value to this action. 

7 In these cases, it is acceptable to use qualitative arguments, which should be evaluated on a 

8 case-by-case basis.

9

10 N.5 Demonstration of “Not Technically Achievable”

11

12 Subpart E, 10 CFR 20.1403(e)(2)(i), addresses circumstances in which a licensee would be 

13 required to demonstrate that further reductions in residual radioactivity are not technically 

14 achievable. Remediation of residual radioactivity is almost always technically achievable, even 

15 if not economically feasible. This provision allows for special cases that may not be 

16 foreseeable; thus, the NRC has no specific guidance on this provision. Instead, the NRC staff 

17 will evaluate licensee submittals on a case-by-case basis. 

18

19 N.6 Demonstration of “Prohibitively Expensive” 

20 Subpart E, 10 CFR 20.1403(e)(2), addresses circumstances in which a licensee would be 

21 required to demonstrate that further reductions in residual radioactivity would be prohibitively 

22 expensive. The licensee can demonstrate this by an analysis like the ALARA analysis 

23 described above but using an increased value of averted dose, VAD. As discussed in Section 

24 N.2.2, NUREG-1530 and NUREG/BR-0058 should be consulted to determine the recommended

25 value for averted dose. For a “prohibitively expense” assessment, this value should be

26 multiplied times 10 prior to being used as VAD in the analysis. This increased value of averted

27 dose reflects the statement in the final rule on radiological criteria for license termination that the

28 NRC considers it appropriate that a remediation would be prohibitively expensive if the cost to

29 avert dose were an order of magnitude more expensive than the cost recommended by the

30 NRC for an ALARA analysis (see 62 FR 39058, p. 39071, July 21, 1997). However, the NRC

31 also stated that “...a lower factor may be appropriate in specific situations when the licensee

32 could become financially incapable of carrying out decommissioning safely.” Thus, values lower

33 than 10 times the value recommended in NUREG-1530 and NUREG/BR-0058 for VAD may be

34 used when remediation actions could otherwise cause the licensee to become financially

35 incapable of safely carrying out decommissioning.

36

37 N.7 Derivation of Main Equations To Calculate ALARA Concentrations for 

38 Unrestricted Use 

39 The following derivation applies to an ALARA evaluation for license termination for unrestricted 

40 use. Additional benefits would apply to restricted use cases (see Sections N.2.1 and N.2.2), 

41 and these equations would be modified. 

42

43 The ALARA analysis compares the monetary value of the desirable effects (benefits) of a

44 remediation action (e.g., the monetary benefit of collective averted dose) with the monetary 

45 value of the undesirable effects (e.g., the costs of waste disposal). If the benefits of a 

46 remediation action would exceed the costs, the licensee should take remediation action to meet 

47 the ALARA requirement.
N-19

1

2 The primary benefit from a remediation action is the collective dose averted in the future 

3 (i.e., the sum over time of the annual doses received by the exposed population). Assume:

4

5 If benefits > costs, the remediation action should be taken (N-17)

6

7 (1) A site proposes remediation and termination for unrestricted use (thus, the benefit to be

8 considered is dose averted only).

9 (2) The site has an area with residual radioactivity at concentration, Conc.

10 (3) The concentration equivalent to 0.25 mSv/y (25 mrem/y) (DCGLW) for the site has been

11 determined (for soil or for building surfaces, as appropriate).

12 (4) The residual radioactivity at a site has been adequately characterized so that the

13 effectiveness of a remediation action can be estimated in terms of the fraction F of the

14 residual radioactivity that the action may remove.

15 (5) The peak dose rate occurs at time 0 and decreases thereafter by radiological decay.

16 The derived concentration guideline (DCGLW) is the concentration of residual radioactivity that 

17 would result in a TED E to an average member of the critical group of 0.25 mSv/y (25 mrem/y). 

18 Therefore, the annual dose D to the average member of the critical group from residual 

19 radioactivity at concentration Conc is:

20

D = 0.025 (

rem

yr ) ×

Conc

DCGLW

21 (N-18)

22

23 If a remediation action would remove a fraction, F, of the residual radioactivity present, the 

24 annual averted dose to the individual, ADindividual, is

25

ADindividual (

rem

yr−person

) = F × 0.025 (

rem

yr ) ×

Conc

DCGLw

26 (N-19)

27

28 The annual collective averted dose, ADcollective, can be calculated by multiplying the individual 

29 averted dose, ADindividual, by the number of people expected to occupy the area, A, containing 

30 the residual radioactivity. The number of people in the area containing the residual radioactivity 

31 is the area, A, times the population density, PD, for the site. 

32

33 Thus:

34

ADcollective = F × 0.025 (

rem

yr ) ×

Conc

DCGLw

35 × A × PD (N-20)

36

37 The annual monetary benefit rate at time 0, B0, from the averted collective dose in dollars per 

38 year can be calculated by multiplying the annual collective averted dose, ADcollective, by the value 

39 of averted dose, VAD:

40

B0 = VAD × F × 0.025 (

rem

yr ) ×

Conc

DCGLw

41 × A × PD (N-21)

42
N-20

1 The total monetary benefit of averted doses can be calculated by integrating the annual benefit 

2 over the exposure time in years, considering both the present worth of future benefits and 

3 radiological decay. It is OMB and NRC policy to use the present worth of both benefits and 

4 costs that occur in the future.

5

6 The equation for the present worth, PWB, of a series of constant future annual benefits, B (in 

7 dollars per year), for N years at a monetary discount rate of r (per year) using continuous 

8 compounding is:

9

PWB = B ×

e

rN − 1

re

rN 10 (N-22)

11

12 The continuous compounding form of the present worth equation is used, because it permits an 

13 easy formulation that includes radiological decay. If the annual benefit rate, B, is not constant 

14 but is decreasing from an original rate, B0, because of radiological decay, the radiological decay 

15 rate acts like an additional discount rate that can be added to the monetary discount rate of 

16 decrease, so that the present worth of the annual benefits PWB becomes:

17

PWB = BO ×

e

(r + λ)N − 1

(r + λ)e

(r + λ)N

18 (N-23)

19

Dividing the numerator and denominator of the right-hand term by e

(r + λ)N 20 yields:

21

PWB = BO ×

1 − e

−(r + λ)N

r + λ

22 (N-24)

23

24 As N→∞, Equation N-24 has the limit:

25

PWB = BO ×

1

r + λ

26 (N-25)

27

28 When the discount rate, r, is zero and the radiological decay rate is very small, so that r + λ → 

29 0, and Equation N-24 has the limit:

30

31 PWB = BO × N (N-26)

32

33 The total benefit from the collective averted dose, Btotal, is the present worth of the annual 

34 benefits. Btotal can be calculated by combining Equations N-21 and N-24:

35

Btotal = VAD × F × 0.025 ×

Conc

DCGLw

× A × PD ×

1 – e

−(r + λ)N

r + λ

36 (N-27)

37

38 Now consider the total cost of a remediation action, CostT. The costs included in CostT are 

39 (1) the direct cost of the remediation action itself, CostR, (2) the cost of waste disposal including

40 its shipping cost, CostWD, (3) the monetary costs of workplace accidents during the remediation,

41 CostACC, (4) the monetary costs of transportation accidents during the shipping of waste, CostTF,

42 (5) the monetary value of the dose that remediation workers receive, CostWDose, (6) the

43 monetary value of the dose to the public from the excavation, transport, and disposal of the

44 waste, CostPDose, and (7) other costs, as appropriate, for the specific site, Costother. Thus,

45

CostT = CostR + CostWD + CostACC + CostTF + CostWDose + CostPDose + CostOther 46 (N-28)

47
N-21

1 What is of interest in this derivation is the concentration, ConcALARA, at which the benefit, Btotal,

2 equals the total cost, CostT. Thus, in Equation N-27, CostT can be substituted for Btotal, and then 

3 Equation N-27 can be solved for the concentration, ConcALARA, relative to the DCGLW, as in 

4 Equation N-29.

5

ConcALARA

DCGLW

=

CostT

VAD × F × 0.025 × PD × A

×

r + λ

1 − e−(r + λ)N

6 (N-29)

7

8 The licensee can use Equation N-29 to determine the concentration in an area for which it 

9 should take a remediation action to meet the ALARA criterion. The equation appears 

10 complicated but can be solved in a few minutes with a hand-held calculator, and it only has to 

11 be done once for each type of remediation action at a site. PD, N, and r are constants. 

12 Section N.2.3 (Table N.2) gives generic values for PD and N, or they may be determined on a 

13 site-specific basis. The only site-specific information that the licensee needs is the total cost, 

14 CostT, and the effectiveness, F, for each remediation action being evaluated.

GUIDANCE FOR THE USE OF COMPOSITE SOIL SAMPLING FOR 

DEMONSTRATING COMPLIANCE WITH RADIOLOGICAL RELEASE 

CRITERIA

O-1

1 O.1 Introduction

2 This appendix provides information on the potential need for composite sampling, as well as the 

3 technical basis and guidance on acceptable approaches that licensees could use to incorporate 

composite sampling strategies into their FSS.

1 4 In addition, this guidance includes appropriate 

5 uses of composite sampling for generating the data for other decommissioning site 

6 investigations and surveys including characterization surveys. This appendix is based on the

7 report titled “Technical Bases and Guidance for the Use of Composite Soil Sampling for 

8 Demonstrating Compliance with Radiological Release Criteria,” (Vitkus, 2012). 

9 MARSSIM, Revision 1, provides guidance on the number of samples needed to have 

10 confidence in the survey results (constrain Type I and II error rates to acceptable levels), as well 

11 as considering the number of samples needed to ensure elevated areas between sample 

12 locations can be detected during scanning. For example, surface scanning specifications for 

13 Class 1 survey units include a required scan MDC (MDCSCAN) that is a function of the sample 

14 spacing and the respective DCGLEMC that provides assurance that elevated areas of concern 

15 are identified through either sampling or scanning. If the required MDCSCAN cannot be met, the 

16 number of samples may need to be increased to reduce the area between samples to match the 

17 actual MDCSCAN (i.e., a greater number of samples and denser sample spacing would lead to a 

18 smaller area between samples; smaller areas generally have higher allowable DCGLEMCs and 

19 therefore, a less restrictive MDCSCAN). For land area survey units (soil), use of scanning to 

20 detect elevated areas relies upon the radionuclides of concern (ROCs) being gamma emitters 

21 and, for the case of non-gamma-emitting ROCs (i.e., hard-to-detect [HTD] ROCs), reliance upon 

22 a surrogate relationship. However, the coupling of sample spacing with a required MDCSCAN

23 cannot specifically be followed when the scenario involves HTD radionuclides as the only ROCs 

24 or when a surrogate relationship cannot be established. In certain cases (e.g., only HTD 

25 radionuclides are present, HTD radionuclides with no surrogate relationship, average survey 

26 unit concentrations that are close to the DCGLW or sites with relatively high variability in residual 

27 radioactivity concentrations), the number of samples required may start to become cost 

28 prohibitive. In these cases, licensees may propose composite sampling to reduce the total 

29 number of samples that are analyzed and hence the analytical cost. 

30

31 The FSS-related portions of this guidance follow MARSSIM processes for demonstrating 

32 compliance with radiological release criteria and the average allowable residual contamination 

33 levels. However, additional evaluations are necessary to guide the use of composite sampling 

34 within the MARSSIM framework for the specific case of ensuring that elevated areas of concern 

35 are addressed when HTD radionuclides are a primary consideration. Both Federal agency and 

36 academic resources are available for supplementing the MARSSIM radiological survey 

37 processes with composite sampling and meeting the challenges presented by HTD 

38 radionuclides. These resources, coupled with the experience from chemically contaminated 

39 sites, provide the bases for the general approaches included in this appendix, that are 

40 applicable to identifying HTD elevated areas. 

41

1 As of the publication date of this volume, limited guidance is available on use of composite sampling in radiological 

surveys. MARSSIM, Revision 1 (2000), recommends that, if an inordinate number of samples are required, then 

the DQOs should be revisited. Chapter 14 of NUREG 1505 briefly introduces the concept of composite sampling 

as a means to reduce the total number of samples requiring analysis (NRC, 1998). 
O-2

1 O.2 Advantages and Disadvantages of Composite Sampling

2 There are scenarios where it could be advantageous for licensees to apply a composite 

3 sampling approach for the MARSSIM-based FSS DQOs and the associated Data Life Cycle. 

4 These scenarios normally will involve a specific set of conditions where the approach could be 

5 beneficial. Generally, such conditions would involve situations where the analytical costs are 

6 high, required MDCSCAN in Class 1 survey units are difficult to achieve, and/or the presence of 

7 HTD radionuclides both increase analytical costs and reduce detection capability. Therefore, 

8 successful implementation of composite sampling requires a well-thought-out plan and is 

9 normally only beneficial when the conditions above exist. Table O.1 summarizes when 

10 composite sampling is advantageous. The advantages numbered 1, 2, and 4 would be 

11 applicable for an FSS, and all of the advantages listed are applicable to other survey types, 

12 such as site characterization. The disadvantages listed must also be considered and addressed 

13 in the planning and data life cycle. Table O.1 also summarizes the uses and considerations that 

14 the licensee should include in its survey plan to ensure that composite sampling would not 

15 interfere with the decision as to whether or not the radiological release criteria are satisfied. 

16

17 O.3 Composite Sample Plan Design 

18 This section discusses the conditions under which a composite sampling approach is 

19 appropriate and would be considered advantageous. This section also discusses conditions 

20 that may affect decisions made during the DQA phase. 
O-3

Table O.1 Composite Sampling Overview

Advantages Disadvantages

1. Reduces analytical costs.

2. Provides a better estimate of mean

concentration in the study area over an

equivalent number of individual samples.

3. Identifies units that have the highest

contaminant levels.

4. With an appropriately adjusted

contaminant benchmark/investigation

level, composite sampling can increase

the ability to detect elevated areas by

increasing the number of locations

sampled.

1. Information is lost on the individual

sample increments that make up a

composite. This loss of information is a

concern when testing to determine if a

ROC exceeds a threshold (e.g., a

DCGLEMC) over a specific area because of

possible dilution to one or more

increments with elevated activity

concentrations by the other composite

increments.

2. Cannot be used when action levels

(DCGLWs) are near analytical detection

limits or the natural background

concentration levels.

3. For nonhomogeneous contaminant

distributions, temporal or spatial variability

information is lost.

4. Cannot be used when integrity of

individual sample values change as a

result of compositing (e.g., loss of volatile

contaminants, due to the physical

compositing mechanism).

Considerations for Applying Composite Sampling

1. Useful when the size of the pattern or feature of interest, such as elevated areas, is

smaller than the spacing between the statistically required random sampling locations.

2. User must account for potential introduction of large additional errors due to

heterogeneous nature of the contaminant in the matrix, or the matrix itself.

3. Aliquots used to form the composite must be of equivalent weight/volume and the

individual aliquots and the composite itself must be well homogenized.

4. Must account for the dilution factor when evaluating the result against a threshold, most

commonly a DCGLEMC. Necessitates a modified investigation level (MIL).

5. In most cases, the user must maintain the ability to retest individual samples (increments)

making up the composite to retrieve potentially lost information.

6. Should not be used to establish surrogate ratios.

7. May not be advisable for Scenario B analyses.
O-4

1 O.3.1 Uses, Considerations, and Limitations for Composite Sampling

2 O.3.1.1 Uses for Composite Sampling

3 • Composite sampling may be used to estimate the mean concentration of a ROC for

4 various decommissioning surveys including scoping, characterization, remedial action

5 support, and final status surveys. Before selecting a composite sampling approach, an

6 analyst should recognize that although an equivalent or better estimate of the mean may

7 result from this approach, there is a loss of crucial information regarding the overall

8 variability of the ROCs within the study area. For example,

9 o 10 locations—numbered 1 through 10—are selected for random sampling from the

10 study area.

11 o The analytically determined concentrations in pCi/g at these 10 locations are as

12 follows:

Sample No. Sample Value Sample No. Sample Value

1 3.1 6 9.7

2 3.9 7 3.4

3 8.9 8 6.8

4 2.0 9 6.6

5 1.9 10 6.9

13

14 o The calculated study area mean and standard deviation (sigma) would be calculated

15 to be 5.3 ± 2.8.

16 o Then, assume two composites are formed and analyzed: Composite 1 from the

17 even numbered locations and Composite 2 from the odd numbered locations.

18 Assuming equal aliquots, the composite concentrations would be 5.7 and 4.6,

19 respectively. The study area mean and sigma would be calculated to be 5.2 ± 0.8.

20 This example illustrates that the mean is accurately estimated with 2 composites; 

21 however, the actual study area variability is significantly underestimated. This factor will 

22 be an important consideration, should data collected during an earlier phase of the RSSI 

23 process be used for planning future investigations (e.g., the data are used to determine 

the number of samples required to ensure adequate power2 24 for hypothesis tests). 

25 • Compositing may be used to estimate the proportion of a population exhibiting a trait,

26 such as the presence or absence of a specific ROC.

2 Used in this context, power is the probability of rejecting the null hypothesis when it is false. For Scenario A, the 

power is the ability to find that a clean site is clean.
O-5

1 • Composite sampling may be used to classify survey areas as containing elevated areas 

2 or identify the spatial distributions of survey area ROC concentration levels, such as 

3 those parts of the site with the highest concentration levels. 

4 • Composite sampling may be used when the contaminant DCGLW, mean, variability,

5 and/or decision error combinations result in a relative shift <1 or otherwise require an

6 inordinate number of samples. The site may elect to use composite sampling to reduce

7 the number of samples requiring analysis, yet still meet the sample number to

8 adequately estimate the survey unit mean/median for the selected statistical test.

9 • Composite sampling may be used as a method to decrease sample spacing. This may

10 occur when the ROC may be a low abundance or low-energy gamma emitter with a high

11 MDCSCAN relative to the required MDCSCAN and hot spot or elevated area identification

12 considerations then become the driver for sample spacing and the respective DCGLEMC.

13 • Composite sampling may be used to reduce analytical costs associated with HTD

14 radionuclides, which typically require more expensive wet chemistry.

15 • Composite sampling may be used during characterization or to provide additional FSS

16 survey unit coverage for Class 2 and 3 areas to ensure proper classification of the unit.

17 Because Class 2 and Class 3 survey units should not have residual contaminant

18 concentrations in excess of the DCGLW when properly classified, under most FSS

19 conditions, there is limited, if any, benefit to composite sampling in properly classified

20 Class 2 or 3 FSS units. Use of composite sampling in these classifications would

21 necessitate application of an MIL that is a fraction of the reclassification investigation

22 level.

23 • Composite sampling may be used for HTD radionuclides for which an actual MDCSCAN

24 cannot be established (e.g., pure beta or alpha emitter in soil) and there are no

25 surrogate radionuclide relationships available. The composite sampling is used as a

26 method to increase the probability of elevated area or hot spot detection and as a means

27 to reduce analytical cost. However, this situation would require considerable evaluations

28 performed on a case-by-case basis. As such, this guidance provides only a general

29 scenario and the associated variables. If composite sampling is proposed to alleviate

30 sampling requirements associated with HTD radionuclides, the licensee should contact

31 the NRC early in the process to discuss the acceptability of the proposal.

32 • Composite sampling may be used when maintaining sample density is important, yet

33 one of the user’s objectives is to cover a larger area without increasing the analytical

34 budget.

35 Each of the above applications will require more rigorous data assessments and may require 

36 retesting of the individual increments comprising a composite in certain cases. These case 

37 requirements are discussed individually in Section O.4.

38 O.3.1.2 General Considerations

39 The justification for incorporating composites into a sampling plan relies upon several factors. 

40 Factors that will assist in properly using composites are discussed below.
O-6

1 (1) Composite sample data may be applied without modification when information on

2 individual samples is not important for the decisions that will be made with the data.

3 When a threshold concentration or IL is of importance (e.g., the DCGLW for a Class 2

4 survey unit or a DCGLEMC for a Class 1 survey unit), then a MIL must be established.

5 The MIL will be a fraction of the concentration threshold based on the number of

6 increments (k) that comprise the composite. There are at least two options for setting

7 the MIL limits. In most of the literature, the MIL is commonly defined as IL/k. Selection

8 of the MIL value is critical. A too high MIL may result in missing discrete samples that

9 exceed the IL (false negative). A too low MIL will conversely result in incorrectly

10 investigating composite results that did not contain increments exceeding the IL (false

11 positives). That is, a high false positive rate would be expected in cases where either k

12 is too high for the site conditions, where there is not an expected substantial difference

13 between the IL and the estimated site concentrations, or a combination of the two. The

14 default MIL proposed under most conditions should be established:

MIL =

IL

k

15

Revisions to the default MIL determination would require technical justification.3 16 In any

17 case, the site must provide both a lower and upper bound of the MIL for which definitive 

18 decisions may be made on a given result being definitely below or above a set action 

19 level.

20 (2) Composites are useful when analytical costs are high; otherwise, composite sampling is

21 generally not likely to be cost effective. Furthermore, additional costs associated with

22 forming composites and packaging and maintaining the increments is factored into the

23 cost differential, before determining the value added of composite sampling.

24 (3) Composite sampling is properly used when it will not affect the analyte integrity; it should

25 not be used, for example, for volatile analytes.

26 (4) The sample matrix must be amenable to homogenization, and each increment must be

27 equally represented in the composite.

28 (5) Analytical detection limits and/or background interferences must be sufficiently low,

29 relative to the proposed MIL, that the probability of misclassifying a composite sample

30 result and obtaining a result less than the MIL is negligible. An example for illustration

31 would be if the Sr-90 NRC screening level DCGLW of 1.1 pCi/g were the IL. A typical soil

32 matrix analytical detection limit is 0.8 pCi/g. Composite sample results would be

33 expected to have a high false negative rate and should not be used unless the objective

34 of the study is to only identify those areas of a site exhibiting high concentrations, as

35 Section O.4.5 discusses.

36 O.3.1.3 Limitations

37 Radiological survey plans that include composite sampling should be reviewed to ensure that 

38 the plan either addresses the following inherent limitations, or the limitation will not adversely 

39 affect the data decisions.

3 R. Correll (2001) described an approach that licensees may benefit from considering. 
O-7

1 (1) Reduction in the information on variability. This limitation is potentially detrimental for

2 situations such as calculating a sufficient number of samples that will satisfy statistical

3 power requirements and DQOs for hypothesis testing, during survey planning and

4 others.

5 (2) Due to loss of information on background variability, composite sampling may not be

6 appropriate for Scenario B analyses. When attempting to show survey unit

7 concentrations are indistinguishable from background (Scenario B analysis), background

8 variability should be assessed and retrospective power should be determined to ensure

9 a sufficiently low Type II (false negative) error rate (low error of retaining the null

10 hypothesis that the site is clean when the site is dirty). This is made more difficult using

11 a composite sampling strategy, which tends to underestimate variability.

12 (3) Potential loss of temporal or spatial information. In some scenarios, specific knowledge

13 of the concentration in a unit area smaller than that represented by the composite is

14 important in the decision process (e.g., for differentiating classifications of a site).

15 (4) Difficulty in homogenizing matrices. Such difficulty may be anticipated with clay soils

16 and/or in cases where the contaminant is present as small particles that cannot be

17 uniformly distributed.

18 (5) Lost information on maximum concentrations. This limitation is important for threshold

19 investigations and may be counteracted with appropriate protocols for retesting the

20 questionable composite sample’s individual increments.

21 (6) Lost information on concentration correlations for two or more ROCs. As a result,

22 composite sampling is not appropriate for determining surrogate relationships, such as

23 estimating the surrogate correlations.

24 O.3.2 Composite Sampling Plan and Data Quality Assessment Review Items

25 The following should be considered when reviewing composite sampling plans:

26 • The objective of the composite protocol, as well as how the resultant data will be

27 assessed is described.

28 • Whether the composite sampling design is implemented following proper

29 procedures/project specific instructions. American Standards and Testing Materials

30 (ASTM D 6051-96, Reapproved 2006), “Standard Guide for Composite Sampling and

31 Field Subsampling for Environmental Waste Management Activities,” provides a

32 potentially useful reference when developing procedures. Procedures should include

33 the following:

34 o Whether each composite increment is collected, and a representative aliquot from

35 each increment is containerized for possible reanalysis.

36 o Whether the remaining portions of each increment are homogenized and the

37 composite sample containerized.

38 o Whether each increment contributes an equivalent volume/weight to the composite.
O-8

1 • Whether the bases for the number of increments, k, per composite sample are provided.

2 Several factors to be considered include, but are not limited to, the physical nature of the

3 samples, the anticipated concentrations of the ROCs relative to the detection limits (to

4 ensure that a single increment above the action level could be detected), and the

5 capability to combine and homogenize the increments adequately. The larger k

6 becomes, the more difficult it will be to adequately homogenize the increments and,

7 therefore, an increase in the sampling error should be expected. To better control the

8 sampling error, there should be a limit to the number of increments. The analytical

9 detection limit (d) will also affect the value of k for a composite sample. This value is

10 defined as:

11 k < IL/d (EPA 1995)

12 When determining the number of increments, k, that the ratio of the analytical cost to 

13 composite acquisition costs, and the estimated ratio between the “error variability” 

14 component of the sample collection and measurement processes and the “inherent 

15 concentration variability” component were considered. Guidance for optimal increment 

16 determination is provided in EPA’s Quality System Document, QA/G-5S (EPA, 2002). 

17 Specific factors that should be considered include the following: 

18 o The analytical MDC relative to the DCGLW and/or MIL (i.e., the MDC should be lower

19 than the MIL).

20 o The ratio of the measurement error standard deviation (includes random error in

21 sample collection and compositing and the measurement error) to the inherent

22 variation standard deviation (variability in the concentration of the target population).

23 The optimal value for k will decrease as this ratio approaches unity. With increasing

24 values of k, it becomes increasingly difficult to prepare a representative composite.

25 Therefore, this guidance recommends minimizing k in most circumstances to

26 between 3 and 6, with a maximum of 10.

27 o Site quality control results that demonstrate that the error in composite formation is

28 minimized. This may be demonstrated via a matrix spike to one of the increments

29 forming a composite. The analytical result for the matrix spike composite sample

30 should approximate 1/k times the matrix spike concentration.

31 o The ratio of the per sample analysis cost and the per sample collection and handling

32 costs. Composite sampling is only cost beneficial if the analysis cost savings and

33 other benefits outweigh the additional sample collection, sample handling, and other

34 costs associated with composite sampling.

35 Once an optimal k is selected, the number of composite samples (n) analyzed will be:

36 n = N/optimized k

37 • Retesting protocols are needed whenever a threshold parameter such as a composite

38 sample MIL is involved. In most cases, the MIL will be a fraction (as a function of k as

39 discussed in Section O.3.1.2.) of the IL for discrete sampling. A composite sample result

40 greater than the MIL would require the licensee to investigate the result and possibly

41 analyze the composite sample increments.
O-9

1 • When used during a FSS, the composite data are evaluated collectively to provide an

2 estimate of the mean/median concentration levels in each survey unit, and individual 

3 composite results are compared to the respective MILs (and DCGLEMCs) for the 

4 composite sample spacing and are used in the statistical tests, as appropriate. 

5 Comparison against MILs provides assurance that release criteria are met.

6 • Provisions should be made in the plan to maintain individual composite increments for

7 retesting.

8 • When composite sampling is used to provide an increased probability of elevated area

9 detection, the maximum elevated area size and concentration of concern should be

10 carefully evaluated and approved before the plan is implemented. Section O.4.5 of this

11 appendix provides additional discussion regarding considerations for elevated areas.

12 • A clear assessment of individual and collective composite sample results should be

13 provided in documentation to ensure conformity with release criteria.

14 O.4 Composite Sampling Designs: Survey Planning Example Evaluations

15 O.4.1 Mean Concentration Estimation

16 A licensee may choose to use composite sampling during site investigations to assist decision17 making in estimating the mean concentration. The reason(s) for doing so may be to reduce 

18 overall analytical costs, to increase site coverage, or both. The target population of interest 

19 could range anywhere from the site as a whole to a specific survey area/unit, depending upon 

20 the decommissioning phase. A licensee may consider composite sampling for mean estimation 

21 during any phase of the site decommissioning, including the FSS. In many cases, composite 

22 sampling results are expected to be compatible with the hypothesis tests recommended in 

MARSSIM (e.g., Sign Test using Scenario A)4 23 . Simulations comparing power for composite 

24 sampling versus discrete sampling could be used to provide support for the method selected.

25 When an estimate of the study area mean is the sampling goal, the sample plan (e.g., ordinary 

26 random, systematic, stratified) is selected based on the expected distribution of the ROC—

27 homogenous, heterogenous, spatially related—and the overall DQOs. The user then calculates 

28 the number of samples (n) necessary to estimate the mean objective, in accordance with the 

29 anticipated data use endpoint. 

30 However, the loss of study area variability information with composite samples should be 

31 considered if composite sampling is intended for other uses besides the mean estimation. If the 

32 overall variability is underestimated, insufficient power and failure of the survey unit may result 

33 due to insufficient sampling for Scenario A. 

34 The NRC staff should review the use of composite samples to estimate the mean concentration 

35 within a study area to ensure that the licensee has provided the information listed below.

4 The number of samples (prior to compositing) should be determined based on agreed upon alpha and beta errors, 

and the relative shift (see MARSSIM, Rev. 1 for additional information). Because the relative shift is a function of 

variability, composite sampling results are not recommended for determining variability for the purpose of 

calculating the relative shift during the planning stage of the FSS due to the tendency for composite sampling to 

underestimate variability. 
O-10

1 • When the composite data are only intended to estimate the mean concentration,

2 information on the individual increments is not important. The mean is calculated using

3 the standard expression(s). However, there are two additional conditions if an IL for an

4 individual increment is a project requirement:

5 o A MIL should be determined

6 o Individual increments must be maintained for reanalysis, if necessary.

7 • When the data are intended for both estimating the mean and determining if a threshold

8 could be exceeded, the reviewer should also ensure that other conditions are satisfied.

9 These begin with determining if the MDC of the selected analyses are several factors

10 less than the expected ROC concentrations or IL, such that the concentrations of

11 interest are not masked by limited analytical sensitivity. An example of such a condition

12 and assumed parameters is provided next.

13

14 O.4.1.1 Composite Sampling Example 1—Evaluation of the Analytical MDC Against the 

15 MIL

16 In this example, the ROC is Th-230. A discrete sample investigation level of 10 pCi/g is

17 established. The selected analysis type is gamma spectroscopy with a 4-hour count time. The 

18 analytical MDC is 5 pCi/g. Five increments will be used to form the composite.

19 With the discrete sample IL of 10 pCi/g, the analytical result must be capable of identifying 

20 composite samples that exceed the MIL. The default MIL is calculated as the 10 pCi/g action 

21 level divided by the number of composite increments, or 10/5 = 2 pCi/g. 

22 Because the analytical MDC is greater than the IL and calculated MIL, composite sampling 

23 cannot be applied without a more sensitive analysis, such as alpha spectroscopy. Alternatively, 

24 if the number of increments, k, is reduced to 2, the default MIL becomes 10/2 = 5 pCi/g and the 

25 MDC requirement is met. 

26 This example may also be applicable for the other composite sampling data end uses discussed 

27 in this guidance. 

28 O.4.2 Identification of a ROC Presence/Absence Trait

29 Early in the site decommissioning process, composite sampling may be beneficial to assist the 

30 user in determining the proportion of the site or another population subset that exhibits a 

31 particular trait. Two specific examples include areal ROC differentiation across the site and 

32 area classification. The resultant data may be used both to provide an estimate of the 

33 proportion of the population that exhibits a specific trait and when additional testing is 

34 necessary.

35 (1) This approach could be used to establish what proportion of the site could be considered

36 for varying classifications. A simple example is determining the approximate proportion

37 of a site that is affected above the analytical detection limit for a ROC that is not present

38 in background.

39 (2) Identification of which ROCs are affecting site populations when the specific ROCs

40 affecting a site are expected to vary over distinct site regions.
O-11

1 With the objective being a presence/absence trait, a binomial distribution may be applied, and 

2 the number of positive composite results can be used to estimate the proportion of the site with 

3 the trait of interest. The estimated proportion (p) of the population with the trait is a function of 

4 the positive results (x) among m composite random samples. p is ≈ x/m. The number of 

5 positive results is dependent upon k. This interrelationship may therefore be expressed as:

p = 1 – (1 – x/m)

1/k 6 (EPA, 2002)

7 This approach may be further expanded to assist in the data gathering required to assess the 

8 presence or absence trait of elevated areas of HTD radionuclides within a specified unit area. 

9 The following considerations should be addressed.

10 (1) The a priori elevated area size and concentration of concern based on a dose

11 consequence evaluation.

12 (2) The probability of not hitting an elevated area of the determined size; defined as 1-β, the

13 acceptable risk is β.

14 (3) The probability of the existence of elevated areas based on sampling results.

15 These are the sample planning considerations. Once the plan is developed, there will be further 

16 evaluations and considerations that should include statistical inferences relative to probabilistic 

17 determination of the elevated area maximum size that could be missed by the plan design. 

18 Regulatory concurrence of unidentified elevated areas of concern would therefore be 

19 necessary. Once data are evaluated, further determinations can be made as to the probability 

20 of any elevated area being in the survey area and statistical inferences determined as to the 

21 maximum probable level. A suggested approach for bounding this maximum probable 

22 concentration is to apply Chebyshev’s inequality. Chebyshev’s inequality guarantees that no 

more than 1/k2 23 of the distribution’s values can depart more than k standard deviation from the 

24 mean (in this instance, k = any real number >0 rather than, as previously defined, the number of 

25 increments). From the maximum level and size inference, the dose consequence may be 

26 calculated and further assessed.

27 O.4.3 Composite Sampling to Augment Required MDCSCAN and/or Analytical Cost

28 Reduction during Final Status Surveys

29 A site may choose composite sampling as a means to resolve an FSS scenario where satisfying 

30 the required MDCSCAN for the applicable DCGLEMC results in an inordinate number of samples 

31 and an associated increase in sample analytical costs, which can be prohibitive. Other 

32 decommissioning survey phases may also select this approach to reduce the analytical costs.

33 Composite sampling can be used to satisfy the sample spacing requirements to detect elevated 

34 areas of concern in Class 1 FSS units for cases where it proves difficult to achieve the required 

35 MDCSCAN conditions. The sample plan should provide sufficient detail to demonstrate that full 

36 consideration has been given to conditions that could deleteriously affect the decision-making 

37 process.
O-12

1 O.4.3.1 Composite Sampling Example 2—Evaluation of MDCSCAN Requirements and 

2 Adjustments to the Number of Samples

3 An example of the approach is presented here. The ROC is Pu-238. The DCGLW for Pu-238 

4 based on dose modeling is 31 pCi/g. The gamma MDCSCAN (FIDLER NaI detector) is 740 pCi/g. 

A survey unit area size of 2000 m2 5 is selected. The Sign test has been selected for the 

6 statistical test. For this problem, the example DCLGEMCs for Pu-238 based on dose modeling 

7 are found in Table O.2.

8 Table O.2 DCGLEMC for Pu-238 

Pu-238 DCGLEMC (pCi/g)

10 m2 20 m2 50 m2 100 m2 200 m2

3,249 1,674 685 347 174

9 For this FSS example, assume a site’s DQO planning inputs result in 15 required samples. The 

maximum size for a Class 1 survey unit is 2000 m2 10 . The respective area represented by each 

sample is therefore 2000/15 or 133 m2. The DCGLEMC for a 133 m2 11 area is determined based 

12 on a fit of the data to an area factor curve and calculated to be 260 pCi/g. The actual MDCSCAN

13 (740 pCi/g) is compared with this required scan MDC (260 pCi/g). Because the actual MDCSCAN

14 exceeds the required MDCSCAN activity concentration level, sample spacing must be reduced to 

15 account for potential elevated areas. Based on the fit of the data, a DCGLEMC of 740 pCi/g 

would be acceptable for a sample spacing of 45 m2 16 . Therefore, 44 samples would be necessary 

17 to ensure the MDCSCAN is adequate to detect elevated areas of the corresponding magnitude. 

18 The site requests that sample sizes remain essentially equivalent to the original plan, due to the 

19 costs associated with analyzing 29 additional samples in multiple Class 1 survey units. 

20 Therefore, composite sampling and retesting are proposed and factored into the FSS plan to 

21 minimize the additional analytical costs. To do so, the DQOs must be planned accordingly and 

22 included in the decision-making process. This example is further developed below as an 

23 illustration.

24 Figure O.1 (left) shows the initial 15-sample design. However, elevated area or hot spot 

25 considerations require that the sample size be increased to 44. The original 15-sample design 

26 is maintained by forming composite samples composed of 2 to 3 increments each, as 

represented in Figure O.1 (right). The revised design sample spacing is 45.5 m2 27 ; the 

3-increment composite therefore represents a sample area of approximately 136.4 m2 28 . This

29 design will provide the necessary assurance that either sampling or scanning will identify any

30 elevated areas of concern. Each of the 44 increments is collected, composites are formed,

31 individual increments are maintained for retesting as needed, and the composites are analyzed.

32 The composite results are then compared with the appropriate MIL. A composite result less

33 than the MIL provides the evidence that elevated areas are below the action threshold and

34 composite results greater than the MIL will require retesting of the increments before a final

35 determination is made as to whether further investigations are required. The example assumes

36 that any elevated areas smaller than 45.5 m2 with activity levels greater than 740 pCi/g would be

37 identified during the scanning phase of the survey. If increment retesting or surface scanning

38 support the need for follow-up investigations, additional measurements will help determine the

39 extent and magnitude of elevations above the DCGLw and to assess whether the elevated

40 areas would cause the release criteria to be exceeded.
O-13

1 Figure O.1 Pu-238 Sample Number Comparison (the Left Figure Shows a Pu-238 Survey

2 Unit with 15 Sample Locations; the Right Figure Shows a Pu-238 Survey Unit 

3 Showing 44 Sample Increment Locations and 15 Composite Groupings)

4 The DQA phase will need added rigor to close the differential between the required MDCSCAN

5 that was addressed by increasing the sampled locations to 44 and forming the 15 composites. 

6 The data assessment would compare each composite concentration result with an appropriate 

7 MIL that accounts for the various scenarios that could exist. There are multiple combinations of 

8 increments adding elevated concentrations of the ROC to the composite results. For this 

9 example, the scenarios to consider for an MIL exceedance could be as follows:

10 (1) one increment adding a high concentration

11 (2) two increments adding moderate-to-high concentrations

12 (3) all three increments adding low-to-high concentrations

Composite 

analyzed

Compare 

result to MIL

< 

MIL

Analyze 

increments

No

es

Perform 

DQA

Yes

Analyze 

increments

No

es

Perform 

DQA

Yes

Analyze 

increments

No

es

Perform 

DQA

Yes

Compare 

result to MIL

<MIL

Analyze 

increments

No

es

Perform 

DQA

Yes

<IL Yes

No

Reassess 

survey unit

Composite 

Sample

Composite 

analyzed

Compare 

result to MIL

Analyze 

increments

No

es

Perform 

DQA

Yes
O-14

1 These assessments will lead to varying decisions, including no further investigation necessary, 

2 failure of the Elevated Measurement Comparison, or retesting increments and comparing the 

3 results to the applicable DCGLEMC. Table O.3 illustrates three scenarios that are based on the 

4 number of contaminated (involved) increments for a composite sample. The scenarios show 

5 the MILs that are a function of the area represented by the number of involved composite 

6 increments with Pu-238 concentrations that equal the respective DCGLEMC. The results show 

7 three differing concentration values where the potential exists for exceeding a DCGLEMC. 

8 Because of these various scenarios, the composite sample MIL must be established at the 

9 lowest concentration value—or 244 pCi/g in the Table O.3 example. Any composite result 

10 greater than 244 pCi/g would need retesting of the individual increments that comprised the 

11 suspect composite sample. 

12

13 The conclusion of the example is that, for the established parameters, using the composite 

14 sampling approach will provide a high degree of certainty that both the average residual Pu-238 

15 can be readily determined and elevated areas of concern will be identified with minimal false 

16 negatives when composite sampling with a defensible MIL is combined with surface scanning. 

17 Lastly, the analytical cost for the survey was maintained, although there will be additional field 

18 labor costs to collect, package, and record the composite sample and the increments.

19 Table O.3 Composite Result Investigation Level Evaluation

Area 

(m2)

DCGLEMC

(pCi/g)

Involved Increments/

% Activity Weighted Contribution at 

the DCGLEMC MIL (pCi/g)

Number of 

Increments

% Activity 

Contribution

136.5 254 3 of 3 1 254

91 378 2 of 3 0.67 253

45.5 740 1 of 3 0.33 244

20 % assumes other increment(s) have no added activity.

21 O.4.4 The Contaminant DCGLW, Mean, Variability, and/or Decision Error Combinations

22 Result in a Relative Shift <1 

23 MARSSIM recommends that the relative shift be maintained between a value of 1 and 3, such 

24 that a reasonable sample size results during the FSS. Site-specific conditions could lead to 

25 situations where the relative shift is less than one, requiring a large number of samples (N) to 

26 achieve adequate statistical power. Such conditions, either individually or in combinations, 

27 could include a high expected mean concentration (LBGR) relative to the DCGLW, a high degree 

28 of variability within the survey unit, or both. Current MARSSIM guidance recommends the 

29 following when this situation occurs: either reducing the value of the LBGR, which affects the 

30 concentrations at which a Type II error could occur or revisiting the DQOs together with the 

31 regulatory authority.

32 The basic two options available to the site for revising the DQOs such that sample sizes are 

33 reduced—other than reducing the LBGR—are increases in either the DCGLW or the Type I 

34 error. Both options require regulatory interactions and approvals. Another option that a site 

35 may consider is composite sampling, where the initial MARSSIM-designed sample size forms 
O-15

1 the basis for the number of increments (k). The site would therefore need to determine the 

2 number of composite samples that will be a function of an optimal k value. Section O.3.2 

3 contains the considerations for k that the licensee must evaluate. 

4 O.4.5 Classifying Survey Areas as Containing Elevated Areas

5 Section O.4.2 introduced the increased probability for elevated area detection, although when 

6 not performed correctly, the opposite effect could occur, where an elevated area is masked. 

7 The advantage for elevated area detection using composite samples is that, through the 

8 reduced analytical costs that can be achieved, more of the budget is available for sampling. 

9 This leads directly to better areal coverage and hence increases the probability that a sample 

10 location will fall on an elevated area. Therefore, a composite sample approach can provide 

11 greater confidence for detecting elevated areas of those ROCs considered to be HTD 

12 radionuclides (e.g., Sr/Yttrium-90, C-14), when a surrogate is not available. Therefore, an 

13 affordable FSS plan can only be developed by first deciding upon an acceptable elevated area

14 size and concentration magnitude. 

15 Again, this guidance provides an approach to overcome the limitations of MARSSIM in cases 

16 where HTD radionuclides are present in elevated areas. The guidance presents an approach 

17 that can more easily meet demanding sample requirements for statistical tests and for 

18 addressing elevated areas when HTD radionuclides are present. For this guidance, such a 

19 technical approach might include composite sampling, composite sampling further 

20 supplemented with rank set sampling, or the adaptation of a probability-based design for 

21 locating elevated areas of a predetermined size and shape (EPA, 2002; G.P. Patil, “Ranked Set 

22 Sampling,” 2002 (Patil, 2002b); J.E. Jozani, “Design Based Estimation for Ranked Set Sampling 

23 in Finite Populations. Environmental and Ecological Statistics,” 2010 (Jozani and 

24 Johnson, 2010; R.O. Gilbert, “Statistical Methods for Environmental Pollution Monitoring,” 1987 

25 (Gilbert, 1987)). However, before preparing either design, an a priori elevated area size of 

26 concern and the associated DCGLEMC must be determined. Once the elevated area size is 

27 determined, then either a discrete or composite sampling approach can be applied to provide a 

28 high level of probability that the elevated area will be sampled. 

29 There remains yet another problem associated with this a priori elevated area. The problem is 

30 how to select the a priori size, as there will likely remain stakeholder concerns for smaller 

31 elevated areas that could again be missed. Therefore, the technical justification should include 

32 additional dose modeling details as to the potential impacts of smaller elevated areas that could 

33 go undetected and the potential contribution to the total dose from all remaining source terms 

34 across the site. Other approaches that may be considered include maximum concentration 

35 bounding scenarios using Chebyshev’s inequality or a Bayesian approach to estimate the 

36 maximum potential contamination level at the site.

37 In summary, the HTD elevated area identification condition will need significant evaluation and 

38 discussions with appropriate survey designers and regulatory authorities.

39 O.5 Conclusion

40 In a “typical” radiological FSS, the reduced probability of identifying elevated areas is accounted 

41 for through the iteration of adjusting sample spacing to satisfy required scan MDC sensitivity. 

42 To limit the increase in sample sizes, more licensees are suggesting composite sampling as 

43 one method to control the increased analytical costs that result. As with most sampling 
O-16

1 approaches, there are both advantages and disadvantages that require evaluation and review to 

2 ensure that the plan accounts for any limitations.

3 With this information, robust radiological sampling plans can be developed that will address 

4 multiple issues while providing assurance on decisions about the average residual 

5 concentrations across the survey unit as well as elevated area considerations. The number of 

6 composite samples and composite increments can be controlled, based on the type of 

7 contaminant, DCGLW levels, analytical MDCs, and the corresponding appropriate MILs. 

8 Additional guidance on optimizing all factors may be found in the references.
O-17

1 O.6 References

2 ASTM International. “Standard Guide for Composite Sampling and Field Subsampling for 

3 Environmental Waste Management Activities. Designation: D 6051-96. West Conshohocken, 

4 PA. October 1, 2006.

5 Carson, Jr., J.H. “Analysis of Composite Sampling Data Using the Principle of Maximum 

6 Entropy.” Environmental and Ecological Statistics, Volume 8, pp. 201–211. Kluwer Academic 

7 Publishers. 2001. 

8 Chapman, J.A., B.A. Watson, A.J. Boerner, E.W. Abelquist. “Spatially-Dependent 

9 Measurements of Surface and Near-Surface Radioactive Material Using In situ Gamma Ray 

10 Spectrometry (ISGRS) for Final Status Surveys.” Oak Ridge Associated Universities, US 

Nuclear Regulatory Commission. Presented at 52 11 nd Annual Meeting of the Health Physics 

12 Society. Portland, Oregon. 2007.

13 Correll, R L. “The Use of Composite Sampling in Contaminated Sites—A Case Study.” 

14 Environmental and Ecological Statistics, Volume 8, pp. 185–200. Kluwer Academic Publishers.

15 2001. 

16 EPA. “EPA Observational Economy Series; Volume 1: Composite Sampling,” 

17 EPA-230-R-95-005. U.S. Environmental Protection Agency. Washington, DC. August 1995.

18 EPA. “Guidance on Choosing a Sampling Design for Environmental Data Collection for Use in 

19 Developing a Quality Assurance Project Plan,” EPA QA/G-5S. U.S. Environmental Protection

20 Agency. Washington, DC. December 2002.

21 Gilbert, R.O. Statistical Methods for Environmental Pollution Monitoring. Van Nostrand 

22 Reinhold. 1987.

23 Jozani, J.E., and B.C. Johnson. “Design Based Estimation for Ranked Set Sampling in Finite

24 Populations.” Environmental and Ecological Statistics. Springer Science & Business Media, 

25 LLC. 2005.

26 NRC. “A Nonparametric Statistical Methodology for the Design and Analysis of Final Status 

27 Decommissioning Surveys,” NUREG-1505; Revision 1, U.S. Nuclear Regulatory Commission. 

28 Washington, DC. June 1998.

29 NRC. “Multi-Agency Radiation Survey and Site Investigation Manual” (MARSSIM), 

30 NUREG-1575, Revision 1. U.S. Nuclear Regulatory Commission. Washington, DC. 

31 August 2000.

32 NRC. “Consolidated Decommissioning Guidance,” NUREG-1757, Revision 2. U.S. Nuclear 

33 Regulatory Commission. Washington, DC. September 2006.

34 Patil, G P. 2002a. “Composite Sampling.” Encyclopedia of Environmetrics, Vol. 1, pp. 387–391. 

35 John Wiley & Sons, Ltd, Chichester, England.

36 Patil, G.P. 2002b. “Ranked Set Sampling.” Encyclopedia of Environmetrics, Vol. 3, 

37 pp. 1684-1690. John Wiley & Sons, Ltd, Chichester, England.
O-18

1 Vitkus, T. “Technical Bases and Guidance for the Use of Composite Soil Sampling for 

2 Demonstrating Compliance with Radiological Release Criteria.” Prepared for the U.S. Nuclear 

3 Regulatory Commission by the Oak Ridge Institute for Science and Education. April 2012.

4 (ADAMS Accession No. ML13101A090.)

5 Watson, B., J. Clements, T. Vitkus, “Decommissioning Composite Sampling and Regulatory 

6 Guidance.” NRC and ORISE. Presented at 56th Annual Meeting of the Health Physics Society. 

7 West Palm Beach, Florida. June 26-30, 2011.
FRAMEWORK FOR USE OF ENGINEERED BARRIERS AT 

DECOMMISSIONING SITES

P-1

1 P.1 Framework for Use of Engineered Barriers at Decommissioning Sites

2 The purpose of this appendix is to provide guidance to licensees considering the use of 

3 engineered barriers to meet radiological criteria for license termination (e.g., engineered covers, 

4 including erosion protection designs, stabilizing cementitious materials, and reactive walls). 

5 Guidance topics include use of a risk-informed, graded approach for selection of engineered 

6 barriers, the engineered barrier analysis process, support for engineered barrier performance, 

7 and the need to consider potential engineered barrier degradation mechanisms. A detailed 

8 example of the application of a graded approach to the design and implementation of 

9 engineered barriers for erosion protection is also provided. 

10

11 The guidance is directed towards licensees pursuing restricted use, since it is envisioned that 

12 this is the situation where engineered barriers will most frequently be used. The guidance is 

13 directed towards designing new engineered barriers; however, it is recognized that some sites 

14 may have in situ engineered features that are part of the existing site being decommissioned, 

15 for which the licensee wants to take account. In these situations, elements of P.1.2 and P.1.3 

16 may be most helpful. The guidance is intended to strike a balance between providing adequate 

17 direction and being overly prescriptive. Because the application of engineered barriers is site18 specific (e.g., due to different radiological source terms, different exposure environments, and 

19 different natural systems), some elements of the guidance may not be applicable to every site.

20

21 P.1.1 Risk-Informed Graded Approach for Engineered Barriers

22 The general need for an engineered barrier at a specific decommissioning site should be 

23 determined by considering if the barrier would be needed for compliance with the LTR dose 

24 criteria (e.g., mitigates the impact of natural processes such as erosion and infiltration). Once 

25 the general need is determined, a risk-informed graded approach should be used for selecting

26 and designing engineered barriers at specific sites and documenting the technical basis for the 

27 selected barrier. The risk-informed, graded approach to engineered barriers is similar to the 

28 risk-informed graded approach for institutional controls described in Section M.2 of Appendix M 

29 of Volume 1, which consists of a general risk framework and associated grades of institutional 

30 controls. The same risk framework is used for both institutional controls and engineered 

barriers and is defined by the hazard level and likelihood of hazard occurrence1 31 . The NRC’s 

32 philosophy is that robust engineered barriers and additional bases for engineered barrier 

33 performance should be provided for higher risk sites (compared to lower risk sites). As 

34 described in Section M.2, higher risk sites are those sites with projected doses greater than 

35 1.0 mSv/y (100 mrem/y) with no institutional controls in place and sites that must rely on 

36 engineered barriers for relatively long periods of time (i.e., longer than 100 years) due to the 

37 presence of risk-significant quantities of long-lived radionuclides. Uncertainty in the 

38 performance of engineered barriers increases with time with limited data available on long-term 

39 performance. This coupled with an increasing likelihood of disruption of the engineered barriers 

40 makes it especially important to design a robust cover. Generally, “robust” means more 

41 substantial, reliable, and sustainable for the time period needed without reliance on ongoing 

42 active maintenance. The term “robust” is similar to the term “durable” used for institutional 

43 controls at higher risk sites (see Section M.2 of Appendix M of Volume 1). Section P.2 provides 

44 a detailed example of the application of the risk-informed graded approach to engineered 

1 It is important to note that hazard is different from risk. Hazard is primarily based on the inventory and dose

conversion factors, while risk also considers the ability of the radioactivity to be released. Therefore, a site can 

have a high hazard but low risk, if the radioactivity is sufficiently contained or access to the radioactivity is 

significantly limited. Engineered barriers can effectively reduce risk for otherwise high-hazard sites.
P-2

1 barriers for erosion protection. Robust engineered barriers are also needed for sites where the 

2 hazards are being significantly mitigated by the functioning of a barrier (i.e., risk reduction). 

3 “Significantly mitigated” is site-specific and can vary due to differences in the source and other 

4 features of the site (see example). In general, an engineered barrier that reduces the risk by 

5 more than a factor of 5 would likely be considered to be significant. The basis for this factor is 

6 that, in many circumstances, the uncertainty in the risk for restricted use sites can approach or 

7 exceed this value.

8

9 Although more frequently used at higher risk restricted use sites, robust engineered barriers 

10 may also be used at lower risk restricted use sites to limit reliance on active monitoring and 

11 maintenance. Such an approach might alleviate institutional controls and maintenance and 

12 thereby reduce licensee costs. In some limited cases, a robust engineered barrier could also be 

13 used at an unrestricted use site, because it has been designed so that its performance does not 

14 rely on active ongoing maintenance.

15

16 The concept of passive performance of an engineered system is not unique to decommissioning 

17 of a radiologically contaminated site. There are many long-lived man-made structures that have 

18 not benefitted from continual monitoring and maintenance. However, passive performance of 

19 an engineered system cannot be assumed. As indicated in the sections that follow, passive 

20 performance must be justified through design, experimentation, analysis, and support, and 

21 uncertainty must be considered in projecting performance, particularly over periods of time that 

22 exceed the experience base. In fact, the experience base is oftentimes much less than the time 

23 period relied on for performance for most waste disposals.

24

25 Uncertainty related to engineered barrier performance over a long temporal scale is also 

26 contextual in that the length of experience in the use of that type of barrier should be considered 

27 relative to how long lived the contamination is. In general, hundreds to thousands of years 

28 would be considered a long time when applying engineered barriers to decommissioning, 

29 because it is almost certain that site-specific experience for the performance of an engineered 

30 system does not extend beyond several decades. Section P.1.2 provides guidance on how the

31 barrier analysis process should be used to determine the performance of the engineered 

32 barriers. The risk-informed graded approach to engineered barriers is linked to the sections on 

33 effective barrier analysis (Section P.1.2) and the technical basis for engineered barrier 

34 performance (Section P.1.3), as well as to Section 17.7.4 of Volume 1, on maintenance and 

35 monitoring. For example, where there is a demonstration that the engineered barriers have 

(Example) Site A and Site B both contain soil contaminated with equal concentrations of 

Sr-90. The primary exposure pathway is from leaching of the contamination to groundwater. 

Both sites design an engineered cover to limit infiltration for the next 100 years. Site A has a 

thick unsaturated zone composed of a clayey soil. The estimated travel time to the 

groundwater, constrained by observations of past releases of Sr-90, is at least 200 years. 

Site B has a thin unsaturated zone composed of a sandy soil and the estimated travel time to 

the groundwater is 60 years. There are no additional data to constrain the numerical estimate 

of the travel time at Site B.

Conclusion: Site B would need a greater technical basis for the performance of its 

engineered cover compared to Site A, because Site B would have less risk reduction from the 

natural system and the travel time estimate (which is directly tied to risk via radioactive decay) 

is more uncertain.
P-3

1 been designed to be robust and not reliant on ongoing active maintenance and repair, the 

2 amount of information and resources needed to support maintenance would be considerably 

3 reduced.

4

The robustness of an engineered barrier and the amount of technical basis provided for an 

engineered barrier should be commensurate with the level of risk for a site (i.e., lower for 

lower risk sites or higher for higher risk sites) and the amount of risk reduction assumed for 

the barrier.

More robust barriers or strong support for engineered barrier performance is needed for 

higher risk sites or where hazards at a site are significantly mitigated by the barrier. 

5

6

7 Engineered barriers should be designed with the goal of remaining effective over the time period

8 needed to achieve compliance (as long as necessary and practical during the 1000-year 

9 compliance period at a minimum), especially for long-lived radionuclides. The following items 

10 should be considered in designing engineered barriers for decommissioning sites:

11

12 • Designs should simplify long-term control and minimize the reliance on active ongoing

13 maintenance and associated costs, especially for sites with long-lived radionuclides.

14 • Designs for sites with long-lived radionuclides (hundred years or more) should place

15 more reliance on natural materials and less reliance on synthetic materials that are less

16 proven over the long term (e.g., less reliance on compacted clay and more reliance on

17 enhanced evapotranspiration). For sites with short-lived radionuclides (tens of years or

18 shorter), synthetic materials may be more advantageous, because there is less

19 variability in their performance.

20 • The determination of adequate financial assurance should consider the cost of

21 monitoring and routine maintenance and the need for potential major repairs of the

22 engineered barrier, based on the time period during which the engineered barrier

23 performance is relied on to achieve compliance.

24 P.1.2 Engineered Barrier Analysis Process

25 This section is directed towards licensees pursuing restricted use. Licensees using engineered 

26 barriers for unrestricted use may find parts of this section useful but should only credit the 

27 passive performance (without monitoring, maintenance, and inspection) for mitigating 

28 radiological impacts. 

29

30 Decisions on approving the use of engineered barriers to meet the LTR will be performance31 based using the dose criteria of the LTR. Therefore, licensees have flexibility in how 

32 engineered barriers may be used together with other remediation activities to achieve the dose 

33 criteria. In addition, staff will base the approval decisions using a risk-informed approach. In 

34 their proposal to use engineered barriers, licensees should include all relevant information 

35 concerning the risks of using the approach versus other remediation alternatives.

36

37 To implement a risk-informed graded approach, licensees should accurately assess the 

38 performance of the engineered barriers. Using engineered barriers for compliance with 

39 Subpart E for restricted use, entails an analysis of the following two items under 10 CFR 

40 20.1403(b) and (e), respectively:
P-4

1

2 • the effectiveness of the engineered barrier in reducing risk when institutional controls are

3 in place (i.e., active monitoring and maintenance are in place to help ensure a properly

4 functioning cover)

5 • the effectiveness of the engineered barrier in reducing risk under a loss of institutional

6 controls (including loss of active monitoring and maintenance) that may lead to more

7 rapid degradation of engineered barrier performance compared to the case where

8 institutional controls are in place (given uncertainty in long-term performance of

9 engineered barriers, degradation should be studied through sensitivity analysis).

10 The first analysis might evaluate the dose to the average member of the critical group who is 

11 assumed to be located at the site boundary due to active institutional controls that prevent 

12 access to the site. The performance of the engineered barriers could include the benefit of 

13 active monitoring and maintenance that might increase the level or length of performance 

14 assumed for the engineered barrier relative to the case with no monitoring and maintenance. 

15 The dose limit specified in 10 CFR 20.1403(b) when institutional controls are in place is 

16 0.25 mSv/y (25 mrem/y).

17

18 The second analysis, which assumes a loss of institutional controls, would typically include 

19 evaluating exposure scenarios involving human receptors occupying or otherwise participating 

20 in onsite activities immediately following license termination or at time=0 years. The dose limit 

21 applied to an average member of the critical group under a loss of institutional controls is 

22 1.0 mSv/y (100 mrem/y) (or 5.0 mSv/y (500 mrem/y), if certain additional criteria specified in 

23 10 CFR 20.1403(e)(2) are met). The analysis should consider the impact of disruption of the 

24 engineered barriers from human activities (e.g., well drilling, home construction, resource 

25 exploitation), and other natural or human-induced degradation mechanisms, events, or 

26 processes (e.g., root penetration into an engineered cover [see for example DOE, 2009] or 

27 infilling of a lateral drainage layer) under a loss of institutional controls. In some cases, a 

28 receptor may disrupt the engineered system that could result in higher doses through a different 

29 pathway (e.g., an agricultural intruder disrupts an engineered cover with plowing but does not 

30 disrupt the contamination directly). Disruption of the cover could result in increased infiltration 

31 and higher groundwater pathway doses.

32

33 The analysis should also consider reasonably foreseeable and less likely but plausible events 

34 and processes that could lead to disruption of the engineered barriers. “Reasonably 

35 foreseeable” disruptive conditions from humans or natural events and processes are those 

36 processes and events expected to have a probable or likely occurrence over the analysis 

37 period. “Less likely but plausible” events and processes are less likely than reasonably 

38 foreseeable events and processes but should also be considered to better understand 

39 challenges to the system, evaluate uncertainty in engineered barrier performance, and provide 

40 additional risk information to the decision-maker. A features, events, and processes (FEPs) 

41 evaluation can be a useful tool for determining what FEPs should be considered. The intent of 

42 performance assessments is to provide reasonable assurance, considering uncertainties in 

43 engineered and natural systems over long time periods that the actual performance of the 

44 disposal facility will comport with the modeled or assumed performance. Natural or human 

45 processes and events are possible stressors to the engineered barrier design. On the other 

46 hand, highly unlikely or implausible processes or events would not be expected to have a 

47 significant influence on risk because of the low probability of occurrence and, therefore, need 

48 not be evaluated. 

49
P-5

1 The potential impact to engineered barriers from reasonably foreseeable disruptive conditions 

2 from humans or natural events and processes can vary by site and by barrier. In some 

3 circumstances, it may be reasonable to take no credit for the engineered barriers. For example, 

4 a hypothetical site installs a geomembrane at the surface to control infiltration at a restricted use 

5 site. In analyzing a loss of institutional controls, it would be reasonable to assume no 

6 performance credit for the geomembrane, because it could be easily removed or damaged by 

7 near surface processes, and this type of barrier is subject to discrete failures. On the other 

8 hand, consider a site in a rural setting using a large, thick earthen cover of low gradation. 

9 Complete removal (and complete degradation) of the earthen cover under these conditions, 

10 while possible, may not be considered reasonably foreseeable. 

11

12 Realistic exposure scenarios for future site use for this example (e.g., residential or farming) 

13 would dictate the reasonably foreseeable disruptive conditions from humans would commonly 

14 be construction of a foundation and a well for a residence that disrupts a portion of the cover, or 

15 farming of the site, which would result in plowing the top layers of the cover and the resultant 

16 potential impacts from the rooting of vegetation and increased infiltration. Because it may be 

17 difficult to (1) estimate the likelihood of disruption of engineered barriers or (2) predict the 

18 long-term performance of engineered barriers, the decision maker should obtain additional risk 

19 information by evaluating less likely but plausible events that could challenge the system, 

20 although it is not necessary to strictly compare the results of these less likely but plausible 

21 events and processes to radiological criteria for license termination.

22

23 With regard to engineered barrier degradation, the licensee should address the two cases 

24 where maintenance is either in place or lost, because the assumption for loss of institutional 

25 controls includes the loss of maintenance of engineered barriers and physical controls such as 

26 fences or signs. It should be noted, however, that, for those cases where an erosion control 

27 cover is designed in accordance with the uranium mill tailings guidance in NUREG-1623, 

28 “Design of Erosion Protection for Long-Term Stabilization,” issued September 2002, a case 

29 might be made for a durable, long-lived engineered barrier that does not rely upon ongoing 

30 active maintenance (i.e., maintenance needed to ensure that the design will meet specified 

31 longevity requirements) and associated future costs to maintain the erosional stability of the site. 

32 For this case, a licensee may be able to demonstrate that no significant degradation of the 

33 erosion control cover is expected.

34

35 The licensee should use the barrier analysis process to determine how much credit is being 

36 taken for engineered barrier performance in reducing risk. To accomplish this, the licensee 

37 should perform an analysis with the engineered system present and functioning. However, for 

38 unrestricted use or when evaluating the loss of institutional controls (without active monitoring 

39 and maintenance), the licensee should assume the performance of the barriers may degrade 

40 over time. An assumption of instantaneous and complete failure of a barrier is not required. 

41 The goal is to clearly identify the expected benefit of the engineered barriers quantitatively in 

42 terms of dose reduction. The analysis should also address the uncertainty in the expected 

43 performance of the barriers. For example, a comparison could be made of the doses without 

44 engineered barriers present (all off) (or with degraded engineered barriers) to the doses with 

45 engineered barriers performing as designed and expected (all on). The assumption to model a 

46 degraded barrier is generally more realistic than the assumption of the absence of a barrier and 

47 actually may lead to higher doses in some situations; for example, partial failure of a cover or 

48 partial failure of a grout wall system can focus water flow and create a “bathtub” effect. Thus, 

49 the licensee should use caution when making simple on-off analyses (i.e., where the barrier is 

50 either assumed to be completely present or completely absent). For some barriers an 
P-6

1 appropriate approach to defining a degraded condition or an end-state of ecological succession 

2 may be to use information from analog sites.

3

4 Caution should also be used when analyzing systems with multiple engineered barriers. When 

5 multiple barriers are present, the performance of one barrier may mask the potential contribution 

6 of another barrier to reducing risk (e.g., when redundant barriers are present). In these 

7 situations, the analysis may need to evaluate various combinations of barrier performance to 

8 determine the individual and cumulative contributions (positive or negative). In some cases, 

9 characteristics of one barrier may challenge or impair the performance of another barrier. 

10 These cases should be carefully assessed (e.g., increased infiltration resulting from eliminating 

11 vegetation by inert covers or the clogging of a drainage layer with a resultant hydraulic head 

12 over the waste). The licensee should determine the type of analysis on a case-by-case basis. 

13 For the complex and higher risk decommissioning sites and those sites with long-lived 

14 radionuclides, the use of a probabilistic analysis should be strongly considered, because a 

15 deterministic analysis may not be able to adequately address the uncertainty in the calculations. 

16 However, for simpler, lower risk sites and sites with short-lived radionuclides, a deterministic 

17 analysis with a sensitivity analysis may be sufficient. Deterministic methods should be used to 

18 select the design flood for the development of long-term erosion controls (see Appendix P). 

19

20 In summary, the analysis of engineered barriers should identify and evaluate those conditions or 

21 processes that are adverse to performance and result in noncompliance. The following should 

22 be provided:

23

24 • engineered barrier analysis with institutional controls in place (i.e., taking credit for active

25 monitoring and maintenance) for restricted use sites

26 • engineered barrier analysis assuming a partial or total loss of institutional controls

27 (i.e., not taking credit for active monitoring and maintenance) for unrestricted and

28 restricted use sites

29 • engineered barrier analysis considering reasonably foreseeable and less likely but

30 plausible disruptive human and natural events and processes that may lead to

31 degradation of the engineered barriers for unrestricted and restricted use

32 P.1.3 Technical Basis for Engineered Barrier Performance

33 Significant uncertainty exists concerning predicting the service life and long-term degradation 

34 rates of most engineered barriers. This section provides guidance on the main elements 

35 necessary to support the analysis and evaluation of engineered barrier performance in 

36 Section P.1.2, including the following:

37

38 • the design, features, and functionality of the engineered barriers

39 • the technical basis for assuming the barriers will meet the dose criteria, including

40 information about the length of time relied on for performance given the radionuclide mix

41 present and considering the degradation mechanisms and the combined and synergistic

42 effects resulting from the real-world conditions expected for the barriers over the

43 performance lifetime (Section P.1.4)

44 • the uncertainty in parameters and models used in assessing barrier performance and

45 the design of engineered barriers
P-7

1 • the suitability of numerical models used for validating engineered barrier performance

2 • a parametric or component sensitivity analysis to identify how much degradation of the 

3 engineered barrier would result in noncompliance (see Section P.1.2)

4 • model support for engineered barrier performance (e.g., analogs, experiments, simple 

5 engineering calculations to demonstrate reasonableness of the results)

6 • quality assurance and quality control (QA/QC) for the design and analysis of engineered 

7 barriers

8

9

10 Many engineered barriers are not amenable to model validation in the true sense; therefore, 

11 multiple lines of evidence are recommended. Model support can come in many different forms, 

12 including but not limited to: analogs, laboratory experiments, field experiments, formal and 

13 informal expert judgment, and engineering calculations to demonstrate the reasonableness of 

14 the results (e.g., hand calculations when numerical models are used). The level of model 

15 support should be commensurate with the risk significance of the engineered barriers to 

16 achieving site decommissioning (see Section P.1.1). If the level of performance needed for an 

17 engineered barrier is consistent with past experience at similar sites, and the engineered 

18 barriers have similar designs and QA, then the model support could be considerably less than 

19 for an engineered barrier with performance objectives that significantly exceed engineering 

20 experience. When considering engineering experience, the licensee must take care to ensure 

21 that the environmental conditions for the relevant degradation mechanisms are reasonably 

22 similar, since many degradation mechanisms can be very sensitive to environmental exposure 

23 conditions.

24

25 For engineered barriers that require very long-term performance, natural analogs should be 

26 considered to help define long-term conditions of the barrier and its environment. The greatest 

27 uncertainties stem from extrapolating the results of short-term tests and observations to long28 term performance. Standard engineering approaches frequently implicitly assume that the initial 

29 conditions persist, while the actual barrier more appropriately could be viewed as an evolving 

30 component of a larger, dynamic ecosystem (Waugh, “Ecology, Design, and Long-Term 

31 Performance of Surface Barriers: Applications at a Uranium Mill Tailings Site,” 1997). In fact, 

32 NUREG/CR-7028, “Effectiveness of Engineered Covers: From Modeling to Performance 

33 Monitoring,” issued December 2011, presents the results of an engineered cover study and 

34 concludes that compacted soil materials used in cover materials at the sites studied did not 

35 retain “as built” properties over periods of regulatory interest. The properties (specifically 

36 saturated hydraulic conductivity) of these materials change to values more typical of 

37 surrounding soils within 5 to 10 years after installation.

38

39 For some types of engineered barriers, natural analogs might provide information about 

40 possible long-term changes to an engineered system and can be thought of as a long-term, 

41 uncontrolled experiment. Some parameters (e.g., permeability, density, and diffusion 

42 coefficients) of natural analogs can be used to help define anticipated end-states of some 

Analysis can be used to understand the impact of uncertainty. For example, 1/4 loss of a 

cover may not result in non-compliance but loss of 3/4 of a cover would. Then determine if 

reasonably foreseeable (or less likely but plausible) natural and human processes might 

cause a loss of 3/4 of a cover. This analysis approach may be an additional way to deal 

with the uncertainty due to lack of data for long-term cover degradation.
P-8

1 simple engineered barriers. Evidence from natural analogs can help demonstrate that there are 

2 real-world complements to the postulated numerical predictions. It is important that the 

3 functionality of an engineered barrier be considered when developing analogs. The example of 

4 earthen mounds constructed by Native Americans, given below, presents a reasonable analog 

5 for the physical stability of an engineered cover. It does not provide analog information for the 

6 ability of engineered covers to limit infiltration or the release of radionuclides. Natural analogs 

7 can demonstrate long-term ecological conditions that may impact barriers such as plant species 

8 and spacing. Analog information is uncertain for a variety of reasons, such as unknown past 

9 environmental conditions. Therefore, analog information should not be envisioned as proof of 

10 future engineered system performance but rather as providing confidence that the engineered 

11 system is likely to perform as designed or to its final degraded end-state. 

12

13 Experience is limited for the long-term performance of most engineered barriers. However, a 

14 number of analogs to engineered covers have shown sustained durability, even for thousands of 

15 years (discussed in more detail below). In addition, some cementitious materials used by the 

16 Romans, for example, are intact after more than a thousand years of exposure to the 

17 environment. In the United States, cements in the Erie Canal and used in colonial settlements 

18 are 200–300 years old. 

19

20 When evaluating analogs, it is important to note that the structures that have persisted are most 

21 likely the most durable structures. That is why it is important to consider analogs that have 

22 persisted, as well as those that may have experienced damage or failure. An additional 

23 complicating factor is that the initial conditions and past exposure environment for the analogs 

24 are not known and may only be estimated. However, developing an understanding of analogs 

25 increases the likelihood that a design may be implemented with sustainable long-term 

26 performance. It should be reiterated that natural analogs should be only one element of the 

27 technical basis for the long-term performance of engineered barriers.

28

29 Monitoring and maintenance are needed to verify the effectiveness, durability, and service life of 

30 the engineered barriers. This monitoring involves both the environmental system surrounding 

31 the engineered facility that could be disruptive and the facility itself. (Section 17.7.4 of Volume 1 

32 discusses the risk-informed approach to monitoring.) Nondestructive monitoring technologies 

33 that included designed and emplaced sensors are preferred to conventional post-failure 

34 monitoring. Novel ideas, such as introducing special dyes or tracers within the engineered 

35 system, may facilitate identification of impending failures. The identification of these and other 

36 measurable performance indicators within a monitoring strategy, coupled with any needed 

37 repair or remediation, can be very important in extending the effective service life of these 

38 facilities and reducing risk. To the extent practicable, engineered barriers should be designed to 

39 support and simplify monitoring and maintenance. NUREG/CR-7169, “Sensors and Monitoring 

40 to Assess Grout and Vault Behavior for Performance Assessment,” issued June 2014, presents 

41 a preliminary evaluation conducted by the National Institute of Standards and Technology on 

42 the state-of-the-art sensors, nondestructive evaluation methods, and geophysical techniques 

43 used to quantify changes to chemical (e.g., redox state) and structural properties of large 

44 engineered systems used for waste isolation. While primarily applicable to waste disposal 

45 facilities, the report provides useful information on techniques that could be used to monitor 

46 engineered cementitious material performance at complex decommissioning sites. The study 

47 focuses on measurement techniques used to detect the onset of cracks and geochemical 

48 changes. Some techniques described in NUREG-2151, “Early Leak Detection External to

49 Structures at Nuclear Power Plants”, issued April 2013, may also be appropriate non-destructive 

50 monitoring techniques to assess the performance of barriers, especially for changes in moisture 

51 content.
P-9

1

2 Example of Natural Analog Model Support—Native American Mounds

3

4 With respect to the durability of earthen covers but not with respect to limiting infiltration or 

5 waste releases in the groundwater pathway, an example of a natural analog may be the various 

6 earthen mounds constructed by Native Americans that have survived for periods of 1,000–5,000 

7 years. Archaeologists have dated the mounds by excavating bones and artifacts from the 

8 mounds and determining the age of the object or the date of its burial. Information on the 

9 mounds is readily available by visiting State or national parks associated with the mounds. 

10 Also, there is considerable information and reference material available on the Internet. 

11 Examples of Native American mounds that have survived relatively intact for very long periods 

12 of time are shown in Table P.1.

13 Table P.1 Examples of Surviving Native American Mounds 

Mound Location Approximate Age (years)

Grave Creek Mounds Moundsville, WV 2,500

Hopewell Culture Mounds Chillicothe, OH 2,000

Cahokia Mounds Collinsville, IL 1,000

Poverty Point Mounds Epps, LA 3,500

Watson Break Mounds Monroe, LA 5,500

14

15 The mounds vary in size, with some being more than 23 meters (m) (76 feet (ft)) high and more 

16 than 100 m (300 ft) long. Mounds of this size reasonably approximate the size of engineered 

17 covers that may be installed at decommissioning sites. Therefore, the long-term stability of 

18 these analog sites provides additional assurance that wastes may be effectively stabilized for 

19 very long periods. It should be noted that stability in this context refers to erosional stability and 

20 not the ability of the mounds to limit infiltration, which is unknown. More importantly, guidance 

21 for earthen cover designs could be developed by examining the causes for failures of any 

22 damaged burial mounds (where only portions of the mounds remain and can be examined 

23 effectively). Understanding the long-term performance of analog systems may allow for 

24 additional safety factors to be used in current designs. 

25

26

27

28 P.1.4 Degradation Mechanisms and Functionality of Common Engineered Barriers

29 The purpose of this section is to identify, for licensee consideration, the common engineered 

30 barriers, the degradation mechanisms for the common barrier types and materials, and their 

31 typical functionality. This information may help the staff or licensees to select and design 

32 appropriate engineered barriers and understand the considerations needed to assess long-term 

33 barrier degradation, to make the overall decommissioning process more efficient.

34

35 The common engineered barriers are those that may be encountered at a decommissioning 

36 site. Because technology evolves, and a site may have unique considerations, this list should 
P-10

1 not be viewed as comprehensive. Engineered waste forms are not explicitly listed as a common 

2 barrier for decommissioning, because, in most instances, engineered waste forms such as 

3 might be present in a low-level waste disposal facility are not used. The assessment of 

4 engineered waste forms has been addressed in low-level waste disposal, and that guidance 

5 should be considered with respect to waste forms used in decommissioning (NRC’s “Technical 

6 Position on Waste Form”, Revision 1, January 1991). Common barriers that may be used at 

7 decommissioning sites and their primary functionality are as follows:

8

9 • Geomembranes—Geomembranes are synthetic materials used primarily to limit

10 infiltration to the waste or residual radioactivity. High-density polyethylene (HDPE)

11 geomembranes are increasingly used for their potential long service life and

12 effectiveness.

13 • Engineered cementitious materials—Concrete/mortar/grout are typically used to stabilize

14 waste or residual radioactivity, provide a chemically favorable environment for retention

15 of radionuclides, limit water contact, prevent erosion, provide shielding, and reduce

16 potential intruder contact with the waste or residual radioactivity.

17 • Engineered Covers—Multilayered and composite engineered covers are typically used

18 to limit infiltration, provide for shielding between the contamination and potential

19 receptors, eliminate exposure scenarios, and limit erosion. Covers for infiltration control

20 are typically either a resistive type or an evapotranspiration type.

21 • Erosion protection barriers—Barriers or parts of a barrier intended to reduce and slow

22 erosion. This could include a riprap layer encompassing part or all of an engineered

23 surface cover (e.g., side slopes of covers that have relatively high gradients).

24 • Vertical barriers—Vertical barriers may be soil-bentonite, soil-cement-bentonite,

25 cement-bentonite, sheet pile (steel or HDPE), and clay barriers and are primarily used to

26 control the horizontal migration of groundwater.

27 • Permeable reactive wall—A contaminant plume is channeled between impervious

28 vertical walls, referred to as the funnel, and flows naturally through a permeable reactive

29 barrier gate, where the pollutants are treated in situ during the flow process.

30 • Interceptor trenches—Used to intercept and collect contaminant releases, they are

31 typically only applicable with monitoring and maintenance.

32 • Chemical barriers—Chemical barriers are used to modify subsurface environmental

33 conditions (e.g., pH, redox potential or Eh) to limit the solubility of radionuclides or to

34 provide a more favorable geochemical environment for sorption. Engineered

35 cementitious materials are a good example (see above). Because of the diversity of

36 chemical barriers that could be applied, the degradation mechanisms and typical levels

37 of functionality are not provided in the following sections but would need to be evaluated

38 on a case-by-case basis.

39 P.1.4.1 Degradation Mechanisms

40 The degradation mechanisms may not be comprehensive, due to the large variability in 

41 conditions and processes from site to site, but they should represent those typically 

42 encountered. Degradation mechanisms depend on both the barrier and site-specific conditions. 
P-11

1 When evaluating degradation mechanisms, licensees should carefully consider whether the 

2 environmental conditions assumed or used in an analysis of long-term performance are 

3 representative of the in-situ conditions expected for the engineered barrier. The main 

4 degradation mechanisms are described below for the different engineered barriers. 

5

6 Degradation of Cement-Based Engineered Barriers 

7 The major environmental degradation processes that affect cement-based engineered barriers 

8 are sulfate attack, carbonation, corrosion of reinforcing steel, alkali-aggregate reactions, and 

9 leaching by acidic subsurface water. Other degradation mechanisms include freeze-thaw 

10 deterioration and microbiological attack. Center for Nuclear Waste Regulatory Analyses 

11 (CNWRA) 2009-001, “Review of Literature and Assessment of Factors Relevant to Performance 

12 of Grouted Systems for Radioactive Waste Disposal,” (issued April 2009) summarizes important 

13 degradation mechanisms for cementitious materials, many of which are discussed below. 

14 Degradation mechanisms can also include poor design and construction of cement-based 

15 structures, and can include differential settlement of the structures, stress concentrations, and 

16 seismic effects, as well as insufficient structural engineering design. To avoid these latter 

17 degradation mechanisms, the structures need to be properly designed and constructed under 

18 strict QA/QC procedures to ensure that they meet their design objectives. The discussions that 

19 follow will not address the design- and construction-related degradation issues of cement-based

20 materials.

21 Sulfate Attack

22 Sulfate attack is a process by which sulfate ions in solution chemically react with compounds in 

23 hydrated cement (e.g., portlandite, calcium aluminate, and calcium-silicate-hydrate) leading to 

24 expansion, strength loss, and potential disintegration of the cement. Naturally occurring or 

25 anthropogenic sources of sulfates of sodium, potassium, calcium, and magnesium are 

26 sometimes found in subsurface water and soils. The cement itself can also be a source of 

27 sulfate. Sulfate attack has occurred in several regions of the United States, particularly in arid 

28 regions such as the Northern Plains and the Southwest States. Localized sources of sulfates in 

29 groundwater include mine workings, mine tailings, blast furnace slag waste piles, and chemical 

30 waste ponds. Water used in irrigation can also be a potential source of sulfate attack, due to 

31 the gradual accumulation of sulfates in the soils. 

32

33 Solutions with sodium and magnesium sulfate can react with portlandite and calcium silicate 

34 hydrate (C-S-H) and lead to the formation of the expansive phase gypsum. C-S-H is the major 

35 binding component of hydrated cement and loss of C-S-H can lead to a loss of strength and 

36 disintegration of the hardened cement paste. Solutions with dissolved calcium sulfate can react 

37 with hydrated cement compounds such as calcium aluminate to form the expansive phase 

38 ettringite, which has a much larger molar volume compared to gypsum.

39

40 Sulfate attack is controlled by diffusion of sulfate ions into the cement material and is, therefore, 

41 lower for cements with lower permeability. Sulfate attack also decreases with the lower 

42 tricalcium aluminate content of dry cement powder or clinker. Cements mixed with significant 

43 quantities of supplementary cement materials, such as fly ash, ground blast furnace slag, and 

44 silica fume, are also less susceptible to sulfate attack. This is attributable to (1) less portlandite 

45 available to react with sulfate ions to form expansive gypsum, and (2) reduced permeability of 

46 the hydrated cement from pozzolanic reactions with the supplementary cement materials.

47
P-12

1 Carbonation

2 Carbonation refers to the general process of reaction of gaseous carbon dioxide (or dissolved 

3 carbonate originating primarily from atmospheric carbon dioxide) with cement, resulting in the 

4 precipitation of calcium carbonate in the form of calcite. Carbonation can lead to physical and 

5 chemical consequences, including carbonation shrinkage; however, carbonation shrinkage 

6 characterized by surface micro-cracking is unlikely to be continuous or lead to a disruption of 

7 the cement matrix. Carbonation can also have a beneficial impact, because calcite has a 

8 greater molar volume than portlandite and can lower the permeability of the cement matrix. 

9 Detrimental impacts of carbonation are typically associated with corrosion of reinforcing steel 

10 (discussed next) due to the reduction in pH. Following carbonation, the pH of the system is 

11 controlled by the solubility of calcite with a pH of around 8.3, compared to a pH of around 12.5 

12 before carbonation. At the lower pH following carbonation, the passive oxide layer on the steel 

13 surface is unstable, thereby allowing corrosion to occur if oxygen and moisture are present.

14 Corrosion of Reinforcing Steel 

15 Corrosion of reinforcing steel embedded in cementitious materials used for isolating residual 

16 radioactivity can lead to the formation of expansive phases, degradation, and cracking. In the 

17 presence of highly alkaline solutions typical of cement environments, reinforcing steel is typically 

18 protected by a passive oxide film. However, breakdown of this protective oxide film can occur 

19 through (1) carbonation of the cement leading to a lower pH of the cement pore water in contact 

20 with the steel, and (2) localized breakdown of the film by accumulation of chloride ions at the 

21 surface of the steel. Formation of expansive corrosion products can lead to pressure buildup 

22 that exceeds the tensile strength of the cement and can cause cracking, spalling, and 

23 delaminating of the material. Although chloride ions are common in nature, and small amounts 

24 are intentionally added in the mix ingredients of concrete to accelerate set times, the principal 

25 sources of chloride ions that cause problems in concrete are from deicing salts, sea water, and 

26 chloride ions in surface runoff.

27

28 An initiation time and propagation stage are typically considered when estimating times to 

29 corrosion of reinforcing steel. Factors that influence chloride-induced corrosion include (1) the 

30 type of cement, (2) water-to-cement ratio, (3) sulfate content, (4) service or environmental 

31 conditions, (5) degree of carbonation, and (6) properties of the steel. Although critical chloride 

32 threshold values have been proposed, due to the multitude of factors influencing the critical 

33 threshold, a more broadly accepted methodology is modeling of diffusion of carbon dioxide into 

34 the cement materials to estimate the initiation time for carbonation-induced corrosion.

35

36 Once the passive film on the reinforcing steel breaks down, the corrosion process can progress. 

37 Corrosion rates are a function of (1) permeability, (2) moisture content, (3) temperature, and 

38 (4) availability of oxygen. Climate also influences the time to degradation of the material caused

39 by steel corrosion, with more humid environments being more corrosive. In the case of

40 chloride-induced corrosion, oxygen diffusion models have been used to estimate corrosion

41 rates. Rates for carbonation-induced corrosion are typically on the high side, approaching a few

42 mils per year (1 mil= 0.0254 mm=0.001 inches (in.)).

43 Alkali-Aggregate Reactions 

44 Alkali-aggregate reactions can lead to the formation of gels that imbibe water and swell, 

45 resulting in expansion, cracking, and degradation of the cement. Alkali-aggregate reactions are 

46 usually internally contained in concrete and are not dependent on the diffusion of an aggressive 

47 solution into the cement-based material. A source of water is required for appreciable amounts 

48 of swelling to occur. Almost all aggregates react to some extent with alkalis in cement. It is 
P-13

1 when the reaction results in the formation of expansive products (e.g., gypsum) that serious 

2 cracking of the concrete occurs. Expansive alkali-aggregate reactions are known to occur when 

3 siliceous (i.e., alkali-silica reactions or ASR) and dolomitic (i.e., alkali-carbonate reactions) 

4 limestone aggregates are used. In addition, the rate of expansive reaction is influenced by the 

5 size of the aggregates. ASR is the most common, with the majority of the reported instances in 

6 the Western States. Alkali-carbonate reactions have occurred in some Midwestern and Eastern 

7 States. ASR has also been observed at the Seabrook nuclear power plant in Seabrook, NH. 

8 Granite aggregates used in the concrete mix at that facility led to cracking. ASR is a slow 

9 process, and its occurrence at Seabrook became evident only decades after the plant was 

10 constructed. Short-term reactivity tests (ASTM C1260-14, “Standard Test Method for Potential 

11 Alkali Reactivity of Aggregates (Mortar-Bar Method),” issued August 2014) are unlikely to 

12 predict the occurrence of ASR over long periods of performance of hundreds to thousands of 

13 years.

14

15 ASR can be mitigated by selecting aggregates that are not susceptible to deleterious reactions 

16 with alkalis in cement. These aggregates are well known and listed in standards and guidance. 

17 Because water is also necessary for the reaction to occur, drying the cement materials and 

18 preventing future contact with water is also an effective method of mitigating reactions.

19 Leaching

20 Buried concrete, in contact with percolating subsurface water, can undergo deterioration by the 

21 dissolution of the common constituents of cement paste (e.g., portlandite and C-S-H) leading to 

22 increased porosity and permeability and the subsequent loss of strength of cement materials. 

23 The rate and extent of the leaching are dependent on the acidity of the water, with groundwaters 

24 below a pH of 5.5 potentially leading to severe attack. Characteristics of particularly aggressive 

25 groundwater with respect to cement leaching have been defined empirically using correlations 

26 with dissolved carbon dioxide (CO2), bicarbonate ion, and calcium ion concentrations. Studies 

27 have shown that a wide range of acidic, CO2 -containing waters have the ability to significantly 

28 enhance dissolution of cement phases, compared to pure water. The potential leaching 

29 capabilities of subsurface water has also been related to the Langelier Index. The Langelier 

30 Index is related to the total dissolved solids, total alkalinity, pH, and calcium content of the

31 water. A positive index indicates that calcium carbonate will be precipitated, while a negative 

32 index indicates lime-deficient water capable of dissolving calcium from the hardened cement 

33 paste.

34

35 The extent of leaching depends on the quality of the cement, water infiltration rates, 

36 temperature, and the chemistry of the groundwater. Higher quality cement and decreased 

37 water flow through the system will help lessen the detrimental impacts of leaching on 

38 cementitious materials.

39
P-14

1 Freeze-Thaw Attack

2 Damage to cement-based materials occurs when water-saturated concrete is exposed to 

3 prolonged cycles of freezing and thawing. Structures most susceptible to freeze-thaw damage 

4 are surfaces of structures where flowing or ponding water can remain in contact with the 

5 concrete structure for extended periods of time. Several precautions should be taken to avoid 

6 freeze-thaw damage, including the following: precluding ponded water from the concrete 

7 structure, incorporating entrained air, placing properly, consolidating, and curing. Cementitious 

8 materials located at depth are also less susceptible to freeze-thaw damage.

9 Microbiological Attack

10 Sulfate-producing bacteria are capable of oxidizing elemental sulfur and sulfides to sulfuric acid 

11 under aerobic conditions, which in turn degrade cementitious materials. Some bacteria can 

12 attack cement-based materials by transforming ammonia into nitrites or nitrates or by producing 

13 lactic acid or butyric acid. In the normal design life (e.g., 40 years) of conventional 

14 cement-based materials, bacterial action does not seem to be a major cause of deterioration. 

15 However, when the chemical durability of cementitious materials must be relied on for hundreds 

16 of years, the impacts of bacterial activity should be assessed, although they may be difficult to 

17 predict.

18 Cracking of Concrete 

19 Cracking can originate within concrete due to a number of mechanisms. During placement, if 

20 the evaporation rate is great enough, the concrete surface can develop tensile stresses 

21 sufficient to crack the concrete. These plastic-shrinkage cracks typically extend through the 

22 entire concrete member. Cracking can also be caused by (i) settlement of the concrete 

23 member, (ii) flexural stresses, and (iii) thermal effects. The continued removal of water by the 

24 hydration process will generate a chemical-shrinkage stress that can initiate autogenously 

25 shrinkage cracks. Subsequent drying due to ambient conditions will also create shrinkage 

26 stress that generates drying-shrinkage cracks. Absent environmental conditions that may cause 

27 concrete-material degradation, cracking can be the most severe degradation mechanism 

28 affecting concrete. Transport through cracks in concrete will only be of consequence if the 

29 cracks extend throughout the concrete member. Relatively large amounts of water can be 

30 transported through a crack, depending on the total potential water-head across the crack, as 

31 well as the crack aperture and density.

32

33 Degradation of Engineered Covers 

34 Engineered surface covers are commonly designed to eliminate or significantly limit infiltration of 

35 subsurface water into the waste but may also be designed to limit gaseous release, provide 

36 shielding, reduce the likelihood of contact with the waste, or create physical stabilization at a 

37 site. From an infiltration management perspective, engineered covers are based on either 

38 resistive or water balance concepts. Covers based on resistive concepts use impermeable 

39 layers to prevent water from contacting the waste. Evapotranspiration covers based on water 

40 balance concepts attempt to mimic natural systems to evaporate and/or transpire water from the 

41 system, typically by allowing plants to grow on the cover. 

42 Degradation of Resistive Covers

43 Resistive engineered cover designs have a wide range of configurations, ranging from a simple 

44 one-layer cover to complex multilayered designs composed of soils and geosynthetics. Soil 

45 materials can include vegetative soils, permeable sand and gravel drainage layers, low 

46 hydraulic conductivity clay soils, and filter soils to preclude the migration of fines from soils 
P-15

1 overlying drainage layers, which cause the drainage layers to clog. Geosynthetic materials 

2 include geomembranes and geosynthetic clay layers. Composite barriers use both soils and 

3 geomembranes/geosynthetic clay layers. Geomembranes are essentially impermeable 

4 polyvinyl chloride (PVC) or polystyrene layers, while geosynthetic clay layers are composed of a 

5 thin layer of bentonite between two geosynthetic textiles, which may be used in conjunction with 

6 geomembranes. The effectiveness of these engineered covers lies in the ability to significantly 

7 limit infiltration of subsurface water into the waste. Cover performance is affected by waste 

8 stability, the degree of settlement, and slope stability of the cover system. 

9

10 Degradation mechanisms unique to engineered cover materials, other than those specific to 

11 inadequate design or construction and QA/QC issues, are discussed in NUREG/CP-0195, 

12 “Proceedings of the Workshop on Engineered Barrier Performance Related to Low-Level 

13 Radioactive Waste, Decommissioning, and Uranium Mill Tailings Facilities,” held August 3–

14 5, 2010. Degradation mechanisms include the following physical, chemical, and biological 

15 processes:

16

17 • physical processes such as freeze-thaw, wet-dry cycle, differential settlement, retention

18 of borrow soil structure (peds or clods) during construction, silting-in of drainage layers,

19 ultraviolet degradation, thermal degradation, erosion, fire, and pedogenesis

20 • chemical processes, such as oxidation of geosynthetic materials and cation exchange

21 mechanisms in sodium bentonites

22 • biological processes that involve unanticipated ecological consequences of designs that,

23 by creating habitat for deep-rooted plants, burrowing animals, and soil microorganisms,

24 can alter soil hydraulic properties

25 NUREG/CR-7028 assessed changes in hydraulic conductivity of 27 cover soils from a variety of 

26 sites across the United States. A major finding of the report was that as-built properties of cover 

27 soils changed significantly over a period of 5 to 10 years regardless of climate or design. For 

28 the covers that were evaluated, measured saturated hydraulic conductivities increased 

29 regardless of the initial saturated hydraulic conductivity. Saturated hydraulic conductivity of 

30 storage and barrier layers for in-service covers was always at least ten times greater than as31 built values. In a few cases in-service values were thousands of times greater. All evaluated 

32 cover types showed alterations in saturated hydraulic conductivity in all climates and for both 

33 barrier and storage layers. Changes in hydraulic conductivity were primarily attributed to cracks 

34 formed by drying events.

35 Compacted Clay Barrier

36 The function of the clay barrier layer in the engineered cover is to prevent and block infiltration 

37 of subsurface water through the cover into the waste. The compacted clay layer is frequently 

specified to have a saturated hydraulic conductivity of 1x10-7 38 cm/sec or less. The longevity and 

39 effectiveness of the engineered cover are influenced by the ability of the clay layer to retain low 

40 permeability characteristics. However, laboratory and field studies have shown that compacted 

41 clay can develop, even within ten years, distinct soil structures such as aggregates and planes 

42 of weakness due to pedogenic processes such as wet-dry and freeze-thaw cycling. These soil 

43 structures, in turn, can significantly increase the hydraulic conductivity of the initial compacted 

44 clay. Larger changes were observed for soils with lower as-built saturated hydraulic 

45 conductivity and soils with a greater proportion of clay particles in the fines fraction. 

46 Degradation of barriers is related to grain size of the material with higher clay content materials 

47 undergoing greater increases in hydraulic conductivity (NUREG/CR-7028). Unanticipated 
P-16

1 ecological processes (e.g., biointrusion), compaction, a drier or wetter than optimum 

2 environment, freeze-thaw cracking, and differential settlement can result in the degradation of 

3 clay and other resistive barriers.

4 Drainage and Filter Soil Layers

5 A common degradation mechanism affecting these drainage and filter components of the 

6 engineered cover system is the potential clogging of these materials by finer soil particles from 

7 overlying soils, including colloidal or biological materials. Clogging of drainage and filter soil 

8 layers can greatly reduce their permeability, allowing more water to move downward through 

9 waste material, and render them unable to perform their intended function to drain subsurface 

10 water.

11 Composite Soil Covers

12 Composite covers comprise a combination of a compacted clay layer, geosynthetic clay layer, 

13 or geomembrane. In general, composite covers have performed well in the time frame of 

14 available field studies (10+ years). Geomembranes protect clay barriers and geosynthetic clay 

15 layers by eliminating or significantly reducing vertical water vapor migration. Geomembranes 

16 (e.g., HDPE) are often placed on top of geosynthetic clay layers to extend the performance life 

17 of the geosynthetic clay layer. Degradation mechanisms of geomembranes include: puncture 

18 by granular soils and construction equipment; behavior of “waves” or fabric wrinkles due to 

19 temperature and overburden stresses; long-term degradation of geomembranes under the 

20 influence of ultraviolet light, chemicals, and radiation effects; the potential for slippage between 

21 geomembranes and adjacent materials; material embrittlement over time; and animal burrowing 

22 and root penetration. Some degradation mechanisms, such as exposure to ultraviolet light, can 

23 be managed with effective QA/QC. 

24

25 Geosynthetic clay layers are sometimes used instead of compacted clay and seem to perform 

26 better with respect to water flow, chemical degradation due to cation exchange, and mass 

27 transport (i.e., diffusion and retardation). As with geomembranes, however, geosynthetic clay 

28 layers have inherent problems due to installation activities (e.g., puncturing and degradation by 

29 construction equipment). Geosynthetic clay layers can experience desiccation cracking similar 

30 to clay layers. In addition, recent research seems to indicate a potential for cation exchange 

31 between commonly available calcium-laden fluids and the sodium in the geosynthetic clay layer 

32 bentonite, thus rendering the geosynthetic clay layer incapable of functioning as a 

33 low-permeability barrier layer in engineered-cover systems (James et al., “Field Performance of 

34 GCL under Ion Exchange Conditions,” issued 1997; Melchoir, “In-Situ Studies on the 

35 Performance of Landfill Caps,” issued 1997; Lin and Benson, “Effect of Wet-Dry Cycling on 

36 Swelling and Hydraulic Conductivity of GCLs,” issued 2000). Research is being conducted to 

37 more carefully study these degradation mechanisms.

38

39 NUREG/CR-7028 evaluates the engineered properties of a variety of covers located around the 

40 Unites States, including those that used composite materials. Under certain conditions, 

41 composite covers were found to perform better than conventional soil covers. For example, 

42 geosynthetic clay layers hydrated to water contents in excess of 50 percent retained relatively 

43 low hydraulic conductivity over the 5–10-year timeframe of the study. To achieve these water 

44 contents, NUREG/CR-7028 recommends geosynthetic clay liners (GCLs) be placed on 

45 subgrades with moisture content greater than 10 percent with an overlying geomembrane to 

46 prevent desiccation.

47 Degradation of Evapotranspiration Covers
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1 Evapotranspiration covers attempt to manipulate the water balance of the source zone by 

2 enhancing soil water storage and evapotranspiration by planted or wild vegetation. The 

3 performance of evapotranspiration covers depends on many factors, especially the climatology, 

4 soil hydrology, and plant ecology at a site. Evapotranspiration covers may be used in a variety 

5 of settings but may be most effective in arid or semiarid climates with high potential 

6 evapotranspiration.

7

8 The effectiveness of an evapotranspiration cover is dependent on the development of a design 

9 that is effective over the range of expected natural and ecological conditions, which are 

10 inherently variable over the timeframe of most decommissioning analyses. As discussed in 

11 Section P.1.4, evapotranspiration covers have been analyzed in detail at a number of sites. 

12 With effective design and development, evapotranspiration covers may be very effective, 

13 especially in arid and semiarid climates. In humid climates, evapotranspiration covers may be 

14 effective at managing a substantial fraction of the infiltration but not achieve design goals. 

15 Infiltration can commonly exceed evapotranspiration in humid climates or in colder climates, 

16 where a large fraction of infiltration may occur as snowmelt when evapotranspiration is low. 

17 Therefore, one of the major lessons learned, albeit not related to physical degradation, is that 

18 the design of an evapotranspiration cover must consider natural and ecological variability over 

19 the period of performance.

20

21 Degradation mechanisms associated with evapotranspiration covers include unanticipated 

22 ecological consequences, evolution of soil properties (e.g., pedogenesis), and effects of 

23 disturbance (e.g., fire, land use) on plant ecology. An example of unanticipated ecological 

24 consequences is the development of deep-rooted plant species that result in pathways for 

25 moisture below the design zone for moisture storage and removal.

26 Degradation of Erosion Protection Barriers

27 The degradation of erosion protection barriers is described in detail in the example provided in 

28 Section P.2.

29 Degradation of Permeable Reactive Barriers 

30 Permeable reactive barriers are in situ constructed walls below the land surface that intercept 

31 contaminated groundwater, which is funneled through it. Reactive materials in the wall can sorb 

32 chemicals and radioactive species on their surfaces and precipitate contaminants dissolved in 

33 the flowing water. In some cases, nutrients and oxygen in a permeable reactive barrier help 

34 microbes in the soil to precipitate contaminants and radioactive species. Experience with 

35 permeable reactive barriers seems to indicate that not all of them are performing well, often due 

36 to poor placement in the groundwater flow field. Material properties, such as grain size of 

37 reactive zeolites, can be changed in the construction process, thereby reducing the hydraulic 

38 conductivity. There seems to be a decrease of barrier performance due to loss of reactivity and 

39 permeability over relatively short periods of time. Systems have experienced challenges 

40 associated with diversion of water around the permeable reactive barrier caused by decreased 

41 permeability of the reactive barrier material in less than 5 years.

42 Degradation of Vertical Barriers 

43 Various types of subsurface vertical barriers are in use. Their primary purpose is to impede or 

44 preclude horizontal groundwater flow. These vertical barriers are placed at depths up to 60 m 

45 (200 ft) and often vary in thickness from 0.6–1.2 m (2–4 ft). The barriers must extend down to 

46 an impermeable natural horizontal barrier, such as a clay zone, to effectively impede 

47 groundwater flow from below. These barriers are often designed as temporary or semi48 permanent remediation techniques to isolate contaminated fluids from migrating to 
P-18

1 uncontaminated surrounding groundwater. Some soil-bentonite mixtures are not able to 

2 withstand attack by chemicals such as strong acids, bases, salt solutions, and certain organic 

3 chemicals. This hastens the deterioration of the barrier. Verification that the vertical wall forms 

4 a continuous barrier is critical to the function of this technology. Although it may be difficult to 

5 identify flaws in the continuity, or gaps in the wall, monitoring is essential to verify their 

6 performance. Although vertical walls have been used for decades, the process of designing the 

7 proper mix of wall materials to contain specific contaminants is less well developed.

8

9 Potential Levels of Functionality and Uncertainty

10 This section gives some general information to help licensees initially consider the use of 

11 engineered barriers. As discussed in Sections P.1.2 and P.1.3, licensees are responsible for 

12 developing acceptable technical bases and conducting analyses of engineered barriers 

13 proposed for a specific site.

14

15 The ranges of functionality or performance for different barriers and the associated uncertainty 

16 are based on a broad consideration of observations and analyses throughout the national and 

17 international community. Potential ranges of functionality are levels of performance that can 

18 likely be supported by technical bases and analyses, not solely based on demonstrated field 

19 experience. As the time scales get longer, past direct observation of the performance of 

20 engineered barriers that can be cited as a basis becomes less likely, and therefore, 

21 performance for longer times becomes more uncertain and based more on inference. The 

22 functionality provided here can be thought of as the level of performance believed to be 

23 reasonably achievable with proper design, analysis (Section P.1.2), technical basis 

24 (Section P.1.3), and implementation (quality), given current understanding and engineering 

25 practice. The ranges for potential functionality help provide direction as to when a less technical 

26 basis may be needed (assume less than typical performance) compared to a more technical 

27 basis (credit is taken for more than typical performance). The level of uncertainty should be 

28 considered in developing a monitoring and maintenance plan. Barriers with less uncertainty 

29 might need less reliance on monitoring and maintenance. In contrast, barriers with higher 

30 uncertainty may need substantial monitoring and maintenance until uncertainties are reduced. 

31

32 The NRC’s discussion in this section is an initial attempt to provide some insights on potential 

33 functionality and uncertainty, using readily available information. This section could be 

34 expanded, based on future studies and input from other programs involved with engineered 

35 barriers. Typical levels of functionality or performance of the main engineered barriers are as 

36 follows.

37

38 Cement-Based Engineered Barriers 

39 The performance of cement-based materials to isolate residual radioactivity can be divided into 

40 two categories:

41

42 • hydrologic effectiveness or physical containment of the wastes to preclude water

43 contacting the waste

44 • chemical effectiveness or the ability of the high pH characteristics of the intact and

45 degraded concrete to limit transport of the radionuclides to the accessible environment

46 Absent environmental degradation factors of concrete (e.g., sulfate attack, chloride corrosion), 

47 full-depth cracking of the concrete member can be the most severe degradation mechanism 
P-19

1 causing contact of water with the waste and the subsequent release of radionuclides. 

2 Accordingly, cement-based physical barrier structures need to be monitored for hydrologic 

3 effectiveness, and the projected service life for the structure should be revised based on an 

4 analysis of the monitoring data. Assuming adequate design and construction practices and 

5 excellent QA/QC, followed by a competent monitoring program, a service life of tens of years to 

6 a few hundred years appears feasible.

7

8 In addition to a physical barrier, Portland cement-based materials provide a chemical barrier by 

9 raising pH of water moving through it. For some radionuclides this can substantially lower their 

10 solubility and reduce releases. Some cementitious materials (e.g. HLW tank backfill grouts) are 

11 designed to provide chemically reducing conditions that can also limit solubility and transport of 

12 certain elements. The length of time chemical barriers, which mitigate releases of radioactivity, 

13 may need to perform to demonstrate compliance with radiological criteria for license termination 

14 is dependent on the length of time residual radioactivity poses a risk to potential receptors. 

15 Chemical barriers to waste release may perform for longer time periods than hydrological 

16 controls. Nonetheless, the longevity of chemical effectiveness is strongly related to the bulk 

17 hydraulic properties of the material, the chemistry of infiltrating water, and the composition of the 

18 cementitious materials (e.g., reductive capacity and cement fraction). This is because the 

19 quantity of water flowing through the cement matrix, the chemical gradients between infiltrating 

20 water and cement pore water, and the reductive and buffering capacity of the cementitious 

21 materials dictate the rate at which cement components that aid in the retention of radioactivity 

22 are leached from the system.

23

24 A cementitious barrier used to limit potential intruder contact with waste, with proper design, 

25 construction, and QA/QC, could be expected to be effective for hundreds of years if it remains 

26 unexposed to aggressive environmental conditions (e.g., high sulfate, excessive freeze-thaw 

27 cycles). Performance of this type of barrier may be enhanced with appropriate monitoring, 

28 repair, and remediation strategies.

29 Engineered Covers 

30 Research indicates that extensive desiccation of clay barriers in soil covers has compromised 

31 the ability of clay barriers to retain low permeability characteristics and prevent infiltration of 

32 water through the cover and into the waste (Albright, “Field Water Balance of Landfill Final 

33 Covers,” issued 2004). Conversely, composite covers composed of a combination of 

34 compacted clay buried at sufficient depth, geomembranes, and GCLs have performed well in 

35 the timeframes of available field studies (10+ years). Continued monitoring may be needed to 

36 verify the effective lifetimes of these covers. Current experience provides evidence that 

37 hydrologic functionality of tens of years for composite covers appears to be feasible. Longer 

38 hydrologic functionality may be feasible with the proper development and implementation of the 

39 elements provided in the previous sections, and with continued research. Existing uncertainty in 

40 long-term functionality could be reduced by additional technical bases, analyses, testing, and 

41 field experience. Geomembranes do not have the experience base of common natural

42 materials or a man-made material like cement, which has been used for hundreds to thousands 

43 of years. Therefore, until the experience base is developed, a cautious approach is needed for 

44 the long-term performance of novel materials in engineered barriers.

45

46 Strategies listed in NUREG/CP-0195 for minimizing the negative impacts of degradation 

47 processes include (1) paying careful attention to construction QA because it is especially 

48 important to the successful short-term performance of the cover, (2) identifying the processes 

49 that have the greatest impact on performance, (3) understanding the total system and planning 

50 accordingly, (4) analyzing each component within context, and (5) identifying and standardizing 
P-20

1 exposure scenarios that ultimately control the performance objectives and cover features 

2 related to these objectives. 

3

4 Unintended consequences must be considered when trying to diminish one form of degradation. 

5 The licensee must clearly determine the function of each cover component to diminish the risk 

6 of unintended consequences. For example, the use of plants to remove moisture from the 

7 barrier and inhibit movement of water down through the waste also results in enhanced radon 

8 transport upward through the barrier (NUREG/CP-0312 (Fuhrmann et al., 2018)). The intended 

9 function of each component must be clearly defined, in addition to the potential risks to the 

10 primary function. This will assist in delineating pathways for managing the total system 

11 performance.

12

13 Designers and analysts should recognize that soil properties may change quickly and therefore 

14 should minimize the consequences of these changes by designing and constructing covers for 

15 sites with long-lived radionuclides that mimic longer-term conditions that are congruent with 

16 nature. The resources required to maintain the engineered system must also be considered. 

17 Instruments can help identify the timing of a degradation process. There is a need to develop 

18 techniques that seek to understand magnitude and direction of natural changes anticipated to 

19 occur. One approach is to develop a catalog of natural analogs.

20

21 Evapotranspiration covers have been studied in the field on large scales, as discussed in 

22 Section P.3. These covers have performed well in arid and semiarid climates when they are 

23 well designed, constructed, and monitored. Because they attempt to mimic natural processes, 

24 the level of performance achieved over the long term would be expected to exceed that of 

25 resistive infiltration barriers. Whereas a conventional (resistive) cover is subject to failure by 

26 fast pathways (likely to be more discrete; either it is functioning, or it is not), an 

27 evapotranspiration cover is more likely to have different degrees of performance (likely to be 

28 more gradual, based on exceeding the design capacity for water management). Comparisons 

29 of evapotranspiration covers and resistive covers discussed in NUREG/CP-0312 show that 

30 evapotranspiration covers perform well in the field on the order of 10 years, and in theory, may 

31 have some passive performance over the very long term.

32 Erosion Protection Barriers

33 The effectiveness of engineered cover components developed for erosion control could exceed 

34 1,000 years. However, erosion control designs may not be adequate to preclude excessive 

35 infiltration. For example, deposition of windblown fines into openings between rocks on rip-rap 

36 covered sites may result in plants taking root and disrupting the clay barriers under the rock. 

37 Appendix Q contains a detailed example, including some of the design considerations.

38

39 Aside from the depth to waste, most engineered covers would not provide a substantial barrier 

to common practices assumed in intruder analysis (e.g., home construction2 40 , well installation). 

41 In some cases, the licensee could present arguments on the likelihood that a well driller would 

42 intrude into an engineered system comprised of large rock for erosion control or systems that 

43 rely on significant quantities of cementitious materials to isolate the radioactivity. For instance, 

44 drillers accustomed to drilling into unconsolidated materials may be alerted to a hazard if 

45 cementitious materials are discovered near the surface.

46 Permeable Reactive Barriers

2 Relatively thick covers may mitigate the impact of basement construction, with basements typically assumed to be 

dug to a depth from ground surface of 3 m, if erosion does not significantly decrease the cover thickness.
P-21

1 There has been limited (approximately 15 years) experience with permeable reactive barriers. 

2 Effective lifetimes of these barriers, from the literature, appear to be limited to less than 

3 10 years.

4

5 Vertical Barriers

6 As noted previously, it is difficult to identify flaws in the continuity and gaps in constructed 

7 vertical walls. In addition, some of these walls are not able to withstand chemicals such as 

8 strong acids, bases, and certain organic materials. Moreover, the process of designing a proper 

9 mix of wall materials to contain specific contaminants (hazardous chemicals and radionuclides) 

10 is less well developed. Accordingly, the effective service life of these structures would likely 

11 range in the low single digits of years and their performance should be demonstrated by field 

12 testing.

13

14 P.2 Example of a Graded Approach for Erosion Protection Covers

15 The following text provides a detailed example of the application of a graded approach to the 

16 design and implementation of engineered barriers for erosion protection. The graded approach 

17 can be readily used for the design of an engineered barrier, specifically a soil or rock cover that 

18 is provided for erosion control. This approach provides significant flexibility and can be adapted 

19 to design covers at a wide variety of sites with a wide range of waste inventories. 

20 The criteria provided in NUREG-1623, “Design of Erosion Protection Covers for Long-Term 

21 Stabilization,” issued September 2002, can be used for lower and higher risk decommissioning 

22 sites. NUREG-1623 has been used by NRC staff to review and approve erosion protection 

23 designs at over 40 different uranium mill tailings impoundments.

24 Erosion control covers for uranium mill tailings impoundments provide examples of design and 

25 construction of robust engineered barriers for long-term protection of radioactive materials. 

26 Requirements for engineered barriers at uranium mill tailings sites are set by the Uranium Mill 

27 Tailings Radiation Control Act (UMTRCA) of 1978, 42 U.S.C. § 7901 et seq.; the U.S. 

28 Environmental Protection Agency’s implementing regulations at 40 CFR Part 192, “Health and 

29 Environmental Protection Standards for Uranium and Thorium Mill Tailings;” and the NRC’s 

30 implementing regulations in 10 CFR Part 40, “Domestic Licensing of Source Material,” 

31 Appendix A, “Criteria Relating to the Operation of Uranium Mills and the Disposition of Tailings 

32 or Wastes Produced by the Extraction or Concentration of Source Material from Ores 

33 Processed Primarily for Their Source Material Content.” Under these regulations, erosion 

34 control covers are required to be designed to remain effective for up to 1,000 years without 

35 reliance on ongoing active maintenance. Although UMTRCA also requires State or DOE 

36 ownership and long-term care of the uranium mill tailings sites, including maintenance of the 

37 covers as needed, the covers are designed to function independently of maintenance. 

38 Therefore, the monitoring or surveillance and maintenance provided by DOE can be thought of 

39 as a backup to the robust design, or a defense-in-depth approach to long-term protection. Over 

40 25 years of experience is available, including the NRC’s guidance and technical basis for the 

41 design of robust erosion control covers and DOE’s construction of these covers and monitoring 

42 of their performance. This program offers an approach and lessons learned for one kind of 

43 robust engineered barrier that could have some application to the design of other types of 

44 robust barriers. Although the graded approach to engineered barriers in this decommissioning 

45 guidance offers greater flexibility to select appropriate options for engineered barriers (including 

46 reliance on maintenance), the UMTRCA covers offer an excellent example of a robust erosion 

47 control barrier to maintain stability for sites with a long-term hazard.
P-22

1 Robust engineered barriers for erosion control should be designed, but a graded approach to 

2 the design should be taken with respect to selection of the design flood, evaluation of rock 

3 durability, and selection of appropriate design factors. Each of these three areas is described 

4 below. 

5 P.2.1 Selection of Design Flood

6 One of the phenomena most likely to affect long-term stability is surface water erosion. To 

7 mitigate the potential effects of surface water erosion, the staff considers it very important to 

8 select an appropriate rainfall event on which to base the erosion protection designs. Further, 

9 the staff considers that the selection of a design flood event should not be based on the 

10 extrapolation of limited historical flood data, due to the unknown level of accuracy associated 

11 with such an extrapolation. The probable maximum precipitation (PMP) is computed by 

12 deterministic methods (rather than statistical methods) and is based on site-specific 

13 hydrometeorological characteristics. The PMP has been defined as the most severe reasonably 

14 possible rainfall event that could occur as a result of a combination of the most severe 

15 meteorological conditions occurring over a watershed. No recurrence interval is normally 

16 assigned to the PMP; however, the staff has concluded that the probability of such an event 

17 being equaled or exceeded during a 1,000-year stability period is very low. Accordingly, the 

18 NRC staff considers the PMP to provide an acceptable design basis.

19 The probable maximum flood (PMF) is based on the occurrence of the PMP and is considered 

20 to represent the most severe flood that can reasonably be expected to occur over a particular 

drainage basin.3 21 There is no probability assigned to the PMF, but the staff will generally not 

22 accept the use of statistically derived floods when the analysis time period significantly exceeds 

23 recorded data, due to the unreliable extrapolation of flood records based on short-term data. 

24 NUREG-1623 contains further discussion of the use of the PMP and PMF.

25 P.2.2 Rock Durability

26 Rock durability is defined as the ability of material to withstand the forces of weathering. 

27 Primary factors that affect rock durability are (1) chemical reactions with water, (2) saturation 

28 time, (3) temperature of the water, (4) scour by sediments, (5) windblown scour, (6) wetting and 

29 drying, and (7) freezing and thawing.

30 For rock to remain effective in controlling erosion, the rock size selected and emplaced should 

31 not be reduced in size by weathering processes. Therefore, if the rock size used for the cover 

32 does not diminish over the 1,000-year compliance time period, its ability to control future erosion 

33 will be sustained. However, uncertainties exist with estimating future rock durability. For 

34 example, quantitative studies of weathering rates of different rock types and minerals are 

35 limited, in general, and not expected to be available for specific rock sources that might be 

36 selected by licensees. As a result, three evaluations of rock durability should be conducted to 

37 provide multiple and complementary lines of evidence and greater confidence in the future 

38 durability of the rock source selected. These evaluations are (1) rock durability testing and 

39 scoring, (2) absence of adverse minerals and heterogeneities, and (3) evidence of resistance to 

40 weathering. A description of each of these evaluations follows. 

3 Although consideration of the PMP is expected to provide a bounding analysis for the PMF in most cases, if other 

features, events or processes (FEPs) could lead to greater impacts, then those FEPs should also be considered. 
P-23

1 P.2.3 Rock Durability Testing and Scoring

2 NUREG-1623 includes a procedure for determining the acceptability of a rock source, which the 

3 NRC staff developed and used for selecting durable rock for erosion covers for uranium mill 

4 tailing sites. This procedure provides a consistent and quantitative way to evaluate rock 

5 sources at NRC-licensed sites using standard parameters that are good indicators of rock 

6 durability. Test samples of the selected rock source should be representative of the specific 

7 rock expected to be used. The number and location of the samples should be determined 

8 based on the variability of the rock source, such as, texture and mineralogy that could affect the 

9 individual test results. The test procedure generally includes the following: 

10 • Test results of four parameters (specific gravity, absorption, sodium sulfate, and Los

11 Angeles abrasion) from representative samples are scored on a scale of 0–10.

12 • The score is multiplied by a weighting factor, which focuses the scoring on those tests

13 that are the most applicable for the particular rock type being tested.

14 • The weighted scores are totaled, divided by the maximum possible score, and multiplied

15 by 100 to determine the rating.

16 • The rock quality scores are then compared to the criteria that are measures of

17 acceptability.

18 After these tests are conducted, an overall rock quality score is determined. Rock scores of 

19 80 percent or greater indicate high-quality rock that can be used for most applications. Rock 

20 scores between 65 percent and 80 percent indicate less durable rock that can also be used for 

21 most applications. Rock scoring less than 65 percent cannot be used for critical areas, such as 

22 diversion ditches or poorly drained toes and aprons. Rock scoring between 50 percent and 

23 65 percent can be used in noncritical areas, such as well-drained side slopes, provided the rock

24 is oversized. Rock scoring less than 50 percent is not recommended for use in any application. 

25 NUREG-1623 contains additional discussion of specific tests and the scoring procedures. For 

26 example, Table F-1 in NUREG-1623 lists a number of American Society for Testing and 

27 Materials (ASTM) standards which should be used for quality and gradation tests.

28 P.2.4 Absence of Adverse Minerals and Heterogeneities

29 Licensees should provide results of petrographic analyses of the selected rock source and 

30 available published data that support the absence of adverse minerals that could cause rapid 

31 degradation of the rock, such as clays, olivine, or calcite cement. If adverse minerals are 

32 present, they should evaluate the potential effect on future weathering of the selected rock. 

33 Licensees should place particular emphasis on selecting rocks that do not have appreciable 

34 clay content or do not contain minerals that could rapidly weather to clays. The staff examined 

35 the causes of rock durability failures in typical applications, such as placement on dam slopes or 

36 stream channels. In most cases, the staff determined that durability failures were caused by the 

37 presence of expanding clay-lattice minerals, which, when exposed to moisture or freeze and 

38 thaw cycles, caused the rapid deterioration of the rock. 

39 Licensees should identify either the absence or presence of heterogeneities that could 

40 adversely affect the selected rock source, such as clay or shale partings, interbeds, fractures, 

41 alteration zones, or vein deposits. Such heterogeneities can be zones of water flow and 
P-24

1 associated chemical alteration or zones of breakage due to freeze and thaw. Licensees should 

2 characterize heterogeneities present and discuss the ability to avoid these adverse 

3 heterogeneities when removing the selected rock for use. They should also evaluate how 

4 heterogeneities could affect achieving the acceptable size of the selected rock. Breakage along 

5 thin interbeds or fractures during rock quarrying, transport, or emplacement could result in 

6 reduction of the rock size necessary for erosion protection. 

7 P.2.5 Evidence of Resistance to Weathering

8 Licensees should provide direct evidence from the selected rock source, such as minerals that 

9 are resistant to weathering, resistant cements, and regional or local geomorphic evidence of 

10 slow weathering of the selected rock that outcrops in other locations. Examples include 

11 rounded boulders or thin weathering rinds. Weathering rind thickness and alteration of minerals 

12 and rock properties from exposures of the weathered selected rock source could provide 

13 insights on the extent and nature of future weathering of the selected rock source from fresh

14 quarry exposures. Identify and describe any available studies of weathering rates of the 

15 selected rock source. 

16 Licensees can also use indirect evidence from other locations where the general rock type is 

17 similar to the selected rock source. For example, evidence of durability from a diabase igneous 

18 rock found in Europe could be used to provide insights on a diabase rock source in Maryland, 

19 because the general mineralogy of diabase is similar, regardless of the location. This approach 

20 allows the use of datable natural or archaeological/historical rock sites (called analog sites) that 

21 could provide general evidence on rock weathering rates or time periods during which rock 

22 types have remained resistant to weathering. For example, as mentioned in Section P.3.3, 

23 numerous datable archaeological sites, such as Stonehenge (constructed about 4,000 years 

24 ago of diabase and silica cemented sandstone), Hadrian's Wall (constructed by the Romans 

25 over 2,000 years ago of primarily diabase), and numerous buildings, monuments, and megaliths 

26 in Europe, could be used to demonstrate that these rock types have been resistant to 

27 weathering over time periods that exceed the 1,000-year period of regulatory compliance for a 

28 robust erosion cover. Historical evidence can also provide useful insights on the durability of 

29 certain rock types. One example is the comparison of dated Civil War photographs of diabase 

30 outcrops in Devil's Den at the Gettysburg Battle Field Park to present-day conditions of the 

31 same outcrop. Such a comparison indicates that this diabase has been resistant to weathering 

32 for about 150 years. Similarly, dated grave markers or historical buildings made from the 

33 selected rock source or similar rock type can provide evidence of resistance to weathering for 

34 100–200 years. Appendix A of NUREG/CR-2642, “Appendix to Long-Term Survivability of 

35 Riprap for Armoring Uranium Mill Tailings and Covers: A Literature Review,” 1982, provides 

36 additional information on rock weathering, durability, and examples of analogs that contain 

37 insights on general weathering rates of various rock types.

38 As previously noted, natural rock sites, such as age-dated glacial erratic, could also be used to 

39 show that the particular rock type has been resistant to weathering for over 10,000 years. 

40 Examples that illustrate this approach include age-dated diabase erratics from New York and 

41 North Dakota that are over 10,000 years old, and river scour features in diabase that have been 

42 preserved for about 20,000 years in Pennsylvania. 

43 For each of the above evaluations, the licensee should provide the appropriate supporting data, 

44 information, and references and identify the location of the selected rock source, along with 

45 regional and local rock source descriptions.
P-25

1 P.2.6 Selection of Appropriate Design Factors

2 In selecting appropriate input parameters for calculating erosion protection size and thickness, it 

3 is important to choose values that reflect the degree of risk and the importance of the rock layer, 

4 as it contributes to overall stability. However, the selection of many input parameters to various 

5 models can sometimes be subjective and will need to be based largely on engineering 

6 judgment. Where there are large ranges in values, or where a parameter cannot be well 

7 defined, or is not well known, it has been the general policy of the NRC staff to accept the use of 

8 reasonable ranges and distributions of input parameters. For well-known or accepted 

9 parameters with narrow empirical distributions or very narrow ranges, expected values should 

10 be used as appropriate. For less well-known parameters, such as those based on little 

11 empirical data or with broad distributions, conservative values should be chosen from within the 

12 observed distributions or estimated range. In any case, there should a reasonable and 

13 defensible technical basis for the choice of a design-basis event or design criteria, and the staff 

14 will accept values that can be justified. Otherwise, reasonably conservative values will be 

15 needed.

16 P.2.7 Erosion Cover Analysis Process

17 At a lower risk site, where engineered barriers are needed to meet applicable requirements for 

18 only about 100 years and there is a lower hazard level should institutional controls and 

19 maintenance fail (up to the public dose limit of 1.0 mSv/y (100 mrem/y)), the principal design 

20 basis and goal is to ensure the relative stability of the contaminated material by providing a 

21 cover design that maintains control of the material. Control is achieved by providing a relatively 

22 robust design that prevents offsite movement (e.g., erosion by natural forces) of the material. 

23 The rock erosion protection barrier could protect a second layer of material; for example, one 

24 that might be a radiation shielding barrier or an infiltration barrier, depending on the 

25 radionuclides at the site and natural processes important to achieving compliance with the dose 

26 criteria. 

27

28 The design should be able to survive the occurrence of relatively rare events, and the erosion 

29 protection should be sufficiently robust to remain effective for about 100 years. Using the 

30 guidance and rationale contained in NUREG-1623 for a 100-year stability period, the barriers 

31 should be designed to resist a flood equivalent to either the regional historic flood of record or 

32 about half of the PMF, whichever is greater. The licensee should provide rainfall, flood, and 

33 erosion analyses that justify the design. A design that meets the suggested flooding and 

34 erosion protection criteria of NUREG-1623 is acceptable. The rock itself should be sufficiently 

35 durable to remain effective for at least 100 years by obtaining a rock quality score of at least 65. 

36 The computations and selection of input parameters should rely on reasonable and justified 

37 estimates.

38

39 For a higher risk site, with long-lived radionuclides, or where failure of institutional controls and 

40 maintenance could result in a higher hazard of 1.0–5.0 mSv/y (100–500 mrem/y), the principal 

41 design basis and goal is to ensure the control and stability of the contaminated material by 

42 providing a robust design that will remain effective for a period of 1,000 years or more by 

43 preventing erosion by natural forces. These covers should be designed to maintain control and 

44 stability with no reliance on active maintenance. However, monitoring and maintenance will be 

45 conducted as a backup to provide defense in depth. 

46 The staff could approve an engineered barrier design that is effective and maintains control of 

47 the material for a period exceeding 1,000 years. Using the guidance and rationale contained in 
P-26

1 NUREG-1623, the barriers should be designed to resist severe localized rainfall events and 

2 large floods on nearby streams. The design rainfall event should be the PMP, and the design 

3 flood should be the PMF. A design that meets the suggested flooding and erosion protection 

4 criteria of NUREG-1623 is acceptable. The rock quality score should be at least 85, and the 

5 selection of input parameters to various models should account for the unknowns associated 

6 with a very long stability period and the high-risk site.

7 For sites like this, if erosion is a significant issue and there are some uncertainties associated 

8 with the magnitude of this erosion, the staff will approve a design that would likely incorporate 

9 (1) covers designed to resist erosion for a stability period exceeding 1,000 years, (2) a long-term

10 surveillance program that monitors the magnitude and rate of erosion, and (3) sufficient funding

11 for the surveillance, repair, and replacement of some of the erosion protection. The staff will

12 work closely with the expected long-term custodian to determine the amount of funding needed.

13 It is important to reiterate that the requirements of 10 CFR Part 40 are very prescriptive and may 

14 have precluded the use of many types of erosion protection designs. The design criteria 

15 suggested in NUREG-1623 may be used at decommissioning sites using approaches different 

16 from those used in uranium mill tailings applications. For example, nearly all tailings sites were 

17 designed with disposal cell side slopes of about 1 vertical (V) on 5 horizontal (H). Based on the 

18 stability of erosion protection placed on much steeper slopes of stream channels, levees, and 

19 dam slopes, there is evidence that may support the use of slopes steeper than 1V on 5H for the 

20 side slopes of disposal cells. The criteria in NUREG-1623 were not developed for use on 

21 specific slopes and may be adapted for steeper side slopes, as necessary. However, minor 

22 changes to construction specifications may need to be considered for steeper slopes, such as

23 emphasizing careful rock placement,. 

24 Table P.2 summarizes the application of the graded approach for the design of erosion 

25 protection for the sites discussed above. 

26 Table P.2 Summary of the Graded Approach for the Design of Erosion Protection 

27 Systems 

Level of Risk Flood Design Basis Rock Durability Score

Confidence in Selection of 

Input Parameters

Lower 1⁄2 PMP / 1⁄2 PMF > 65 Reasonable

Higher PMP/ PMF > 85 Very High

28

29 P.2.8 Technical Basis for Design and Performance of Erosion Protection Covers

30 The use of the guidance presented in this document and in NUREG-1623 will result in designs 

31 that provide acceptable long-term erosion protection at decommissioning sites. These 

32 documents contain design criteria that incorporate a strong technical basis, including (1) use of 

33 NRC experience and lessons learned at various sites to develop and implement erosion 

34 protection guidance, (2) use of appropriately conservative design bases and computational 

35 procedures for erosion protection designs, and (3) an extensive archaeological and natural 

36 basis for long-term stability.

37
P-27

1 P.2.8.1 NRC Experience and Lessons Learned

2 The NRC staff has over 20 years of experience in the design and review of erosion protection 

3 covers. This experience includes activities associated with (1) development of design guidance 

4 for uranium mill tailings reclamation, including use of technical studies sponsored by the NRC to 

5 address specific design problems, (2) review of reclamation plans and DPs at about 50 sites, 

6 including review of specific problem sites where challenging erosion problems were addressed, 

7 and (3) review and inspections of construction problems and deficiencies. The staff has applied 

8 this experience in its review of uranium mill reclamation plans and DPs on numerous occasions.

9 P.2.8.2 Guidance Development

10 UMTRCA regulations established design standards to be met at uranium mill sites for Title I 

11 (inactive sites) and Title II (licensed active sites), as required by UMTRCA. Specifically, the 

12 design standard for long-term stability was established to be 1,000 years to the extent 

13 reasonably achievable, or in any case at least 200 years. When the regulations were 

14 developed, there was very little experience associated with providing designs that would remain 

15 effective for such long periods of time. To address this problem, the NRC staff worked closely 

16 with DOE and various contractors in the 1980s to establish design criteria and guidance for 

17 long-term stabilization. The NRC held several joint meetings with DOE and DOE contractors to 

18 formulate guidance and design procedures to address the long-term stability of erosion 

19 protection covers. These procedures were published in various technical and construction 

20 documents that were developed by DOE, DOE contractors, and the NRC staff.

21 During the development of design guidance, it was recognized that some of the procedures 

22 normally used for the design of erosion protection were not necessarily appropriate for long23 term stabilization or for conditions that would be encountered at various sites. To address such 

24 issues, the NRC sponsored extensive technical studies. These studies were conducted in the 

25 1990s at Colorado State University and included prototype flume studies to determine rock 

26 sizing procedures for overland flows; flume studies to determine rock sizing and volume 

27 requirements for aprons and toes of slopes; gully studies to determine rates, magnitude, and 

28 location of gully development on reclaimed slopes; and rock durability studies to address rock 

29 weathering rates and tests needed to ensure adequate rock quality. The results of the studies 

30 were published in nationally recognized journals and peer reviewed by experts in the field of 

31 erosion protection design. One widely used example includes detailed guidance for the design 

32 of riprap for flood flows that would be expected down the side slope of a disposal cell (S.R. Abt 

33 and T.L. Johnson, “Riprap Design for Overtopping Flow,” American Society of Civil Engineers, 

34 Journal of Hydraulic Engineering, 1991 (Abt and Johnson, 1991)). NUREG-1623 includes 

35 results from this and other publications.

36 P.2.8.3 NRC Staff Review of Challenging Sites

37 Vegetated soil covers, rock covers, and composite covers (soil and rock) were used extensively 

38 in the reclamation and stabilization of all uranium mill tailings sites. In general, most of these 

39 sites presented no significant erosion problems, and DOE and several NRC licensees had a 

40 great deal of success in designing and constructing these covers. However, several sites 

41 presented challenging erosion problems. Table P.3 provides examples of several significant 

42 erosion problems that were solved with engineered barriers, using design criteria in 

43 NUREG-1623.

44
P-28

1 Table P.3 Examples of Engineered Barriers Used to Solve Erosion Problems 

Site Erosion Problem Solution to Problem

Maybell 

UMTRCA Site

Downstream gullies that could 

scour to a depth of about 20 feet 

could encroach on the disposal 

cell. Unstable local base levels 

could cause further increase in 

the potential for gullying.

DOE provided extensive rock aprons to 

check erosion advance to the disposal 

cell, using criteria for scour depth and 

rock sizing provided in NUREG-1623.

Grand Junction 

UMTRCA Site

Deep gullies with relatively large 

drainage areas existed in the 

approximate center of the 

proposed disposal cell. 

Hydraulically steep slopes were 

present near the toe of the cell.

DOE provided diversion channels and 

channel outlets with very large riprap to 

safely convey flows around the disposal 

cell, using criteria found in 

NUREG-1623.

Rifle (Estes 

Gulch) 

UMTRCA Site

The disposal cell was excavated 

in a steep gully located in the 

center of the site, requiring 

diversion of flood flows.

An upstream diversion channel was 

constructed to convey flows around the 

cell into a different drainage basin.

Atlas Title II Site The disposal cell is immediately 

adjacent to the Colorado River, 

with a drainage area of 

thousands of square miles. 

There is a potential for the river 

to erode and migrate towards 

the cell, where high river 

velocities could impinge on the 

cell.

The licensee provided a design to resist 

the erosional forces associated with 

river migration, proposing to construct a 

large rock apron in accordance with the 

suggested criteria of NUREG-1623. (It 

should be noted that a decision was 

later made to completely move the 

entire tailings pile.)

2

3 A large amount of experience was gained at various sites where erosion was found to be a 

4 significant problem. The staff analyzed and became familiar with various design solutions to 

5 mitigate significant erosion problems. Based on this experience, the staff considers that the 

6 erosion protection criteria suggested in NUREG-1623, combined with the graded regulatory 

7 approach discussed in this example, will provide a significant degree of flexibility in solving 

8 difficult erosion problems at complex sites.

9

10 P.2.8.4 NRC Staff Inspections and Reviews of Construction Deficiencies

11 For the last 15–20 years, NRC staff has routinely inspected sites in the Title I and Title II 

12 programs. The inspections included evaluations of soil cover and rock cover placement during 

13 construction and final closeout inspections of the completed work. 

14 During these inspections, the staff noted that adequate placement of rock riprap layers was 

15 difficult for many contractors to achieve. There were numerous instances where rock layers 

16 were not placed to correct and consistent design thicknesses, in-place riprap layers did not have 

17 correct gradations and varied from tested samples, and areas of segregation where rock sizes 

18 were much smaller than required over large areas of the disposal cell. In many cases, the staff 

19 required corrective actions. Based on these problems encountered in staff reviews and its 
P-29

1 experience with corrective actions, the staff developed guidance for proper rock placement. 

2 The staff used its construction experience to develop suggested quality control procedures for 

3 use by contractors. Appendix D to NUREG-1623 contains the suggested guidance.

4 At the present time, the staff also routinely accompanies DOE and/or DOE contractors on 

5 annual inspections at some sites where DOE now has custody and licensed responsibilities. 

6 These inspections have generally shown that covers are performing rather well and regulatory 

7 requirements continue to be met in all cases.

8 The staff also has experience with rock durability problems at uranium mill tailings reclamation 

9 sites. These problems occurred, for example, at sites using rock sources consisting of minerals 

10 susceptible to weathering (such as olivine basalts) or sources that had undergone hydrothermal 

11 alteration. In addition, the staff reviewed information on 149 case histories (E.E. Esmiol, (“Rock 

12 as Upstream Slope Protection for Earth Dams—149 Case Histories,” issued September 1967 

13 (Esmiol, 1967)), associated with rock durability problems at facilities constructed by other 

14 Federal agencies (such as Bureau of Reclamation dam sites). The NRC developed the rock 

15 durability criteria suggested in NUREG-1623 using the lessons learned from both uranium mill 

16 sites and other Federal sites. 

17 P.2.8.5 Use of Design Bases Appropriate for Long Stability Periods

18 As discussed in previous sections, erosion protection designs that meet the suggested criteria 

19 and guidance in this document and NUREG-1623 will provide adequate protection against 

20 extreme erosion events that could occur over the period of regulatory interest. The guidance 

21 reflects staff review and construction experience, past practices with regard to the selection of 

22 design bases, and good engineering practices. The staff considers that it has provided 

23 appropriate guidance on (1) selection of conservative rainfall and flooding events that reflect the 

24 long stability periods needed to meet regulatory requirements, (2) selection of parameters for 

25 determining flood discharges that account for the uncertainties associated with flood 

26 calculations, (3) computation of flood discharges using appropriate and/or conservative 

27 methods, (4) computation of appropriate flood levels and flood forces associated with the design 

28 discharge, (5) use of widely accepted, state-of-the-art, and standardized methods for 

29 determining erosion protection sizes and thicknesses, (6) selection of a rock type for the riprap 

30 layer that will be durable and maintain its size and ability to provide protection for a long period 

31 of time, and (7) placement of riprap layers in accordance with accepted engineering practice 

32 and in accordance with appropriate testing and QA controls.

33

34 P.2.8.6 Archaeological and Natural Bases for Long-Term Stability

35 A strong archaeological basis exists to demonstrate the long-term stability of erosion protection 

36 materials. NUREG/CR-2642 presents substantial information to demonstrate the long-term 

37 survivability of various rock structures. NUREG-1623 provides information on long-term 

38 weathering rates, based on observations of rock petroglyphs that could be dated to a period of 

39 nearly 1,000 years before present. Further, as discussed in Section P.1.3 of this document, 

40 observations of Native American burial mounds in West Virginia, Ohio, Illinois, and Louisiana 

41 illustrate the survivability of man-made earthen structures for long periods of time (1,000 to 

42 5,500 years).

43

44 Numerous other archaeological sites, such as Stonehenge (constructed about 4,000 years ago 

45 of diabase and silica cemented sandstone), Hadrian's Wall (constructed by the Romans over 

46 2,000 years ago of primarily diabase), and numerous buildings in Europe demonstrate that 
P-30

1 certain rock types used in these structures have been resistant to weathering over time periods 

2 greater than the 1,000-year period of regulatory compliance. In addition to archeological and 

3 historical evidence, natural rock sites can offer further insights that certain rock types and 

4 sources have been resistant to weathering. For example, dating of glacial erratics (rocks 

5 transported by glaciers) and preservation of river scour features in diabase results from high 

6 flow during glacial melting demonstrate that these rocks have been resistant to weathering and 

7 preserved for over 10,000 years.

8

9 The natural and archaeological insights noted above serve to demonstrate that engineered 

10 structures and construction materials can survive for very long periods of time. The intent of this 

11 guidance is to develop procedures for improving long-term stability by further enhancing design 

12 concepts where structures have generally remained intact for many years.

13

14 P.2.9 Degradation Mechanisms for Erosion Control Covers

15 The erosion control cover at a typical decommissioning site could consist of either a rock layer 

16 or a soil layer and underlying rock layer. One of the most likely degradation mechanisms would 

17 be gully erosion. To account for that process, the design consideration that should be analyzed 

18 (and is considered by the staff to be the most likely) is the formation of a gully in the top soil 

19 cover, caused by surface erosion, flow concentrations, and/or the uprooting of large trees. The 

20 erosion should be assumed to continue and be deep enough to expose the rock layer, and thus 

21 the rock layer would need to be designed to resist further erosion and down-cutting of the gully. 

22 The licensee should design the soil cover, as described above, to be stable for rainfall events 

23 and runoff as large as the PMP/PMF. Further, the rock layer should be designed as a separate 

24 backup system for the soil cover and should also be designed for the PMP/PMF occurring in 

25 that gully, with flow concentrations produced by the growth of a drainage network to the gully. 

26 Further, the rock should meet the durability criteria suggested in NUREG-1623, with particular 

27 emphasis placed on the petrographic examination that indicates that no clay minerals are 

28 present (see Table P.4).

29

30 P.3 Summary of Existing Guidance and Reference Information 

31 Table P.4 summarizes existing guidance and reference information that may have some 

32 relevance to the application of engineered barriers at decommissioning sites. Early contact with 

33 the NRC staff is encouraged to discuss which portions of these referenced reports may be 

34 appropriate for the site and for the intended purpose of the engineered barriers. NUREG-1620, 

35 “Standard Review Plan for the Review of a Reclamation Plan for Mill Tailings Sites Under Title II 

36 of the Uranium Mill Tailings Radiation Control Act, Draft,” Revision 1, issued June 2003, and 

37 NUREG-1623 contain guidance on the design of engineered disposal cells for uranium mill 

38 tailings sites. For sites considering engineered disposal cells for long-term stability 

39 (e.g., erosion control), this guidance may be somewhat useful. However, the standards in 

40 10 CFR Part 40, Appendix A, applicable to uranium mills, are more prescriptive than the 

41 performance-based dose criteria of 10 CFR Part 20, Subpart E. Licensees using the uranium 

42 mill guidance should also consider how the guidance can be adapted for applicability to 

43 compliance with 10 CFR Part 20, Subpart E. As discussed below, in some cases, traditional 

44 designs have not performed particularly well with respect to infiltration.

45

46 A variety of programs have evaluated and continue to evaluate engineered barrier technology 

47 for waste containment. A comprehensive summary of engineered barrier research is not 

48 attempted in this document. However, some good examples of programs to understand, 
P-31

1 design, and support engineered barrier performance applicable to decommissioning sites 

2 include the following:

3

4 • the U.S. Environmental Protection Agency (EPA), through the Alternative Cover

5 Assessment Program (ACAP), has supported the field-scale evaluation of engineered

6 covers

7 • the U.S. Department of Energy (DOE), through the Alternative Landfill Cover

8 Demonstration (ALCD) completed a large-scale field demonstration comparing six

9 landfill cover designs

10 • DOE has instrumented engineered covers at some Uranium Mill Tailings Remedial

11 Action (UMTRA) sites to understand and evaluate their performance

12 • the University of Wisconsin conducted follow-up tests using exhumed ACAP covers

13 under an NRC contract with the U.S. Geological Survey (USGS). The findings of the

14 research project are documented in NUREG/CR-7028

15 • NUREG/CP-0195 summarizes experience with engineered barriers and

16 recommendations from Federal, State, and tribal agencies, and academic and industry

17 experts

18 • summarizes results of studies of barriers at four in-service Uranium Mill Tailings

19 Radiation Control Act (UMTRCA) sites, focusing on hydraulic conductivity and radon flux

20 EPA’s ACAP evaluated 27 test covers at 12 sites in eight states to characterize the field 

21 hydrology of water balance and conventional covers (Albright et al., 2004, NUREG/CR-7028). 

22 The evaluation included 12 conventional covers (7 composite and 5 clay) and 15 water balance 

23 covers (9 monolithic and 6 capillary barriers). Nine of the sites had side-by-side comparisons of 

24 conventional and alternative covers. Large-scale lysimeters (approximately 10 m by 20 m (30 ft 

25 by 70 ft) areal extent) were installed and instrumented to collect detailed water balance 

26 information. The main lessons learned for water balance covers were that (1) percolation rates 

27 in semiarid and sub-humid climates can be very low (less than 1 mm/y), provided that there is 

28 adequate storage capacity and that the vegetation effectively removes stored water each year, 

29 (2) there is a need to better understand the phenology of plants and the response to

30 meteorological and geotechnical conditions, and (3) low percolation rates may not be achieved

31 at sites with water balance covers, in particular at humid sites, but the water balance covers

32 may still provide some performance benefit. The main lessons learned for composite covers

33 were that (1) composite covers may be effective at limiting percolation to less than 1 mm/y while

34 the geomembranes or geosynthetics are intact, and (2) clay covers are prone to damage over

35 very short periods of time and can transmit percolation at much higher rates than anticipated.

36 The ACAP program provides an excellent example of developing the technical basis for

37 engineered barrier performance.

38

39 DOE’s ALCD evaluated six cover designs at Sandia National Laboratory in Albuquerque, NM, to 

40 obtain large-scale water balance field data subjected to identical field and climatic conditions 

41 (Dwyer, “Water Balance Measurements and Computer Simulations of Landfill Covers,” issued 

42 2003). The study included a Resource Conservation and Recovery Act (RCRA) Subtitle D 

43 cover, a geosynthetic clay layer cover, a RCRA Subtitle C cover, an anisotropic barrier cover, a 

44 capillary barrier cover, and an evapotranspiration cover. The RCRA Subtitle D cover had the 

45 highest percolation rate (above the 1 mm/y goal), and the geosynthetic clay layer cover had the 
P-32

1 second highest average percolation rate. Various damage processes, such as desiccation 

2 cracking, led to preferential flow through the RCRA Subtitle D cover. The field data from this 

3 project were interpreted to suggest the geosynthetic clay layer cover performance was suspect 

4 in this application, potentially due to desiccation and ion exchange. The other four cover types 

5 all had average annual fluxes of less than 0.2 mm/y (0.01 in/y).

6

7 Disposal cell covers for uranium mill tailings wastes at UMTRCA sites have been developed 

8 over the past 20 years at a variety of sites with different climates. From an erosion control and 

9 stability perspective, these covers have required little to no maintenance to prevent erosional 

10 release of radioactive materials although there have been problems at some sites with localized 

11 erosion events, such as slumping and ponding of water on the cover. At some covers, DOE has 

12 removed vegetation; however, these actions have primarily been undertaken because of 

13 concerns with the impact of the vegetation on water management (e.g., infiltration). Recent 

14 work shows that the presence of large deep-rooted plants both on rip-rap and vegetated covers, 

15 can result in increased releases of radon due to drying of the radon barrier while nearby 

16 unvegetated control areas have low radon fluxes (NUREG/CP-0312). DOE has instrumented 

17 some of the covers to understand and evaluate their performance. In particular, some covers 

18 based on resistive-type designs (e.g., impermeable layers), similar to unlined RCRA Subtitle D 

19 covers, have not achieved the design values for hydraulic conductivity measured in the 

20 laboratory and therefore appear to have much higher infiltration rates than anticipated (Waugh, 

21 “Design, Performance, and Sustainability of Engineered Covers for Uranium Mill Tailings,” 

22 issued 2004). In fact, at a number of sites, the in situ hydraulic conductivity was measured to be 

23 more than two orders of magnitude higher than the design target. However, monitoring data 

24 from evapotranspiration covers suggest that design infiltration rates have been achieved. One 

25 of the lessons learned was that seemingly subtle differences in soil types, sources, and 

26 compaction can result in significant differences in performance.

27

28 A number of the covers that were part of the ACAP study were exhumed and studied following 

29 termination of the program. The University of Wisconsin conducted field and laboratory tests 

30 under an NRC contract with USGS to study the change in engineered properties of the cover 

31 materials during their 5- to 10-year service life. NUREG/CR-7028 documents the findings of the 

32 research project. An important conclusion of the report is that compacted soil materials used in 

33 cover materials at the sites studied did not retain “as built” properties over periods of regulatory 

34 interest. The properties of these materials change to values more typical of surrounding soils 

35 within 5 to 10 years after installation. Changes in low permeable cover soils can be rapid and 

36 can result in an increase to the saturated hydraulic conductivity by three to four orders of 

37 magnitude. With respect to drainage layers, greater reductions in transmissivity and 

38 permeability were observed for drainage layers covered with soils having higher fines content. 

39 However, this effect was modest, and all of the drainage layers functioned as anticipated. It is 

40 important to note that the report did not address cover elements designed for erosion protection. 

41

42 With regard to covers that use composite materials, NUREG/CR-7028 found that these types of 

43 covers do not appear to significantly degrade over the short term. An analysis in the report 

showed that GCLs have very low saturated hydraulic conductivity (less than 5x10 44 -11 m/s) when 

45 placed on a moist subgrade (water content greater than 10 percent) and covered with a 

46 geomembrane and cover soil soon after installation, although GCLs installed under other 

47 conditions can be much more permeable. In addition, changes in geomembranes and 

48 geosynthetic drainage layers during the short period of the study were modest or small. GCLs 

49 that underwent complete hydration maintained low hydraulic conductivity, even when the native 

50 sodium (Na) was replaced by calcium (Ca) and magnesium (Mg). However, GCLs installed 

51 under other conditions were much more permeable; therefore, QA during cover construction is 
P-33

1 of great importance. Analysis of antioxidants in the geomembranes showed that antioxidant 

2 depletion was consistent with expectations based on first-order kinetics and laboratory3 measured depletion rates. Based on these rates, the minimum service life of geomembranes 

4 was estimated to be on the order of 50–125 years. The report concluded that actual service 

5 lives are likely to be longer but are difficult to estimate. Because covers change over time, it 

6 was recommended that they be monitored to ensure that they are functioning as intended. 

7 Recommendations included monitoring using large pan lysimeters, combined with secondary 

8 measurements collected for interpretive purposes (e.g., water content, temperature, vegetation 

9 surveys), as well as studying analogs of natural environments.

10

11 Given the risk-significant findings of NUREG/CR-7028, the NRC staff review assessed the risk 

12 associated with various licensed facilities, including a single decommissioning site with an 

13 operating cover. The risks associated with underperformance of an engineered cover is related 

14 to a variety of factors, including climate, evapotranspiration rates, hydrogeological properties of 

15 vadose zone materials, depth to groundwater, thickness of overburden materials, depth of 

16 resistive barriers (with shallower covers potentially more susceptible to surface degradation 

17 processes), existence of a liner and leachate collection system, and the extent to which 

18 groundwater is used as a resource. The NRC staff recommendations focus on uranium 

19 recovery facilities and include continued monitoring; data collection; stakeholder 

20 communication; and research, including field studies to continue to evaluate degradation 

21 mechanisms and to study the uncertainty associated with long-term performance of engineered 

22 covers. 

23

24 In August 2010, the NRC sponsored a workshop at its headquarters office in Rockville, MD 

25 (NUREG/CP-0195). The purpose was to provide Federal, State, and tribal agencies, and other 

26 interested stakeholders a forum to share information on engineered surface covers and bottom 

27 liners used for waste containment at low-level waste (LLW) disposal facilities, decommissioning 

28 sites, and uranium mill tailings sites. Stated objectives of the workshop included (1) facilitation 

29 of communications among Federal and State staff and contractors and selected experts on 

30 current engineered barrier issues and technical and regulatory experiences, (2) discussion of

31 lessons learned and approaches for monitoring and modeling, (3) preparation of 

32 recommendations to address maintenance of engineered barrier performance over time, and 

33 (4) identification of topics for future research and the potential need to update technical

34 guidance. At the conclusion of the workshop, it was noted that consistent themes repeatedly

35 mentioned by the workshop participants were (1) the need to more actively monitor and quantify

36 engineered system behavior so as to know the record of performance and better understand the

37 key processes effecting performance, and (2) the need for better communication and exchange

38 between the specialists of various fields, (e.g., engineers, pedologists, ecologists, modelers)

39 involved with the design, construction, and maintenance of engineered barriers. These insights

40 were supported by technical presentation details and discussions by all engaged participants

41 and were repeated throughout the 3-day workshop.

42

43 Following the August 2010 workshop additional studies were conducted to address workshop 

44 recommendations. For example, NUREG/CR-7200 investigates the coupling of erosion and 

45 hydrology and its impact on the performance of waste covers. Hydrologic and landform 

46 evolution models were used to evaluate various engineered cover designs for LLW disposal 

47 with respect to the ability of the cover designs to limit erosion and percolation of water through 

48 the waste. Climate and vegetation were found to have the most influence on erosion with 

49 erosion depths in semi-arid climates being 4 m greater compared to humid climates for 

50 simulations with a rip-rap or gravel admixture surface. Vegetation was also found to decrease 

51 erosion and percolation through the cover. Short slope, low grade, and small grade differences 
P-34

1 (for semi-arid climates) or terraced (for humid climates) slopes were found to be optimal with 

2 respect to limiting erosion. Although rip-rap surface materials decreased erosion, use of rip-rap 

3 led to higher percolation through the cover. In contrast, a gravel admixture surface had slightly 

4 greater erosion but was much were effective in limiting percolation through the cover.

5

6 A workshop was held at NRC headquarters on July 25-26, 2018, to present and discuss findings 

7 from a research project in which the performance of radon barriers at UMTRCA sites was 

8 studied after having been in service for about 20 years. Four mill tailing sites were visited by the 

9 research team: Falls City in Texas, Bluewater in New Mexico, Shirley Basin South in Wyoming, 

10 and Lakeview in Oregon. Small areas on these sites were excavated, radon fluxes were 

11 measured, numerous observations were made, and samples taken for a variety of parameters 

12 such as root counts, moisture, density, Pb-210 concentrations, hydraulic conductivity, soil 

13 texture, chemistry, and nematode counts. As part of the project, a comparison of the current 

14 state of these engineered covers to their as-built condition was made. In some cases, the 

15 saturated hydraulic conductivity and radon flux of the covers increased over time to higher 

16 values closer to values observed in natural analogs. In one case, the saturated hydraulic 

17 conductivity of the cover remained fairly stable over the time period of the study (approximately 

18 20 years), likely due to the high percentage of expandable clays in the cover material. Biotic 

19 activity, especially in the form of root intrusion into the radon barrier, was found to have a 

20 significant deleterious impact on radon barrier performance. The conference proceedings 

21 NUREG/CP-0312 contains extended abstracts and slides from the presentations, and a 

22 summary of the results and recommendations from the workshop.

23

24 The programs cited above, and the documents listed below are not intended to be 

25 comprehensive; rather, they are intended to provide appropriate examples of studies 

26 undertaken to understand and support engineered barrier performance. The examples provided 

27 above focused on engineered covers, which are only one type of barrier addressed in this 

28 guidance. The documents listed below provide information on a variety of different barriers.

29
P-35

1 Table P.4 Summary of Existing Key Documents Related to Engineered Barriers

Document Brief Summary

Walton, J.C., L.E. Plansky, and R.W. Smith, 

“Models for Estimation of Service Life of 

Concrete Barriers in Low-Level Radioactive 

Waste Disposal,” NUREG/CR-5542, 

U.S. Nuclear Regulatory Commission, 

Washington, DC, September 1990.

Provides primarily empirically based models 

for typical concrete formulations to estimate 

degradation rates.

Bennett, R.D., “Recommendations to the NRC 

for Soil Cover Systems Over Uranium Mill 

Tailings and Low-Level Radioactive Wastes—

Identification and Ranking of Soils for 

Disposal Facility Covers,” NUREG/CR-5432, 

U.S. Nuclear Regulatory Commission, 

Washington, DC, February 1991.

Discusses (1) selecting soil materials, 

(2) laboratory and field tests for covers, and

(3) construction methods.

NISTIR 5612, “4SIGHT, Manual: A Computer 

Program for Modeling Degradation of 

Underground LLW Concrete Vaults,” National 

Institute of Standards and Technology, 

Gaithersburg, MD, 1995.

User manual for numerical computer 

modeling of concrete degradation, 4SIGHT, 

to facilitate assessment of concrete vaults for 

isolating radioactive waste in low-level waste 

disposal applications.

NISTIR 89-4086, NUREG/CR-5466, “Service 

Life of Concrete,” National Institute of 

Standards and Technology, Gaithersburg, 

MD, 1995. 

Examines degradation processes in cementbased materials and discusses 

considerations of their occurrence, extent of 

potential damage, and mechanisms.

NUREG-1532, “Final Technical Evaluation 

Report for the Proposed Revised Reclamation 

Plan for the Atlas Corporation Moab Mill,” 

U.S. Nuclear Regulatory Commission, 

Washington, DC, March 1997.

Section 4 provides an example of the staff 

review of a reclamation design and discusses 

staff bases for acceptability of rock riprap 

erosion protection and input parameters used 

for those designs.

NISTIR 6519, “Effect of Drying Shrinkage 

Cracks and Flexural Cracks on Concrete Bulk 

Permeability,” National Institute of Standards 

and Technology, Gaithersburg, MD, 2000.

Discusses a model for predicting both the 

width and spacing of flexural and dryingshrinkage cracks to estimate composite 

(intact and cracked) concrete structure 

permeability.

U.S. NRC, “A Performance Assessment 

Methodology for Low-Level Radioactive 

Waste Disposal Facilities, Recommendations 

of NRC’s Performance Assessment Working 

Group,” NUREG-1573, U.S. Nuclear 

Regulatory Commission, Washington, DC, 

October 2000.

Provides general information pertinent to 

modeling and assessment of engineered 

barriers. Provides a bibliography of reports 

related to engineered barriers.

NISTIR 6747, “Validation and Modification of 

the 4SIGHT Computer Program,” National 

Institute of Standards and Technology, 

Gaithersburg, MD, 2001.

Discusses the validation and verification of 

the fluid transport mechanisms incorporated 

in the concrete degradation code 4SIGHT 

using reference and laboratory data.
P-36

Table P.4 Summary of Existing Key Documents Related to Engineered Barriers (cont.)

Document Brief Summary

EPA/600/R-02/099, “Assessment and 

Recommendations for Improving the 

Performance of Waste Containment

Systems,” U.S. Environmental Protection 

Agency, 2002.

Discusses issues related to the design, 

construction, and performance of waste 

containment systems used in landfills, surface 

impoundments, and waste piles and in the 

remediation of contaminated sites.

Johnson, T.L., “Design of Erosion Protection 

for Long-Term Stabilization,” NUREG-1623, 

U.S. Nuclear Regulatory Commission, 

Washington, DC, September 2002. 

Provides guidance on methods to achieve 

erosion controls for long-term stabilization. 

Provides a list of key references, including the 

technical work supporting the guidance.

Dwyer, Stephen F., “Water Balance 

Measurements and Computer Simulations of 

Landfill Covers,” PhD Dissertation, University 

of New Mexico, 2003.

Provides a comprehensive summary of data 

collection, analysis, and computer simulations 

associated with DOE’s ALCD program. Also 

includes a summary of measurements of 

infiltration at various sites with engineered 

covers.

“Handbook of Groundwater Remediation 

Using Permeable Reactive Barriers,” edited 

by D.L. Naftz, S.J. Morrison, C. Fuller, and 

J.A. Davis, 2003.

Discusses cost-effective remedies and 

provides case studies on treating 

contaminated groundwater using permeable 

reactive barriers, including discussions on 

construction, development of reactive 

materials, and operable chemical and 

biological reactions.

Interstate Technology & Regulatory Council 

(ITRC), “Technical and Regulatory Guidance 

for Design, Installation, and Monitoring of 

Alternative Final Landfill Covers,” 

Washington, DC, 2003.

A guidance document primarily written for 

decisionmakers associated with plan 

development, review, and implementation of 

alternative covers, focuses on the decisions 

and facilitating the decision processes related 

to the design, evaluation, construction, and 

postclosure care associated with alternative 

covers.

NAS, “Assessment of the Performance of 

Engineered Waste Containment Barriers,” 

National Academy of Sciences, Washington, 

D.C., 2007.

Identifies engineered barrier systems used for 

surface and subsurface waste containment, 

defines and evaluates performance of 

barriers, assesses methodologies to predict 

and monitor performance, and identifies 

information needs to fill knowledge gaps. The 

report provides a series of recommendations 

on data collection and distribution; models; 

monitoring periods, and performance criteria.

NISTIR 7026, “Condition Assessment of 

Concrete Nuclear Structures Considered for 

Entombment,” National Institute of Standards 

and Technology, Gaithersburg, MD, 2003. 

Provides assessment of cement-based 

engineered barrier structures based on

characterization of intact concrete and crack 

properties. Material property uncertainties 

are incorporated into a Monte Carlo 

simulation.
P-37

Table P.4 Summary of Existing Key Documents Related to Engineered Barriers (cont.)

Document Brief Summary

Albright, W.H., et al., “Field Water Balance of 

Landfill Final Covers,” Journal of 

Environmental Quality, 33(6), 2317–

2332, 2004.

Results of large-scale field research study to 

assess the ability of landfill final covers to 

control infiltration into underlying waste. A 

comprehensive publication summarizing 

ACAP experience.

Waugh, W.J., “Design, Performance, and 

Sustainability of Engineered Covers for 

Uranium Mill Tailings,” Proceedings of 

Long-Term Performance Monitoring of Metals 

and Radionuclides in the Subsurface: 

Strategies, Tools, and Case Studies. 

U.S. Environmental Protection Agency, 

U.S. Department of Energy, U.S. Geological 

Survey, U.S. Nuclear Regulatory 

Commission, April 21–22, 2004, Reston, VA, 

2004.

Provides information on experiences with 

cover designs for DOE’s UMTRA Project sites 

of conventional and alternative covers.

Interstate Technology & Regulatory Council, 

“Permeable Reactive Barriers: Lessons 

Learned/New Directions,” Washington, DC, 

2005.

Summarizes the understanding and 

experience with permeable reactive barriers, 

including numerous case studies.

“Long-Term Performance of Permeable 

Reactive Barriers, edited by K.E. Roehl, T. 

Meggyes, F.G. Simon, and D.I. Stewart, 

2005.

Describes methods for evaluating and 

enhancing the long-term performance of 

permeable reactive barrier systems, focused 

on systems that treat uranium, and organic 

contaminants by sorption and/or precipitation 

mechanisms and discusses a number of 

topics, including (1) selection, 

characterization, and development of suitable 

reactive materials, (2) identification of 

attenuation mechanisms, (3) accelerated 

testing, (4) evaluation of the influence of site 

conditions, and (5) monitoring techniques.

National Research Council, National 

Academy of Sciences, “Assessment of the 

Performance of Engineered Waste 

Containment Barriers,” 2007.

Examines current knowledge and gaps in the 

understanding of the performance of 

engineered waste containment barriers and 

recommends research needs in the area. 

Walter, G.R., P. Dubreuilh, “Evaluation of 

Approaches to Simulate Engineered Cover 

Performance and Degradation,” Prepared for 

U.S. NRC Contract NRC-02-07-006, Center 

for Nuclear Waste Regulatory Analyses, San 

Antonio, TX, April 2007. 

Reviews a report of 21 codes, consisting of 

(1) hydrologic codes for evaluating

percolation through the cover, (2) generalized

and localized erosion codes for evaluating

long-term stability of the cover, and

(3) miscellaneous codes for evaluating

degradation of covers.
P-38

Table P.4 Summary of Existing Key Documents Related to Engineered Barriers (cont.)

Document Brief Summary

Albright, W.H., Benson, C.H., Waugh, W.J., 

“Water Balance Covers for Waste 

Containment: Principles and Practice,” 

American Society of Civil Engineers, Reston, 

VA, 2010.

Textbook on evapotranspiration covers using 

information from numerous studies examining 

performance and degradation mechanisms of 

various engineered surface barrier 

components over time.

Arlt, H., “A Regulatory Perspective of 

Monitoring and Assessing Performance of 

Engineered Surface Barriers,” Proceedings of 

Waste Management Symposium, Phoenix, 

AZ, 2010.

Paper discussing function and type of 

engineered surface barriers, as well as 

approaches to monitoring and assessing 

time-dependent performance and degradation 

from a regulatory perspective.

Benson, C.H., et al., “Engineered Covers for 

Waste Containment: Changes in Engineering 

Properties & Implications for Long-Term 

Performance Assessment,” NUREG/CR7028, U.S. Nuclear Regulatory Commission, 

Washington, DC, December 2011.

Examines changes in properties of 

engineered covers used for waste 

containment and discusses practical 

implications on long-term performance 

assessment, as well as providing 

recommendations.

Arlt, Hans, R. L. Johnson, D. Mandeville, G. 

Alexander, M. Meyer, J. Philip, J. Kanney, M. 

Fuhrmann, and T. Johnson, 2011,

Documentation of the Engineered Covers 

Technical Group (ECTG) Activities. 

ML112300105

A qualitative assessment of existing sites to

identify and prioritize by risk existing and 

planned covers that may be impacted by the 

findings in NUREG/CR-7028. Processes 

contributing to risk include both radon flux 

from the disposal cell and effects to the 

groundwater due to the potential of an 

increased rate of water infiltration through the 

covers. 

Interstate Technology and Regulatory 

Council, “Permeable Reactive Barrier: 

Technology Update Version 1.01,” 2011. 

Provides consolidated information from

previously issued ITRC documentation on 

additional types of reactive media, 

contaminants amenable to treatment, 

longevity of barriers, and approaches to 

construction and installation of barriers.

Nicholson, T.J., and H.D. Arlt, “Proceedings 

of the Workshop on Engineered Barrier 

Performance Related to Low-Level 

Radioactive Waste, Decommissioning, and 

Uranium Mill Tailings Facilities: Held at the 

U.S. Nuclear Regulatory Commission 

Headquarters, Rockville, MD, August 3-5, 

2010,” NUREG/CP-0195, U.S. Nuclear 

Regulatory Commission, Washington, DC, 

August 2011.

Summarizes experience with engineered 

barriers and recommendations from Federal 

and State agencies, Tribal Nations, and 

academic and industry experts compiled 

during engineered barrier workshop at the 

NRC.
P-39

Table P.4 Summary of Existing Key Documents Related to Engineered Barriers (cont.)

Document Brief Summary

Dinwiddie, C., S. Stothoff, “Methods for 

Monitoring Net Inflow Through Soil Covers of 

Uranium Mill Tailings Impoundments,” 

Prepared for U.S. Nuclear Regulatory 

Commission by Center for Nuclear Waste 

Regulatory Analyses, San Antonio, Texas, 

November 2013.

The report examines approaches for 

implementing effective postclosure monitoring 

programs for estimating inflow through 

uranium mill tailings impoundment covers, 

focusing on U.S. Nuclear Regulatory 

Commission (NRC) licensed Title-II-in-closure 

sites.

Snyder, K.A., and W.J. Weiss, “Sensors and 

Monitoring to Assess Grout and Vault 

Behavior for Performance Assessment,” 

NUREG/CR-7169, U.S. Nuclear Regulatory 

Commission, Washington, DC, June 2014.

Preliminarily evaluates state-of-the-art of 

sensors, nondestructive evaluation methods, 

and any relevant geophysical techniques that 

may be used to quantify changes to the 

intended chemical (e.g., redox state) and 

structural properties (e.g., crack initiation, 

development, and propagation) of large 

engineered waste isolation systems.

Sagar, B., et al., “Uranium Tailings 

Impoundments—Bottom Liners and Their 

Performance,” Prepared for U.S. Nuclear 

Regulatory Commission by Center for 

Nuclear Waste Regulatory Analyses, San 

Antonio, Texas, January 2015.

The focus of the report is on providing 

information to address some uncertainties 

associated with the use of bottom liners under 

tailings impoundments at uranium recovery 

facilities. In addition, the information can be 

used to provide insights regarding the longterm performance expectations of bottom 

liners at tailings impoundments.

Smith, C.L., C.H. Benson, “Influence of 

Coupling Erosion and Hydrology on the LongTerm Performance of Engineered Surface 

Barriers,” NUREG/CR-7200, U.S. Nuclear 

Regulatory Commission, Washington, D.C., 

May 2016.

Hydrologic and landform evolution models 

were used to evaluate various engineered 

cover designs for LLW disposal facilities with 

respect to the ability of the cover designs to 

limit erosion and percolation rates. 

Fuhrmann, M., C.H. Benson, J. Waugh, M.

Williams, and H. Arlt, “Proceedings of the 

Radon Barriers Workshop,” NUREG/CP0312, NRC: Washington, DC. July 25-26, 

2018.

NUREG/CP-0312 summarizes results of 

studies of barriers at four in-service UMTRCA 

sites, focusing on hydraulic conductivity and 

radon flux.

1
P-40

1 P.4 References

2 Abt, S. R. and T. L. Johnson. “Riprap Design for Overtopping Flow.” American Society of Civil 

3 Engineers. Journal of Hydraulic Engineering, Vol. 117, No. 8. 1991.

4 Albright, W., et al. “Field Water Balance of Landfill Final Covers,” Journal of Environmental 

5 Quality, 33(6), 2317–2332. 2004.

6 ASTM C1260-14, “Standard Test Method for Potential Alkali Reactivity of Aggregates (Mortar7 Bar Method).” ASTM International: West Conshohocken, PA. 2014. 

8 Code of Federal Regulations. 40 CFR Part 192, “Health and Environmental Protection

9 Standards for Uranium and Thorium Mill Tailings”.

10 DOE. “Plant Encroachment on the Burrell, Pennsylvania, Disposal Cell: Evaluation of Long11 Term Performance and Risk.” DOE, Grand Junction Office: Grand Junction, CO. 1999.

12 Dwyer, Stephen F. “Water Balance Measurements and Computer Simulations of Landfill 

13 Covers,” PhD Dissertation, University of New Mexico, 2003.

14 Esmiol, E.E. “Rock as Upstream Slope Protection for Earth Dams—149 Case Histories,” Report 

15 No. DD-3, U.S. Department of the Interior, Bureau of Reclamation. 1967.

16

17 Fuhrmann, M., C.H. Benson, J. Waugh, M. Williams, and H. Arlt, “Proceedings of the Radon 

18 Barriers Workshop,” NUREG/CP-0312. NRC: Washington, DC. 2018.

19 James, A., et al. “Field Performance of GCL under Ion Exchange Conditions,” Journal of 

20 Geotechnical and Geoenvironmental Engineering, American Society of Civil Engineers, Vol. 13, 

21 No. 10, pp. 897–901. 1997.

22 Lin, L., and C.H. Benson. “Effect of Wet-Dry Cycling on Swelling and Hydraulic Conductivity of 

23 GCLs.” Journal of Geotechnical and Geoenvironmental Engineering, pp. 40–49. 2000.

24 Lindsey, C.G. et, al. “Long-Term Survivability of Riprap for Armoring Uranium Mill Tailings and 

25 Covers: A Literature Review,” NUREG/CR-2642 (also PNL-4225). NRC: Washington, DC. 

26 1982. 

27

28 Melchoir, S. “In-situ studies on the performance of landfill caps,” In Proceedings of the 

29 International Containment Technology Conference, pp. 365–373. DOE: Germantown, MD.

30 1997.

31 NRC. “Technical Position on Waste Form (Revision 1).” Division of Low-Level Waste 

32 Management and Decommissioning. NRC: Washington, DC. 1991.

33 — — — — —. “Design of Erosion Protection for Long-Term Stabilization, Draft Report for 

34 Comment,” NUREG-1623. NRC: Washington, DC. 2002.

35

36 — — — — —. “Effectiveness of Engineered Covers: From Modeling to Performance 

37 Monitoring,” NUREG/CR-7028. NRC: Washington, DC. 2011a.

38 — — — — —. “Proceedings of the Workshop on Engineered Barrier Performance Related to 
P-41

1 Low-Level Radioactive Waste, Decommissioning, and Uranium Mill Tailings Facilities,”

2 NUREG/CP-0195. NRC: Washington, DC. 2011b.

3

4

5 — — — — —. “Early Leak Detection External to Structures at Nuclear Power Plants,” 

6 NUREG-2151. NRC: Washington, DC. 2013.

7 — — — — —. “Sensors and Monitoring to Assess Grout and Vault Behavior for Performance 

8 Assessment,” NUREG/CR-7169. NRC: Washington, DC. 2014.

9

10 Pabalan, R.T., et al., “Review of Literature and Assessment of Factors Relevent to Performance 

11 of Grouted Systems for Radioactive Waste Disposal,” CNWRA 2009-001, Prepared for Nuclear 

12 Regulatory Commission Contract NRC NRC-02-07-006. Center for Nuclear Waste Regulatory 

13 Analyses: San Antonio, Texas. 2009.

14 Smith, C.L., C.H. Benson. “Influence of Coupling Erosion and Hydrology on the Long-Term 

15 Performance of Engineered Surface Barriers,” NUREG/CR-7200, U.S. Nuclear Regulatory 

16 Commission: Washington, DC. 2016.

17

18 Waugh, W.J., et al. “Ecology, Design, and Long-Term Performance of Surface Barriers: 

19 Applications at a Uranium Mill Tailings Site,” Presented in “Barrier Technologies for 

20 Environmental Management,” Summary of a Workshop, National Research Council, National 

21 Academy of Sciences. 1997.

22 Waugh, W.J., “Design, Performance, and Sustainability of Engineered Covers for Uranium Mill 

23 Tailings.” Environmental Sciences Laboratory, S.M. Stoller Corporation: Grand Junction, 

24 Colorado. 2004.

UNCERTAINTY IN PERFORMANCE AND DOSE ASSESSMENTS

Q-1

1 Q.1 Introduction

2 The objective of this appendix is to provide information and guidance to the NRC staff on the 

3 treatment and representation of uncertainty in site-specific dose or performance assessments, 

4 DPs, and technical analyses. This guidance discusses types of uncertainty, methods to 

5 understand the impact of uncertainty, and different approaches to incorporate uncertainty in 

6 these assessments. The appendix provides a generic example of the results of different 

7 methods to represent uncertainty in dose and performance assessments. 

8

9 Q.2 Background

10 A performance assessment is a projection of what can happen, how likely it is, and the 

11 associated consequences for potentially many thousands of years into the future. Dose

12 assessments typically incorporate uncertainty and variability in data and parameters. Although 

13 dose or performance assessments may be probabilistic or deterministic, modern computational 

14 advances have allowed greater use of probabilistic assessments. Dose and performance 

15 assessments can be data intensive, and in many cases, data can be sparse. Sparse site16 specific data typically lead to the use of generic sources of information, where the basis may 

17 range from limited to fairly complete data sets. If caution is not taken, misinterpretation or 

18 misrepresentation of the variance in information, particularly when site-specific supporting 

19 information is sparse, may result in ineffective decision-making. Therefore, results produced 

20 from models with sparse data and limited site-specific information can require careful 

21 interpretation. 

22

23 Figure Q.1 provides a conceptual overview of the dose and performance assessment process. 

24 The central portion of the figure has four main steps: collect data, develop conceptual models, 

25 develop numerical and computer models, and combine models and estimate effects. Each step 

26 in the process can be influenced by uncertainty in different ways. Usually consideration of 

27 uncertainty in dose and performance assessments is focused on data uncertainty and, in 

28 particular, a subset of overall data uncertainty (i.e., observed data uncertainty). This appendix 

29 covers data uncertainty, as well as uncertainty about process and conceptual models (model 

30 uncertainty) and model representation or computational uncertainty.

31

32 It is important to consider uncertainty in dose and performance assessments, because the

33 assessments are designed to provide estimates of future system performance for consideration 

34 against regulatory criteria. If uncertainty has not been appropriately represented in the results 

35 of the assessments, regulatory decision-making may be impaired or inappropriately biased. 

36 This appendix is organized into the following main sections:

37

38 • Section Q.2 of this appendix provides examples of different types of uncertainty that may

39 be encountered in a dose or performance assessment.

40 • Section Q.3 discusses high-level topics and complications that can arise as a result of

41 incorporating uncertainty in a dose or performance assessment.

42 • Section Q.4 provides general questions that a reviewer should consider when evaluating

43 uncertainty in a dose or performance assessment.

44 • Section Q.5 discusses different methods for evaluating the impact of different

45 uncertainties on dose and performance assessment results.
Q-2

1 • Section Q.6 briefly discusses different uncertainty and sensitivity analysis methods. 

2

3

4 Figure Q.1 Conceptual Overview of the Performance Assessment Process
Q-3

1 Q.3 Types of Uncertainty

2 A variety of different sources of uncertainty can affect dose and performance assessment

3 modeling. Figure Q.2 is a conceptual representation of the dose and performance assessment

4 process and significant sources of uncertainty associated with the different steps in the process. 

5 At the fundamental level, uncertainty can be classified as epistemic or aleatory. Epistemic 

6 uncertainty is sometimes referred to as reducible uncertainty or lack of knowledge uncertainty. 

7 In theory, this type of uncertainty can be reduced by investing resources to develop 

8 understanding. In practice, reduction in epistemic uncertainty may not be warranted (i.e., not 

9 risk significant) or may be too expensive. An example of epistemic uncertainty is the porosity of 

10 concrete in a waste disposal vault. Aleatory uncertainty arises from a perceived randomness in 

11 the behavior of the system under study. Aleatory uncertainty involves the characterization of 

12 the likelihood of events that have a real possibility of occurrence due to fundamental natural 

13 processes. The appropriate separation of aleatory uncertainty and epistemic uncertainty can be 

14 an important part of the conceptual structure and associated computational analysis of a 

15 complex system. However, aside from the most complex PAs the staff may review, a 

16 separation of uncertainty is rarely made. In some cases, uncertainties cannot be readily 

17 classified as either epistemic or aleatory. A process that results in an observation that is 

18 perceived as random may not be understood, such that the effect of the process is classified as 

19 aleatory whereas ultimately it is the process that is epistemic.

20

21 Figure Q.2 is a conceptual figure to communicate the major sources of uncertainty in the dose 

22 and performance assessment process. Most PAs explicitly account for data uncertainty and will 

23 implicitly account for model uncertainty. Few explicitly or implicitly account for some subtypes of 

24 representational uncertainty, although some projects with rigorous QA programs may have 

25 requirements that broadly address representational uncertainty. There are also subtypes of 

26 uncertainty (e.g., integration uncertainty) within broader classes (model uncertainty). 
Q-4

1

2

3

4 Figure Q.2 Uncertainty Types Associated with Performance Assessment

5 The following definitions are applied in this appendix:

6

7 Uncertainty—Uncertainty is a multifaceted characterization about data or predictions made from 

8 data that may include several concepts, including error, accuracy, validity, quality, noise, and 

9 confidence.

10

11 Aleatory uncertainty—Irreducible uncertainties stemming from an inherent randomness in 

12 processes.

13

14 Epistemic uncertainty—Reducible uncertainty or lack of knowledge uncertainty.

15

16 Performance assessment model—The integrated model of numerous submodels used to 

17 simulate the range of future performance.

18

19 Submodel—A component of a dose or performance assessment model (e.g., wasteform model, 

20 groundwater transport model) that may be an abstraction of a detailed process model.

21

22 Abstraction—A simplification of a process model that represents the essential behavior of the 

23 process model (e.g., lookup tables, response surfaces, and parameter distributions).

24

25 Process model—A detailed “physics-based” model used to simulate some aspect of the dose or 

26 performance assessment model, usually involving a solution of some form of differential 

27 equations, with a high (relative to a dose or performance assessment model) degree of spatial 

28 and temporal resolution, noting that computational resources may be high for using a process 

29 model.

Data Uncertainty:

Information (what)

Measurement (how)

Variability (where)

Validity (why) Model Uncertainty:

Conceptual/alternatives

Support

Integration of submodels

Representational 

Uncertainty:

Abstractions

Verification (response)
Q-5

1 Verification—The process of ensuring that the calculated values are correct, noting that a 

2 standard QA program includes verification of all technical aspects of the dose and performance 

3 assessment process.

4

5 Model support—Multiple lines of evidence used to support the results of a dose or performance 

6 assessment model, including but not limited to laboratory experiments, field experiments, field 

7 observations, natural analogs, expert elicitation, and comparison to alternative models.

8

9 Q.4 General Considerations

10 The licensee should evaluate data, model, and representational uncertainty as part of the dose 

11 and performance assessment process. In general, a systematic and proactive approach to 

12 managing uncertainties is better than an ad hoc and reactive approach. In particular, for dose 

13 and performance assessment models, uncertainties may not be well understood. Limited 

14 information, as well as inherent complexity, can impair the development of understanding and 

15 appropriate management techniques.

16

17 Deterministic modeling is sometimes used for PAs. Typically, these types of PAs will attempt to 

18 use “conservative” parameters and models and evaluate uncertainties with one-at-a-time 

19 parameter sensitivity analyses. However, because dose and performance assessment models 

20 are collections of integrated submodels, local and subjective definitions of “conservative” may 

21 not be robust because of the inherent complexity in dose and performance assessment model 

22 response. In addition, one-at-a-time parameter sensitivity analyses are a local measure of the 

23 impact of an uncertainty. The term local is with respect to the dimensions of the uncertainty 

24 space. Deterministic modeling that can overcome these disadvantages can be an effective tool 

25 to provide information for decision-making.

26

27 As computational resources have improved, probabilistic modeling has become more common. 

28 Probabilistic modeling combined with global uncertainty analyses can provide more information 

29 with respect to the importance of different uncertainties and submodels compared to 

30 deterministic modeling. It can also be used to make more robust “conservative” modeling 

31 decisions. A primary drawback can occur when probabilistic modeling is used for a site-specific 

32 analysis with very limited site-specific information. Risk dilution occurs when favorable 

33 conditions (compared to the specific site) are included in parameter distributions and models 

34 that are not representative of the site-specific conditions.

35

36 The overall objective of dose and performance assessment modeling should be to make the 

37 models as simple as possible but no simpler. Stated another way, the complexity of the dose 

38 and performance assessment model should be consistent with available supporting information 

39 and be able to include the response to uncertainties. Even with advances in computational 

40 resources, complexity is not always warranted or desired. Adding complexity that does not 

41 address risk or advance decision-making impedes the review and understanding process and 

42 dilutes effort on the more significant aspects. On the other hand, a model without adequate 

43 complexity and model support may miss important processes or interactions. The dose and 

44 performance assessment process should be iterative, which allows the analyst to arrive at the 

45 appropriate level of complexity in an iterative manner. The complexity of a model will need to 

46 reflect a balance between needing to include essential behavior and avoiding the burden of 

47 nonessential behavior. If model support information is available, it can be used to help 

48 determine when the level of detail of the modeling is sufficient.

49
Q-6

1 Q.5 Review Procedures

2 This material is intended to be used in a risk-informed manner. For sites where the risk is 

3 confidently low or the analysis is clearly conservative, consideration of uncertainties is less 

4 important, and these procedures can be applied on a selective basis.

5

6 Q.5.1 General Considerations

7 The following general considerations may be used to perform a high-level evaluation of the 

8 scope of the uncertainty assessment. These are written in question form to allow a reviewer to 

9 step through them and determine how the topics were addressed.

10

11 • Is the data (e.g., range, distribution type) representative of the site-specific conditions?

12 o Does it vary spatially and temporally? If so, how has that variability been

13 characterized in the data?

14 o Can bounds to the data be established?

15 o Has the data been measured? What is the measurement error? Is there bias in the

16 measurements?

17 o Did the data have outliers? How were they treated?

18 • Are correlations in the data needed?

19 • Have generic datasets been used?

20 o Has the full range of the data been used?

21 o Has risk dilution been introduced?

22 o Is the generic dataset representative of site-specific conditions?

23 o If a subset of the generic dataset has been used, how was it selected?

24 • Has adequate model support been provided or are there alternative conceptual models?

25 o Have the alternative models been evaluated?

26 o Have combinations of alterative models been evaluated?

27 o Does the selected model have higher or lower risk?

28 o Is the selected model most plausible? Why?

29 • Is there uncertainty associated with how submodels have been integrated?

30 o How have the connections been established?

31 o Is any of the information that is passed an observable metric?
Q-7

1 o Are there feedbacks? If yes, how were they established?

2 • Were abstractions used in the modeling?

3 o Does the abstraction adequately represent the process model?

4 o What comparisons were made (e.g., visual, statistical)?

5 o Has verification of the abstraction been performed?

6 o Does the abstraction preserve important responses that would otherwise be present

7 in the process model?

8 By addressing these high-level general questions, the reviewer should ensure that uncertainty 

9 has been appropriately handled in the modeling. Not all of these general considerations will 

10 apply in every review; however, in the majority of cases, these questions can be used as a form 

11 of checklist to evaluate the treatment of uncertainty.

12

13 Q.5.2 Examples

14 The following are intended to provide illustrative examples of the technical topics contained in 

15 the previous section. The examples may be related to a specific discipline (e.g., geochemistry) 

16 but are to be used more generically by the reviewer. In other words, they are intended to 

17 provide practical examples of concepts.

18

19 Q.5.2.1 Data Representativeness

20 Data must be representative of the site-specific conditions and be consistent with the 

21 observations. If not, use of data that are not representative could introduce errors into the dose 

22 and performance assessment. 

23

24 The following example is for the liquid saturation in a layer of the unsaturated zone that is 

25 sufficiently below the surface to not be subject to diurnal variation. However, seasonal variation 

26 is significant. Figure Q.3 is a plot of 10 years of the liquid saturation of the layer in question 

27 showing the times that samples were collected. Over 100 samples were collected with a mean 

28 liquid saturation of 0.88 and a standard deviation of 0.11. Figure Q.4 is a contour plot of an 

29 areal view of the liquid saturation in a layer, showing the sample locations. 
Q-8

1

2

3

4 Figure Q.3 Ten Years of Simulated Liquid Saturation Data

5

6

7 Figure Q.4 Contour Plot of Liquid Saturation at a Given Point in Time

8

Sampling window

Sampling 

location
Q-9

1 Annual liquid saturation samples were collected in the early spring months each year, from fixed 

2 sampling locations. The example is intended to convey the importance of the 

3 representativeness of data. Even with many samples, if those samples do not characterize the 

4 temporal and spatial variability in the data, then simulations based on the data may be 

5 inaccurate. Temporal variability in environmental data may be driven by diurnal, seasonal, and 

6 longer-term fluctuations. Data used in a dose or performance assessment must adequately 

7 account for the different types of variability. Spatial variability may also play an important role. 

8 In the example shown in Figure Q.4, if sampling were limited to the region of the dashed circle, 

9 the derived values might not be appropriate to simulate the overall performance of the site. It is 

10 important to consider the scale of the variability (e.g., the correlation length) compared to the 

11 scale of the disposal facility. Short correlation lengths in comparison to the scale of the facility 

12 are not generally a cause for concern. It can be especially important to ensure that statistical 

13 techniques used to develop data have preserved real observed variability.

14

15 Q.5.2.2 Data Correlations

16 Since PAs attempt to simulate many different processes (e.g., infiltration, waste release, 

17 transport), it is important to assess the need for correlation of different parameters. For 

18 computational reasons, PAs commonly use abstractions. An abstraction is a simplification of a 

19 process model that represents the essential elements of the process model. The abstractions 

20 or submodels of a dose or performance assessment may not directly simulate global processes 

21 but can be commonly influenced by a global process or condition. If correlations should have 

22 been used but were not, the net effect is that the likelihood of favorable or unfavorable 

23 conditions may not be appropriately represented. Two examples are used below to illustrate the 

24 concept. 

25

26 The first example is for consumption data in a biosphere model. Consumption data are needed 

27 when food pathways are considered in an assessment. Usually data are provided for the 

28 individual consumption rate of different items (e.g., beef, pork, chicken, eggs, milk), using an 

29 average member of the critical group construct. Problems can arise if the analyst does not 

30 impose correlation in the data or constraints on consumption. The average member will have a 

31 defined range of caloric intake that should be consistent with the assumptions about the 

32 individual. If all consumption parameters are sampled independently, at the extremes it can 

33 result in nonphysical overall caloric or fluid intake rates, both high and low.

34

35 The second example involves corrosion of a carbon steel disposal container and transport 

36 through underlying soil. The corrosion rate of the carbon steel could be input into the dose or 

performance assessment as a simple rate (mils/y)1 37 that characterizes the observed values of 

38 corrosion rates. Likewise, the transport rate through the soil can be characterized by a 

39 retardation factor, which is, itself, a function of a distribution coefficient. Both of these seemingly 

40 different processes will be influenced by moisture content and pH. Independent simulation of 

41 the processes without correlation in the parameters (analogous to a lumped parameter model) 

42 may result in nonphysical results. For instance, the corrosion rate may be higher, and the 

43 distribution coefficient may be lower at low pH values. If pH was sampled and both the 

44 corrosion rate and distribution coefficient were correlated to it, the response of the system 

45 model could be appropriately captured. If state variables are defined and the submodels or 

46 abstractions are based on the state variables, the need for complex correlation structures can 

47 be reduced or eliminated.

48

1 1 mil = 0.001 inches 
Q-10

1 Q.5.2.3 Use of Generic Data

2 Generic data are commonly used in PAs. In most applications, the goal of the dose or 

3 performance assessment is to provide estimates of site-specific performance. As discussed in 

4 Section Q.4.2.1, the representativeness of generic data must be considered. A generic dataset 

5 may contain—in fact, is likely to contain—values that are not representative of site-specific 

6 conditions. Some parameters contribute mainly to the magnitude of the output metric 

7 (e.g., drinking water ingestion rate), whereas others can contribute to both timing and magnitude 

8 (e.g., Kd values). If the generic data are unbiased compared to the site-specific data and only 

9 affect the magnitude, the impact is mainly on the perceived variance in the output but not on the 

10 mean of the output. However, skewed generic data can lead to unreliable estimates. 

11

12 Generic data used in site-specific PAs that can influence the magnitude and timing of the output 

13 require careful consideration. Figure Q.5 shows a hypothetical distribution coefficient (Kd) 

14 parameter distribution derived from a generic reference such as “Default Soil Solid/Liquid 

15 Partition Coefficients, Kds, for Four Major Soil Types: A Compendium,” (Sheppard and 

16 Thibault, 1990). The generic distribution may represent all measurements performed over all 

17 sites. It contains within the distribution all observed sources of uncertainty and variability, such 

18 as mineralogy, geochemistry (pH, Eh), and other sources of variability (e.g., measurement error, 

19 measurement technique). The color-coded symbols represent “measurements” at three 

20 different sites. The sites may have different mineralogy (sand, loam, clay) and different 

21 geochemical conditions. Inclusion of the portion of the generic distribution that is not 

22 representative of the site-specific conditions can result in risk dilution. For instance, the sorption 

23 associated with the “x” symbols in the figure is much stronger on average than the sorption 

24 associated with the square symbols. If the specific site being analyzed has conditions 

25 comparable to the squares, inclusion of the other data “dilutes” the risk in the analysis. For the 

26 analyst, the challenge is determining when generic datasets have been used appropriately and 

27 when they have not been.

28

29 Different approaches can be used to understand the impact of using generic datasets. As 

30 discussed in Section Q.6, sensitivity and uncertainty analyses can identify which parameters are 

31 risk significant. If a parameter derived from a generic dataset is identified as being risk 

32 significant, a number of actions should be considered:

33

34 • Collect site-specific information to replace or constrain the generic dataset.

35 • Introduce conservative bias in the representation of the dataset in the dose or

36 performance assessment.

37 • Use an alternative metric for decision-making, such as mean of the peaks, rather than

38 the normal probabilistic model output metric, which is peak of the means.
Q-11

1

2

3 Figure Q.5 Kd Distribution with Values Plotted from Measurements at Three Different 

4 Hypothetical Sites

5 Figure Q.6 is from a paper entitled, “The Impact of Uncertainty Type and Representation in 

6 Performance Assessment,” 2012 (Esh, Grossman, and Parks, 2012), which used a simplified 

7 dose or performance assessment model to examine the impact of sparse information and 

8 generic datasets on assessment model output for different analysis techniques. As shown in 

9 (a), risk dilution can occur in probabilistic modeling using generic datasets for a site-specific 

10 evaluation. In addition, when information is sparse to establish parameter distributions, there is 

11 a higher likelihood that the distribution assigned may not be representative and may be in error. 

12 The analyst should consider how much information is available to assign parameter distributions 

13 and how representative it may be. 
Q-12

(a)

(b)

1 Figure Q.6 

2

3

Influence of Different Amounts of Site-Specific Data on Performance 

Assessment Model Output (a) is for Uranium whereas (b) is for Sr-90; due to 

Risk Dilution, even a Very High Percentile of the Probabilistic Modeling Results 

4 may be outside of the Range Expected for a Site-Specific Analysis

U238

0.0E+00

2.0E-07

4.0E-07

6.0E-07

8.0E-07

1.0E-06

1.2E-06

1.4E-06

1.6E-06

1.8E-06

0 5,000 10,000 15,000 20,000 25,000

Time(yr) Dose (Sv/yr)

D1

D2

D3

D4

D5

D6

D7

D8

D9

D10

CD1

CD2

CD3

CD4

CD5

CD6

CD7

CD8

CD9

CD10

Base 95%

Base Mean

Base 5%

Full 95%

Full Mean

Full 5%

Conservative Deterministic - Low

Base Probabilistic - Low

Full Probabilistic

Deterministic - Low

Sr90

1.0E-10

1.0E-09

1.0E-08

1.0E-07

1.0E-06

1.0E-05

1.0E-04

1.0E-03

1.0E-02

1.0E-01

1.0E+00

0 200 400 600 800 1,000

Time(yr) Dose (Sv/yr)

D1

D2

D3

D4

D5

D6

D7

D8

D9

D10

CD1

CD2

CD3

CD4

CD5

CD6

CD7

CD8

CD9

CD10

Base 95%

Base Mean

Base 5%

Full 95%

Full Mean

Full 5%

Conservative Deterministic - Low

Base Probabilistic - Low

Full Probabilistic

Deterministic - Low
Q-13

1 Q.5.2.4 Alternative Conceptual Models

2 Model uncertainty is common in system modeling, especially for complex systems. Dose and 

3 performance assessment models cannot be validated in the traditional sense, which results in 

4 model uncertainty. Even the submodels of a dose or performance assessment, which may be 

5 focused on select aspects of the overall modeling effort, usually cannot be validated because of 

6 the projections over very long timescales. Model support may vary between submodels such 

7 that some models have high confidence whereas others have low confidence. Model 

8 uncertainty is directly tied to model support; when model support is limited, it can be expected 

9 that model uncertainty will be higher. 

10

11 Even with a strong effort to develop model support, there may be multiple conceptual models 

12 that are consistent with the available supporting information. Table Q.1 provides an example of 

13 a dose or performance assessment model broken down into submodels. In this example, there 

14 are multiple submodels that could be used to represent different parts of the dose or 

15 performance assessment model. There are a total of 24 different combinations of submodels. 

16 A challenge arises because the relative importance of an individual submodel may or may not 

17 be conditional on the other submodels with which it is combined. For example, waste release 

submodel 1 (WR1) may be of relative importance2 18 1 when combined with solubility 1 (SOL1) but 

19 could be of relative importance 3 when combined with solubility 2 (SOL2). 

20 Table Q.1 Example of Alternative Conceptual Models in a Performance Assessment

Submodel Alternative 1 Alternative 2 Alternative 3

Infiltration INF1 INF2

Waste Release WR1 WR2 WR3

Solubility SOL1 SOL2

UZ Transport UZ1 UZ2

SZ Transport SZ1

21

22 Different methods can be used to assess the importance of alternative conceptual models. The 

23 analyst should base the safety decision on a conceptual model that is adequately supported or 

24 sufficiently conservative to account for the effects of model uncertainty on the decision. Three 

25 methods to address model uncertainty are the following:

26

27 (1) iterative dose and performance assessment using increased model support on risk

28 significant submodels to reduce model uncertainty

29 (2) analysis of submodel combinations; use of conservative combinations

30 (3) probability weighting of submodels based on the degree of belief of each submodel and

31 performing a probabilistic assessment

32

33 All of these methods require some amount of quantitative representation of each submodel, 

34 although abstractions may be used to reduce the resources required to perform the 

35 assessment. The preferred method to deal with model uncertainty is to use an iterative dose or 

2 Relative importance is defined as the ratio of risk using the submodel compared to the relative risk using the 

submodel that provides the lowest risk.
Q-14

1 performance assessment. First, sensitivity and uncertainty analyses combined with barrier

2 analyses are used to identify the most risk-significant submodels (note: risk significance can 

3 result in an increase or decrease to the output metrics of concern). Next, model support is 

4 developed to constrain the conceptual models and identify which submodels are most accurate. 

5 If enough model support can be developed to identify the most accurate submodels, then model 

6 uncertainty has been effectively reduced.

7

8 The second method for determining model uncertainty is to evaluate each combination of 

9 submodels individually and evaluate the results. The most conservative combination likely will 

10 not reflect “risk,” but it may provide a bounding argument to make a regulatory decision. This 

11 method avoids the problem of assigning a degree of belief to the different submodels. This 

12 method can also be used by selecting a conservative combination for decision-making 

13 purposes, although the conservatism is subjective and should be presented as such.

14

15 The third method is similar to the second except a degree of belief is assigned to each 

16 submodel. This makes the importance of the submodels more risk informed, if the degrees of 

17 belief are based on an unbiased process. The challenge is that, when model support is limited, 

18 the subjective degrees of belief may have significant error rates associated with them. It is 

19 advisable to include uncertainty in the degrees of belief, however they are obtained. One 

20 method is to perform independent expert elicitation.

21

22 Q.5.2.5 Model Integration

23 There may be uncertainty associated with how submodels should be integrated, especially 

24 when model support is limited. Model integration involves developing the connections between 

25 submodels by defining what information is passed and how it should be passed. Ideally, the 

26 connections between submodels should be based on observable information 

27 (e.g., concentration of radionuclides, flux of radionuclides) but that may not always be possible, 

28 especially for a facility under design. The information that needs to be passed is primarily 

29 addressed initially by the subject matter experts developing the model and is confirmed by 

30 model support activities.

31

32 It is important to ensure that the spatial and temporal distribution of the information being 

33 passed is at an appropriate resolution such that data reduction has not eliminated important 

34 model response. The following example shows the flow of chloride concentrations through 

35 three different submodels (soil layer, engineered cover, waste zone) in a waste disposal system. 

36 The ultimate goal of the model is to estimate the corrosion rate of a steel liner located below the 

37 waste zone. The corrosion rate is a function of the chloride concentration:

38

C (mils/y) = a (Cl-) 39 b where a and b are constants

40

41 The flux rate and concentration of chloride is represented with ten cells in the soil submodel

42 (see Figure Q.7); the concentrations vary from cell to cell. Those concentrations (fluxes) are 

43 translated through the system until they are used to estimate failure times of the steel liner. The 

44 estimated failure times may not be accurate if the temporal and spatial resolution of the data 

45 has not been preserved during submodel integration. Assuming the chloride concentration at 

46 location 1 is five times the value in the neighboring cell (1.1) in the top soil layer, the 

47 concentration of chloride assuming only vertical flow will be approximately 1.7 times less when it 

48 reaches the liner (region 4).
Q-15

1

2

3 Figure Q.7 Hypothetical Waste Disposal System Showing Representation of Different 

4 Submodels

5 For the expression given above, if b = 1.7, then the total corrosion estimated at region 4 would 

6 be approximately 68 mils (from the red dashed line to the left), compared to 170 mils in the 

7 highlighted region. In one case, failure may not occur whereas, in the other case, releases may 

8 occur, albeit over a smaller area. This example highlights the importance of proper spatial and 

9 temporal integration of submodels. Comparison with observed values can help ensure that the 

10 temporal and spatial resolution was appropriate for the problem.

11

12 Q.5.2.6 Abstractions

13 It is expected that abstractions may be needed in dose or performance assessment modeling. 

14 An abstraction is a simplification of a process model that represents the essential behavior of 

15 the process model. Examples include but are not limited to lookup tables, response surfaces, 

16 and parameter distributions. An abstraction preserves the essential behavior in a form that can 

17 be implemented in a computationally efficient manner. The use of abstractions can result in 

18 reduced complexity that has many benefits, such as easier review and interpretation. The main 

19 disadvantage is that complex responses of integrated models may not be preserved. The 

20 questions a reviewer may ask with respect to abstractions are found in Section Q.4.1.

21

22 The following example is used to illustrate abstractions in the form of parameter distributions 

23 and how model performance may be improved by ensuring the abstractions are consistent with 

24 the underlying state variable. Consider a system in which the solubility of a radionuclide and the 

25 distribution coefficient are represented with a parameter distribution. The underlying model 

26 response is equivalent to:

27

28 Y = a (solubility) / (Kd)

29

30 However, both solubility and the distribution coefficient are driven by the underlying state 

31 variable pH in the example. Figure Q.8 provides the model response as a function of Kd for two 

32 different representations: (1) the abstractions for solubility and Kd (parameter distributions) are 

1.1
Q-16

1 based on the observed ranges and are sampled independently, also defined as a Level 0 

2 abstraction, and (2) the abstractions for solubility and Kd still use parameter distributions but 

3 those distributions are based on an underlying state variable pH (in effect the parameter 

distributions are correlated), which is known as a Level 1 abstraction.

3 4 As can be seen from the 

5 figure, the overall model response is more constrained for higher order abstraction. It may not 

6 always be possible to define higher order abstractions, because the state variables themselves 

7 or the responses to the state variables may not be known. However, if higher order abstractions 

8 can be defined, they should be used, as they preserve more of the underlying process model 

9 response. 

10

11

12 Figure Q.8 Model Response as a Function of Kd Using a Level 0 or Level 1 Abstraction 

13 Q.6 Sensitivity and Uncertainty Analyses

14 Sensitivity and uncertainty analyses can and should be used to understand dose or 

15 performance assessment models; to identify risk-significant parameters, models, and 

16 assumptions; and to evaluate the incorporation of uncertainty in the assessment. In the high17 level waste program, the NRC has over 20 years of experience performing sensitivity and 

18 uncertainty analyses (CNWRA, 2011). The Center for Nuclear Waste Regulatory Analyses 

19 (CNWRA) developed a compilation of the different techniques and the lessons learned 

20 associated with the use of those techniques. That report should be considered as an essential 

3 In the Level 1 abstraction, the Kd and solubility are both a function of the state variable pH, which is uncertain. 
Q-17

1 reference to complete or review sensitivity and uncertainty analyses of PAs. The report 

2 includes sections on:

3

4 Uncertainty Analysis—Provides discussion on topics such as parameters, sampling, maximum 

5 entropy, correlations, convergence, output metrics, and risk dilution.

6

7 Sensitivity Analysis—Provides discussion on sensitivity analyses applied to parameters, 

8 barriers, and submodels. It discusses traditional regression-based methods, as well as 

9 nonparametric methods.

10

11 Advanced and Special-Case Sensitivity Techniques—Provides a detailed summary of many 

12 methods that have been developed over time to support the high-level waste dose or 

13 performance assessment. However, a fair number of the techniques have been used in other 

14 programs. Techniques include but are not limited to Cumulative Distribution Function,

15 Sensitivity Analyses Methods, Genetic Algorithms with Cascaded Variable Selection, 

16 Regionalized Sensitivity Analyses, Fraction Factorial Methods, and many more.

17

18 Q.7 Additional Resources

19 A number of resources cover the topic of uncertainty applied to PAs. This guide is intended to 

20 complement those resources. For additional information, the user may want to consult the 

21 following, which are listed in the reference section:

22

23 NUREG-1854 (2007)—Provides review procedures for data uncertainty, model uncertainty, and

24 model support (among other topics) applied to the review of U.S. Department of Energy waste 

25 determinations. Specific review procedures associated with individual topics are provided 

26 (e.g., radionuclide transport) as well as generic review procedures for PAs.

27

28 NUREG-1573 (2000)—Provides recommendations of the NRC’s Performance Assessment

29 Working Group on acceptable methodologies for PAs of low-level radioactive waste disposal 

30 facilities. Section 3.2.4. contains guidance on the treatment of sensitivity and uncertainty in 

31 low-level waste. Section 3.3.2 includes a description of sources of uncertainty, issues, and 

32 recommended approaches.

33

34 Mohanty, S. et al (2011)—Provides a robust compilation of sensitivity and uncertainty analyses

35 techniques that have been used for over two decades in the high-level waste program. It 

36 includes many lessons learned and valuable tips that are generally applicable to other systems.

37

38
Q-18

1 Q.8 References

2 Esh, D., C. Grossman, and L. Parks, “The Impact of Uncertainty Type and Representation in 

3 Performance Assessment,” 11th International Probabilistic Safety Assessment and 

4 Management Conference, Helsinki, Finland, 2012.

5

6 Mohanty, S. et al, “History and Value of Uncertainty and Sensitivity Analyses at the Nuclear 

7 Regulatory Commission and Center for Nuclear Waste Regulatory Analyses,” Center for 

8 Nuclear Waste Regulatory Analyses, Southwest Research Institute, San Antonio, TX, 2011. 

9 Agencywide Documents Access and Management System (ADAMS) Accession 

10 No. ML112720123.

11

12 Schmidt, D. W. et al, “Consolidated NMSS Decommissioning Guidance: Characterization, 

13 Survey, and Determination of Radiological Criteria,” NUREG-1757, Vol. 2, 2003. ADAMS 

14 Accession No. ML032530405.

15

16 Sheppard, M.I. and D.H. Thibault. Default Soil Solid/Liquid Partition Coefficients, Kds, for Four 

17 Major Soil Types: A Compendium. Health Physics. Vol. 59, pp. 471–482, 1990.

18

19 U.S. Nuclear Regulatory Commission, “NRC Staff Guidance for Activities Related to 

20 U.S. Department of Energy Waste Determinations,” NUREG-1854 (Draft Final Report for Interim 

21 Use), 2007. ADAMS Accession No. ML072360184.

22

23 U.S. Nuclear Regulatory Commission, “A Performance Assessment Methodology for Low-Level 

24 Radioactive Waste Disposal Facilities: Recommendations of NRC’s Performance Assessment 

25 Working Group.” NUREG-1573, October 2000.
Technical 



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