DOE/SRS-WD-2014-001 

Revision 0 

Basis for Section 3116 Determination 

for 

Closure of H-Tank Farm 

at the 

Savannah River Site 

December 2014 
Basis for Section 3116 Determination DOE/SRS-WD-2014-001 

for Closure of H-Tank Farm Revision 0 

at the Savannah River Site December 2014 

Page ii 

 

REVISION SUMMARY 

REV. # DESCRIPTION DATE OF ISSUE 

0 Initial Issue December 2014 

 
Basis for Section 3116 Determination DOE/SRS-WD-2014-001 

for Closure of H-Tank Farm Revision 0 

at the Savannah River Site December 2014 

Page iii 

 

TABLE OF CONTENTS 

Page 

REVISION SUMMARY ................................................................................................................................. ii

LIST OF TABLES ........................................................................................................................................ vi

LIST OF FIGURES ..................................................................................................................................... vii

APPENDICES .............................................................................................................................................. ix

ACRONYMS / ABREVIATIONS ................................................................................................................... x

1.0 INTRODUCTION AND PURPOSE ............................................................................................... 1-1

1.1 Introduction ................................................................................................................................... 1-1

1.2 Purpose and Scope ....................................................................................................................... 1-4

1.3 Schedule and Plans for Closing Waste Tanks .............................................................................. 1-4

1.4 NDAA Section 3116(a) Consultation Summary ............................................................................ 1-5

1.5 Outline of HTF 3116 Basis Document .......................................................................................... 1-7

2.0 BACKGROUND ............................................................................................................................ 2-1

2.1 Savannah River Site and H-Tank Farm Facility Overview ............................................................ 2-1

2.1.1 Geography and Demography ........................................................................................... 2-1

2.1.2 Meteorology and Climatology .......................................................................................... 2-6

2.1.3 Ecology ............................................................................................................................ 2-8

2.1.4 Geology, Seismology and Volcanology ......................................................................... 2-10

2.1.5 Hydrogeology ................................................................................................................. 2-17

2.1.6 Geochemistry ................................................................................................................. 2-28

2.1.7 Natural Resources ......................................................................................................... 2-28

2.1.8 Natural and Background Radiation ................................................................................ 2-30

2.1.9 Tank Farm Operations ................................................................................................... 2-31

2.1.10 H-Tank Farm .................................................................................................................. 2-32

2.1.11 Waste Tanks .................................................................................................................. 2-33

2.1.12 Ancillary Structures ........................................................................................................ 2-47

2.2 H-Tank Farm Wastes .................................................................................................................. 2-54

2.2.1 Sources of the Waste ..................................................................................................... 2-55

2.2.2 Waste Management ....................................................................................................... 2-57

2.3 Waste Removal Approach for Waste Tanks, Annuli and Ancillary Structures ............................ 2-65

2.3.1 Tank Waste Removal History ........................................................................................ 2-65

2.3.2 Waste Removal Technologies ....................................................................................... 2-66

2.3.3 Chemical Cleaning ......................................................................................................... 2-68

2.3.4 Vacuum Heel Removal Cleaning ................................................................................... 2-69

2.3.5 Removal of Residual Waste from Failed Cooling Coils ................................................. 2-71

2.3.6 Annulus Cleaning ........................................................................................................... 2-71

2.3.7 Potential Future Waste Tank Cleaning Technology ...................................................... 2-72

2.3.8 Ancillary Structures Cleaning ......................................................................................... 2-72

2.4 Radionuclide Inventory in H-Tank Farm Facility Systems, Structures and Components ........... 2-74

2.4.1 Residual Inventory for Waste Tank Primary Tanks ....................................................... 2-74

2.4.2 Residual Inventory for Tank Annuli ................................................................................ 2-77

2.4.3 Residual Inventory for Type II Sand Layers ................................................................... 2-79

2.4.4 Residual Inventory for Failed Cooling Coils and Internal Waste Tank Surfaces ........... 2-79

2.4.5 Residual Waste Inventory for Transfer Lines, Pump Tanks and Catch Tanks and 

Evaporator Systems ....................................................................................................... 2-80
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2.5 Residual Waste Stabilization ...................................................................................................... 2-80

2.6 Closure Cap ................................................................................................................................ 2-82

2.6.1 Function of Closure Cap Layers .................................................................................... 2-87

3.0 SECTION 3116 OF THE RONALD W. REAGAN NATIONAL DEFENSE 

AUTHORIZATION ACT FOR FISCAL YEAR 2005 ..................................................................... 3-1

4.0 WASTE DOES NOT REQUIRE PERMANENT ISOLATION IN A DEEP GEOLOGIC 

REPOSITORY FOR SPENT FUEL OR HIGH-LEVEL RADIOACTIVE WASTE ......................... 4-1

5.0 WASTE HAS HAD HIGHLY RADIOACTIVE RADIONUCLIDES REMOVED TO THE 

MAXIMUM EXTENT PRACTICAL ............................................................................................... 5-1

5.1 Highly Radioactive Radionuclides ................................................................................................. 5-1

5.1.1 Methodology ..................................................................................................................... 5-1

5.1.2 Performance Assessment Radionuclides ........................................................................ 5-3

5.1.3 Highly Radioactive Radionuclides Based on 100-Meter Groundwater Analysis 

(For Member of the Public Following Closure) ................................................................ 5-4

5.1.4 Highly Radioactive Radionuclides Based on Air Pathway Analysis (For Member 

of the Public Following Closure) ...................................................................................... 5-5

5.1.5 Highly Radioactive Radionuclides Based on Intruder Pathway Analysis ........................ 5-6

5.1.6 Highly Radioactive Radionuclides Based on Uncertainty and Sensitivity Analyses ........ 5-7

5.1.7 Highly Radioactive Radionuclides Summary ................................................................... 5-8

5.2 Removal of Highly Radioactive Radionuclides to the Maximum Extent Practical ...................... 5-10

5.2.1 Current Status of H-Tank Farm Waste Removal Activities ............................................ 5-11

5.2.2 Waste Removal Technologies ....................................................................................... 5-11

5.2.3 Optimization of Existing Technologies ........................................................................... 5-19

5.3 Removal to the Maximum Extent Practical ................................................................................. 5-19

5.4 Conclusion .................................................................................................................................. 5-29

6.0 RADIONUCLIDE CONCENTRATIONS OF STABILIZED RESIDUALS, TANKS AND 

ANCILLARY STRUCTURES ........................................................................................................ 6-1

6.1 Waste Concentrations ................................................................................................................... 6-2

6.2 Approach to Waste Concentrations for H-Tank Farm Residuals .................................................. 6-3

6.3 Methodology .................................................................................................................................. 6-4

6.3.1 Methodology Inputs .......................................................................................................... 6-5

6.3.2 Site-Specific HTF Waste Concentration Calculation Averaging Expressions ................. 6-5

6.4 Waste Concentration Calculations .............................................................................................. 6-11

6.4.1 Waste Tank Concentration Calculation.......................................................................... 6-11

6.4.2 Transfer Line Concentration Calculation ....................................................................... 6-12

6.5 Conclusion .................................................................................................................................. 6-14

7.0 THE WASTE WILL BE DISPOSED OF IN ACCORDANCE WITH THE PERFORMANCE 

OBJECTIVES SET OUT IN 10 CFR 61, SUBPART C ................................................................ 7-1

7.1 10 CFR 61.41 ................................................................................................................................ 7-4

7.1.1 General Approach ............................................................................................................ 7-4

7.1.2 Public Release Pathways Dose Analysis ........................................................................ 7-4

7.1.3 Results of the Analyses ................................................................................................... 7-8

7.1.4 As Low As Reasonably Achievable ............................................................................... 7-12

7.1.5 Conclusion ..................................................................................................................... 7-13

7.2 10 CFR 61.42 .............................................................................................................................. 7-13

7.2.1 General Approach .......................................................................................................... 7-13

7.2.2 Acute Intruder-Drilling Scenario ..................................................................................... 7-13
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7.2.3 Chronic Intruder-Agricultural (Post-Drilling) Scenario .................................................... 7-14

7.2.4 Intruder Release Pathways Dose Analysis .................................................................... 7-14

7.2.5 Results of the Analysis ................................................................................................... 7-17

7.2.6 Conclusion ..................................................................................................................... 7-20

7.3 10 CFR 61.43 .............................................................................................................................. 7-20

7.3.1 Air Emissions Limit for Individual Member of the Public (10 CFR 20.1101(d)) .............. 7-21

7.3.2 Total Effective Dose Equivalent Limit for Adult Workers (10 CFR 20.1201(a)(1)(i)) ..... 7-22

7.3.3 Any Individual Organ or Tissue Dose Limit for Adult Workers (10 CFR 

20.1201(a)(1)(ii)) ............................................................................................................ 7-22

7.3.4 Annual Dose Limit to the Lens of the Eye for Adult Workers (10 CFR 

20.1201(a)(2)(i)) ............................................................................................................. 7-22

7.3.5 Annual Dose Limit to the Skin of the Whole Body and to the Skin of the 

Extremities for Adult Workers (10 CFR 20.1201(a)(2)(ii)) .............................................. 7-23

7.3.6 Limit on Soluble Uranium Intake (10 CFR 20.1201(e)) .................................................. 7-23

7.3.7 Dose Equivalent to an Embryo/Fetus (10 CFR 20.1208(a)) .......................................... 7-23

7.3.8 Total Effective Dose Equivalent Limit for Individual Members of the Public (10 

CFR 20.1301(a)(1)) ........................................................................................................ 7-24

7.3.9 Dose Limits for Individual Members of the Public in Unrestricted Areas (10 CFR 

20.1301(a)(2)) ................................................................................................................ 7-24

7.3.10 Dose Limits for Individual Members of the Public in Controlled Areas (10 CFR 

20.1301(b)) ..................................................................................................................... 7-25

7.3.11 As Low As Reasonably Achievable (10 CFR 20.1003) ................................................. 7-25

7.3.12 Reasonable Assurance .................................................................................................. 7-25

7.4 10 CFR 61.44 .............................................................................................................................. 7-28

7.4.1 Siting .............................................................................................................................. 7-28

7.4.2 Design ............................................................................................................................ 7-29

7.4.3 Use/Operation ................................................................................................................ 7-29

7.4.4 Closure ........................................................................................................................... 7-29

7.4.5 Conclusion ..................................................................................................................... 7-30

8.0 STATE-APPROVED CLOSURE PLAN ....................................................................................... 8-1

8.1 State-Approved Closure Plan ....................................................................................................... 8-1

8.2 Conclusion .................................................................................................................................... 8-2

9.0 CONCLUSION .............................................................................................................................. 9-1

10.0 REFERENCES ............................................................................................................................ 10-1

11.0 GLOSSARY ................................................................................................................................ 11-1
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LIST OF TABLES 

Page 

Table 2.1-1: Population Distribution and Percent of Region of Influence (% ROI) for Counties and 

Selected Communities ....................................................................................................................... 2-6

Table 2.1-2: Characterization and Monitoring Data in the GSAD ........................................................... 2-25

Table 2.1-3: Water Quality in the Savannah River Upstream and Downstream from SRS (Calendar 

Year 2010) ....................................................................................................................................... 2-29

Table 2.1-4: Water Quality in Selected SRS Streams (Calendar Year 2010) ........................................ 2-29

Table 2.1-5: Well Monitoring Results for Major Areas within SRS, 2007–2008 ..................................... 2-30

Table 2.1-6: Waste Tank Nominal Capacities, Nominal Dimensions and Other Features ..................... 2-33

Table 2.1-7: Type II Tanks Primary Tank Liner Plate Data ..................................................................... 2-37

Table 2.1-8: Type III/IIIA Tanks Primary Tank and Secondary Liner Plate Data .................................... 2-42

Table 2.4-1: Waste Tank Groupings ....................................................................................................... 2-77

Table 2.6-1: HTF Closure Cap Layers .................................................................................................... 2-88

Table 2.6-2: Function of the HTF Closure Cap Layers ........................................................................... 2-89

Table 5.1-1: HTF Highly Radioactive Radionuclides ................................................................................ 5-2

Table 5.1-2: Radionuclides Contained in HTF PA Initial Inventory ........................................................... 5-8

Table 6.1-1: 10 CFR 61.55 Table 1 Class C Concentration Limits ........................................................... 6-2

Table 6.1-2: 10 CFR 61.55 Table 2 Class C Concentration Limits ........................................................... 6-3

Table 6.3-1: HTF Waste Tank Site-Specific Factors for 10 CFR 61.55 Table 1 Radionuclides ............... 6-9

Table 6.3-2: HTF Waste Tank Site-Specific Factors for 10 CFR 61.55 Table 2 Radionuclides ............. 6-10

Table 6.3-3: HTF Transfer Line Site-Specific Factors for 10 CFR 61.55 Table 1 Radionuclides ........... 6-10

Table 6.3-4: HTF Transfer Line Site-Specific Factors for 10 CFR 61.55 Table 2 Radionuclides ........... 6-11

Table 6.4-1: Sum of the Fractions Calculation Using the HTF PA, Revision 1 Inventory for Tank 32 

(10 CFR 61.55 Table 1 Radionuclides) ........................................................................................... 6-12

Table 6.4-2: Sum of the Fractions Calculation Using the HTF PA, Revision 1 Inventory for Tank 32 

(10 CFR 61.55 Table 2 Radionuclides) ........................................................................................... 6-12

Table 6.4-3: Sum of the Fractions Calculation Using the HTF PA, Revision 1 Inventory for a 

Three-inch Transfer Line (10 CFR 61.55 Table 1 Radionuclides) .................................................. 6-13

Table 6.4-4: Sum of the Fractions Calculation Using the HTF PA, Revision 1 Inventory for a 

Three-inch Transfer Line (10 CFR 61.55 Table 2 Radionuclides) .................................................. 6-14

Table 7.3-1: Crosswalk Between Applicable 10 CFR 20 Standards and DOE Requirements ............... 7-21
Basis for Section 3116 Determination DOE/SRS-WD-2014-001 

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LIST OF FIGURES 

Page 

Figure 2.1-1: Physical Location of SRS .................................................................................................... 2-2

Figure 2.1-2: Location of SRS and Adjacent Areas .................................................................................. 2-3

Figure 2.1-3: Predominant SRS Operational Area Location Map ............................................................. 2-4

Figure 2.1-4: Layout of the GSA ............................................................................................................... 2-5

Figure 2.1-5: General Layout of HTF ........................................................................................................ 2-5

Figure 2.1-6: SRS Meteorological Monitoring Network ............................................................................. 2-7

Figure 2.1-7: Regional Geological Provinces of Eastern United States ................................................. 2-10

Figure 2.1-8: Regional Physiographic Provinces of South Carolina ....................................................... 2-11

Figure 2.1-9: GSA Topography ............................................................................................................... 2-11

Figure 2.1-10: General Soil Associations for SRS .................................................................................. 2-12

Figure 2.1-11: Historical Seismic Events in the Southeast ..................................................................... 2-14

Figure 2.1-12: Seismic Events within a 50-Mile Radius of SRS ............................................................. 2-15

Figure 2.1-13: Regional Scale Faults for SRS and Vicinity .................................................................... 2-16

Figure 2.1-14: Regional NW to SE Cross Section .................................................................................. 2-18

Figure 2.1-15: Comparison of Chronostratigraphic, Lithostratigraphic, and Hydrostratigraphic Units 

in the SRS Region ........................................................................................................................... 2-19

Figure 2.1-16: Potentiometric Surface of the UTRA ............................................................................... 2-20

Figure 2.1-17: Potentiometric Surface of the Gordon Aquifer ................................................................. 2-21

Figure 2.1-18: Hydrostratigraphic Picks in GSAD ................................................................................... 2-22

Figure 2.1-19: Laboratory Determined Permeability Data in GSAD ....................................................... 2-23

Figure 2.1-20: Multiple Well Pump Test Data in GSAD .......................................................................... 2-23

Figure 2.1-21: Single Well Pump Test Data in GSAD ............................................................................ 2-24

Figure 2.1-22: Slug Test Data in GSAD .................................................................................................. 2-24

Figure 2.1-23: Conceptual Diagram of Groundwater Flow beneath the GSA ........................................ 2-25

Figure 2.1-24: Savannah River Basin Dams ........................................................................................... 2-26

Figure 2.1-25: SRS Watershed Boundaries and Major Tributaries ........................................................ 2-27

Figure 2.1-26: GSA Gauging Stations .................................................................................................... 2-27

Figure 2.1-27: Major Sources of Radiation Exposure in the Vicinity of SRS .......................................... 2-30

Figure 2.1-28: General Layout of HTF .................................................................................................... 2-32

Figure 2.1-29: Typical Type I Tank Cross Section .................................................................................. 2-34

Figure 2.1-30: Type I Tank Basemat Construction ................................................................................. 2-35

Figure 2.1-31: Type I Tank Primary and Secondary Liner Construction ................................................ 2-35

Figure 2.1-32: Typical Type I Tank Column Support .............................................................................. 2-35

Figure 2.1-33: Typical Type I Tank Cooling Coil Obstacles .................................................................... 2-36

Figure 2.1-34: Type I Waste Access Area for Waste Removal Equipment Diagram ............................. 2-36

Figure 2.1-35: Typical Type II Tank Cross Section ................................................................................. 2-37

Figure 2.1-36: Type II Tank Basemat Construction ................................................................................ 2-37

Figure 2.1-37: Type II Tank Primary and Secondary Liner Construction ............................................... 2-37

Figure 2.1-38: Soil Hydration System Below Type II Tanks ................................................................... 2-38
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Figure 2.1-39: Typical Type II Support Column Dimension Details ........................................................ 2-38

Figure 2.1-40: Typical Type II Tank Cooling Coil Obstacles ................................................................... 2-39

Figure 2.1-41: Type II Waste Access Area for Waste Removal Equipment Diagram ............................ 2-39

Figure 2.1-42: Typical HTF Type III Tank Cross Section ....................................................................... 2-40

Figure 2.1-43: Typical Type IIIA Tank Cross Section ............................................................................. 2-40

Figure 2.1-44: Typical Type III/IIIA Tank Basemat Construction ............................................................ 2-41

Figure 2.1-45: Typical Type IIIA Tank Primary Tank and Secondary Liner - Late Construction (Vault 

Wall Not Constructed) ..................................................................................................................... 2-41

Figure 2.1-46: Typical Type III/IIIA Tank Radial Air Grooves ................................................................. 2-42

Figure 2.1-47: Typical Type III Tank Insertable Cooling Coils ................................................................ 2-43

Figure 2.1-48: Typical Type IIIA Tank Cooling Coils During Construction ............................................ 2-43

Figure 2.1-49: Typical Type III Tank Access Area for Waste Removal Equipment Diagram ................. 2-44

Figure 2.1-50: Typical Type IIIA Tank Access Area for Waste Removal Equipment Diagram ............... 2-44

Figure 2.1-51: Typical HTF Type IV Tank Cross Section ....................................................................... 2-44

Figure 2.1-52: Typical Type IV Tank Stiffener Angles ............................................................................ 2-45

Figure 2.1-53: Typical Type IV Tank Shotcrete Application ................................................................... 2-46

Figure 2.1-54: Typical Type IV Tank Domes and Risers During Construction ....................................... 2-46

Figure 2.1-55: Typical Type IV Tank Access Area for Waste Removal Equipment Diagram ................. 2-46

Figure 2.1-56: General Layout of HTF .................................................................................................... 2-47

Figure 2.1-57: HTF Waste Tank Transfer Line During Construction ...................................................... 2-48

Figure 2.1-58: Interior View of HPP-3 ..................................................................................................... 2-48

Figure 2.1-59: Typical Diversion Box and Pump Pit Design ................................................................... 2-49

Figure 2.1-60: Typical Catch Tank Cross Section .................................................................................. 2-49

Figure 2.1-61: 242-H Evaporator System Schematic ............................................................................. 2-50

Figure 2.1-62: 242-16H Evaporator System Schematic ......................................................................... 2-51

Figure 2.1-63: 242-25H Evaporator System Schematic ......................................................................... 2-52

Figure 2.2-1: Saltcake at Bottom of Waste Tank .................................................................................... 2-56

Figure 2.3-1: Typical Standard Slurry Pump ........................................................................................... 2-65

Figure 2.3-2: Submersible Mixer Pump Waste Removal Diagram ......................................................... 2-67

Figure 2.3-3: Submersible Mixing Pump and Submersible Transfer Pump ............................................ 2-68

Figure 2.3-4: Submersible Mixing Pump in a Test Tank ......................................................................... 2-68

Figure 2.3-5: Mantis ................................................................................................................................ 2-70

Figure 2.3-6: Waste Deposits on Exterior Primary Tank Wall Before Cleaning (Tank 12) ..................... 2-71

Figure 2.3-7: Waste Deposits on Annulus Floor Before Cleaning (Tank 9) ............................................ 2-71

Figure 2.6-1: HTF Closure Cap General Concept .................................................................................. 2-83

Figure 2.6-2: HTF Closure Cap Conceptual Design, Cap Footprint and Grading Cap .......................... 2-84

Figure 2.6-3: HTF Closure Cap Conceptual Design, Sections A-A and B-B .......................................... 2-85

Figure 2.6-4: HTF Closure Cap Conceptual Design, Sections C-C and D-D ......................................... 2-86

Figure 2.6-5: HTF Closure Cap Conceptual Design, Sections E-E and F-F .......................................... 2-87

Figure 5.2-1: HTF Waste Tank Composite Inventory ............................................................................. 5-11

Figure 5.2-2: Typical Standard Slurry Pump ........................................................................................... 5-13

Figure 5.2-3: Submersible Mixer Pump Waste Removal Diagram ......................................................... 5-14
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Figure 5.2-4: Submersible Mixing Pump and Submersible Transfer Pump ............................................ 5-15

Figure 5.2-5: Submersible Mixing Pump in a Test Tank ......................................................................... 5-15

Figure 5.2-6: Mantis ................................................................................................................................ 5-18

Figure 5.2-7: Robotic Sampler at Test Facility ........................................................................................ 5-19

Figure 5.3-1: Tank 16 Waste Removal ................................................................................................... 5-20

Figure 5.3-2: Tank 12 Waste Removal ................................................................................................... 5-22

Figure 5.3-3: Tank 18 Waste Removal ................................................................................................... 5-24

Figure 5.3-4: Tank 19 Waste Removal ................................................................................................... 5-24

Figure 5.3-5: Tank 5 Waste Removal ..................................................................................................... 5-26

Figure 5.3-6: Tank 6 Waste Removal ..................................................................................................... 5-27

Figure 6.2-1: Intruder-Driller Scenarios for Concentration Calculations ................................................... 6-3

Figure 7.0-1: HTF Boundary and Buffer Zone .......................................................................................... 7-3

Figure 7.1-1: Scenario with Well Water as Primary Source ...................................................................... 7-6

Figure 7.1-2: Peak All-Pathways Dose to a Member of the Public Within 10,000 Years (Base Case 

Deterministic Analysis) ...................................................................................................................... 7-8

Figure 7.1-3: Peak All-Pathways Dose to a Member of the Public Within 100,000 Years (Base Case 

Deterministic Analysis) ...................................................................................................................... 7-9

Figure 7.1-4: Statistical Summary of Time History of Total Member of the Public Dose, at the Well of 

Maximum Concentration, Within 10,000 Years (All Cases Probabilistic Analysis) ......................... 7-10

Figure 7.1-5: Statistical Summary of Time History of Total Member of the Public Dose, at the Well of 

Maximum Concentration Within 100,000 Years (All Cases Probabilistic Analysis) ........................ 7-11

Figure 7.2-1: Acute Intruder-Drilling Scenario ......................................................................................... 7-15

Figure 7.2-2: Chronic Intruder-Agricultural (Post-Drilling) Scenario ....................................................... 7-16

Figure 7.2-3: Peak Acute Dose to the Inadvertent Intruder (Base Case Deterministic Analysis) ........... 7-18

Figure 7.2-4: Peak Chronic Dose to the Inadvertent Intruder (Base Case Deterministic Analysis) ....... 7-19

APPENDICES 

Page 

Appendix A: Liquid Waste System Description ......................................................................................... A-1 

Appendix B: Approach to Documenting Removal of Radionuclides to Support DOE Closure 

Authorization .................................................................................................................................... B-1 

Appendix C: H-Tank Farm Performance Assessment Dose Summary ................................................... C-1 
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ACRONYMS / ABREVIATIONS 

ALARA As Low As Reasonably Achievable 

ARP Actinide Removal Process 

CA Composite Analysis 

CFR Code of Federal Regulations 

Ci curie(s)

CLM Central Climatology 

cm centimeter(s) 

CSRA Central Savannah River Area 

CSSX Caustic Side Solvent Extraction

CTS Concentrate Transfer System 

DDA Deliquification, Dissolution and Adjustment

DOE United States Department of Energy

DOE-SR United States Department of Energy-Savannah River Operations Office 

DSA Documented Safety Analysis 

DWPF Defense Waste Processing Facility

Eh Measure of reduction (or oxidation) potential

EIS Environmental Impact Statement

EPA United States Environmental Protection Agency 

ETF Effluent Treatment Facility

o

F degree Fahrenheit 

FEPs Features, Events and Processes 

FFA Federal Facility Agreement

ft foot 

FTF F-Tank Farm

g gram(s) 

GCL Geosynthetic Clay Liner 

GCP General Closure Plan 

GSA General Separations Area

GSAD General Separations Area Database 

GWSB Glass Waste Storage Building

HA Hazard Analysis 

HAW High Activity Waste

HDPE High Density Polyethylene 

HELP Hydrologic Evaluation of Landfill Performance 

HLW High-Level Waste 

HM H-Modified

HEPA High Efficiency Particulate Air 

hr hour(s) 

HRR Highly Radioactive Radionuclide 

HTF H-Tank Farm

ISWLF Industrial Solid Waste Landfill Facility 

ITP In-Tank Precipitation 

Kd Distribution Coefficient 

kV kilovolt 

L liter(s) 

LAW Low Activity Waste 
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m meter(s) 

MCi million curie(s) 

MCU Modular Caustic Side Solvent Extraction Unit 

MEP Maximum Extent Practical 

mg milligram(s) 

Mgal million gallon(s) 

mrem millirem(s) 

MSL Mean Sea Level 

mSv milliSievert(s) 

NAS National Academy of Sciences 

NASA National Aeronautics and Space Administration 

nCi nanocurie(s)

NCRP National Council on Radiation Protection and Measurements 

NDAA Ronald W. Reagan National Defense Authorization Act for Fiscal Year 2005 

NEPA National Environmental Policy Act 

NPDES National Pollutant Discharge Elimination System 

NRC United States Nuclear Regulatory Commission 

NRMP Natural Resources Management Plan 

OD Outside Diameter 

OSHA Occupational Safety and Health Administration 

PA Performance Assessment 

PC Performance Category 

pCi picocurie(s)

PEL Permissible Exposure Limit 

pH Measure of acidity or alkalinity of a solution 

ppb parts per billion 

psi pounds per square inch 

PUREX Plutonium Uranium Extraction 

RBA Radiological Buffer Area 

RBOF Receiving Basin for Offsite Fuels 

ROD Record of Decision 

ROI Region of Influence 

RPP Radiation Protection Program 

RRF Resin Regeneration Facility 

SCDHEC South Carolina Department of Health and Environmental Control 

SCGNAC South Carolina Governor’s Nuclear Advisory Council 

SDF Saltstone Disposal Facility 

SMP Submersible Mixer Pump 

SPF Saltstone Production Facility 

S/RID Standards/Requirement Identification Document 

SREL Savannah River Ecology Laboratory 

SRNL Savannah River National Laboratory 

SRS Savannah River Site 

STP Submersible Transfer Pump 

SWPF Salt Waste Processing Facility 

TED Total Effective Dose 

TEDE Total Effective Dose Equivalent 
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TER Technical Evaluation Report 

U.S. United States 

UTR Upper Three Runs 

UTRA Upper Three Runs Aquifer 

UTRA-LZ Upper Three Runs Aquifer-Lower Zone 

UTRA-UZ Upper Three Runs Aquifer-Upper Zone 

WCS Waste Characterization System 

WD Waste Determination 

WMC Waste Mixing Chamber 

yr year(s) 
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1.0 INTRODUCTION AND PURPOSE 

Section Purpose 

This section provides the purpose and scope of this document, titled Basis for Section 

3116 Determination for Closure of H-Tank Farm at the Savannah River Site (hereinafter 

referred to as: HTF 3116 Basis Document). 

Section Contents 

This section contains a brief introduction to the Savannah River Site (SRS) and H-Tank 

Farm (HTF) and describes the purpose and scope of this HTF 3116 Basis Document. 

Key Points 

 The United States Department of Energy (DOE) is issuing this HTF 3116 Basis 

Document to provide a basis for a determination by the Secretary of Energy, in 

consultation with the United States Nuclear Regulatory Commission (NRC), pursuant 

to Section 3116 of the Ronald W. Reagan National Defense Authorization Act (NDAA) 

for Fiscal Year 2005 (hereinafter referred to as: NDAA Section 3116). [NDAA_3116] 

 This HTF 3116 Basis Document concerns stabilized residuals in waste tanks1 and 

ancillary structures, those waste tanks, and the ancillary structures (including integral 

equipment) at the SRS HTF at the time of closure. 

 The HTF is a 45-acre site consisting of underground radioactive waste storage tanks 

and supporting ancillary structures. 

 The HTF tank waste storage and removal operations are performed in accordance with 

a State-issued industrial wastewater construction permit. Removal from service and 

stabilization of the HTF waste tanks and ancillary structures will be carried out 

pursuant to a State-approved closure plan and will be consistent with the SRS Federal 

Facility Agreement (FFA). [WSRC-OS-94-42] 

 After completion of waste removal activities, the HTF waste tanks will be stabilized by 

filling the waste tanks with grout. Ancillary structures will be filled, as necessary, to 

prevent subsidence. 

 Stabilization of individual HTF waste tanks and ancillary structures is anticipated to 

take place after individual component cleaning is complete.

 This HTF 3116 Basis Document takes into account comments received from the NRC 

during consultation with DOE under NDAA Section 3116(a), as well as comments from 

the United States Environmental Protection Agency (EPA) and the State of South 

Carolina Governor’s Nuclear Advisory Council (SCGNAC) that were submitted as part 

of the public comment opportunity. No other comments were received from the 

public.

1.1 Introduction 

In accordance with NDAA Section 3116, certain waste from reprocessing of spent nuclear fuel is not highlevel waste if the Secretary of Energy, in consultation with the NRC, determines that the criteria in NDAA 

Section 3116(a) are met. This HTF 3116 Basis Document shows that those criteria are satisfied, to 

support a potential determination that the Secretary may make pursuant to NDAA Section 3116. This 

HTF 3116 Basis Document concerns the stabilized residuals in waste tanks and ancillary structures, 

those waste tanks, and the ancillary structures (including integral equipment) at the SRS HTF at the time 

of closure. 

The HTF is a 45-acre site containing 29 underground radioactive waste storage tanks2

 and supporting 

ancillary structures, such as evaporators, pump pits, pump tanks, transfer valve boxes and piping. Most 

 

1

 The HTF has 29 waste tanks and cleaning and waste removal efforts are well underway for several of the waste tanks. All 29 

waste tanks in HTF (and their residuals) are addressed by this HTF 3116 Basis Document, and, unless otherwise specified, 

references in this document to “the waste tanks” or “HTF waste tanks” refers to all 29 of the HTF waste tanks. 
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of the waste in these waste tanks and ancillary structures originated in the SRS H-Canyon Separations 

Facility, which, in the past, primarily reprocessed used uranium fuel for the recovery of uranium but also 

recovered other nuclear materials produced in the site’s nuclear production reactors. H Canyon 

continues to generate radioactive liquid waste when performing stabilization missions, such as recovering 

and blending highly enriched uranium for non-defense related use. DOE carefully controls waste 

transferred to the HTF waste tanks and ancillary structures through a waste compliance program and 

waste acceptance criteria which establish the physical, chemical and radionuclide limits for all waste 

entering the waste tanks and ancillary structures.3

 

The DOE is in the process of closing HTF, and is engaged in an expansive campaign to clean, stabilize 

and close all HTF underground, radioactive waste storage tanks, as well as supporting ancillary 

structures (evaporators, pump pits, pump tanks, diversion boxes, transfer valve boxes and piping), used 

to store, treat and transfer waste generated, in part, by prior reprocessing of spent nuclear fuel. The 

waste tanks and ancillary structures are several decades old, and several of the aging HTF waste tanks 

are approaching 60 years of service life – well beyond their design life.4

 Given the risks inherent in 

exhuming the aging waste tanks, DOE is instead pursuing a campaign to clean, stabilize and close the 

HTF waste tanks and ancillary structures in place, to reduce the risks to the workers, the public and the 

environment associated with the aging waste tanks and ancillary structures at HTF. 

To date, DOE has over two decades of experience in successfully cleaning waste tanks of different types 

at SRS, using a variety of mechanical and chemical technologies. The DOE’s experience has 

encompassed some of the largest and most challenging waste tanks at SRS, including waste tanks that 

contain the most complex internal structures (cooling coil arrays) and the most challenging zeolite-laden 

wastes.5

 The cleaning campaigns have spanned a number of years and involved multiple phases, and 

have included some of the most difficult waste tanks at SRS.6 Following cleaning, waste “residuals”, that 

cannot be practically removed, remain on the floors and internal surfaces of the waste tanks and ancillary 

structures. After cleaning, the waste tanks and applicable ancillary structures, along with their residuals, 

are stabilized in place with a grout matrix to prevent subsidence. At HTF, waste removal and cleaning 

efforts are well underway for several waste tanks. These waste removal efforts will continue for the waste 

tanks and ancillary structures for which cleaning has not been completed, to be followed by grouting. 

Based on extensive experience in cleaning the SRS waste tanks, DOE is confident that, although 

additional waste removal efforts remain to be completed, DOE can appropriately demonstrate at this time 

that the criteria in NDAA Section 3116(a) will be met at closure for the HTF waste tanks, ancillary 

structures and residuals.7

 

 2 The HTF waste tanks have capacities ranging from 750,000 gallons to 1,300,000 gallons, depending on the waste tank, as discussed further in 

Section 2.0 of this HTF 3116 Basis Document. Eight waste tanks (Tanks 9 through 16), including tanks which have leaked in the past, along with some 

ancillary structures, are either fully submerged or partially submerged in the water table. 

3

 All generating facilities, including H Canyon, are required to develop and maintain a Waste Compliance Plan which describes the controls or 

procedures imposed within the facility to ensure that the HTF waste acceptance criteria are met for any material transferred to the HTF waste tanks 

and ancillary structures. [X-SD-G-00001] 

4

 Some of these waste tanks do not have a secondary containment structure (called an annulus) and some have developed leaks in the primary tanks, 

which have leaked into the secondary containment annuli. To date, waste has escaped to the surrounding soils due to a leak site in one primary tank. 

This event occurred in 1960 and was associated with Tank 16 in HTF. [DPSOX-5954] 5

 Zeolite was used in the past to capture cesium and is present from spent zeolite resin in some of the HTF waste tanks. The presence of zeolite 

poses significant challenges to removal of waste from those waste tanks, due to chemical changes to the zeolite resin over time, the relatively large 

particle size and high density of zeolite that results in quick settling, and the negative effect of cleaning acids on the zeolite, which causes 

conglomerates that cannot be readily removed from the waste tank and may retain radionuclides. 6

 In HTF, the DOE has conducted heel removal and annulus cleaning activities for the primary tank and annulus for Tank 16 and for the Tank 12 

primary tank and has completed bulk waste removal efforts on Tank 11. In addition, bulk waste removal efforts are currently underway in Tank 13 and 

Tank 10. DOE also has successfully removed sludge and saltcake from several other HTF waste tanks; although available space in those tanks has 

been re-filled and the tanks are or will be cleaned further for closure, this previous experience illustrates DOE’s expertise in removing waste from HTF 

tanks. DOE also has extensive waste removal experience for F-Tank Farm (FTF) waste tanks, which provides additional indicia of DOE’s experience 

and expertise with waste removal at SRS, although the FTF waste tanks are not within the scope of this HTF 3116 Basis Document. In FTF, DOE has 

cleaned six waste tanks (Tanks 17, 20, 18, 19, 5 and 6) and completed bulk waste removal efforts on Tanks 4, 7 and 8. DOE’s experience to date has 

encompassed several designs or types of waste tanks. In this regard, the HTF contains four basic designs of waste tanks, called Type I, Type II, Type 

IV, and Type III/IIIA tanks, whereas FTF has Types I, IV and III/IIIA tanks. In total, DOE has completed cleaning of four Type IV tanks, three Type I 

tanks and one Type II tank, and has completed bulk waste removal efforts on an additional four Type I tanks (Tank 11 in HTF and Tanks 4, 7 and 8 in 

FTF). Waste removal is primarily influenced by tank type and waste type (e.g., liquid and solid salt waste versus sludge), and the factors are not 

primarily tank farm dependent. 

7

 Moreover, the Secretary of Energy, in consultation with the NRC, previously determined that similar stabilized residuals, waste tanks and ancillary 

structures at closure in FTF at SRS are not high-level radioactive waste and may be disposed of in place at SRS. [DOE-WD-2012-001] Considering 

the high level of interdependency between the HTF and FTF and the importance of continuing to move forward with reducing the risk associated with 

the aging waste tanks and ancillary structures in both tank farms, DOE believes it is imperative at this time for the Secretary of Energy, in consultation 

with the NRC, to determine whether the criteria in Section 3116(a) are met for HTF at closure, so as to make progress as expeditiously as possible 

toward the closure of HTF, in a manner which is protective of human health and safety as well as the environment.
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The DOE is predicating this HTF 3116 Basis Document on extensive analyses and scientific rationale, 

including the Performance Assessment for the H-Area Tank Farm at the Savannah River Site, SRRCWDA-2010-00128 (hereinafter referred to as: HTF PA).8

 These analyses and rationale, including the 

HTF PA, demonstrate that there is reasonable assurance9 that the maximum all-pathways dose to any 

member of the public from closure of HTF will be well below the 25 mrem/yr performance objective 

referenced in NDAA Section 3116. For example, as explained further and graphically displayed in 

Section 7.0 of this HTF 3116 Basis Document, the HTF PA projects a maximum all-pathways dose to a 

member of the public of approximately 0.3 mrem/yr within 1,000 years after tank farm closure and 

approximately 4 mrem/yr within 10,000 years after tank farm closure, using the models and assumptions 

of the HTF PA Base Case.10 In addition, the HTF PA projects the peak inadvertent intruder (i.e., 

individual within the HTF boundary) doses to be less than 500 mrem/yr, thus demonstrating the 

performance objective referenced in NDAA Section 3116 regarding the inadvertent intruder will also be 

met. Consistent with the above, DOE is confident that the dose limits to the public and the inadvertent 

intruder, applicable under Section 3116(a), would not be exceeded. Under these circumstances and 

given DOE’s extensive waste removal experience noted above, DOE believes it is appropriate to proceed 

at this time with this HTF 3116 Basis Document, which, along with its supporting references, 

demonstrates that the criteria in NDAA Section 3116(a) are met for the waste tanks, ancillary structures 

and residuals at HTF at closure. 

On February 6, 2013, DOE issued the Draft Basis for Section 3116 Determination for Closure of H-Tank 

Farm at the Savannah River Site (hereinafter referred to as: Draft HTF 3116 Basis Document) for NRC 

consultative review, as part of DOE’s consultation with NRC under NDAA Section 3116. [DOE/SRS-WD2013-001] During review of the Draft HTF 3116 Basis Document, NRC requested additional information 

from DOE. [ML13196A135] The DOE responded to the NRC request for additional information. [SRRCWDA-2013-00106, SRR-CWDA-2013-00144] The DOE also provided other additional information as 

requested by the NRC, and DOE and NRC held six consultative telephone calls and three meetings 

between March 14, 2013 and August 29, 2013. [ML13086A080, ML13106A338, ML13120A496, 

ML13154A327, ML13193A072, ML13183A410, ML13260A275, ML13199A413]. 

On June 17, 2014, the NRC issued its Technical Evaluation Report For H-Area Tank Farm Facility, 

Savannah River Site, South Carolina. [ML14094A496] The Technical Evaluation Report (TER) presents 

NRC’s consultative observations and recommendations.11 The DOE has considered the NRC 

observations and recommendations in preparing this HTF 3116 Basis Document. [SRR-CWDA-2014-

00080] 

NDAA Section 3116(b) requires that the NRC, in coordination with the State of South Carolina, monitor 

disposal actions taken by DOE for the purposes of assessing compliance with the performance objectives 

of Title 10 Code of Federal Regulations (CFR) Part 61, Subpart C. In its TER, NRC staff states that the 

NRC will coordinate with the State of South Carolina to develop a plan by which NRC and the State of 

South Carolina will monitor DOE’s disposal actions. The DOE will work with the NRC and the State of 

South Carolina to facilitate their efforts to conduct monitoring under NDAA Section 3116(b). 

 

8

 A performance assessment is a multi-disciplined assessment, (e.g., geochemistry, hydrogeology, materials science, health 

physics) which uses a variety of computational modeling codes to evaluate groundwater concentrations and doses at various points 

of assessment over time. In doing this evaluation, DOE assesses the impact of natural features (e.g., hydrogeology, soil properties, 

groundwater infiltration) and engineered barriers (e.g., closure cap, fill grout, waste tank design) on the release of radionuclides, to 

estimate the potential dose to a hypothetical member of the public and a hypothetical inadvertent intruder. 9

 Section 3116 cross-references the NRC performance objectives in 10 CFR Part 61, Subpart C. The first performance objective at

10 CFR 61.40 sets forth a general requirement that “reasonable assurance” exists that exposures to humans are within the limits of 

the performance objectives in 10 CFR 61.41 through 61.44. The HTF PA was conducted and will be maintained in accordance with 

DOE Manual 435.1-1 and DOE guidance, pursuant to DOE’s authority and responsibilities to protect human health and safety under 

the Atomic Energy Act of 1954, as amended. DOE Manual 435.1-1 uses the phrase “reasonable expectation”, which is analogous 

to “reasonable assurance”. For convenience, this HTF 3116 Basis Document uses the phrase “reasonable assurance”. 10 See Section 7.0 of this HTF 3116 Basis Document for additional discussion. The results of the HTF PA, as reported here, should 

not be considered limits or thresholds. As required by DOE Manual 435.1-1, maintenance of the HTF PA will include future 

performance assessment revisions or special analyses to incorporate new information, update model codes and reflect analysis of

actual residual inventories. 

11 On page xiv of the TER, the NRC notes, “The NRC staff’s review results and recommendations are being provided to DOE for 

consideration only and are not intended to represent any regulatory authority related to DOE’s WD activities. The Secretary of 

Energy, in consultation with the NRC, is responsible for determining whether the waste streams addressed in the draft basis for the 

WD for HTF are not HLW, and therefore, satisfy the requirements in Section 3116 of the NDAA.” [ML14094A496] 
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Although not required by NDAA Section 3116, DOE also made available the Draft HTF 3116 Basis 

Document for a three-month public review and comment period beginning in February 2013. [FR Doc. 

2013-03305] Comments on the document were received from the EPA and the SCGNAC. [EPA_04-30-

2013, GNAC-OS-2013-04-19-01] These comments were considered in preparing this HTF 3116 Basis 

Document. Comment responses were prepared and provided to the EPA, and have been posted on the 

following websites:12 [SRR-CWDA-2014-00057] 

 http://sro.srs.gov/f_htankfarmsdocuments.htm 

 www.em.doe.gov 

No other comments were received. 

1.2 Purpose and Scope 

The purpose of this HTF 3116 Basis Document is to demonstrate and document that, after final 

stabilization activities are complete, the stabilized residuals in the HTF waste tanks and ancillary 

structures, the waste tanks, and the ancillary structures (including integral equipment) at the time of 

closure meet the NDAA Section 3116(a) criteria and, therefore are not high-level waste. 

The scope of this HTF 3116 Basis Document specifically addresses the stabilized residuals in the HTF 

waste tanks and ancillary structures, the waste tanks, and the ancillary structures (including integral 

equipment) at the time of HTF closure.13 This HTF 3116 Basis Document does not include other SRS 

facilities or systems, or waste removed from the waste tanks and ancillary structures. 

This HTF 3116 Basis Document is premised on the facts, assumptions and analyses contained or 

referenced herein. Accordingly, a NDAA Section 3116(a) Secretarial determination made in reliance on 

this HTF 3116 Basis Document only covers situations consistent with those facts, assumptions and 

analyses. 

1.3 Schedule and Plans for Closing Waste Tanks 

The HTF waste tanks14 are closed in accordance with the SRS FFA, a formal agreement between DOE, 

Region 4 of the EPA and the South Carolina Department of Health and Environmental Control 

(SCDHEC). The FFA establishes that, among other things, the SRS waste tanks that do not meet 

secondary containment standards (older style tanks, specifically the Type I, Type II and Type IV tanks in 

HTF) must be removed from service according to the FFA schedule. The current FFA calls for staggered 

operational closure of the twelve HTF and ten F-Tank Farm (FTF) older style waste tanks (tank numbers 

not specified in the SRS FFA) by September 2022.15 In this regard, the current FFA requires the 

operational closure of four additional waste tanks by September 2015, which likely will be split between 

HTF and FTF. [WSRC-OS-94-42] DOE addresses the closure of the remaining HTF waste tanks (Type 

III/IIIA tanks) and ancillary structures in the SRS Liquid Waste System Plan.16 [SRR-LWP-2009-00001] 

 

12 Due to the nature of the comments provided, DOE did not prepare a specific comment response document to address the 

SCGNAC comments. SCGNAC did not recommend any specific changes to the Draft HTF 3116 Basis Document nor any of the 

information contained within the document. In their comments, SCGNAC concludes, “...we believe that the conservative approach 

used by DOE and its contractors in developing (and the NRC in reviewing) this document ensures that tank closures will protect the 

environment, the public and the workforce under the H-Tank Farm closure plan.” [GNAC-0S-2013-04-19-01] 13 For the purpose of this HTF 3116 Basis Document, the residual waste remaining in a waste tank or ancillary structure – following 

successful completion of waste removal activities and removal of the highly radioactive radionuclides to the maximum extent practical – is 

referred to as “residuals.” Stabilization of these residuals within the HTF waste tanks will be carried out by filling the waste tanks with grout 

after completion of waste removal activities. Ancillary structures will be filled, as necessary, to prevent subsidence of the structure or final 

closure cap. The DOE does not plan to add fill material to the HTF transfer lines. 

14 The HTF contains four basic designs (types) of tanks, called Type I, Type II, Type IV, and Type III/IIIA tanks. The HTF waste tanks are 

also numbered, although not sequentially. The tank types are discussed in further detail in Section 2.0 of this HTF 3116 Basis Document. 15 The FFA includes procedures for revision of the FFA schedule. 16 The Liquid Waste System Plan is updated periodically as appropriate. 

According to the Performance Assessment for the H-Area Tank Farm at the Savannah River Site (SRR-CWDA-2010-00128), DOE’s 

planned sequence for closure of HTF is as follows: 

 Closure of Types I, II and IV tanks and finally the Type III and IIIA tanks. The ancillary structures (such as transfer lines) are planned to 

be closed as appropriate with a goal of closing HTF in stages. 

 Following closure of a geographic section (such as Type IV tanks and evaporator area), the section may be left in an interim closure 

state in preparation for final closure. 

 Following closure of all HTF waste tanks and ancillary structures, HTF will undergo final closure in accordance with the FFA. 
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The DOE will, pursuant to its authority, including that under the Atomic Energy Act of 1954, as amended, 

and applicable DOE Orders, manuals and policies, pursue closure of the HTF and monitor its activities to 

ensure compliance with all requirements. Furthermore, DOE uses a documented process to review and 

resolve any disposal questions and develop any mitigation measures, as appropriate. 

Tank waste storage and removal operations are governed by a SCDHEC industrial wastewater 

construction permit. [DHEC_01-25-1993] Removal from service and stabilization of the HTF waste tanks 

and ancillary structures will be carried out pursuant to a State-approved closure plan, the HTF General 

Closure Plan (GCP), which contains the overall plan for removing from service and stabilizing the HTF 

waste tanks and ancillary structures. [SRR-CWDA-2011-00022] A specific Closure Module for each 

waste tank or ancillary structure or groupings of waste tanks and ancillary structures will be developed 

and submitted to SCDHEC for approval. Final waste tank stabilization activities will not proceed until 

SCDHEC grants approval. Stabilization of individual HTF waste tanks and ancillary structures is 

anticipated to take place after individual component cleaning is complete. 

In the Savannah River Site High-Level Waste Tank Closure Environmental Impact Statement Record of 

Decision, DOE selected the alternative to fill the waste tanks with reducing grout to stabilize the residual 

material. This method was the most preferred environmental alternative. [DOE/EIS-0303 ROD] The 

stabilized grout form will provide a chemical environment to reduce migration of contaminants into the 

environment, prevent inadvertent intrusion and minimize void spaces in the waste tanks. 

Since the publication of the Environmental Impact Statement (EIS) and issuance of the Record of 

Decision (ROD), DOE developed new waste removal and cleaning technologies and developed the HTF 

PA. [SRR-CWDA-2010-00128] In light of these changes, DOE prepared the High-Level Waste Tank 

Closure Final Environmental Impact Statement for the Savannah River Site Supplement Analysis, to 

determine whether there were substantial changes or significant new information relevant to 

environmental concerns that would require a new or supplemental EIS. Based on the Supplement 

Analysis, DOE determined that the proposed actions in the FTF, the HTF and the tank closure program 

did not constitute substantial changes from those evaluated in the original EIS, did not present significant 

new circumstances or information relevant to environmental concerns, and that no further National 

Environmental Policy Act (NEPA) documentation was required to proceed with closure activities in HTF.17 

[DOE/EIS-0303-SA-02] 

1.4 NDAA Section 3116(a) Consultation Summary 

To provide the statutory context for the ensuing discussion, NDAA Section 3116(a) provides that certain 

waste from the reprocessing of spent nuclear fuel is not high-level waste if the Secretary of Energy, in 

consultation with the NRC, determines that the waste meets the criteria specified in NDAA Section 

3116(a). Those criteria are, in relevant part: 

1) the waste does not require permanent isolation in a deep geologic repository for spent fuel or 

high-level waste; 

2) the waste has had highly radioactive radionuclides (HRRs) removed to the maximum extent 

practical (MEP); and 

3) (A) the waste does not exceed concentration limits for Class C low-level waste as set out in 10 

Code of Federal Regulations (CFR) 61.55, and will be disposed of (i) in compliance with the 

performance objectives set out in 10 CFR 61, Subpart C; and (ii) pursuant to a Stateapproved closure plan or State-issued permit; or 

(B) the waste exceeds concentration limits for Class C low-level waste as set out in 10 CFR 

61.55, but will be disposed of (i) in compliance with the performance objectives set out in 10 

CFR 61, Subpart C; (ii) pursuant to a State-approved closure plan or State-issued permit; and 

(iii) pursuant to plans developed by the Secretary in consultation with NRC. 

 

17 Similarly, DOE previously prepared the High-Level Waste Tank Closure Final Environmental Impact Statement for the Savannah 

River Site Supplement Analysis for FTF, to determine whether there were substantial changes or significant new information 

relevant to environmental concerns that would require a new or supplemental EIS. Based on the Supplement Analysis, DOE 

determined that a supplemental or new EIS was not required, and that no further NEPA documentation was required to proceed with

closure activities in the FTF. [DOE/EIS-0303-SA-01] 
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NDAA Section 3116(a) is set forth in its entirety in Section 3.0 of this HTF 3116 Basis Document. 

As discussed in Section 1.1 of this HTF 3116 Basis Document, DOE has consulted extensively with the 

NRC pursuant to NDAA Section 3116(a). The NRC documented its consultative observations and 

recommendations in the TER. [ML14094A496] 

In the Transmittal Letter for the TER, NRC summarized its TER observations and recommendations as 

follows: [ML14094A496] 

 For Criterion 118, NRC concluded that, “provided DOE can meet the remaining NDAA criteria, 

there is no indication that other considerations would warrant the disposal of the waste in a 

geologic repository.” 

 With regard to Criterion 2, NRC noted that “DOE has robust processes for identification and 

characterization of HRRs. Furthermore, the DOE has identified strategies for HRR removal to 

MEP for future tank cleaning. However, because most of the tanks have yet to be cleaned, the 

information is prospective.” In the text of the TER, NRC “notes several areas where the DOE 

could provide additional information to support its conclusions regarding the ability of the HTF to 

meet the NDAA Section 3116 criteria as the tank cleaning process continues.” 

 Finally, concerning Criterion 3, the NRC states that it “is not making a conclusion at this time on 

the ability of the HTF to meet the requirements of the performance objectives in 10 CFR Part 61, 

Subpart C due to uncertainty in the final inventories for the remaining tanks as well as 

uncertainties associated with contaminants that have breached primary containment in several 

tanks.” The NRC staff go on to recommend that “the DOE conduct waste release experiments to 

increase support for key modeling assumptions.” 

In the Executive Summary of the TER, the NRC summarized its findings for both Criterion 2 and Criterion 

3, and provided “key recommendations” (eight for Criterion 2 and six for Criterion 3), concerning ongoing 

or planned DOE activities, which NRC suggested should be continued or undertaken in the future to, for 

example, enhance DOE’s documentation, technology, and certain models for the HTF PA. NRC 

highlighted Recommendation ES-3.1 (concerning Criterion 3) as NRC’s “primary recommendation”, which 

“reiterates its [NRC’s] FTF recommendation that DOE conduct waste release experiments.” In Section 5 

of the TER, titled Overall NRC Review Results and Recommendations, NRC also stated, “Given Tank 16 

will be one of the first HTF tanks to be closed, the extent of contamination in its annulus, and the previous 

release of waste from secondary containment, the NRC staff recommends that DOE should conduct a 

more comprehensive analysis of contaminant release from the annulus prior to grouting Tank 16.” 

DOE has carefully considered NRC’s recommendations. Those recommendations, suggestions and 

observations are discussed and addressed throughout the applicable sections and appendices of this 

HTF 3116 Basis Document, and in the Nuclear Regulatory Commission’s H-Tank Farm Technical 

Evaluation Report’s Recommendations – Department of Energy’s Activity Summary Matrix referenced in 

this HTF 3116 Basis Document. [SRR-CWDA-2014-00080] Notably, DOE has resumed its plans to 

undertake additional waste-release research, experiments, and testing, as appropriate, which DOE will 

continue to share with NRC and SCDHEC.19 Moreover, DOE has, and will continue to use, existing and 

active processes currently in place to update information and to support DOE final approvals to stabilize 

(grout) and close applicable HTF tanks, including Tank 16 and its annulus. A critical component of these 

processes includes a formal program to update and maintain performance assessments for the Liquid 

Waste facilities at SRS, including the HTF PA, pursuant to DOE Manual 435.1-1. [DOE M 435.1-1] DOE 

requires its contractors to issue an annual update to its performance assessment maintenance program 

implementation plan.20

 

18 NRC referred to NDAA Section 3116(a)(1) as “Criterion 1”, NDAA Section 3116(a)(2) as “Criterion 2”, and NDAA Section 3116(a)(3) as “Criterion 3”. 19 Section 3116 and the performance objectives referenced therein do not mandate waste-release experiments, research or the like. Nevertheless, 

DOE has considered NRC’s suggestions, and is performing appropriate waste-release activities pursuant to DOE’s authority and responsibilities under 

the Atomic Energy Act of 1954, in a manner that is fully protective of health, safety and the environment. 20 The most recent update, Savannah River Site Liquid Waste Facilities Performance Assessment Maintenance Program, FY2014 Implementation 

Plan, was issued in January 2014. [SRR-CWDA-2013-00133] In Section 3.3.1.2, Waste Release Studies, of this Implementation Plan, a description of 

current and future activities related to tank residual waste release testing and research is described. These activities specifically support the “primary 

recommendation” highlighted in the TER. The Fiscal Year 2014 Implementation Plan noted that activities related to waste release testing were “onhold pending available funding”. Subsequent to issuance of the plan, funding was identified and work continued in this area during fiscal year 2014. A 

status of this work will be included in the Fiscal Year 2015 Implementation Plan. 
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Another such process, DOE’s Approach to Documenting Removal of Radionuclides to Support DOE 

Closure Authorization, is described in detail in Appendix B of this HTF 3116 Basis Document. Appendix 

B outlines the activities that are required to inform the DOE decision to proceed with closure (e.g., 

stabilization activities) of a specific waste tank, ancillary structure or group. Tank 16 is likely to be one of 

the first waste tanks in HTF to be stabilized. As required by the process outlined in Appendix B of this 

HTF 3116 Basis Document, prior to final approval to grout Tank 16 (referred to as the “Tier 2 Closure 

Authorization”), a Tank 16 Special Analysis will be developed and formally reviewed and approved that 

will consider the final characterization of the residuals remaining in both the primary and annulus, and that 

will incorporate, as appropriate, associated recommendations in the TER for the comprehensive analysis 

of contaminant release from the Tank 16 annulus. The data from the Tank 16 Special Analysis, including 

the comprehensive analysis of contaminant release from the Tank 16 annulus, will be considered in the 

preparation and analyses contained in the Tank 16 HRR Removal Report. As noted in footnote 144 in 

Appendix B, “The results of these special analyses will be used to support the cost-benefit analyses and 

subsequent removal reports generated for each of the waste tanks and ancillary structures.” 

In sum, DOE has consulted with the NRC and taken into consideration the recommendations, observation 

and suggestions from the NRC, as well as comments from the public. In light of these considerations and 

based on the analysis, rationale and demonstrations set forth in the ensuing sections of this HTF 3116 

Basis Document, DOE is confident that the stabilized residuals, waste tanks, and ancillary structures 

(including integral equipment) at the SRS HTF at the time of closure satisfy the provisions in NDAA 

3116(a). 

1.5 Outline of HTF 3116 Basis Document 

To support closure of the HTF at the SRS, this HTF 3116 Basis Document demonstrates that the 

stabilized residuals within the HTF waste tanks and ancillary structures, those waste tanks, and the 

ancillary structures (including integral equipment) at the time of closure resulting from, in part, prior 

reprocessing of spent nuclear fuel meet the criteria in NDAA Section 3116(a) and thus are not high-level 

waste. 

Section 2.0 of this HTF 3116 Basis Document provides an overview of SRS and the HTF. In addition, 

extensive descriptions of the HTF and waste processing facilities are provided in the HTF PA. [SRRCWDA-2010-00128] Section 3.0 of this HTF 3116 Basis Document provides the specific language and 

criteria of NDAA Section 3116(a). Subsequently, Section 4.0 through Section 8.0 provides the basis to 

support a determination that the Secretary of Energy, in consultation with the NRC, may make pursuant 

to NDAA Section 3116(a). 

As discussed in Section 4.0 of this document, the stabilized residuals within the waste tanks and ancillary 

structures, the waste tanks, and the ancillary structures (including integral equipment) do not raise any 

unique considerations that, notwithstanding the demonstration that all other NDAA Section 3116(a) 

criteria have been met, require permanent isolation in a deep geologic repository. Accordingly, the HTF 

residual waste, waste tanks and ancillary structures at closure satisfy the criterion in NDAA Section 

3116(a)(1). 

The information provided in Section 5.0 and Appendix B demonstrates that the HTF waste tanks, ancillary 

structures and their associated stabilized residuals will have had HRRs removed to the MEP at the time of 

closure. Removal of HRRs to the MEP in HTF waste tanks and ancillary structures occurs through a 

systematic progression of waste removal and cleaning activities using proven technologies to a point 

where further removal of HRRs is not sensible or useful in light of the overall benefit to human health, 

safety and the environment. 

As discussed in Section 6.0, the stabilized HTF wastes at closure are anticipated to meet concentration 

limits for Class C low-level waste as set out in 10 CFR 61.55. Nevertheless, DOE has consulted with the 

NRC on DOE’s disposal plans for HTF pursuant to the consultation process in NDAA Section 

3116(a)(3)(B)(iii) to take full advantage of the consultation process established by NDAA Section 3116.21 

 

21 On page 4-8 of its TER, NRC states “The NRC staff has reviewed DOE’s preliminary disposal plans for the HTF waste as part of 

the consultation process that is documented in this TER, thereby satisfying the requirements of Section 3116(a)(3)(B)(iii) for tanks 

containing residual waste that might exceed Class C concentrations at the time of preparation of this TER. Should DOE find that
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This document demonstrates that the stabilized residuals within the waste tanks and ancillary structures, 

the waste tanks, and the ancillary structures (including integral equipment) located at HTF at the time of 

closure will meet the 10 CFR 61, Subpart C performance objectives so as to provide for the protection of 

the public health and the environment. These performance objectives address protection of the general 

population from radioactivity releases, individuals from inadvertent intrusion on the disposal site, 

protection of workers and the public during disposal facility operations, and the stability of the disposal 

site after closure. 

As discussed in Sections 7.1 and 7.2, through use of the performance assessment process, DOE has 

analyzed the possible methods by which a future member of the public or inadvertent intruder could be 

exposed to the HTF residuals and has demonstrated that there is reasonable assurance that the 

performance objectives at 10 CFR 61.41 and 10 CFR 61.42 are met. The results of the HTF PA show 

that there is reasonable assurance the annual peak doses for a future hypothetical member of the public 

and a hypothetical inadvertent intruder will be below 25 mrem and 500 mrem, respectively. As discussed 

previously, a systematic and deliberative process is in place to ascertain that the conclusions cited above 

remain valid once final residual characterization data is available for each waste tank and ancillary 

structure in HTF. 

The DOE has programs in place to ensure protection of workers and the public during facility operations. 

As demonstrated in Section 7.3 of this HTF 3116 Basis Document, the DOE requirements for 

occupational radiological protection and those for radiological protection of the public and the 

environment are equivalent to the requirements contained in the performance objective at 10 CFR 61.43. 

Section 7.4 demonstrates that the HTF at closure meets the performance objective at 10 CFR 61.44, 

concerning long-term site stability. DOE reviewed the site characteristics, including demography, 

geography, meteorology, climatology, ecology, geology, seismology and hydrogeology. As demonstrated 

in Section 7.4 of this HTF 3116 Basis Document, the site conditions do not present hazards that threaten 

long-term HTF stability. In addition, the HTF closure methods will result in a facility closure that does not 

require ongoing maintenance. 

As described in Section 8.0, the HTF waste tanks and ancillary structures will be removed from service 

(operationally closed) and stabilized pursuant to State-approved Closure Modules, consistent with the 

HTF GCP that has been approved by SCDHEC. Per the SRS FFA, the waste tanks will be cleaned until 

the United States Department of Energy-Savannah River Operations Office (DOE-SR), SCDHEC and 

EPA agree that waste removal may cease. 

As summarized above and as discussed more fully in the following sections, this HTF 3116 Basis 

Document demonstrates that the HTF waste tanks, ancillary structures and residuals at closure meet the 

criteria in NDAA Section 3116(a). Moreover, DOE has engaged in extensive consultation with the NRC, 

as discussed previously. Accordingly, this document provides the basis for the Secretary of Energy, in 

consultation with the NRC, to determine that the NDAA Section 3116(a) criteria are met and, thus, the 

HTF waste is not high-level waste. 

 HTF waste exceeds Class C limits based on final inventories developed in the future, the NRC staff could provide further 

recommendations related to DOE’s final disposal plans, if requested by DOE.” [ML14094A496] 
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2.0 BACKGROUND 

Section Purpose 

The purpose of this section is to provide background information to support discussions in 

later sections which demonstrate that the provisions in NDAA Section 3116(a) are met. 

Section Contents 

Section 2.1 provides an overview of HTF with descriptions of the different waste tank 

designs and ancillary structures. Section 2.2 identifies the sources of the waste managed 

in HTF and summarizes the history of each of the waste tanks. Section 2.3 describes 

waste tank closure activities and status. Section 2.4 describes the residual 

characterization process. Section 2.5 discusses stabilization of the waste tanks. Section 

2.6 describes the HTF Closure Cap. 

Key Points 

 The HTF occupies 45 acres in the General Separations Area (GSA) near the center of 

the SRS. 

 The HTF contains 29 carbon steel waste tanks of four different basic designs, four with 

a nominal capacity of 750,000 gallons per tank, four with a nominal capacity of 

1,030,000 and 21 with a nominal capacity of 1,300,000 gallons per tank. 

 Most of the waste in these waste tanks originated in the SRS H-Canyon Separations 

Facility, which primarily reprocessed used uranium fuel for the recovery of uranium 

but also recovered other nuclear materials produced in the site’s nuclear production 

reactors. 

 In addition to the waste tanks, HTF contains ancillary structures with a residual 

radiological inventory that is accounted for as part of HTF closure. 

 Estimated radionuclide concentrations for residual material in HTF at closure were 

determined by sample analyses, process knowledge data maintained in the Waste 

Characterization System (WCS), and special analyses; the associated risks were 

assessed in the HTF PA. 

 After waste removal, HTF waste tanks will be filled with grout to provide long-term 

stability and minimize the mobility and migration of radionuclides. 

2.1 Savannah River Site and H-Tank Farm Facility Overview 

This section provides brief descriptions of the site and the HTF.22

2.1.1 Geography and Demography 

2.1.1.1 SRS Site Description 

Construction of SRS (one of the facilities in the DOE complex) started in the early 1950s to produce 

nuclear materials (such as Pu-239 and tritium). The site covers approximately 310 square miles in South 

Carolina and borders the Savannah River. The SRS encompasses approximately 198,000 acres in 

Aiken, Allendale and Barnwell counties of South Carolina. The site is approximately 12 miles south of 

Aiken, South Carolina, and 15 miles southeast of Augusta, Georgia, as shown in Figure 2.1-1. [SRNSSTI-2011-00059] 

 

22 Sections 1.0 and 2.0, as well as Appendix A, of this HTF 3116 Basis Document, contain information to further inform the reader. 

DOE views such information, to the extent it is not otherwise relied upon in this HTF 3116 Basis Document, as outside the scope of 

this HTF 3116 Basis Document and not included as NDAA Section 3116 requirements or criteria.
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Figure 2.1-1: Physical Location of SRS 

Prominent geographic features within 30 miles of SRS include the Savannah River and Clarks Hill Lake 

(also known as Thurmond Lake), shown in Figure 2.1-2. The Savannah River forms the southwest 

boundary of SRS. Clarks Hill Lake is the largest nearby public recreational area. This reservoir lies on 

the Savannah River approximately 40 miles upstream of the center of SRS. 

 
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Figure 2.1-2: Location of SRS and Adjacent Areas 

Within the SRS boundary, prominent water features include PAR Pond and L Lake, shown in Figure 2.1-

3. PAR Pond, a former reactor cooling water impoundment, covers approximately 2,700 acres and lies in 

the eastern sector of SRS. L Lake, another former reactor cooling water impoundment, covers 

approximately 1,000 acres and lies in the southern sector of SRS. [WSRC-IM-2004-00008] 

Figure 2.1-3 also shows the major operational areas at SRS. Prominent operational areas, both past and 

present, include, Separations (F and H Areas), Waste Management Operations (E Area), Liquid Waste 

(F, H, J, S and Z Areas) and the Reactor Areas (C, K, L, P and R). The Savannah River National 

Laboratory (SRNL) and Savannah River Ecology Laboratory (SREL) are located in A Area. 

Administrative and support services are located in B Area and construction administration activities are 

located in N Area. D Area, a heavy water facility, M Area, a fuel and target area, and TNX, a testing 

facility, have undergone deactivation and decommissioning. 

 
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Figure 2.1-3: Predominant SRS Operational Area Location Map 

 
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2.1.1.2 Closure Site Description 

The HTF is in H Area, which is 

located in the central region of SRS. 

Figure 2.1-4 presents the area 

known as the GSA. The GSA is 

located atop a ridge that runs 

southwest to northeast forming the 

drainage divide between Upper 

Three Runs (UTR) to the north and 

Fourmile Branch to the south. The 

GSA contains the F-Area and HArea Separations Facilities, the SArea Defense Waste Processing 

Facility (DWPF), the Z-Area 

Saltstone Facility, the J-Area Salt 

Waste Processing Facility (SWPF) 

and the E-Area low-level waste 

disposal facilities.23 The HTF is an 

active facility consisting of 29 

carbon steel waste tanks (Figure 

2.1-5) in varying degrees of service 

or waste removal activities. The 

waste was generated primarily from 

the H-Canyon chemical separations processes. The HTF design features (e.g., waste tanks, transfer 

lines, evaporator systems) are discussed in more detail in Sections 2.1.11 and 2.1.12. 

2.1.1.3 Population Distribution 

According to the United States (U.S.) Census Bureau data, the estimated 2010 population in the eightcounty region of influence (ROI) was 571,637. Four of the counties lie in South Carolina and include 

Aiken, Allendale, Bamberg and Barnwell. The other four counties lie in Georgia and include Burke, 

Columbia, Richmond and Screven 

(Figure 2.1-2). The ROI includes 

the counties immediately adjacent 

to SRS and the counties where 

the majority of SRS workers 

reside. Approximately 85 % of the 

population in the ROI lives in the 

following three counties: Aiken 

(28.0 %), Richmond (35.1 %) 

and Columbia (21.7 %). Only 

approximately 15 % of the 

population in the ROI lives in the 

remaining counties as shown in 

Table 2.1-1. [SRR-LWDL-2012-

00001] 

From 2000 to 2010, the population 

in the eight-county region grew an 

estimated 9.8 %. Columbia 

County had the highest growth at 

38.9 % followed by Aiken County 

with a growth of 12.3 % and Burke 

County with a growth of 4.8 %. 

 

23 See Appendix A for a brief description of DWPF, SWPF and Saltstone Facility operations. 

Figure 2.1-4: Layout of the GSA 

Figure 2.1-5: General Layout of HTF 
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Allendale, Bamberg, Barnwell and Screven Counties experienced a net population loss. [SRR-LWDL2012-00001] 

The High-Level Waste Tank Closure Final Environmental Impact Statement contains population 

projections and further information regarding the region around SRS. [DOE/EIS-0303] 

2.1.1.4 Land Use Present and 

Planned 

Land within a five-mile radius of 

the HTF is entirely within SRS 

boundaries and its current use is 

for industrial purposes or as 

forested land. The classification 

of the current land use within the 

entire GSA is heavy nuclear 

industrial. Two key planning 

documents contain the plans for 

the future of SRS and are 

identified below. 

 The Savannah River Site 

End State Vision, PIT-MISC-0089

 The Savannah River Site 

Long Range Comprehensive 

Plan, PIT-MISC-0041 

The Long Range Comprehensive Plan assumes that the entire site will be owned and controlled by the 

Federal Government in perpetuity.24

2.1.2 Meteorology and Climatology 

2.1.2.1 General SRS Climate 

The SRS region has a humid subtropical climate characterized by relatively short, mild winters and 

extended, hot and humid summers. Summer-like conditions (including mid to late summer heat waves) 

typically last from May through September when the area is frequently under the influence of a western 

extension of the semi-permanent subtropical high-pressure system, most commonly known in North 

America as the Bermuda High. Winds in summer are light and cold fronts generally remain well north of 

the area. On average, greater than one-half of the days register temperatures in excess of 90°F during 

the summer months. As this maritime tropical mass comes inland, it rises and forms localized scattered 

afternoon and evening thunderstorms that are often intense. The influence of the Bermuda High begins 

to diminish during the fall as continental air masses become more prevalent, resulting in lower humidity 

and more moderate temperatures. 

Average rainfall during the fall is usually the least of the four seasons. In the winter months, mid-latitude 

low-pressure systems and associated fronts often migrate through the region. As a result, conditions 

frequently alternate between warm, moist, subtropical air from the Gulf of Mexico region and cool, dry 

polar air. The Appalachian Mountains to the north and northwest of SRS help to moderate the extremely 

cold temperatures that are associated with occasional outbreaks of Arctic air. Consequently, less than 

one-third of winter days have minimum temperatures below freezing on average, and days with 

temperatures below 20°F are infrequent. Measurable snowfall occurs on average once every two years. 

Tornadoes occur more frequently in spring than the other seasons of the year. Although spring weather 

is somewhat windy, temperatures are usually mild and humidity is relatively low. [WSRC-TR-2007-00118] 

 

24 For the purposes of the HTF PA, no federal protection is assumed beyond a 100-year period of institutional control. The 100-year 

period of institutional control is assumed to begin in the year 2032. 

Table 2.1-1: Population Distribution and Percent of Region of 

Influence (% ROI) for Counties and Selected Communities 

Jurisdiction 

2000 

Population 

2010 

Population 

% 

Change 

2010 % of 

Region 

SOUTH CAROLINA 

Aiken County 142,552 160,099 12.3 28.0 

Allendale County 11,211 10,419 -7.1 1.8 

Bamberg County 16,658 15,987 -4.0 2.8 

Barnwell County 23,478 22,621 -3.7 4.0 

GEORGIA 

Burke County 22,243 23,316 4.8 4.1 

Columbia County 89,288 124,053 38.9 21.7 

Richmond County 199,775 200,549 0.4 35.1 

Screven County 15,374 14,593 -5.1 2.6 

Eight-County Total 520,579 571,637 9.8 

[SRR-LWDL-2012-00001] 
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2.1.2.2 Meteorological Data Collection 

The collection of SRS meteorological data is from a network of nine primary monitoring stations (Figure 

2.1-6). Towers located adjacent to each of eight areas (A, C, D, F, H, K, L and P Areas) are equipped to 

measure wind direction and wind speed at 201.3 feet above ground and to measure temperature and dew 

point at both 6.6 feet and 201.3 feet above ground. A ninth tower near N Area, known as the Central 

Climatology (CLM) 

site, is instrumented 

with wind, temperature 

and dew point sensors 

at four levels: 6.6 feet 

(13.2 feet for wind), 

59.4 feet, 118.8 feet 

and 201.3 feet. The 

CLM site is also 

equipped with an 

automated tipping 

bucket rain gauge, a 

barometric pressure 

sensor and a solar 

radiometer near the 

tower at ground level. 

Data acquisition units 

at each station record 

a measurement from 

each instrument at 

one-second intervals. 

Every 15 minutes, 900 

data points are 

processed to generate 

statistical summaries 

for each variable, 

including averages 

and instantaneous 

maxima. The results 

are uploaded to a 

relational database for 

permanent archival. 

[WSRC-TR-2007-

00118] 

In addition, the Tall 

Tower facility near Beech Island, South Carolina, provides a set of high quality meteorological 

measurements that is unique to the Southeastern United States. This facility utilizes fast-response sonic 

anemometers, water vapor sensors, barometric pressure sensors, slow-response temperature sensors 

and relative humidity sensors. The data are collected at 100 feet, 200 feet and 1,000 feet above ground 

level. Spread-spectrum modems at each measurement level transmit raw data to a redundant set of 

personal computers at the SRNL. Data processing software on the personal computers determine mean 

values and other statistical quantities every 15 minutes and uploads the results to the relational database. 

Collection of precipitation measurements are from a network of 13 rain gauges across SRS (Figure 2.1-

6). Twelve of these gauges are read manually by site personnel once daily, usually around 6:00 A.M. The 

daily data are reported to the SRNL Atmospheric Technologies Center, where it is technically reviewed 

and manually entered into a permanent electronic database. The other is an automated rain gauge at the 

CLM site previously addressed above. 

Figure 2.1-6: SRS Meteorological Monitoring Network

[WSRC-TR-2007-00118] 
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2.1.2.3 Data Pertinent To PA Modeling 

Weather data pertinent to the PA modeling are atmospheric dispersion, precipitation and air temperature. 

Each is discussed below. 

2.1.2.3.1 Atmospheric Dispersion 

Since the mid-1970s, a five-year database of meteorological conditions at SRS has been updated in 

order to support dose calculations for accident or routine release scenarios for on-site and off-site 

populations. The meteorological database includes wind speed, wind direction, temperature, dew point 

and horizontal and vertical turbulence intensities. The most recent database is for the time period 

January 1, 2002 through December 31, 2006, and consists of one-hour time averages of temperature and 

dew point; wind speed, direction and turbulence. [WSRC-STI-2007-00613] These data are for 

determining dose release factors in the evaluation for air pathways dose modeling described in Section 

4.5 of the HTF PA, and was reported in SRNL-STI-2010-00018. 

2.1.2.3.2 Precipitation 

Compilations of rainfall data obtained from meteorological data collection described above for years 1952 

through 2006 for the site and for years 1961 through 2006 obtained from the 200-F weather station are in 

WSRC-STI-2007-00184. An average precipitation level result of 48.5 inches/year was gathered from the 

55-year monitoring period for the site and 49 inches/year from the 46-year monitoring period for F Area. 

These data are for determining appropriate rainfall assumptions for the performance evaluation of 

infiltration through the closure cap described in Section 2.6 and evaluated in WSRC-STI-2007-00184. 

2.1.2.3.3 Air Temperature 

A compilation of air temperature data obtained from meteorological data collection (described above) for 

years 1968 through 2005 is in WSRC-STI-2007-00184. For this 37-year period, the annual average air 

temperature was approximately 64°F with an average monthly air temperature from a low of 

approximately 46°F, to a high of approximately 81°F. These data are for determining appropriate 

assumptions for the performance evaluation of infiltration through the closure cap described in 

Section 2.6 and evaluated in WSRC-STI-2007-00184. 

2.1.3 Ecology 

Comprehensive descriptions of SRS ecological resources and wildlife are in SRS Ecology Environmental 

Information Document and briefly discussed in this section. [WSRC-TR-2005-00201] 

The SRS supports abundant terrestrial and semi-aquatic wildlife, as well as a number of species 

considered threatened or endangered. Since the early 1950s, the site has changed from 67 % forest and 

33 % agriculture to 94 % forest, with the remainder in aquatic habitats and developed areas. Wildlife 

populations correspondingly shifted from forest-farm edge utilizing species to a predominance of forestdwelling species. The SRS now supports 44 species of amphibians, 60 species of reptiles, 255 species 

of birds, and 55 species of mammals. These populations include urban wildlife, several commercially and 

recreationally important species, and a few threatened or endangered species. Protection and restoration 

of all flora and fauna to a point where their existence is not jeopardized are principal goals of federal and 

state environmental programs. Those species of plants and animals afforded governmental protection 

are referred to collectively as “species of concern.” [WSRC-TR-2005-00201] 

The SRS has extensive, widely distributed wetlands, most of which are associated with floodplains, 

creeks or impoundments. In addition, approximately 200 Carolina bays occur on SRS. Carolina bays are 

unique wetland features of the Southeastern United States. They are isolated wetland habitats dispersed 

throughout the uplands of SRS. The approximately 200 Carolina bays on SRS exhibit extremely variable 

hydrogeology and a range of plant communities from herbaceous marsh to forested wetland. [DOE/EIS0303] 

The Savannah River bounds SRS to the southwest for approximately 20 miles. The river floodplain 

supports an extensive swamp, covering approximately 15 square miles of SRS with a natural levee 

separating the swamp from the river. Timber was cut in the swamp from the turn of the century until 

1951, when the Atomic Energy Commission assumed control of the area. At present, the swamp forest is 

comprised of two kinds of forested wetland communities. Areas that are slightly elevated and well 
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drained are characterized by a mixture of oak species, as well as red maple, sweet gum and other 

hardwood species. Low-lying areas that are continuously flooded are dominated by second-growth bald 

cypress and water tupelo. [DOE/EIS-0303] 

The SRS supports abundant herpetofauna because of its temperate climate and diverse habitats. The 

species of herpetofauna include 17 salamanders, 27 frogs and toads, one crocodilian, 13 turtles, nine 

lizards and 36 snakes. The class Amphibia is represented on-site by two orders, 11 families, 16 genera 

and 44 species. The Reptilia are represented by three orders, 12 families, 41 genera and 59 species. 

[WSRC-TR-2005-00201] 

More than 255 species of birds can be found at SRS. Waterfowl and wading birds, as well as many 

upland species, use SRS aquatic habitats year round. The site’s Carolina bays and emergent marshes 

are used by 67 % of these birds. This type of habitat is used by 68 % of the upland species. Edge or 

shoreline areas account for high numbers of upland birds at the Carolina bays and emergent marshes, 

stream, and small drainage corridors, and river swamp habitats. The aquatic birds are most common in 

open water habitats. [WSRC-TR-2005-00201] 

Large mammals inhabiting the site include white-tailed deer and feral hogs. Raccoon, beaver and otter 

are relatively common throughout the wetlands of SRS. In addition, the gray fox, opossum, bobcat, gray 

squirrel, fox squirrel, eastern cottontail, mourning dove, northern bobwhite and eastern wild turkey are 

common at SRS. Threatened or endangered plant and animal species known to exist or that might be 

found on the overall site include the smooth purple coneflower, wood stork, red-cockaded woodpecker 

and short-nose sturgeon. [WSRC-TR-2005-00201] 

The HTF is located within a densely developed, industrialized area of SRS. The immediate area provides 

habitat for only those animal species typically classified as urban wildlife. Species commonly 

encountered in this type of urban landscape include the Southern toad, green anole, rat snake, rock dove, 

European starling, house mouse, opossum and feral cats and dogs. Grasses and landscaped areas 

within the GSA in proximity to the HTF also provide some marginal terrestrial wildlife habitat. A number of 

ground-foraging bird species (e.g., American robin, killdeer and mourning dove) and small mammals 

(e.g., cotton mouse, cotton rat and Eastern cottontail) that use lawns and landscaped areas around 

buildings may be present at certain times of the year, depending on the level of human activity (e.g., 

frequency of mowing). Pine plantations managed for timber production by the U.S. Forest Service (under 

an interagency agreement with DOE) occupy surrounding areas. 

The Fourmile Branch seepline area is located in a bottomland, hardwood forest community. The canopy 

layer of this bottomland forest is dominated by sweet gum, red maple and red bay with an occasional 

sweet bay throughout. The understory consists largely of saplings of these same species, as well as an 

herbaceous layer of smilax, dog hobble, giant cane, poison ivy, chain fern and hepatica. At the seepline 

upland edge, scattered American holly and white oak occur. Dominant along Fourmile Branch in this 

area are tag alder, willow, sweet gum and wax myrtle. The seepline is located in a similar bottomland, 

hardwood forest community. [DOE/EIS-0303] 

No endangered or threatened fish or wildlife species have been recorded near the UTR and Fourmile 

Branch seeplines. The seeplines and associated bottomland community do not provide habitat favored 

by endangered or threatened fish and wildlife species known to occur at SRS. The American alligator is 

the only federally protected species that could potentially occur in the area of the seeplines. Fourmile 

Branch does support a small population of American alligator in its lower reaches, where the stream 

enters the Savannah River swamp. [DOE/EIS-0303] 

According to summaries on UTR studies documented in the SRS Ecology Environmental Information 

Document, the macroinvertebrate communities of UTR drainage are unusual. [WSRC-TR-2005-00201] 

They include many rare species and species not often found living together in the same freshwater 

system. Since UTR is a spring-fed stream and is colder and generally clearer than most surface water at 

its low elevation, species typical of unpolluted streams in northern North America or the southern 

Appalachian Mountains are found here along with lowland (Atlantic Coastal Plain) species. 

The fish community of UTR is typical of third and higher order streams on SRS that have not been greatly 

affected by industrial operations, with shiners and sunfish dominating collections. The smaller tributaries 

of UTR are dominated by shiners and other small-bodied species (i.e., pirate perch, madtoms and 
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darters) indicative of un-impacted streams in the Atlantic Coastal Plain. In the 1970s, the U.S. Geological 

Survey designated UTR as a National Hydrological Benchmark Stream due to its high water quality and 

rich fauna. However, this designation was rescinded in 1992 due to increased development of the UTR 

watershed north of SRS site boundaries. [DOE/EIS-0303] 

2.1.4 Geology, Seismology and Volcanology 

Regional and local information on the geologic and seismic characteristics of the HTF are presented in 

this section. Because SRS is not located within a region of active-plate tectonics characterized by 

volcanism, volcanology is not an issue of concern in the HTF PA, and thus further discussion of this topic 

is omitted from the following discussion. [WSRC-IM-2004-00008] 

2.1.4.1 Regional and Site-Specific Topography 

The SRS is on the Atlantic Coastal Plain Physiographic Province approximately 25 miles southeast of the 

Fall Line that separates the relatively unconsolidated coastal plain sediments from the underlying 

Piedmont Physiographic Province. Beneath the coastal plain, sedimentary sequences reveal two 

geologic terrains. One is the Dunbarton 

basin, a Triassic-Jurassic Rift basin filled 

with lithified terrigenous and lacustrine 

sediments. The other is a crystalline 

terrain of metamorphosed sedimentary 

and igneous rock that may range in age 

from Precambrian to late Paleozoic 

derived from the crystalline igneous and 

metamorphic rocks of possibly late 

Precambrian to late Paleozoic age in the 

Piedmont Province. Early to middle 

Mesozoic (Triassic to Jurassic) rocks 

occur in isolated fault-bounded valleys 

either exposed within the crystalline 

belts or buried beneath the coastal plain 

sediments. The coastal plain sediments 

were derived from erosion of the 

crystalline rocks during late Mesozoic 

(Cretaceous) in stream and river valleys, 

and are represented locally by gravel 

deposits adjacent to present-day 

streams and by sediments filling upland 

depressions (sinks and Carolina bays). 

The Cretaceous and younger sediments 

are not significantly indurated. The total 

thickness of the sediment package at 

SRS varies between approximately 700 

feet at the northwest boundary and 

1,200 feet at the southeast boundary. 

[WSRC-TR-95-0046] 

Figure 2.1-7 shows the relationship of 

SRS to overall regional geological 

provinces, and Figure 2.1-8 details the 

regional physiographic provinces in 

South Carolina. As can be seen on 

Figure 2.1-8, much of SRS lies within the Aiken Plateau, and this plateau has an approximate 5 % slope 

to the southeast. Savannah and Congaree Rivers bound the plateau, which extends from the Fall Line to 

the Orangeburg escarpment. The highly dissected surface of the Aiken Plateau is characterized by broad 

interfluvial areas with narrow, steep-sided valleys. Local relief can be as much as 300 feet. Figure 2.1-9 

shows the topography and 20-foot contour lines of the GSA. [WSRC-TR-95-0046] 

Figure 2.1-7: Regional Geological Provinces of Eastern 

United States 

[WSRC-TR-2000-00310, Figure 1]
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Figure 2.1-8: Regional Physiographic Provinces of South Carolina 

[WSRC-TR-95-0046, Figure 2-3] 

 

Figure 2.1-9: GSA Topography 
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Currently, HTF storm water drainage is directed to an outfall, which will be unaffected by HTF operations 

and waste tank closure activities. The installation of the HTF closure cap (see Section 2.6) will 

necessitate changes to the HTF drainage system, which will be designed later as part of the overall 

closure of HTF. 

2.1.4.2 Local Geology and Soils 

SRS is comprised of seven major soil associations: Chastain-Tawcaw-Shellbluff, Rembert-Hornsville, 

Blanton-Lakeland, Fuquay-Blanton-Dothan, Orangeburg, Vaucluse-Ailey and Troup-Pinkney-Lucy. 

Figure 2.1-10 delineates the 

general soil associations for 

SRS. Details regarding these 

associations may be found in 

the Soil Survey of the 

Savannah River Plant, Parts 

of Aiken, Barnwell, and 

Allendale Counties, South 

Carolina. [PIT-MISC-0104] 

The overall general soil 

association for H Area is the 

Fuquay-Blanton-Dothan. The 

most predominant soil types 

within H Area are classified 

as Udorthents. Udorthents 

consist of well-drained soils 

that formed in heterogeneous 

materials, which are the spoil 

or refuse from excavations 

and major construction 

operations. Udorthents range 

from sandy to clayey, 

depending upon the source of 

material or geologic parent 

material. Udorthents are 

most commonly associated 

with well drained to 

excessively drained upland 

soils. A few small, poorly 

drained areas that have spoil 

are also included. Typical 

profiles for Udorthents are not 

shown due to the lack of 

consolidation within short 

distances. Clayey soil has demonstrated good retention for most radionuclides. There are also areas 

that consist of cross-bedded, poorly sorted sand with lenses and layers of silt and clay. 

The uppermost geologic unit in the HTF is comprised of the middle to late Miocene-age Upland Unit, 

which extends over much of SRS (see Section 2.1.5.1). The term “Upland Unit” is an informal name used 

to describe sediments at higher elevations located in the Upper Coastal Plain in southwestern South 

Carolina. This area has also been referred to as the Aiken Plateau. The Upland Unit includes the vadose 

zone and a portion of the UTR Aquifer-Upper Zone (UTRA-UZ). The occurrence of cross-bedded, poorly 

sorted sands with clay lenses in the Aiken Plateau indicates fluvial deposition (high-energy channel 

deposits to channel-fill deposits) with occasional transitional marine influence. This depositional 

environment results in wide differences in lithology and presents a very complex system of transmissive 

and confining beds or zones. The lower surface of the Upland Unit is very irregular due to erosion of the 

underlying formations. [DOE/EIS-0303] 

Figure 2.1-10: General Soil Associations for SRS 

[PIT-MISC-0104]
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A notable feature of the Upland Unit is its compositional variability. This formation predominantly consists 

of red-brown to yellow-orange, gray and tan colored, coarse to fine grained sand, pebbly sand with lenses 

and beds of sandy clay and clay. Generally vertically upward through the unit, sorting of grains becomes 

poorer, clay beds become more abundant and thicker, and sands become more argillaceous and 

indurated. In some areas, small-scale joints and fractures, both of which are commonly filled with sand or 

silt, traverse the unit. The mineralogy of the sands and pebbles primarily consists of quartz, with some 

feldspars. In areas to the east-southeast, sediments may become more phosphatic and dolomitic. The 

soils in the Upland Unit may contain as much as 20 % to 40 % clay. [DOE/EIS-0303] 

Below the Upland Unit lies the Tobacco Road Formation, consisting of red, brown, tan, purple, and 

orange quartz sands, and clayey quartz sands. These sands are fine to coarse moderately to poorly 

sorted, with minor clay laminae. In general, the sands of the Tobacco Road Formation are muddier, more 

micaceous and more highly colored than the sands of the underlying Dry Branch Formation. The base of 

the Tobacco Road Formation is marked in places by a coarse layer that contains flat quartz pebbles. 

Clay laminae in the upper part of the formation suggest that some of the unit was deposited in a 

transitional, low-energy environment, such as a tidal flat. The Tobacco Road Formation is approximately 

20-foot thick and is part of the UTRA-UZ. [SRNL-STI-2010-00148] 

Underlying the Tobacco Road Formation is the Dry Branch Formation, consisting of variably colored, 

poorly sorted to well-sorted sand with the interbedded tan to gray clay. The upper portion of the Dry 

Branch Formation is within the UTRA-UZ. The middle to lower portion of the Dry Branch Formation 

includes the Twiggs Clay; a semi-confining clay layer also designated as the Tan Clay Confining Zone, 

which separates the UTRA-UZ from the UTRA-Lower Zone (UTRA-LZ). 

Below the Twiggs Clay are the Clinchfield and Santee Formations. In H Area, the Santee Formation is 

composed of mixed clastic and carbonate materials, with clastic material being dominant; the interpreted 

depositional scenario is a moderate energy, middle shelf environment, with input of both clastic and 

carbonate sediments. Lithologic and petrographic studies have divided the Santee Formation in the GSA 

into eight microfacies, quartz sand (stone), terrigenous mud (stone), skeletal lime mudstone, skeletal 

wackestone, skeletal packstone, skeletal grainstone, microsparite and siliceous mudstone. [WSRC-RP94-54] None of these depositional environments contains significant amounts of limestone that would be 

conducive to the formation of large subsurface voids, karst or caves within the vicinity of HTF. 

The calcareous zones, located within the Santee Formation, contain “soft zones.” Characterization 

activities reported in various early documents describe potential voids, drilling fluid losses and grout takes 

associated with the Santee Formation at SRS. Soft zones have been encountered beneath most of SRS, 

but are less common in the northwest (updip) and more common in the southeastern (downdip near 

K Area) regions. This distribution appears to correlate with the well-documented pattern of increasing 

carbonate content in the Santee Formation to the southeast. This lateral variation in carbonate content 

reflects the original range of depositional environments, from nearshore and inner shelf environments with 

primarily terrigenous input in the northwest, to quiet water, outer shelf conditions of carbonate 

accumulation in the southeast (in the vicinity of K Area). [WSRC-RP-94-54, WSRC-TR-99-4083] 

A recent evaluation of more than 60 years of investigation and research into the occurrence, origin and 

behavior of soft zones confirms that soft zones beneath SRS are not cavernous voids, but are small, 

isolated, poorly connected three-dimensional features filled with loose, fine-grained, water saturated 

sediment. [SRNL-TR-2012-00160] 

In the GSA, which includes HTF, there is no evidence of actual subsurface voids, karst or caves that 

would act as open flow conduits. In historical and recent literature, no documentation was found of void 

spaces or other phenomena that would influence contaminant migration in a manner not already captured 

in the GSA Database (GSAD). As described in Section 2.1.5.2, the GSAD was developed using field 

data and interpretations for the GSA and vicinity and is a subset of site-wide data sets of soil lithology and 

groundwater information. The GSAD is used as the basis of hydrogeologic input values into the 

computational model for groundwater flow and contaminant transport as described in Section 7.1.2. 

Underlying the Santee Formation is the Warley Hill Formation, often referred to as the “Green Clay”, 

which forms the hydrologic barrier separating the UTRA-LZ from the underlying Gordon Aquifer of the 

Congaree Formation. 
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A more detailed description of the geology and soils of the H Area can be found in a report titled 

Hydrogeologic Framework of West-Central South Carolina. [PIT-MISC-0112] 

2.1.4.3 Seismology 

The seismic history of the Southeastern United States (of which SRS is a part) spans a period of nearly 

three centuries, and is dominated by the Charleston earthquake of August 31, 1886 (estimated magnitude 

of 7.0). The historical database for the region is essentially composed of two data sets extending back to 

as early as 1698. The first set is comprised of pre-network, mostly qualitative data (1698 to 1974), and 

the second set covers the relatively recent period of instrumentally recorded or post-network seismicity, 

1974 through April 2009. Figure 2.1-11 shows the locations of historical seismic events in the Southeast. 

Figure 2.1-12 denotes the epicenter locations of seismic events within a 50-mile radius of SRS. [SRRCWDA-2010-00128] 

Figure 2.1-11: Historical Seismic Events in the Southeast 

[SRR-CWDA-2010-00128]
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Figure 2.1-12: Seismic Events within a 50-Mile Radius of SRS

[SRR-CWDA-2010-00128]

The most recent seismic event occurring within a 50-mile radius of SRS was on February 14, 2014, with a 

magnitude of 4.1 and February 16, 2014, with a magnitude of 3.0. [SRR-LWDL-2014-00008, SRRLWDL-2014-00009] No damage to SRS was recorded. However, there have been four earthquakes with 

epicenter locations within SRS. They occurred June 9, 1985 (magnitude of 2.6); August 5, 1988 

(magnitude of 2.0); May 17, 1997 (magnitude of 2.5), and October 8, 2001 (magnitude of 2.6). No strong 

motion accelerometers were triggered because of these earthquakes. Note that additional seismic events 

with epicenter locations within SRS occurred shortly after the October 2001 earthquake, these seismic 

events were attributed to aftershocks. [WSRC-MS-2003-00617] 

The regional faults within SRS and vicinity are shown in Figure 2.1-13. A study entitled Comparison of 

Cenozoic Faulting at the Savannah River Site to Fault Characteristics of the Atlantic Coast Fault 

Province: Implications for Fault Capability (WSRC-TR-2000-00310) provides additional data. This study 

concludes that these regional faults exhibit the same general characteristics, are closely associated with 

the faults of the Atlantic Coastal Fault Province, and thus are part of the Atlantic Coastal Fault Province. 

Several faults of the Atlantic Coastal Fault Province have been the subject of detailed investigations. In 

all cases, the conclusion has been reached that these faults have not had a movement within the past 

35,000 years and no movement of a recurring nature within the past 500,000 years. Inclusion in the 

Atlantic Coastal Fault Province means that the historical precedent established by decades of previous 

studies on the seismic hazard potential for the Atlantic Coastal Fault Province is relevant to faulting at the 

SRS. [WSRC-TR-2000-00310] 
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In 1976, a short-period seismic network was established. In 1999, a 10-station strong motion 

accelerometers network was installed throughout the complex. Detailed information regarding seismic 

characteristics at SRS can be found in the Documented Safety Analysis document, WSRC-IM-2004-

00008. 

As noted in Section 2.1.4.2, soft 

zones have been reported in various 

early documents. The soft zones 

described in these documents are 

described as voids, drilling fluid losses 

and grout takes associated with the 

Santee Formation beneath SRS that 

may be susceptible to seismic activity. 

However, in spite of their under 

consolidated nature, soft zones have 

survived for a very long time and 

remain structurally competent in the 

presence of significant overburden 

stresses. [SRNL-TR-2012-00160] 

The predicted behavior of soft zones 

under both static and dynamic 

conditions has been modeled for 

numerous SRS facilities. These 

calculations show soft zones to be 

stable under static conditions; 

dynamic analyses predict that soft 

zones will not collapse in response to 

a design basis earthquake. [WSRCTR-99-4083] The design basis 

earthquake and associated ground 

motion, measured in peak ground 

acceleration for construction of 

facilities at the SRS (ground motion 

0.2 force of gravity) is based on 

historic seismic events in the region, 

the geologic literature and attenuation 

relations. [WSRC-TR-90-0284] 

As a conservative approach, the design for some SRS facilities assumes that soft zones will collapse 

(compress) in response to applied stress. An analysis for a proposed facility (Actinide Packaging and 

Storage Facility) within the GSA calculated that collapse of a relatively thick (approximately 8 inches) twolayer soft zone would only cause a ground surface settlement of about 4 inches. [K-CLC-F-00034] 

Although such conservative calculations are an important aspect of nuclear safety evaluations, it is 

noteworthy that soft zones in the Eocene age Santee Formation have survived without collapsing for tens 

of millions of years and have presumably persisted in spite of many earthquakes, including design basis 

earthquakes, and less frequent events of even greater magnitude. [SRR-CWDA-2011-00054] 

A structural assessment was prepared for operationally closed waste tanks. Waste tank settlement can 

occur due to two loads, static load and seismic loads. Static settlement is likely to occur due to the large 

overburden load from the closure cap. This settlement is expected to be relatively uniform. Any static 

differential settlement would be small in magnitude and cause a grout-filled waste tank to rotate as a rigid 

body. Small magnitudes of rigid body rotation will induce only small lateral forces that can be considered 

negligible. Therefore, static differential settlement is not considered further. [T-CLC-F-00421] 

Seismic differential settlement can occur due to liquefaction and soft zone settlement. Soft zones are 

often areas of under-consolidated material in a stronger matrix material that essentially forms a soil arch, 

allowing the soft zones to remain under-consolidated. A large seismic event could cause the soil arch to 

Figure 2.1-13: Regional Scale Faults for SRS and Vicinity 

[WSRC-TR-2000-00310, Figure 10] 
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fail resulting in settlement as consolidation occurs in the under consolidated material until it is normally 

consolidated. The maximum tensile stresses resulting from this consolidation on the grout-filled waste 

tanks is an overstress of 4 %, occurring for a small depth. Since this stress occurs only for an extreme 

settlement case and due to the many bounding assumptions made in the structural assessment, it was 

concluded that there is very high confidence the grout filled waste tanks will not crack. [T-CLC-F-00421] 

In addition, for the E-Area vaults in the GSA, a structural degradation study was prepared. This study 

included an evaluation of ground motion effects on the vaults. Ground motion magnitudes were 

extrapolated from SRS Performance Category (PC)-1 to PC-4 site-specific seismic criteria.25 For 

horizontal acceleration, a 0.45 force of gravity value was obtained by extrapolating the zero period 

accelerations (i.e., peak ground acceleration) of the SRS design response for PC-1 to PC-4. [T-CLC-E00018] 

For vertical acceleration (2.0 force of gravity), a bounding approach was taken by extrapolating the peak 

of the SRS horizontal design response spectra for PC-1 to PC-4. This approach results in the large 

discrepancy between horizontal and vertical acceleration. This bounding approach for vertical 

acceleration was used in the structural degradation study because the item of concern was a buried roof 

slab with voids below. Therefore, the E-Area vault roof could respond differently than the ground (i.e., not 

peak ground acceleration). As the stabilized waste tanks will have no significant voids after grouting, this 

issue is not a concern. [T-CLC-E-00018] 

Due to the lack of vertical/horizontal studies for low probability of exceedance events at SRS, the same 

bounding criteria used in the E-Area study were used for the structural assessment for closed waste 

tanks. [T-CLC-F-00421] However, it is recognized that 2.0 force of gravity is a bounding number. It is 

not a realistic number for ground acceleration at SRS. At the nearby Vogtle Electric Generating Plant, the 

vertical/horizontal ratio for the maximum considered event was 1.0, so a similar ratio should be 

considered acceptable for the SRS tank farms. [NUREG-1923] Based on a vertical/horizontal ratio of 

1.0, the maximum vertical acceleration would be 0.45 force of gravity, much less than 2.0 assumed. 

The PC-3 return period is 2,500 years (probability of exceedance 4.0E-04), and the PC-4 return period is 

10,000 years (probability of exceedance 1.0E-04). In the E-Area analysis, one-dimensional soil analyses 

indicated the differential lateral displacement between the top and bottom elevations of the E-Area vault 

(approximately 28 feet in height) were 0.05 inches for a PC-3 event and 0.09 inches for a PC-4 event. 

The height differential in the E-Area vaults is similar to the height differential of the waste tanks. 

Extrapolating to probability of exceedance 1.0E-06 (a very low probability event) gives a maximum lateral 

differential displacement of 0.22 inches. For this small amount of deformation, the soil would deform 

locally at the boundaries of the grout-filled waste tank and stresses induced in the waste tank structure 

will be minimal. [T-CLC-E-00018] 

The impacts from seismic events are considered in the conceptual model. To simulate potential 

conditions in the HTF closure system, multiple waste tank configurations were identified for analysis. 

While the configurations and the potential seismic events are not explicitly linked, the types of cracks 

caused by the credible seismic events at the HTF are assumed bounded by the configurations and the 

occurrence probability associated with the configurations in the stochastic modeling. 

Seismic considerations are also included in the design of the conceptual closure cap to ensure seismic 

induced degradation mechanisms are addressed. Section 2.6 discusses the conceptual closure cap 

design, which will further consider and handle static loading induced settlement, seismic induced 

liquefaction and subsequent settlement, and seismic induced slope instability. 

2.1.5 Hydrogeology 

An understanding of the hydrogeology of the HTF is required in order for an estimate of the fate and 

transport of the residual HTF contaminants to be modeled. Characterization and monitoring data in the 

 

25 Performance category classification is a graded approach used to establish design and evaluation requirements for structures, 

systems and components. Performance categories range from 0 to 4 in order of increasingly stringent mitigation and performance

requirements and with decreasing values of annual probability of exceedance of acceptable behavior limits. Performance 

categories are developed with regards to acts of nature (e.g., earthquake, wind, hurricane, tornado, flood, rain or snow precipitation, 

volcanic eruption, lightning strike or extreme cold or heat) which may threaten workers, the public or the environment by potential 

damage to structures, systems and components. 
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SRS GSA is extensive and provides a clear understanding of hydrogeology containing the HTF, and 

permitted generation of the GSAD. Additional background information supporting this conclusion is 

presented in Section 2.1.5.2. 

2.1.5.1 Regional Hydrogeology 

The SRS lies in the Atlantic Coastal Plain, a southeast-dipping wedge of unconsolidated and semiconsolidated sediment, which extends from its contact with the Piedmont Province at the Fall Line to the 

continental shelf edge. Sediments range in geologic age from late Cretaceous to recent and include 

sands, clays, limestones and gravels. This sedimentary sequence ranges in thickness from essentially 

zero at the Fall Line to more than 4,000 feet at the Atlantic Coast. At SRS, coastal plain sediments 

thicken from approximately 700 feet at the northwestern boundary to approximately 1,400 feet at the 

southeastern boundary of the site and form a series of aquifers and confining or semi-confining units. 

Aquifer systems include the Floridan and Dublin-Midville systems. [WSRC-STI-2006-00198] 

Figure 2.1-14 shows a generalized cross section of the sedimentary strata and their corresponding 

depositional environments for the Upper Coastal Plain down-dip through SRS into the Lower Coastal 

Plain. Figure 2.1-15 shows the regional lithologic units discussed in Section 2.1.4.2 and their 

corresponding hydrostratigraphic units at SRS. This classification system is consistent with the 

established system and is now widely used as SRS standard. [SRNL-STI-2010-00148] 

Figure 2.1-14: Regional NW to SE Cross Section

[WSRC-STI-2006-00198]
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Figure 2.1-15: Comparison of Chronostratigraphic, Lithostratigraphic, and Hydrostratigraphic 

Units in the SRS Region 

[SRNL-STI-2010-00148]

Figure 2.1-16 and Figure 2.1-17 illustrate potentiometric maps of the UTRA and Gordon Aquifer. 

Groundwater within the Floridan Aquifer system flows toward streams and swamps and into the 

Savannah River at rates ranging from inches to several hundred feet per year. The depth to which 

nearby streams cut into sediments, the lithology of the sediments and the orientation of the sediment 

formations control the horizontal and vertical movement of the groundwater. The valleys of smaller 

perennial streams in the GSA, such as Fourmile Branch, McQueen Branch and Crouch Branch, allow 

discharge from the shallow saturated geologic formations. The valleys of major tributaries of the 

Savannah River (e.g., UTR) drain formations of greater depth. With the release of water to the streams, 

the hydraulic head of the aquifer unit releasing the water can become less than that of the underlying unit. 

If this occurs, groundwater has the potential to migrate upward from the lower unit to the overlying unit. 

[DOE/EIS-0303] 
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Figure 2.1-16: Potentiometric Surface of the UTRA 

[SRNS-STI-2011-00059]
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Figure 2.1-17: Potentiometric Surface of the Gordon Aquifer 

[SRNS-STI-2011-00059]

 
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2.1.5.2 Characterization of Local Hydrogeology 

The GSA has been the focus of numerous geological and hydrogeological investigations. Early work 

included installation of monitoring wells in the 1950s and 1960s. Further characterization and monitoring 

were conducted in the area during the 1970s through present time, largely to support groundwater 

monitoring and decommissioning activities. The GSAD was developed using field data and 

interpretations for the GSA and vicinity through 1996. Although characterization and monitoring have 

been ongoing, the additional data has not altered fundamental understanding of groundwater flow 

patterns and gradients in the GSA. The GSAD is a subset of site-wide data sets of soil lithology and 

groundwater information. Figure 2.1-18 shows the location of all hydrostratigraphic picks used in the 

GSAD. Picks were made based on a combination of geophysical logs, Cone Penetration Test logs and 

core descriptions. Figures 2.1-19 through 2.1-22 show the locations of laboratory permeability data, 

multiple well pump tests, single well pump tests and slug test data used in the GSAD. Table 2.1-2 

presents a summary of the characterization and monitoring data in the GSAD. These data provide 

detailed understanding of local hydrogeology beneath the HTF. See WSRC-TR-96-0399, Volumes 1 and 

2, for a more comprehensive discussion of the data set. The GSAD, comprising SRS characterization 

and monitoring data and interpretations, is used as the basis of hydrogeologic input values into the 

computational model for groundwater flow and contaminant transport as described in Section 4.2.2.1.3 of 

the HTF PA. [SRR-CWDA-2010-00128] 

 

Figure 2.1-18: Hydrostratigraphic Picks in GSAD 

[WSRC-TR-96-0399-Vol. 1] 
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Figure 2.1-19: Laboratory Determined Permeability Data in GSAD 

[SRNL-ESB-2007-00035] 

 

Figure 2.1-20: Multiple Well Pump Test Data in GSAD 

[WSRC-TR-96-0399-Vol. 2] 
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Figure 2.1-21: Single Well Pump Test Data in GSAD 

[WSRC-TR-96-0399-Vol. 2] 

 

Figure 2.1-22: Slug Test Data in GSAD 

[WSRC-TR-96-0399-Vol. 2] 
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As described in Section 3.1.4.2 of the HTF PA, calcareous zones within the UTRA-LZ have been 

documented to contain soft zones, often related to dissolution of carbonate material. Soft zones at the 

SRS have not been studied using tracer tests; however, no unusual hydraulic gradients or unexpected 

flow conditions have been documented in the HTF or the GSA. Soft zones have, however, been the 

subject of many general and facility-specific investigations. These studies have shown that the soft zones 

are isolated, discrete, poorly connected, non-uniformly distributed features within the UTRA-LZ. Although 

their size and shape vary, their average 

thickness is generally only a few feet 

with a postulated maximum lateral 

dimension approximately 10 to 20 feet 

or less. [K-ESR-G-00013] 

2.1.5.3 Groundwater Flow in 

 the GSA 

The aquifers of primary interest for HTF 

modeling are the UTRA-UZ/LZ and 

Gordon Aquifer. Plate 17 of the 

Hydrogeological Framework of WestCentral South Carolina (PIT-MISC-0112) 

gives the leakance coefficient of the 

Crouch Branch Confining Unit (of the 

Meyers Branch Confining System) as 

roughly 3E-06 per day, which 

corresponds to 0.13 inch/year for every 

10 feet of head difference. The 

measurement of head difference across 

the Crouch Branch Confining Unit is 

zero to 20 feet causing an upward flow 

averaging 0.13 inch/year. [PIT-MISC0112] Flow across the unit is therefore 

a small fraction of total recharge, and is 

negligible in the HTF modeling. 

Potential contamination from the HTF is not expected to enter the deeper Crouch Branch Aquifer because 

an upward gradient exists between the Crouch Branch and Gordon Aquifers near UTRA. Figure 2.1-23 is 

Table 2.1-2: Characterization and Monitoring Data in the GSAD 

Data Type Quantity Reference 

Sediment Core Descriptions 204 Locations; 

~37,500 feet WSRC-TR-96-0399-Vol. 1, App. B 

Tops of Hydrostratigraphic Units/Zones 

Crouch Branch Confining Unit 52 Locations 

WSRC-TR-96-0399-Vol. 1, 

App. A-3 

Gordon Aquifer Unit 146 Locations 

Gordon Confining Unit 161 Locations 

Upper Three Runs – Lower Zone Aquifer 222 Locations 

Tan Clay Confining Zone 225 Locations 

Permeability Measurements 

Pump Tests 85 Values 

Slug Tests 481 Values WSRC-TR-96-0399-Vol. 2, App. B 

Laboratory Permeability 258 Values 

Water Levels 

Gordon Aquifer Unit 79 Locations 

Upper Three Runs – Lower Zone Aquifer 173 Locations WSRC-TR-96-0399-Vol. 2, App. C 

Upper Three Runs – Upper Zone Aquifer 387 Locations 

Figure 2.1-23: Conceptual Diagram of Groundwater Flow 

beneath the GSA 
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a cross-sectional schematic representation of groundwater flow patterns in the UTRA and Gordon Aquifer 

along a north-south cross section running through the center of HTF, shown with significant vertical 

exaggeration. Section 4.2.2.1.3 of the HTF PA provides the modeling inputs associated with groundwater 

flow characteristics obtained from the GSAD. [SRR-CWDA-2010-00128] 

Although calcareous zones containing soft zones have been identified in the UTRA-LZ (see Section 

3.1.4.2 of the HTF PA), during the 20-year period spanned by investigations used to populate the GSAD 

at more than 15 locations near HTF, no open flow conduits or other factors have been identified that 

would critically influence contaminant transport. 

In addition, a further evaluation of more than 60 years of onsite investigation and research into soft zone 

occurrence, origin and behavior concludes that soft zones at the SRS appear not to be a critical influence 

on either groundwater flow or contaminant transport. [SRNL-TR-2012-00160] 

Calcareous zones and associated soft zones are not treated separately in the flow model because they 

are isolated and discontinuous in the GSA, representing only a small fraction of the UTRA-LZ. These 

features occur near the base of the UTRA-LZ in the GSA and do not extend through the entire thickness 

of the aquifer. [WSRC-TR-99-4083] 

2.1.5.4 Surface-Water Flow 

in the GSA 

The Savannah River, which 

forms the boundary between 

Georgia and South Carolina, is 

the principal surface-water 

system near SRS. The river 

adjoins the site along its 

southwestern boundary for a 

distance of approximately 20 

miles and the site is 160 rivermiles from the Atlantic Ocean. 

Five upstream reservoirs – 

Jocassee, Keowee, Hartwell, 

Richard B. Russell and Clarks 

Hill Lake (also known as 

Thurmond Lake), minimize the 

effects from droughts and the 

impacts of low flow on 

downstream water quality and 

fish and wildlife resources in 

the river. Figure 2.1-24 shows 

the Savannah River Basin 

dams. The long-term yearly 

Savannah River flow at SRS 

averages approximately 10,400 

cubic feet per second at SRS. 

[WSRC-TR-2005-00201, Table 

4-24] For 2010, the average 

annual measured flow rate was 

6,603 cubic feet per second. 

[SRNS-STI-2011-00059] 

The major tributaries that occur on SRS are UTR, Fourmile Branch, Pen Branch, Steel Creek and Lower 

Three Runs (Figure 2.1-25). These tributaries drain all of SRS with the exception of a small area on the 

northeast side, which drains to a tributary of the Salkehatchie River. Each of these streams originates on 

the Aiken Plateau in the coastal plain and descends 50 to 200 feet before discharging into the river. The 

source of most of the surface water on SRS is either natural rainfall (see Section 2.1.2), water pumped 

Figure 2.1-24: Savannah River Basin Dams 
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from the Savannah River and used for 

cooling site facilities, or groundwater 

discharging to surface streams. The 

streams, which historically have received 

varying amounts of effluent from SRS 

operations, are not commercial sources of 

water. Downstream of SRS, the river 

supplies domestic water and is used for 

commercial and sport fishing, boating and 

other recreational activities. [DOE/EIS0303] 

The natural flow of SRS streams range 

from eight cubic feet per second in smaller 

streams to 245 cubic feet per second in 

UTR. [WSRC-IM-2004-00008] Gauging 

stations located in the GSA (Figure 2.1-

26) monitor flows for UTR and Fourmile 

Branch. Both Fourmile Branch and UTR 

are measured monthly for water flow, 

temperature and quality. The annual 

Savannah River Site Environmental 

Report for 2010 contains detailed 

information on flow rates and water quality 

of the Savannah River and SRS streams. 

[SRNS-STI-2011-00059] 

The SCDHEC regulates the physical 

properties and concentrations of 

chemicals and metals in SRS effluents 

under the National Pollutant Discharge 

Elimination System (NPDES) program. 

Also regulated by SCDHEC, biological 

water quality standards for SRS waters, have classified the Savannah River and SRS streams as 

“Freshwaters.” [DOE/EIS-0303] Freshwaters are described as suitable for primary and secondary 

contact recreation and as a source for drinking water supply after treatment in accordance with SCDHEC 

requirements. Freshwaters are suitable for 

fishing, for the survival and propagation of a 

balanced indigenous aquatic community of 

fauna and flora and for industrial and 

agricultural uses. [SCDHEC R.61-68] 

The longest of SRS streams, UTR is a large 

blackwater stream in the northern part of 

SRS that discharges to the Savannah River. 

It drains an area of over 195 square miles 

and is approximately 25 miles long, with its 

lower 17 miles within SRS boundaries. This 

stream receives more water from 

underground sources than other SRS 

streams and is the only stream with 

headwaters arising outside the site. The 

UTR is the only major tributary on SRS that 

has not received thermal discharges. The 

UTR valley has meandering channels, 

 

Figure 2.1-25: SRS Watershed Boundaries and Major 

Tributaries 

[WSRC-STI-2008-00057]

Figure 2.1-26: GSA Gauging Stations 
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especially in the lower reaches, and its floodplain ranges in width from 0.25 to 1 mile. It has a steep 

southeastern side and gently sloping northwestern sides. [DOE/EIS-0303] 

Fourmile Branch is a blackwater stream that originates near the center of the SRS and flows southwest 

for 15 miles before emptying into the Savannah River. It drains an area of approximately 22 square miles 

inside SRS including much of F, H and C Areas. Fourmile Branch flow is generally perpendicular to the 

Savannah River behind natural levees and enters the river through a breach downstream from Beaver 

Dam Creek. In its lower reaches, Fourmile Branch broadens and flows via braided channels through a 

delta formed by the deposition of sediments eroded from upstream during high flows. Downstream from 

the delta, the channels rejoin into one main channel. Most of the flow discharges into the Savannah River 

while a small portion flows west and enters Beaver Dam Creek. The valley is V-shaped, with sides 

varying from steep to gently sloping. The floodplain is up to 1,000 feet wide. [DOE/EIS-0303] 

Flood hazard recurrence frequencies have been calculated for the various SRS site areas. The 

calculated flood water levels for Fourmile Branch near H Area, for the probability of 100-year, 1,000-year 

and 10,000-year returns are about 234.3, 235.2 and 235.8 feet above mean sea level (MSL), 

respectively. As shown in Section 2.1.11, the lowest elevation of any waste tank basemat in HTF is 239.9

feet above MSL; thus, the highest flood water level of approximately 236 feet above MSL is below the 

lowest elevation of residual radioactive material. In addition, the lowest elevation of the lower foundation 

layer of the proposed closure cap is 280 feet above MSL, which is about 44 feet above the highest flood 

water level of 236 feet. Therefore, flooding will not affect the HTF and is therefore not considered in the 

HTF PA. [WSRC-TR-99-00369, SRNL-ESB-2008-00023] 

2.1.6 Geochemistry 

The migration of radionuclides in the subsurface environment is dependent on physical and chemical 

parameters or properties of cementitious materials, soils and groundwater. Studies and analyses have 

been conducted to determine appropriate distribution coefficient values. The data used in the 

radionuclide transport model is presented in Section 4.2.2 and 4.2.3.2 of the HTF PA specific to the GSA 

and is not reproduced in this section. [SRR-CWDA-2010-00128] 

2.1.7 Natural Resources 

Natural resources at SRS are managed under the Natural Resources Management Plan for the 

Savannah River Site (NRMP) prepared for the DOE by the U.S. Department of Agriculture Forest Service 

- Savannah River. [NRMP-2005] The NRMP, which governs SRS natural resource management, was 

updated in May 2005 and fosters the following principles : 

 All work will be done in accordance with integrated safety management components found in 

DOE Policy 450.4A, Integrated Safety Management Policy.

 Environmental stewardship activities will be compatible with future SRS missions. 

 The SRS will continue to protect and manage SRS natural resources. 

 Sustainable resource management will be applied to SRS natural resources. 

 Close cooperation will be maintained among organizations when managing and protecting SRS 

natural resources. 

 The results of research, monitoring and operational findings will be used in the management of 

SRS natural resources. 

 Restoration of native communities and species will continue. 

 Employees, customers, stakeholders, state natural resource officials and regulators will be invited 

to participate in the natural resource planning process. 

 The SRS will maintain the area as a National Environmental Research Park. 

2.1.7.1 Water Resources 

The SRS monitors non-radioactive liquid discharges to surface waters through the NPDES, as mandated 

by the Clean Water Act. As required by EPA and SCDHEC, SRS has NPDES permits in place for 

discharges to the waters of the United States and South Carolina. These permits establish the specific 

sites to be monitored, parameters to be tested, and monitoring frequency, as well as analytical, reporting, 

and collection methods. Continuous surveillance monitoring of site streams occurs downstream of 
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several process areas to detect and quantify levels of radioactivity in effluents transported to the 

Savannah River. [SRNS-STI-2011-00059] 

Table 2.1-3 characterizes Savannah River water quality both upstream and downstream of SRS. Table 

2.1-4 characterizes water quality in UTR and Fourmile Branch downstream of the GSA. 

Table 2.1-3: Water Quality in the Savannah River Upstream and Downstream from SRS (Calendar 

Year 2010) 

Parameter Unit of Measure 

Upstreamb

 Downstreamc

Minimum Maximuma Minimum Maximuma

Aluminum mg/L 0.105 0.487 0.11 0.57 

Cadmium mg/L ND ND ND ND 

Chromium mg/L ND ND ND ND 

Copper mg/L ND 0.517 ND 0.0083 

Dissolved Oxygen mg/L 4.6 19.9 4.28 11.31 

Gross Alpha 

Radioactivity pCi/L ND 1.59 ND 1.31 

Lead mg/L ND ND ND 0.0023 

Mercury mg/L ND 0.000023 ND 0.000024 

Nickel mg/L ND ND ND 0.0066 

Nitrate (as N) mg/L 0.18 0.38 0.12 0.58 

pH pH units 6.25 7.32 6.42 7.41 

Phosphate mg/L 0.095 0.17 0.079 0.17 

Suspended solids mg/L 2 10 5 20 

Temperature °F 44.2 75.9 43.3 79.2 

Tritium pCi/L ND 265 98.6 957 

Zinc mg/L 0.0022 0.0087 0.0013 0.0352 

a

 The maximum listed concentration is the highest single result found during one sampling event. b

 Data from sampling location RM-160. 

c

 Data from sampling location RM-118.8. 

Note: Information extracted from SRNS-STI-2011-00059 accompanying data files. Parameters are those DOE routinely measures as a regulatory 

requirement, or as part of ongoing monitoring programs. 

ND - Non Detectable 

Table 2.1-4: Water Quality in Selected SRS Streams (Calendar Year 2010) 

Temperature (o

F) pH 

Dissolved 

Oxygen (mg/L) 

Total Suspended 

Solids (mg/L) 

Sampling Location: Fourmile Branch (Downstream of GSA)a

Mean 60.2 6.8 7.0 2.9 

Range 39.0 – 77.2 6.4 – 7.2 3.3 – 11.5 0 - 6 

Sampling Location: Upper Three Runs (Downstream of GSA)b

Mean 58.2 6.2 7.8 6.1 

Range 42.5 – 75.0 5.7 - 7.2 4.9 – 16.1 1 - 12 

a

 Data from sampling location FM-6. 

b

 Data from sampling location U3R-4. 

Note: All data extracted from SRNS-STI-2011-00059 accompanying data files 

2.1.7.1.1 Groundwater 

The Safe Drinking Water Act was enacted in 1974 to protect public drinking water supplies. SRS 

domestic water is supplied by 17 separate systems, all of which utilize groundwater sources. The A-Area 

and D-Area drinking water facilities are actively regulated by SCDHEC, while the remaining smaller water 

systems receive a reduced level of regulatory oversight. The K-Area drinking water system was 
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incorporated into the A Area system in 2010, and removed from SCDHEC’s water system inventory. 

[SRNS-STI-2011-00059] 

Table 2.1-5 provides the summary of maximum groundwater monitoring results for those areas that most 

likely discharge to UTR or Fourmile Branch obtained from the Savannah River Site Environmental Report 

for 2008, which represents the latest annual summary of well monitoring results summarized by area. 

[SRNS-STI-2009-00190] The groundwater in these areas is not being consumed and active remediation 

projects are in progress to address the groundwater conditions. 

Table 2.1-5: Well Monitoring Results for Major Areas within SRS, 2007–2008 

Location 

Major 

Contaminants Units 

2007 

Maximum MCL 

2008 

Maximum Likely Outcrop Point 

E Area Tritium 

TCE 

pCi/L 

ppb 

30,800,000 

370 

20,000 

0.5 

29,200,000 

460 

UTR/Crouch Branch in 

North; Fourmile Branch in 

South 

F Area 

TCE 

Tritium 

Gross alpha 

Beta 

ppb 

pCi/L 

pCi/L 

pCi/L 

52.2 

73,000 

2,120 

380 

5.0 

20,000 

15 

4 mrem/yra

60 

130,000 

1,470 

628 

UTR/Crouch Branch in 

North; Fourmile Branch in 

South 

F-Area 

Seepage 

Basin 

Tritium 

Gross alpha 

Beta 

pCi/L 

pCi/L 

pCi/L 

5,710,000 

523 

1,870 

20,000 

15 

4 mrem/hra

4,810,000 

777 

2,100 

Fourmile Branch 

H Area 

Tritium 

Gross alpha 

Beta 

pCi/L 

pCi/L 

pCi/L 

67,200 

25.5 

55.6 

20,000 

15 

4 mrem/yra

74,800 

14.9 

81.9 

UTR/Crouch Branch in 

North; Fourmile Branch in 

South 

H-Area 

Seepage 

Basins 

Tritium 

Gross alpha 

Beta 

pCi/L 

pCi/L 

pCi/L 

3,020,000 

88.4 

2,970 

20,000 

15 

4 mrem/yra

3,120,000 

85 

2,050 

Fourmile Branch 

a

 The activity (pCi/L) equivalent to 4 mrem/yr varies according to which specific beta emitters are present in the sample. [SRNS-STI-2009-00190] 

2.1.8 Natural and Background Radiation 

All human beings are exposed to sources of 

ionizing radiation that include naturally occurring 

and man-made sources. An individual’s average 

dose contribution estimates from various sources 

were obtained from information presented in 

National Council on Radiation Protection and 

Measurements (NCRP) Report 160 and are shown 

in Figure 2.1-27. [NCRP-160] On average, a 

person living in the United States or the Central 

Savannah River Area (CSRA) receives 

approximately the same annual radiation dose of 

620 mrem/yr. The dose from SRS operations to the 

maximally exposed off-site individual during 

calendar year 2010 was estimated to be 0.1 mrem. 

[SRNS-STI-2011-00059] 

The major sources of radiation exposure to an 

average member of the public in the CSRA is 

attributed to naturally occurring radiation 

(311 mrem/yr) and medical exposure (300 

mrem/yr). This naturally occurring radiation is often 

referred to as natural background radiation and 

includes dose from background radon and its decay 

products (228 mrem/yr), cosmic radiation (33 

mrem/yr), internal radionuclides occurring naturally 

in the body (29 mrem/yr) and natural radioactive 

Figure 2.1-27: Major Sources of Radiation 

Exposure in the Vicinity of SRS 
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material in the ground (21 mrem/yr). The dominant medical sources include dose from computed 

tomography (147 mrem/yr), nuclear medicine (77 mrem/yr) and radiography/fluoroscopy (77 mrem/yr). 

The remainder of the dose is from consumer products (13 mrem/yr), industrial/educational/research 

activities (less than 1 mrem/yr) and occupational exposure (less than 1 mrem/yr). [NCRP-160] 

2.1.9 Tank Farm Operations 

Since initiation of operations at SRS, the tank farms combined have received over 140,000,000 gallons of 

liquid waste. The vast majority of this liquid waste originated from the chemical separation processes in F 

and H Canyons associated with three major missions: 1) reprocessing of spent nuclear fuel; 2) production 

of nuclear materials for weapons; and 3) production of material for National Aeronautics and Space 

Administration (NASA) space missions. [SRR-LWP-2009-00001] In addition to the Canyon waste 

streams, the DWPF returns a liquid waste stream (referred to as “DWPF recycle”) to the FTF and HTF. 

This “DWPF recycle” liquid waste stream is a by-product of the production of waste canisters in DWPF. 

These canisters contain SRS high-level waste stabilized in borosilicate glass. [DOE/EIS-0303] 

The F- and H-Canyon chemical separations processes used acids to dissolve irradiated reactor target or 

fuel assemblies to prepare the desired products for extraction. The DWPF process also uses an acidbased process. The resultant waste stream from all of these processes is acidic. Before transferring the 

waste material from the F and H Canyons and DWPF to the tank farms, sodium hydroxide is added 

adjusting the waste to a high alkaline state to prevent corrosion of the carbon steel waste tanks. This 

chemical adjustment results in the precipitation of solids. These solids settle in the waste tanks forming a 

layer that is commonly referred to as “sludge.” These solids are comprised of fine particles of settled 

metal oxides, including, in small part, uranium, strontium and plutonium hydroxides. These solids are 

insoluble due to the chemical nature of the solution. After settling of the solids has occurred, the liquid 

salt waste solution (supernate) above this sludge layer is transferred out of the waste tank. To maximize 

the space available in the waste tanks for storing additional waste, DOE’s practice at SRS has been to 

use the tank farm evaporator systems to reduce the volume of the decanted supernate by concentrating 

the waste. [HLW-2002-00025] 

During the evaporation process, the liquid salt waste is concentrated. After the concentrated salt waste is 

returned to the waste tank, the concentrated salt waste forms two distinct phases (collectively called salt 

waste): 1) concentrated supernate solution and 2) solid saltcake. The predominant radionuclide present 

in the salt waste is Cs-137. Because of the high solubility of Cs-137, approximately 95 % of the Cs-137 is 

present in the concentrated supernate solution and the liquid found within the interstitial spaces in 

saltcake. The solid saltcake is composed predominantly of nitrate and nitrite salts and contains relatively 

small quantities of insoluble radioactive solids. When saltcake is dissolved and removed from the tank, 

these entrained insoluble solids eventually settle on the waste tank bottom adding to the sludge inventory. 

[SRR-LWP-2014-00030] 

As of July 1, 2014, as the result of the evaporation process, the combined total of more than 140,000,000 

gallons of liquid waste originally received in FTF and HTF had been reduced to approximately 36,800,000 

gallons. [SRR-LWP-2009-00001, SRR-LWP-2014-00030] The SRS no longer conducts weapons or 

NASA-related material recovery activities or the weapons-related spent nuclear fuel reprocessing that 

generated the original waste. The DOE has deactivated the Plutonium Uranium Extraction (PUREX) 

process in F-Canyon (one of the two chemical separations canyon facilities) and is no longer generating 

radioactive liquid waste for storage in the FTF. H Canyon continues to generate radioactive liquid waste 

when performing stabilization missions such as recovering and blending highly enriched uranium for nondefense related use. DOE carefully controls waste transferred to the HTF waste tanks and ancillary 

structures through a waste compliance program and waste acceptance criteria which establish the 

physical, chemical and radionuclide limits for all waste entering the waste tanks and ancillary structures. 

All generating facilities, including H Canyon, are required to develop and maintain a Waste Compliance 

Plan which describes the controls or procedures imposed within the facility to ensure that the HTF waste 

acceptance criteria are met for any material transferred to the HTF waste tanks and ancillary structures. 

[X-SD-G-00001] 

Ongoing SRS operations continue to require the need for waste tank space. Most of the SRS Type III/IIIA 

waste storage tanks, tanks that meet current secondary containment and leak detection standards and 

have no prior leak sites, are already at or near full capacity. Projected available waste tank space is 
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carefully tracked to ensure the tank farms do not become “water logged,” a term meaning that so much of 

the useable compliant waste tank space has been filled that normal operations, waste removal and waste 

processing, cannot effectively continue. Substantial amounts of waste tank space are required to safely 

and effectively remove tank waste and prepare it for disposal. This includes the preparation of highactivity sludge waste for vitrification in DWPF. 

The preparation of saltcake for disposition also requires significant waste tank space because the solid 

saltcake must be dissolved to make it mobile for processing. The dissolution of saltcake typically requires 

a ratio of approximately three gallons of water to one gallon of saltcake to properly dissolve the saltcake 

into liquid salt solution. [CBU-PIT-2005-00031] 

Waste tank space for this liquid addition to the tank farm inventory must be available to allow for efficient 

salt processing and disposition and, ultimately, tank closure activities. A portion of the available waste 

tank space must also be reserved as contingency space in the event a new waste tank leak occurs. The 

tank farms also receive new waste from the H-Canyon and other waste streams as discussed in Section 

2.2.1 of this HTF 3116 Basis Document. 

2.1.10 H-Tank Farm 

The H Area occupies approximately 395 acres. It includes HTF and the H-Canyon Separations Facility. 

The HTF is located in the central region of H Area and occupies approximately 45 acres. The HTF is an 

active waste storage facility used to store liquid radioactive waste generated primarily during prior 

operations of H-Canyon. Figure 2.1-28 shows an aerial view of the HTF. 

Figure 2.1-28: General Layout of HTF 
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The HTF consists of: 

 29 carbon steel waste tanks (i.e., Tanks 9 through 16, Tanks 21 through 24, Tanks 29 through 

32, Tanks 35 through 43 and Tanks 48 through 51), 

 ancillary structures: 

− the HTF transfer line system including approximately 74,800 linear feet of underground waste 

transfer lines, 

− 10 pump pits (HPP-1 through HPP-10), 

− nine pump tanks (HPT-2 through HPT-10), 

− one 11,700 gallon catch tank, 

− three evaporator systems (242-H, 242-16H and 242-25H), 

− the 242-3H and 242-18H concentrate transfer systems (CTSs), 

− eight diversion boxes (HDB-1 through HDB-8), and 

− 11 valve boxes (valve box 15/16, Tanks 21 and 22 valve boxes, Tank 40 valve box, Tank 40 

drain valve box, Tank 42 valve box, Tank 49 valve box, Tank 50 valve box, Tank 51 valve 

box, Tank 51 drain valve box and 241-96H valve box). 

This equipment is discussed separately as follows: 

 waste tanks (Type I, II, III/IIIA and IV) 

 ancillary structures 

2.1.11 Waste Tanks 

All 29 of the HTF waste tanks are built of carbon steel and reinforced concrete, but the designs vary. 

There are four principal types of waste tanks in the HTF, designated as Type I, II, III/IIIA and IV tanks. 

The waste tanks were numbered sequentially based on time of design and siting, and are not tank farm 

specific. For example, the original twelve tanks constructed at SRS were of Type I design and were 

numbered Tanks 1 through 8 in FTF and Tanks 9 through 12 in HTF. In HTF, eight waste tanks, Tanks 9 

through 16, along with some ancillary structures, are either fully submerged or partially submerged in the 

water table. [SRR-CWDA-2010-00128] Table 2.1-6 summarizes the important design features. 

Table 2.1-6: Waste Tank Nominal Capacities, Nominal Dimensions and Other Features 

Waste Tank Numbers 

Diameter 

(ft) 

Height

(ft) 

Capacity 

(gallons) 

Cooling 

Coils 

Secondary 

Containment 

Elevation 

(ft)a

Distance 

from 

Basemat to 

Water 

Table (ft)b

Type I 

9, 10, 11, 12 75.0 24.5 750,000 Yes Yes 239 - 241 -36 

Type II 

13, 14, 15, 16 85.0 27.0 1,030,000 Yes Yes 270 -7 

Type III/IIIAc

29, 30, 31, 32, 35, 

36, 37, 38, 39, 40, 

41, 42, 43, 48, 49, 

50, 51 

85.0 33.0 1,300,000 Yes Yes 280 - 293 8 - 18 

Type IV 

21, 22, 23, 24 85.0 34.5 1,300,000 No No 281 7 

a Approximate feet above MSL for the waste tank basemat. b Approximate distance based on GSAD water table elevations. Negative value reflects distance below water table. c Tanks 29 through 32 are Type III tanks. The remaining waste tanks are Type IIIA. 

The general design features of the waste tanks are summarized below, followed by brief descriptions and 

illustrations of the different waste tanks types. 
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2.1.11.1 General Tank Design Features 

Each Type I, II, III and IIIA tank has a primary tank and a carbon steel secondary containment liner. The 

secondary liner for Type III and IIIA tanks extends the full height of the primary tank. The Type I and II 

tank secondary liner extends only 5 feet above the bottom of primary tank floor and is sometimes referred 

to as an “annulus pan.” Because the secondary liner is larger in diameter than the primary tank, an 

annular space exists between them. This waste tank annulus, which differs in size and capacity for each 

waste tank type, provides a collection point for any potential leakage from the primary tank, as well as a 

method for heating or cooling the primary tank wall in conjunction with the annulus ventilation system. 

Reinforced concrete vaults surround the secondary liner. 

The Type IV waste tank consists of a reinforced concrete vault surrounding a single carbon steel liner. 

The Type IV tanks do not have an annulus. The reinforced concrete vault provides both structural 

support and radiation shielding for all the waste tanks. The bottom part of the vault is called the basemat. 

Underneath the basemat of the Type I, II, III and IIIA tanks lies a concrete working slab. The Type IV 

waste tanks do not have a concrete working slab below the basemat. 

Chromate cooling water provides the primary cooling for the waste stored in the waste tanks, which flows 

through cooling coils located inside the primary waste tank. The cooling coils are installed in Type I, II, III 

and IIIA tanks and vary in design for each waste tank type. Type IV tanks do not contain cooling coils. 

Risers provide access to the primary waste tank and annulus interiors. Risers are used primarily for 

inspections, level detection, dip samples and the installation of equipment such as annulus jets, dip tubes, 

thermocouples, conductivity probes, ventilation inlet and outlets, reel tapes, hydrogen monitors and waste 

removal equipment. Lead or concrete plugs are inserted in the riser opening if no equipment is installed. 

The riser structures are made of concrete and lined with carbon steel. Riser layout is dependent on the 

specific waste tank. 

2.1.11.2 Type I Tanks 

Type I tanks (Tanks 9 through 12) were constructed in the early 1950s. These waste tanks are 75 feet in 

diameter and 24.5 feet in height with a nominal operating capacity of 750,000 gallons. The tank tops are 

approximately 9 feet below grade. All Type I tanks have a secondary carbon steel liner 80 feet in 

diameter and 5 feet high (2.5 feet annulus space). All Type I tanks have similarly configured vertical and 

horizontal cooling coils. A typical26 Type I tank cross section is shown in Figure 2.1-29, waste tank 

concrete basemat construction is shown in Figure 2.1-30 and a waste tank primary tank and secondary 

liner construction is shown in Figure 2.1-31. Additional Type I tank details are provided in Section 3.0 of 

the HTF PA. [SRR-CWDA-2010-00128] 

Figure 2.1-29: Typical Type I Tank Cross Section 

 

26 The word “typical” as used throughout this section refers to representative design features of the system being described. 
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The primary tank is made of 0.5-inch thick carbon steel. The walls are joined to the roof and floor of the 

primary tank by curved knuckle plates made of the same material and are welded in place. The 

secondary liner is also made of 0.5-inch thick carbon steel. Transfer line penetrations allow three-inch 

diameter inlet waste transfer lines to enter the primary waste tank near the top through the top knuckle. 

Each transfer line is enclosed in a four-inch diameter carbon steel jacket pipe where it bridges the waste 

tank annulus. [SRR-CWDA-2010-00128] 

Figure 2.1-30: Type I Tank Basemat 

Construction 

Figure 2.1-31: Type I Tank Primary and 

Secondary Liner Construction 

The waste tank vault is constructed of 22-inch thick reinforced concrete with an inner diameter of 80 feet. 

Approximately 9 feet of soil covers the vault roof as shown in Figure 2.1-29. [SRR-CWDA-2010-00128] 

Each Type I tank has 12 concrete filled steel columns to support the roof. These columns have an outer 

diameter of 2 feet of 0.5-inch carbon steel pipe filled with concrete and welded to the top and bottom of 

the primary tank. A waste tank column at the time of construction is shown in Figure 2.1-32. [SRRCWDA-2010-00128] 

Cooling coils in Type I tanks are configured in both a 

horizontal and a vertical array, which creates obstacles 

to waste removal and other activities inside the waste 

tank (Figure 2.1-33). Each Type I tank contains 34 

vertical cooling coils that are supported from the 

primary tank roof by hanger and guide rods, which are 

welded to the primary tank. All combined, the vertical 

coils consist of 604 vertical sections 18.5 feet long with 

604 loops (half circle with a 24-inch radius) that connect 

the vertical sections. Two horizontal cooling coils 

(upper and lower) extend across the bottom of the 

waste tanks and are supported by guide rods welded to 

the primary tank floor. The lower horizontal cooling coil 

is approximately 1 inch above the tank floor and the 

upper horizontal cooling coil is approximately 4 inches 

above the primary tank floor. The horizontal coils 

consist of 26 horizontal sections and 26 loops (half 

circle with a 24-inch radius) that connect the horizontal 

sections. The horizontal cooling coil runs at the bottom 

of the Type I tanks were “field to fit” during the time of 

waste tank construction. In addition, there are supply 

pipes that connect the tank top cooling water system to 

the cooling coils. There are approximately 22,800 

linear feet of two-inch carbon steel pipe cooling coils in 

a Type I tank. [D116048, C-CLC-G-00364, D116001] 

Figure 2.1-32: Typical Type I Tank Column 

Support 
Basis for Section 3116 Determination DOE/SRS-WD-2014-001 

for Closure of H-Tank Farm Revision 0 

at the Savannah River Site December 2014 

Page 2-36 

 

Visual and equipment manipulation access within the Type I tank is limited by the tank riser design 

configuration. Riser configuration, above the tank top, limits direct access to equipment and allows a 

limited view of the primary tank floor as shown in Figure 2.1-34. Additionally, the size of the access ports 

limits the manipulation of long-handled mechanical tools. Due to access port geometry, choices are 

limited as to the types of remote equipment that can be successfully deployed. Type I tanks have a 

42-inch diameter center riser and eight 24-inch perimeter risers. Each riser is approximately 10-feet-10-

inches in length (Figure 2.1-34). [W146625] 

Figure 2.1-33: Typical Type I Tank Cooling 

Coil Obstacles 

Figure 2.1-34: Type I Waste Access Area for 

Waste Removal Equipment Diagram 

2.1.11.3 Type II Tanks 

Type II tanks (Tanks 13 through 16) were constructed between 1955 and 1956. These waste tanks are 

85 feet in diameter and 27 feet in height with a nominal operating capacity of 1,030,000 gallons. The 

backfill around the waste tanks was brought to an elevation level with the top of the waste tanks 

(approximately 395 feet above MSL) and extended laterally for a minimum of 21 feet. The backfill was 

then sloped down at an angle less than 1:1 for a lateral distance of 31 feet, reaching final grade at an 

elevation of 300 feet above MSL. All Type II tanks have a secondary carbon steel liner 90-foot 3-inch 

diameter and 5 feet high (2-foot 6.125-inch annulus space). All Type II tanks have similarly configured 

vertical and horizontal cooling coils. A typical27 Type II tank cross section is shown in Figure 2.1-35, 

waste tank concrete basemat construction is shown in Figure 2.1-36 and a waste tank primary tank and 

secondary liner construction is shown in Figure 2.1-37. Additional Type II tank details are provided in 

Section 3.0 of the HTF PA. [SRR-CWDA-2010-00128] 

 

27 The word “typical” as used throughout this section refers to representative design features of the system being described. 
Basis for Section 3116 Determination DOE/SRS-WD-2014-001 

for Closure of H-Tank Farm Revision 0 

at the Savannah River Site December 2014 

Page 2-37 

 

Figure 2.1-36: Type II Tank Basemat 

Construction 

Figure 2.1-37: Type II Tank Primary and 

Secondary Liner Construction 

The primary tank is made of carbon steel with varying thicknesses shown in Table 2.1-7. The walls are 

joined to the roof and floor of the primary tank by curved knuckle plates made of the same material and 

are welded in place. The secondary liner is 

also made of 0.5-inch thick carbon steel. 

Transfer line penetrations allow three-inch 

diameter inlet waste transfer lines to enter the 

primary waste tank near the top through the 

top knuckle. Each transfer line is enclosed in a 

four-inch diameter carbon steel jacket pipe 

where it bridges the waste tank annulus. 

[SRR-CWDA-2010-00128] 

A soil hydration system and five feed wells were installed beneath the Type II tanks to address potential 

issues with soil shrinkage and settlement. The hydration system consists of an interconnecting grid 

comprised of four-inch diameter drain tile (perforated piping) located 18 inches below the working slab 

(Figure 2.1-38). The soil hydration system was never used for soil hydration since the water table under 

the Type II tanks is higher than anticipated and soil dehydration is not a problem. However, in the past, 

this soil hydration system was used to monitor groundwater levels. The soil hydration system wells were 

Figure 2.1-35: Typical Type II Tank Cross Section 

Table 2.1-7: Type II Tanks Primary Tank Liner 

Plate Data 

Location Thickness (inch) 

Primary tank roof 0.5 

Primary tank floor 0.5 

Upper knuckle 0.562 

Primary tank wall 0.625 

Lower knuckle 0.875 
Basis for Section 3116 Determination DOE/SRS-WD-2014-001 

for Closure of H-Tank Farm Revision 0 

at the Savannah River Site December 2014 

Page 2-38 

 

used in the 1960s and 1970s to pump water from beneath Tank 16 as part of the Tank 16 groundwater 

monitoring effort. 

The waste tank vault is constructed of 33-inch thick 

reinforced concrete walls and 45-inch thick 

reinforced concrete roof with an outer diameter of 

95 feet 8.5 inches. [SRR-CWDA-2010-00128] 

Each Type II tank has one central filled steel 

column to support the roof. This column has an 

inner diameter of 6 feet 8 inches of 0.5-inch thick 

carbon steel that was welded to the bottom of the 

primary tank and filled with concrete. A waste tank 

support column dimension detail is shown in Figure 

2.1-39. [SRR-CWDA-2010-00128] 

Cooling coils in Type II tanks are configured in both 

a horizontal and a vertical array, which creates 

obstacles to waste removal and other activities 

inside the waste tank (Figure 2.1-40). Each Type II 

tank contains 40 vertical cooling coils (20 

operating, 20 auxiliary) that are supported from the 

primary tank roof by hanger and guide rods, which 

are welded to the primary tank. The vertical coils 

consist of approximately 20 foot-long vertical 

sections connected with 24-inch radius half circle 

loops. Four horizontal cooling coils (two upper 

operating, two lower auxiliary) extend across the 

bottom of the waste tanks and are supported by 

guide rods welded to the primary tank floor. The 

lower horizontal cooling coil is approximately 1 inch 

above the tank floor and the upper horizontal 

cooling coil is approximately 4 inches above the 

primary tank floor. The horizontal coils consist of 

40 horizontal sections and 36 loops (half circle with 

a 24-inch radius) that connect the horizontal 

sections. In addition, there are supply pipes that 

connect the tank top cooling water system to the 

cooling coils. There are approximately 29,400 

linear feet of two-inch carbon steel pipe cooling 

coils in a Type II tank. [SRR-CWDA-2010-00128, 

W163593, W163658] 

Visual and equipment manipulation access within 

the Type II tank is limited by the tank riser design 

configuration. Riser configuration, above the tank 

top, limits direct access to equipment and allows a 

limited view of the primary tank floor as shown in 

Figure 2.1-41. Additionally, the size of the access 

ports limits the manipulation of long-handled 

mechanical tools. Due to access port geometry, 

choices are limited as to the types of remote 

equipment that can be successfully deployed. Type II tanks were constructed with ten 24-inch risers and 

one 42-inch riser. Each riser is approximately 3-feet 9-inches in length (Figure 2.1-41). [SRR-CWDA2010-00128, W163012] 

 

Figure 2.1-38: Soil Hydration System Below 

Type II Tanks 

Figure 2.1-39: Typical Type II Support Column 

Dimension Details 
Basis for Section 3116 Determination DOE/SRS-WD-2014-001 

for Closure of H-Tank Farm Revision 0 

at the Savannah River Site December 2014 

Page 2-39 

 

Figure 2.1-40: Typical Type II Tank Cooling 

Coil Obstacles 

Figure 2.1-41: Type II Waste Access Area for 

Waste Removal Equipment Diagram 

2.1.11.4 Type III/IIIA Tanks 

The Type III tanks were constructed between 1966 and 1970 (Tanks 29 through 32). The Type IIIA tanks 

were complete between 1974 and 1981 (Tanks 35 through 37, 38 through 43 and 48 through 51) All 

Type III/IIIA tanks have full secondary containment by a secondary steel liner, which is 90 feet in diameter 

and 33 feet in height (2.5-foot annulus). Note that Tanks 35, 36 and 37 have been designated as Type 

IIIA tanks, but they differ from Figure 2.1-43 in that these waste tanks have a flat roof with a uniform fourfoot concrete thickness, similar to the Type III tanks. Additionally, Tank 35 has insertable cooling coils 

rather than permanently installed cooling coils. [SRR-CWDA-2010-00128] 

Each Type III tank has a minimum 3-foot-6-inch thick reinforced concrete basemat placed on top of a 

six-inch concrete working slab. The Type IIIA tanks have a minimum 3-foot-7-inch thick reinforced 

concrete basemat placed on top of a four-inch concrete working slab. Both Type III and IIIA tanks have a 

center drop panel that varies the basemat thickness for each tank type. Additional Type III/IIIA tank 

details are available in Section 3.0 of the HTF PA. [SRR-CWDA-2010-00128] 

Tanks 35 through 37 each had 56 thermocouples installed on the outside of the primary tank. Tank 38 

has 34 thermocouples that were installed on the outside of the primary tank. Each of the remaining Type 

IIIA tanks (39 through 43 and 48 through 51) have 22 thermocouples that were installed on the outside of 

the primary tank. All Type IIIA tanks have a thermocouple located on the top of the basemat and another 

thermocouple located just below the working slab. Tanks 35 through 37 have an additional thermocouple 

that was installed approximately 10 feet below the working slab. [SRR-CWDA-2010-00128] 

A typical HTF Type III tank cross section is shown in Figure 2.1-42, typical Type IIIA tank cross section is 

shown in Figure 2.1-43 and Typical Type III/IIIA tank basemat construction is shown in Figure 2.1-44. 
Basis for Section 3116 Determination DOE/SRS-WD-2014-001 

for Closure of H-Tank Farm Revision 0 

at the Savannah River Site December 2014 

Page 2-40 

 

Figure 2.1-42: Typical HTF Type III Tank Cross Section 

Figure 2.1-43: Typical Type IIIA Tank Cross Section 

 
Basis for Section 3116 Determination DOE/SRS-WD-2014-001 

for Closure of H-Tank Farm Revision 0 

at the Savannah River Site December 2014 

Page 2-41 

 

The Type III/IIIA tank primary tanks are 

made of carbon steel. The walls are 

joined to the roof and floor plates by 

curved knuckle plates made of the same 

material. The primary tank and 

secondary liner for a Type IIIA tank late 

in the construction phase are shown in 

Figure 2.1-45. The Type III/IIIA tank 

primary tanks were fully stress-relieved 

by heating after fabrication. [SRR-STI2014-00243] The waste tank primary 

tank and secondary liner plate data are 

summarized in Table 2.1-8. [SRRCWDA-2010-00128] 

Both Type III and IIIA tanks contain 

multiple penetrations through a carbon 

steel primary tank (e.g., two-inch or 

three-inch diameter pipe in six-inch or 

ten-inch diameter sleeves, respectively) 

near the top of the waste tank for 

transfer lines into and out of the tank. 

[SRR-CWDA-2010-00128] 

Figure 2.1-44: Typical Type III/IIIA Tank Basemat 

Construction 

Figure 2.1-45: Typical Type IIIA Tank Primary Tank and Secondary Liner - Late Construction 

(Vault Wall Not Constructed) 

Upper Knuckle

Basemat 

Secondary Liner

Primary Tank Wall
Basis for Section 3116 Determination DOE/SRS-WD-2014-001 

for Closure of H-Tank Farm Revision 0 

at the Savannah River Site December 2014 

Page 2-42 

 

The Type III/IIIA 

primary waste tanks 

are completely 

enclosed in a 

concrete vault. The 

vault roof is at least 

48 inches thick and 

the walls are 30 

inches thick; 

therefore, there is 

no earthen cover for 

shielding on top of 

these waste tanks. 

Type III/IIIA tanks 

have both a center 

and outer annulus. The center annulus is formed between the primary waste tank wall and the roof 

support column. This design allows for ventilation airflow to the underside of the primary tank floor and 

then out to the outer annulus through the radial air grooves, as shown in Figure 2.1-46. 

Type IIIA tank tops are 5 feet thick at the waste tank center, 4 feet thick at the edge and sloped to allow 

rainwater drainage. In the Type III/IIIA designs, the primary waste tank roof is supported by a steel-lined 

center support column that is integrated into the concrete basemat (Figure 2.1-42 and Figure 2.1-43). 

Ventilation systems are embedded in the center column. Air flows through the column, into the center 

annulus, through the radial air grooves and exits through the outer annulus. [SRR-CWDA-2010-00128] 

Type III tanks cooling coil piping 

consists of insertable coolers 

(consolidated bundles, conical 

coolers and cylindrical coolers as 

shown in Figure 2.1-47). These 

coolers were inserted through the 

risers in the closed position and 

in the case of the conical and 

cylindrical coolers deployed 

(opened) once inside the waste 

tank. The coolers are supported 

by the tank top. The bottoms of 

the waste tanks are cooled by 

the air passing through the 

annulus and radial air grooves. 

Type IIIA tanks, except Tank 35, 

have permanently installed 

cooling coils as shown in Figure 2.1-48. The Type IIIA cooling coils have top and bottom supported 

vertical coils. There are 246 vertical coils mounted nine inches off the bottom of the waste tank and 

spaced on three 8-foot centers. These are made of two-inch diameter carbon steel pipe. [SRR-CWDA2010-00128] 

 

Table 2.1-8: Type III/IIIA Tanks Primary Tank and Secondary Liner Plate Data 

Location 

Tanks 

29 - 32 

Thickness 

(inch) 

Tanks 

35 - 37 

Thickness 

(inch) 

Tanks 

38 - 43 

Thickness 

(inch) 

Tanks 

48 - 51 

Thickness 

(inch) 

Primary tank roof 0.5 0.5 0.5 0.5 

Primary tank floor 0.5 0.5 0.5 0.5 

Upper knuckle 0.5 0.5 0.5 0.5 

Secondary liner - upper band 0.5 0.5 0.5 0.5 

Secondary liner - middle band 0.5 0.625 0.625 0.625 

Secondary liner - lower band 0.5 0.5 0.875 0.875 

Primary tank wall - upper band 0.5 0.625 0.5 0.5 

Primary tank wall - lower band 0.5 0.875 0.625 0.625 

Lower knuckle - secondary 1.0 0.875 0.875 0.875 

Lower knuckle - primary 0.625 0.625 0.625 0.625 

Figure 2.1-46: Typical Type III/IIIA Tank Radial Air Grooves 
Basis for Section 3116 Determination DOE/SRS-WD-2014-001 

for Closure of H-Tank Farm Revision 0 

at the Savannah River Site December 2014 

Page 2-43 

 

Figure 2.1-47: Typical Type III Tank Insertable Cooling Coils 

Figure 2.1-48: Typical Type IIIA Tank Cooling Coils 

During Construction 

The waste tank riser design configuration restricts visual and equipment access into Type III/IIIA tanks. 

These waste tanks have 37 to 42 risers, ranging from 8 to 52 inches in diameter, and are approximately 

4.5 feet in length as shown in Figure 2.1-49 and Figure 2.1-50. [W704010, W449795] 

 

Consolidated Conical Cylindrical 

Bundle
Basis for Section 3116 Determination DOE/SRS-WD-2014-001 

for Closure of H-Tank Farm Revision 0 

at the Savannah River Site December 2014 

Page 2-44 

 

Figure 2.1-49: Typical Type III Tank Access 

Area for Waste Removal Equipment Diagram 

Figure 2.1-50: Typical Type IIIA Tank Access 

Area for Waste Removal Equipment Diagram 

2.1.11.5 Type IV Tanks 

The Type IV tanks were 

constructed between 1958 

and 1962. Type IV tanks are 

85 feet in diameter and 

approximately 34 feet in 

height at the side wall with a 

nominal operating capacity of 

1,300,000 gallons. Type IV 

tanks do not have a 

secondary liner or cooling 

coils. These waste tanks 

have a single carbon steel 

liner with a self-supporting, 

reinforced concrete dome 

roof. A typical HTF Type IV 

tank cross section is shown 

in Figure 2.1-51. Additional 

Type IV tank details are 

available in Section 3.0 of the 

HTF PA. [SRR-CWDA-2010-

00128] 

 

Figure 2.1-51: Typical HTF Type IV Tank Cross Section 
Basis for Section 3116 Determination DOE/SRS-WD-2014-001 

for Closure of H-Tank Farm Revision 0 

at the Savannah River Site December 2014 

Page 2-45 

 

The Type IV tank primary liner is an open-top carbon steel cylinder. The sides and bottom are formed of 

0.375-inch plates with 0.4375-inch thick knuckle plates, as shown in Figure 2.1-51. The primary waste 

tank liner is reinforced internally with three, circumferential, 4-inch carbon steel stiffener angles, as shown 

in Figure 2.1-52, and is anchored externally to the concrete wall. Each waste tank has concrete wall 

penetrations through the vault and waste tank liner located just below the dome for transfer lines into and 

out of the waste tank. [SRR-CWDA-2010-00128] 

Figure 2.1-52: Typical Type IV Tank Stiffener Angles 

Each Type IV tank is enclosed in a concrete vault. Each vault was constructed in layers using a shotcrete 

technique as shown in Figure 2.1-53. No secondary containment structure or annulus exists with this 

design. 

The concrete vault walls are cylindrical with an inside diameter of 85 feet and a height of 34 feet 3.75 

inches at the springline surmounted by a dome ring. The core wall is 7 inches at the top and 11 inches at 

the bottom. This core wall was prestressed with steel bands that remained in place and were covered 

with sprayed-on concrete. [SRR-CWDA-2010-00128] 

Type IV concrete vault sidewalls are surrounded by three layers of backfill. Bags of vermiculite were 

placed around the vault in brick-like pattern, covered with a special manually-compacted soil and topped 

with test controlled, compacted soil fill. 

The dome roof is 7 inches to 10 inches thick, with the greater thickness near the risers, and is reinforced 

throughout with steel bars. The dome has an internal curvature radius of 90-foot-4-inches and a rise of 

10-foot-7.5-inches above the springline. The Type IV tank dome and risers are shown, near the end of 

the concrete construction phase, in Figure 2.1-54. [SRR-CWDA-2010-00128] 

The concrete roof of a Type IV waste tank is not lined with carbon steel on the inside. Each waste tank 

has six peripheral risers which have an inner diameter of 2 feet and are approximately 5 feet in length. 

The waste tank riser design configuration provides limited access to the tank interior as shown in Figure 

2.1-55. [SRR-CWDA-2010-00128, W231206] 

 

Stiffener Angle 

Stiffener Angle 
Basis for Section 3116 Determination DOE/SRS-WD-2014-001 

for Closure of H-Tank Farm Revision 0 

at the Savannah River Site December 2014 

Page 2-46 

 

Figure 2.1-53: Typical Type IV Tank 

Shotcrete Application 

Figure 2.1-54: Typical Type IV Tank Domes and 

Risers During Construction 

Figure 2.1-55: Typical Type IV Tank Access Area for 

Waste Removal Equipment Diagram 
Basis for Section 3116 Determination DOE/SRS-WD-2014-001 

for Closure of H-Tank Farm Revision 0 

at the Savannah River Site December 2014 

Page 2-47 

 

2.1.12 Ancillary Structures 

In addition to the waste tanks, HTF contains ancillary structures with a residual inventory that must be 

accounted for as part of HTF closure. These ancillary structures include buried pipe (transfer lines), 

pump tanks and evaporators, all of which have been in contact with liquid waste. The ancillary structures 

are used in the HTF to transfer waste (e.g., transfer lines, pump tanks) and reduce waste volume through 

evaporation. The amount of contamination on these components depends on such factors as the service 

life of the component, its materials of construction and the contamination medium in contact with the 

component. Figure 2.1-56 identifies locations of HTF-specific ancillary structures. 

A description of the HTF ancillary structures is provided as follows: 

 the HTF transfer line system (approximately 74,800 linear feet of underground waste transfer 

lines), including transfer line jackets, leak detection boxes, modified leak detection boxes and the 

Type I tank transfer line encasements, 

 the HTF pump tanks (HPT-2 through HPT-10, CTS PT-242-3H, and CTS PT-242-18H), HTF 

pump pits (HPP-1 through HPP-10, CTS PP-242-3H, and CTS PP-242-18H) and the HTF catch 

tank, 

 the 242-H Evaporator System, including the evaporator cell and support tanks (e.g., mercury 

collection tank, cesium removal column pump tank and overheads tanks), 

 the 242-16H Evaporator System, including the evaporator cell and support tanks (e.g., mercury 

collection tank, cesium removal column pump tank and overheads tanks), 

 the 242-25H Evaporator System, including the evaporator cell and support tanks (e.g., 

condenser, mercury collection tank and overheads tanks), 

 the HTF diversion boxes (HDB-1 through HDB-8), and 

 the HTF valve boxes (valve box 15/16, Tanks 21 and 22 valve boxes, Tank 40 valve box, Tank 40 

drain valve box, Tank 42 valve box, Tank 49 valve box, Tank 50 valve box, Tank 51 valve box, 

Tank 51 drain valve box and 241-96H valve box). 

[SRR-CWDA-2010-00128] 

2.1.12.1 Waste Transfer Lines 

There are over 74,800 linear feet 

of waste transfer line in HTF, with 

the line segments ranging from a 

few feet to over 3,400 feet in 

length. The HTF waste transfer 

lines are typically constructed of 

a stainless steel primary core 

pipe and are normally located 

below ground. Those lines that 

are above ground or near the 

surface are shielded to minimize 

radiation exposure to personnel. 

Most primary waste transfer lines 

have some type of secondary 

containments. The majority of 

primary transfer lines are 

surrounded by another pipe 

(jacket) constructed of carbon 

steel, stainless steel or cementasbestos. These jackets drain to 

leak detection boxes, modified 

leak detection boxes or to another primary or secondary containment such as a waste tank. A few 

primary transfer lines are located inside a covered, concrete encasement. [SRR-CWDA-2010-00128] 

Figure 2.1-56: General Layout of HTF 
Basis for Section 3116 Determination DOE/SRS-WD-2014-001 

for Closure of H-Tank Farm Revision 0 

at the Savannah River Site December 2014 

Page 2-48 

 

Waste transfer lines are typically 

sloped to be self-draining and, 

where a pipe transitions from one 

size to another, the bottom of the 

pipe is generally aligned to allow 

for draining. The line segments are 

supported using rod or disk type 

core pipe spacers, core pipe 

supports, jacket supports, jacket 

guides or other approved methods. 

Typically, core pipe spacers and 

supports are of stainless steel and 

welded to the core pipe and jacket, 

while the jacket supports and 

guides are of stainless steel with a 

concrete support. Figure 2.1-57 

shows waste transfer lines during 

construction. [SRR-CWDA-2010-

00128] 

2.1.12.2 Pump Pits, Pump Tanks 

and Catch Tanks 

The pump pits are shielded 

reinforced concrete structures 

located below grade at the low 

points of transfer lines and are 

lined with stainless steel. The pump pit walls are approximately 2 to 3 feet thick, sloped floors are 

approximately 3 feet thick and 

cell covers are concrete slabs 

approximately 2 to 4 feet thick. 

All HTF pump pits house a pump 

tank (with the exception of HPP1) with the pump pits providing 

secondary containment for pump 

tanks. Figure 2.1-56 shows the 

location of the pump pits in HTF. 

HPP-1 through HPP-4 and HPP7 through HPP-10 are co-located 

with a diversion box. [SRRCWDA-2010-00128] 

The pump pits were often 

constructed in conjunction with a 

diversion box (diversion box 

details are discussed later in this 

section). Figure 2.1-58 is a 

photograph of the interior view of 

HPP-3. A typical diversion 

box/pump pit configuration is 

depicted in Figure 2.1-59. 

 

Figure 2.1-57: HTF Waste Tank Transfer Line During 

Construction 

Figure 2.1-58: Interior View of HPP-3 
Basis for Section 3116 Determination DOE/SRS-WD-2014-001 

for Closure of H-Tank Farm Revision 0 

at the Savannah River Site December 2014 

Page 2-49 

 

Figure 2.1-59: Typical Diversion Box and Pump Pit Design 

The following is a description of pump pit and pump tank features. 

 HPP-1 through HPP-4/HPT-2 through HPT-4: These pump pits have an interior floor that is 15 

feet square. HPT-2 through HPT-4 have an approximately 7,200-gallon capacity. There is not a 

pump tank in HPP-1. 

 HPP-5 and HPP-6/HPT-5 and HPT-6: These pump pits have an interior floor that is 18 feet x 15 

feet. HPT-5 and HPT-6 have an approximately 7,200-gallon capacity. 

 HPP-7 through HPP-10/HPT-7 through HPT-10: These pump pits have an interior floor that is 

18 feet square. HPT-7 through HPT-10 have an approximately 6,000-gallon capacity. 

 CTS PP 242-3H and CTS PP 242-18H/CTS PT 242-3H and CTS PT 242-18H: These pump pits 

have an interior floor that is 14 feet square. The CTS PT 242-3H and CTS PT 242-18H have an 

approximately 3,000-gallon capacity. CTS PP 242-3H was retired from service in 1979 and 

replaced with CTS PP 242-18H

to accommodate additional 

waste tanks. 

[SRR-CWDA-2010-00128] 

There is a single catch tank in HTF 

designed to collect drainage from HDB-1 

and the Type I tank transfer line 

encasements. The stainless steel catch 

tank capacity is approximately 11,700 

gallons and is located in an underground 

reinforced concrete cell. The catch tank 

encasement has walls that are 2 feet 8 

inches thick, a cover that is 2 feet 11 

inches thick and a floor that is 3 feet 10 

inches thick. (Figure 2.1-60). [SRRCWDA-2010-00128] 

Figure 2.1-60: Typical Catch Tank Cross Section 
Basis for Section 3116 Determination DOE/SRS-WD-2014-001 

for Closure of H-Tank Farm Revision 0 

at the Savannah River Site December 2014 

Page 2-50 

 

2.1.12.3 Evaporator Systems 

There are three evaporator systems in the HTF, the 242-H Evaporator System (1H Evaporator), 242-16H 

Evaporator System (2H Evaporator) and the 242-25H Evaporator System (3H Evaporator). Evaporators 

are used to reduce the volume of liquid radioactive waste within HTF by driving off a portion of the water 

in the waste. The evaporator systems are principally comprised of the evaporator, the overheads system 

and the condenser. [SRR-CWDA-2010-00128] 

2.1.12.3.1 242-H Evaporator System 

The 242-H Evaporator Facility was constructed and placed into service in 1963 and was removed from 

service in 1994. [HLW-2002-00025] Figure 2.1-61 provides a schematic of the 242-H Evaporator 

System. 

 242-H Evaporator 

Cells: The 

evaporator cell is a 

cuboid with a 16-

foot x 15-foot base 

and a height of 25 

feet. The cell 

includes a floor 

sump 2 feet x 2 feet 

x 2.5 inches deep. 

The cell provided 

containment for the 

evaporator and 

served as shielding 

for personnel 

protection. The cell 

includes a stainless 

steel liner. 

 242-H Evaporator 

Vessel: The 

evaporator vessel, 

located inside the 242-H evaporator cell, is a stainless steel, cylindrical vessel with a cone 

bottom. The cylindrical portion is 8 feet in diameter and the overall height is 15 feet. The 

evaporator was used to concentrate liquid by evaporating water from the waste to reduce waste 

volumes. 

 242-H Overheads System: The receiver cell is cuboid with a 15 feet x 8 feet x 10-inch base and 

a height of 16-foot-6-inches. The receiver cell includes a floor sump with a 1.5 feet x 1.5 feet 

base and a depth of 1.5 feet. The receiver cell provided containment for the two overheads 

vessels. The overheads vessels functioned as receipt tanks for liquids condensed from 

evaporator vapors via the 242-H Condenser. 

[SRR-CWDA-2010-00128]

2.1.12.3.2 242-16H Evaporator System 

The 242-16H Evaporator System was placed into service in 1982 and continues to operate. [HLW-2002-

00025] The 242-16H evaporator system is arranged into three cells and a gang valve house. The 

evaporator cell contains the evaporator, the condenser cell contains the condenser and an overheads cell 

contains overheads system components other than the condenser. Figure 2.1-62 provides a schematic 

of the 242-16H Evaporator System. 

 242-16H Evaporator Cells: The evaporator cell is a cuboid with a 16 feet x 16 feet base and a 

height of 25 feet, with walls constructed of stainless steel-lined, grooved concrete that is 3.5 feet 

thick and a roof 1 foot thick, composed of concrete slab sections with a sloped, galvanized steel 

rain cover with access points. The evaporator cell is stainless steel-lined for collecting leakage 

Figure 2.1-61: 242-H Evaporator System Schematic 
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from equipment inside the evaporator or condenser cells, leakage from the lift/lance/evaporator 

cell sump-gang valve vent header and liquid from cell spray operations. An evaporator underliner 

sump collects any leakage through the concrete or stainless steel liner. 

The condenser cell is 10 feet x 6 feet x 9 feet 8 inches and 15 feet 6 inches high with walls 

constructed of two-foot thick concrete and has a roof composed of one-foot thick concrete slab 

sections, and a 

sloped, 

galvanized steel 

rain cover with 

access ports. 

The condenser 

cell contains a 1-

foot high stainless 

steel liner pan on 

a sloped floor. 

The condenser 

cell has an 

opening to the 

evaporator cell for 

the deentrainment 

column piping and 

permits airflow to 

the evaporator 

cell. 

The overheads 

cell (which is 

open to the 

environment) is 

15 feet x 21 feet 

and 21 feet high. 

The walls are constructed of concrete and the cell contains the mercury removal tank, cesium 

removal column feed tank, two cesium removal column pumps and two overheads tanks, and two 

overheads pumps. 

 242-16H Evaporator Vessel: The evaporator vessel is 8 feet in diameter with a height of 19 

feet, and a cone-shaped bottom. The vessel is constructed of 0.5-inch stainless steel. There are 

multiple evaporator vessel service/equipment lines installed in, or penetrating, the vessel, 

including the feed inlet nozzle, steam tube bundle, warming coil, lift lines, de-entrainment column, 

lance lines and the seal pot. 

 242-16H Overheads System: The overheads system includes the condenser, mercury removal 

tank, cesium removal column feed tank, two cesium removal column pumps, two overheads 

tanks and two overhead pumps. The condenser is a vertical, single-pass, counter-flow tube and 

shell type heat exchanger located in the condenser cell. The mercury removal tank receives 

condensed overheads from the condenser. When full, the stainless steel tank overflows to the 

cesium removal column feed tank, permitting the heavier mercury to settle out and remain in the 

tank. The tank vents to the condenser cell, which vents and drains to the evaporator cell. The 

path from the evaporator vessel to the overheads tanks travels through a stainless steel cesium 

removal column feed tank. 

[SRR-CWDA-2010-00128] 

2.1.12.3.3 242-25H Evaporator System 

The 242-25H Evaporator System was placed into service in 2000 and continues to operate. [HLW-2002-

00025] The system houses the evaporator vessel, the condenser cell and condenser, and an overheads 

cell, which contains the overheads system (that includes the mercury removal tank, mercury removal 

Figure 2.1-62: 242-16H Evaporator System Schematic 
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station, two overheads tanks, and two overheads pumps). Figure 2.1-63 provides a schematic of the 

242-25H Evaporator System. 

 242-25H Evaporator Cells: The evaporator cell is a cuboid with a 27 feet 6 inches x 20 feet base 

and a height of 32 feet 9 inches with walls constructed of 3-foot 6-inch thick concrete and has a 

roof composed of 3-foot 6-inch thick concrete slab sections. The evaporator cell is stainless 

steel-lined. 

The condenser 

cell is 10 feet 9 

inches x 19 feet 

and 18 feet high 

with walls 

constructed of 

two-foot thick 

concrete and has 

a roof composed 

of two-foot thick 

concrete slabs. 

The condenser 

cell has an 

opening to the 

evaporator cell for 

the deentrainment 

column piping and 

permits airflow to 

the evaporator 

cell. 

The overheads 

cell (which is 

open to the environment) is 25 feet x 24 feet and 23 feet high. This cell contains a mercury 

removal tank, two overheads tanks, an overheads tank sample system, and two overheads 

pumps. 

 242-25H Evaporator Vessel: The evaporator vessel has a capacity of approximately 19,000 

gallons. The insulated vessel is 14 feet in diameter with a height of 26 feet 6.375 inches, and a 

cone-shaped bottom. The vessel is constructed of 0.5625-inch stainless steel and the cone is 

comprised of 0.4038-inch stainless steel. There are multiple evaporator vessel 

service/equipment lines installed in, or penetrating, the vessel, including the feed inlet nozzle, 

steam tube bundle, warming coil, lift lines, de-entrainment column, lance lines and the seal pot. 

 242-25H Overheads System: The overheads system includes the condenser, mercury removal 

tank, two overheads tanks and two overheads pumps. The condenser is a vertical, single-pass, 

counter-flow tube and shell type heat exchanger located in the condenser cell. The mercury 

removal tank receives condensed overheads from the condenser. A drain valve leads from the 

bottom of the removal tank to the mercury collection station located in the overheads receiver 

cell. The overheads are pumped to the Effluent Treatment Facility (ETF) by one of the tworecirculation pumps. The removal tank vents to the condenser cell, which vents and drains to the 

evaporator cell. 

[SRR-CWDA-2010-00128]

2.1.12.4 Diversion Boxes 

The diversion boxes are shielded, reinforced concrete structures containing transfer line nozzles to which 

jumpers are connected to direct waste transfers to the desired location. The diversion boxes are often 

constructed in conjunction with a pump pit. 

The following is a description the features associated with the HTF diversion boxes: [SRR-CWDA-2010-

00128] 

Figure 2.1-63: 242-25H Evaporator System Schematic 
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 HDB-1 is 78 feet long x 7 feet wide and 21 feet high. The walls are reinforced concrete and are a 

minimum of 1 foot 6 inches thick and taper to accommodate the two layers of concrete slabs that 

form a roof approximately 2 feet 8 inches thick. The sloped, reinforced concrete floor is a 

minimum of 2 feet 6 inches thick. 

 HDB-2 is 26 feet long x 15 feet wide that is incorporated with HPP-1 through HPP-4. The walls 

are reinforced concrete a minimum of 3 feet thick with a sloped floor of reinforced concrete 

approximately 4 feet 6 inches thick. The diversion box covers are concrete slabs and the walls 

and floor are lined with stainless steel. 

 HDB-3 is a square with outside dimensions of 6 feet 8 inches. The concrete walls and floor are 

10 inches thick and the concrete slabs that comprise the roof are 8 inches thick. 

 HDB-4 is an octagon with an outer dimension of 10 feet and inside diameter of 7 feet. It is 

comprised of reinforced concrete walls a minimum of 18 inches thick and a sloped, reinforced 

concrete floor that is approximately 2 feet 6 inches thick. The cover is a reinforced concrete plug 

7 feet 8 inches inside diameter and 3 feet thick. Stainless steel plate covers the walls, floor, and 

sump. 

 HDB-5 is an octagon with an outer dimension of 10 feet and an inside diameter of 7 feet. It is 

comprised of reinforced concrete walls that are a minimum of 18 inches thick and a sloped, 

reinforced concrete floor that is approximately 2 feet 4 inches thick. The cover is a reinforced 

concrete plug with an inside diameter of 7 feet 8 inches and a thickness of 3 feet. Stainless steel 

plate covers the walls, floor, and sump. 

 HDB-6 is a 15 foot square with walls and floor that are comprised of reinforced concrete. The 

walls are minimum 18 inches thick and the sloped floor is approximately 2 feet 11 inches thick. 

The cover for the diversion box is comprised of reinforced concrete slabs that are 3 feet thick. 

Stainless steel sheets cover the wall, floor, and sump. 

 HDB-7 is a 25-foot long x 19-foot wide rectangle with walls and floor that are comprised of 

reinforced concrete. The walls are minimum 2 feet 6 inches thick and the sloped floor is 

approximately 3 feet 4 inches thick. The cover for the diversion is comprised of reinforced 

concrete slabs that are 3 feet thick. Stainless steel covers the walls, floor, and sump. 

 HDB-8 is a 20-foot long x 24-foot wide rectangle that is incorporated with HPP-7 through HPP-10. 

The HDB-8 walls are reinforced concrete with a minimum thickness of 3 feet. The floor is 

reinforced concrete that is 3 feet thick. The diversion box cover is comprised of reinforced 

concrete slabs approximately 4 feet 3 inches thick. The walls, floor, and sump are lined with 

stainless steel. 

[SRR-CWDA-2010-00128] 

2.1.12.5 Transfer Valve Boxes 

Transfer valve boxes facilitate specific waste transfers that are conducted frequently. The valves are 

generally manual ball valves in removable jumpers with flush water connections on the transfer lines. For 

HTF valve boxes, leakage collects in the valve box and drains back to the associated waste tank, 

diversion box or leak detection box. Valve boxes are generally located adjacent to the waste tanks they 

serve. [SRR-CWDA-2010-00128] 

 Valve Box 15/16 is located near Tank 16 and is shared by Tanks 15 and 16. It contains a 

transfer line connection to HDB-2 and allows this line to be lined up to receive waste from either 

Tank 14, 15 or 16. Any liquid accumulation can be drained through the HDB-2 encasement to the 

H-Catch Tank. The total volume for this valve box is approximately 60 gallons. 

 Tanks 21 and 22 Valve Boxes contain isolation valves that connect with a transfer line to 

HDB-5. Both valve boxes drain to Tank 22. These valve boxes have an approximate volume of 

400 gallons each. 

 Tank 40 Valve Box and Tank 40 Drain Valve Box - The Tank 40 valve box is used mainly for 

transfers to DWPF. The Tank 40 drain valve box drains to a leak detection box. The Tank 40 

drain valve box was installed because low points in the transfer piping were created that could not 

be drained. The transfer valve box and drain valve box have approximate volumes of 1,020 

gallons and 350 gallons, respectively. 
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 Tank 42 Valve Box leakage drains to a leak detection box and then to the HDB-7 sump. This 

valve box has an approximate volume of 1,460 gallons. 

 Tank 49 Valve Box leakage drains to the leak detection box drain cell. This valve box has an 

approximate volume of 1,300 gallons. 

 Tank 50 Valve Box leakage drains back to Tank 50. This valve box has an approximate volume 

of 1,530 gallons. 

 Tank 51 Valve Box and Tank 51 Drain Valve Box - The Tank 51 valve box is used for transfers 

in and out of Tank 51. The Tank 51 drain valve box drains to a leak detection box. The Tank 51 

drain valve box was installed because low points in the transfer piping were created that could not 

be drained. The transfer valve box and drain valve box have approximate volumes of 1,030 

gallons and 630 gallons, respectively. 

 241-96H Valve Box allows transfers in and out of the building 241-96H monosodium titanate 

strike tanks. Its leakage drains to the leak detection box drain cell. This valve box has an 

approximate volume of 1,916 gallons. 

[WHJISB03.STGD000100] 

2.1.12.6 Other Ancillary Structures 

The leak detection boxes provide for the collection and detection of leakage from a transfer line. Drain 

piping is run from a transfer line jacket to a leak detection box. The leak detection boxes typically have 

conductivity probe leak detection and drain and overflow plugs. Drain piping for the leak detection boxes 

is provided so that liquid from the leak sites is diverted to a diversion box or pump pit. No leakage into 

transfer line secondary containment (e.g., transfer line jacket and leak detection box) due to primary line 

failure has been detected. [SRR-CWDA-2010-00128] 

The modified leak detection boxes serve the same purpose as the leak detection boxes but are larger and 

are installed at low points that cannot be gravity drained to a collection point. In addition to a conductivity 

probe, modified leak detection boxes also include a vent line to a diversion box or pump pit, an aboveground pressure gage to monitor for potential over-pressurization and a smear/cleanout pipe for 

measuring levels and manual pump-out of leakage into the box. [SRR-CWDA-2010-00128] 

2.2 H-Tank Farm Wastes 

As discussed in the following subsections, HTF has received waste from a number of different sources. 

Most of the waste managed in HTF originated in the H-Canyon Separations Facility, which primarily made 

use of the H-Modified (HM) chemical separations process. The HM process, which is similar to the 

PUREX process formerly utilized in the F-Canyon Separations Facility,28 was developed to support 

processing of enriched uranium fuels beginning in 1959. Typically, in this process, uranium was 

extracted from fuel that was irradiated in the SRS nuclear production reactors. Subsequent modifications 

made possible the recovery of Np-237 and the efficient processing of moderately enriched uranium fuels 

from off-site reactors. From its startup in 1955 until introduction of the HM process in 1959, the H-Canyon 

Separations Facility was used for processing of natural uranium fuels utilizing the PUREX process. 

Before transfer of the waste from the H-Canyon to the tank farms, sodium hydroxide was and continues 

to be added to adjust the waste to a high alkaline state to prevent corrosion of the carbon steel waste 

tanks. This chemical adjustment resulted in the precipitation of solids. These solids settled in the waste 

tanks forming a layer that is commonly referred to as “sludge.” These solids are comprised of fine 

particles of settled metal oxides including strontium, uranium and plutonium hydroxides. These solids are 

insoluble due to the chemical conditions of the solution. After settling of the solids occurred, the liquid salt 

waste solution (supernate) above this sludge layer was transferred out of the waste tanks. To maximize 

the space available in the waste tanks for storing additional waste, DOE’s practice at SRS has been to 

use the tank farm evaporator systems to reduce the volume of the decanted supernate by concentrating 

the waste. [HLW-2002-00025] 

During the evaporation process, the liquid salt waste is concentrated. After the concentrated salt waste is 

returned to the waste tank, the concentrated salt waste forms two distinct phases (collectively called salt 

 

28 The final transfer of waste from F-Canyon to the FTF occurred in August 2005. Additional information on the separations facilities 

is provided in Appendix A of this HTF 3116 Basis Document. 
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waste): 1) concentrated supernate solution and 2) solid saltcake. The predominant radionuclide present 

in the salt waste is Cs-137. Because of the high solubility of Cs-137, approximately 95 % of the Cs-137 is 

present in the concentrated supernate solution and the liquid found within the interstitial spaces in 

saltcake. The solid saltcake is composed predominantly of nitrate and nitrite salts and contains relatively 

small quantities of insoluble radioactive solids. When saltcake is dissolved and removed from the tank, 

these entrained sludge particles eventually settle on the waste tank bottom adding to the sludge 

inventory. [SRR-LWP-2010-00040] 

The solids on the waste tank bottom collectively behave as a non-ideal fluid and generally have a 

consistency similar to peanut butter. Thus, a significant amount of mixing energy is required to suspend 

the sludge in solution to transfer it out of a waste tank. 

The waste stored in HTF waste tanks is comprised of a combination of sludge, supernate and saltcake. 

2.2.1 Sources of the Waste 

This section describes the origin of the wastes managed in HTF and explains how these wastes are 

managed. 

2.2.1.1 The H-Modified Process 

This subsection briefly describes the HM process that generated most of the HTF waste. [DP-1500] 

As discussed previously, most of the waste managed in HTF originated in the H-Canyon Separations 

Facility, which made use of the HM process. The first major step in the HM process involved dissolving 

the reactor fuels in nitric acid. This dissolver solution was processed to remove silica solids and the 

clarified solution was then fed to the first solvent extraction cycle. In this cycle, uranium, neptunium and 

plutonium were separated from the fission products by being extracted into the solvent. Approximately 

95 % of the fission products remained in the aqueous phase, with the balance of the fission products 

being contained in the solvent along with the uranium, neptunium and plutonium. 

The solvent product stream underwent additional processing cycles to separate the uranium, neptunium 

and plutonium and to purify these materials. The enriched uranium was recovered and transported offsite for further processing. Neptunium was concentrated for recovery from the solution. The relatively 

small quantities of plutonium were normally discarded in the aqueous waste, unless the content of Pu-238 

was high enough to make its recovery desirable. The aqueous waste stream containing 95 % of the 

fission products, and the plutonium stream if not being recovered, was evaporated to reduce the waste 

volume and recover the nitric acid. The waste stream was then neutralized and transferred to one of the 

HTF waste tanks. 

This neutralized waste stream was known as high-heat waste. Waste from the rest of the process and 

from other operations within the facility contain the remainder of the radionuclides and was categorized as 

low-heat waste. [DP-1500] 

2.2.1.2 DWPF Recycle 

As previously described in Section 2.1.9, the DWPF returns “DWPF recycle” to the HTF. The DWPF 

recycle stream is a generally very low-activity stream that consists of condensate from chemical 

processing and melter operations, waste from decontamination activities, and waste from miscellaneous 

drains and sumps in DWPF. Before the recycle stream is transferred to HTF, it is chemically adjusted to 

a high alkaline state to prevent corrosion of the carbon steel waste tanks. DWPF recycle is currently the 

largest influent stream received by HTF. Disposition of the DWPF recycle stream is handled through 

evaporation and through beneficial reuse (e.g., salt solution molarity adjustment, salt dissolution, heel 

removal). [SRR-LWP-2009-00001] 
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2.2.1.3 Evaporator Products 

The three HTF evaporators concentrated the 

liquid waste generated in the PUREX and HM 

processes to remove excess water, a process 

which generated concentrated supernate and 

evaporator overheads (i.e., condensed 

steam). The concentrated supernate is 

transferred to a waste tank (typically referred 

to as the “evaporator concentrate receipt 

tank”) where this liquid cools. As the material 

cools salt crystals precipitate thus forming 

“saltcake,” a solid material comprised 

principally of inert salts such as nitrates and 

nitrites. Figure 2.2-1 shows the saltcake at 

the bottom of a waste tank. The remaining 

concentrated supernate is recycled back to 

the “evaporator feed tank” so that it can be processed through the evaporator again, further concentrating 

the supernate. This recycling continues through the evaporator system until no additional water can be 

readily removed from the solution. Since cesium has a higher solubility than sodium, the majority of the 

cesium remains in solution resulting in the concentrated supernate typically containing relatively high 

concentrations of Cs-137. This concentrated supernate will eventually be treated through the Liquid 

Waste System salt waste treatment processes (as described in Appendix A). Overheads are the excess 

water removed from the waste stream by the evaporation process. In the evaporator pot, the supernate 

is heated to the point that water is driven off the waste in the form of steam. This steam is passed 

through a condenser to return it to the liquid phase and sampled to ensure that there is minimal carryover 

of radioactive material (typically Cs-137). If the cesium levels are found to exceed predetermined 

thresholds, the overheads are returned to a waste tank to repeat the evaporation process. These 

thresholds are used to protect the waste acceptance criteria for the ETF where very low levels of 

radioactivity are further removed from the overheads prior to their eventual release to a permitted outfall 

at UTR.29 

2.2.1.4 Zeolite Resin 

Some waste tanks contain zeolite resin from ion exchange columns, which were also known as cesium 

removal columns. Tanks 24, 32 and 42 were equipped with these columns to remove cesium from the 

HTF evaporator overheads waste streams, with the spent zeolite containing captured Cs-137 being 

discharged into the waste tanks (zeolite is a natural mineral known for its ability to capture and retain 

cesium). The cesium removal columns were removed from service when ETF became operational. As 

discussed in the individual waste tank histories in Section 2.2.2.2, Tanks 24, 32, 38, 40, 42 and 51 

contain zeolite resin. Tank 38, 40 and 51 zeolite content is a result of waste transfer receipts from Tanks 

24, 42 and FTF. [CBU-PIT-2005-00099] 

2.2.1.5 Miscellaneous Waste Streams 

As described below, some HTF waste tanks have also received limited amounts of radioactive waste from 

other sources such as FTF, which received waste from the PUREX process in F-Canyon, Receiving 

Basin for Offsite Fuels (RBOF), Resin Regeneration Facility (RRF), In-Tank Precipitation (ITP) Project 

and Interim Salt Processing.30 

 

29 Both the Liquid Waste System salt waste treatment processes and the activities at the ETF are outside the scope of this HTF 

3116 Basis Document. 

30 As FTF waste tanks and ancillary structures are closed, waste currently stored in FTF, if not evaporated, will be transferred to 

HTF for additional processing through DWPF or salt processing (See Appendix A for additional discussion on DWPF and salt 

processing). The waste streams from the RBOF and RRF facilities are not wastes generated from the reprocessing of spent nuclear 

fuel and therefore are not, by themselves, subject to NDAA Section 3116. However, to the extent that these waste streams were 

added to waste tanks which did contain waste generated from the reprocessing of spent nuclear fuel and the waste streams became

mixed, such mixed tank wastes are within the scope of this HTF 3116 Basis Document. 

Figure 2.2-1: Saltcake at Bottom of Waste Tank 
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2.2.2 Waste Management 

This section addresses processes used for waste storage and volume reduction and provides a capsule 

history of each HTF waste tank. 

2.2.2.1 Waste Storage and Volume Reduction 

As discussed previously, the primary function of HTF has been to support the H-Canyon operation by 

storing waste produced primarily from the HM process. Management of waste stored in HTF has been 

complex due to the large volumes of waste produced by H-Canyon operations, efforts to reduce the 

volume of waste and waste removal from waste tanks. Because the HM process was based on a nitric 

acid flowsheet and because the waste tanks in HTF were constructed of carbon steel, the pH of the waste 

originating from H-Canyon must be chemically adjusted from an acidic solution to an alkaline solution 

through the addition of sodium hydroxide prior to transfer to HTF to protect the integrity of carbon steel 

waste tanks. 

The volume of waste managed in HTF has been significantly reduced using the HTF Evaporator 

Systems. The 242-H Evaporator System operated from 1963 through 1994. The 242-16H Evaporator 

Facility entered service in 1982 and continues to operate. The 242-25H Evaporator Facility entered 

service in 2000 and continues to operate. Operation of an evaporator involves the use of a waste tank to 

feed the waste to the evaporator and other waste tanks to receive the concentrated salt waste from the 

evaporation process. 

Waste removal from the waste tanks as a precursor to future closure activities results in waste being 

transferred to other waste tanks. The waste removal process includes adding large volumes of water to 

aid in dissolving saltcake and suspending sludge into slurry for transfer. These factors make it necessary 

to carefully manage waste tank space to prevent the tank farm from becoming “water logged.” 

The waste tank histories in Section 2.2.2.2 describe how individual waste tanks have been used to 

receive waste directly from H-Canyon and in waste volume reduction and waste tank cleaning efforts. 

2.2.2.2 Waste Tank History 

The following waste tank histories are based on daily and monthly data report summaries, personal logs 

and tank-to-tank transfer data, unless otherwise specifically noted. [SRR-LWP-2012-00061] Sludge and 

saltcake volumes include the interstitial liquid (i.e., liquid within the sludge and saltcake matrix) associated 

with those phases. Leak site information is included for the Type I and Type IV waste tanks known to 

have a history of leakage. No primary waste tank leakage has been detected in Type III/IIIA secondary 

liners or concrete vaults. 

Type I Tanks 

Tank 9 was constructed between 1951 and 1953 and entered service in 1955 as the first H-Canyon 

waste receipt tank. This waste tank remained active and operational until 1973. The largest volume of 

waste stored in Tank 9 has been approximately 740,000 gallons. [DPSP 58-1-7-S_p27] As of July 1, 

2014, Tank 9 contained approximately 13,600 gallons of supernate, 2,710 gallons of sludge and 534,000 

gallons of saltcake. [SRR-LWP-2014-00030] In 1955, Tank 9 received high-heat waste from the 

H-Canyon PUREX process, prior to H-Canyon conversion to the HM process. In 1965 and 1966, the 

supernate and sludge were removed from Tank 9 to allow the waste tank to serve as a concentrate 

receipt tank for the 242-H evaporator. Tank 9 supported the 242-H evaporator until 1973. Since 1973, 

the waste tank has served as a salt waste storage tank. [DPSPU 79-11-1] Tank 9 is known to have 

leaked and approximately 8 to 10 inches of salt deposits has been observed on the annulus floor. 

However, the exact location of the leak sites has not been identified. The secondary containment system 

has worked as designed; the waste has been contained within the annular pan and has not reached the 

surrounding soils. [SRR-STI-2014-00243] As shown in Table 2.1-6, Tank 9 is submerged in the water 

table. 

Tank 10 was constructed between 1951 and 1953 and entered service in 1955 as an H-Canyon waste 

receipt tank. The largest volume of waste stored in Tank 10 has been approximately 727,000 gallons. 

[DPSP 59-1-4-S_p42] As of July 1, 2014, Tank 10 contained approximately 270 gallons of supernate, 

2,710 gallons of sludge and 203,000 gallons of saltcake. [SRR-LWP-2014-00030] From 1955 through 
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1959, Tank 10 received high-heat waste from the H-Canyon PUREX process, prior to H-Canyon 

conversion to the HM process. In 1967, the supernate and majority of sludge were removed from Tank 

10 to allow the waste tank to serve as a concentrate receipt tank for the 242-H evaporator. During the 

sludge removal operation, a spill of approximately 40 to 50 gallons occurred near Riser 5 on Tank 10 

[DPSP 67-1-2-S_p39_p41]. From 1967 until 1974, Tank 10 served as a concentrate receipt tank for the 

242-H evaporator and, during this time, underwent several salt removal campaigns. The remaining 

saltcake resulting from receipts of the evaporator bottoms was left in storage in Tank 10. [DPSPU 78-11-

11] Additional salt removal campaigns were performed in Tank 10 beginning in 1979 and 1982. [DPSP 

79-21-5_p8, DPSP 82-21-12_p14, DPSP 83-21-2_p13] In 1985, water additions to Tank 10 were made 

to support tank operations and, during the late 1980s, several transfers of supernate were made from 

Tank 10. [SRR-LWP-2012-00061] In 2013, a salt removal campaign was performed in Tank 10 to 

support preparation of feed for ARP/MCU operations, approximately 100,000 gallons of dissolved salt 

solution was transferred out of the waste tank. [SRR-LWE-2013-00099] Tank 10 is known to have 

leaked and approximately 2 to 3 inches of salt deposits has been observed on the annulus floor. 

However, the exact location of the leak sites has not been identified. The secondary containment system 

has worked as designed; the waste has been contained within the annular pan and has not reached the 

surrounding soils. [SRR-STI-2014-00243] As shown in Table 2.1-6, Tank 10 is submerged in the water 

table. 

Tank 11 was constructed between 1951 and 1953 and entered service in 1955 as an H-Canyon waste 

receipt tank. The largest volume of waste stored in Tank 11 has been approximately 744,000 gallons. 

[DPSP 60-1-9-S_p38] As of July 1, 2014, Tank 11 contained approximately 116,000 gallons of supernate 

and 19,300 gallons of sludge. [SRR-LWP-2014-00030] From 1955 through 1956, Tank 11 received lowheat waste from the H-Canyon PUREX process, prior to H-Canyon conversion to the HM process. In 

1961, supernate was transferred from Tank 11 and, from 1961 through 1968, the waste tank received 

high-heat waste from the HM process and Thorex. In 1968, a sludge removal campaign was performed 

in Tank 11 to allow the waste tank to serve as concentrate receipt tank for the 242-H evaporator. After a 

short period of time as the concentrate receipt tank, Tank 11 again became a receipt tank for high-heat 

waste from the HM process and served in that role until 1982. [DPSPU 78-11-12, DPSP 82-21-1_p8] In 

2004, a sludge removal campaign was performed in Tank 11. [CBU-PIT-2004-00002] In 2008, aluminum 

rich supernate from a Low Temperature Aluminum Dissolution campaign conducted in Tank 51 was 

transferred into Tank 11 for storage. [SRR-LWP-2010-00007] In 2012, a portion of the aluminum rich 

supernate was transferred from Tank 11 to Tank 8 to be staged for future processing. [SRR-LWP-2012-

00061] Tank 11 is known to have leaked at two identified leak sites and trace amounts of waste are 

present on the walls near the leak sites and on the annulus floor. [C-ESR-G-00003] The secondary 

containment system has worked as designed; the waste has been contained within the annular pan and 

has not reached the surrounding soils. As shown in Table 2.1-6, Tank 11 is submerged in the water 

table. Tank 11 has a history of groundwater in-leakage to the annulus. The amount of in-leakage is such 

that under normal operating conditions, the waste tank annulus ventilation system is sufficient to 

evaporate any groundwater which leaks into the annulus and prevent any accumulation of groundwater 

on the floor of the annulus. 

Tank 12 was constructed between 1951 and 1953 and entered service in 1955 as an H-Canyon waste 

receipt tank. The largest volume of waste stored in Tank 12 has been approximately 736,000 gallons. 

[DPSPU 78-11-9] From 1956 through 1974, Tank 12 received high-heat waste from both the H-Canyon 

PUREX process, prior to H-Canyon conversion to the HM process, and the HM process. During this time 

period the supernate was removed five times, leaving the sludge. [DPSPU 78-11-9] From 1975 until 

2004, Tank 12 was essentially idle with only several large transfers to remove portions of the Tank 12 

supernate being performed. In 2004, water additions were made to Tank 12 in order to re-wet the dry 

sludge in preparation for waste removal. From 2009 through 2011, three mechanical sludge removal 

campaigns were conducted in Tank 12 to support sludge batch preparation for DWPF. [SRR-LWP-2012-

00061] The solids remaining in the waste tank at the conclusion of the mechanical sludge removal 

campaigns was estimated to be 13,700 gallons. A chemical cleaning campaign using Low Temperature 

Aluminum Dissolution, completed in 2011, reduced the residual solids volume to approximately 7,800 

gallons. Additional mechanical sludge removal performed in 2012, following Low Temperature Aluminum 

Dissolution, further reduced the residual solids volume to approximately 4,400 gallons. A second 

chemical cleaning campaign, using oxalic acid, performed in 2013, reduced the residual solids volume to 
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less than 2,000 gallons. [SRR-CWDA-2013-00125] Based on a preliminary volume estimate performed 

after the oxalic acid cleaning campaign, the residual solids volume is estimated to be approximately 1,400 

gallons, however, a final volume determination for Tank 12 has not been completed. [U-ESR-H-00109, 

M-CLC-H-03256] Tank 12 is known to have leaked at 15 identified leak sites and a small quantity of 

waste has been observed on the walls and on the annulus floor. [C-ESR-G-00003] The secondary 

containment system has worked as designed; the waste has been contained within the annular pan and 

has not reached the surrounding soils. As shown in Table 2.1-6, Tank 12 is submerged in the water 

table. 

Type II Tanks 

Tank 13 was constructed between 1955 and 1956 and entered service in 1956 as an H-Canyon waste 

receipt tank. The largest volume of waste stored in Tank 13 has been approximately 1,064,000 gallons. 

[SRR-CWDA-2012-00164] As of July 1, 2014, Tank 13 contained approximately 649,000 gallons of 

supernate and 38,400 gallons of sludge. [SRR-LWP-2014-00030] From 1956 until 1959, Tank 13 

received low-heat waste from the H-Canyon PUREX process. In 1961, supernate was removed from 

Tank 13 and, until early 1976, Tank 13 served as a transfer tank for high-heat supernate waste being 

transferred to the 242-16H evaporator feed tank. In addition, from 1966 through 1969, Tank 13 received 

sludge from Tanks 9, 10, 11 and 14. [DPSPU 78-11-2] From 1976 through 1994, Tank 13 served as the 

feed tank for the 242-H evaporator. [SRR-CWDA-2012-00164, WSRC-RP-94-383-2] Tank 13 remained 

idle until 2012 when a waste removal campaign was initiated in Tank 12 to support sludge batch 

preparation for DWPF. [SRR-LWP-2012-00025, SRR-LWP-2012-00061] The Tank 13 waste removal 

campaign is currently ongoing. Tank 13 is known to have leaked at two identified leak sites and trace 

amounts of waste are present on the walls near the leak sites and on the annulus floor. [C-ESR-G00003] The secondary containment system has worked as designed; the waste has been contained 

within the annular pan and has not reached the surrounding soils. As shown in Table 2.1-6, Tank 13 is 

partially submerged in the water table. 

Tank 14 was constructed between 1955 and 1956 and entered service in 1957 as an H-Canyon waste 

receipt tank. The largest volume of waste stored in Tank 14 has been approximately 1,061,000 gallons. 

[DPSPU 77-11-19] As of July 1, 2014, Tank 14 contained approximately 28,000 gallons of sludge and 

130,000 gallons of saltcake. [SRR-LWP-2014-00030] From 1957 through 1959, Tank 14 received highheat waste from the H-Canyon PUREX process until waste leakage into the Tank 14 annulus was 

detected and fresh high-heat waste receipts were diverted to another waste tank. From 1959 through 

1965, Tank 14 was used sparingly to receive high-heat waste from the HM process and high-heat waste 

that had leaked into the Tank 16 annulus. In 1968, supernate was removed from Tank 14 in preparation 

for a sludge removal campaign and, in 1969, a sludge removal campaign was initiated. From 1969 

through 1970, Tank 14 received concentrated supernate from several waste tanks and, in 1972, 14,000 

gallons were siphoned from the primary tank into the Tank 14 annulus via the annulus jet, the waste was 

immediately returned to the primary tank. [DPSPU 77-11-19] Tank 14 remained idle until 1977 when 

waste from the Tank 16 annulus was transferred into Tank 14. [SRR-CWDA-2011-00126] Since 1977, 

the waste tank has remained idle. Tank 14 is known to have leaked at 33 identified leak sites, and it is 

estimated that there are about 50 leak sites in this waste tank. Approximately 12 to 13 inches of salt 

deposits has been observed on the annulus floor. [SRR-STI-2014-00243, C-ESR-G-00003] The 

secondary containment system has worked as designed; the waste has been contained within the 

annular pan and has not reached the surrounding soils. As shown in Table 2.1-6, Tank 14 is partially 

submerged in the water table.

Tank 15 was constructed between 1955 and 1956 and entered service in 1960 as a receipt tank for 

supernate from Tank 16. The largest volume of waste stored in Tank 15 has been approximately 

1,075,000 gallons. [SRR-CWDA-2012-00164] As of July 1, 2014, Tank 15 contained approximately 

159,000 gallons of sludge and 76,200 gallons of saltcake. [SRR-LWP-2014-00030] In 1960, in response 

to leakage in Tank 16, Tank 15 was placed into service to initially receive supernate from Tank 16. From 

1960 through 1978, Tank 15 received high-heat waste from the HM process and, during this time, 

supernate was periodically removed from Tank 15 to allow continued receipt of waste from H-Canyon. 

[DPSPU 77-11-26] From 1978 to 1980, Tank 15 received sludge slurry transfers from Tank 16. [SRRCWDA-2011-00126] In 1982, a sludge removal campaign was performed in Tank 15. Since 1982, the 

waste tank has remained idle. [DPSP 82-21-3_p13] Tank 15 is known to have leaked at 20 identified 
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leak sites and trace amounts of waste are present on the walls near the leak sites and on the annulus 

floor. In 1973, 12 new annulus risers were installed to provide nearly 100 % waste tank wall area 

surveillance capability. [DPSPU 77-11-26, C-ESR-G-00003] The secondary containment system has 

worked as designed; the waste has been contained within the annular pan and has not reached the 

surrounding soils. As shown in Table 2.1-6, Tank 15 is partially submerged in the water table. 

Tank 16 was constructed between 1955 and 1956 and entered service in 1959 as an H-Canyon waste 

receipt tank. The largest volume of waste stored in Tank 16 has been approximately 1,061,000 gallons. 

[DPSPU 77-11-17] From 1959 through 1960, Tank 16 received high-heat waste from the HM process 

and, in 1960, due to leakage in the primary tank, approximately one-half of the contents of Tank 16 were 

transferred to Tanks 14 and 15. From 1967 through 1972, Tank 16 received low-heat waste from the HM 

process and concentrate from the 242-H evaporator, however, in 1972 leakage of the Tank 16 primary 

tank resumed and the supernate in Tank 16 was transferred to Tank 13. The sludge remained in the 

waste tank and the use of Tank 16 for any additional waste receipts was discontinued. [DPSPU 77-11-

17] In 1978, an annulus cleaning campaign was performed for Tank 16. Waste removal from the Tank 

16 primary tank was performed from 1978 through 1980. [SRR-CWDA-2011-00126] Tank 16 is known to 

have leaked at approximately 300 to 350 identified leak sites. Waste leakage into the annulus was first 

discovered in 1959, and in September 1960, liquid waste overflowed the annulus pan. A few tens of 

gallons of waste are estimated to have breached the waste tank concrete encasement and entered the 

surrounding soil. Leakage essentially stopped after the waste tank liquid level was lowered below the 

middle horizontal weld. In 1962 and 1974, sandblasting was used to support detailed inspections of the 

leak sites. Metallurgical examination indicated the cause of the cracks was nitrate-induced stress 

corrosion. [DPSPU 77-11-17, DP-1358] Based on the final volume determination for Tank 16, there are 

330 gallons of residual solids remaining in the Tank 16 primary tank and 1,910 gallons remaining in the 

Tank 16 annulus. [U-ESR-H-00113] As shown in Table 2.1-6, Tank 16 is partially submerged in the 

water table. 

Type IV Tanks 

Tank 21 was constructed in 1961 and entered service in 1961 as a receipt tank for supernate from Tank 

13. The largest volume of waste stored in Tank 21 has been approximately 1,345,000 gallons. [DPSP 

62-1-6-S_p42] As of July 1, 2014, Tank 21 contained approximately 424,000 gallons of supernate, 

46,400 gallons of sludge and.3,540 gallons of saltcake. [SRR-LWP-2014-00030] In 1961, Tank 21 

received low-heat waste transferred from Tank 13 and, subsequently, received additional low-heat waste 

from Tank 11. From 1963 through 1976, Tank 21 served as the feed tank for the 242-H evaporator and, 

during this time, stored both low-heat and high-heat waste. [DPSPU 78-11-10] From 1976 through 1986, 

Tank 21 received waste from the HM process and, in 1986, a sludge removal campaign was performed in 

the waste tank. [MO-RPT-86-SEP2_p20] In addition, from 1963 to 1994, Tank 21 received contaminated 

water from building 244-H, RBOF, and 245-H, RRF. From 2001 to 2005, Tank 21 served as a receipt 

tank for DWPF recycle. [CBU-SPT-2003-00181] Since 2005, Tank 21 has been utilized to assemble salt 

batches in preparation for treatment via ARP/MCU Interim Salt Processing.31 Visual examinations of the 

steel liner have shown no evidence of failure, significant surface corrosion or other anomalies. [SRR-STI2014-00243] 

Tank 22 was constructed between 1961 and 1962 and entered service in 1965 as a receipt tank for 

supernate from Tank 21. The largest volume of waste stored in Tank 22 has been approximately 

1,338,000 gallons. [DPSP 65-1-6-S_p53] As of July 1, 2014, Tank 22 contained approximately 666,000 

gallons of supernate and 110,000 gallons of sludge. [SRR-LWP-2014-00030] In 1965, Tank 22 received 

high-heat waste transferred from Tank 21 and, subsequently, the material was transferred back to Tank 

21, processed through the 242-H evaporator and returned to Tank 22. The recycled concentrate from 

Tank 21 and from other waste tanks was stored in the waste tank until 1971. From 1971 through 1974, 

supernate was removed from Tank 22 and a salt removal campaign was performed. [DPSPU 79-11-5] 

From 1974 until 1980, Tank 22 received low-heat waste from the HM process. In 1986, two sludge 

removal campaigns were completed in Tank 22 to provide feedstock for DWPF. In addition, from 1971 to 

1989, Tank 22 received waste from building 244-H, RBOF, and 245-H, RRF via Tank 23. [SRR-LWP2012-00061] Since 2000, Tank 22 has served as a receipt tank for DWPF recycle. [CBU-SPT-2003-

 

31 A description of Interim Salt Processing is provided in Appendix A of this HTF 3116 Basis Document. 
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00181] Visual examinations of the steel liner have shown no evidence of failure, significant surface 

corrosion or other anomalies. [SRR-STI-2014-00243] 

Tank 23 was constructed between 1961 and 1962 and entered service in 1964 as a receipt tank for 

contaminated water from RBOF and RRF. The largest volume of waste stored in Tank 23 has been 

approximately 1,319,000 gallons. [MO-RPT-87-AUG3_p5] As of July 1, 2014, Tank 23 contained 

approximately 307,000 gallons of supernate and 90,600 gallons of sludge. [SRR-LWP-2014-00030] In 

1964, Tank 23 began receiving contaminated water from building 244-H, RBOF, and 245-H, RRF. Waste 

from RBOF and RRF typically contained 1 to 2 % solids and 0.1 to 1.0 % of the amount of radioactivity 

contained in typical low-heat waste. [DPSPU 79-11-7] The waste tank remained in this service until 1995 

when the RBOF facility was shutdown. The Tank 23 waste was processed initially by the 242-H 

evaporator; but beginning in 1966 the waste was processed through a zeolite bed to remove Cs-137 and 

small amounts of other radioisotopes, and discarded to seepage basins. [DPSPU 79-11-7] This practice 

was discontinued in 1988 when the Tank 23 material began being processed at ETF. In 2008, the 

material in Tank 23 was used to adjust salt solution from Tank 41 salt dissolution for processing at SDF 

as part of Deliquification, Dissolution and Adjustment (DDA) salt processing. [LWO-LWP-2008-00007] 

Since 2008, Tank 23 has been used to assemble salt batches in preparation for treatment via ARP/MCU 

Interim Salt Processing.32 Visual examinations of the steel liner have revealed corrosion but no evidence 

of failure. [SRR-STI-2014-00243] 

Tank 24 was constructed between 1961 and 1962 and entered service in 1963 as a concentrate receipt 

tank for the 242-H evaporator. The largest volume of waste stored in Tank 24 was approximately 

1,326,000 gallons. [DPSP 68-1-5_p106] As of July 1, 2014, Tank 24 contained approximately 1,240,000 

gallons of supernate and 3,540 gallons of sludge. [SRR-LWP-2014-00030] From 1963 through 1967, 

Tank 24 served as a concentrate receipt tank for the 242-H evaporator. Tank 24 remained mostly idle 

until 1981 when it began to receive waste, via Tank 23, from building 244-H, RBOF, and 245-H, RRF. 

The waste tank remained in this service until 1987. [SRR-LWP-2012-00061] In addition, in 1965, Tank 

24 was equipped with a cesium removal column used to remove cesium from the evaporator overheads 

and the waste streams from RBOF/RRF. From 1966 through 1984, Tank 24 received approximately 

44,000 gallons of zeolite. In 1982, a salt removal campaign was performed in Tank 24 and, in 1983, 

additional salt solution was removed from the waste tank. Also in 1983, a zeolite slurry transfer was sent 

from Tank 24 to Tank 38. In 1985, an oxalic acid demonstration for zeolite removal was performed in 

Tank 24 and it was estimated that approximately 5,500 gallons of zeolite remained in the waste tank. 

[CBU-PIT-2005-00099] In 2005, Tank 24 was used to assemble salt batches in preparation for treatment 

via ARP/MCU Interim Salt Processing. [CBU-PIT-2005-00269] Since 2010, Tank 24 has been used as a 

concentrated supernate storage tank. [SRR-LWP-2011-00001] Visual examinations of the steel liner 

have shown no evidence of failure, significant surface corrosion or other anomalies. [SRR-STI-2014-

00243] 

Type III Tanks 

No primary waste tank leakage has been detected in Type III secondary liners or concrete vaults. [SRRSTI-2014-00243] 

Tank 29 was constructed between 1967 and 1970 and entered service in 1971 as a concentrate receipt 

tank for the 242-H evaporator. The largest volume of waste stored in Tank 29 has been approximately 

1,295,000 gallons. [SRR-CWDA-2012-00164] As of July 1, 2014, Tank 29 contained approximately 

237,000 gallons of supernate and 1,020,000 gallons of saltcake. [SRR-LWP-2014-00030] From 1971 

through 1988, Tank 29 served as a concentrate receipt tank for the 242-H evaporator and, in 1980, also 

received concentrated supernate from Tank 10. [DPSP 80-21-5_p23] After 1988, three transfers were 

made from Tank 29 to Tank 32 and only one transfer, from Tank 42, was received. [SRR-LWP-2012-

00061] 

Tank 30 was constructed between 1967 and 1970 and entered service in 1974 as a concentrate receipt 

tank for the 242-H evaporator. The largest volume of waste stored in Tank 30 has been approximately 

1,278,000 gallons. [SRR-LWP-2012-00025] As of July 1, 2014, Tank 30 contained approximately 

841,000 gallons of supernate, 620 gallons of sludge and 329,000 gallons of saltcake. [SRR-LWP-2014-

 

32 A description of Interim Salt Processing is provided in Appendix A of this HTF 3116 Basis Document. 
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at the Savannah River Site December 2014 

Page 2-62 

 

00030] From 1974 through 1995, Tank 30 served as a concentrate receipt tank for the 242-H evaporator. 

Tank 30 was converted to a concentrate receipt tank for the 242-25H evaporator in 2000. [SRR-LWP2012-00061] 

Tank 31 was constructed between 1967 and 1970 and entered service in 1973 as a concentrate receipt 

tank for the 242-H evaporator. The largest volume of waste stored in Tank 31 has been approximately 

1,288,000 gallons. [SRR-CWDA-2012-00164] As of July 1, 2014, Tank 31 contained approximately 

100,000 gallons of supernate and 1,150,000 gallons of saltcake. [SRR-LWP-2014-00030] From 1973 

through 1983, Tank 31 served as a concentrate receipt tank for the 242-H evaporator. Since then, Tank 

31 has not received a waste transfer. [SRR-LWP-2012-00061] 

Tank 32 was constructed between 1967 and 1970 and entered service in 1971 as an H-Canyon waste 

receipt tank. The largest volume of waste stored in Tank 32 has been approximately 1,277,000 gallons. 

[SRR-CWDA-2012-00164] As of July 1, 2014, Tank 32 contained approximately 706,000 gallons of 

supernate, 104,000 gallons of sludge and 186,000 gallons of saltcake. [SRR-LWP-2014-00030] From 

1971 through 1988, Tank 32 received high-heat waste from the HM process. In addition, Tank 32 was 

equipped with a cesium removal column used to remove cesium from the evaporator overheads and, 

from 1985 to 1988, received approximately 6,700 gallons of spent zeolite. [CBU-PIT-2005-00099] Tank 

32 was converted to a feed tank for the 242-25H evaporator in 2000. [SRR-LWP-2012-00061] 

Type IIIA Tanks 

No primary waste tank leakage has been detected in Type IIIA secondary liners or concrete vaults. 

[SRR-STI-2014-00243] 

Tank 35 was constructed between 1974 and 1977 and entered service in 1977 as an H-Canyon waste 

receipt tank. The largest volume of waste stored in Tank 35 has been approximately 1,274,000 gallons. 

[WSRC-RP-92-78-7B] As of July 1, 2014, Tank 35 contained approximately 1,060,000 gallons of 

supernate and 89,200 gallons of sludge. [SRR-LWP-2014-00030] From 1977 through 1990, Tank 35 

received high-heat waste from the HM process. After 1990, Tank 35 received salt solution from Tank 37 

saltcake removal campaigns in 2002, 2005 and 2010. [SRR-LWP-2012-00061] 

Tank 36 was constructed between 1974 and 1977 and entered service in 1977 as concentrate receipt 

tank for the 242-H evaporator. The largest volume of waste stored in Tank 36 has been approximately 

1,285,000 gallons. [SRR-LWP-2012-00061] As of July 1, 2014, Tank 36 contained approximately 

124,000 gallons of supernate, 186 gallons of sludge and 1,150,000 gallons of saltcake. [SRR-LWP-2014-

00030] From 1977 through 1988, Tank 36 served as a concentrate receipt tank for the 242-H evaporator. 

Since then, Tank 36 has not received a waste transfer. [SRR-LWP-2012-00061] 

Tank 37 was constructed between 1974 and 1977 and entered service in 1978 as a concentrate receipt 

tank for the 242-H evaporator. The largest volume of waste stored in Tank 37 has been approximately 

1,288,000 gallons. [SRR-LWP-2010-00007] As of July 1, 2014, Tank 37 contained approximately 

144,000 gallons of supernate and 1,110,000 gallons of saltcake. [SRR-LWP-2014-00030] From 1978 

through 1988, Tank 37 served as a concentrate receipt tank for the 242-H evaporator. In 2002, saltcake 

removal operations were performed in Tank 37 to allow the tank to serve as a concentrate receipt tank for 

the 242-25H evaporator. Additional saltcake removal campaigns were carried out in 2005 and 2010 

allowing Tank 37 to continue serving as a concentrate receipt tank for the 242-25H evaporator. [SRRLWP-2012-00061] 

Tank 38 was constructed between 1976 and 1980 and entered service in 1981 as a receipt tank for a 

variety of waste streams in support of the H-Area Liquid Waste program. The largest volume of waste 

stored in Tank 38 has been approximately 1,263,000 gallons. [SRR-LWP-2012-00061] As of July 1, 

2014, Tank 38 contained approximately 268,000 gallons of supernate and 813,000 gallons of saltcake. 

[SRR-LWP-2014-00030] From 1981 through 1986, Tank 36 received a variety of waste streams in 

support of the H-Area Liquid Waste program, including, in 1983 and 1984, zeolite slurry from Tank 24. 

Tank 38 received a total of approximately 38,000 gallons of zeolite from Tank 24. [CBU-PIT-2005-00099] 

Tank 38 was converted to a concentrate receipt tank for the 242-16H evaporator in 1986. [SRR-LWP2012-00061] 

Tank 39 was constructed between 1976 and 1980 and entered service in 1982 as an H-Canyon waste 

receipt tank. The largest volume of waste stored in Tank 39 has been approximately 1,280,000 gallons. 
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[WSRC-RP-89-78-9B_p4] As of July 1, 2014, Tank 39 contained approximately 769,000 gallons of 

supernate and 179,000 gallons of sludge. [SRR-LWP-2014-00030] From 1982 through 1994, Tank 39 

received high-heat waste from the HM process and, in 1987, served as a concentrate receipt tank for the 

242-16H evaporator. After 1997, Tank 39 served as a receipt tank for H-Canyon waste, waste from the 

299-H maintenance facility and waste from other Liquid Waste program initiatives. Transfers out of Tank 

39 since 1996 have primarily been to the SRS FTF evaporator system. [SRR-LWP-2012-00061] 

Tank 40 was constructed between 1976 and 1980 and entered service in 1986 as a receipt tank for 

sludge slurry transfers from Tanks 18 and 22. The largest volume of waste stored in Tank 40 has been 

approximately 1,252,000 gallons. [MO-RPT-87-MAR3_p4] As of July 1, 2014, Tank 40 contained 

approximately 623,000 gallons of sludge. [SRR-LWP-2014-00030] From 1986 through 2001, Tank 40 

served as a receipt tank for sludge slurry transfers from Tanks 8, 18, 22 and 42 and, in 2001, the contents 

of Tank 40 were prepared as the second sludge batch for DWPF. Tank 40 began sending sludge to 

DWPF in 2002 and has been serving as the DWPF feed tank since that time. [SRR-LWP-2012-00061] 

Tank 40 has received zeolite material from both Tanks 42 and 51. [CBU-PIT-2005-00099] 

Tank 41 was constructed between 1976 and 1980 and entered service in 1981 as a receipt tank for 

supernate from Tank 38. The largest volume of waste stored in Tank 41 has been approximately 

1,281,000 gallons. [WSRC-RP-91-78-8B] As of July 1, 2014, Tank 41 contained approximately 522,000 

gallons of supernate, 17,300 gallons of sludge and 469,000 gallons of saltcake. [SRR-LWP-2014-00030] 

In 1981, Tank 41 received concentrated supernate from Tank 38. From 1982 through 1987, Tank 41 

served as a concentrate receipt tank for the 242-16H evaporator. In 2002, salt removal efforts in Tank 41 

were initiated to prepare low-level waste feed for the Saltstone Production Facility (SPF). Tank 41 

supernate was transferred to Tank 49 and, interstitial liquid was removed from Tank 41 and transferred to 

Tanks 49 and 39. Salt dissolution activities were performed in Tank 41 in 2005 and again in 2008. The 

salt solution from the Tank 41 saltcake removal campaigns was eventually transferred to the SPF as lowlevel waste having been treated by the DDA process. In 2008, Tank 41 began receiving salt solution from 

salt removal activities in Tank 25 in support of ARP/MCU Interim Salt Processing.33 [SRR-LWP-2012-

00061] 

Tank 42 was constructed between 1976 and 1980 and entered service in 1982 primarily as a receipt tank 

for sludge slurry from sludge removal campaigns in Tanks 15, 24, 18 and 21. The largest volume of 

waste stored in Tank 42 has been approximately 1,278,000 gallons. [LWO-PIT-2006-00063] As of July 

1, 2014, Tank 42 contained approximately 548,000 gallons of supernate and 17,600 gallons of sludge. 

[SRR-LWP-2014-00030] From 1982 through 1986, Tank 42 served as a receipt tank for sludge slurry 

from sludge removal campaigns in Tanks 15, 24, 18 and 21. In addition, Tank 42 was equipped with a 

cesium removal column used to remove cesium from the evaporator overheads and, from 1982 to 1990, 

received approximately 5,100 gallons of spent zeolite. [CBU-PIT-2005-00099] During this time period, 

and continuing until 1993, supernate was transferred out of Tank 42 to several waste tanks. From 1996 

to 1999, Tank 42 supported the sludge batch preparation process by sending sludge slurry to Tank 51. 

Tank 42 was mostly idle until it was used to support various Liquid Waste program initiatives and the 242-

25H evaporator from 2004 through 2006. In 2011, Tank 42 began to support sludge batch preparation 

and other waste removal initiatives. [SRR-LWP-2012-00061] 

Tank 43 was constructed between 1976 and 1980 and entered service in 1982 as the 242-16H 

evaporator feed tank. The largest volume of waste stored in Tank 43 has been approximately 1,260,000 

gallons. [CBU-SPT-2003-00181] As of July 1, 2014, Tank 43 contained approximately 913,000 gallons 

of supernate and 238,000 gallons of sludge. [SRR-LWP-2014-00030] Tank 43 entered service in 1982 

as the 242-16H evaporator feed tank, and it continues to serve in that capacity today. From 1982 until 

the startup of DWPF in 1996, Tank 43 received material from various waste tanks, the 299-H 

maintenance facility and low-heat waste from the HM process. From 1996 until the present time, the 242-

16H evaporator system (Tanks 43 and 38) has primarily supported the DWPF facility by volume reducing 

the DWPF recycle stream sent to HTF. Initially the DWPF recycle stream was sent directly to Tank 43; 

however, due to complications with processing this material in the 242-16H evaporator, the DWPF recycle 

stream was redirected to Tanks 21 and 22 beginning in 1999 to allow for solids settling to occur. 

 

33 A description of Interim Salt Processing is provided in Appendix A of this HTF 3116 Basis Document. 
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Transfers from Tanks 21 and 22 to Tank 43 are periodically made to support the 242-16H evaporator and 

DWPF. [SRR-LWP-2012-00061] 

Tank 48 was constructed between 1978 and 1981 and entered service in 1983 as part of the ITP project 

demonstration. The largest volume of waste stored in Tank 48 has been approximately 483,000 gallons. 

[WSRC-RP-95-841-6] As of July 1, 2014, Tank 48 contained approximately 255,000 gallons of 

supernate. [SRR-LWP-2014-00030] In 1983, cold chemicals were added to Tank 48 prior to receiving a 

transfer of salt solution from Tank 24. The initial batch of material in Tank 48 formed the base to conduct 

the ITP project demonstration. During the demonstration, a portion of the Tank 48 waste was treated via 

the ITP process and transferred to Tank 49. The ITP project ultimately was discontinued. Due to the 

residual organics that remained in Tank 48 after the demonstration project, Tank 48 has strict restrictions 

on what additions can be made to the waste tank. Tank 48 remains in this status at the present time 

while processes are being developed to disposition the contents of Tank 48. [SRR-LWP-2012-00061] 

Tank 49 was constructed between 1978 and 1981 and entered service in 1983 as part of the ITP project 

demonstration. The largest volume of waste stored in Tank 49 has been approximately 1,250,000 

gallons. [SRR-LWP-2011-00001] As of July 1, 2014, Tank 49 contained approximately 1,250,000 

gallons of supernate and 7,020 gallons of saltcake. [SRR-LWP-2014-00030] In 1983, Tank 49 received 

two transfers of ITP washed precipitant during the ITP project demonstration. The ITP project ultimately 

was discontinued. Due to the residual organics that remained in Tank 49 after the demonstration project, 

Tank 49 had strict restrictions on what additions could be made to the waste tank. Tank 49 remained idle 

until 2001 when the residual waste from the ITP project demonstration was dispositioned. In 2005, Tank 

49 received salt solution from salt dissolution activities performed in Tank 41. The salt solution from the 

Tank 41 saltcake removal campaigns was eventually adjusted with other tank farm material and 

transferred, via tank 50, to the SPF in 2007 and 2008 as low-level waste after having been treated by the 

DDA process. In 2009, Tank 49 became the batch feed tank for ARP/MCU Interim Salt Processing.34 

[SRR-LWP-2012-00061] 

Tank 50 was constructed between 1978 and 1981 and entered service in 1983 as part of the ITP project 

demonstration. The largest volume of waste stored in Tank 50 has been approximately 1,232,000 

gallons. [SRR-LWP-2012-00025] As of July 1, 2014, Tank 50 contained approximately 561,000 gallons 

of supernate. [SRR-LWP-2014-00030] In 1983, Tank 50 was placed into service as the feed tank for the 

SPF as part of the ITP project demonstration and, received approximately 518,000 gallons of ITP 

decontaminated salt solution during the demonstration. The ITP project ultimately was discontinued. In 

1988, Tank 50 began receiving ETF evaporator bottoms for final disposition to the Saltstone Disposal 

Facility (SDF) and, continues to serve as the feed tank for low-level waste streams being transferred to 

the SPF for final disposition. In addition to the ETF low-level waste stream, Tank 50 has processed other 

low-level waste streams to SDF such as ITP project demonstration material from Tank 49, DDA material 

originating from Tank 41, low-level waste streams from H-Canyon and decontaminated salt solution from 

ARP/MCU Interim Salt Processing. [SRR-LWP-2012-00061] 

Tank 51 was constructed between 1978 and 1981 and entered service in 1986 as a receipt tank for 

sludge slurry from sludge removal campaigns in Tanks 18, 21 and 22. The largest volume of waste 

stored in Tank 51 has been approximately 1,257,000 gallons. [MO-RPT-87-FEB2_p4] As of July 1, 

2014, Tank 51 contained approximately 799,000 gallons of supernate and 250,000 gallons of sludge. 

[SRR-LWP-2014-00030] From 1986 through 1993, Tank 51 served as a receipt tank for sludge slurry 

from sludge removal campaigns in Tanks 18, 21 and 22 and, in addition, received transfers from Tanks 

40 and 42. From 1996, when DWPF started up, until 2001, Tank 51 served as the feed tank for DWPF, 

the DWPF feed tank role was switched to Tank 40 in 2001. Tank 51 began to collect sludge slurry 

transfers from a variety of waste tanks to support future sludge batch preparation and remains in this role 

presently. [SRR-LWP-2012-00061] Tank 51 has received zeolite from Tank 42 and FTF Tank 7. [CBUPIT-2005-00099] 

 

34 A description of Interim Salt Processing is provided in Appendix A of this HTF 3116 Basis Document. 
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2.3 Waste Removal Approach for Waste Tanks, Annuli and Ancillary Structures 

The closure process for the tank systems begins with removing the waste using mechanical, chemical, 

and/or vacuum waste removal techniques, or other methods of comparable or greater effectiveness, 

discussed below. 

Bulk waste removal is the first step toward waste tank closure and typically employs agitation/mixer 

pumps to suspend solids and potentially dissolve soluble material. If the tank contains saltcake, then well 

water, DWPF recycle or chemically treated water is added to dissolve the waste which results in a 

considerable impact to the tank farm waste storage space. In order to make the salt mobile for 

transferring and to adjust the salt concentrations (molarity) for processing and disposal, significant 

quantities of dissolution liquid must be added. [DOE-WD-2005-001] Bulk sludge and/or salt waste is 

transferred from the tank, leaving behind a heel. 

Mechanical heel removal employs techniques such as 

agitation, spraying, lancing, pulse jet mixing, vacuum 

retrieval, mechanical manipulators, robotic devices or recycle 

systems to augment existing waste removal equipment to 

reduce the heel volume. These technologies are described in 

Section 2.3.2. 

Chemical heel removal employs oxalic acid or other 

specialized chemical treatment of the heel to dissolve solids. 

The oxalic acid may be sprayed into the tank to clean 

contaminants from the internal tank surfaces (e.g., walls, 

cooling coils, support columns, equipment), as practical. 

If necessary, at the conclusion of chemical heel removal, the 

interior of the waste tank may be washed with water to rinse 

oxalic acid from internal surfaces and dislodge loose 

contamination. The wash water will then be removed. 

For waste tanks that have leaked waste from primary to 

secondary containment, waste may be removed as 

applicable from the annulus as described in Section 2.3.6. 

Waste tank, annulus and ancillary structure cleaning is 

subject to a variety of operating constraints including: 

 maintaining emergency tank space, 

 meeting safety basis requirements, 

 controlling tank chemistry and radionuclide inventory, 

 requirements to remove waste from tanks with a 

leakage history and tanks that do not meet secondary 

containment and leak detection requirements, and 

 preparing waste for downstream waste treatment 

facilities. 

The complex interdependency of safety and process 

requirements of the various Liquid Waste facilities drive the 

sequencing of the tanks undergoing waste removal and tank 

cleaning. Plans for bulk waste removal, mechanical, 

chemical and vacuum cleaning and tank closure are 

summarized in the Liquid Waste System Plan. [SRR-LWP2009-00001] See Section 1.3 for discussion on the closure 

schedule. 

2.3.1 Tank Waste Removal History 

Early waste removal efforts at SRS (1960s) employed hydraulic mining and sluicing techniques using 

once-through water at high pressure. The practice was discontinued because insufficient tank storage 

Figure 2.3-1: Typical Standard 

Slurry Pump 

[WSRC-TR-2001-00313]
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space was available to accommodate the large volume of water added to the tank farm system. The 

technique was modified to use existing waste supernate as the slurry liquid media and slurry pumps for 

breaking up and suspending the sludge. Using this technique, several slurry pumps are installed in the 

waste tank being cleaned in lieu of the external pumps formerly used. This change allows the slurrying 

operation to be repeated as often as necessary to suspend the sludge without adding significant new 

waste volume to the tank farm. Figure 2.3-1 presents a typical standard slurry pump design. A slurry 

pump is a vertical shafted centrifugal pump with the drive motor mounted topside. A coupled shaft 

connects the motor and pump. Suction is drawn into the pump and discharged from two nozzles (aimed 

in opposite directions from each side of the pump). The nozzles are shaped such that high velocity jets 

are ejected into the liquid. The pump rotates on a turntable, thereby allowing the jets to spin in the 

horizontal plane if desired. The pumps are typically installed in available risers such that the circular 

pattern of suspended solids, or effective cleaning radius, of each individual pump overlaps with the 

adjacent pump to maximize effectiveness. The initial elevation of the pump suction is typically positioned 

just above the sludge layer. Water may be added to the tank if there is not enough supernate to use as 

the slurry media. The pumps typically suspend sludge that can be suspended (at that slurry pump 

elevation setting) within a few days. The slurry pumps are then lowered typically in 10- to 17-inch 

increments, more water is added, if needed, and the next layer of sludge is suspended. This process is 

repeated until the slurry pumps are at the lowest elevation practical, typically 10 inches above the waste 

tank floor. The transfer pump is then lowered to the desired elevation, typically 6 inches above the tank 

floor. Interferences in the waste tanks such as “field to fit” horizontal cooling coils may limit the depth to 

which individual pumps may be lowered.35 The sludge slurry is then transferred out of the tank. To 

obtain the proper weight percent of suspended solids in the resulting sludge batch, more than one 

transfer may be required. Examples of HTF tanks in which this technique has been successfully used for 

bulk sludge removal include Tanks 11, 12, 13, 15, 16, 21, 22 and 42. [HLW-2002-00025] 

Slurry pumps have also been used for bulk waste removal in tanks containing saltcake. To remove 

saltcake, the pumps are positioned just above the saltcake and water is added to the tank. The water is 

stirred by the pumps to dissolve the top layer of saltcake. Once the resulting solution becomes nearly 

saturated with dissolved salt it is transferred out of the tank. The slurry pumps are then lowered, water is 

added and the process is repeated. In HTF, this technique has been successfully used for bulk salt 

removal in Tank 24. [HLW-2002-00025] 

Saltcake can also be dissolved without agitation. To dissolve the saltcake and create salt solution 

batches, the waste tank is filled with dissolution liquid, which, if necessary, is water chemically treated to 

prevent corrosion of the carbon steel waste tanks, until the saltcake surface is flooded. The dissolution 

liquid dissolves a portion of the saltcake forming a salt solution. The salt solution is then transferred out 

of the tank. In HTF, this process has been used for saltcake removal in Tanks 10, 22, 37 and 41. [HLW2002-00025] 

Processing the salt waste requires significant available waste tank space. To make the salt mobile for 

transferring and to adjust the salt concentrations (molarity) for processing and disposal, significant 

quantities of dissolution liquid must be added. Type III/IIIA tank space must be available for receiving and 

adjusting these solutions. 

2.3.2 Waste Removal Technologies 

Slurry pumps previously installed in HTF waste tanks may continue to be used in waste removal activities. 

For example, HTF Tank 12 underwent bulk waste removal efforts utilizing standard slurry pumps. The 

slurry pumps installed in Tank 12 were then used to support heel removal activities performed within the 

waste tank. [SRR-LWE-2012-00059] In addition, a new generation of waste removal equipment has 

been developed and are available as described in the following subsections. 

2.3.2.1 Submersible Mixer Pumps - Type I, II and III and IIIA Tanks 

In 2003, SRS used a systematic process to identify, evaluate and select equipment for waste removal 

tasks to accelerate clean up. This process is documented in a Systems Engineering Evaluation. [G-

 

35 Location and spacing of vertical and horizontal cooling coil loops are specified on waste tank drawings. However, the location 

and routing of supply and return piping for the cooling coil loops was left to the discretion of the construction crews during 

installation. 
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ESR-G-00051] The study investigated options for bulk waste mixing, waste transfer and heel removal. 

The study graded the options on weighted selection criteria such as technical maturity, effectiveness, 

reliability, reusability, radiological control requirements, integration with the tank farm system and cost. 

Knowledgeable tank farm operations, engineering, plant support and maintenance personnel identified 

potential technology candidates based on experience, literature, world wide web research and contacts 

with other knowledgeable personnel in the DOE complex and commercial industry. 

The team recommended using floor-mounted canned Submersible Mixer Pumps (SMPs) for bulk waste 

mixing, a mast-mounted Submersible Transfer Pump (STP) for waste transfer, chemical cleaning using 

oxalic acid for final heel removal and an air-driven submersible pump to enhance final heel removal if 

centrifugal STPs are insufficient to remove the final residual material. The technology consists of a 

mobile substation that provides power, a Mobile Waste Removal Control Center that provides local 

control and monitoring capabilities, SMPs for mixing and suspending waste solids and an STP for waste 

transfer. These mobile units have the capability of being co-located near any tank or tanks scheduled for 

waste removal. This concept efficiently performs waste removal using mobile and reusable equipment 

(Figure 2.3-2).36 [G-ESR-G-00051] 

To date, the SMPs have been used to support bulk waste removal efforts on HTF Tank 13 and FTF 

Tanks 4, 5 and 637 and support mechanical and chemical cleaning of residual heels in Tanks 5 and 6. 

The SMPs are variable speed, single-stage centrifugal pumps with a 305-horsepower motor that can 

operate up to 1,600 revolutions per minute. The SMPs utilize the tank liquid waste to cool the motor and 

lubricate the upper and lower bearings. Two discharge nozzles give the SMPs the capability to produce 

an effective cleaning radius of up to 50 feet. The SMPs are rotated by a turntable assembly that provides 

the motive force for oscillation, or allows for stationary indexing operation. The SMPs have a rotating foot 

attached to the lower end of the pump, which allows the SMP to rest on the tank floor and oscillate. [MCLC-G-00349] 

 

36 Figure 2.3-2 depicts two SMPs located in the waste tank, however, during actual cleaning operations DOE may deploy from one 

to four SMPs within a waste tank based upon the particular waste tank configuration and waste characteristics. 37 Discussion on Tanks 4, 5 and 6 cleaning is included for information; however, Tanks 4, 5 and 6 are located in the FTF at SRS and 

are not within the scope of this HTF 3116 Basis Document. 

Figure 2.3-2: Submersible Mixer Pump Waste Removal Diagram 
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Waste transfers during HTF Tanks 11 and 12, and FTF Tanks 4, 5 and 6 bulk waste removal, as well as 

Tanks 5 and 6 mechanical cleaning and HTF Tank 12 heel removal activities were achieved by using a 

STP. In addition, current bulk waste removal efforts in HTF Tank 13 are utilizing a STP. The STP is a 15 

horsepower, 3,450 revolutions per minute, 250 gallons per minute centrifugal pump that has the capability 

of being located at any elevation within the waste tank. The pump is typically located inside a 22-inch 

diameter sleeve pipe (caisson) that rests on the tank floor. The caisson protects the STP from direct 

discharge from the SMPs. Pump configurations are shown in Figure 2.3-3 and Figure 2.3-4. 

During waste removal supported by SMPs, supernate may be used as the slurry media to minimize the 

amount of water added to the tank farm system. Minimization of water additions to the tank farm is 

important because water introduced into a waste tank becomes additional waste that occupies the limited 

available waste storage space and must be ultimately treated and disposed of. The SMPs are operated 

using various strategies depending on the configuration of the waste in that particular tank. To keep the 

solids suspended in the slurry, mixing continues as long as possible while the slurry is transferred out of 

the tank. The SMPs are required to be shut down as the liquid level approaches the elevation of the 

discharge nozzles to prevent waste spraying, which may cause filters in the tank ventilation system to 

become inoperable and cause potentially high exposure to workers. Typically, the STP continues to 

operate until it loses the ability to pump out any additional waste. After the transfer, the residual solids 

configuration and volume are assessed based on the condition of the waste remaining in the tank. Liquid 

may be transferred back into the waste tank for additional mixing and transfer cycles. 

2.3.3 Chemical Cleaning 

Chemically-aided cleaning techniques have been evaluated for additional levels of waste removal 

following mechanical heel removal. A team of knowledgeable and experienced engineers and scientists 

assessed the current knowledge base and collected and evaluated information available on chemicalbased methods for removing residual solids from the waste tanks. [WSRC-TR-2003-00401] As part of 

this study, the team developed recommendations for chemical treatments to remove residual solids. The 

cleaning agents identified included: 

 oxalic acid, 

 a mixture of oxalic acid and citric acid, 

 a combination of oxalic acid with hydrogen peroxide, 

 nitric acid, 

Figure 2.3-3: Submersible Mixing Pump and Submersible 

Transfer Pump 

Figure 2.3-4: Submersible 

Mixing Pump in a Test 

Tank 
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 formic acid, and 

 organics. 

The results of the evaluation support oxalic acid as the optimal chemical cleaning agent. Nitric acid, 

formic acid and oxalic acid with hydrogen peroxide were all closely grouped for the next best choice. The 

mixture of oxalic acid and citric acid rated poorly (primarily due to the fact that it performed less well than 

oxalic acid and the presence of citrate could adversely impact downstream operations, such as the SWPF 

and the DWPF). Organics rated even more poorly due to large uncertainties in performance and 

downstream impacts. 

The use of oxalic acid was recommended for a number of reasons. First, oxalic acid has been widely 

studied and used several times to clean waste tanks at SRS and at other sites within the DOE Complex. 

Second, the oxalic acid has been shown to be effective for a wide variety of sludge types and outperformed nitric acid and other chemical cleaning agents in head-to-head tests. Lastly, oxalic acid is less 

corrosive to the carbon steel tank than nitric acid or a combination of oxalic acid and hydrogen peroxide. 

[WSRC-TR-2003-00401] 

Chemical cleaning using bulk oxalic acid is the method currently being considered for heel removal in 

HTF waste tanks, as appropriate. However, other cleaning solutions or methods may be used if they 

have comparable or greater effectiveness. The contemplated chemical cleaning methods are similar to 

Bulk Oxalic Acid Cleaning methods used in HTF Tanks 12 and 16 and FTF Tanks 5 and 6, consisting of 

several oxalic acid strikes, use of agitation to facilitate particle-acid contact and a final clean water mixing 

operation. As appropriate, oxalic acid could be sprayed into the tank to clean contaminants from internal 

tank surfaces (e.g., walls, cooling coils, support columns, equipment, etc.). 

Oxalic acid cleaning produces sodium oxalates in the solids slurry that will be added to sludge batches 

that feed the DWPF. Because of sodium limits and oxalate restrictions on the DWPF feed, preparation of 

the feed results in a significant amount of additional material being generated that eventually must be 

processed through SWPF and disposed of in the SDF.38 Therefore, the oxalic acid flowsheet evaluations 

have considered the effects of oxalates on DWPF and salt processing to optimize the quantities of oxalic 

acid introduced into the Liquid Waste System. [WSRC-TR-2004-00317] Due to the downstream effect of 

oxalic acid on DWPF and salt processing, the selection of a chemical cleaning method will be considered 

on each individual application basis. Environmental conditions to which the waste has been exposed also 

affect its dissolution characteristics; therefore, in future chemical cleaning planning, each waste tank (or 

groups of tanks with similar waste and similar historical conditioning) will be considered individually. 

Another example of a chemical cleaning method is Low Temperature Aluminum Dissolution which DOE 

has deployed at HTF to help remove residual solids. This methodology was utilized to help reduce the 

solids volume in Tank 12. [X-CLC-H-00921] Under this process, aluminum is dissolved from sludge 

waste into the supernate by treatment with caustic, followed by decantation and water washing to 

subsequently remove aluminum. Aluminum solids in the sludge are believed to be present in primarily 

three compounds – aluminum trihydrate or gibbsite, alumina monohydrate or boehmite, and 

aluminosilicate. With caustic treatment, the gibbsite form dissolves readily at the relatively low dissolving 

temperatures possible in the waste tanks. The boehmite form dissolves much more slowly and is 

somewhat less soluble than gibbsite, but can still be dissolved at relatively low temperatures, given 

enough time. The aluminosilicate has such a low solubility in waste slurries that it is generally considered 

insoluble. [SRNS-STI-2008-00021] 

2.3.4 Vacuum Heel Removal Cleaning 

As the result of a March 2006 DOE-sponsored Tank Cleaning Technical Exchange, a new vacuum 

technology was identified. The DOE has adapted and successfully used this new technology in the 

unobstructed Type IV tanks, Tanks 18 and 19,39 located at SRS FTF. To deploy this technology in Tank 

18 and Tank 19, DOE utilized a cleaning device, called a Mantis, which consists of a mechanical crawler 

 

 

38 See Appendix A for a brief description of DWPF, SWPF and SDF operations. 39 Discussion of Tanks 18 and 19 is included for information regarding the use of the Mantis technology; however, Tanks 18 and 19

are located in FTF at SRS and are not within the scope of this HTF 3116 Basis Document. 
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and an eductor assembly that made up a retrieval system utilizing an ultra-high-pressure water eductor to 

vacuum residual solids and transport the slurry to a receipt tank (Figure 2.3-5). The process system 

consists of a remotely controlled, in-tank Mantis, an umbilical hose containing hydraulic supply lines and 

the high-pressure water hoses, in-tank waste retrieval hose, a diesel-driven ultra-high-pressure water 

pump, a motor-driven high pressure water pump, hydraulic pump skid, a diesel generator, above-ground 

hose-in-hose transfer lines, Waste Mixing Chamber (WMC) and support equipment. The device was 

inserted into the tank through a 24-inch riser in a folded position. Once inside the tank, the device was 

unfolded into its operational configuration. 

The Mantis was remotely driven around the waste tank bottom by an operator located in the Control 

Center. A high pressure hydro-lance at its front was used to break up waste mounds and an eductor was 

used to vacuum waste from the floor of the waste tanks. The waste traveled through the eductor in-tank 

waste retrieval hose up into a tee spool piece located on top of the tank riser and then through an aboveground transfer line that terminated inside a WMC installed inside a riser on the receipt tank. An 

immersion mill, located near the bottom of the WMC, size-reduced solid waste particles so that the 

particles can be more easily re-suspended in future waste removal activities. [WSRC-TR-2007-00327] 

The cooling coils in Type I, Type II, Type III and Type IIIA tanks precludes the use of large tethered 

mechanical crawlers such as the Mantis platform. However, DOE recognizes the potential for future use 

of vacuum technology deployed on other platforms specifically tailored for applications in tanks with 

obstructions. 

Figure 2.3-5: Mantis 
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2.3.5 Removal of Residual Waste from Failed Cooling Coils 

DOE recognizes that some waste tank cooling coils have failed and that additional failures during waste 

removal are likely. Potential failure mechanisms include: 

 pitting and cracking due to corrosion, 

 cracking due to thermal expansion and contraction, and 

 breaks due to forces imposed on the coils from mixer operation during bulk waste removal and 

heel removal. 

Once a coil has failed, the coil may become 

internally contaminated with waste. Cooling coils 

with the potential for residual waste holdup will be 

flushed, as appropriate. 

2.3.6 Annulus Cleaning 

Currently, eight waste tanks in HTF have leak 

sites on the primary tank walls. The waste tanks 

include all of the Type I tanks (i.e., Tanks 9, 10, 11 

and 12) and all of the Type II tanks (i.e., Tanks 13, 

14, 15 and 16). [SRR-STI-2014-00243] To date, 

waste escaped to the surrounding soils once due 

to a leak site in a primary tank wall. This event 

occurred in 1960 and was associated with Tank 

16. Based on a review of historical waste levels in 

the primary tank and annulus, and subsequent soil 

and groundwater sampling that was performed, it 

was determined that “only a few tens of gallons of 

waste” reached the surrounding soils. [DP-1358] 

Details on the number of leak sites and volume of 

waste on the annulus floor for all of the HTF waste 

tanks is provided in the individual waste tank 

histories contained in Section 2.2.2.2 of this HTF 

3116 Basis Document. Figure 2.3-6 provides an 

example of an HTF waste tank with waste deposits 

on the exterior primary tank wall and Figure 2.3-7 

provides an example of waste deposits on the 

annulus floor. 

Figure 2.3-6: Waste Deposits on Exterior 

Primary Tank Wall Before Cleaning (Tank 12) 

Figure 2.3-7: Waste Deposits on Annulus 

Floor Before Cleaning (Tank 9) 
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To date in HTF, waste removal activities on Tank 16 have included steps to remove material from the 

floor of the waste tank annulus. Annulus cleaning activities in Tank 16 were undertaken in the 1970’s. 

The annulus historical maximum volume was 25,680 gallons. After liquids removal, approximately 6,000 

gallons of solids remained in the annulus. The outside of the primary tank wall was sandblasted to 

facilitate leak site inspection and sand was introduced into the annulus. In 1976, vacuum operations 

removed some of the sand from the annulus. Water was introduced in the annulus to dissolve remaining 

solids and steam jets were used to increase the temperature and promote circulation. Approximately 

1,400 gallons of additional material were estimated to have been removed from the annulus. [SRRCWDA-2011-00126] In 2007, DOE investigated the potential use of vendor supplied robotic equipment to 

remove additional material from the Tank 16 annulus. [LWO-LWE-2007-00204] A proof of concept test 

was completed, however, the project was suspended due to funding constraints. Beginning in 2010, DOE 

initiated a project to re-evaluate this technology and other alternate technologies to determine the 

practicality of additional waste removal from the Tank 16 annulus. DOE evaluated a number of 

technologies and selected water sluicing as the most viable option for additional waste removal from the 

Tank 16 annulus. In 2012 a full scale demonstration test of the water sluicing technology was performed. 

Based on a qualitative assessment utilizing the results of the testing, among other factors, DOE 

determined that it appeared probable that implementation and execution of the technology would not be 

practical. [SRR-CWDA-2013-00041] As required by the HTF GCP (SRR-CWDA-2011-00022), DOE 

reviewed this information with SCDHEC and the EPA. The three agencies (DOE, SCDHEC, and EPA) 

concurred and DOE suspended waste removal activities and initiated final sampling and analysis of the 

Tank 16 residuals. [WDPD-13-40, EPA_04-10-2013, DHEC_04-11-2013] Final residual characterization 

for Tank 16 is currently in progress. As described in Appendix B of this HTF 3116 Basis Document, DOE 

will utilize the final residual characterization results to perform a cost-benefit analysis of additional waste 

removal and will document the cost-benefit analysis, among other things, in a removal report. The 

removal report will combine documentation from the entire waste removal process to provide the 

complete demonstration of radionuclide removal to support Tier 2 Closure Authorization for Tank 16. For 

all HTF waste tanks, the annular regions of the waste tank will be inspected prior to ceasing waste 

removal activities to verify previously documented annulus conditions and determine if any additional 

leakage from the primary waste tank has occurred during the waste removal process. For those waste 

tanks with residual salt waste present in the annulus, the annulus of applicable HTF waste tanks will be 

cleaned to the extent practical. 

2.3.7 Potential Future Waste Tank Cleaning Technology 

DOE will continue to review and consider technological developments relevant to waste tank cleaning and 

will evaluate technologies of comparable, or greater, effectiveness than those discussed above.40 A 

range of potential technologies for evaluation will potentially include proven technologies developed 

and/or used at other DOE sites, in domestic commercial industry and in international applications. Waste 

tank cleaning technologies that will potentially be evaluated include, but are not necessarily limited to, 

sluicing, mixing, chemical cleaning, vacuum retrieval techniques, mechanical manipulators, robotic 

devices and processes that remove (chemically extract) the radionuclides from the residual material that 

may remain in the tank.41 

2.3.8 Ancillary Structures Cleaning 

Ancillary structures are described in Section 2.1.12. 

 

40 On page xvii of its TER, under Key Recommendation ES-2.5, NRC states, “The NRC staff recommends that DOE provide more 

emphasis on removal of HRRs in its technology selection process and provide a clear linkage between the HTF PA results, 

including information regarding the long-term risks associated with the HTF facility, and the demonstration that HRRs have been

removed to the MEP per Criterion 2.” [ML14094A496] Future technology selection and optimization will be informed by the results 

of the HTF PA and DOE will take into consideration specific technologies, with emphasis on HRRs, as described in Appendix B of 

this HTF 3116 Basis Document. 

41 On page xvii of its TER, under Key Recommendation ES-2.4, NRC states, “The NRC staff recommends that DOE continue to 

participate in technology exchanges and evaluate new cleaning technologies as they become available, rather than defaulting to 

previously selected technologies or relying on previous evaluations for technology selection.” In addition, under Key 

Recommendation ES-2.6, NRC notes, “The NRC staff recommends that DOE consider how it might better assess and optimize the 

effectiveness of selected technologies (e.g., obtain better baseline information).” [ML14094A496] DOE agrees with these 

recommendations and will continue to evaluate these areas under DOE Manual 435.1-1, pursuant to DOE’s responsibilities under 

the Atomic Energy Act of 1954, as amended. In addition, DOE will continue to evaluate and optimize the effectiveness of selected 

technologies as described in Appendix B of this HTF 3116 Basis Document. 
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Flushing the transfer lines after use has long been practiced for waste transfers to prevent material build 

up within the systems. Transfer line core pipe flushing has been part of operations of the tank farms from 

at least the mid-1970s, and there is also indication that some level of flushing has always been a part of 

transfer system operations. The rigor to which flushing has been applied has increased over the years. 

The SRS Tank Farm Transfer Control Program requires transfer line flushing based on two factors: 1) the 

radioactivity of the potential residual waste in the core pipe as indicated by the inhalation dose potential 

and 2) the potential for salt solids formation to cause pluggage in the core pipe. The inhalation dose 

potential criteria for flushing have recently been reduced from 9.8E+07 rem/gallon to 3.5E+07 rem/gallon. 

[WSRC-TR-2002-00403] Typically,42 transfer line core pipe flushing based on potential residual material 

radioactivity is required within thirty days after completing a waste transfer. At least three line volumes of 

water are required to be flushed through the core pipe following a transfer identified to be a “sludge slurry 

transfer.” Sludge slurry transfers are those transfers that may result in a one weight percent or greater, 

concentration of sludge during the transfer. Three line volumes at a normal flush water system rate is 

expected to dilute the sludge slurry by 99 %. At least one line volume of water is required following other 

types of transfers and for flushes required by potential salt solids formation. In addition to transfer line 

flushing based on the potential residual waste radioactivity and salt solids formation potential, transfer 

lines are also typically43 flushed if there is a section of the transfer line that may not be self-draining (low 

point). [WSRC-TR-2002-00403] 

Transfer lines may be cut and capped during the isolation and closure process for an associated waste 

tank. Typically,44 the transfer lines will be flushed to remove contamination prior to cutting and capping as 

an as low as reasonably achievable (ALARA) practice to reduce the potential for worker dose. In 

summary, transfer line flushing is always performed when the transferred material meets the criteria for 

residual radioactivity or the criteria for salt solids formation. Additionally, transfer lines are typically45

flushed if there is a low point in the transfer route and at other times to reduce potential worker dose. 

In addition to operational practices of flushing, specific design practices have contributed to removing the 

waste from waste transfer line piping systems. The installation of stainless steel for the waste transfer 

core piping, the transfer piping sloping toward a waste tank with minimal valves and the layout of turn 

radii are specific design features that prevent waste accumulation in the piping systems. 

Secondary containment (e.g., transfer line jackets, leak detection box encasements) is provided for 

transfer line core pipes. No leakage of waste from primary core pipes into secondary containment has 

been identified. Transfer lines currently in service are tested, or evaluated, as part of the Structural 

Integrity Program. [S-TSR-G-00001]. The structural integrity air or helium testing of transfer lines 

procedurally requires radiological surveys to check for indications of contamination. [SW10.6-SVP-5] This 

testing has not identified any significant contamination in HTF transfer line secondary containment. 

Therefore, no residual waste is expected in these structures at the time of closure. 

As described in Section 2.1.12, there are three evaporator systems in HTF, the 242-H Evaporator 

System, the 242-16H Evaporator System and the 242-25H Evaporator System. The 242-H evaporator 

and associated concentrate transfer system were shutdown in 1994. [HLW-2002-00025] The 242-16H 

and 242-25H Evaporator Systems remain operational. 

The final residual levels in the three HTF Evaporator Systems are expected to be comparable to the 

residual levels achieved in the 242-F Evaporator System due to similar design, operational histories, and 

comparable cleaning techniques. The SRS FTF 242-F evaporator and associated concentrate transfer 

system were shut down in 1988. Various waste removal campaigns were completed in 1991, 1992 and 

again in 2004. During the 2004 waste removal campaign, various mixing and transfer cycles were 

performed in the evaporator vessel and concentrate transfer system waste tank. Following the 2004 

 

42 Flushing will normally be carried out as described, however, there may be some instances in which flushing is not, or cannot, be 

performed. For example, if the time between multiple transfers is less than thirty days, flushing may not be performed until after the 

last transfer. Also, equipment issues could impact the ability to perform routine flushing following transfers. Similarly, although 

transfer lines would normally be flushed prior to cutting and capping, there may be instances where lines cannot be flushed due to 

such things as the condition of the transfer line or the transfer line configuration (e.g., no route for flushing liquid, no access to 

transfer line). 

43 See footnote 42. 44 See footnote 42. 45 See footnote 42. 
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waste removal campaign, the evaporator vessel and the concentrate transfer system tank were inspected 

and the contents were sampled and characterized. [CBU-LTS-2004-00078] 

There are nine stainless steel pump tanks in HTF: HPT-2 through HPT-10. There is a single catch tank 

in HTF designed to collect drainage from HDB-1 and the Type I tank transfer line encasements. These 

stainless steel tanks are accessible for waste removal. [SRR-CWDA-2010-00023] 

Prior to closure, secondary containment structures (e.g., pump pits, diversion boxes, leak detection 

boxes, modified leak detection boxes and valve boxes) will be inspected and exposure to waste over their 

service life determined. The pump pits are shielded reinforced concrete structures located below grade at 

the low points of transfer lines and are lined with stainless steel. These structures are secondary 

containments that house the pump tanks. The diversion boxes are shielded reinforced concrete 

structures containing transfer line nozzles to which jumpers are connected to direct waste transfers to the 

desired location. The majority of the diversion boxes are located below ground and are either stainless 

steel-lined or sealed with waterproofing compounds. Valve boxes provide secondary containment for 

valves and transfer line jumpers to facilitate specific waste transfers. The valves are generally manual 

ball valves in removable jumpers with flush water connections on the transfer piping. Since these 

structures do not provide primary containment, they are expected to contain only incidental radionuclide 

contamination at the time of closure. [SRR-CWDA-2010-00128] 

2.4 Radionuclide Inventory in H-Tank Farm Facility Systems, Structures and Components 

Minimal quantities of residual material are expected to remain in HTF at the time of closure following 

waste tank and ancillary structures cleaning activities. Data developed for the HTF PA, Revision 1,46

provided the estimated HTF residual radionuclide inventory used in this HTF 3116 Basis Document. 

[SRR-CWDA-2010-00128] Estimated radionuclide concentrations used to develop the HTF PA, Revision 

1, residual radionuclide inventory were determined by three methods: 

 sample analysis 

 process knowledge data maintained in a controlled database (WCS) 

 special analysis 

The inventory of each waste tank was used to establish the characterization of the estimated residual 

material in the ancillary structures and was decayed to year 2032.47 These estimates should not be 

considered limits associated with future cleaning activities nor should they be considered binding 

estimates. The following subsections describe the process for estimating residual material inventory at 

closure for the HTF waste tanks and ancillary structures, as provided in the HTF PA, Revision 1.48 [SRRCWDA-2010-00128] 

2.4.1 Residual Inventory for Waste Tank Primary Tanks 

A methodical approach was used to construct reasonably conservative estimates of HTF waste tank 

closure inventories to be used in performance assessment modeling. Independent steps were developed 

to systematically construct the estimated HTF tank inventories, with each step adjusting inventory either 

by tank or by radionuclide. The steps used in inventory development were as follows.49 [SRR-CWDA2010-00023] 

1. The initial list of radionuclides to be included in the HTF tank inventories was established. This 

list was developed beginning with an initial listing of 849 radionuclides compiled from a variety of 

published resources including those from Tables 1 and 2 of 10 CFR 61.55. [CBU-PIT-2005-

00228] 

 

46 Unless otherwise noted, the information in Section 2.4 and subsequent Sections 2.5 and 2.6 is based on information developed in 

support of the HTF PA, Revision 1. As required by DOE Manual 435.1-1, maintenance of the HTF PA will include future updates to

incorporate new information, updated model codes, analysis of actual residual inventories, and other information as appropriate. 

[SRR-CWDA-2010-00128] 

47 The year 2032 corresponds to the year the HTF PA assumes, for the purposes of analysis, that a 100-year period of institutional 

control will begin. 

48 On page 3-55 of its TER, NRC staff states, “DOE’s approach to developing inventories for tanks that have not been cleaned is 

reasonable and appears to be generally conservative (tends to over- rather than under-predict inventory).” [ML14094A496] 49 Removal experience in Tanks 5, 6, 18 and 19 was used to support development of the HTF PA residual inventory estimates; 

however, Tanks 5, 6, 18 and 19 are located in the FTF at SRS and are not within the scope of this HTF 3116 Basis Document. 
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2. The contaminant screening process consisted of several steps to arrive at an appropriate list of 

54 isotopes to be included in the HTF waste tank closure inventory estimates to be used in the 

HTF PA modeling. This initial screening process considered the following information: 

 Physical properties of each radionuclide such as half-life and decay mechanism, 

 Waste source and handling based on radionuclide production mechanisms and time since 

the radionuclide was produced, and 

 Screening factors for radionuclide ground disposal developed in NCRP-123 which convert a 

quantity of each radionuclide to a dose. 

This initial screening reduced the radionuclide list from 849 down to 159 radionuclides. [CBUPIT-2005-00228] 

Additional screening of the 159 radionuclides was performed to identify the radionuclides, from 

this list of 159, to be considered in the initial HTF PA inventory. The screening criteria included 

the following: 

 Radionuclides were screened out if there were no ancestors present from the specific decay 

chain or no decay source for the radionuclide. 

 Evaluation of HTF waste production history information for the potential for a specific 

radionuclide. This criterion screened out radionuclides not present within the HTF waste. 

 In general, radionuclides present only due to ingrowth from a decay series were screened 

out, however, production history was used to retain those radionuclides present at a greater 

proportion than from the decay series. This criterion eliminated radionuclides that are 

present only due to the decay of their parent radionuclide. The inventory of these 

radionuclides can be controlled by removing the parent radionuclide(s). 

 Radionuclides with less than a five year half-life were screened out. This criterion reflects 

that active institutional control will be maintained over the site for 100 years after facility 

closure. The inventories of these radionuclides will be significantly diminished due to the 

amount of radioactive decay that will occur during the 100-year institutional control period. 

Based on these screening criteria, an additional 105 radionuclides were screened out, thus 

reducing the HTF PA modeling initial inventory radionuclide number to 54. [SRR-CWDA-2010-

00128] 

3. The concentrations for each of the 54 radionuclides were estimated by using the WCS database 

[DOE-WD-2005-001] or physical relationships. The WCS is an electronic information system that 

tracks waste tank data, including projected radionuclide inventories based on sample analyses, 

process histories, composition studies and theoretical relationships. The WCS, initially developed 

in 1995, and populated with historical information, tracks, among other things, the dry sludge 

concentrations of radionuclides in each of the SRS waste tanks. The primary purpose for 

developing WCS was to provide reasonable estimates on which to base safety analysis 

evaluations such as criticality issues in the tanks farms. Safety analysis evaluations, in general, 

build in a degree of conservatism, consequently, a level of conservatism for some materials is 

reflected in the data. The radionuclides tracked in the WCS were selected primarily based on 

their impact on waste tank safety basis source term, inhalation dose potential or on the E-Area 

Vault waste acceptance criteria. For the radionuclides with information available in WCS, the 

concentration was calculated by dividing the dry sludge activity by the corresponding settled 

sludge volume extracted from WCS. 

In the WCS, a subset of the HTF waste tanks required additional estimating where no input was 

available for various radionuclides. In addition, alternate methods provided estimates for 

additional radionuclides generated as activation products. Certain HTF waste tanks contain 

zeolite in addition to the sludge material. The estimated radionuclide concentration in Tanks 24, 

32, 38, 40, 42 and 51 were adjusted to account for zeolite and corresponding captured cesium. 

Affected radionuclides and the methods used to estimate their concentrations are detailed in 

SRR-CWDA-2010-00023. 

4. The estimated residual volume for the HTF waste tanks, for the purpose of developing the HTF 

PA inventory, was based on waste removal experience in Tanks 5, 6, 16, 18 and 19. Based on 
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this experience, the projection for residual material volume remaining in the primary tank was 

conservatively estimated to be 4,000 gallons for each HTF waste tank with the exception of Tank 

16. To provide a reasonably conservative estimate for Tank 16, 1,000 gallons was used as an 

estimate of residual volume in the primary tank. 

5. To determine the initial radionuclide inventory estimates for each of the HTF waste tanks, the 

concentrations, discussed in Step 3 above, were multiplied by the residual volume estimates, 

discussed in Step 4 above, to determine the waste tank inventory for each of the radionuclides. 

The radionuclide inventories of residual liquid were assumed to be included in the final closure 

solids inventory as the residual liquid has typically evaporated by the time samples are taken from 

the waste tank. 

6. Following the development of the initial inventory estimates, adjustments were made to add a 

reasonable conservatism to the inventory estimates. The steps outlined below summarize the 

adjustments made to the initial inventory estimates. 

7. Allowing for a more efficient and cost effective means of confirming concentrations for 

radionuclides with a limited potential impact to dose, the inventories for these radionuclides were 

adjusted to either one curie or used an analytical detection limit. If the radionuclide initial 

estimated inventory was less than the detection limit, then it was adjusted up to the detection 

limit. However, if the radionuclide initial estimated inventory was at least at the detection limit, 

then it was adjusted up to one curie. Recent sample analyses from Tanks 5, 18 and 19 were 

reviewed for appropriate detection limits. [SRNL-STI-2012-00034, SRNL-STI-2010-00386, 

SRNL-STI-2010-00439] The purpose of this adjustment was to bound those specific 

radionuclides within the HTF PA analysis which will allow for easier comparison of the impact of 

the final residual inventory versus the HTF PA analysis without artificially increasing the overall 

estimated doses within the HTF PA. Evaluating a higher than anticipated value in the HTF PA 

allows for a higher analytical detection limit to be established for those radionuclides which do not 

impact dose. In general, during laboratory analysis of waste tank residuals, the lower the 

concentration value of a specific radionuclide that is required to be measured (i.e., analytical 

detection limit), the more time consuming and costly the analysis becomes. The adjustments to 

either the detection limit or to one curie exclusively increased residual inventory estimates. 

Inventory estimates were not adjusted lower, only higher, from the initial estimated inventory. 

Those radionuclides with initial estimated inventories of greater than one curie were not adjusted 

in this step. 

For those radionuclides that have been observed (through previous analyses or scoping studies) 

to have greater potential impact on the overall dose, if the initial estimated inventory was less 

than the detection limit the inventory was adjusted to the analytical detection limit. However, if 

the radionuclide initial estimated inventory was at least at the detection limit, then it was not 

adjusted further. The reason for not making adjustments to these specific radionuclides was that 

artificially increasing the inventory for these radionuclides, for the purpose of making residual 

material analysis more efficient and cost-effective, may artificially increase the estimated doses 

within the HTF PA and mask the potential impact of other radionuclides. 

8. Based on the differences in concentrations observed due to the waste removal process in Tank 5, 

decreases in concentrations are anticipated for cesium, strontium and zirconium. [WSRC-STI2007-00192, SRNL-STI-2009-00492] Based on this observation, all the waste tank inventories of 

these radionuclides were adjusted one order of magnitude lower to reflect the effect of the waste 

removal process. 

9. Based on results from the Tank 5 inventory determination and FTF Performance Assessment 

(PA) (SRS-REG-2007-00002), inventory estimates required adjustment for Tc-99 and Zr-93. For 

Tc-99, the initial estimated inventories were believed to be overly conservative. The FTF PA 

estimated inventory for Tc-99 was close to three orders of magnitude higher than the final 

inventory. To reduce the overestimate while maintaining a reasonably conservative inventory 

estimate for Tc-99, each waste tank’s initial estimated inventory was reduced by one order of 

magnitude. In previous versions of HTF PA inventory estimates, Zr-93 inventory estimates were 

based on a detection limit value. In Tanks 5, 18 and 19, Zr-93 was measured. Therefore, a 
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method to estimate Zr-93 inventories was needed. Using these sample results, Zr-93 was 

estimated by using a ratio to the Sr-90 concentrations due to similar fission yields. 

10. The waste tanks were binned according to waste 

tank use and design. The tank type generally 

had an effect on the type of waste received and 

therefore guided the group selection. In general, 

each waste tank was built at approximately the 

same time as others of the same type. The bins 

for the waste tanks are presented in Table 2.4-1. 

As each waste tank is cleaned, the waste 

removed from within the tank during waste 

removal will be transferred to other tanks. Due 

to transfer line configuration, the material in one 

tank will typically pass through other tanks of the same type prior to exiting the HTF. Due to the 

uncertain order of tank waste removal and closure activities, the maximum inventory 

concentrations associated with an individual tank within a tank group were applied to the other 

tanks within the bin. The maximum adjusted inventory, to this point, was used as the estimated 

inventory for all tanks within each waste tank group. 

11. Within each waste tank group, for each of the radiounuclides, after adjusting the initial estimated 

inventories, as described above, the maximum radionuclide inventory for any one waste tank was 

used to estimate that radionuclides’ inventory for the other waste tanks within the grouping. The 

one exception to this was that for adjustments to Pu-238 inventories within the Type III/IIIA tank 

group, the waste tanks were further grouped by salt and sludge tanks prior to making this 

adjustment. 

Based on experience with previous performance assessments, overestimating the Pu-238 

inventories can ultimately exaggerate the projected overall dose due to the associated ingrowth of 

Ra-226. To reduce this exaggeration for the purpose of estimating Pu-238 inventories in the 

Type III/IIIA tanks, the grouping was split based on the two different waste types (salt and 

sludge). 

At the completion of waste removal for each of the tanks, the estimated residual inventory identified for 

that tank in the HTF PA will be compared and evaluated against the actual residual inventory determined 

during final residual characterization after the tank has been cleaned. [SRR-CWDA-2010-00128] The 

actual residual inventory will be developed from a determination of the residual material volume combined 

with analytical data from a statistically based sampling program of the residual material. Additional 

information regarding the comparison and subsequent evaluation between estimated and actual residual 

inventories is provided in Appendix B.50

2.4.2 Residual Inventory for Tank Annuli 

Wall inspections of the waste tanks have found cracks where material has seeped into the secondary 

containment or annulus. The amount of material contained in the waste tank’s annulus has been 

estimated. Based on these estimates, inventories within the appropriate annuli were estimated. 

To estimate each tank’s annulus inventory, estimates of the residual volume and constituent 

concentrations were prepared. 

The most current estimates, at the time the inventory was developed, of the amount of material within the 

tank annuli were used to estimate the residual volumes at closure. [C-ESR-G-00003] The Type I and 

 

50 On page xvii of its TER, under Key Recommendation ES-2.1, NRC states with respect to Criterion 2 of NDAA Section 3116 

(NDAA Section 3116(a)(2), “The NRC staff recommends that DOE explore methods to improve the process by which residual waste 

volumes and associated uncertainty are estimated. The NRC staff also recommends that DOE evaluate and clearly communicate 

the relative contributions of various forms of uncertainty related to the radionuclide sampling process.” [ML14094A496] DOE will 

continue to incorporate actual residual waste characterization information in special analyses to update the HTF PA, and explore 

appropriate methods to reduce uncertainty in the HTF PA, as part of the HTF PA maintenance process under DOE Manual 435.1-1, 

pursuant to DOE’s responsibilities under the Atomic Energy Act of 1954, as amended. 

Table 2.4-1: Waste Tank Groupings 

Type I & II Type III/IIIA Type IV 

9, 10, 11, 12, 

13, 14, 15 

29, 30, 31, 32, 

35, 36, 37, 38, 

39, 40 ,41 ,42, 

43, 48, 49, 50, 

51 

21, 22, 23, 

24 

Note: Tank 16 is a special case with its own grouping 
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Type II tanks are known to have leak sites and material in their annuli. Type IV tanks do not have annuli 

and the Type III/IIIA tanks have no residual material within their annuli. 

The amount of material in the Tank 16 annulus was estimated at that time to be 3,300 gallons.51 [SRRLWE-2012-00039] For other annuli with significant volume, this volume was also used. Except for Tank 

16, the material in the annuli is expected to be highly soluble. This is due to the material originally being 

supernate that leaked into each annulus and dried. Therefore, the 3,300-gallon estimate for all other 

waste tanks is believed to be reasonably conservative. Tank 16 was expected to be an exception due to 

the mixture of silicon from sand blasting activities carried out in the Tank 16 annulus. This material was 

expected to limit the quantity of material removed or limit the practicality of waste removal from the 

annulus and, was therefore the reason the 3,300-gallon volume estimate was used as the reasonably 

conservative estimate for the appropriate annuli volume. 

For those waste tank annuli with a trace amount of material, a reasonably conservative volume of 100 

gallons was used. 

Characterization of the material within the various annuli is based on available samples taken from 

annulus material. Recently, samples were collected from the Tank 16 annulus. Four samples were taken 

at various locations within the annulus and numerous constituents analyzed. These sample results were 

used to inform the inventory estimates for the HTF waste tanks with annulus material. [SRNL-STI-2012-

00178] For those constituents analyzed, the concentration reported provided the estimate of that 

constituent’s concentration. Since the sample analysis did not include all constituents of concern, the 

remaining constituents were estimated, as explained below. The type of adjustment made to each 

radionuclide in the primary tank residual inventory estimate (see Section 2.4.1) determined the method to 

estimate each radionuclide’s concentration for the associated annulus. 

Those constituents, whose Tank 16 primary tank inventory was estimated by using detection limits (see 

Section 2.4.1), concentrations were estimated by taking a ratio of the primary tank residual volume 

estimate to the annulus residual volume estimate. For example, Cl-36 was estimated in the Tank 16 

primary tank residual material via a detection limit; a ratio of the Tank 16 primary tank residual volume 

estimate (1,000 gallons) to the annulus residual volume estimate (3,300 gallons) multiplied by the primary 

tank residual inventory estimate was used to estimate the Cl-36 concentration in the annulus material. 

For those constituents estimated in the Tank 16 primary tank via the unit curie adjustment methodology 

(see Section 2.4.1), a unit curie was also assigned to the annulus inventory. For example, Ac-227 was 

estimated in the Tank 16 primary tank inventory estimate at one curie. Therefore, the Tank 16 annulus 

inventory estimate of Ac-227 was also one curie. 

For those constituents with sample analysis results or considered significant to dose analysis, 

concentrations were based on a ratio of a chemically similar element, within the analysis, and the 

applicable primary tank residual estimate. A ratio to the Pu-238 analysis was used for the radionuclides 

that would tend to be insoluble, while a ratio to the Tc-99 analysis was used for the more soluble 

components. For example, Am-241 was not analyzed in the Tank 16 annulus samples. Considering 

Am-241 can be significant to dose analysis, its annulus concentration was estimated by using the Pu-238 

sample concentration and a ratio of the Am-241 to Pu-238 primary inventories. The Am-241 

concentration in the Tank 16 annulus material was estimated by multiplying the Pu-238 annulus samples 

concentration by a ratio of the Am-241 Tank 16 primary tank inventory to the Pu-238 Tank 16 primary 

tank inventory.52

The annulus inventories were estimated by multiplying the volume and concentration estimates. The 

decay date for these inventories is 2032.53

 

51 After further review including activities associated with sampling the Tank 16 annulus, the final volume determination for the Tank 

16 annulus was 1,910 gallons. [U-ESR-H-00113] 

52 Final residual characterization for Tank 16 is currently in progress. Following the process outlined in Appendix B of this HTF 3116 

Basis Document, a Tank 16 Special Analysis will be developed and formally reviewed and approved that will consider the final 

characterization of the residuals remaining in both the primary tank and the annulus. The data from the Tank 16 Special Analysis 

will support a cost-benefit analysis and subsequently, the development of a removal report for Tank 16. 53 The year 2032 corresponds to the year the HTF PA assumes, for the purposes of analysis, that a 100-year period of institutional 

control will begin. 
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2.4.3 Residual Inventory for Type II Sand Layers 

All Type II tanks have both a primary and secondary sand layer. The one-inch thick primary sand layer is 

between the primary tank and secondary liner and the one-inch thick secondary sand layer is between 

the secondary liner (annular pan) and the basemat. Due to the material that leaked from the Type II 

primary tanks into each tank’s annulus, residual material has been assumed to be present within the 

respective sand layers. The sand layer inventories were estimated by multiplying the residual 

concentrations by the estimated residual quantity, as explained further below. 

The residual material within the sand layer was assumed to have the same concentrations as determined 

for the annulus material (see Section 2.4.2). 

The quantity estimate within the Type II tank sand layers was based on the operational history of each 

waste tank. For Tanks 14 and 16, a significant quantity of material leaked from the primary tank into the 

secondary containment and was sufficient to deposit material at a depth of several inches. For Tanks 13 

and 15, a minimal quantity of material has leaked from the primary tank. This is based on the inspections 

of the annulus floor where negligible quantities of material have been observed. [SRR-CWDA-2010-

00023] 

The Type II tanks have a grout pad that surrounds the primary sand layers. This grout pad would limit the 

flow of material into the sand layer. The top part of the grout pad meets the bottom of the primary tank 

liner. For material to reach the primary sand layer, the liquid level in the annulus would need to be higher 

than the top part of the grout pad. In Tanks 13 and 15, since the quantity of material that leaked from the 

primary tank is limited, the amount of material that moved into the sand layer is considered negligible. 

Following the reasonably conservative approach, 100 gallons was assumed to be present within the 

primary sand layer for Tanks 13 and 15. For Tanks 14 and 16, the depth of material in the annulus 

suggests the possibility of material movement into the sand layer. It is thought that little liquid material 

moved into the sand layer due to the tight, although not sealed, fit between the grout pad and primary 

tank liner. However, to be reasonably conservative, the void space within these sand layers was 

estimated to be completely saturated with residual material. [SRR-CWDA-2010-00023] 

The Type II tanks also have a secondary sand layer that is beneath the secondary liner or annulus. 

Tank 16 experienced the largest quantity of material leaving the primary tank and gathering in the 

annulus. In 1960, the level in the annulus was above the top of the 5 foot high annular pan, liquid was in 

direct contact with the concrete vault wall. Based on detailed analysis, it is estimated that tens of gallons 

escaped from the Tank 16 system and reached the surrounding soils. [DPSPU 77-11-17] For the 

purpose of this inventory evaluation, it is assumed that all of the material (26 gallons) that is estimated to 

have escaped the Tank 16 system entered the secondary sand layer below Tank 16. For Tanks 13 

through 15, no material has leaked beyond the secondary containment; therefore, it is assumed that the 

secondary sand layers below these waste tanks contain no inventory. [SRR-CWDA-2010-00023] 

The sand layer inventories were estimated by multiplying the volume and concentration estimates. The 

decay date for these inventories is 2032.54 [SRR-CWDA-2010-00023] 

2.4.4 Residual Inventory for Failed Cooling Coils and Internal Waste Tank Surfaces 

When a solution is in contact with a solid phase, the constituents may partition between the liquid and 

solid. In the waste tank, liquid waste is in contact with corrosion products on the tank interior wall, so 

small quantities of radioactive material could be held on the corrosion products (i.e., iron oxides). [SRTWPT-2005-00049] 

The inventories for inside failed cooling coils and on the surface of the waste tank walls, cooling coils and 

columns are encompassed by the estimated total tank inventories. Cooling coils with the potential for 

residual waste holdup will be evaluated and flushed with water as appropriate. Flushing is expected to 

remove residual waste that may have entered damaged coils. The volume of cooling coils represents 

less than 1 % of the entire waste tank volume. [SRR-CWDA-2010-00023] 

 

54 The year 2032 corresponds to the year the HTF PA assumes, for the purposes of analysis, that a 100-year period of institutional 

control will begin. 
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2.4.5 Residual Waste Inventory for Transfer Lines, Pump Tanks and Catch Tanks and 

Evaporator Systems 

Ancillary structures include transfer lines, transfer line secondary containment, pump tanks, pump pits, 

the HTF catch tank, diversion boxes, valve boxes, concentrate transfer system tanks and the evaporator 

systems. Over the operating life of the facility, radioactive waste comes into physical contact with some 

of these components, contaminating them and hence, leaves contamination on the components. The 

degree of contamination depends on many factors, which include, but are not limited to, the service life of 

the component, the material of construction and the type of waste in contact with the component. Some 

of the listed equipment serves only as secondary containment and may not have contacted the waste. 

[SRR-CWDA-2010-00023] 

Ancillary structures inventories are estimated for the following categories: 

 transfer lines, 

 pump tanks, 

 concentrate transfer system tanks, and 

 evaporator vessels. 

To estimate the inventory associated with the transfer lines, pump tanks and concentrate transfer system 

tanks, the inventory of each waste tank is used to establish the characterization of the residual material in 

the ancillary structures. The results of a review of waste transfers within HTF and between FTF and HTF 

have been sorted to determine the percentage of the volume of all waste transfers that can be attributed 

to each HTF waste tank. The representative tank’s concentration was then determined by applying a 

weighted average to each isotopic distribution in the HTF waste tanks. The characterization of dry sludge 

was used for each waste tank for conservatism. It is important to note that while the sludge 

concentrations were used, dry sludge is only a small portion of the total waste that passes through the 

transfer lines that are routinely flushed with a high volume of supernate. Using the dry sludge 

concentrations provides a conservative representation of the actinides and long-lived isotopes. The 

short-lived isotopes, which are more concentrated in the supernate than the sludge, will have decayed 

significantly during the 100-year active institutional control period and are not expected to be a significant 

portion of the residual material in the ancillary structures. [SRR-CWDA-2010-00023] 

To estimate the residual inventory associated with the HTF evaporators, sample characterization from the 

FTF 242-F Evaporator System was used. At this time, no residual sample characterization (i.e., after 

waste removal) exists for any of the HTF evaporators but, based on the similarities of evaporator designs 

and the anticipated similarities between evaporator flushing and cleaning techniques, the final residual 

source term is not expected to differ significantly from the 242-F system. 

2.5 Residual Waste Stabilization 

In May 2002, DOE issued an EIS on waste tank cleaning and stabilization alternatives. DOE studied five 

alternatives: 

 empty, clean and fill tank with grout, 

 empty, clean and fill tank with sand, 

 empty, clean and fill tank with saltstone, 

 clean and remove tanks, and 

 no action. 

The EIS concluded the Fill with Grout option was preferred. DOE also issued a ROD selecting the Fill 

with Grout alternative for SRS waste tank closure. [DOE/EIS-0303 ROD, SRR-CWDA-2010-00128] 

Evaluations described in the EIS showed the Fill with Grout alternative to be the best approach to 

minimize human health and safety risks associated with closure of the waste tanks. This alternative 

offers several advantages over the other alternatives evaluated such as the following: 

 provides greater long-term stability of the tanks and their stabilized contaminants than the sandfill approach,
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 provides for retaining radionuclide within the tanks by use of reducing agents in a fashion that the 

sand-fill would not,

 avoids the technical complexities and additional worker radiation exposure that the fill-withsaltstone approach would entail,

 produces smaller impacts due to radiological contaminant transport than the sand- and saltstonefill alternatives, and

 avoids the excessive personnel radiation exposure and greater occupational safety impact that 

would be associated with the clean and remove alternative. 

Cementitious materials are often used to stabilize radioactive wastes. Grout has been one of the most 

commonly used materials for solidifying and stabilizing radioactive wastes and the technology is at a 

mature stage of development. Grout is a mixture of primarily cement and water proportioned to produce 

a pourable consistency. Stabilization is needed to maintain the waste tank structure and minimize water 

infiltration over an extended period of time, thereby impeding release of stabilized contaminants into the 

environment. 

The waste tank fill grout will likely be reducing grout, which has low Eh minimizing the mobility of certain 

radionuclides after closure. All grout formulas are alkaline because grout is a cement-based material that 

naturally has a high pH, which is compatible with the tank carbon steel. The tank fill grout will have a 

relatively high compressive strength and low permeability, which enhances its ability to limit the migration 

of contaminants after closure. The grout formulas must be fluid to allow a near-level placement. [SRRCWDA-2010-00128] 

The grout properties studied consisted of two major states, cured and fresh. [WSRC-STI-2007-00369] 

The major requirements for cured properties of grout include compressive strength, saturated hydraulic 

conductivity, porosity and dry bulk density. The fresh grout properties include flow, bleed water, set time, 

air content and wet unit weight (density). [SRNL-STI-2011-00551, WSRC-STI-2007-00641] 

The fluidity of the mixtures is one of the main requirements. Grout requirements consist of both 

mechanical and chemical properties. The mechanical requirements of the grout consist of adequate 

compressive strength to withstand the overburden load and provide a physical barrier to discourage 

future intruders. The chemical requirements of grout include a high pH and a low Eh to create an 

environment that makes contaminants less soluble and less mobile. Grout with a high pH can accept 

protons to lower the pH of contaminants to make them less soluble. Grout with low Eh has the tendency 

to donate electrons and thereby reduce contaminants to make them less mobile. 

Reducing grout will likely be used to fill the Type I, II, III and IIIA tanks. In the Type IV tanks, reducing 

grout will likely be used to fill the waste tank volume and the tank dome will be filled with a strong grout 

(i.e., a grout with compressive strength properties in excess of 2,000 psi) to deter intrusion. However, if 

the final reducing grout recipe provides equivalent compressive strength (2,000 psi minimum) as the 

strong grout, then only reducing grout will be used for the Type IV waste tanks. Type I, II, III and IIIA 

tanks have sufficient thickness of reinforced concrete roofs to deter such intrusion. 

For waste tank types with cooling coils and annuli, the cooling coils and annuli will be grouted to minimize 

void spaces, to minimize fast flow pathways and for stability. Annulus risers and ductwork will be filled 

with grout up to grade level and closed and capped. 

Various pieces of equipment will remain in the waste tanks at the time of closure. This equipment 

consists of items such as transfer jets, thermowells, level instrumentation, a leak detection system, 

transfer piping out of the waste tank and equipment directly used in tank closure operations (such as 

submersible mixers and pumps, cables, temporary transfer hoses). These various pieces of equipment, 

in both the primary tanks and the annuli, will be grouted to the extent practical. In addition, steel tapes 

and other miscellaneous debris will remain in the waste tank after closure. These components will be 

entombed in the grout as part of the closure process. 

The intent of the in-tank equipment grouting process using highly flowable grout is to minimize fast flow 

paths that would potentially be present due to void spaces in equipment that extend vertically from the 

waste tank top down through the grout to close proximity to the residual waste at the bottom of the 

grouted waste tank. The configuration of the grouted waste tank, annulus and equipment is intended to 
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eliminate fast flow paths (i.e., significant vertical voids that provide a pathway for infiltrating water to 

bypass the grout layer and impact the contaminant zone), consistent with the Base Case (Case A) 

presented in the HTF PA. [SRR-CWDA-2010-00128] During grout planning, equipment to be entombed 

in the grouted waste tank will be identified and documented. This identification will take into consideration 

the location of the equipment in the waste tank, the state of the equipment in question (e.g., is the 

equipment failed) and the practicality of removing the equipment (e.g., potential worker dose 

considerations). Waste tank top modifications to equipment will be performed to provide access to deliver 

grout to the void spaces of equipment that will be entombed in the grouted waste tank. For example, the 

motor at the top of abandoned standard slurry pumps (Figure 2.3-1) may be removed to provide access 

to deliver grout into the pump column. Efforts will be made to assess the completeness of filling 

equipment void spaces. For example, as practical, grout may continue to be pumped into the supply pipe 

of a transfer jet until grout is observed exiting the discharge pipe of the jet to demonstrate the void space 

has been filled with grout. Small equipment such as sample crawlers have minimal void spaces and 

grout will generally flow into horizontal spaces. 

The DOE has successfully developed and tested highly flowable grout in preparation for filling void space 

in equipment that will be entombed inside of waste tanks at closure. In full scale tests, simulated 

horizontal and vertical cooling coils were filled with a grout formulation consisting of slag and cable grout. 

Examination of grouted simulated cooling coil test samples indicated that air entrainment and resulting 

void space was minimal (much less than 4 %). The use of this type of grout will maximize the ability to fill 

voids in equipment in the waste tanks at closure. [WSRC-STI-2008-00172, WSRC-STI-2008-00298] 

In general, equipment that extends to the tank top will have its void spaces filled with grout directly, while 

equipment that does not extend to the tank top will have its voids grouted indirectly through 

encapsulation. The void space of some equipment inside the waste tanks, that do not extend to the top 

of the waste tank, cannot be fully grouted. For example, the void space in transfer pumps located at 

various elevations within FTF Tanks 18 and 19 were not fully grouted because of limited or inadequate 

grout delivery access.55 However, void spaces of equipment, such as these transfer pumps, entombed 

inside the waste tanks do not extend to the top of the waste tank and therefore do not provide a vertical 

void space of significant length to create a fast flow path through the grouted tank. For example, the 

Pitbull pump in FTF Tank 19 is approximately 25 feet below the northeast riser; and the transfer pump in 

FTF Tank 18 (used for Tank 18 to Tank 19 transfers) is approximately 29 feet below the west riser. The 

other transfer pumps and dewatering pumps in the northeast risers of both Tanks 18 and 19 are 

positioned in very close approximation to the waste tank floor. As described above, because this 

equipment does not extend to the top of the waste tank and does not provide a vertical void space of 

significant length, these void spaces would not provide a fast flow path to the residual inventory. 

With the exception of the transfer lines, ancillary structures (e.g., diversion boxes, pump tanks, pump pits) 

associated with HTF will be filled, as necessary, at final closure to prevent subsidence. Since any 

residual radioactive waste would be on the interior wall of the transfer lines and the leach rate would not 

be significantly influenced by grout, there is no environmental or human health and safety benefit to 

grouting these small diameter transfer lines. [SRR-CWDA-2010-00128] 

2.6 Closure Cap 

An engineered closure cap will be installed over the HTF following the closure of the waste tanks and 

ancillary equipment. The HTF conceptual closure cap design is presented in SRNL-ESB-2008-00023. 

Because of the similar characteristics of the HTF design to the FTF design presented in the WSRC-STI2007-00184, the FTF infiltration rates are considered applicable. The design information being provided 

is for planning purposes sufficient to support evaluation of the closure cap as part of the integrated site 

conceptual model being evaluated in the HTF PA and is adopted from the detailed FTF closure cap report 

WSRC-STI-2007-00184 for HTF. The closure cap design will be finalized closer to the time of HTF 

closure, to take advantage of possible advances in materials and closure cap technology that could be 

used to improve the design. 

 

55 Discussion on Tanks 18 and 19 equipment is included for examples of DOE plans for grouting of in-tank equipment; however, 

Tanks 18 and 19 are located in the FTF at SRS and are not within the scope of this HTF 3116 Basis Document. 
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A detailed discussion on closure cap design is provided in Section 3.0 of the HTF PA. An overview of the 

general design features and figures from the HTF PA are provided below. [SRR-CWDA-2010-00128] 

Figure 2.6-1 presents the general design of the closure cap above a closed waste tank. Figure 2.6-2 

presents the closure cap footprint and the elevations of the closure cap surfaces and the grading plan. 

Figure 2.6-3 through Figure 2.6-5 present cross sections of the closure cap conceptual design. 

Figure 2.6-1: HTF Closure Cap General Concept 
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Figure 2.6-2: HTF Closure Cap Conceptual Design, Cap Footprint and Grading Cap 

[SRNL-ESB-2008-00023, Figure 1]
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Figure 2.6-3: HTF Closure Cap Conceptual Design, Sections A-A and B-B 

NOTE: Vertical scale of sections has been exaggerated five times in order to show all closure cap layers. 

[SRNL-ESB-2008-00023, Figure 2] 

Feet Above MSL Feet Above MSL 
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Figure 2.6-4: HTF Closure Cap Conceptual Design, Sections C-C and D-D 

NOTE: Vertical scale of sections has been exaggerated five times in order to show all closure cap layers. 

See legend on Figure 2.6-3.

[SRNL-ESB-2008-00023, Figure 3] Feet Above MSL Feet Above MSL 
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Figure 2.6-5: HTF Closure Cap Conceptual Design, Sections E-E and F-F 

NOTE: Vertical scale of sections has been exaggerated five times in order to show all closure cap layers. 

See legend on Figure 2.6-3. 

[SRNL-ESB-2008-00023, Figure 3]

2.6.1 Function of Closure Cap Layers 

The HTF conceptual closure cap design has a 2 % maximum surface slope that is less than 585 feet in 

length. Therefore, the calculations for the FTF conceptual closure cap design documented in WSRC-STI2007-00184 are applicable for the HTF. 

Using the 585 feet maximum slope length and 2 % maximum slope, initial infiltration estimates through 

the conceptual closure cap case were made utilizing the Hydrologic Evaluation of Landfill Performance 

(HELP) Model. Based upon the initial estimates, detailed water balances were produced. Table 2.6-1 

presents the pertinent closure cap layers for HTF modeling and the resulting average initial infiltration 

rate. Table 2.6-2 summarizes the function of each of these layers. Feet Above MSL Feet Above MSL 
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Details on the input development required for the HELP modeling are provided in WSRC-STI-2007-

00184. For the purposes of this modeling, synthetic daily weather data for precipitation, temperature, and 

solar radiation over 100 years were generated based upon the HELP data for Augusta, Georgia, and 

modified with SRS-specific average monthly precipitation and temperature data reported in WSRC-STI2007-00184. 

 

Table 2.6-1: HTF Closure Cap Layers 

Parameter Configuration 

Layer (depth) Vegetative Cover (NA) 

Layer (depth) Topsoil (6 inches) 

Layer (depth) Upper Backfill (30 inches) 

Layer (depth) Erosion Barrier (12 inches) [soil infill] 

Layer (depth) Geotextile Fabric (NA) 

Layer (depth) Middle Backfill (12 inches) 

Layer (depth) Geotextile Filter Fabric (NA) 

Layer (depth) Lateral Drainage Layer (12 inches) [soil infill] 

Layer (depth) Geotextile Fabric (NA) 

Layer (depth) High Density Polyethylene (HDPE) Geomembrane (0.06 inch) 

Layer (depth) Geosynthetic Clay Liner (GCL) (0.2 inch) 

Layer (depth) Upper Foundation Layer (12 inches) 

Layer (depth) Lower Foundation Layer (72 inches minimum) 

Average Infiltration Rate 0.00088 in/yr (through the GCL) 

Average Change in Water Storage 0.06 in/yr 

- Negligible 

NA - Not Applicable 

[WSRC-STI-2007-00184, Table 11]

 
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Table 2.6-2: Function of the HTF Closure Cap Layers 

Layer Function 

Vegetative 

Cover 

The vegetative cover will promote runoff, minimize erosion, and promote evapotranspiration. 

The initial vegetative cover will be a persistent grass such as Bahia. If it is determined that 

bamboo is a climax species that prevents or greatly slows the intrusion of pine trees, bamboo 

will be planted as the final vegetative cover at the end of the 100-year institutional control period. 

Bamboo is not assumed in present design calculations and modeling. 

Topsoil The topsoil is designed to support a vegetative cover, promote runoff, prevent the initiation of 

gullying, and provide water storage for the promotion of evapotranspiration. 

Upper Backfill The upper backfill is designed to increase the elevation of the closure cap to that necessary for 

placement of the topsoil and to provide water storage for the promotion of evapotranspiration. 

Erosion Barrier The erosion barrier is designed to prevent riprap movement during a probable maximum 

precipitation event and therefore form a barrier to further erosion and gully formation (i.e., 

provide closure cap physical stability). It is used to maintain a minimum 10 feet of clean material 

above the tanks and significant ancillary equipment to act as an intruder deterrent. It also 

provides minimal water storage for the promotion of evapotranspiration. 

Geotextile 

Fabric 

This geotextile fabric is designed to prevent the penetration of erosion barrier stone into the 

underlying middle backfill and to prevent piping of the middle backfill through the erosion barrier 

voids. 

Middle Backfill The middle backfill provides water storage for the promotion of evapotranspiration in the event 

that the topsoil and upper backfill are eroded away since the overlying erosion barrier provides 

only minimal water storage. 

Geotextile 

Filter Fabric 

This geotextile fabric is designed to provide filtration between the overlying middle backfill layer 

and the underlying lateral drainage layer. This filtration allows water to freely flow from the 

middle backfill to the lateral drainage layer while preventing the migration of soil from the middle 

backfill to the lateral drainage layer. 

Lateral 

Drainage 

Layer

The lateral drainage layer is a coarse sand layer designed to: 

 divert infiltrating water away from the underlying tanks and ancillary equipment and transport 

the water to the perimeter drainage system, in conjunction with the underlying composite 

hydraulic barrier (i.e., HDPE geomembrane and GCL) and 

 provide the necessary confining pressures to allow the underlying GCL to hydrate properly. 

Geotextile 

Fabric 

This geotextile fabric is a nonwoven geotextile fabric designed to protect the underlying HDPE 

geomembrane from puncture or tear during placement of the overlying lateral drainage layer. 

HDPE 

Geomembrane 

The HDPE geomembrane forms a composite hydraulic barrier in conjunction with the GCL. The 

composite hydraulic barrier is designed to promote lateral drainage through the overlying lateral 

drainage layer and minimize infiltration to the tanks and ancillary equipment. 

GCL The GCL forms a composite hydraulic barrier described above in conjunction with the HDPE 

geomembrane. As part of the composite hydraulic barrier the GCL is designed to hydraulically 

plug any holes that may develop in the HDPE geomembrane. 

Upper 

Foundation 

Layer 

Lower 

Foundation 

Layer

The foundation layers are designed to: 

 provide structural support for the rest of the overlying closure cap, 

 produce the required contours and produce a slope of 2 % for the overlying layers, 

 produce the maximum 3:1 side slopes of the closure cap, 

 provide a suitable surface for installation of the GCL (i.e., a soil with a moderately low 

permeability and a smooth surface, free from deleterious materials), 

 promote drainage of infiltrating water away from and around the tanks and ancillary 

equipment, and 

 contain utilities, equipment, facilities, etc., that are not removed from above current grade 

prior to installation of the closure cap. 

[WSRC-STI-2007-00184, Table 12] 
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3.0 SECTION 3116 OF THE RONALD W. REAGAN NATIONAL DEFENSE AUTHORIZATION ACT 

FOR FISCAL YEAR 2005 

The NDAA Section 3116(a) provides that radioactive waste that results from reprocessing is not “highlevel radioactive waste” if the Secretary of Energy determines, in consultation with the NRC, that the 

waste meets certain specified criteria. 

The NDAA Section 3116(a) provides in pertinent part: 

In General – Notwithstanding the provisions of the Nuclear Waste Policy Act of 1982, the 

requirements of section 202 of the Energy Reorganization Act of 1974, and other laws that define 

classes of radioactive waste, with respect to material stored at a Department of Energy site at 

which activities are regulated by a covered State pursuant to approved closure plans or permits 

issued by the State, the term “high-level radioactive waste” does not include radioactive waste 

resulting from the reprocessing of spent nuclear fuel that the Secretary of Energy (in this section 

referred to as the “Secretary”), in consultation with the Nuclear Regulatory Commission (in this 

section referred to as the “Commission”), determines – 

(1) does not require permanent isolation in a deep geologic repository for spent fuel or highlevel radioactive waste; 

(2) has had highly radioactive radionuclides removed to the maximum extent practical; and 

(3) (A) does not exceed concentration limits for Class C low-level waste as set out in section 

61.55 of title 10, Code of Federal Regulations, and will be disposed of – 

(i) in compliance with the performance objectives set out in subpart C of part 61 of title 

10, Code of Federal Regulations; and 

(ii) pursuant to a State-approved closure plan or State-issued permit, authority for the 

approval or issuance of which is conferred on the State outside of this section; or 

 (B) exceeds concentration limits for Class C low-level waste as set out in section 61.55 of 

title 10, Code of Federal Regulations, but will be disposed of – 

(i) in compliance with the performance objectives set out in subpart C of part 61 of title 

10, Code of Federal Regulations; 

(ii) pursuant to a State-approved closure plan or State-issued permit, authority for the 

approval or issuance of which is conferred on the State outside of this section; and 

(iii) pursuant to plans developed by the Secretary in consultation with the Commission. 
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4.0 WASTE DOES NOT REQUIRE PERMANENT ISOLATION IN A DEEP GEOLOGIC 

REPOSITORY FOR SPENT FUEL OR HIGH-LEVEL RADIOACTIVE WASTE 

The NDAA Section 3116(a) provides in pertinent part: 

 [T]he term “high-level radioactive waste” does not include radioactive waste resulting from the 

reprocessing of spent nuclear fuel that the Secretary of Energy..., in consultation with the Nuclear 

Regulatory Commission..., determines – 

(1) does not require permanent isolation in a deep geologic repository for spent fuel or highlevel radioactive waste. 

Under NDAA Section 3116(a), certain wastes from reprocessing are not “high-level radioactive waste” if 

the Secretary of Energy, in consultation with the NRC, determines that certain criteria are met. The 

NDAA Section 3116(a) sets out three criteria. Criterion (2), which is set forth in NDAA Section 3116(a)(2), 

requires removal of highly radioactive radionuclides to the maximum extent practical. Criterion (3) 

generally mirrors the regulatory criteria that the NRC has established for determining whether waste 

qualifies for land disposal as low-level waste. That criterion provides that disposal of the waste must 

meet the NRC performance objectives at 10 CFR Part 61, Subpart C, that the waste must not exceed 

concentration limits for Class C waste in 10 CFR 61.55 or must be disposed of pursuant to plans 

developed by the Secretary in consultation with NRC, and that disposal must be pursuant to a Stateapproved closure plan or permit. Criteria (2) and (3) will be discussed in subsequent sections of this HTF 

3116 Basis Document, which demonstrate that those criteria are satisfied. 

Criterion (1), quoted above, is a broader criterion that allows the Secretary of Energy, in consultation with 

the NRC, to consider whether there are other considerations that, in the Secretary of Energy’s judgment, 

warrant permanent isolation of the radioactive waste in a deep geologic repository. Generally, such 

considerations would be an unusual case because waste that meets the third criterion would be waste 

that will be disposed of in a manner that meets the 10 CFR 61, Subpart C performance objectives and 

either falls within one of the classes set out in 10 CFR 61.55 that the NRC has specified are considered 

“generally acceptable for near-surface disposal” or for which the Secretary of Energy has consulted with 

NRC concerning DOE’s disposal plans.56 Normally, it follows that if disposal of a waste stream in a facility 

that is not a deep geologic repository will meet these objectives, in the ordinary case, that waste does not 

“require permanent isolation in a deep geologic repository” because non-repository disposal will be 

protective of public health and safety. 

However, it is possible that in rare circumstances a waste stream that meets the third criterion might have 

some other unique radiological characteristic or may raise unique policy considerations that warrant its 

disposal in a deep geologic repository. Clause (1) of NDAA Section 3116(a) is an acknowledgment by 

Congress of that possibility. For example, the waste stream could contain material that, while not 

presenting a health and safety danger if disposed of at the near- or intermediate-surface, nevertheless 

presents non-proliferation risks that the Secretary concludes cannot be adequately guarded against 

absent deep geologic disposal.57 Criterion (1) allows the Secretary of Energy, in consultation with the 

NRC, to consider such factors in determining whether waste that meets the other two criteria may need 

permanent isolation in a deep geologic repository in light of these considerations. 

This is not the case here. As demonstrated later in this document, the stabilized residuals within the 

waste tanks and ancillary structures, the waste tanks, and the ancillary structures (including integral 

 

56 As the NRC explained in In the Matter of Louisiana Energy Services, L.P. (National Enrichment Services), [CLI-05-05], the 10 

CFR 61, Subpart C performance objectives in turn “set forth the ultimate standards and radiation limits for (1) protection of the 

general population from releases of radioactivity; (2) protection of individuals from inadvertent intrusion; (3) protection of individuals 

during operations; and (4) stability of the disposal site after closure.” 57 In NUREG-1854, NRC Staff Guidance for Activities Related to U.S. Department of Energy Waste Determinations, the NRC 

similarly explains: “In general, there is reasonable assurance that this criterion can be met if the two other NDAA criteria can be met. 

In other words, if highly radioactive radionuclides have been removed to the maximum extent practical and the waste will be 

disposed of in compliance with the performance objectives in 10 CFR Part 61, Subpart C (which are the same performance 

objectives NRC uses for disposal of low-level waste), then this supports a conclusion that the waste does not require disposal in a 

deep geologic repository. However, this criterion allows for the consideration that waste may require disposal in a geologic 

repository even though the two other NDAA criteria may be met. Those circumstances under which geologic disposal is warranted 

to protect public health and safety and the environment could be considered; for example, unique radiological characteristics of 

waste or nonproliferation concerns for particular types of material.” 
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equipment) located at HTF at the time of closure will meet the performance objectives of 10 CFR 61, 

Subpart C so as to provide for the protection of the public health and the environment. Accordingly, the 

waste does not require disposal in a deep geologic repository due to the risk to public health and safety. 

Furthermore, the stabilized residuals within the waste tanks and ancillary structures, the waste tanks, and 

the ancillary structures (including integral equipment) do not raise any unique considerations that, 

notwithstanding these demonstrations, require permanent isolation in a deep geologic repository. 

Accordingly, the stabilized residuals within the waste tanks and ancillary structures, the waste tanks, and 

the ancillary structures (including integral equipment), meet the criterion of Clause (1) of NDAA Section 

3116(a).58 

 

58 On page xv of the TER, the NRC concludes that, “The NRC staff notes in Chapter 2 of this TER that it believes that DOE can 

meet NDAA Criterion 1 provided Criteria 2 and 3 are demonstrated for the HTF. That is, the cleaned HTF tanks will not require 

exhumation and disposal in a geologic repository.” Similarly, on page 2-1 the NRC states, “Provided that DOE can meet the 

remaining NDAA criteria (Criteria 2 and 3), there is no indication that other considerations would warrant disposal of the waste in 

geologic repository because there appears to be no special properties of the waste and there are no proliferation concerns. The

NRC staff believes that DOE can meet Criterion 1 for the waste at HTF.” Likewise, in the Transmittal Letter transmitting the TER to 

DOE, NRC states, “With regard to Criterion 1, provided that the DOE can meet the remaining NDAA criteria (Criteria 2 and 3), there 

is no indication that other considerations would warrant the disposal of the waste in a geologic repository.” [ML14094A496] 
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5.0 WASTE HAS HAD HIGHLY RADIOACTIVE RADIONUCLIDES REMOVED TO THE MAXIMUM 

EXTENT PRACTICAL 

Section Purpose 

The NDAA Section 3116(a) provides that certain waste resulting from reprocessing is not 

high-level waste if the Secretary of Energy, in consultation with the NRC, determines, 

among other things, that the waste has had HRRs removed “to the maximum extent 

practical”. This section demonstrates that the HTF residual waste, tanks and ancillary 

structures, upon completion of waste removal activities at closure, will have had HRRs 

removed to the MEP and meet this criterion. 

Section Contents 

Section 5.1 identifies the HRRs for the purpose of this HTF 3116 Basis Document. 

Section 5.2 describes the removal processes used to remove HRRs to the MEP. 

Section 5.3 demonstrates that, at closure, the HRRs will have been removed to the MEP. 

Key Points 

 The list of HRRs for HTF describes the radionuclides that could reasonably be expected 

to exist in the HTF waste tanks and ancillary structures and that, using a risk-informed 

approach, contribute significantly to the radiological risk to workers, the public and the 

environment, taking into account scientific principles, knowledge and expertise. 

 The list of HRRs for HTF includes all radionuclides important to meeting the 

performance objectives in 10 CFR Part 61, Subpart C, and all radionuclides in Tables 1 

and 2 of 10 CFR 61.55 were considered. 

 Cleaning methodologies are expected to collectively remove approximately 99 % of 

HRRs, based on a starting point of the maximum historical radionuclide inventory in 

the overall HTF, although individual waste tanks or ancillary structures may not 

achieve this level of HRR removal on an individual basis.

The NDAA Section 3116(a) provides in pertinent part: 

 [T]he term “high-level radioactive waste” does not include radioactive waste resulting from the 

reprocessing of spent nuclear fuel that the Secretary of Energy ..., in consultation with the 

Nuclear Regulatory Commission ..., determines — ... 

(2) has had highly radioactive radionuclides removed to the maximum extent practical. 

5.1 Highly Radioactive Radionuclides 

5.1.1 Methodology 

Based on consultation with the NRC, DOE views “highly radioactive radionuclides” to be those 

radionuclides that, using a risk-informed approach, contribute most significantly to radiological risk to 

workers, the public and the environment. Strontium-90, Tc-99, I-129, Cs-137, U-233, U-234, U-235, Np237, Pu-238, Pu-239, Pu-240, Am-241 and Am-243 are the HRRs in the HTF stabilized residuals, HTF 

waste tanks and HTF ancillary structures at the closure of HTF that DOE has determined contribute 

significantly to radiological risk to workers, the public and the environment, taking into account scientific 

principles, knowledge and expertise.59

The list of HRRs, Table 5.1-1, was developed beginning with an initial listing of 849 radionuclides 

compiled from a variety of published resources (e.g., NCRP information), and included radionuclides not 

necessarily present in the projected HTF inventory at closure. [CBU-PIT-2005-00228] The initial listing of 

 

59 Some of the radionuclides listed as HRRs in this HTF 3116 Basis Document may not be listed in other NDAA Section 3116 basis 

documents if such radionuclides are not present in the waste or do not contribute significantly to dose to the worker, the public or the 

inadvertent intruder. 
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849 radionuclides included those radionuclides from Tables 1 and 2 of 10 CFR 61.55.60 DOE reviewed 

this initial list and identified those radionuclides that were present in the waste and may be important in 

meeting performance objectives in 10 CFR Part 61, Subpart C because they contribute to the dose to the 

workers, the public and/or the inadvertent intruder (for one or more reasonable intruder scenarios) in the 

HTF PA Base Case and sensitivity and uncertainty analyses. In DOE’s view, this approach results in a 

risk-informed list of HRRs that includes: 

those short-lived radionuclides that may 

present risk because they produce 

radiation emissions that, without 

shielding or controls, may harm humans 

simply by proximity to humans without 

inhalation or ingestion; and those longlived radionuclides that persist well into 

the future, may be mobile in the 

environment or may pose a risk to 

humans if inhaled or ingested.61

The list of HTF HRRs in Table 5.1-1 

account for approximately 99 % of the 

current radioactivity in the HTF waste. 

The short-lived fission products Cs-137 

and Sr-90 and their equilibrium daughter 

products, Ba-137m and Y-90, are by far 

the predominant sources of radioactivity 

present in the HTF waste today. Cs-137, 

and its daughter Ba-137m, are typically 

considered as a single radionuclide for 

human health protection purposes 

because the half-life of Ba-137m is so 

short that it only exists when Cs-137 is 

present. The same is true for Sr-90 and 

its daughter Y-90. Accordingly, the 

discussions that follow in this HTF 3116 

Basis Document focus on Cs-137 or Sr90 since approaches that are effective in removing Cs-137 and Sr-90 also remove Ba-137m and Y-90, 

respectively. Approximately 96 % of the current radioactivity in the HTF waste is associated with these 

two radionuclides and their short-lived daughters. [SRR-LWP-2014-00034] Moreover, Cs-137, Sr-90 and 

their daughters are present in sufficient concentrations in the HTF waste that, without shielding and 

controls, they produce radiation emissions that would present risk to humans simply due to their proximity 

 

60 Although Tables 1 and 2 in 10 CFR 61.55 specify concentration limits for certain radionuclides in the form of activated metal, DOE 

includes such radionuclides without regard to whether such radionuclides are in the form of activated metal. 61 On page 3-55 of the TER, NRC states that, “DOE’s process for identification of HRRs is reasonable.” The NRC notes on page 

xvii, under Key Recommendation ES-2.2, and similarly on page 3-55, that, “The NRC staff recommends that DOE continue to 

evaluate its HRR list and provide sufficient justification for any changes as additional information becomes available. The HRR list 

should be evaluated especially where it is used to inform decisions, such as the selection of radionuclides characterized in residual 

waste, selection of treatment technologies, and the screening of radionuclides for the purpose of detailed PA calculations.” 

Additionally, on page xviii, under Key Recommendation ES-2.3, and similarly on page 3-55, NRC states, “The NRC staff 

recommends that DOE assess, through future tank residual characterization, the validity of prior assumptions and the resulting 

impacts to the list of HRRs.” [ML14094A496] DOE agrees that the list of HRRs is reasonable as demonstrated in this HTF 3116 

Basis Document and supporting references. In DOE’s view, continued confirmation of the list of HRRs following a NDAA Section 

3116(a) Secretarial determination is not required by NDAA Section 3116(a) and is not part of NRC’s statutorily-prescribed 

monitoring, in coordination with the State of South Carolina, under NDAA Section 3116(b). Nevertheless, DOE will continue to 

characterize and confirm the actual residuals after cleaning, with an emphasis on HRRs, for the waste tanks and ancillary 

structures, as described in this HTF 3116 Basis Document (including Appendix B) and supporting references. DOE also will 

continue to incorporate actual residual waste characterization information, and explore associated refinement of assumptions, in 

special analyses and future revisions of the HTF PA as part of the HTF PA maintenance and monitoring processes under DOE 

Manual 435.1-1, pursuant to DOE responsibilities under the Atomic Energy Act of 1954, as amended. 

Table 5.1-1: HTF Highly Radioactive Radionuclides 

Radionuclide 

Radionuclide 

Half-Life (yr) 

Potential 

Long-Term 

Radiological 

Hazards 

Potential 

Short-Term 

Radiological 

Hazards 

Sr-90 b, c, d, f 2.89E+01 X 

Tc-99 a, b, c, e 2.11E+05 X 

I-129 a, b, e 1.57E+07 X 

Cs-137 c, d, f 3.00E+01 X 

U-233 b

 1.59E+05 X 

U-234 a, b, c 2.46E+05 X 

U-235 a

 7.04E+08 X 

Np-237 a, c, e 2.14E+06 X 

Pu-238 a, b, e 8.77E+01 X 

Pu-239 b, c, e 2.41E+04 X 

Pu-240 b, c, e 6.56E+03 X 

Am-241a, b, c, e 4.32E+02 X 

Am-243 b, e 7.37E+03 X 

a

 HRRs based on groundwater analyses results from the HTF PA. [SRRCWDA-2010-00128] b

 HRRs based on intruder analysis results from the HTF PA. [SRR-CWDA2010-00128] c

 HRRs based on uncertainty and sensitivity run results from the HTF PA. 

[SRR-CWDA-2010-00128] d

 HRRs based on potential contribution to worker dose. e Included in Table 1 of 10 CFR 61.55. f Included in Table 2 of 10 CFR 61.55. 
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without direct inhalation or ingestion. Accordingly, they are of potential acute hazard to occupational 

workers, the public and the environment. 

The remainder of the radionuclides listed are the long-lived isotopes that may pose the greatest risk in the 

future to human health because of their long life and because they present human health risk if directly 

inhaled or ingested. These long-lived isotopes combined account for less than 1 % of the current 

radioactivity in the HTF waste today.62 [SRR-LWP-2014-00034] 

5.1.2 Performance Assessment Radionuclides 

As explained above, DOE has included in the list of HRRs those radionuclides that may be important to 

meeting the performance objectives of 10 CFR 61, Subpart C because they contribute to the dose to 

workers, the public and/or the inadvertent intruder based on the HTF PA, which includes sensitivity and 

uncertainty analyses. The HTF PA applied a rigorous, documented, multi-step, multi-factor screening 

methodology (including consideration of NCRP information) to 849 radionuclides compiled from 

radionuclides present in the HTF (based on the SRS Waste Characterization System, process knowledge 

and available sampling data) as well as radionuclides reported in a variety of published resources. 

Specifically, DOE used the following approach in the HTF PA to focus on those radionuclides that 

contribute to the dose for various pathways. [SRR-CWDA-2010-00128] 

For the purpose of determining which radionuclides should be evaluated in the HTF PA, an initial 

radionuclide screening process was developed and performed, evaluating an initial list of 849 

radionuclides compiled from a variety of published resources including the following: 

 Screening Models for Releases of Radionuclides to Atmosphere, Surface Water and Ground, 

Volumes I and II, National Council on Radiation Protection and Measurements, which identifies 

826 nuclides of interest in determining radiation exposure due to releases to the air, water and 

ground. [NCRP-123] 

 The EPA Risk Assessment Web Site, which provides conversion factors for 824 nuclides which 

are of interest in determining human health cancer risk from radionuclides in the air, soil and 

water. [CBU-PIT-2005-00228] 

 Fission-Product Yields from Neutron-Induced Fission, Nucleonic - Reference Data Manual, which 

details 148 radionuclides produced from thermal neutron induced fission of U-235. 

[Nucleonics_1960] 

 Integrated Data Base Report-1996: U.S. Spent Nuclear Fuel and Radioactive Waste Inventories, 

Projections and Characteristics, DOE, which identified 168 nuclides of issue in making material 

disposition decisions. [DOE/RW-0006] 

 Derivation of Initial Radionuclide Inventories for the Safety Assessment of the Disposal of Used 

CANDU Fuel, Atomic Energy of Canada Limited, which lists 211 nuclides of interest in making 

decisions about disposal of spent fuel from heavy water reactors (note that the five reactors 

operated at SRS were also heavy water reactors). [AECL-9881] 

This initial screening process considered the following information: 

 physical properties of each radionuclide such as half-life and decay mechanism, 

 waste source and handling based on radionuclide production mechanisms and time since the 

radionuclide was produced, and 

 screening factors for radionuclide ground disposal developed in Screening Models for Releases 

of Radionuclides to Atmosphere, Surface Water and Ground [NCRP-123] which convert a 

quantity of each radionuclide to a dose. 

This initial screening reduced the radionuclide list from 849 down to 159 radionuclides. [CBU-PIT-2005-

00228] 

Additional screening of the 159 radionuclides was performed to identify the radionuclides to be 

considered in the initial HTF PA inventory. The screening and adjustment included the following: 

 

62 The remaining radionuclides in the HTF waste in combination contribute approximately 3 % of the current radioactivity. 
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 Radionuclides were screened out if there were no ancestors present from the specific decay 

chain or no decay source for the radionuclide. 

 The HTF waste production history information was evaluated for the potential for a specific 

radionuclide to be present in HTF. This step screened out radionuclides not present within the 

HTF waste. 

 In general, radionuclides present due to ingrowth from a decay series were screened out, 

however, production history was used to retain those radionuclides present at a greater 

proportion than from the decay series (i.e., Np-237). This screening step eliminated radionuclides 

that are present only due to the decay of their parent radionuclide (i.e., Ba-137m, Y-90, Ra-226 

and Th-229). The inventory of these radionuclides can be controlled by removing the parent 

radionuclide(s) to the “maximum extent practical.” 

 Radionuclides with less than a five-year half-life were screened out. This screening reflects that 

active institutional control will be maintained over the site for 100 years after facility closure. The 

inventories of these radionuclides will be significantly diminished due to the amount of radioactive 

decay that will occur during the assumed 100-year institutional control period. 

Based on the above screening and adjustment, an additional 105 radionuclides were screened out, thus 

reducing the HTF PA modeling initial inventory radionuclide number to 54. For the waste tanks and 

ancillary structures, an initial inventory value for these 54 radionuclides was developed and was used as 

input into the HTF PA modeling.63 The results of the HTF PA analyses were then evaluated using the 

methodology described below to determine which of the 54 radionuclides would be considered HRRs for 

HTF. [SRR-CWDA-2010-00023, SRR-CWDA-2010-00128] 

5.1.3 Highly Radioactive Radionuclides Based on 100-Meter Groundwater Analysis (For Member 

of the Public Following Closure) 

To determine which radionuclides are HRRs, DOE considered the doses estimated in the HTF PA, as 

well as the NRC guidance in NUREG-1854, the NRC guidance in Volume 2 of NUREG-1757 (referenced 

in NUREG-1854), and recommendations by the NRC during previous consultation.64 The approach 

followed by DOE in this HTF 3116 Basis Document is identical to the approach followed in the FTF 3116 

Basis Document. It is a thorough and methodical approach that reflects NRC guidance and 

recommendations and results in the identification of those radionuclides that may provide more than an 

insignificant contribution to dose. [DOE/SRS-WD-2012-001] 

Several radionuclides have been included on the HRR list based on an evaluation of the HTF 100-meter 

groundwater dose, using the groundwater dose results calculated in the HTF PA. For the HTF PA, the 

100-meter point is the point of maximum exposure at, or outside of, the HTF 100-meter buffer zone. The 

pathways for release to a member of the public considered in the HTF PA analyses are discussed in 

Section 7.1.2. The groundwater analysis in the HTF PA utilized the initial inventory of 54 radionuclides 

resulting from the screening analysis described in the previous section as input for the HTF PA model. 

The model used to perform the groundwater analysis accounted for radioactive decay and ingrowth 

throughout the assessment period. [SRR-CWDA-2010-00128] The results of the groundwater analysis 

were then evaluated to determine the HRRs. 

DOE examined the resulting dose contributions to the groundwater analysis, shown in the HTF PA, from 

all individual radionuclides in the HTF PA initial inventory at the time of closure. Those radionuclides 

which, in aggregate, would not contribute greater than 1.25 mrem/yr were not considered HRRs. The first 

step in this groundwater evaluation was to review the resulting doses from the HTF PA groundwater 

analysis at any time within 20,000 years.65 The individual doses from each radionuclide were then listed 

 

63 On page 3-16 of the TER, NRC states that, “The NRC staff has reviewed the screening approach and finds the list of 

radionuclides for which DOE has developed inventories to be reasonable.” In addition, on page 3-55 of the TER, NRC notes that,

“DOE’s approach to developing inventories for tanks that have not been cleaned is reasonable and appears to be generally 

conservative (tends to over- rather than under-predict inventory).” [ML14094A496] 64 NRC suggested during a public Draft FTF 3116 Basis Document scoping meeting held on July 13-14, 2010 that DOE consider the 

guidance in NUREG-1757 in establishing these evaluation thresholds. [SRR-CWDA-2010-00091] 65 To ensure that a conservative approach is taken in selection of the HRRs, the calculation of groundwater dose for 20,000 years

was used to account for waste tank degradation for Type I, Type II and Type III/IIIA tanks. DOE is using a 20,000-year evaluation 

period for determination of HRRs due to the expected timing of tank liner failures and resultant peak doses. This tank liner failure 

analysis is specific to the SRS HTF Type I, Type II and Type III/IIIA tanks. 
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in order of contribution. DOE conducted a quantitative analysis to determine those radionuclides with an 

aggregate contribution to dose of less than or equal to 1.25 mrem/yr.66 These radionuclides were 

screened from the HRR list. Based on this evaluation, the remaining radionuclides – Ni-59, Tc-99, I-129, 

Ra-226, Pa-231 and Np-237 - were initially identified for consideration. For those radionuclides, the 

projected inventories (as shown in the HTF PA) at the time of HTF closure were reviewed. For those 

radionuclides with a relatively insignificant initial inventory (i.e., Ra-226 and Pa-231) the associated decay 

chains were examined. Reduction of these radionuclides is accomplished through the removal of the 

respective parent radionuclides, i.e., U-234 (for Ra-226), U-235 (for Pa-231), Pu-238 (for Ra-226), Pu239 (for Pa-231). An additional parent radionuclide, Am-241, was included based on progeny ingrowth of 

Np-237. 

Nickel-59 is one of the radionuclides identified in the initial screening described above. When listing the 

individual doses from each radionuclide in order of contribution, Ni-59 is the radionuclide which causes 

the aggregate dose to exceed 1.25 mrem/yr, Ni-59 causes the aggregate dose in this screening step to 

increase from 1.1 mrem/yr to 1.83 mrem/yr. The dose contribution from Ni-59 is attributed to the 

estimated Ni-59 residual inventory in the annuli of the Type I waste tanks. The estimated Ni-59 inventory 

in the Type I tank annuli was estimated by setting the annulus residual inventory equal to the primary tank 

residual inventory. The actual annulus inventories for Ni-59 in the Type I tanks is expected to be 

considerably lower than the inventory used in the HTF PA modeling. Based on this consideration, the 

associated dose from Ni-59, in aggregate with the previously screened radionuclides, are not expected to 

contribute greater than 1.25 mrem/yr to the groundwater dose. Additional conservatism for Ni-59 dose is 

also provided by the fact that the HTF PA PORFLOW model does not account for isotope dilution on 

solubility. In addition to the Ni-59 present in the tank waste, stable nickel is also present. The amount of 

stable nickel significantly surpasses the amount of Ni-59. The PORFLOW model treats each constituent 

individually, which ignores other constituents of the same element and their impact on solubility. There is 

more than two orders of magnitude more stable nickel than Ni-59. There would be a proportional 

decrease in the dose associated with Ni-59. 

A review of the decay chain for Pa-231 provides that both U-235 and Pu-239 could be potential 

contributors to progeny ingrowth of Pa-231. A review of the dose results shows that decay from the initial 

inventory of U-235 contributes greater than 85 % of the future Pa-231 inventory. Therefore, Pu-239, 

although a contributor to the future Pa-231 inventory, is not considered an HRR based on the 

groundwater pathway. 

DOE believes that the screening of radionuclides whose dose contribution in aggregate is less than or 

equal to 1.25 mrem/yr is sufficiently low, compared to the 25 mrem/yr all-pathways dose limit, to capture 

all risk-significant radionuclides in those that remain. For the purpose of this HTF 3116 Basis Document, 

Tc-99, I-129, U-234, U-235, Np-237, Pu-238 and Am-241 were included in the HRR list based on the 

groundwater pathway. 

5.1.4 Highly Radioactive Radionuclides Based on Air Pathway Analysis (For Member of the 

Public Following Closure) 

In a similar manner, radionuclides were evaluated for inclusion on the HRR list based on the HTF 100-

meter dose from airborne radionuclides, using the air pathway dose results calculated in the HTF PA. 

The pathways for release to a member of the public considered in the HTF PA analyses are discussed in 

Section 7.1.2 of this HTF 3116 Basis Document. In the HTF PA, radionuclides contained in the initial 

inventory that are susceptible to volatilization were considered in the air pathways analysis. These 

radionuclides included H-3, C-14, Cl-36, Se-79, Tc-99, Sb-125, Sn-126 and I-129. For the HTF PA, the 

100-meter point is the point of maximum exposure at or outside of the HTF 100-meter buffer zone. The 

HTF PA shows that the air pathway is not a significant contributor to dose, and contributes, in aggregate 

 

66 This approach is consistent with the guidance and general approach in Volume 2 of NUREG-1757, Consolidated 

Decommissioning Guidance (NUREG-1757), which explains that “NRC staff considers radionuclides and exposure pathways that 

contribute no greater than 10% of the dose criteria to be insignificant contributors.” The above-referenced NUREG, which applies to 

NRC licensees, is being used only as general guidance, and DOE’s use of this NUREG as guidance should not be construed to 

suggest that it is a requirement under NDAA Section 3116 or that the NUREG is applicable in the 3116 context. To ensure that 

selection of the HRRs is sufficiently conservative, DOE has used 5 percent (i.e., 1.25 mrem/yr) of the 25 mrem/yr all-pathways dose 

limit.
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less than 0.1 mrem/yr peak dose. [SRR-CWDA-2010-00128] Therefore, for the purpose of this HTF 

3116 Basis Document, no radionuclides were included in the HRR list based on the air pathway. 

5.1.5 Highly Radioactive Radionuclides Based on Intruder Pathway Analysis 

Several radionuclides have been included on the HRR list based on an evaluation of the HTF inadvertent 

intruder dose, using the intruder dose results calculated in the HTF PA. The intruder scenarios and 

associated exposure pathways used for this evaluation are discussed in Section 7.2.2 of this HTF 3116 

Basis Document. The intruder analysis in the HTF PA utilized the initial inventory of 54 radionuclides 

resulting from the screening analysis described previously as input for the HTF PA model. The model 

used to perform the intruder analysis did account for radioactive decay and ingrowth throughout the 

assessment period. [SRR-CWDA-2010-00128] The results of the intruder analysis were then evaluated 

to determine the HRRs. 

DOE examined the resulting dose contributions to the inadvertent intruder analysis, shown in the HTF PA, 

from all individual radionuclides in the HTF PA initial inventory at the time of closure. Those radionuclides 

which, in aggregate, would not contribute greater than 25 mrem/yr were not considered HRRs. The first 

step in this intruder evaluation was to review the resulting doses from the HTF PA intruder analysis at any 

time within 20,000 years.67 The individual doses from each radionuclide were then listed in order of 

contribution. DOE conducted a quantitative analysis to determine those radionuclides with an aggregate 

contribution to dose of less than or equal to 25 mrem/yr.68 These radionuclides were screened from the 

HRR list. Based on this evaluation the remaining radionuclides - Sr-90, Y-90, Tc-99, I-129, Ra-226, Th229, U-233, U-234, Pu-239, Pu-240, Pu-244, Am-241, Am-243, Cm-245, Cm-247, Cm-248 and Pb-210 - 

were initially identified for consideration. For those radionuclides, the projected inventories (as shown in 

the HTF PA) at the time of HTF closure were reviewed. For radionuclides with a relatively insignificant 

initial inventory (i.e., Ra-226, Th-229, Pb-210), the associated decay chain was examined. Reduction of 

these radionuclides is accomplished through the removal (during the cleaning and waste removal 

process) of the respective parent radionuclides, i.e., U-233 (for Th-229), U-234 (for Ra-226), Pu-238 (for 

Ra-226), Ra-226 (for Pb-210). 

For a number of radionuclides in the projected inventory with limited potential impact to dose, inventories 

were adjusted to either one curie or the analytical detection limit. If the radionuclide inventory was less 

than the detection limit, then it was adjusted to the detection limit. However, if the radionuclide inventory 

estimated was at least at the detection limit, then it was adjusted up to one curie. For the list of 

radionuclides identified in the initial screening of the HTF PA intruder doses described above Pu-244, 

Cm-245, Cm-247 and Cm-248 all had their projected HTF PA inventory set to one curie, at least one 

order of magnitude higher than the inventory based on estimates utilizing Tanks 5, 6, 18 and 19 sample 

analyses and WCS estimates (i.e., Pu-244 greater than two orders of magnitude, Cm-245 greater than 

one order of magnitude, Cm-247 greater than five orders of magnitude, Cm-248 greater than three orders 

of magnitude). [SRR-CWDA-2010-00023] Based on the HTF PA projected inventory for Pu-244, Cm245, Cm-247 and Cm-248, the associated dose from these radionuclides, in aggregate with the 

previously screened radionuclides, are not expected to contribute greater than 25 mrem/yr to the 

inadvertent intruder dose. 

Strontium-90, and its daughter Y-90, are typically considered as a single radionuclide for human health 

protection purposes because the half-life of Y-90 is so short that it only exists when Sr-90 is present. 

Removal of Y-90 is accomplished through removal of the parent radionuclide Sr-90 during the waste 

 

67 To ensure that a conservative approach is taken in selection of the HRRs, the calculation of inadvertent intruder dose for 20,000 

years was used to account for waste tank degradation for Type I, Type II and Type III/IIIA tanks. DOE is using a 20,000-year 

evaluation period for determination of HRRs due to the expected timing of tank liner failures and resultant peak doses. This tank 

liner failure analysis is specific to the SRS HTF Type I, Type II and Type III/IIIA tanks. 68 This approach is consistent with the guidance and general approach in Volume 2 of NUREG-1757, Consolidated 

Decommissioning Guidance (NUREG-1757), which explains that “NRC staff considers radionuclides and exposure pathways that 

contribute no greater than 10% of the dose criteria to be insignificant contributors.” The above-referenced NUREG, which applies to 

NRC licensees, is being used only as general guidance, and DOE’s use of this NUREG as guidance should not be construed to 

suggest that it is a requirement under NDAA Section 3116 or that the NUREG is applicable in the 3116 context. To ensure that 

selection of the HRRs is sufficiently conservative, DOE has used 5 percent (25 mrem/yr) of the 500 mrem/yr all-pathways dose 

recommended in NRC guidance. [NUREG-0945 and NUREG-1854] 
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removal process. Therefore, Y-90 is not considered an HRR based on contribution to the inadvertent 

intruder dose. 

DOE believes the screening of radionuclides whose dose contribution, in aggregate, is less than or equal 

to 25 mrem/yr is sufficiently low, compared to the 500 mrem/yr peak intruder dose recommended in NRC 

guidance (NUREG-0945 and NUREG-1854), to capture all risk significant radionuclides in those that 

remain. For the purpose of this HTF 3116 Basis Document, Sr-90, Tc-99, I-129, U-233, U-234, Pu-239, 

Pu-240, Am-241 and Am-243 were included in the HRR list based on the intruder pathway. 

5.1.6 Highly Radioactive Radionuclides Based on Uncertainty and Sensitivity Analyses 

Some radionuclides have been included on the HRR list based on an evaluation of the uncertainty and 

sensitivity analyses included in the HTF PA. [SRR-CWDA-2010-00128] The HTF PA uncertainty and 

sensitivity analyses were reviewed to identify those radionuclides shown to have the most influence on 

the model results. 

The purpose of the uncertainty and sensitivity analyses was to consider the effects of uncertainties in the 

conceptual models used and sensitivities in the parameters used in the mathematical models. While the 

uncertainty and sensitivity analyses were primarily performed using a probabilistic model, some additional 

single parameter sensitivity analyses were performed through deterministic modeling. The probabilistic 

model allows for variability of multiple parameters simultaneously, so concurrent effect of changes in the 

model can be analyzed. The deterministic model single parameter analysis provides a method to 

evaluate the importance of the uncertainty around a single parameter of concern. Using both probabilistic 

and deterministic models for sensitivity analysis versus a single approach provides additional information 

concerning which parameters are of most importance to the HTF PA model. In addition, as part of the 

sensitivity analyses, modeling was performed for a period of 100,000 years. [SRR-CWDA-2010-00128] 

The HTF PA considers the uncertainties and sensitivities associated with the projected dose results to a 

member of the public through uncertainty analysis of the HTF probabilistic model, through sensitivity 

analysis using the HTF probabilistic model and through sensitivity analysis using the HTF deterministic 

model. A review of the uncertainty analysis realizations with the highest peak doses show Sr-90 and Tc99 to be significant to dose. A review of the sensitivity analysis performed using the HTF probabilistic 

model shows Tc-99 and Ra-226 to be significant. A review of the sensitivity analysis performed using the 

HTF deterministic model, including a barrier analysis, show Tc-99, I-129, Ra-226 and Np-237 to be 

significant to dose. For Ra-226, which has a relatively insignificant initial inventory, the associated decay 

chain was examined. Reduction of Ra-226 is accomplished through the removal of its parent 

radionuclides U-234 and Pu-238 during the waste removal process. An additional parent radionuclide, 

Am-241, was included for progeny ingrowth of Np-237. Based on an evaluation of the HTF PA results for 

these different analyses, Sr-90, Tc-99, I-129, U-234, Np-237, Pu-238 and Am-241 were included in the 

HRR list. 

The HTF PA also considered the effects on the Intruder Analysis of uncertainties in the conceptual 

models used and sensitivities in the parameters used in the mathematical models. Deterministic intruder 

sensitivity analyses performed identified Ra-226 and U-234 as potentially significant to the inadvertent 

intruder results. For Ra-226, which has a relatively insignificant initial inventory, the associated decay 

chain was examined. Reduction of Ra-226 is accomplished through the removal of its parent 

radionuclides U-234 and Pu-238 during the waste removal process. A review of the HTF intruder 

sensitivity analyses identified Sr-90, Y-90, Tc-99, Cs-137 and Ba-137m as potentially significant to the 

inadvertent intruder results. Cesium-137, and its daughter Ba-137m, are typically considered as a single 

radionuclide for human health protection purposes because the half-life of Ba-137m is so short that it only 

exists when Cs-137 is present. The same is true for Sr-90 and its daughter Y-90. Removal of Ba-137m 

and Y-90 is accomplished through removal of the parent radionuclides Cs-137 and Sr-90, respectively, 

during the waste removal process. Therefore, Ba-137m and Y-90 are not considered HRRs based on an 

evaluation of the uncertainty and sensitivity analyses. Strontium-90, Tc-99, Cs-137, U-234 and Pu-238 

were included in the HRR list based on an evaluation of the results for the HTF PA inadvertent intruder 

uncertainty and sensitivity analyses.

Based on the HTF PA uncertainty and sensitivity analyses, Sr-90, Tc-99, I-129, Cs-137, U-234, Np-237, 

Pu-238 and Am-241 were included in the HRR list. 
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5.1.7 Highly Radioactive Radionuclides Summary 

The results of the HRR evaluation are summarized in Table 5.1-2. The table provides the results of the 

evaluation for each of the 54 radionuclides contained in the initial HTF PA inventory. Radionuclides were 

considered HRRs based on the evaluations and considerations discussed above. Based on these 

evaluations and considerations, Sr-90, Tc-99, I-129, Cs-137, U-233, U-234, U-235, Np-237, Pu-238, Pu239, Pu-240, Am-241 and Am-243 are considered the HRRs in the HTF stabilized residuals, HTF waste 

tanks and HTF ancillary structures at the closure of HTF. 

Table 5.1-2: Radionuclides Contained in HTF PA Initial Inventory Radionuclide with Initial Inventory Half-Life (yr) 

Evaluation Resultsa

Groundwater 

Analysis 

Air Pathway 

Analysis 

Intruder 

Analysis 

Dose > 1.25 mrem/yr Impact on Progeny > 1.25 mrem/yr Dose > 1.25 mrem/yr Impact on Progeny > 1.25 mrem/yr Dose > 25.0 mrem/yr Impact on Progeny > 25.0 mrem/yr Uncertainty & Sensitivity Analysis Contribution to Worker Dose 

H-3c 1.23E+01 x x x x x x x x 

C-14b 5.70E+03 x x x x x x x x 

Al-26 7.17E+05 x x x x x x x x 

Cl-36 3.01E+05 x x x x x x x x 

K-40 1.25E+09 x x x x x x x x 

Ni-59b

 7.60E+04 x x x x x x x x 

Ni-63c

 1.00E+02 x x x x x x x x 

Co-60c

 5.30E+00 x x x x x x x x 

Se-79 2.95E+05 x x x x x x x x 

Sr-90c 2.89E+01 x x x x  x  

Y-90 7.31E-03 x x x x x x x x 

Zr-93 1.53E+06 x x x x x x x x 

Nb-93m 1.61E+01 x x x x x x x x 

Nb-94b

 2.03E+04 x x x x x x x x 

Tc-99b 2.11E+05  x x x  x  x 

Pd-107 6.50E+06 x x x x x x x x 

Sn-126 2.30E+05 x x x x x x x x 

Sb-126 1.24E+01 x x x x x x x x 

Sb-126m 1.92E+01 x x x x x x x x 

I-129b 1.57E+07  x x x  x  x 

Cs-135 2.36E+06 x x x x x x x x 

Cs-137c 3.00E+01 x x x x x x  

Ba-137m 4.85E-06 x x x x x x x x 

Sm-151 9.00E+01 x x x x x x x x 

Eu-152 1.35E+01 x x x x x x x x 

Eu-154 8.59E+00 x x x x x x x x 

 Included on the HRR list based on specific evaluation threshold. 

x Does not meet specific evaluation threshold. 

a

 Radionuclides considered HRRs if one or more evaluation thresholds are met. 

b

 Included in Table 1 of 10 CFR 61.55. 

c

 Included in Table 2 of 10 CFR 61.55. 

Note: HRRs for this HTF 3116 Basis Document are highlighted. 

[SRR-CWDA-2010-00128] 
Basis for Section 3116 Determination DOE/SRS-WD-2014-001 

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Page 5-9 

Table 5.1-2: Radionuclides Contained in HTF PA Initial Inventory (Continued) Radionuclide with Initial Inventory Half-Life (yr) 

Evaluation Resultsa

Groundwater 

Analysis 

Air Pathway 

Analysis 

Intruder 

Analysis 

Dose > 1.25 mrem/yr Impact on Progeny > 1.25 mrem/yr Dose > 1.25 mrem/yr Impact on Progeny > 1.25 mrem/yr Dose > 25.0 mrem/yr Impact on Progeny > 25.0 mrem/yr Uncertainty & Sensitivity Analysis Contribution to Worker Dose 

Pt-193 5.00E+01 x x x x x x x x 

Ra-226 1.60E+03 x x x x x x x x 

Ac-227 2.17E+01 x x x x x x x x 

Th-229 7.88E+03 x x x x x x x x 

Th-230 7.54E+04 x x x x x x x x 

Pa-231 3.27E+04 x x x x x x x x 

U-232 6.89E+01 x x x x x x x x 

U-233 1.59E+05 x x x x   x x 

U-234 2.46E+05   x x    x 

U-235 7.04E+08 x  x x x x x x 

U-236 2.34E+07 x x x x x x x x 

U-238 4.47E+09 x x x x x x x x 

Np-237b 2.14E+06  x x x x x  x 

Pu-238b 8.77E+01 x  x x x   x 

Pu-239b 2.41E+04 x x x x  x x x 

Pu-240b 6.56E+03 x x x x  x x x 

Pu-241b

 1.43E+01 x x x x x x x x 

Pu-242b

 3.75E+05 x x x x x x x x 

Pu-244b

 8.00E+07 x x x x x x x x 

Am-241b 4.32E+02 x  x x  x  x 

Am-242mb

 1.41E+02 x x x x x x x x 

Am-243b 7.37E+03 x x x x  x x x 

Cm-243b

 2.91E+01 x x x x x x x x 

Cm-244b 1.81E+01 x x x x x x x x 

Cm-245b

 8.50E+03 x x x x x x x x 

Cm-247b

 1.56E+07 x x x x x x x x 

Cm-248b

 3.48E+05 x x x x x x x x 

Cf-249b

 3.51E+02 x x x x x x x x 

 Included on the HRR list based on specific evaluation threshold. 

x Does not meet specific evaluation threshold. 

a

 Radionuclides considered HRRs if one or more evaluation thresholds are met. 

b

 Included in Table 1 of 10 CFR 61.55. 

c

 Included in Table 2 of 10 CFR 61.55. 

Note: HRRs for this HTF 3116 Basis Document are highlighted. 

[SRR-CWDA-2010-00128] 
Basis for Section 3116 Determination DOE/SRS-WD-2014-001 

for Closure of H-Tank Farm Revision 0 

at the Savannah River Site December 2014 

Page 5-10 

5.2 Removal of Highly Radioactive Radionuclides to the Maximum Extent Practical 

The NDAA Section 3116(a) provides that certain waste resulting from reprocessing is not high-level waste 

if the Secretary of Energy, in consultation with the NRC, determines, among other things, that the waste 

has had HRRs removed “to the maximum extent practical”.69 Section 5.2 and 5.3 demonstrates that the 

HTF residual waste, tanks and ancillary structures will have had HRRs removed to the MEP upon 

cessation of waste removal activities. Removal to the maximum extent “practical” is not removal to the 

extent theoretically “possible.” Rather, a “practical” approach to removal is one that is “adapted to actual 

conditions” (A Dictionary of Modern English Usage); “adapted or designed for actual use” 

(http://infoplease.com/ipd/A0598638.html); “useful” (http://infoplease.com/ipd/A0598638.html); selected 

“mindful of the results, usefulness, advantages or disadvantages, etc., of [the] action or procedure” 

(http://infoplease.com/ipd/A0598638.html); fitted to “the needs of a particular situation in a helpful way” 

(http://dictionary.cambridge.org/define.asp?key=practical*2+0&dict=A); “effective or suitable” 

(http://dictionary.cambridge.org/define.asp?key=practical*2+0&dict=A). Therefore, the determination as 

to whether a particular HRR will be removed to the MEP will vary from situation to situation, based not 

only on the available technologies but also on the overall costs and benefits70 of deploying a technology 

with respect to the conditions in a specific HTF waste tank or ancillary structure. The MEP standard 

contemplates room for exercising expert judgment in weighing several factors. Such factors may include, 

but are not limited to, environmental, health, timing or other exigencies; the risks and benefits to public 

health, safety and the environment arising from further HRR removal as compared with countervailing 

considerations that may ensue from not removing or delaying removal; the reasonable availability of 

proven technologies; the usefulness of such technologies; and the sensibleness of using such 

technologies. What may be removal to MEP in a particular situation or at one point in time may not be 

that which, on balance, is practical, feasible or sensible in another situation or at a prior or later point in 

time. 

Moreover, it may not be practical to undertake further removal of certain radionuclides because further 

removal is not sensible or useful in light of the overall benefit to human health and the environment. As a 

general matter, such a situation may arise if certain radionuclides are present in such extremely low 

quantities that they make an insignificant contribution71 to potential doses to workers, the public, and the 

hypothetical human intruder. 

The HRRs have been and, for tanks and equipment to be cleaned in the future, will be removed from HTF 

waste tanks and ancillary structures to the MEP for the purpose of removal from service72 and eventual 

closure of the waste tanks and ancillary structures. Removal of HRRs to the MEP in HTF waste tanks 

and ancillary structures occurs through a systematic progression of waste removal and cleaning activities 

using proven technologies to a point where further removal of HRRs is not sensible or useful in light of the 

overall benefit to human health, safety and the environment. 

 

69 The NDAA Section 3116 does not specify “remedial goals” or other numerical objectives, and does not require DOE to develop 

any such removal goals or objectives. 

70 While prior NRC and DOE requirements for waste determinations called for removal “to the maximum extent technically and 

economically practical” [NRC_03-02-93; DOE M 435.1-1], NDAA Section 3116 omits these adverbs, thereby suggesting that a broad 

range of considerations, including but not limited to technical and economic practicalities, may appropriately be taken into account in 

determining the extent of removal that is practical. 

71 The DOE normally would view radionuclides as making a clearly insignificant contribution if the contribution to dose from those 

radionuclides, in both the expected case and considering sensitivity analyses, does not exceed any of the following: (1) 10% of the 

25-mrem/yr all-pathways annual dose to the public, (2) 10% of the DOE 100-mrem annual dose limit to the intruder (under all 

reasonable intruder scenarios), (3) 10% of the DOE 500-mrem acute dose limit to the intruder (under all intruder scenarios), and 

(4) 10% of the annual worker dose in the relevant provisions of 10 CFR 20. This methodology is based on NRC consultation and is

intended to be consistent with the guidance and general approach in Volume 2 of NUREG-1757, Consolidated Decommissioning 

Guidance (NUREG-1757), which explains that “NRC staff considers radionuclides and exposure pathways that contribute no greater 

than 10% of the dose criteria to be insignificant contributors.” The above-reference NUREG, which applies to NRC licensees, is

being used only as general guidance, and DOE’s use of this NUREG as guidance should not be construed to suggest that it is a 

requirement under NDAA Section 3116 or that either the NUREG or 10 CFR 20, Subpart E is applicable in the 3116 context. 72 The term “removal from service” refers to the protocols set forth in the State-approved GCP, as described in Section 8.0 of this 

HTF 3116 Basis Document, to stabilize the waste tank or ancillary structure. 
Basis for Section 3116 Determination DOE/SRS-WD-2014-001 

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Page 5-11 

5.2.1 Current Status of H-Tank Farm Waste Removal Activities 

The HTF includes 29 waste storage tanks. The Type I, Type II and Type IV waste tanks in HTF are no 

longer actively receiving fresh canyon waste and the Type I and Type II tanks are currently undergoing 

waste removal activities. Tanks 12 and 16 have undergone extensive waste removal campaigns. [SRRCWDA-2011-00126, SRR-CWDA-2013-00125, SRR-CWDA-2014-00005] As of July 1, 2014, Tank 

residual samples from Tank 16 have been collected and analyses is ongoing. Preparation for sampling of 

Tank 12 residuals is ongoing. It is anticipated that Tank 16 and Tank 12 will be the first two waste tanks 

removed from service in HTF. Bulk waste removal efforts have been completed for Tank 11, and are 

currently underway in Tanks 10 and 13; preparation for bulk waste removal efforts in Tank 15 are in 

progress. Waste removal activities, for the purposes of final waste tank closure, have not been initiated in 

any of the 17 Type III/IIIA tanks or four Type IV tanks located in HTF. However, DOE has previously 

performed additional waste removal activities in various HTF waste tanks to support efficient 

management of ongoing tank farm operations and preparation of feed batches for DWPF. Bulk sludge 

removal activities have been performed in HTF Tanks 9, 10, 11, 14, 15, 21, 22 and 42. Bulk saltcake 

removal has been performed in HTF Tanks 10, 22, 24, 37 and 41. After completion of the waste removal 

activities in these tanks, the available storage space created within these tanks was subsequently re-used 

to support ongoing tank farm operations.73

As of July 1, 2014, HTF stored approximately 185,000,000 curies in approximately 21,700,000 gallons of 

waste. The sludge portion of this waste represents approximately 9 % of the volume but contains 

approximately 54 % of the radioactivity. Of the approximately 19,700,000 gallons of salt waste, 

approximately 7,180,000 gallons is in the form of saltcake with the remaining approximately 12,500,000 

gallons being concentrated supernate. The concentrated supernate accounts for approximately 43 % of 

the total radioactivity in the HTF. Figure 5.2-1 graphically presents the approximate breakdown of the 

waste in HTF in terms of both volume and curies for each of the three primary waste types. [SRR-LWP2014-00034, SRR-LWP-2014-00030]

Figure 5.2-1: HTF Waste Tank Composite Inventory 

5.2.2 Waste Removal Technologies 

The DOE has considered a large number of different technologies in recent years in its efforts to identify 

the best available technologies to remove waste from the SRS tanks, and a new generation of waste 

removal equipment has been selected for use at SRS. DOE has over 40 years of experience in 

successfully removing waste from the SRS waste tanks. This experience encompasses removal of all 

waste types (supernate, sludge and saltcake). 

 

73 Following this re-use, DOE is currently removing salt waste from Tank 10 as part of bulk waste removal efforts as discussed 

previously, and will eventually be initiating bulk waste removal efforts from the other re-used tanks. 

Volume

21.7 Mgal

Curies

84 MCi

(46 %)

101 MCi

(54 %)

185 MCi

79 MCi

(43 %)

Sludge

19.7 Mgal

(91 %)

2.0 Mgal

(9 %)

12.5 Mgal

(58 %)

HTF Inventory values as of July 1, 2014

Salt Supernate

5 MCi

(3 %) Saltcake

7.2 Mgal

(33 %)
Basis for Section 3116 Determination DOE/SRS-WD-2014-001 

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Page 5-12 

In 2003, DOE used a systematic process to identify, evaluate and select equipment for waste removal 

tasks to accelerate waste removal and the removal of waste tanks from service. This process is formally 

documented in a Systems Engineering Evaluation. The evaluation investigated options for bulk waste 

mixing, waste transfer and residual waste heel removal. The evaluation graded the options on weighted 

selection criteria such as technical maturity, effectiveness, reliability, reusability, radiological control 

requirements, integration with the tank farm system and cost. Knowledgeable tank farm operations, 

engineering, plant support and maintenance personnel identified potential technology candidates based 

on experience, literature, worldwide web research and contacts with other knowledgeable personnel in 

the DOE Complex and commercial industry. The team recommended using a combination of mechanical 

removal technologies and chemical removal technologies, if necessary, to perform waste removal in the 

tanks. [G-ESR-G-00051] 

In addition to the mechanical and chemical waste removal technologies, as the result of a March 2006 

DOE-sponsored Tank Cleaning Technical Exchange, DOE identified a new vacuum removal technology 

for heel removal applications. [CBU-PIT-2006-00067] 

DOE also conducted a new Systems Engineering Evaluation in 2012, which evaluated potential new 

technologies and technology enhancements, as documented in the appendix accompanying the CostBenefit Analysis for Removal of Additional Highly Radioactive Radionuclides from Tank 18. [SRR-CWDA2012-00026] This 2012 Systems Engineering Evaluation examined over 50 potential technologies and 

chemical cleaning methods, with four analyzed in detail, to identify any new or enhanced technologies to 

remove HRRs, as discussed further in Section 5.3 of this HTF 3116 Basis Document. 

The DOE is utilizing mechanical, chemical and vacuum heel removal technologies at SRS in various 

combinations and sequences depending on the unique characteristics of the waste and conditions in 

each tank. The DOE will continue to consider new technological developments relevant to waste tank 

cleaning, and has a detailed process in place to evaluate potential new technologies and technology 

optimization, with a focus on HRRs, as each tank, ancillary structure, or group is cleaned, as described in 

Appendix B to this HTF 3116 Basis Document. A range of potential technologies for evaluation will 

potentially include technologies developed and/or used at other DOE sites, in domestic commercial 

industry and in international applications.74 [SRR-CWDA-2014-00003] A recent example of how DOE 

evaluates potential technologies is documented in the Systems Engineering Evaluation accompanying 

the Cost-Benefit Analysis for Removal of Additional Highly Radioactive Radionuclides from Tank 18, 

discussed above and in Section 5.3. [SRR-CWDA-2012-00026] Another example of DOE’s ongoing 

review of new technologies is provided in Annual SCDHEC Technology Briefing. [SRR-LWE-2014-

00055] 

The following discusses the current technologies and how they are being utilized at SRS. 

5.2.2.1 Mechanical Cleaning 

Over the past several decades at HTF, DOE has successfully used, and is continuing to use, mechanical 

cleaning approaches for several types of waste tanks in HTF. These mechanical approaches are 

described below by HTF waste tank, for bulk waste removal and, depending on the waste tank, heel 

removal. 

 Tank 10 – bulk waste removal initiated using water dissolution to remove saltcake. 

 Tank 11 – bulk waste removal completed using several slurry pumps; awaiting heel removal. 

 Tank 12 – bulk waste removal completed using several slurry pumps; heel removal successfully 

completed using a combination of slurry pumps and chemical cleaning. 

 Tank 13 – bulk waste removal ongoing using several SMPs; heel removal will be completed after 

bulk waste removal. 

 Tank 15 – technical design currently underway for mechanical bulk waste removal; after 

completion of technical design, bulk waste removal and heel removal will be completed. 

 Tank 16 – bulk waste removal completed using several slurry pumps; heel removal successfully 

completed using a combination of slurry pumps and chemical cleaning. 

 

74 NRC acknowledged, on page 3-55 of its HTF TER, that, “DOE has a program in place to identify, evaluate, and implement 

cleaning technologies to remove HRRs to the MEP.” [ML14094A496] 
Basis for Section 3116 Determination DOE/SRS-WD-2014-001 

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Page 5-13 

In addition to the tanks listed above, DOE previously used slurry pumps successfully in HTF for bulk 

sludge removal in Tanks 21, 22 and 42, and, prior to the development of slurry pumps utilized highpressure water jets to remove sludge from Tanks 9, 10, 11 and 14. DOE used a combination of water 

dissolution and mechanical pumps as necessary to remove saltcake from Tanks 22, 24, 37 and 41 in 

HTF. Although these waste tanks were subsequently re-used and re-filled, they serve as further indicia of 

DOE experience and expertise using mechanical cleaning approaches at HTF. Furthermore, DOE has 

effectively implemented mechanical approaches to clean waste tanks in FTF, including Tanks 4, 5, 6, 17, 

18, 19 and 20.75 Removal of waste is influenced primarily by tank type and waste type, rather than the 

tank farm in which a particular tank is located, and DOE’s successful use of mechanical cleaning 

approaches in FTF further evinces DOE’s extensive experience and expertise in successfully using 

mechanical cleaning methods. The mechanical cleaning approaches for the SRS tank farms, and the 

predicate for their selection, are described in the ensuing 

discussion. 

In the late 1960’s, to consolidate sludge into a few waste 

tanks, DOE removed sludge from several Type I tanks using 

high-pressure water jets to disperse the sludge into slurry that 

was then removed using centrifugal transfer pumps. 

Although effective, the method created significant amounts of 

new waste (approximately five gallons of water for every 

gallon of sludge removed). Sludge removal utilizing this 

method was performed in several Type I tanks including HTF 

Tanks 9, 10, 11 and 14. In the 1970’s, DOE began 

investigating ways to remove sludge without utilizing large 

volumes of water and encumbering the available waste tank 

space, which led to development of slurry pumps. [SRRCWDA-2014-00003] 

Since the late 1970’s, mechanical mixing in HTF waste tanks 

has been performed utilizing standard slurry pumps or some 

variation thereof. The slurry pump is an older style mixing 

pump, with the motor located above the tank top. The pump 

connects to the motor with a long shaft through a water-filled 

column. The first successful test runs of a slurry pump were 

conducted in HTF Tank 16 in 1978 and 1979 and were 

utilized for the duration of the Tank 16 heel removal effort 

which included both mechanical and chemical heel removal. 

Mechanical heel removal in Tank 16 was carried out in five 

mixing and pumping campaigns utilizing one slurry pump for 

the first two campaigns and three slurry pumps for the 

remainder of the campaigns. The mechanical heel removal 

effort in Tank 16 was successful in reducing the waste 

volume in Tank 16 to an estimated 5,250 gallons prior to 

chemical heel removal. [SRR-CWDA-2011-00126] Since 

1979, 16 waste tanks at SRS have used variations of the 

slurry pump design for sludge removal, salt dissolution and 

sludge mixing for DWPF feed preparation. The most recent 

design of the slurry pump, Figure 5.2-2, is referred to as the 

standard slurry pump. In the late 1990s, reliability problems 

with the standard slurry pumps coupled with the need for 

extensive tank top modifications and a separate support 

system (i.e., bearing water) led to an investigation of alternate 

mixing technologies and eventual development of the SMP. 

[SRR-CWDA-2014-00003] Standard slurry pumps previously installed in HTF waste tanks may continue 

 

75 FTF waste tanks are not within the scope of this HTF 3116 Basis Document, but are discussed here to demonstrate DOE’s 

successful experience and extensive expertise in removing waste, including HRRs, from the SRS waste tanks. 

Figure 5.2-2: Typical Standard 

Slurry Pump 

[WSRC-TR-2001-00313]
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at the Savannah River Site December 2014 

Page 5-14 

to be used in heel removal activities. For example, HTF Tanks 11 and 12 underwent bulk waste removal 

efforts utilizing standard slurry pumps. The standard slurry pumps installed in Tank 12 were used to 

support heel removal activities within the waste tank. [SRR-CWDA-2013-00125, SRR-CWDA-2014-

00005] 

As documented in the 2003 Systems Engineering Evaluation (for HTF and FTF), the team of 

knowledgeable and experienced engineers and operations personnel recommended using floor-mounted, 

canned SMPs for bulk waste mixing and a mast-mounted STP for waste transfer. 

Based on the recommendations of the team, DOE selected and implemented mechanical removal 

techniques that use liquid (water76 and/or supernate) as the media for mixing. Mixer pumps used for bulk 

waste removal are also used in the subsequent heel removal process and may be augmented by 

spraying and lancing. Spraying and lancing within the waste tanks is performed by inserting a nozzle 

through an open riser in the waste tank and directing the liquid at a targeted location. Lancing typically is 

used to refer to a higher pressure, more concentrated spray pattern aimed at breaking-up or moving the 

solids within the waste tank. A recycle system, also referred to as a “feed and bleed” system, may be 

employed to enhance the efficiency and effectiveness of the mechanical removal of the solids, if practical. 

The technology consists of a Mobile Substation that provides power, a Mobile Waste Removal Control 

Center that provides local control and monitoring capabilities, SMPs for mixing and suspending waste 

solids and an STP for waste transfer. These mobile units have the capability of being co-located near 

any tank or tanks scheduled for waste removal. This concept efficiently performs waste removal using 

mobile and reusable equipment (Figure 5.2-3).77 

A key component of the mechanical waste removal equipment is the SMP because the ability to mix and 

suspend waste solids has a direct impact on the volume of solids remaining after mechanical heel 

removal. The SMPs are variable speed, single-stage centrifugal pumps with a 305-horsepower motor 

that can operate up to 1,600 revolutions per minute. The SMPs utilize the tank liquid waste to cool the 

 

76 Chemically-treated water is typically utilized when significant volumes will be added to the waste tanks to minimize the potential 

corrosion to the carbon steel primary tank walls and floors and the secondary annular pans, as applicable. 77 Figure 5.2-2 depicts two SMPs located in the waste tank. However, during actual cleaning operations DOE may deploy from one 

to four SMPs within a waste tank based upon the particular waste tank configuration and waste characteristics. 

Figure 5.2-3: Submersible Mixer Pump Waste Removal Diagram 
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Page 5-15 

motor and lubricate the upper and lower bearings. The SMPs are rotated by a turntable assembly that 

provides the motive force for oscillation or allows for stationary indexing operation. The SMPs have a 

rotating foot attached to the lower end of the pump, which allows the SMP to rest on the tank floor and 

oscillate. SMPs used for bulk waste removal are also used in the subsequent residual heel removal 

process and can be augmented by spraying or lancing. In some cases, a recycle system or feed and 

bleed system may be employed to enhance mechanical heel removal. Two discharge nozzles give the 

SMPs the capability to produce an effective cleaning radius of up to 50 feet. [M-CLC-G-00349] While 

obstructions such as cooling coils and support columns can significantly degrade the effective cleaning 

radius of SMPs, the use of multiple SMPs in an obstructed tank can make a significant contribution to 

removal of waste. To date, SMPs have been used to support bulk waste removal efforts on HTF Tank 13 

and FTF Tanks 4, 5 and 678 and to support mechanical and chemical cleaning of residual heels in Tanks 

5 and 6. The waste tank designs in HTF and FTF are essentially identical and the experience in FTF 

Tank 5 and Tank 6 mechanical heel removal has demonstrated that two or three SMPs can successfully 

suspend the majority of the sludge solids in a waste tank with internal obstructions. A Type I tank, such 

as Tank 5 and Tank 6, represents some of the most challenging tanks for waste removal activities due, in 

part, to a limited number of access points compared to a Type III/IIIA tank, the presence of roof support 

columns in the Type I tanks, and horizontal coiling coil runs at the bottom of the waste tank including 

stacked horizontal runs (often referred to as “fences”) that were “field to fit” during the time of waste tank 

construction versus only having vertical cooling coils in the waste zone in Type III/IIIA tanks. Pump 

configurations are shown in Figure 5.2-4 and Figure 5.2-5. 

The first deployments and operations of SMPs at SRS led to successful removal of sludge from two 

Type I tanks. Seven mechanical heel removal phases using SMPs in Tank 5 reduced the volume of 

sludge solids from approximately 34,000 gallons to approximately 3,500 gallons. In Tank 6, SMP 

operations during eleven mechanical heel removal phases resulted in the reduction of sludge solids from 

approximately 25,000 gallons to approximately 6,000 gallons. These first uses of SMPs provided the 

opportunity to gather and evaluate data to refine and enhance operational parameters such as mixer 

speed, mixer orientation and strategy (oscillation and fixed position) and coordination of mixer and 

transfer pump operations to optimize waste removal effectiveness in future tanks. Mechanical heel 

removal using SMPs successfully reduced the volume of sludge to the level required for chemical heel 

 

78 Discussion on Tanks 4, 5 and 6 cleaning is included for information; however, Tanks 4, 5 and 6 are located in the FTF at SRS and 

are not within the scope of this HTF 3116 Basis Document. 

Figure 5.2-4: Submersible Mixing Pump and Submersible 

Transfer Pump 

 Figure 5.2-5: Submersible 

Mixing Pump in a Test Tank 
Basis for Section 3116 Determination DOE/SRS-WD-2014-001 

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removal in Tank 5 and Tank 6. Following the chemical cleaning campaigns in Tanks 5 and 6, mechanical 

cleaning utilizing a feed and bleed process with three SMPs was carried out in each of the waste tanks. 

[SRR-CWDA-2011-00033, SRR-CWDA-2011-00005, M-ESR-F-00107, M-ESR-F-00147, M-ESR-F00132] 

5.2.2.2 Chemical Heel Removal Cleaning 

The DOE also successfully uses chemical heel removal technology that employs oxalic acid for chemical 

treatment of the heel to dissolve solids that cannot be removed by mechanical methods and water 

addition alone. As practical, the oxalic acid also may be sprayed into the tank to further clean 

contaminants from the internal tank surfaces (e.g., walls, cooling coils, support columns, equipment). At 

the conclusion of chemical heel removal, the interior of the waste tank are washed with water to rinse 

oxalic acid from internal surfaces and dislodge loose contamination. 

Chemically-aided cleaning techniques have been evaluated for additional levels of waste removal 

following mechanical heel removal. A team of knowledgeable and experienced engineers and scientists 

assessed the current knowledge base and collected and evaluated information available on chemicalbased methods for removing residual solids from the waste tanks. [WSRC-TR-2003-00401] As part of 

this study, the team developed recommendations for chemical treatments to remove residual solids. The 

cleaning agents identified included: 

 oxalic acid, 

 a mixture of oxalic acid and citric acid, 

 a combination of oxalic acid with hydrogen peroxide, 

 nitric acid, 

 formic acid, and 

 organics. 

The results of the evaluation support oxalic acid as the cleaning agent of choice. Nitric acid, formic acid 

and oxalic acid with hydrogen peroxide were all closely grouped for the next best choice. The mixture of 

oxalic acid and citric acid rated poorly (primarily due to the fact that it performed less well than oxalic acid 

and the presence of citrate could adversely impact downstream treatment operations, such as the 

DWPF). Organics rated even more poorly due to large uncertainties in performance and downstream 

impacts. 

The use of oxalic acid was recommended for a number of reasons. First, oxalic acid has been widely 

studied and used several times to clean waste tanks at SRS and at other sites within the DOE Complex. 

Its effect on downstream waste treatment process (e.g., DWPF) and evaporator operations is better 

known. Oxalic acid has been shown to be effective for a wide variety of sludge types and out-performed 

nitric acid and other chemical cleaning agents in head-to-head laboratory tests. Lastly, oxalic acid is less 

corrosive to the carbon steel tanks than nitric acid or a combination of oxalic acid and hydrogen peroxide. 

[WSRC-TR-2003-00401] 

Oxalic acid cleaning of tanks was successfully demonstrated through the cleaning of Tank 16 in the early 

1980s. Oxalic acid cleaning in Tank 16 (part of an overall waste removal program which also employed 

mechanical cleaning) included, along with various water washes, three chemical heel removal campaigns 

each involving the addition of four weight percent oxalic acid to the waste tank, mixing of the waste tank 

contents and subsequent transfer of the mixed solution out of the waste tank. The volume of the 

residuals in the primary tank at the conclusion of the oxalic acid cleaning and water washes was 

determined to be 330 gallons. [U-ESR-H-00113] 

Tanks 5 and 679 were cleaned using three large batches of eight weight percent oxalic acid similar to 

Tank 16 cleaning. This process is referred to as Bulk Oxalic Acid Cleaning. If needed, oxalic acid may 

be sprayed into the tank to clean contaminants from internal tank surfaces (e.g., walls, cooling coils, 

support columns, equipment, etc.). The internal surfaces of Tanks 5 and 6 were sprayed with oxalic acid 

during the cleaning process on those waste tanks. [SRR-CWDA-2011-00033, SRR-CWDA-2011-00005] 

Oxalic acid was applied multiple times in each tank using various methods (e.g., downcomer, spray 

 

79 Discussion on Tanks 5 and 6 cleaning is included to demonstrate successful deployment of the oxalic acid cleaning process; 

however, Tank 5 and 6 are located in the FTF at SRS and is not within the scope of this HTF 3116 Basis Document. 
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nozzle, mixer agitation and non-agitation soak times) which provided data for process evaluation, 

improved effectiveness and overall process optimization. Oxalic acid cleaning in Tank 5 reduced the 

volume of residual solids to approximately 3,300 gallons, while the residual solids volume in Tank 6 was 

reduced to approximately 3,500 gallons. [SRR-CWDA-2011-00033, SRR-CWDA-2011-00005, M-ESR-F00160, M-ESR-F-00165] A significant portion of the remaining waste is non-radioactive oxalate 

compounds that formed during the chemical cleaning process. Formation of these non-radioactive 

oxalate compounds is demonstrated by a greater than 30 % increase in waste volume between the first 

and second chemical cleaning cycles and the second and third chemical cleaning cycles for Tank 6 and 

between the first and second chemical cleaning cycles in Tank 5. [SRR-CWDA-2011-00033, SRRCWDA-2011-00005, M-ESR-F-00158, M-ESR-F-00165] 

Oxalic acid cleaning was conducted in Tank 12 in 2013. Prior to executing the oxalic acid cleaning, the 

Tank 12 flowsheet was revised to incorporate lessons learned from previous chemical cleaning 

campaigns (i.e., Tanks 16, 24, 5 and 6). The improvement to the flowsheet included such things as 

washing the Tank 12 material prior to cleaning to reduce the sodium concentration, modifying the mixing 

strategy to improve the contact rate of the acid with the material, monitoring pH of the solution and 

maintaining temperature of the material at the maximum level allowed. Overall, the Tank 12 oxalic acid 

cleaning campaign reduced the residual material volume by greater than 75 %. [SRR-CWDA-2014-

00003] The volume of the residuals in the primary tank has been estimated to be approximately 1,400 

gallons based on preliminary volume estimates, however, final volume determination has not been 

completed. [U-ESR-H-00109, M-CLC-H-03256] 

As discussed above, chemical cleaning of waste tanks using oxalic acid produces sodium oxalates in the 

solids slurry that will be eventually included as feed to DWPF. Because of sodium limits and oxalate 

restrictions on the DWPF feed, preparation of the feed results in a significant amount of additional 

material being generated that eventually must be processed through SWPF and disposed of in the SDF.80 

The oxalic acid flowsheet evaluations have considered the downstream effects of oxalates on the DWPF 

process and salt processing to determine quantities of oxalic acid that can be tolerated by the Liquid 

Waste System. Based on a cleaning strategy similar to that used on Tank 12, a waste tank cleaned with 

oxalic acid is projected to add approximately 32,000 kg (i.e., 5,550 gallons) of new sodium oxalates (a 

non-radioactive compound) to the Liquid Waste System. In addition, approximately 500,000 gallons of 

salt waste will be created. These quantities of oxalates result in additional wash cycles for DWPF feed, 

increased likelihood of feed breaks to DWPF and extension of the operating life of the entire Liquid Waste 

System. [SRR-CWDA-2014-00003] The oxalates are also anticipated to create evaporator foaming and 

scaling problems. [LWO-SPT-2008-00033] Due to these downstream impacts, the amount of Bulk Oxalic 

Acid Cleaning in HTF will be carefully controlled to optimize the cleaning effectiveness and the 

downstream waste treatment processes. 

Another example of a chemical cleaning method is Low Temperature Aluminum Dissolution which DOE 

has deployed at HTF to help remove residual solids. This methodology has been utilized to help reduce 

the solids volume in SRS Tank 12 and is applicable to sludges that have a high percentage of aluminum, 

such as sludge that originated from H-Canyon. [X-CLC-H-00921] Under this process, aluminum is 

dissolved from sludge waste into the supernate by treatment with caustic, followed by decantation and 

water washing to subsequently remove aluminum. Aluminum solids in the sludge are believed to be 

present in primarily three compounds – aluminum trihydrate or gibbsite, alumina monohydrate or 

boehmite, and aluminosilicate. With caustic treatment, the gibbsite form dissolves readily at the relatively 

low dissolving temperatures possible in the waste tanks. The boehmite form dissolves much more slowly 

and is somewhat less soluble than gibbsite, but can still be dissolved at relatively low temperatures, given 

enough time. The aluminosilicate has such a low solubility in waste slurries that it is generally considered 

insoluble. [SRNS-STI-2008-00021] 

5.2.2.3 Vacuum Heel Removal Cleaning 

As the result of a March 2006 DOE-sponsored Tank Cleaning Technical Exchange, a new vacuum 

technology was identified. The DOE has adapted and successfully used this new technology in 

unobstructed Type IV tanks, for which there are four in HTF, Tanks 21, 22, 23 and 24. This technology 

 

80 See Appendix A for a brief description of DWPF, SWPF and SDF operations. 
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used an ultra-high-pressure water eductor to vacuum residual solids and transport the slurry to a receipt 

tank. This technology was initially deployed in Type IV tanks with no internal obstructions due the size of 

the device and the large accompanying tether system. 

To deploy this technology in Tank 18 and Tank 19,81 located at SRS FTF, DOE utilized a cleaning device, 

called a Mantis, which consists of a mechanical crawler and an eductor assembly that made up a retrieval 

system utilizing an ultra-high-pressure water eductor to vacuum residual solids and transport the slurry to 

a receipt tank (Figure 5.2-6). The process system consists of a remotely controlled, in-tank Mantis, an 

umbilical hose containing hydraulic supply lines and the high-pressure water hoses, in-tank waste 

retrieval hose, a diesel-driven ultra-high-pressure water pump, a motor-driven high pressure water pump, 

hydraulic pump skid, a diesel generator, above-ground hose-in-hose transfer lines, WMC and support 

equipment. The device was inserted into the tank through a 24-inch riser in a folded position. Once 

inside the tank, the device was unfolded into its operational configuration. 

Figure 5.2-6: Mantis 

The Mantis was remotely driven around the waste tank bottom by an operator located in the Control 

Center. A high pressure hydro-lance at its front was used to break up waste mounds and an eductor was 

used to vacuum waste from the floor of the waste tanks. The waste traveled through the eductor in-tank 

waste retrieval hose up into a tee spool piece located on top of the tank riser and then through an aboveground transfer line that terminated inside a WMC installed inside a riser on the receipt tank. An 

immersion mill, located near the bottom of the WMC, size-reduced solid waste particles so that the 

particles can be more easily re-suspended in future waste removal activities. [WSRC-TR-2007-00327] 

The deployment of this new cleaning device allowed removal of waste from Tanks 18 and 19 to a greater 

extent than the technologies available when waste removal was previously discontinued due to 

diminishing returns in 2003 and 2001, respectively. 

The working end effecter of the Mantis, the ultra-high-pressure eductor system, effectively vacuumed 

residual solids and transported the slurry to the receipt tank. The cooling coils in Type I, Type II, Type III 

and Type IIIA tanks precludes the use of large tethered mechanical crawlers such as the Mantis platform. 

However, DOE recognizes the potential for future use of vacuum technology deployed on other platforms 

 

81 Discussion of Tanks 18 and 19 is included for information regarding the use of the Mantis technology; however, Tanks 18 and 19

are located in FTF at SRS and are not within the scope of this HTF 3116 Basis Document. 
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and continues to evaluate potential deployment on platforms such as alternate mechanical crawlers or 

robotic arm-based technologies. [SRR-CWDA-2014-00003, SRR-LWE-2014-00055] 

5.2.3 Optimization of Existing Technologies 

DOE continues to optimize its existing technologies that have been successfully deployed for waste 

removal by applying lessons learned from previous waste removal campaigns when establishing future 

waste removal strategies. [SRR-CWDA-2014-00003] DOE is continuing to pursue small-scale robotic 

technologies for waste removal and sampling applications in tanks with extensive internal obstructions. 

For example, a small robotic crawler 

was developed and utilized to sample 

both Type I and Type IV tanks at SRS 

(Figure 5.2-7), and DOE continues to 

evaluate available robotic technologies 

for applications in future waste tanks. 

[SRR-LWE-2014-00055] Such tactical 

applications of tailored robotic 

platforms will continue to be used in 

future waste removal activities in HTF. 

5.3 Removal to the Maximum 

Extent Practical 

As described above, extensive waste 

removal operations have occurred at 

SRS. Based on waste removal 

experience to date and anticipated new 

technologies, HTF waste removal 

activities will result in significant 

collective removal of waste including 

HRRs.82 For example, waste removal activities in HTF Tank 16 resulted in removal of over 99 % of the 

waste volume from the tank.83 [SRR-CWDA-2013-0041] In HTF Tank 12, waste removal activities also 

resulted in removal of over 99 % of the waste volume.84 [SRR-CWDA-2013-00125] Furthermore, 

experience to date with waste removal for FTF Tanks 18 and 19 resulted in removal of over 99 % of the 

waste volume from Tanks 18 and 19, and approximately 99 % of the HRR inventory from Tank 18 and 

greater than 99 % of the HRR inventory from Tank 19, based on a starting point of the maximum 

historical radionuclide inventory in those tanks. [SRR-CWDA-2011-00091] In FTF Tanks 5 and 6, waste 

removal activities resulted in removal of over 99 % of the waste volume from the tanks. [SRR-CWDA2012-00138] 

Removal of HRR’s begins with the removal of the solids and liquid from a waste tank or ancillary structure 

in a bulk waste removal phase.85 Following bulk waste removal, heel removal is performed using a mix of 

 

82 In this regard, NDAA Section 3116 does not specify “remedial goals” or other numerical objectives and does not require DOE to develop 

any such removal goals or objectives. Although the cleaning methodologies are expected to collectively remove approximately 99% of 

HRRs, based on a starting point of the maximum historical radionuclide inventory in the overall HTF, individual waste tanks or ancillary 

structures may not achieve this level of HRR removal on an individual basis. Demonstration that waste removal within a particular waste 

tank or ancillary structure has achieved 99% removal of HRRs is not, by itself, a justification for stopping HRR removal activities. In addition, 

demonstration that residual radionuclide inventory of a given waste tank or ancillary structure is below that assumed in the HTF PA is not 

sole justification to conclude cleaning activities on an individual waste tank or ancillary structure. 

83 Final Tank 16 primary tank residual volume was determined to be 330 gallons. [U-ESR-H-00113] The final residual volume was lower 

than the 1,000 gallons utilized in the initial qualitative estimate. [SRR-CWDA-2011-00126] 

84 Tank 12 final volume estimate is based on a qualitative estimate of residuals; final volume determination is ongoing. 85 Bulk waste removal efforts have been completed for Tank 12 (Type I tank), and bulk waste removal efforts are currently underway in Tank 

10 (a Type I tank) and Tank 13 (a Type II tank); preparation for bulk waste removal efforts in Tank 15 (a Type II tank) are in progress. In 

addition, DOE has previously performed waste removal activities in various HTF waste tanks to support efficient management of ongoing 

tank farm operations and preparation of feed batches for DWPF. Bulk sludge removal activities have been performed in HTF Tanks 9, 10, 

11, 14, 15, 21, 22 and 42. Bulk saltcake removal has been performed in HTF Tanks 22, 24, 37 and 41. After completion of the waste 

removal activities in these tanks, the available storage space created within these tanks was subsequently re-used to support ongoing tank 

farm operations. Although DOE is or will be further removing waste from these tanks to support closure, DOE’s prior experience in bulk 

waste (sludge and saltcake) removal confirms DOE’s expertise in effectively using proven technologies to remove tank waste and associated 

HRRs during the bulk waste removal phase. 

Figure 5.2-7: Robotic Sampler at Test Facility 
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technologies described above, as appropriate, accounting for the physical configuration of the tank and 

the chemical characteristics of the waste. 

Throughout the heel removal process, DOE continually evaluates the ongoing effectiveness of the 

technology being implemented and optimizes the existing technologies. In addition, DOE evaluates the 

usefulness and practicality of additional technology deployment once the existing technology has reached 

the point of diminished effectiveness for HRR removal. The DOE’s approach consists of the following 

phases: initial technology selection, technology implementation, technology execution, technology 

effectiveness evaluation and additional technology evaluation.86 [DOE/SRS-WD-2011-001] Additional 

description of these phases is provided in Appendix B. 

Tank 16, a Type II tank in HTF, has undergone extensive waste removal and tank cleaning activities. In 

1972, due to continued leakage from the primary tank to the annulus, the supernate in Tank 16 was 

transferred out of the waste tank leaving approximately 77,000 gallons of sludge solids in the Tank 16 

primary tank. In 1978, Tank 16 became the first tank to utilize slurry pumps to perform waste removal 

activities. Bulk sludge removal from Tank 16 was carried out in five mixing and pumping campaigns 

utilizing one slurry pump for the first two campaigns and three slurry pumps for the remainder of the 

campaigns. The mechanical heel removal effort in Tank 16 was successful in reducing the waste volume 

in Tank 16 to an estimated 5,250 gallons. In 1980, Tank 16 became the first waste tank to undergo 

chemical cleaning utilizing oxalic acid and underwent a series of chemical cleaning campaigns using 

oxalic acid, along with various water washes. During the chemical cleaning activities, three chemical heel 

removal campaigns each involving the addition of four weight percent oxalic acid to the waste tank, 

mixing of the waste tank contents and subsequent transfer of the mixed solution out of the waste tank 

were carried out. The volume of the residuals in the primary tank at the conclusion of the oxalic acid 

cleaning and water washes was estimated to be approximately 1,000 gallons, however, final inventory 

determination had not been completed. Figure 5.3-1 shows the overall waste removal effectiveness for 

the Tank 16 primary tank waste removal activities. [SRR-CWDA-2011-00126] Based on final inspection 

of the Tank 16 primary tank, the residual inventory of the primary tank has been determined to be 330 

gallons. [U-ESR-H-00113] 

 

86 Per the FFA, the waste tanks will be cleaned until DOE-SR, SCDHEC and EPA agree that waste removal may cease. 

Figure 5.3-1: Tank 16 Waste Removal 
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Tank 16 was put into service in 1959 and shortly thereafter leak sites were discovered in the primary tank. 

In 1960, the majority of waste that had leaked into the annulus was transferred out to another waste tank. 

Between 1960 and 1971, periodic inspection of the annulus revealed an increase in leak sites and slow 

but continued seepage into the annulus. In 1972, additional liquid was extracted from the annulus and at 

that time it was estimated that 6,000 gallons of saltcake remained in the annulus. In 1974, the outside of 

the primary tank wall was sandblasted to facilitate leak site inspection and about 20 cubic feet of sand 

accumulated in the annulus. In 1976, vacuum operations removed some of the sand from the annulus. 

In 1977, water was introduced in the annulus to dissolve remaining solids and steam jets were used to 

increase the temperature and promote circulation. Approximately 1,400 gallons of material were 

estimated to have been removed from the annulus. [SRR-CWDA-2011-00126] In 2007 and again in 

2010, DOE investigated the potential use of vendor supplied robotic equipment to remove additional 

material from the Tank 16 annulus. DOE evaluated a number of technologies and selected water sluicing 

as the most viable option for additional waste removal from the Tank 16 annulus. In 2012 a full scale 

demonstration test of the water sluicing technology was performed. The results of the testing confirmed 

an airborne radiological hazard associated with the technology, showed a moderate ability to move waste 

within the spray zone and a limited ability to move solids to the pump suction. Based on laboratory 

testing showing low solubility of key radionuclides, the Tank 16 residuals being much less soluble than 

typical salt waste, difficulty with breaking-up and moving residuals to the pump suction and radiological 

hazards associated with the sluicing technology, DOE determined, based on a qualitative assessment, 

that it appeared probable that implementation and execution of the technology would not be practical. 

[SRR-CWDA-2013-00041] As required by the HTF GCP (SRR-CWDA-2011-00022), DOE reviewed this 

information with SCDHEC and the EPA. The three agencies (DOE, SCDHEC, EPA) concurred and DOE 

suspended waste removal activities and initiated final sampling and analysis of the Tank 16 residuals. 

[WDPD-13-40, EPA_04-10-2013, DHEC_04-11-2013] Final residual characterization for Tank 16 is 

currently in progress. Following the process outlined in Appendix B of this HTF 3116 Basis Document, a 

Tank 16 Special Analysis will be developed and formally reviewed and approved that will consider the 

final characterization of the residuals remaining in both the primary tank and the annulus. The data from 

the Tank 16 Special Analysis will support a cost-benefit analysis and subsequently, the development of a 

removal report for Tank 16.87 The removal report will combine documentation from the entire waste 

removal process to provide the complete demonstration of radionuclide removal to support Tier 2 Closure 

Authorization for Tank 16.88

Tank 12, a Type I tank in HTF, has undergone extensive waste removal and tank cleaning activities. The 

largest volume of waste stored in Tank 12 has been approximately 736,000 gallons. [DPSPU 78-11-9] 

Bulk waste removal efforts within the tank were initiated with supernate removal in 1974. Design and 

installation of bulk sludge removal equipment began in 2003 and resulted in the installation of four slurry 

pumps to provide mechanical mixing of the sludge. Bulk waste removal efforts resumed in 2009, were 

completed in 2010 and involved multiple mixing and transfer sludge removal campaigns. The total 

volume of solids at the beginning of these campaigns was estimated to be approximately 203,000 

gallons. Upon completion of the bulk waste removal efforts in Tank 12 the total volume of sludge 

remaining was estimated to be approximately 22,000 gallons. [U-ESR-H-00093] From 2009 through 

2011, three mechanical sludge removal campaigns were conducted in Tank 12 to support sludge batch 

preparation for DWPF. [SRR-LWP-2012-00061] The solids remaining in the waste tank at the conclusion 

of the mechanical sludge removal campaigns was estimated to be 13,700 gallons. A chemical cleaning 

 

87 As noted in footnote 144 in Appendix B of this HTF 3116 Basis Document, “The results of these special analyses will be used to

support the cost-benefit analyses and subsequent removal reports generated for each of the waste tanks and ancillary structures.” 88 As Key Recommendation ES-2.8, on page xviii of the TER the NRC states, “Given the potential risk significance of the waste 

remaining in the Tank 16 annulus, the NRC staff recommends that DOE more fully evaluate the practicality of additional radionuclide 

removal from the Tank 16 annulus versus the long-term benefit of reduced risk considering uncertainty in the releases of 

radionuclides from the Tank 16 annulus. While DOE’s HTF PA demonstrates that the risk from waste remaining in the annulus is 

reasonable, alternative waste release models may lead to higher risk estimates.” The NRC acknowledges DOE’s process on page 

3-54 of the TER, “DOE will provide more detailed documentation on the practicality of removing additional waste from the Tank 16 

annulus in a Tank 16 Removal Report that will be issued after the completion of final residual characterization and subsequent costbenefit analysis. The NRC staff recommends that DOE evaluate a waste release scenario due to groundwater in-leakage into and 

out of the annular region and contacting the high-solubility waste in the annuli of those tanks with residual material in the annulus 

and sandpads, including Tank 16. The NRC staff also acknowledges that it may be impractical to further clean the annulus of Tank 

16 due to the recalcitrant nature of some of the annular contamination and nuclear safety concerns.” As part of the Tank 16 Special 

Analysis, DOE will evaluate contaminant release from the Type I and Type II annuli, including Tank 16. 
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campaign using Low Temperature Aluminum Dissolution, completed in 2011, reduced the residual solids 

volume to approximately 7,800 gallons. Additional mechanical sludge removal performed in 2012, 

following Low Temperature Aluminum Dissolution, further reduced the residual solids volume to 

approximately 4,400 gallons. A second chemical cleaning campaign, using oxalic acid, was performed in 

2013. Prior to executing the oxalic acid cleaning, the Tank 12 flowsheet was revised to incorporate 

lessons learned from previous chemical cleaning campaigns (i.e., Tanks 16, 24, 5 and 6). The 

improvement to the flowsheet included such things as washing the Tank 12 material prior to cleaning to 

reduce the sodium concentration, modifying the mixing strategy to improve the contact rate of the acid 

with the material, monitoring pH of the solution and maintaining the temperature of the material at the 

maximum level allowed taking into consideration tank corrosion chemistry requirements. Overall, the 

Tank 12 oxalic acid cleaning campaign reduced the residual material volume by greater than 75 %. 

[SRR-CWDA-2014-00003] 

A qualitative assessment of additional waste removal options for Tank 12 indicated that additional waste 

removal is not practical. Additional Bulk Oxalic Acid Cleaning would have minimal benefits expected 

based on an evaluation of the flowsheet and process samples. A new mixing and transfer strategy using 

SMPs would provide minimal return on investment due to cost of disassembly, removal and relocation of 

SMPs from FTF. Furthermore, the effectiveness of additional waste removal utilizing SMPs is uncertain 

considering amount and location of residual solids remaining in Tank 12. Vacuum technology was 

considered, but the proven Mantis vacuum technology which has been deployed in Type IV tanks (i.e., 

Tanks 18 and 19) is not feasible due to in-tank obstructions (e.g., cooling coils) in the Type I tanks. 

Smaller robotic vacuum technology has not reached a technical maturity that would support in-tank 

deployment for waste removal.89 Both Low Temperature Aluminum Dissolution and Bulk Oxalic Acid 

Chemical Cleaning were deployed in Tank 12; no other chemical cleaning technology is technically 

mature enough for deployment. The Tank 12 annulus is estimated to contain less than 30 gallons of salt 

waste, minimal benefit is expected from annulus cleaning. [SRR-CWDA-2013-00125] 

Figure 5.3-2 depicts the overall waste removal effectiveness for waste removal activities to date in Tank 

12. The volume of the residuals in the primary tank at the conclusion of heel removal operations has 

been estimated to be approximately 1,400 gallons based on preliminary volume estimates, however, final 

volume determination has not been completed. [U-ESR-H-00109, M-CLC-H-03256] 

 

89 As discussed in Section 5.2.3 of this document, DOE has utilized small robotic crawlers for sampling of waste tank residuals. 

Figure 5.3-2: Tank 12 Waste Removal 
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DOE reviewed the above information with SCDHEC and EPA as required by the HTF GCP, and the three 

agencies reached concurrence to suspend waste removal activities and move into final sampling and 

analysis. [SRR-CWDA-2013-00125, WDPD-14-20, EPA-OS-2014-02-19-01, DHEC-OS-2014-02-18-01] 

Sampling of the waste tank is currently in progress. As outlined in Appendix B of this document, DOE will 

be documenting final waste tank inventories, radionuclide removal effectiveness, with an emphasis on 

HRRs, and final cost-benefit analysis in the removal report. 

The FTF Type IV tanks, Tanks 17, 18, 19 and 20, which are essentially the same design as the HTF Type 

IV tanks, have all undergone waste removal and tank cleaning activities resulting in a relatively small 

quantity of resultant tank residuals. For example, in 1998 following bulk waste removal in Tank 19, DOE 

used a systematic selection process which was documented in a Systems Engineering Evaluation to 

select the best available technology at the time for heel removal activities in Tank 19. [PIT-MISC-0040] 

Heel removal activities using the selected mechanical removal technology were carried out in 2001. In 

2001, a similar selection process was also used to select a removal technology for the heel in Tank 18, 

similarly a Type IV waste tank. [WSRC-RP-2001-00024] This Tank 18 technology selection took into 

account additional technology studies conducted since the issuance of the Tank 19 Systems Engineering 

Evaluation. Heel removal activities were carried out in Tank 18 in 2002 utilizing a different mechanical 

removal technology. In these initial Tank 18 and Tank 19 heel removal campaigns a series of waste 

removal phases were carried out in each of the tanks until it was no longer practical to continue with the 

mechanical removal technologies that were being utilized. In 2006, following initial heel removal 

campaigns using the tailored mechanical removal techniques, it was determined that it was practical to 

deploy an alternative vacuum technology, the Mantis (as described above in Section 5.2.2.3), that could 

result in significant additional waste removal within these tanks. 

Throughout the heel removal activities in Tank 18 and Tank 19 utilizing the Mantis, DOE continually 

worked to optimize the effectiveness of the Mantis and minimize the impact on the rest of the Liquid 

Waste System by adjusting how the sprays were utilized, attempting different vacuuming patterns, using 

the hose/cable bundle to drag the solids into a concentrated area, or turning off the sprays when possible 

to improve removal efficiency and reduce space impacts on the receipt tank. During the Mantis 

campaigns on both Tank 18 and Tank 19, the Mantis became ineffective due to failure of one or more of 

the in-tank components on the equipment. In both cases, DOE evaluated the costs of repairing the 

Mantis and the anticipated effectiveness once repaired and determined that it was practical to make the 

repairs and continue the heel removal operations. The Mantis equipment was utilized within Tank 18 and 

Tank 19 until it was no longer effective. Factors leading to this decision included such things as: visual 

observation of remaining tank residuals, transfer line radiation readings, significant increase in ratio of 

water additions to solids removed and significant equipment degradation. Figure 5.3-3 and Figure 5.3-4 

depict the overall waste removal effectiveness for Tanks 18 and 19 respectively. Mantis operations in 

Tanks 18 and 19 reduced the volume of residual solids to approximately 4,000 gallons in Tank 18 and 

2,000 gallons in Tank 19. [U-ESR-F-00041, U-ESR-F-00042] 

 
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Figure 5.3-4: Tank 19 Waste Removal 

 

Figure 5.3-3: Tank 18 Waste Removal 
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Once it was determined that the existing Mantis equipment was no longer effective, alternative HRR 

removal technologies were reviewed to determine practicality for design, construction, deployment and 

operation. An evaluation was completed to determine if it was useful to develop and deploy another 

cleaning technology assuming such a technology could be identified and safely deployed. This 

evaluation considered whether the costs, such as monetary costs, delays in higher- risk reducing 

activities, or occupational exposure of site workers to hazardous or potentially hazardous materials, 

including radioactive materials, outweighed the potential benefits associated with further waste removal in 

Tanks 18 and 19. Though no new practical technology was identified in the technology evaluation, an 

upgraded Mantis was considered to have the highest likelihood of success in removing additional residual 

waste, could be deployed in the least amount of time, and would be the least costly technology alternative 

to implement. A cost-benefit analysis was performed to determine the potential risk reduction from 

removing additional residual waste from Tanks 18 and 19. The cost-benefit analysis considered a broad 

range of costs including resultant schedule impacts on other ongoing cleaning activities and waste 

disposition activities, as well as the current state of waste removal capabilities and technologies. As a 

result of the cost-benefit analysis, DOE concluded that the relatively insignificant benefits of removing 

additional waste from Tank 18 or Tank 19 do not outweigh the costs of implementation or the detrimental 

impacts to ongoing and future FTF cleaning and stabilization activities to reduce risks to the public, the 

workers and the environment. Therefore, even if a technology could be identified and deployed, the 

relatively insignificant reduction of risk associated with further removal of residuals from Tank 18 or Tank 

19 would not justify the associated additional costs. [SRR-CWDA-2011-00091] 

In consideration of an NRC recommendation in the FTF TER, DOE conducted extensive additional 

analysis, including an additional Tank 18 waste removal Systems Engineering Evaluation and a new costbenefit analysis, to provide further demonstration of removal of HRRs to the MEP from Tank 18. 

[ML112371715, SRR-CWDA-2012-00026] The Systems Engineering Evaluation was performed to 

identify the most promising technologies for removal of additional HRRs from Tank 18, and included a 

review of previous tank cleaning evaluations and assessed the potential for deployment of new or 

emerging technologies. The team performing the Systems Engineering Evaluation did not identify any 

new or emerging tank cleaning technologies that had not been previously considered by DOE, and 

identified no other new developments that appeared to be on the horizon. However, the team determined 

that the tools necessary to implement various cleaning technologies have become more mature in recent 

years, that more vendors can supply these tools, and that four technologies merited further study. The 

results of the new Tank 18 waste removal Systems Engineering Evaluation and the new cost-benefit 

analysis (SRR-CWDA-2012-00026) further supported DOE’s conclusion reached in the previous costbenefit analysis (SRR-CWDA-2011-00091) that removing additional HRRs from Tank 18 would not justify 

the associated additional costs.

Two Type I tanks have also undergone extensive heel removal campaigns. These two Type I tanks, 

Tank 5 and Tank 6 in FTF, which are essentially the same design as the HTF Type I tanks, originally 

contained approximately 34,000 and 25,000 gallons of sludge solids, respectively, at the conclusion of 

their bulk waste removal campaigns. A Type I tank, such as Tank 5 and Tank 6, represents the most 

challenging tank for waste removal activities due, in part, to a limited number of access points, horizontal 

cooling coil runs at the bottom of the waste tank including stacked horizontal runs (often referred to as 

“fences”) that were “field to fit” during the time of waste tank construction and the presence of roof support 

columns. Experience in Tank 5 and Tank 6 demonstrates DOE’s successful deployment of innovative 

technologies capable of removing HRRs even under the most challenging conditions. [SRR-CWDA2012-00138, SRR-CWDA-2011-00033, SRR-CWDA-2011-00005] As described in Section 5.2.2, in 2003, 

DOE performed a Systems Engineering Evaluation to identify, evaluate and select equipment for waste 

removal tasks to accelerate waste removal and the removal of waste tanks from service. This evaluation 

resulted in the selection of new mechanical removal technologies in combination with chemical removal 

technologies, if necessary. 

These new mechanical technologies, in combination with the chemical removal technology, were first 

utilized in Tank 5 and Tank 6. Throughout the initial mechanical heel removal operations in both Tank 5 

and Tank 6, DOE performed numerous waste removal phases consisting of liquid addition followed by 

SMP operation and then pump down of the slurried waste. Prior to each of these phases, DOE evaluated 

the results from the previous phases and optimized the removal effectiveness by adjusting the indexing of 
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the SMPs or utilizing a hydro-lance to disperse the solids from areas the SMPs were not effectively 

cleaning. Once it was determined that the existing mechanical cleaning method had reached a point of 

diminished effectiveness, chemical heel removal cleaning, described in Section 5.2.2.2, was utilized until 

it too reached a point of diminished effectiveness. The initial plan for both Tank 5 and Tank 6 was to first 

perform mechanical cleaning of the tank heels and follow that with chemical heel removal. However, at 

the conclusion of the chemical heel removal cleaning campaign in both tanks, DOE evaluated the tank 

conditions and determined that it would be practical to deploy additional cleaning methods within the 

tanks. In Tank 5, for example, following the implementation of three chemical cleaning cycles using a 

bulk oxalic acid flowsheet, a new mechanical cleaning method utilizing the three existing SMPs in the 

tank was deployed. Instead of adding liquid, attempting to slurry the solids and then pumping out the 

solution in a batch fashion, which required the SMP to be turned off at a certain point in the pump down, 

the operations were modified to allow for a continuous “feed and bleed” to occur. This new methodology 

was deployed until it was determined that the new methodology had reached diminished returns. A 

modified version of this “feed and bleed” mechanical cleaning technology was also utilized in Tank 6 as a 

final cleaning campaign until it had reached diminished returns. [SRR-CWDA-2011-00033, SRR-CWDA2011-00005] 

A qualitative assessment of additional waste removal options for Tanks 5 and 6 indicated that additional 

waste removal was not practical. DOE reviewed the information with SCDHEC and EPA as required by 

the FTF GCP, and the three agencies reached concurrence to suspend waste removal activities and 

move into final sampling and analysis. [SRR-CWDA-2010-00157, EPA_12-08-2010, DHEC_11-22-2010] 

Sampling and analysis of the waste tanks was completed and, as outlined in Appendix B of this 

document, DOE documented the final waste tank inventories, radionuclide removal effectiveness, with an 

emphasis on HRRs, and final cost-benefit analysis in a removal report for Tanks 5 and 6. [SRR-CWDA2012-00138] 

Figure 5.3-5 and Figure 5.3-6 depict the overall waste removal effectiveness for Tanks 5 and 6 

respectively. Heel removal operations reduced the residual solids to approximately 1,900 gallons and 

3,000 gallons respectively. [SRR-CWDA-2012-00138] 

Figure 5.3-5: Tank 5 Waste Removal 
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Once it was determined that the extent of technology had been reached for the “feed and bleed” 

mechanical cleaning technology that had been deployed, alternative HRR removal technologies were 

reviewed to determine practicality for design, construction, deployment and operation. An evaluation was 

completed to determine if it was useful to develop and deploy another cleaning technology assuming 

such a technology could be identified and safely deployed. This evaluation considered whether the costs, 

such as monetary costs, delays in higher-risk reducing activities, or occupational exposure of site workers 

to hazardous or potentially hazardous materials, including radioactive materials, outweighed the potential 

benefits associated with further waste removal in Tanks 5 and 6. No new practical technology was 

identified in the technology evaluation. A cost-benefit analysis, incorporating the results of a Tank 5 and 

6 Special Analysis, was performed to determine the potential risk reduction from removing additional 

residual waste from Tanks 5 and 6. The cost-benefit analysis considered a broad range of costs 

including resultant schedule impacts on other ongoing cleaning activities and waste disposition activities, 

as well as the current state of waste removal capabilities and technologies. As a result of the cost-benefit 

analysis, DOE concluded that the relatively insignificant benefits of removing additional waste from Tank 

5 or Tank 6 do not outweigh the costs of implementation or the detrimental impacts to ongoing and future 

FTF cleaning and stabilization activities to reduce risks to the public, the workers and the environment. 

Therefore, DOE concluded even if a technology could be identified and deployed, the relatively 

insignificant reduction of risk associated with further removal of residuals from Tank 5 or Tank 6 would not 

justify the associated additional costs. [SRR-CWDA-2012-00138] 

Effective radionuclide removal is expected to be achieved during cleaning of the HTF tanks, and the 

cleaning and/or flushing of ancillary equipment, for a number of reasons. As discussed above, DOE has 

extensive experience at the SRS Tanks Farms in removing waste, including HRRs, from various types of 

waste tanks. In HTF, Tank 16 (a Type II tank) and Tank 12 (a Type I tank) have undergone extensive 

waste removal campaigns and are currently in the sample and analysis phase. Bulk waste removal 

efforts have been completed for Tank 11 (a Type I tank), and bulk waste removal efforts are currently 

underway in Tank 10 (a Type I tank) and Tank 13 (a Type II tank); similarly, preparation for bulk waste 

removal efforts in Tank 15 (a Type II tank) are in progress. In addition, DOE has previously removed 

waste (and associated radionuclides) successfully from various HTF waste tanks to support ongoing tank 

Figure 5.3-6: Tank 6 Waste Removal 
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farm and DWPF operations including bulk sludge removal from HTF Tanks 9, 10, 11, 14, 21, 22 and 42, 

and bulk saltcake removal in HTF Tanks 22, 24, 37 and 41.90 Furthermore, DOE has successfully 

removed waste from FTF tanks, including Type IV tanks (including Tanks 18 and 19) as well as Type I 

tanks (Tanks 5 and 6), which present the most challenging conditions in the SRS Tank Farms. DOE 

anticipates achieving comparable removal of HRRs from the HTF waste tanks and ancillary structures at 

the time of closure. Waste removal is primarily influenced by tank type and waste type (e.g., salt waste 

versus sludge), and the factors are primarily not tank farm dependent. The waste tank types between the 

two tank farms, FTF and HTF, are essentially identical and the waste removal experience in the FTF Type 

I and Type IV tanks that have been cleaned provides indicia of the level of waste removal that DOE 

anticipates achieving in the HTF waste tanks. In addition, DOE’s experience to date in the cleaning of 

Tank 16, a Type II tank, and Tank 12, a Type I tank, in HTF, have shown similar levels of waste removal 

to that achieved in FTF. The cleaning process employed is thorough, and the process is reviewed and 

documented during cleaning to maximize practical effectiveness, as explained in Appendix B. 

DOE will continue to use such measures as visual (remote) observation of remaining tank residuals 

against benchmarks in the waste tank primary tank or annulus, as applicable (or ancillary equipment), 

transfer line radiation readings, sampling and analysis, radiation monitoring, and equipment operating 

parameters to evaluate efficiency and effectiveness of cleaning operations. Moreover, removal activities 

on a given tank or ancillary structure will not be considered complete until it is clearly demonstrated and 

documented, for each individual tank or ancillary structure, that further deployment of the technology is no 

longer useful or sensible, and that other proven technologies have been evaluated and would not be 

practical. These documented considerations will take into account a variety of factors including such 

things as the conditions in the specific waste tank or ancillary structure, the status of the HTF and the 

overall Liquid Waste System (e.g., available waste tank volume), available proven technologies, the 

potential benefits from long-term risk reduction from continued HRR removal, increased radiation 

exposure to site workers or the public due to removal activities, increased risk associated with impacts to 

other DOE missions involving risk-reducing activities, direct monetary expenditures and effectiveness of 

available technologies.91 [DOE/SRS-WD-2011-001] 

Furthermore, DOE has a well-documented, contractually required process in place, which requires that 

throughout the waste tank or ancillary structure cleaning process, numerous reports, evaluations, 

analyses, data, operational documents, and cost-benefit analyses must be developed for each waste tank 

and applicable ancillary structure to support completion of waste removal activities. This process is 

described in detail in Appendix B and has been successfully used to date.92 As outlined in Appendix B, 

DOE documents final waste inventories, radionuclide removal effectiveness (with an emphasis on HRRs), 

and final cost-benefit analysis in a final removal report for each waste tank and applicable ancillary 

structure. Documentation and information collected from each phase of the removal process eventually 

contributes to the final removal report, with an emphasis on removal of HRRs, which supports and is 

required before DOE provides authorization and approval to stabilize (grout) each tank or ancillary 

structure.93 This process provides further confidence that HRRs will have been removed to the MEP from 

the HTF waste tanks and applicable ancillary structures at closure. 

 

90 As explained previously, the available storage space created within these tanks was subsequently re-used to support ongoing 

tank farm operations. DOE is or will be further removing waste, including HRRs, from these tanks to support closure. 91 Typically, the cost-benefit analysis will be relatively simple and will focus on the financial costs for implementation of new

technologies versus the decrease in the potential future doses resulting from the additional removal of residuals. [NUREG-1854] 

For future cost-benefit analyses for HTF tanks, DOE will follow a similar approach to that documented in Documentation of Removal 

of Highly Radioactive Radionuclides in Tanks 5 and 6 (SRR-CWDA-2012-00138) to address NRC’s observations on TER pages 3-

52 that “It remains unclear to the NRC staff if DOE will follow a consistent format with the appropriate content for the future costbenefit analyses”, and NRC’s Key Recommendation ES-2.7 on page xviii which states, “The NRC staff recommends that DOE 

continue to better define the documented process to be used to demonstrate removal to the MEP to ensure consistent (nonarbitrary) application of the criterion.” [ML14094A496] 

92 This process has been implemented successfully for Tanks 5, 6, 18 and 19 in FTF. 93 In addition, SCDHEC and EPA must concur in the suspension of waste removal activities and closure of each tank and applicable 

ancillary structure in accordance with the FFA. 
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5.4 Conclusion 

Removal of HRRs to the MEP in HTF waste tanks and ancillary structures occurs through a systematic 

progression of waste removal and cleaning activities using proven technologies to a point where further 

removal of HRRs is not sensible or useful in light of the overall benefit to human health, safety and the 

environment. The preceding subsections demonstrate that the HTF waste tanks, ancillary structures and 

their associated stabilized residuals will have had HRRs removed to the MEP at the time of closure. 
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6.0 RADIONUCLIDE CONCENTRATIONS OF STABILIZED RESIDUALS, TANKS AND 

ANCILLARY STRUCTURES 

Section Purpose 

The purpose of this section is to demonstrate whether the HTF stabilized residuals at 

closure will meet concentration limits for Class C low-level waste as set out in 10 CFR Part 

61, Section 61.55. 

Section Contents 

This section provides the methodology and assumptions to demonstrate whether the HTF 

stabilized residuals at closure meet Class C concentration limits. 

Key Points 

 DOE is using the NRC guidance in NUREG-1854, Category 3 – Site-Specific Averaging 

in its approach to determining whether the stabilized residuals meet Class C 

concentration limits. 

 The Category 3 approach involves the use of the site-specific intruder-driller scenarios 

analyzed in the HTF PA. 

 In addition, DOE, in previous consultation with the NRC,94 has derived site-specific 

concentration averaging expressions for HTF waste based upon the site-specific 

intruder-driller scenarios and the guidance in NUREG-1854. 

 While DOE believes there is a reasonable basis to conclude that none of the stabilized 

residuals, tanks and ancillary structures will exceed the Class C concentration limits in 

10 CFR 61.55, DOE nevertheless has consulted with the NRC on DOE’s disposal plans, 

as described in this HTF 3116 Basis Document, to take full advantage of the NDAA 

Section 3116 consultation process. 

The NDAA Section 3116(a) provides in pertinent part: 

[T]he term “high-level radioactive waste” does not include radioactive waste resulting from the 

reprocessing of spent nuclear fuel that the Secretary of Energy..., in consultation with the Nuclear 

Regulatory Commission..., determines – 

(3)(A) does not exceed concentration limits for Class C low-level waste as set out in section 

61.55 of title 10, Code of Federal Regulations, and will be disposed of— 

(i) in compliance with the performance objectives set out in subpart C of part 61 of title 

10, Code of Federal Regulations; and 

(ii) pursuant to a State-approved closure plan or State-issued permit, authority for the 

approval or issuance of which is conferred on the State outside of this section; or 

(B) exceeds concentration limits for Class C low-level waste as set out in section 61.55 of 

title 10, Code of Federal Regulations, but will be disposed of— 

(i) in compliance with the performance objectives set out in subpart C of part 61 of title 

10, Code of Federal Regulations; and 

(ii) pursuant to a State-approved closure plan or State-issued permit, authority for the 

approval or issuance of which is conferred on the State outside of this section; and 

(iii) pursuant to plans developed by the Secretary in consultation with the Commission. 

 

94 NRC suggested during a public Draft FTF 3116 Basis Document scoping meeting held on July 13-14, 2010 that DOE consider 

this approach. [SRR-CWDA-2010-00091] On page 85 of its FTF TER, NRC stated, “NRC staff has evaluated DOE’s methodology 

for classifying waste and finds the approach an acceptable application of category 3 in NUREG-1854.” [ML112371715] The 

methodology described in this HTF 3116 Basis Document is consistent with the methodology used to support the FTF 3116 Basis 

Document. 
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6.1 Waste Concentrations 

For the purposes of making a determination under NDAA Section 3116(a)(3), regardless of whether the 

waste exceeds or does not exceed the concentration limits for Class C low-level waste as set out in 10 

CFR 61.55, the Secretary of Energy, in consultation with the NRC, must determine that the waste will be 

disposed of in compliance with the performance objectives of 10 CFR 61, Subpart C, and that the waste 

will be disposed of in accordance with State-approved closure plans. In Section 7.0 of this HTF 3116 

Basis Document, information is presented that demonstrates that the waste will be disposed of in 

compliance with the performance objectives of 10 CFR 61, Subpart C. In Section 8.0 of this HTF 3116 

Basis Document, information is presented that demonstrates that the waste will be disposed of in 

compliance with State-approved closure plans. 

In situations where the waste exceeds the concentration limits for Class C low-level waste, NDAA Section 

3116(a)(3)(B)(iii) provides for consultation with NRC about the disposal plans for the waste. 

[NDAA_3116] 

As discussed in this section, under DOE’s disposal plans, the stabilized residuals within the waste tanks 

and ancillary structures, the waste tanks, and the ancillary structures (including integral equipment) in the 

HTF are not expected to exceed concentration limits for Class C low-level waste. Nevertheless, DOE has 

consulted with the NRC pursuant to the consultation process in NDAA Section 3116(a)(3)(B)(iii) to take 

full advantage of the consultation process established by NDAA Section 3116. In this regard, DOE 

specifically requested in the Draft HTF 3116 Basis Document that NRC identify what changes, if any, 

NRC would recommend to DOE’s disposal plans as described in the Draft HTF 3116 Basis Document. 

[DOE/SRS-WD-2013-001] As part of NRC’s consultative review of the Draft HTF 3116 Basis Document, 

the NRC reviewed DOE’s disposal plans, did not recommend any specific changes to the disposal plans, 

and concluded that the 

requirements of 3116(a)(3)(B)(iii) 

were satisfied, should the waste 

exceed concentration limits for 

Class C low-level waste.95 In the 

following subsections, the 

methodology for comparison to 

the Class C concentration limits 

for radionuclides included in 10 

CFR 61.55 is presented. The 

radionuclides and their 

associated limits are specified in 

two separate tables within 10 

CFR 61.55 which are reproduced 

in Table 6.1-1 and Table 6.1-2. 

 

 

95 On page 4-8 of its TER, NRC states “The NRC staff has reviewed DOE’s preliminary disposal plans for the HTF waste as part of 

the consultation process that is documented in this TER, thereby satisfying the requirements of Section 3116(a)(3)(B)(iii) for tanks 

containing residual waste that might exceed Class C concentrations at the time of preparation of this TER. Should DOE find that

HTF waste exceeds Class C limits based on final inventories developed in the future, the NRC staff could provide further 

recommendations related to DOE’s final disposal plans, if requested by DOE.” [ML14094A496] 

Table 6.1-1: 10 CFR 61.55 Table 1 Class C Concentration Limits 

Radionuclides (Long-lived) Concentration (Ci/m3) 

C-14 8 

C-14 in activated metal 80 

Ni-59 in activated metal 220 

Nb-94 in activated metal 0.2 

Tc-99 3 

I-129 0.08 

Alpha Emitting Transuranic nuclides with half-life 

greater than five years 

1

100 

Pu-241 1

3,500 

Cm-242 1

20,000 

(1) Units are in nanocuries per gram. 

 [10 CFR 61]
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Table 6.1-2: 10 CFR 61.55 Table 2 Class C Concentration Limits 

Concentration (Ci/m3

) 

Radionuclides (Short-lived) 

Column 1 

[Class A] 

Column 2 

[Class B] 

Column 3 

[Class C] 

Total of all nuclides with less than 5 year half-life 700 (1) (1) 

H-3 40 (1) (1) 

Co-60 700 (1

) (1) 

Ni-63 3.5 70 700 

Ni-63 in activated metal 35 700 7000 

Sr-90 0.04 150 7000 

Cs-137 1 44 4600 

(1) There are no limits established for these radionuclides in Class B or C wastes. Practical considerations such as the effects of external radiation 

and internal heat generation on transportation, handling and disposal will limit the concentrations for these wastes. These wastes shall be Class 

B unless the concentrations of other nuclides in the table determine the waste to be Class C independent of these nuclides. 

 [10 CFR 61]

6.2 Approach to Waste Concentrations for H-Tank Farm Residuals 

Prior NRC guidance to determine concentrations for comparison with Class C concentration limits of 10 

CFR 61.55 was based on excavation as the likely pathway to expose an inadvertent member of the public 

to waste in a commercial shallow land burial site. [NUREG-1854] Due to the disposal depth of the HTF 

stabilized residuals in the waste tanks and the ancillary structures, the basement excavation scenario 

associated with development of 10 CFR 61.55 Table 1 and 2 is not applicable to the HTF waste tanks 

and ancillary structures. A more appropriate scenario for the purposes of calculation and comparison 

with Class C concentration limits is one that assumes the inadvertent intruder drills a groundwater well 

and drills through a waste tank or ancillary structure (Figure 6.2-1). 

Consistent with more recent NRC staff guidance, this HTF 3116 Basis Document follows the Category 

3—Site-Specific Averaging approach set forth in NUREG-1854, using the intruder-drilling scenario. 

[NUREG-1854] This 

approach utilizes a riskinformed approach that 

takes into consideration 

such things as the 

specific conditions of the 

HTF site, the final form 

of the stabilized 

residuals, site-specific 

parameters and the final 

closure configuration. 

The following section 

describes the 

methodology, and 

presents the inputs and 

assumptions, DOE used 

to compare the 

concentration of the 

stabilized residuals in 

HTF at closure to the 

Class C concentration 

limits. 

 

Figure 6.2-1: Intruder-Driller Scenarios for Concentration Calculations 
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6.3 Methodology 

The Category 3—Site-Specific Averaging approach to concentration averaging reflects site-specific 

conditions of HTF and the final form of the stabilized residuals to account for the volume, concentration 

and accessibility of the residual material. In order to account for the site-specific conditions relative to 

HTF, DOE has developed, consistent with the Category 3—Site-Specific Averaging methodology, 

averaging expressions for HTF based on the results of the inadvertent intruder analysis performed within 

the HTF PA. [SRR-CWDA-2010-00128] As discussed in the following sections, the concentrations of the 

stabilized residuals have been compared, utilizing these averaging expressions, against the concentration 

limits for Class C low-level waste as set out in 10 CFR 61.55 Table 1 and Table 2. For the waste tanks 

and the ancillary structures, this comparison was based on the projected inventories at closure in the HTF 

PA. 

For purposes of comparison to the Class C concentration limits, and to align with the inputs used in 

developing the averaging expressions for HTF, the residual inventory used for these calculations are 

decayed to the inventory that will be present at the time of closure (assumed to be 2032 for the purposes 

of analysis in the HTF PA). [SRR-CWDA-2010-00128] As discussed below, the radionuclide 

concentrations of the stabilized residuals are compared, using the sum of fractions methodology and the 

HTF averaging expressions, to the concentration limits for Class C low-level waste as set out in 10 CFR 

61.55 Table 1 and Table 2. 

In order to demonstrate compliance with, among other things, the performance objectives set out in 10 

CFR Part 61, Subpart C, as required by NDAA Section 3116(a)(3), DOE developed a performance 

assessment covering closure activities within HTF. [SRR-CWDA-2010-00128] To demonstrate 

compliance with 10 CFR 61.42, the HTF PA is used to demonstrate that there is reasonable assurance 

the dose to an inadvertent intruder will remain below 500 mrem/yr taking into consideration a variety of 

intruder scenarios. DOE utilized the inadvertent human intruder analysis in the HTF PA to develop the 

HTF averaging expressions used for waste classification. 

The HTF PA models used to simulate the performance of the HTF closure system take into account the 

release of radiological contaminants from the waste tanks and the associated ancillary structures in the 

HTF and simulates transport of the radiological contaminants through soil and groundwater to the 

assessment point. The models use numerous HTF-specific input parameters to represent the HTF 

closure system behavior over time. Many of the input parameters are based on site-specific data (e.g., 

soil and cementitious materials distribution coefficient (Kd) values) used in transport modeling. In 

addition, site-specific information is used to model the behavior of individual barriers within the HTF, such 

as the waste tank carbon steel primary tanks and secondary liners (as applicable), and cementitious 

barriers. Numerous bioaccumulation factors (e.g., soil-to-plant transfer factors), human health exposure 

parameters (e.g., water ingestion rates, vegetable consumption data) and dose conversion factors are 

used in the computer modeling to calculate doses for each of the exposure pathways. All of these 

parameters factor into development of the HTF averaging expressions. A detailed discussion of the HTF 

PA intruder analyses is provided in Section 7.2 of this HTF 3116 Basis Document and the HTF PA. 

[SRR-CWDA-2010-00128] 

The stabilized contaminant materials after HTF closure will be primarily located in areas protected by 

significant materials (e.g., grouted waste tanks, diversion box cell covers and valve box shielding) which 

are clearly distinguishable from the surrounding soil and make drilling an unlikely scenario based on 

regional drilling practices. Regional drilling conditions are such that a barrier such as the closure cap 

erosion barrier, tank top or grout fill are situations that would cause drillers to stop operations and move 

drilling location. The most vulnerable location for stabilized residuals is in a transfer line which may be 

near grade-level prior to closure and are of a small size (typically a three-inch diameter or less) which 

makes them the most credible stabilized contaminants vulnerable during any intruder drilling operations 

although the probability of hitting a transfer line is small due to the small surface area of transfer lines 

versus the large HTF footprint. However, for the purposes of developing averaging expressions for HTF, 

it is assumed that the structures would be penetrated and that construction of the well would be 

completed. 

The following subsections describe how the sum of fractions is calculated for the HTF waste tanks and 

ancillary structures. 
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6.3.1 Methodology Inputs 

The following inputs are used for the concentration calculations. Generally, the inputs and assumptions 

underestimate or do not take credit for certain masses or volumes that would lower the calculated 

radionuclide contribution to the sum of fractions. 

 The residual inventory used for the concentration calculations is the total inventory of the residual 

material within the waste tank or ancillary structure and includes all material on the floor, walls, 

annulus, sand layer, piping, cooling coils and any structure that will be left in the tank. 

 The residual material layer in the waste tanks and ancillary structures, with the exception of 

transfer lines, is assumed to be spread evenly across the floor of the waste tank or ancillary 

structure. The residual material within transfer lines is assumed to be spread evenly over the 

internal surface of the transfer line. 

 The volume of the residual material used in the calculations will be determined on an individual 

basis for each waste tank or ancillary structure at the time it is being removed from service. 

 For the purpose of calculating mass-based concentrations, the density of the residual material 

within the waste tanks is assumed to be the same as the grout used for stabilization. 

 Site-specific averaging expressions for HTF, as described in Section 6.3.2, are utilized for 

comparison against the concentration limits for Class C low-level waste as set out in Table 1 and 

Table 2 of 10 CFR 61.55. 

 The HTF averaging expressions utilizing transfer line volumes, mass and surface area, have 

been established on the basis of one linear foot of transfer line piping. This basis was used only 

for the purpose of standardizing the format for use in performing the calculation and does not 

impact the calculated radionuclide contributions to the sum of fractions. The ratio between these 

parameters is constant and the length of pipe does not impact the results of the equations. 

 The projected inventories for the ancillary structures (other than the transfer lines discussed 

above) are bounded by the residuals projected for the waste tanks. [SRR-CWDA-2010-00128] 

6.3.2 Site-Specific HTF Waste Concentration Calculation Averaging Expressions 

As described above, the Category 3—Site-Specific Averaging approach to concentration averaging 

contemplates consideration of site-specific conditions of HTF and the stabilized residuals. In 

development of Table 1 and 2 of 10 CFR 61.55, the underlying assumption was that the concentration 

limits and disposal requirements ensure that an inadvertent intruder (e.g., assuming excavation to a depth 

of 10 feet for construction of a house) would not receive a dose exceeding an equivalent of 500 mrem/yr 

to the whole body.96 At closure, the depth of the stabilized residuals within the HTF waste tanks and 

ancillary structures will be well below (i.e., greater than 10 feet) the HTF closure cap and a robust intruder 

barrier (e.g., grouted waste tanks, diversion box cell covers and valve box shielding), as described in the 

HTF PA, will be in place. [SRR-CWDA-2010-00128] Therefore, the intruder-construction scenario is 

considered inapplicable. Based on the depth to the stabilized residuals and the presence of a robust 

intruder barrier, the “Deep waste, intruder barrier” scenario from Table 3-2 of NUREG-1854 is being 

utilized. In order to account for the site-specific conditions relative to HTF, site-specific averaging 

expressions for HTF, based on the results of the Inadvertent Intruder Analyses performed within the HTF 

PA, have been developed. 

The HTF PA provides the estimated dose to an intruder who resides within the boundary of the HTF after 

the period of institutional control (100 years). The intruder is assumed to be exposed via various 

pathways from water collected from a one-meter well and from drill cuttings. The groundwater associated 

with the one-meter well is contaminated from all the sources within the HTF (waste tanks, transfer lines 

and other ancillary structures). In addition, drill cuttings that pull up contamination from striking a transfer 

line are deposited on the ground surface. The Base Case in the HTF PA assumes that a three-inch 

transfer line is penetrated by a driller and the cuttings are spread among the garden – thus an additional 

 

96 NUREG-1854 states, “Although multiple scenarios were considered, the limiting intruder scenario that was used (in deriving the

concentration limits for waste classification found in Table 1 and Table 2 of 10 CFR 61.55) was an intruder construction scenario 

(NRC, 1981). This scenario involved excavation of a foundation for a house. Approximately 232 m3

 [8,190 ft3] of waste was 

assumed to be exhumed, and the excavation was assumed to be to a depth of 3 m [10 ft]...The underlying assumption of the values 

in Tables 1 and 2 of 10 CFR 61.55 is that the concentration limits and disposal requirements ensure that an inadvertent intruder 

would not receive a dose exceeding an equivalent of 5 mSv [500 mrem] to the whole body.” [NUREG-1854] 
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source is added to the contaminated well source. The impact of drilling into a four-inch transfer line has 

been presented in the HTF PA with respect to the chronic intruder. The impact of drilling into a waste 

tank was also considered in the HTF PA with respect to the acute intruder, the well driller. Since the 

likelihood of a well driller penetrating a waste tank is very remote based on local drilling practices that 

would terminate the drilling once significant resistance is encountered, a chronic intruder was not 

assessed. [SRR-CWDA-2010-00128] 

Because the stabilized residuals in HTF at closure are expected to have multiple radionuclides from 

Table 1 and Table 2 of 10 CFR 61.55, the sum of fractions approach for comparing to Class C 

concentration limits was applied. The sum of fractions approach requires that the concentration of each 

of the Table 1 and Table 2 radionuclides contained in the stabilized residuals be divided by the 

appropriate Table 1 or Table 2 Class C concentration limit. The resulting fraction for each of the 

radionuclides are then totaled for the applicable 10 CFR 61.55 Table 1 or Table 2 radionuclides. If the 

sum of the fractions is less than 1.0 for the individual tables, the waste is below the Class C concentration 

limits set out in 10 CFR 61.55. Consistent with the Category 3—Site-Specific Averaging approach, the 

averaging expressions used to determine the individual radionuclide contribution to the sum of fractions is 

represented by the following equation: 

i

i

R

i x SiteFactor

Table Value

C SOF _ 

where: 

SOFi

 = Radionuclide “i” contribution to the sum of the fractions 

CR = Concentration of the drilled source for radionuclide “i” at closure

Table_Valuei

 = Class C concentration limit from 10 CFR 61.55 Table 1 or 

Table 2 for radionuclide “i” 

Site Factori

 = Site-specific factor for radionuclide “i” based on site-specific 

conditions within the HTF after closure 

The HTF averaging expressions, based on the above equation, that DOE is utilizing are shown below. 

6.3.2.1 HTF Waste Tank Waste Concentration Calculation Averaging Expressions 

For HTF waste tanks individual radionuclide concentrations for the sum of the fractions calculations are 

determined with the following equations: 

For volume-based concentrations: 

i

R

R

i

i x Site Factor

V

I

x

Table Value

SOF _

1 

where: 

SOFi

 = Radionuclide “i” contribution to the sum of the fractions 

Table_Valuei

 = Class C concentration limit in Ci/m3 from 10 CFR 61.55 Table 1 

or Table 2 for radionuclide “i” 

IR = Total tank residuals inventory for radionuclide “i” decayed to date 

of closure (i.e., 2032), units in curies

VR = Total volume of residuals remaining in the waste tank, units in m3

 

Site Factori

 = Site-specific factor for radionuclide “i” at closure (Table 6.3-1 or 

Table 6.3-2, see Section 6.3.2.3 for derivation of site-specific 

factors) 
Basis for Section 3116 Determination DOE/SRS-WD-2014-001 

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at the Savannah River Site December 2014 

Page 6-7 

For mass-based concentrations: 

i

R G

R

i

i x Site Factor

V

I

x

Table Value

SOF _ ( ) ( ) (1,000,000)

1

   

where: 

SOFi

 = Radionuclide “i” contribution to the sum of the fractions 

Table_Valuei

 = Class C concentration limit in nCi/g from 10 CFR 61.55 Table 1 

for radionuclide “i” 

IR = Total tank residuals inventory for radionuclide “i” decayed to date 

of closure (i.e., 2032), units in nanocuries 

VR = Total volume of residuals remaining in the waste tank, units in m3 

ρG = Density of stabilizing grout, units in g/cm3 

Site Factori

 = Site-specific factor for radionuclide “i” at closure (Table 6.3-1, 

see Section 6.3.2.3 for derivation of site-specific factors) 

The calculated fractions are totaled for the applicable 10 CFR 61.55 Table 1 or Table 2 radionuclides. If 

the sum of the fractions is less than 1.0 for the individual tables, the waste is below the Class C 

concentration limits set out in 10 CFR 61.55. 

6.3.2.2 HTF Transfer Line Waste Concentration Calculation Averaging Expressions 

For HTF transfer lines the individual radionuclide concentrations for the sum of the fractions calculations 

are determined with the following equations: 

For volume-based concentrations: 

i

TL

R

i

i x Site Factor

V

I Surface Area

x

Table Value

SOF ( ) ( )

_

1  

where: 

SOFi

 = Radionuclide “i” contribution to the sum of the fractions 

Table_Valuei

 = Class C concentration limit in Ci/m3 from 10 CFR 61.55 Table 1 

or Table 2 for radionuclide “i” 

IR = Transfer line residuals inventory for radionuclide “i” decayed to 

date of closure (i.e., 2032), units in Ci/ft2

Surface Area = Transfer line internal surface area for one linear-foot of transfer 

line piping, units in ft2

VTL = Total volume of piping material for one linear-foot of transfer line 

piping, units in m3 

Site Factori = Site-specific factor for radionuclide “i” at closure (Table 6.3-3 or 

Table 6.3-4, see Section 6.3.2.3 for derivation of site-specific 

factors) 

 
Basis for Section 3116 Determination DOE/SRS-WD-2014-001 

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at the Savannah River Site December 2014 

Page 6-8 

For mass-based concentrations: 

i

TL

R

i

i x Site Factor

M

I Surface Area

x

Table Value

SOF ( ) ( )

_

1  

where: 

SOFi

 = Radionuclide “i” contribution to the sum of the fractions 

Table_Valuei

 = Class C concentration limit in nCi/g from 10 CFR 61.55 Table 1 

for radionuclide “i” 

IR = Transfer line residuals inventory for radionuclide “i” decayed to 

date of closure (i.e., 2032), units in nCi/ft2

Surface Area = Transfer line internal surface area for one linear-foot of transfer 

line piping, units in ft2

MTL = Total mass of piping material for one linear-foot of transfer line 

piping, units in g 

Site Factori

 = Site-specific factor for radionuclide “i” at closure (Table 6.3-3, 

see Section 6.3.2.3 for derivation of site-specific factors) 

The calculated fractions are totaled for the applicable 10 CFR 61.55 Table 1 or Table 2 radionuclides. If 

the sum of the fractions is less than 1.0 for the individual tables, the waste is below the Class C 

concentration limits set out in 10 CFR 61.55. 

6.3.2.3 Site-Specific Factors for Use in HTF Averaging Expressions 

The site-specific factors in the HTF averaging expressions discussed previously were developed to 

account for site-specific conditions while ensuring the same protection as the concentration limits in 

Table 1 and Table 2 and the Part 61.55 analysis provides. To develop the site-specific factors, the results 

of the HTF PA inadvertent intruder analyses along with the HTF PA inventory at closure were utilized. 

The HTF PA deterministic model and its associated dose calculation methodology was utilized to 

determine the dose to the chronic intruder assuming the one-meter well contaminated source and one of 

three drill cutting sources including a three-inch diameter transfer line, a four-inch diameter transfer line or 

a waste tank. Because it is a major contributor to the peak dose from the one-meter well in the HTF, 

Tank 13 was used for determining waste tank site-specific factors. The peak dose for each radionuclide, 

regardless of the time of the peak, was determined and site-specific factors were developed based on the 

assumed concentrations at closure from the drill cutting source. For the waste tank analysis, the timeperiod evaluated started at 500 years after closure and for transfer lines the time-period started at 100 

years after closure. Although the HTF closure design does provide a robust intruder barrier that would 

prevent intrusion into the waste for the first 500 years after closure, for conservatism, the HTF PA 

evaluated the transfer line scenario beginning at 100 years after closure. Therefore, in selecting the peak 

doses for the transfer line scenarios, the peak dose was not just the dose after 500 years but also 

included any individual radionuclide peaks which may have occurred between the 100- and 500-year 

period. 

To determine, based on the inadvertent intruder analysis performed within the HTF PA, the individual 

radionuclide site-specific factors that would result in an inadvertent intruder under the HTF site-specific 

conditions receiving an equivalent dose, 500 mrem/yr, to that used in developing the 10 CFR 61.55 

concentration limits, the following equation was used: 

mrem yr

Dose

x

C

Table Value Site Factor i

PA

i

i 500 /

_ 

 
Basis for Section 3116 Determination DOE/SRS-WD-2014-001 

for Closure of H-Tank Farm Revision 0 

at the Savannah River Site December 2014 

Page 6-9 

where: 

Site Factori

 = Site-specific factor for radionuclide “i” at closure 

Table_Valuei = Class C concentration limit from 10 CFR 61.55 Table 1 or Table 

2 for radionuclide “i” 

CPA = Concentration, based on the HTF PA inventory at closure, of the 

drilled source for radionuclide “i” 

Dosei = Peak dose, based on results of the HTF PA, that occurs beyond 

100 years (for transfer lines) or beyond 500 years (for waste 

tank) after closure, for radionuclide “i”, units in mrem/yr 

Using values for the HTF PA closure inventory along with the mass, for mass based limits, and volume, 

for volume based limits, of the residual material, the radionuclide concentrations at closure were 

determined. The calculated concentration for each radionuclide, the peak dose for each radionuclide and 

the equation developed above, were then used to determine the site-specific factors, on a radionuclide 

basis, for the three different sources (i.e., three-inch diameter transfer line, a four-inch diameter transfer 

line or a waste tank) to be used in the site-specific averaging expression as shown in Section 6.3.2.1 and 

6.3.2.2. 

Recognizing that the peak dose for a specific radionuclide may be dominated by the contaminated 

groundwater source and not the drilling source, the site-specific factor for the transfer lines was set at the 

limiting value based on either the three-inch diameter transfer line or the four-inch diameter transfer line. 

Table 6.3-1 and Table 6.3-2 provide the HTF waste tank site-specific factors for 10 CFR Table 1 and 

Table 2 radionuclides, respectively, which are used in the HTF averaging expressions. Table 6.3-3 and 

Table 6.3-4 provide the HTF transfer line site-specific factors for 10 CFR Table 1 and Table 2 

radionuclides, respectively, which are used in the HTF averaging expressions. [SRR-CWDA-2012-00109] 

Table 6.3-1: HTF Waste Tank Site-Specific Factors for 10 CFR 61.55 Table 1 

Radionuclides 

Table 1 

Radionuclide 

Site-Specific 

Factor 

C-14 3.0E-02 

Ni-59 1.0E-01 

Nb-94 8.4E-03 

Tc-99 5.8E-01 

I-129 7.0E+00 

Np-237 9.3E-03 

Pu-238 9.5E-06 

Pu-239 2.2E-03 

Pu-240 9.2E-04 

Pu-241 4.9E-04 

Pu-242 1.3E-02 

Pu-244 1.3E-02 

Am-241 4.5E-04 

Am-242m 2.0E-03 

Am-243 2.5E-02 

Cm-243 1.4E-06 

Cm-244 4.8E-06 

Cm-245 1.8E-02 

Cm-247 3.1E-02 

Cm-248 2.5E-01 

Cf-249 3.9E-04 

Cf-251 6.1E-04 

Note: Radionuclides listed are from SRR-CWDA-2010-00023. Only the 10 CFR 61.55 Table 1 radionuclides in the 

tank inventory are listed. 
Basis for Section 3116 Determination DOE/SRS-WD-2014-001 

for Closure of H-Tank Farm Revision 0 

at the Savannah River Site December 2014 

Page 6-10 

Table 6.3-2: HTF Waste Tank Site-Specific Factors for 10 CFR 61.55 Table 2 

Radionuclides 

Table 2 

Radionuclide 

Site-Specific 

Factor 

H-3 (1) 

Co-60 (1)

Ni-63 1.2E-03 

Sr-90 3.1E-03 

Cs-137 8.4E-04 

(1) There are no limits established for these radionuclides in Class B or C wastes. Practical considerations such 

as the effects of external radiation and internal heat generation on transportation, handling and disposal will 

limit the concentrations for these wastes. These wastes shall be Class B unless the concentrations of other 

nuclides in the table determine the waste to be Class C independent of these nuclides. 

Note: Radionuclides listed are from SRR-CWDA-2010-00023. Only the 10 CFR 61.55 Table 2 radionuclides in the 

tank inventory are listed.

Table 6.3-3: HTF Transfer Line Site-Specific Factors for 10 CFR 61.55 

Table 1 Radionuclides 

Table 1 

Radionuclide 

Site-Specific 

Factor 

C-14 1.6E+03 

Ni-59 2.0E+02 

Nb-94 1.7E+02 

Tc-99 2.1E+00 

I-129 8.7E+03 

Np-237 2.0E+01 

Pu-238 3.7E-04 

Pu-239 1.1E+00 

Pu-240 5.3E-01 

Pu-241 4.5E-02 

Pu-242 1.6E+02 

Pu-244 3.7E+04 

Am-241 5.3E-02 

Am-242m 1.5E-02 

Am-243 1.4E+01 

Cm-243 6.4E-05 

Cm-244 9.4E-06 

Cm-245 1.0E+03 

Cm-247 7.3E+11 

Cm-248 5.7E+12 

Cf-249 3.9E+01 

Cf-251 4.7E+08 

Note: Radionuclides listed are from SRR-CWDA-2010-00023. Only the 10 CFR 61.55 Table 1 radionuclides in the 

transfer line inventory are listed. 

 
Basis for Section 3116 Determination DOE/SRS-WD-2014-001 

for Closure of H-Tank Farm Revision 0 

at the Savannah River Site December 2014 

Page 6-11 

Table 6.3-4: HTF Transfer Line Site-Specific Factors for 10 CFR 61.55 

Table 2 Radionuclides 

Table 2 

Radionuclide 

Site-Specific 

Factor 

H-3 (1) 

Co-60 (1)

Ni-63 1.5E-02 

Sr-90 1.5E+01 

Cs-137 4.3E+00 

(1) There are no limits established for these radionuclides in Class B or C wastes. Practical considerations such 

as the effects of external radiation and internal heat generation on transportation, handling and disposal will 

limit the concentrations for these wastes. These wastes shall be Class B unless the concentrations of other 

nuclides in the table determine the waste to be Class C independent of these nuclides. 

Note: Radionuclides listed are from SRR-CWDA-2010-00023. Only the 10 CFR 61.55 Table 2 radionuclides in the 

transfer line inventory are listed. 

6.4 Waste Concentration Calculations 

The following subsections provide calculations of radionuclide concentrations and compare those 

concentrations to the Class C concentration limits set out in 10 CFR 61.55. 

6.4.1 Waste Tank Concentration Calculation 

For this calculation, the best estimate residual radionuclide inventory and residual volume for Tank 32 

based on the HTF PA projected inventory at closure is used. [SRR-CWDA-2010-00023] This tank was 

selected on the basis that it represents the waste tank grouping (i.e., Type III/IIIA-Sludge) that has the 

highest resulting sum of the fractions based on the HTF PA projected inventories. [SRR-CWDA-2012-

00129] The contribution of each radionuclide to the sum of the fractions was calculated using the HTF 

averaging expressions presented in Section 6.3.2.1 for a mass or volume basis as necessary. For 

example, using the inventory values for C-14 and Pu-241, the mass- and volume-based averaging 

expressions from Section 6.3.2.1 become: 

For the volume-based C-14 fraction of Class C concentration limit: 

3.0E 02

15.1

(1.0E 00 )

8.0E 00 /

1

[ 14] 3 3  

   x

m

Ci

x

Ci m

SOFC = 2.5E-04 

For the mass-based Pu-241 fraction of Class C concentration limit: 

4.9E 04

(15.1 ) (1.97 / ) (1.0E 06 / )

(4.6E 12 )

3.5 03 /

1

[ 241] 3 3 3 3    



   x

m g cm cm m

nCi

x

E nCi g

SOF Pu = 2.2E-02 

The remainder of the Table 1 and Table 2 radionuclides are calculated similarly and the results are 

presented in Table 6.4-1 and Table 6.4-2. 

 
Basis for Section 3116 Determination DOE/SRS-WD-2014-001 

for Closure of H-Tank Farm Revision 0 

at the Savannah River Site December 2014 

Page 6-12 

Table 6.4-1: Sum of the Fractions Calculation Using the HTF PA, Revision 1 Inventory for 

Tank 32 (10 CFR 61.55 Table 1 Radionuclides) 

Table 1 

Radionuclide 

Tank 

Inventory (Ci) 

Class C 

Concentration Limita Fraction of Class C 

Concentration Limit 

C-14 1.0E+00 8.0E+00 Ci/m3 2.5E-04 

Ni-59 1.0E+00 2.2E+02 Ci/m3 3.0E-05 

Nb-94 1.1E-01 2.0E-01 Ci/m3 3.1E-04 

Tc-99 9.7E+00 3.0E+00 Ci/m3 1.2E-01 

I-129 6.7E-03 8.0E-02 Ci/m3 3.9E-02 

Np-237 4.0E-01 1.0E+02 nCi/g 1.2E-03 

Pu-238 1.5E+04 1.0E+02 nCi/g 4.8E-02 

Pu-239 2.4E+02 1.0E+02 nCi/g 1.8E-01 

Pu-240 1.5E+02 1.0E+02 nCi/g 4.6E-02 

Pu-241 4.6E+03 3.5E+03 nCi/g 2.2E-02 

Pu-242 1.0E+00 1.0E+02 nCi/g 4.4E-03 

Pu-244 1.0E+00 1.0E+02 nCi/g 4.4E-03 

Am-241 1.1E+03 1.0E+02 nCi/g 1.7E-01 

Am-242m 1.0E+00 1.0E+02 nCi/g 6.7E-04 

Am-243 1.0E+00 1.0E+02 nCi/g 8.4E-03 

Cm-243 1.0E+00 1.0E+02 nCi/g 4.7E-07 

Cm-244 2.2E+03 1.0E+02 nCi/g 3.5E-03 

Cm-245 1.0E+00 1.0E+02 nCi/g 6.0E-03 

Cm-247 1.0E+00 1.0E+02 nCi/g 1.0E-02 

Cm-248 1.0E+00 1.0E+02 nCi/g 8.4E-02 

Cf-249 1.0E+00 1.0E+02 nCi/g 1.3E-04 

Cf-251 1.0E+00 1.0E+02 nCi/g 2.0E-04 

Sum of the Fractions 7.4E-01 

a

 Values from 10 CFR 61.55 Table 1. 

Note: Inventory values are from SRR-CWDA-2010-00023. Only the 10 CFR 61.55 Table 1 radionuclides in the tank inventory are listed. 

 

Table 6.4-2: Sum of the Fractions Calculation Using the HTF PA, Revision 1 Inventory for 

Tank 32 (10 CFR 61.55 Table 2 Radionuclides) 

Table 2 

Radionuclide Tank Inventory (Ci) 

Class C Concentration 

Limit(Ci/m3

)

a Fraction of Class C 

Concentration Limit 

H-3 1.0E+00 (1) NA 

Co-60 1.0E+00 (1) NA 

Ni-63 7.9E+02 7.0E+02 8.9E-05 

Sr-90 2.0E+04 7.0E+03 5.9E-04 

Cs-137 5.5E+03 4.6E+03 6.6E-05 

Sum of the Fractions 7.4E-04 

a

 Values from 10 CFR 61.55 Table 2. 

(1) There are no limits established for these radionuclides in Class C waste. 

NA - Not Applicable 

Note: Inventory values are from SRR-CWDA-2010-00023. Only the 10 CFR 61.55 Table 2 radionuclides in the tank inventory are listed. 

The sum of the fractions from Table 6.4-1 and Table 6.4-2 are 7.4E-01 and 7.4E-04, respectively. Using 

the inventory values assumed for this calculation, the stabilized residuals would not exceed Class C 

concentration limits. 

6.4.2 Transfer Line Concentration Calculation 

For this calculation, the HTF PA estimated residual inventory for a three-inch transfer line is used.97 

[SRR-CWDA-2010-00023] This inventory was selected on the basis that over 98 % of the transfer lines 

located in HTF are three-inch transfer lines or smaller. The contribution of each radionuclide to the sum 

 

97 The inventory used for this calculation is based on the three-inch transfer line inventory developed to support the HTF PA. [SRRCWDA-2010-00023] 
Basis for Section 3116 Determination DOE/SRS-WD-2014-001 

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at the Savannah River Site December 2014 

Page 6-13 

of the fractions was calculated using the averaging expressions presented in Section 6.3.2.2 for a mass 

or volume basis as necessary. For example, using the inventory values for C-14 and Pu-241, and 

assuming schedule 40 three-inch piping, the mass- and volume-based averaging expressions from 

Section 6.3.2.2 become: 

For the volume-based C-14 fraction of Class C concentration limit: 

1.6E 03

4.38E 04

(1.8E 09 / ) (8.03E 01 )

8.0E 00 /

1

3

2 2

[ 14] 3  

  

   xm

Ci ft ft x

Ci m

SOF C = 6.6E-04 

For the mass-based Pu-241 fraction of Class C concentration limit: 

4.5E 02

3,438

(2.2E 05 / ) (8.03E 01 )

3.5 03 /

1 2 2

[ 241]    

   xg

nCi ft ft x

E nCi g

SOF Pu = 6.6E-04 

The remainder of the Table 1 and Table 2 radionuclides are calculated similarly and the results are 

presented in Table 6.4-3 and Table 6.4-4. 

Table 6.4-3: Sum of the Fractions Calculation Using the HTF PA, Revision 1 Inventory for a 

Three-inch Transfer Line (10 CFR 61.55 Table 1 Radionuclides) 

Table 1 

Radionuclide 

3-inch Transfer Line 

Inventory98 (Ci/ft2

) 

Class C 

Concentration Limita Fraction of Class C Concentration Limit 

C-14 1.8E-09 8.0E+00 Ci/m3 6.6E-04 

Ni-59 1.7E-06 2.2E+02 Ci/m3 2.8E-03 

Nb-94 3.9E-10 2.0E-01 Ci/m3 6.1E-04 

Tc-99 1.7E-05 3.0E+00 Ci/m3 2.2E-02 

I-129 1.8E-10 8.0E-02 Ci/m3 3.6E-02 

Np-237 7.5E-08 1.0E+02 nCi/g 3.5E-03 

Pu-238 1.1E-03 1.0E+02 nCi/g 9.5E-04 

Pu-239 1.8E-05 1.0E+02 nCi/g 4.6E-02 

Pu-240 1.1E-05 1.0E+02 nCi/g 1.4E-02 

Pu-241 2.2E-04 3.5E+03 nCi/g 6.6E-04 

Pu-242 3.1E-08 1.0E+02 nCi/g 1.2E-02 

Pu-244 1.4E-10 1.0E+02 nCi/g 1.2E-02 

Am-241 1.4E-04 1.0E+02 nCi/g 1.7E-02 

Am-242m 1.0E-07 1.0E+02 nCi/g 3.5E-06 

Am-243 2.2E-06 1.0E+02 nCi/g 7.2E-02 

Cm-243 5.3E-08 1.0E+02 nCi/g 7.9E-09 

Cm-244 1.8E-05 1.0E+02 nCi/g 4.0E-07 

Cm-245 7.3E-09 1.0E+02 nCi/g 1.7E-02 

Cm-247 1.7E-17 1.0E+02 nCi/g 2.9E-02 

Cm-248 1.8E-17 1.0E+02 nCi/g 2.4E-01 

Cf-249 9.5E-17 1.0E+02 nCi/g 8.7E-12 

Cf-251 3.3E-18 1.0E+02 nCi/g 3.6E-06 

Sum of the Fractions 5.3E-01 

a

 Values from 10 CFR 61.55 Table 1. 

Note: Inventory values are from SRR-CWDA-2010-00023. Only the 10 CFR 61.55 Table 1 radionuclides in the transfer line inventory are listed. 

 

 

98 See footnote 97. 
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for Closure of H-Tank Farm Revision 0 

at the Savannah River Site December 2014 

Page 6-14 

Table 6.4-4: Sum of the Fractions Calculation Using the HTF PA, Revision 1 Inventory for a 

Three-inch Transfer Line (10 CFR 61.55 Table 2 Radionuclides) 

Table 2 

Radionuclide 

3-inch Transfer Line 

Inventory99 (Ci/ft2

) 

Class C Concentration 

Limit(Ci/m3)

 a Fraction of Class C 

Concentration Limit 

H-3 5.3E-07 (1) NA 

Co-60 3.7E-06 (1) NA 

Ni-63 1.2E-04 7.0E+02 4.7E-06 

Sr-90 2.7E-02 7.0E+03 1.1E-01 

Cs-137 6.1E-03 4.6E+03 1.0E-02 

Sum of the Fractions 1.2E-01 

a

 Values from 10 CFR 61.55 Table 2. 

(1) There are no limits established for these radionuclides in Class C waste. 

NA - Not Applicable 

Note: Inventory values are from SRR-CWDA-2010-00023. Only the 10 CFR 61.55 Table 2 radionuclides in the tank inventory are listed. 

The sum of the fractions from Table 6.4-3 and Table 6.4-4 are 5.3E-01 and 1.2E-01, respectively. Using 

the inventory values assumed for this calculation, the stabilized residuals in this three-inch transfer line 

would not exceed Class C concentration limits. 

6.5 Conclusion 

As demonstrated by the above discussion, the stabilized HTF wastes at closure are anticipated to meet 

concentration limits for Class C low-level waste as set out in 10 CFR 61.55. Nevertheless, DOE has 

consulted with the NRC on its disposal plans for HTF pursuant to the consultation process in NDAA 

Section 3116(a)(3)(B)(iii) to take full advantage of the consultation process established by NDAA Section 

3116. Therefore, DOE concludes that NDAA criteria can be met under either NDAA Section 

3116(a)(3)(A) or (a)(3)(B).100

 

99 See footnote 97. 100 On page 4-8 of its TER, NRC states, “The NRC staff has reviewed DOE’s preliminary disposal plans for the HTF waste as part of 

the consultation process that is documented in this TER, thereby satisfying the requirements of Section 3116(a)(3)(B)(iii) for tanks 

containing residual waste that might exceed Class C concentrations at the time of preparation of this TER. Should DOE find that

HTF waste exceeds Class C limits based on final inventories developed in the future, the NRC staff could provide further 

recommendations related to DOE’s final disposal plans, if requested by DOE.” [ML14094A496] 
Basis for Section 3116 Determination DOE/SRS-WD-2014-001 

for Closure of H-Tank Farm Revision 0 

at the Savannah River Site December 2014 

Page 7-1 

7.0 THE WASTE WILL BE DISPOSED OF IN ACCORDANCE WITH THE PERFORMANCE 

OBJECTIVES SET OUT IN 10 CFR 61, SUBPART C 

Section Purpose 

The purpose of this section is to demonstrate that the stabilized residuals in the HTF 

waste tanks and ancillary structures, the waste tanks, and the ancillary structures 

(including integral equipment) will be disposed of in compliance with the performance 

objectives for land disposal of low-level waste found in 10 CFR Part 61, Subpart C, 

Sections 61.41 through 61.44. 

Section Contents 

This section describes key parameters and results from the HTF PA that demonstrate 

compliance with the performance objectives in 10 CFR 61.41 and 10 CFR 61.42, DOE 

regulatory and contractual requirements which ensure compliance with 10 CFR 61.43 and 

relevant factors of HTF siting, design, use, operation and closure, which ensure 

compliance with 10 CFR 61.44. 

Key Points 

 Based on the HTF PA there is reasonable assurance that the 10 CFR 61.41 and 10 CFR 

61.42 performance objectives will be met. 

 The DOE is using an assumed institutional control period of 100 years for the purpose 

of analysis although the SRS Land Use Plan [PIT-MISC-0041] calls for federal 

ownership in perpetuity. 

 For the purpose of calculating doses to a member of the public, a 100-meter buffer 

zone around the HTF boundary is assumed. 

 The HTF PA analysis demonstrates compliance with the performance objective in 10 

CFR 61.41 based on compliance with a 25 mrem/yr peak all-pathways Total Effective 

Dose Equivalent (TEDE) to a hypothetical member of the public. 

 The HTF PA analysis demonstrates compliance with the performance objective in 10 

CFR 61.42 based on consideration of a dose of 500 mrem/yr to a future hypothetical 

inadvertent intruder of the closed HTF.

 The DOE regulatory and contractual requirements for HTF facilities and activities 

establish dose limits based on 10 CFR 835 and relevant DOE Orders. These dose 

limits correspond to the radiation protection standards set out in 10 CFR 20, as crossreferenced in 10 CFR 61.43.

 The HTF waste tanks will be filled with grout to provide long-term stability. 

 The HTF ancillary structures may be filled with appropriate fill materials, as necessary, 

to provide long-term stability. There are currently no plans to grout or fill the HTF 

transfer lines. 

The NDAA Section 3116(a) provides in pertinent part: 

[T]he term “high-level radioactive waste” does not include radioactive waste resulting from the 

reprocessing of spent nuclear fuel that the Secretary of Energy..., in consultation with the Nuclear 

Regulatory Commission..., determines – 

(3)(A)(i) [Will be disposed of] in compliance with the performance objectives set out in subpart 

C of part 61 of title 10, Code of Federal Regulations. 

The 10 CFR 61, Subpart C, Sections 61.41 through 61.44 detail performance objectives the NRC 

established for land disposal of radioactive waste. These performance objectives address protection of 

the general population from radioactivity releases (10 CFR 61.41), individuals from inadvertent intrusion 

on the disposal site (10 CFR 61.42), protection of workers and the public during disposal facility 

operations (10 CFR 61.43) and the stability of the disposal site after closure (10 CFR 61.44). 
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10 CFR 61.40 states: 

Land disposal facilities must be sited, designed, operated, closed, and controlled after closure so 

that reasonable assurance exists that exposures to humans are within the limits established in 

the performance objectives in §§61.41 through 61.44. 

10 CFR 61.40 requires “reasonable assurance” that exposures are within the limits of the subsequent 

performance objectives for 10 CFR 61.41 through 10 CFR 61.44.101 

The DOE has developed an HTF PA which provides the technical basis and results demonstrating there 

is reasonable assurance that the 10 CFR 61.41 and 10 CFR 61.42 performance objectives will be met 

after HTF closure. These analyses were performed using a variety of modeling codes including the 

PORFLOW deterministic code and GoldSim probabilistic code. As required by the DOE Manual 435.1-1, 

maintenance of the HTF PA will include future revisions as needed (e.g., to incorporate new information 

and update model codes).102 

The HTF PA modeling consisted of a hybrid approach using both deterministic modeling (Base Case — 

also called Case A — and sensitivity analyses) as well as probabilistic modeling for certain sensitivity and 

uncertainty analyses. The HTF PA includes deterministic and probabilistic analyses for 100,000 years 

after HTF closure. This approach envelopes both the 1,000-year period after closure, as described in 

DOE Manual 435.1-1 for performance assessments for DOE low-level waste disposal facilities, as well as 

the 10,000-year period suggested in NRC’s NUREG-1854. DOE is providing all modeling results (i.e., up 

to 100,000 years) for purposes of consulting with NRC under Section 3116 (a) of the NDAA. 

The HTF PA details the analyses performed to provide “reasonable assurance” that the stabilized 

residuals, waste tanks and ancillary structures will be disposed of in compliance with the 10 CFR 61.41 

and 61.42 performance objectives in conjunction with closure of the HTF. Individual HTF system 

behaviors are evaluated within the HTF PA for various waste tank and ancillary structure configurations, 

including a Base Case, which provides results reflecting the closure system behavior. The HTF PA 

provides the development and calculation of the following doses: 

 potential radiological doses to a hypothetical member of the public and 

 potential radiological doses to a hypothetical inadvertent intruder. 

 

101 The general requirement at 10 CFR 61.40 provides that land disposal facilities must be “controlled after closure”. Consistent with 

this requirement, the HTF will be “controlled after closure” through a number of mechanisms, including the following: 

 DOE Order 435.1 and its associated Manual 435.1-1 requires the DOE to manage radioactive wastes and associated facilities 

in accordance with DOE requirements, in a manner that protects the public, workers and the environment, and that complies 

with applicable federal, state and local laws. Among other things, DOE Manual 435.1-1 requires the development, review, 

approval, and implementation of closure plans for radioactive waste management facilities, and specifies the requirements that 

such plans must meet. DOE Manual 435.1-1 also requires continued monitoring of DOE facilities after closure. [DOE O 435.1, 

DOE M 435.1-1] 

 The SRS FFA, a formal agreement between DOE, Region 4 of the EPA and SCDHEC provides for a comprehensive 

remediation of SRS, governs the corrective/remedial action process from site investigation through site remediation, and 

describes procedures for that process. The FFA, in conjunction with applicable South Carolina law and regulation, establish 

the framework for the operation, new construction, removal from service, and any appropriate RCRA/CERCLA response action 

related to the waste tank systems. The FFA provides timetables for the removal from service of waste tanks that do not meet 

the secondary containment standards of FFA Section IX.C, or that leak or have leaked, as well as provisions for new 

construction and prevention and mitigation of releases or potential releases from the waste tank systems. [WSRC-OS-94-42] 

 DOE will maintain active institutional control and ownership of HTF such that HTF facility maintenance and controls will be 

performed to prevent inadvertent intrusion and protect public health and the environment. For the purposes of the HTF PA, the 

period of active institutional control is assumed to be for 100 years to ensure a conservative analysis relative to potential public 

risk, although the SRS Land Use Plan requires federal ownership and control of the site well beyond 100 years after closure. 102 The NRC TER included several recommendations in Chapter 4 regarding potential model refinements or model support, which 

NRC recommended DOE consider in future performance assessment revisions. NRC suggested that the recommendations be 

considered in the “intermediate term” or as part of “long-term/maintenance”, and designated such recommendations as “low” or 

“medium-risk significance”. DOE will evaluate the recommendations as part of performance assessment maintenance under DOE 

Manual 435.1-1. 
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These calculations were performed to provide information regarding potential peak doses from the closed 

HTF. In addition, uncertainty and sensitivity analyses were used to ensure reasonably conservative 

information is available to develop risk-informed conclusions related to the closure of HTF.103 

The following general definitions are used in the HTF PA and will serve as the basis for future HTF PA 

revisions. 

HTF Boundary: The HTF boundary is the line of demarcation enclosing the HTF waste tanks (Figure 

7.0-1). 

Buffer Zone: The buffer zone is the radial area that encompasses the HTF 100 meters from its boundary 

(Figure 7.0-1). 

Institutional Control: Institutional control is a 100-year period in which DOE retains ownership and 

control of HTF such that HTF facility maintenance and controls will be performed to prevent inadvertent 

intrusion and protect public health and the environment.104

Base Case: The waste tank system 

configuration modeling case within 

the HTF PA that represents the 

most probable and defensible 

estimate of expected conditions for 

the HTF closure system based on 

currently available information.105

Uncertainty and Sensitivity: 

Uncertainty and sensitivity analyses 

are employed to consider the effects 

of uncertainties in the conceptual 

models and sensitivity of simulation 

results to the parameters in the 

mathematical models. The 

sensitivity analyses consider 

sensitivity of results to parameters 

both individually and collectively. 

The HTF PA includes uncertainty 

and sensitivity analyses. 

For additional information, the 

Savannah River Site DOE 435.1 

Composite Analysis presents the 

results of a site-wide radiological 

assessment of SRS.106 [SRNL-STI2009-00512] The Composite 

Analysis (CA) documents the projected cumulative impacts to future members of the public from the 

disposal of low-level radioactive waste, closure of radioactive liquid waste tanks, and all other sources of 

 

103 DOE predicated the HTF PA and this HTF 3116 Basis Document on reasonable assumptions and reasonably foreseeable natural 

processes, recognizing that, in DOE’s view, validation of assumptions over extreme periods of time is not consistent with a riskinformed approach under DOE requirements and NRC guidance. 

104 To ensure a conservative analysis relative to potential public risk, DOE is using an institutional control period of 100 years. As 

described in Section 7.4.4 of this HTF 3116 Basis Document, the SRS Land Use Plan requires federal ownership and control of the

site well beyond 100 years after closure. 

105 To demonstrate in this HTF 3116 Basis Document that there is “reasonable assurance” (as called for by 10 CFR 61.40) that the 

performance objectives at 10 CFR 61.41 and 61.42 are met, DOE utilizes the most probable and defensible HTF PA estimates of 

expected conditions (the HTF PA Base Case, also called Case A) based on available information and reasonable assumptions, and 

informed by the uncertainty and sensitivity analyses in the HTF PA. 

106 Information about this CA is included to further inform the reader. DOE has prepared the CA under DOE Manual 435.1-1, which 

accompanies DOE Order 435.1, pursuant to DOE’s responsibilities under the Atomic Energy Act of 1954, as amended. The CA is 

not a NDAA Section 3116 requirement, and is not relied upon in this HTF 3116 Basis Document to demonstrate compliance with the 

NDAA Section 3116 criteria. As such, the CA should be viewed as outside the scope of both this HTF 3116 Basis Document and 

NDAA Section 3116. 

Figure 7.0-1: HTF Boundary and Buffer Zone 

Legend: 

 Boundary Buffer Zone (100 meters from HTF) 

Red Line = Demarcation line from which the one-meter and 100-meter 

concentrations are calculated.
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residual radioactive material projected to be left at SRS that could interact with the disposal facilities and 

closure sites to affect the future radiological dose to a member of the public. The CA uses methodologies 

and assumptions similar to, or derived from, the performance assessments for E Area, SDF, FTF, HTF 

and additional radioactive source terms from other SRS production and deactivated facilities (e.g., F- and 

H-Canyons, reactors and decommissioned facilities). In summary, the CA identified that during the 

10,025-year period following the projected site end-state date (2025), the maximum expected member of 

the public dose would be approximately 3 mrem/yr compared to DOE’s 30 mrem/yr dose constraint to 

ensure compliance with the 100 mrem/yr dose limit. [SRNL-STI-2009-00512] DOE will periodically 

update and maintain the CA, in accordance with DOE policy and DOE Manual 435.1-1. 

The following subsections discuss the 10 CFR 61.41 (see Section 7.1), 10 CFR 61.42 (see Section 7.2), 

10 CFR 61.43 (see Section 7.3) and 10 CFR 61.44 (see Section 7.4) performance objectives. 

7.1 10 CFR 61.41 

10 CFR 61.41 states: 

Concentrations of radioactive material which may be released to the general environment in 

ground water, surface water, air, soil, plants, or animals must not result in an annual dose 

exceeding an equivalent of 25 millirems to the whole body, 75 millirems to the thyroid, and 25 

millirems to any other organ of any member of the public. Reasonable effort should be made to 

maintain releases of radioactivity in effluents to the general environment as low as is reasonably 

achievable. 

7.1.1 General Approach 

To demonstrate compliance with this performance objective, a 25 mrem/yr peak all-pathways TEDE is 

used, rather than individual organ doses. The NRC states in NUREG-1854 that use of the 25 mrem/yr 

all-pathways TEDE is used by the NRC in making the assessment for compliance with the whole body, 

thyroid and any other organ limits in 10 CFR 61.41 and is protective of human health and the 

environment.107 

In addition NUREG-1854 states: 

...incidental waste determinations may use total effective dose equivalent (TEDE) without specific 

consideration of individual organ doses. Intruder calculations should be based on 5 mSv [500 

mrem] TEDE limit, without specific consideration of individual organ doses, to ensure consistency 

between 10 CFR 61.41 and 10 CFR 61.43. Because of the tissue weighting factors and the 

magnitude of the TEDE limit, specific organ dose limits are not necessary for protection from 

deterministic effects. 

The hypothetical future member of the public is assumed to be located at the boundary of the DOE 

controlled area until the assumed active institutional control period ends (i.e., 100 years after closure), at 

which point the receptor is assumed to move to the point of maximum exposure at or outside of the HTF 

100-meter buffer zone. For the purposes of demonstrating reasonable assurance that the performance 

objective at 10 CFR 61.41 will be met, the peak all-pathways dose at or outside of the 100-meter buffer 

zone is used. 

The pathways for release to a member of the public considered in the HTF PA analyses are discussed 

below. The scenarios are not assumed to occur until after the assumed 100-year institutional control 

period ends. [SRR-CWDA-2010-00128] 

7.1.2 Public Release Pathways Dose Analysis 

The primary water sources for the member of the public release pathways are either a well drilled into the 

groundwater aquifers or a GSA stream. The bounding dose scenario and associated exposure pathways 

for the member of the public was determined to be an agricultural resident who uses water from a well for 

domestic purposes. The bounding public dose scenario and associated exposure pathways are 

 

107 At page 4-9 of the TER, NRC explains “A 0.25 mSv/yr (25 mrem/yr) total effective dose equivalent (TEDE) limit applies for the 

post closure period of a disposal facility (NUREG-1854).” [ML14094A496] 
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documented in the HTF PA. The following exposure pathways involving the use of contaminated108 well 

water were considered (Figure 7.1-1): 

 direct ingestion of well water, 

 ingestion of milk and meat from livestock (e.g., dairy and beef cattle) that drink well water, 

 ingestion of meat and eggs from poultry that drink well water, 

 ingestion of vegetables grown in garden soil irrigated with well water, 

 ingestion of milk and meat from livestock (e.g., dairy and beef cattle) that eat fodder from a 

pasture irrigated with well water, 

 ingestion of meat and eggs from poultry that eat fodder from a pasture irrigated with well water, 

and 

 ingestion and inhalation of well water while showering. 

The following exposure pathways involving the use of contaminated surface water (from the applicable 

stream) for recreational use are assumed to occur: 

 direct irradiation during recreational activities (e.g., swimming, fishing, boating) from stream 

water, 

 dermal contact with stream water during recreational activities (e.g., swimming, fishing), 

 incidental ingestion and inhalation of stream water during recreational activities, and 

 ingestion of fish from the stream water. 

Additional exposure pathways could involve releases of radionuclides into the air from the water taken 

from the well (i.e., volatile radionuclides such as C-14 and I-129). Exposures from the air pathway in the 

HTF PA are: 

 direct plume shine and 

 inhalation. 

Secondary and indirect pathways that contribute relatively minor doses to a receptor when compared to 

direct pathways (e.g., ingestion of milk and meat) include: 

 inhalation of well water used for irrigation, 

 inhalation of dust from the soil that was irrigated with well water, 

 ingestion of or dermal contact with soil that was irrigated with well water, and 

 direct radiation exposure from radionuclides deposited on the soil that was irrigated with well 

water. 

The point of assessment for the groundwater wells used in the member of the public scenario is located 

100 meters from the HTF, as shown in Figure 7.0-1. The peak concentrations used to determine the 

peak doses for the member of the public exposure pathways are calculated and documented in the HTF 

PA. The groundwater concentrations used are peak concentrations for each radionuclide at the given 

point of assessment, from any of the aquifers. 

The groundwater concentrations were calculated based on the HTF PA conceptual model. The 

conceptual model is used to simulate the performance of the HTF closure system following HTF closure 

and is comprised of both near-field and far-field models that represent the HTF closure system and the 

environmental media through which radionuclides may migrate. The conceptual model was used to 

simulate transport of the radiological contaminants through soil and groundwater to the 100-meter 

assessment point and nearby streams. 

As discussed in the HTF PA and Section 2 of this HTF 3116 Basis Document, eight waste tanks and 

certain ancillary structures, are either fully submerged or partially submerged in the water table. The fact 

that some waste tanks are fully or partially submerged can impact particular aspects of the HTF PA 

modeling such as waste release modeling and transport modeling. The impacts of the waste tanks or 

ancillary structures being fully or partially submerged are addressed in detail in the HTF PA (e.g., HTF PA 

Sections 4.2.1.1.2 and 4.4.4.1.2). 

 

108 Contaminated in this context refers to radioactive materials that have been projected to migrate from the closed HTF. 
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Figure 7.1-1: Scenario with Well Water as Primary Source 
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The conceptual model used numerous HTF-specific input parameters to represent the HTF closure 

system behavior over time. Many of the input parameters are based on site-specific data (e.g., soil and 

cementitious materials Kd values) used in transport modeling. In addition, site-specific information is used 

to model the behavior of individual barriers within the HTF conceptual model, such as the waste tank 

carbon steel primary tanks and secondary liners or annular pans, if applicable, and cementitious barriers. 

The models and model inputs used in the HTF conceptual model to calculate groundwater concentrations 

are described in detail in the HTF PA. 

The groundwater peak dose for the member of the public is calculated in the HTF PA using site-specific 

input parameters, and the bounding dose scenario exposure pathways and peak concentrations 

discussed previously. Numerous bioaccumulation factors (e.g., soil-to-plant transfer factors), human 

health exposure parameters (e.g., water ingestion rates, vegetable consumption data) and dose 

conversion factors are used in the computer modeling to calculate doses for each of the exposure 

pathways, and these parameters are documented in the HTF PA. An air-pathway analysis was also 

performed in addition to the groundwater analysis to determine the dose contribution from the air 

pathway. This analysis utilized an atmospheric screening methodology to identify radionuclides for airpathway modeling based on waste tank radionuclide projected inventories and the limited number of 

radionuclides susceptible to volatilization. Computer modeling was performed to calculate the transport 

of radionuclides through the stabilized waste form and the closure cap to the surface of HTF. An airpathway dose was then calculated based on the specific curies of each radionuclide assumed to be 

transported to the surface of HTF. The air pathway analysis and groundwater analysis are combined to 

determine an all-pathways peak dose for a member of the public. 

In addition to the deterministic all-pathways peak dose Base Case analysis, additional analyses are 

provided in the HTF PA to characterize the context of uncertainty and sensitivity surrounding the HTF PA 

all-pathways peak dose results. These evaluations focused on the key uncertainties and sensitivities 

identified during calculation of the member of the public dose. The uncertainty analyses provide 

information regarding how collective uncertainty in model input parameters is propagated through the 

model to the various model results. The sensitivity analyses provide information as to how various 

individual input parameters affect dose results. Together the uncertainty and sensitivity analyses provide 

assurance that the impacts of variability and uncertainty in the member of the public dose analyses are 

understood and addressed. 

The uncertainty and sensitivity analyses were primarily performed using a probabilistic model, with some 

additional single parameter sensitivity analyses (e.g., solubility value sensitivity analysis, liner failure 

sensitivity analysis, alternate configuration sensitivity analysis) performed through deterministic modeling. 

The probabilistic model allows for variability of multiple parameters simultaneously, including variability of 

the flow data, so concurrent effects of changes in the model can be analyzed. The deterministic model 

single parameter analyses provide a method to evaluate the importance of the uncertainty around a 

single parameter of concern. The deterministic model single parameter analyses included a 

comprehensive barrier analyses that identified barriers to waste migration and evaluated the capabilities 

of each barrier as understood from the results of the HTF PA. The barrier analyses assessed the 

contribution of individual barriers (e.g., closure cap, grout, contamination zone, waste tank liner and waste 

tank concrete) by comparing contaminant flux results under various barrier conditions. Using both 

probabilistic and deterministic models for sensitivity analysis versus a single approach provides additional 

information, to inform the decision making process, concerning which parameters are of most importance 

to the HTF PA model. [SRR-CWDA-2010-00128] DOE performed a review of potential Features, Events 

and Processes (FEPs), and analyzed and documented how the FEPs are considered in the HTF PA. As 

a result of this evaluation, DOE found that all applicable FEPs have been covered by existing analyses in 

the HTF PA.109 [SRR-CWDA-2012-00044, SRR-CWDA-2012-00011] 

 

109 On page 4-24 of the TER, NRC states, “DOE should improve its documentation of the screening and disposition of FEPs in the 

HTF PA to enhance transparency and traceability including the use of expert judgment during the screening and implementation of

included FEPs in the HTF PA. (Low-to-Medium Risk-Significance, Intermediate-Term)” DOE will evaluate this recommendation as 

part of performance assessment maintenance under DOE Manual 435.1-1. [ML14094A496] 
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7.1.3 Results of the Analyses 

The HTF PA modeling was used to determine an all-pathways dose to a member of the public for 

comparison with the 10 CFR 61.41 performance objectives. The deterministic Base Case analysis in the 

HTF PA, Revision 1, projected the peak all-pathways dose to the HTF public receptor (i.e., individual 

greater than or equal to 100 meters from the HTF) to be less than the 25 mrem/yr performance objective 

as shown in Figure 7.1-2.110 The peak all-pathways dose projection captured in Figure 7.1-2 includes the 

groundwater pathways and air pathways associated with all 29 HTF waste tanks and associated ancillary 

structures with the groundwater pathway being the most significant contributor. The peak doses to the 

member of the public depicted in Figure 7.1-2 are primarily from Tc-99. 

 

 

110 The Figure 7.1-2 peak all-pathways dose curve reflects the current results of the Base Case through deterministic (PORFLOW) 

modeling for HTF PA, Revision 1, and reflects dose projections following closure of HTF. The peak HTF all-pathways Base Case 

dose results provided in the HTF PA are not to be considered limits. The horizontal green lines in Figures 7.1-2 through 7.1-5 are 

presented for visual illustration only. As required by DOE Manual 435.1-1, maintenance of the HTF PA will include future updates to 

incorporate new information, update model codes, analysis of actual residual inventories, and other information, as appropriate. 

Figure 7.1-2: Peak All-Pathways Dose to a Member of the Public Within 10,000 Years 

(Base Case Deterministic Analysis) 

[SRR-CWDA-2010-00128] 
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For additional information, Figure 7.1-3 displays the peak all-pathways dose within 100,000 years after 

closure to demonstrate consideration of the uncertainty inherent in the timing of the peak dose. [SRRCWDA-2010-00128] Later dose peaks, beyond the initial 10,000 years after HTF closure, are associated 

with loss of containment due to failure of the primary tank walls.111 Loss of the primary tank wall initiates 

changes to the chemistry and radionuclide holding capability of the grout, which directly affects 

radionuclide release rates. Peak doses to the member of the public within 100,000 years are primarily 

from Ra-226.112 

As discussed previously, additional analyses are provided in the HTF PA to characterize the context of 

uncertainty and sensitivity surrounding the HTF PA peak dose results. The HTF PA probabilistic 

modeling performed for the uncertainty and sensitivity analyses was used to better understand the 

projected dose to the HTF public receptor for the Base Case, as well as other tank configurations, over a 

wide range of variability in input parameters. The HTF PA, Revision 1, uncertainty and sensitivity analysis 

projected various statistical results associated with the All Cases113 dose (i.e., plots of dose versus time 

 

111 The Base Case for HTF PA, Revision 1, assumes that the primary tank walls for Tanks 12, 14, 15 and 16 are completely failed at

the time of HTF closure. Therefore, Tanks 12, 14, 15 and 16 are modeled as having no primary tank wall and no annular pan in 

place at the time of closure. The primary tank walls for Type IV tanks fail at approximately year 3,600, Type I tanks, not including 

Tank 12, at approximately year 11,400, Tank 13 at approximately year 12,700 and Type III/IIIA tanks at approximately year 12,800. 

[SRR-CWDA-2010-00128] 

112 Dose contribution from Ra-226 is not a result of initial inventories of this radionuclide at the time of closure but is a result of 

ingrowth from decay of parent radionuclides, i.e., U-234 and Pu-238. 113 As part of the probabilistic analyses, a set of realizations was performed to collectively evaluate the effects of all postulated waste 

tank configurations. In the “All Cases” run, every waste tank model independently sampled the possible waste tank configurations 

during each realization, allowing the probabilistic analysis to consider tank configuration variability. 

Figure 7.1-3: Peak All-Pathways Dose to a Member of the Public Within 100,000 Years 

(Base Case Deterministic Analysis) 

[SRR-CWDA-2010-00128] 
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showing 5th percentile, 25th percentile, median, mean, 75th percentile and 95th percentile doses) to a 

member of the public within the initial 10,000 years, as shown in Figure 7.1-4, for additional 

information.114 Figure 7.1-5 displays the uncertainty analyses dose results within 100,000 years after 

HTF closure to reflect consideration of the uncertainty inherent in the timing of the dose results. The fact 

that the peak of the means dose from the HTF probabilistic analyses is higher than the deterministic peak 

dose is not unexpected, since many of the stochastic distributions used in the probabilistic modeling are 

reasonably conservative, driving the peak of the means higher than the deterministic peak dose. This is 

demonstrated in Figure 7.1-4 by the fact that the mean dose consistently exceeds the median dose. 

[SRR-CWDA-2010-00128] 

 

114 The Figure 7.1-4 probabilistic dose statistics reflect the current results of the uncertainty analyses projected through probablilistic 

modeling for HTF PA, Revision 1. The probabilistic dose results provided in the HTF PA are not to be considered limits. As 

required by DOE Manual 435.1-1, maintenance of the HTF PA will include future updates to incorporate new information, update 

model codes, analysis of actual residual inventories, and other information, as appropriate. 

Figure 7.1-4: Statistical Summary of Time History of Total Member of the Public Dose, at the Well 

of Maximum Concentration, Within 10,000 Years (All Cases Probabilistic Analysis) 

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Since there are over 40 unique and independent inventory sources modeled in the HTF model, there is 

significant temporal and spatial complexity inherent in the modeling system. The uncertainty and 

sensitivity analyses demonstrated that the impact of individual parameters and/or specific barriers can be 

variable, with the impact depending to a great extent upon the tank type and/or radionuclide involved.115 

[SRR-CWDA-2010-00128] Additional discussion regarding the radionuclides most impacting the dose 

results can be found in Section 5.1 of this HTF 3116 Basis Document. 

The HTF all-pathways Base Case dose results, calculated in the HTF PA remain below the 25 mrem/yr 

peak all-pathways dose limit, as shown in Figure 7.1-2. In addition, the uncertainty and sensitivity 

analyses included in the HTF PA provide sufficient information on parameter sensitivities and modeling 

uncertainties to provide reasonable assurance that the 25 mrem/yr all-pathways dose limit will be met.116 

 

115 Detailed discussion of the uncertainty and sensitivity analyses can be found in Section 5.6 of the HTF PA. The impact of the various integrated 

conceptual model segments on the dose results is detailed in Section 7.1 of the HTF PA. [SRR-CWDA-2010-00128] 116 The NRC TER included several recommendations in Chapter 4 regarding potential model refinements or model support, which NRC recommended 

DOE consider during the monitoring period or in future performance assessment revisions. NRC suggested that the recommendations be considered 

in the “intermediate term” or as part of “long-term/maintenance”, and designated such recommendations as “low” or “medium-risk significance”. These 

recommendations include Key Recommendation ES-3.4, “DOE should continue to evaluate the appropriateness of selected transport parameters (e.g., 

cementitious material and soil Kds) and the selection of sorption models during the monitoring period (Medium Risk Significance, Intermediate Term)” 

and Key Recommendation ES-3.5, “DOE should improve the calibration of the far-field model, particularly local to H-Area and continue to study 

uncertainty in calibration targets (Medium Risk Significance, Intermediate Term),” both of which are noted on page xx of the NRC TER. The NRC TER 

contains additional recommendations related to potential model refinements which the NRC considers to be of lower risk significance. DOE has 

considered these recommendations as well in preparing this HTF 3116 Basis Document. [SRR-CWDA-2014-00080] DOE will evaluate and 

incorporate, as appropriate, the recommendations as part of performance assessment maintenance under DOE Manual 435.1-1. This means that 

DOE will undertake, in a focused manner, the applicable evaluation of HTF PA model support activities or experimental work recommended in the TER 

─ while cleaning activities are ongoing at HTF ─ taking into account the availability of other information, alternative means to obtain the information, the 

potential reduction to risk and uncertainty, worker and public exposure, costs, the availability of appropriated funds to do the work, and like 

considerations. This approach to performance assessment issues is consistent with NRC guidance in NUREG-1573, A Performance Assessment 

Methodology for Low-Level Radioactive Waste Disposal Facilities, which recognizes that a variety of approaches are appropriate to provide enhanced 

justification and support for performance assessments.

Figure 7.1-5: Statistical Summary of Time History of Total Member of the Public Dose, at the Well 

of Maximum Concentration Within 100,000 Years (All Cases Probabilistic Analysis) 

[SRR-CWDA-2010-00128] 
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[SRR-CWDA-2010-00128] Additional discussion on HTF PA dose results is provided in Appendix C of 

this HTF 3116 Basis Document. 

7.1.4 As Low As Reasonably Achievable 

The NRC performance objective in 10 CFR 61.41 also provides that reasonable effort shall be made to 

maintain releases of radioactivity in effluents to the environment ALARA. The HTF PA was developed in 

accordance with the comparable requirement in DOE Manual 435.1-1: 

Performance assessments shall include a demonstration that projected releases of radionuclides 

to the environment shall be maintained as low as reasonably achievable (ALARA).

As discussed previously, the HTF PA provides the information to demonstrate compliance with the 

25 mrem all-pathways dose performance objective, including stabilization117 of the residual waste using 

grout to minimize releases to the environment. [SRR-CWDA-2010-00128] Section 5.2 of this HTF 3116 

Basis Document provides the information to show that HRRs in the waste tanks and ancillary structures 

will have been removed to the maximum extent practical at closure. 

In addition to removal of HRRs to the maximum extent practical, other HTF closure design features also 

serve to support the ALARA objective set forth in 10 CFR 61.41. The closure design of the HTF stabilizes 

the residual waste, minimizes infiltration of water through the waste tanks and ancillary structures, and 

provides long-term stability. These features of the HTF closure design serve to impede release of 

stabilized contaminants into the general environment. 

The residual material remaining in the waste tanks after HRRs have been removed to the maximum 

extent practical will be stabilized with reducing grout, a chemically reducing environment known to 

minimize the mobility of the contaminants after closure. The waste tank fill grout will also have low 

permeability, which enhances its ability to limit the migration of contaminants after closure. 

There are multiple elements of the HTF design that will serve to minimize infiltration of water through the 

waste tanks and ancillary structures. The waste tank concrete vaults and primary and secondary steel 

liners serve to significantly retard water flow through the waste tanks. In addition, the waste tank liners 

and annular space, if applicable, are filled with cementitious material, which will further serve to limit the 

amount of water infiltration into the waste tanks. The concrete structures, steel wall liners, if applicable, 

and transfer line encasements or outer jackets will serve to significantly retard water flow into ancillary 

structures. In addition, the waste tanks and ancillary structures are expected to be covered with a closure 

cap,118 which further limits the water infiltration. 

Final HTF closure will also support long-term stability consistent with the ALARA objective set forth in 10 

CFR 61.41. Because the waste tanks will be filled with grout at closure, significant structural failure (i.e., 

collapse) is not likely. Ancillary structures such as diversion boxes, pump pits, and pump tanks are 

expected to be filled with appropriate fill materials, as necessary, to prevent subsidence. Additionally, the 

engineered closure cap will also provide physical stabilization of the closed site. 

The design features described above serve to impede the release of stabilized contaminants into the 

general environment. These features, along with the removal of HRRs to the maximum extent practical, 

are consistent with the ALARA objective in 10 CFR Part 61.41 to maintain releases of radioactivity in 

effluents to the general environment ALARA. 

Sections 7.3.11 and 7.3.12 provide discussion relative to compliance with the ALARA objective set forth 

in 10 CFR Part 61.43. 

 

117 Stabilization of the HTF waste tanks will be carried out by filling the waste tanks with grout after completion of waste removal 

activities. Ancillary structures will be filled, as necessary, to prevent subsidence of the structure or final closure cap. The DOE 

currently does not plan to grout the HTF transfer lines. 

118 The closure cap design described in the HTF PA is based on the best information available at the time the HTF PA was 

developed. [SRR-CWDA-2010-00128] The design information utilized is for planning purposes sufficient to support evaluation of

the closure cap as part of the integrated site conceptual model evaluated in the HTF PA. Any actual closure cap design will be

finalized closer to the time of HTF closure in accordance to the FFA for SRS (e.g., Section IX.E.(2).) [WSRC-OS-94-42], to take 

advantage of possible advances in materials and closure cap technology that could be used to improve the design. The final 

closure cap design will minimize water infiltration into the waste tanks and ancillary structures, and the likelihood of intrusion into the 

waste. 
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7.1.5 Conclusion 

As demonstrated in the preceding discussion, reasonable assurance is provided that the performance 

objective at 10 CFR 61.41 will not be exceeded.119

7.2 10 CFR 61.42 

Provisions in 10 CFR 61.42 require: 

Design, operation, and closure of the land disposal facility must ensure protection of any 

individual inadvertently intruding into the disposal site and occupying the site or contacting the 

waste at any time after active institutional controls over the disposal site are removed. 

7.2.1 General Approach 

The requirement of 10 CFR 61.42 exhibits the NRC’s intent to protect persons who inadvertently intrude 

on the waste. While the performance objective does not establish quantitative limits on exposure, the 10 

CFR 61 Final EIS suggests a dose limit of 500 mrem/yr for the waste classification scheme in 10 CFR 

61.55. By way of guidance, the NRC uses 500 mrem/yr dose limit for evaluating impacts to an 

inadvertent intruder for purposes of 10 CFR 61.42.120 [NUREG-0945, NUREG-1854] For the purposes of 

demonstrating reasonable assurance that the performance objective at 10 CFR 61.42 will be met, the 500 

mrem/yr peak intruder dose is used. 

The 10 CFR 61.42 regulations do not specify use of a particular scenario to demonstrate compliance. In 

developing intruder scenarios, the DOE assumes that humans will continue land use activities, which are 

consistent with past (e.g., recent decades) and present regional practices, after the end of the assumed 

active institutional control period. 

To calculate the dose to an inadvertent intruder, potential intruder scenarios were considered in the HTF 

PA and the bounding Acute Intruder and Chronic Intruder dose scenarios were determined to be the 

Acute Intruder-Drilling Scenario and Chronic Intruder-Agricultural (Post-Drilling) Scenario respectively. 

7.2.2 Acute Intruder-Drilling Scenario 

The bounding Acute Intruder scenario analyzed in the HTF PA is an Acute Intruder-Drilling Scenario. 

This scenario assumes that after the end of active institutional controls a well is drilled within the HTF 

buffer zone. The well is assumed to be used for domestic water use and irrigation. Because no other 

natural resources have been identified in the HTF, no additional drilling scenarios are considered. In a 

drilling scenario, an Acute Intruder is assumed to be the person or persons who install the well and are 

exposed to drill cuttings during well installation. 

 

 

119 On page xxi of the TER, NRC states, “Because DOE is still in the early stage of HTF cleaning and significant technical 

uncertainties remain regarding DOEs ability to meet the performance objectives, it would be premature for the NRC to draw 

conclusions in that regard now. However, the NRC is confident that, if DOE implements the recommendations contained in this 

TER, the likelihood of meeting the 10 CFR Part 61 performance objectives is enhanced.” [ML14094A496] DOE has carefully 

considered the NRC’s recommendations as discussed within this section of, and throughout, this HTF 3116 Basis Document. NRC 

Key Recommendation ES-3.1, which the NRC notes as its “primary recommendation,” re-iterates the NRC’s FTF recommendation 

that DOE conduct waste release testing. DOE is currently addressing this recommendation through its performance assessment 

maintenance program implementation plan. [SRR-CWDA-2013-00133] As stated in the implementation plan, DOE is performing 

waste release testing and research with the overall objective, “[T]o provide additional information regarding the residual waste 

solubility assumptions used in the FTF and HTF PA waste release models by developing a series of analytical methods to be used 

to test the solubility of plutonium, neptunium, uranium, and technetium under various simulated waste tank conditions using actual 

waste tank residuals.” Although noted within the implementation plan that testing was “on-hold pending available funding,” 

subsequent to issuance of the plan, funding was available and activities continued in fiscal year 2014. NRC Key Recommendation

ES-3.2 on page xx of the TER states, “DOE should conduct a more comprehensive analysis of contaminant release from the 

annular regions of Type I and II tanks (Medium-to-High Risk Significance, Short and Intermediate Term).” As discussed in Section 

5.3 of this HTF 3116 Basis Document, the process outlined in Appendix B will be followed to support DOE Tier 2 Closure 

Authorization for all waste tanks and ancillary structures. As noted in footnote 88 in Section 5.3, as part of the Tank 16 Special 

Analysis which will be completed to support Tier 2 Closure Authorization for Tank 16, DOE will be evaluating contaminant release 

from the Type I and Type II annuli. DOE is performing these activities as part of performance assessment maintenance under DOE

Manual 435.1-1. [DOE M 435.1-1] 

120 For additional information, DOE Manual 435.1-1 also establishes a 100 mrem/yr chronic dose limit for evaluating impacts of an

inadvertent intruder. 
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The exposure pathways for this acute drilling scenario include (Figure 7.2-1): 

 inhalation of resuspended drill cuttings, 

 external exposure to the drill cuttings, and 

 inadvertent drill cuttings ingestion. 

7.2.3 Chronic Intruder-Agricultural (Post-Drilling) Scenario 

The bounding chronic intruder scenario analyzed in the HTF PA is Chronic Intruder-Agricultural (PostDrilling) Scenario. This scenario assumes that after the end of active institutional controls, a farmer lives 

within the HTF buffer zone and consumes food crops grown, and meat and milk from animals raised 

there, using water from a well drilled within the HTF buffer zone. The Chronic Intruder-Agricultural 

Scenario (i.e., post-drilling) is an extension of the Acute Intruder-Drilling Scenario. This scenario 

assumes that an intruder lives in a building near the well drilled as part of the intruder-drilling scenario 

and engages in agricultural activities within the HTF buffer zone. Excavation to the surface of the 

stabilized contaminants in the waste tanks was not considered credible because its depth is more than 40 

feet below the closure cap. Therefore, the intruder-agricultural scenario was retained for the ancillary 

structures inventory and specifically a waste transfer line. This is because it is less protected than a 

diversion box, valve box or pump pit, which are protected by thick shield covers, equaling several feet of 

concrete. The soil used for agricultural purposes is assumed to be contaminated by both drill cuttings 

and well water used for irrigation. 

The intruder is exposed to (Figure 7.2-2): 

 direct ingestion of well water, 

 ingestion of milk and meat from livestock (e.g., dairy and beef cattle) that drink well water, 

 ingestion of meat and eggs from poultry that drink well water, 

 ingestion of vegetables grown in garden soil irrigated with well water and containing contaminated 

drill cuttings, 

 ingestion of milk and meat from livestock (e.g., dairy and beef cattle) that eat fodder from a 

pasture irrigated with well water, 

 ingestion of meat and eggs from poultry that eat fodder from a pasture irrigated with well water, 

 ingestion and inhalation of well water while showering, 

 direct irradiation during recreation activities (e.g., swimming, fishing, boating) from stream water, 

 dermal contact with stream water during recreational activities (e.g., swimming, fishing), 

 incidental ingestion and inhalation of stream water during recreational activities, 

 ingestion of fish from the stream water, 

 direct plume shine, 

 inhalation, 

 inhalation of well water used for irrigation, 

 inhalation of dust from the soil that was contaminated by drill cuttings and irrigated with well 

water, 

 ingestion of soil that was contaminated by drill cuttings and irrigated with well water, and 

 direct radiation exposure from radionuclides deposited on the soil that was contaminated by drill 

cuttings and irrigated with well water. 

The intruder may also be exposed to a release of volatile radionuclides (e.g., C-14 and I-129) from the 

drill cuttings and contaminated well water. These pathways include direct plume shine and inhalation. 

7.2.4 Intruder Release Pathways Dose Analysis 

As discussed previously, the bounding Acute Intruder and Chronic Intruder dose scenarios are the Acute 

Intruder-Drilling Scenario and Chronic Intruder-Agricultural (Post-Drilling) Scenario, respectively. These 

bounding intruder dose scenarios and associated exposure pathways are documented in the HTF PA. 

The water source for the intruder release pathways is a well drilled into the groundwater aquifers. The 

contaminated drill cuttings in the intruder release pathways are from drilling into a waste transfer line. 
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Figure 7.2-1: Acute Intruder-Drilling Scenario 
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Figure 7.2-2: Chronic Intruder-Agricultural (Post-Drilling) Scenario 
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The point of assessment for the groundwater wells used in the intruder scenario is located one meter 

from the HTF boundary (Figure 7.0-1). The peak concentrations used to determine the peak doses for 

the intruder release exposure pathways are calculated and documented in the HTF PA. The groundwater 

concentrations used are peak concentrations for each radionuclide at the given point of assessment, from 

any of the aquifers. 

The groundwater concentrations were calculated based on the HTF PA conceptual model. The 

conceptual model is used to simulate the performance of the closed HTF for the purpose of informing 

closure actions associated with HTF waste tanks and ancillary structures. The conceptual model is 

comprised of models that represent the HTF closure system and the environmental media through which 

radionuclides may migrate. The conceptual model was used to simulate transport of the radiological 

contaminants through soil and groundwater. 

The conceptual model used numerous HTF-specific input parameters to represent the HTF closure 

system behavior over time. Many of the input parameters are based on site-specific data (e.g., soil and 

cementitious materials Kd values) used in transport modeling. In addition, site-specific information is used 

to model the behavior of individual barriers within the HTF conceptual model, such as the waste tank 

carbon steel primary tanks and secondary tanks or annular pans (as applicable), and cementitious 

barriers. The models and model inputs used in the HTF conceptual model to calculate groundwater 

concentrations and the waste transfer line drill cutting inventory are described in detail in the HTF PA. 

The peak intruder dose is calculated in the HTF PA using site-specific input parameters and the bounding 

dose scenario exposure pathways and peak concentrations discussed previously. Numerous 

bioaccumulation factors (e.g., soil-to-plant transfer factors), human health exposure parameters (e.g., 

water ingestion rates, vegetable consumption data) and dose conversion factors are used in the computer 

modeling to calculate doses for each of the exposure pathways, and these parameters are documented in 

the HTF PA. 

In addition to the intruder peak dose analyses, additional analyses are provided in the HTF PA to 

characterize the context of uncertainty and sensitivity surrounding the HTF PA intruder peak dose results. 

These evaluations focused on the key uncertainties and sensitivities identified during calculation of the 

intruder dose. The uncertainty analyses provide information regarding how collective uncertainty in 

model input parameters is propagated through the model to the various model results. The sensitivity 

analyses provide information as to how various individual input parameters affect dose results. Together 

the uncertainty and sensitivity analyses provide assurance that the impacts of variability and uncertainty 

in the intruder dose analyses are understood and addressed. 

The uncertainty and sensitivity analyses were primarily performed using a probabilistic model, with some 

additional single parameter sensitivity analyses (e.g., alternate configuration sensitivity analysis) 

performed through deterministic modeling. The probabilistic model allows for variability of multiple 

parameters simultaneously, including variability in flow data, so concurrent effects of changes in the 

model can be analyzed. The deterministic model single parameter analyses provide a method to 

evaluate the importance of the uncertainty around a single parameter of concern. The deterministic 

model single parameter analyses included comprehensive barrier analyses that identified barriers to 

waste migration and evaluated the capabilities of each barrier as understood from the results of the HTF 

PA. The barrier analyses assessed the contribution of individual barriers (e.g., closure cap, grout, 

contamination zone, waste tank liner and waste tank concrete) by comparing contaminant flux results 

under various barrier conditions. Using both probabilistic and deterministic models for sensitivity analyses 

versus a single approach provides additional information to inform the decision making process 

concerning which parameters are of most importance to the HTF PA model. [SRR-CWDA-2010-00128] 

DOE performed a review of potential FEPs and analyzed how the FEPs are considered in the HTF PA. 

As a result of this evaluation, it was determined that all applicable FEPs have been covered by existing 

analyses in the HTF PA. [SRR-CWDA-2012-00044, SRR-CWDA-2012-00011] 

7.2.5 Results of the Analysis 

The HTF PA deterministic model was used to determine an inadvertent intruder dose for comparison with 

the 10 CFR 61.42 performance objective. The HTF PA projected the peak inadvertent intruder (i.e., 

individual within the HTF boundary) acute and chronic doses to be less than the 500 mrem/yr 
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performance measure as shown in Figures 7.2-3 and 7.2-4, respectively.121 Figure 7.2-4 displays the 

peak chronic dose for 20,000 years in consideration of the uncertainty inherent in the timing of the peak 

dose. The 500 mrem/yr inadvertent intruder dose considers releases associated with the closure of all 29 

waste tanks and related ancillary structures within HTF. 

 

 

121 The acute and chronic dose results shown in Figures 7.2-3 and 7.2-4 reflect results from the Base Case through deterministic 

(PORFLOW) modeling for HTF PA and reflect dose projections out to 10,000 and 20,000 years, respectively, following closure of 

HTF. The peak HTF inadvertent intruder dose results provided in the HTF PA are not to be considered as limits. The green text

box in Figure 7.2-3 and horizontal green line in Figure 7.2-4 are presented for visual illustration only. As required by DOE Manual 

435.1 1, maintenance of the HTF PA will include future updates to incorporate new information, update model codes, analysis of 

actual residual inventories, and other information, as appropriate. 

Figure 7.2-3: Peak Acute Dose to the Inadvertent Intruder 

(Base Case Deterministic Analysis) 

[SRR-CWDA-2010-00128]
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The HTF PA modeling performed for the uncertainty and sensitivity analyses was used to determine the 

projected dose to an inadvertent intruder for the Base Case, as well as other tank configurations, over a 

wide range of variability in input parameters. The HTF PA uncertainty analysis projected a 760 mrem/yr 

peak of the mean chronic dose to an inadvertent intruder within a 10,000-year period following HTF 

closure for 1,000 Base Case realizations. The peak of the mean dose is greater than 500 mrem/yr due to 

the tendency for many of the stochastic distributions used in the uncertainty analysis to be conservatively 

biased high, thereby skewing the uncertainty analysis results to the high dose side of their distributions.122 

Since there are over 40 unique and independent inventory sources modeled in the HTF model, there is 

significant temporal and spatial complexity inherent in the modeling system. The uncertainty and 

sensitivity analyses demonstrated that the impact of individual parameters and/or specific barriers can be 

variable, with the impact depending to a great extent upon the tank type and/or radionuclide involved.123 

Additional discussion regarding the radionuclides most impacting the dose results can be found in Section 

5.1 of this HTF 3116 Basis Document. 

 

122 The 760 mrem/yr dose reflects the current results of the uncertainty analyses results projected through probablilistic modeling for 

HTF PA. The peak of the means dose provided in the HTF PA is not to be considered a limit. As required by DOE Manual 435.1-1,

maintenance of the HTF PA will include future updates to incorporate new information, update model codes, analysis of actual 

residual inventories, and other information, as appropriate. 

123 Detailed discussion of the uncertainty and sensitivity analyses can be found in Section 5.6 of the HTF PA. The impact of the 

various integrated conceptual model segments on the dose results is detailed in Section 7.1 of the HTF PA. [SRR-CWDA-2010-

00128] 

Figure 7.2-4: Peak Chronic Dose to the Inadvertent Intruder 

(Base Case Deterministic Analysis) 

[SRR-CWDA-2010-00128]
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Demonstration of compliance with the performance objective concerning an inadvertent intruder is 

provided by the fact that peak HTF Base Case inadvertent intruder dose calculated in the HTF PA is less 

than 500 mrem/yr. In addition, the uncertainty and sensitivity analyses included in the HTF PA provide 

sufficient information on parameter sensitivities and modeling uncertainties. [SRR-CWDA-2010-00128] 

Additional discussion on HTF PA dose results is provided in Appendix C of this HTF 3116 Basis 

Document. 

7.2.6 Conclusion 

The preceding discussion demonstrates that there is reasonable assurance that the 10 CFR 61.42 

performance objective will not be exceeded after HTF closure.124 

7.3 10 CFR 61.43 

Provisions in 10 CFR 61.43 states: 

Operations at the land disposal facility must be conducted in compliance with the standards for 

radiation protection set out in part 20 of this chapter, except for releases of radioactivity in 

effluents from the land disposal facility, which shall be governed by §61.41 of this part. Every 

reasonable effort shall be made to maintain radiation exposures as low as is reasonably 

achievable. 

This requirement references 10 CFR 20, which contains radiological protection standards for workers and 

the public. The DOE requirements for occupational radiological protection are provided in 10 CFR 835 

and those for radiological protection of the public and the environment are provided in DOE Order 458.1. 

Consistent with NDAA Section 3116(a), the cross-referenced “standards for radiation protection” in 10 

CFR 20 that are considered in detail in this HTF 3116 Basis Document are the dose limits for the public 

and the workers during disposal operations set forth in 10 CFR 20.1101(d), 10 CFR 20.1201(a)(1)(i), 10 

CFR 20.1201(a)(1)(ii), 10 CFR 20.1201(a)(2)(i), 10 CFR 20.1201(a)(2)(ii), 10 CFR 20.1201(e), 10 CFR 

20.1208(a), 10 CFR 20.1301(a)(1), 10 CFR 20.1301(a)(2) and 10 CFR 20.1301(b).125 [NDAA_3116] 

Consistent with NUREG-1854, the following sections explain that these dose limits correspond to the 

dose limits in 10 CFR 835 and relevant DOE Orders which establish DOE regulatory and contractual 

 

124 On page xix of the TER, NRC states: “DOE develops reasonable exposure scenarios to evaluate potential exposures of the inadvertent 

intruder to radiation from the residual waste and demonstrate that performance objectives in 10 CFR 61.42 could be met. However, the NRC 

staff notes that the probabilistic analyses for inadvertent intrusion indicate that projected doses could exceed 500 mrem/yr within 10,000 

years.” [ML14094A496] As noted in Appendix C of this HTF 3116 Basis Document, the probabilistic (i.e., GoldSim) model intruder results 

were conservative (e.g., about three times higher than the deterministic PORFLOW results) due to differences in the models. The 

inadvertent intruder peak of the mean dose of 760 mrem/yr is higher than the median peak dose of 500 mrem/yr, illustrating that the mean is 

greater than the median and indicating that a few realizations sampled at the tail of the parameter distributions can cause the mean to be 

high. DOE’s conclusion that there is reasonable assurance that the inadvertent intruder performance objective is met is based on doses 

associated with the Base Case, taking into account other cases and additional sensitivity analyses. DOE considers the probabilistic analyses 

to be a conservative analysis which further supports DOE’s conclusion of reasonable assurance the performance objective is not exceeded. 

The NRC also states on page xix of the TER, “[B]ecause the groundwater pathway is a key contributor to demonstrating compliance with 

both 10 CFR 61.41 and 10 CFR 61.42, a demonstration of compliance with 10 CFR 61.42 is dependent upon resolution of technical issues 

associated with the 10 CFR 61.41 analysis.“ [ML14094A496] With regards to the 10 CFR 61.42 performance objective, DOE has taken into 

consideration addressing the NRC concerns relative to the 10 CFR 61.41 analysis as described in Section 7.1. 

125 The introductory “notwithstanding” phrase to NDAA Section 3116 makes it clear that the provisions of NDAA Section 3116(a) are to apply 

in lieu of other laws that “define classes of radioactive waste.” As is evident from the plain language of this introductory “notwithstanding” 

phrase, NDAA Section 3116(a) pertains to classification and disposal, and radiation protection standards for disposal, of certain waste at 

certain DOE sites. Thus, the factors for consideration set forth in NDAA Section 3116(a)(1) through NDAA Section 3116(a)(3) are those 

which pertain to classification and disposal of waste, and the radiation protection standards for disposal. The Joint Explanatory Statement of 

the Committee of Conference in Conference Report 108-767, accompanying H.R. 4200 (the NDAA), also confirms that NDAA Section 

3116(a) concerns classification, disposal, and radiation protection standards associated with disposal, and does not concern general 

environmental laws or laws regulating radioactive waste for purposes other than disposal. Moreover, in the plain language of NDAA Section 

3116, Congress directed that the Secretary of Energy consult with the NRC but did not mandate that DOE obtain a license or any other 

authorization from NRC, and did not grant NRC any general regulatory, administrative, or enforcement authority for disposal of the DOE 

wastes covered by NDAA Section 3116. As such, the “standards for radiation protection” in 10 CFR Part 20 (as cross-referenced in the 

performance objective at 10 CFR 61.43), which are relevant in the context of NDAA Section 3116, are the dose limits for radiation protection 

of the public and the workers during disposal operations, and not those which address general licensing, administrative, programmatic, or 

enforcement matters administered by NRC for NRC licensees. Accordingly, this HTF 3116 Basis Document addresses in detail the radiation 

dose limits for the public and the workers during disposal operations that are contained in the provisions of 10 CFR Part 20 referenced 

above. Although 10 CFR 20.1206(e) contains limits for planned special exposures for adult workers, there will not be any such planned 

special exposures for closure operations at HTF. Therefore, this limit is not discussed further in this HTF 3116 Basis Document. Likewise, 

10 CFR 20.1207 specifies occupational dose limits for minors. However, there will not be minors working at HTF who will receive an 

occupational dose. Therefore, this limit is not discussed further in this HTF 3116 Basis Document. 
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requirements for DOE facilities and activities.126 The following subsections show the HTF closure meets 

these dose limits and that doses will be maintained ALARA.127 Table 7.3-1 provides a crosswalk between 

the standards set forth in 10 CFR 20 and the applicable DOE requirements. 

Table 7.3-1: Crosswalk Between Applicable 10 CFR 20 Standards and DOE Requirements 

10 CFR 20 Standard DOE Requirement 

Basis 

Document 

Section Title 

10 CFR 20.1101(d) DOE Order 458.1, Chg. 3 7.3.1 Air Emissions Limit for Individual Member of the 

Public 

10 CFR 20.1201(a)(1)(i) 10 CFR 835.202 (a)(1) 7.3.2 Total Effective Dose Equivalent Limit for Adult 

Workers 

10 CFR 20.1201(a)(1)(ii) 10 CFR 835.202 (a)(2) 7.3.3 Any Individual Organ or Tissue Dose Limit for 

Adult Workers 

10 CFR 20.1201(a)(2)(i) 10 CFR 835.202 (a)(3) 7.3.4 Annual Dose Limit to the Lens of the Eye for 

Adult Workers 

10 CFR 20.1201(a)(2)(ii) 10 CFR 835.202 (a)(4) 7.3.5 Annual Dose Limit to the Skin of the Whole Body 

and to the Skin of the Extremities for Adult 

Workers 

10 CFR 20.1201(e) DOE Order 440.1B 7.3.6 Limit on Soluble Uranium Intake 

10 CFR 20.1208(a) 10 CFR 835.206 (a) 7.3.7 Dose Equivalent to an Embryo/Fetus 

10 CFR 20.1301(a)(1) DOE Order 458.1 7.3.8 Total Effective Dose Equivalent Limit for 

Individual Members of the Public 

10 CFR 20.1301(a)(2) 10 CFR 835.602 

10 CFR 835.603 

7.3.9 Dose Limits for Individual Members of the Public 

in Unrestricted Areas 

10 CFR 20.1301(b) 10 CFR 835.208 7.3.10 Dose Limits for Individual Members of the Public 

in Controlled Areas 

7.3.1 Air Emissions Limit for Individual Member of the Public (10 CFR 20.1101(d)) 

The NRC regulation at 10 CFR 20.1101(d) provides in relevant part: 

[A] constraint on air emissions of radioactive material to the environment, excluding Radon-222 

and its daughters, shall be established ... such that the individual member of the public likely to 

receive the highest dose will not be expected to receive a total effective dose equivalent in 

excess of 10 mrem( 0.1 mSv), per year from these emissions. 

The DOE similarly limits effective dose equivalent from air emissions to the public at 10 mrem/yr in DOE 

Order 458.1, Chg. 3 to comply with the EPA requirement in 40 CFR 61.92, which has the same limit.128 

The estimated dose per year from airborne emissions to the maximally exposed individual member of the 

public located at or beyond the SRS boundary from all operations at SRS ranged from 0.04 mrem to 0.11 

mrem from 1997 through 2010. [WSRC-TR-97-00322, WSRC-TR-98-00312, WSRC-TR-99-00299, 

WSRC-TR-2000-00328, WSRC-TR-2001-00474, WSRC-TR-2003-00026, WSRC-TR-2004-00015, 

WSRC-TR-2005-00005, WSRC-TR-2006-00007, WSRC-TR-2007-00008, WSRC-STI-2008-00057, 

SRNS-STI-2009-00190, SRNS-STI-2010-00175, SRNS-STI-2011-00059] These values (0.04 mrem to 

0.11 mrem from 1997 to 2010) for the SRS operations, not only HTF closure operations, are well below 

the dose limit specified in 10 CFR 20.1101(d) of 10 mrem, 0.1 mSv per year. 

 

126 On page xix of the TER, NRC concluded, “Because DOE regulations provide an equivalent level of protection as is provided by 10 

CFR Part 20, DOE has procedures and processes in place to demonstrate compliance with protection of individuals during operations (10 

CFR 61.43).” [ML14094A496]

127 In addition, 10 CFR Part 835, like Part 20 for NRC licensees, includes requirements that do not set dose limits, such as 

requirements for radiation protection programs, monitoring, entrance controls for radiation areas, posting, records, reporting or 

training. 

128 40 CFR 61.92 provides in relevant part as follows: “Emissions of radionuclides to the ambient air from Department of Energy 

facilities shall not exceed those amounts that would cause any member of the public to receive in any year an effective dose 

equivalent of 10 mrem/yr.” 
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7.3.2 Total Effective Dose Equivalent Limit for Adult Workers (10 CFR 20.1201(a)(1)(i)) 

The NRC regulation at 10 CFR 20.1201(a), concerning occupational dose limits for adults, provides in 

relevant part: 

(a) [C]ontrol the occupational dose to individual adults, except for planned special exposures...to 

the following dose limits. 

(1) An annual limit, which is the more limiting of – 

(i) The total effective dose equivalent being equal to 5 rems (0.05 Sv). 

The DOE regulation in 10 CFR 835.202 (a)(1) has the same annual dose limit for the annual occupational 

dose to general employees.129 For the occupational dose to adults during HTF closure, the total effective 

dose (TED) per year will be controlled using the ALARA principles, and will be below 5 rem as described 

in 5Q Manual, Chapter 2, Radiological Standards. Occupational doses to workers have been well below 

the annual limits specified in 10 CFR 20.1201(a)(1)(i) for all SRS work activities. Since 1995, the highest 

annual dose received by an SRS worker is 1,808 mrem in 2003. [SRR-CWDA-2012-00097] The highest 

total dose received by an HTF worker from 1995 - 2011 was 687 mrem in 2005. [PIT-MISC-0062, SRRCWDA-2012-00097] Considering the dose histories for HTF and the doses received during waste tank 

closure activities in FTF the TED to workers from HTF closure is expected to remain well below the 

DOE/NRC limit. Because of the similarities between HTF and FTF waste removal processes, doses 

received during FTF waste removal activities provide insight into the magnitude of the doses that can be 

anticipated during HTF waste removal activities. F-Tank Farm Tank 17 and FTF Tank 20 were 

operationally closed in 1997. Given that the highest FTF worker dose was 215 mrem in 1997, reasonable 

assurance is provided that doses received by a worker during closure activities will be below 5 rem. [PITMISC-0062, SRR-CWDA-2012-00097] 

7.3.3 Any Individual Organ or Tissue Dose Limit for Adult Workers (10 CFR 20.1201(a)(1)(ii)) 

The NRC regulation at 10 CFR 20.1201(a), concerning occupational dose limits for adults, provides in 

relevant part:

(a) [C]ontrol the occupational dose to individual adults, except for planned special exposures...to 

the following dose limits. 

(1) An annual limit, which is the more limiting of – 

(ii) The sum of the deep-dose equivalent and the committed dose equivalent to any 

individual organ or tissue other than the lens of the eye being equal to 50 rems (0.5 

Sv). 

The dose limit specified in 10 CFR 20.1201(a)(1)(ii) is similar130 to the dose limit specified in 10 CFR 

835.202 (a)(2). For the occupational dose to adults during HTF closure, the sum of the deep-dose 

equivalent and the committed dose equivalent to any individual organ or tissue other than the lens of the 

eye will be controlled to ALARA, below a maximum of 50 rem/yr. The SRS Engineering Standard 

Number 01064, Radiological Design Requirements, provides the design basis annual occupational 

exposure limits for any organ or tissue, other than the eye, cannot exceed 10 rem/yr, which is well below 

the NRC limit of 50 rem/yr. [5Q Manual, Chapter 2, WSRC-TM-95-1] 

7.3.4 Annual Dose Limit to the Lens of the Eye for Adult Workers (10 CFR 20.1201(a)(2)(i)) 

The NRC regulation at 10 CFR 20.1201(a), concerning occupational dose limits for adults, provides in 

relevant part: 

(a) [C]ontrol the occupational dose to individual adults, except for planned special exposures...to 

the following dose limits. 

 

129 The DOE regulation requires that the occupational dose per year for general employees shall not exceed both a TED of 5 rems 

which is the sum of the equivalent dose to the whole body for external exposures and the committed effective dose, which includes 

the weighted internal exposures to any other organ or tissue other than the skin or the lens of the eye. 130 10 CFR 835.202(a)(2) also excludes exposure to skin as well as exposure to the lens of the eye and the dose term is Committed 

Equivalent Dose. 
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(2) The annual limits to the lens of the eye, to the skin of the whole body, and to the skin of 

the extremities, which are: 

(i) A lens dose equivalent of 15 rems (0.15 Sv). 

The dose limit specified in 10 CFR 20.1201(a)(2)(i) is the same as that specified in the DOE regulation at 

10 CFR 835.202 (a)(3). For the occupational dose to adults during HTF closure, the annual dose limit to 

the lens of the eye will be controlled using the ALARA principles, and will be below 15 rem/yr. The SRS 

Engineering Standard Number 01064 provides the design basis annual occupational exposure limits for 

the lens of the eye cannot exceed 3 rem/yr, which is well below the NRC limit of 15 rem/yr. [5Q Manual, 

Chapter 2, WSRC-TM-95-1] 

7.3.5 Annual Dose Limit to the Skin of the Whole Body and to the Skin of the Extremities for 

Adult Workers (10 CFR 20.1201(a)(2)(ii)) 

The NRC regulation at 10 CFR 20.1201(a), concerning occupational dose limits for adults, provides in 

relevant part: 

(a) [C]ontrol the occupational dose to individual adults, except for planned special exposures...to 

the following dose limits. 

(2) The annual limits to the lens of the eye, the skin of the whole body, or to the skin of the 

extremities, which are: 

(ii) A shallow-dose equivalent of 50 rem (0.5 Sv) to the skin of the whole body or to the 

skin of any extremity. 

This NRC dose limit specified in 10 CFR 20.1201(a)(2)(ii) is the same as the DOE dose limit specified at 

10 CFR 835.202 (a)(4). For the occupational dose to adults during HTF closure, which involve limited 

hands-on activity, the annual dose limit to the skin of the whole body or to the skin of any extremity will be 

controlled using the ALARA principles, and will be below a shallow-dose equivalent of 50 rem/yr. [5Q 

Manual, Chapter 2] 

7.3.6 Limit on Soluble Uranium Intake (10 CFR 20.1201(e)) 

The NRC regulation at 10 CFR 20.1201(e), concerning occupational dose limits for adults, provides in 

relevant part: 

(e) In addition to the annual dose limits,...limit the soluble uranium intake by an individual to 10 

milligrams in a week in consideration of chemical toxicity[.] 

In addition to the adult annual dose limits during HTF closure, the soluble uranium intake by an individual 

is controlled to less than 10 milligrams per week. DOE Order 440.1B specifies that soluble uranium 

intake requirements are the more restrictive concentrations in the American Conference of Governmental 

Industrial Hygienists Threshold Limit Values (0.2 milligrams per cubic meter, same as noted in 10 CFR 20 

Appendix B footnote 3) or the Occupational Safety and Health Administration (OSHA) Permissible 

Exposure Limit (PEL) (0.05 milligrams per cubic meter). The soluble uranium OSHA PEL limit, which 

equates to a soluble uranium intake of 2.4 milligrams per week, is the more restrictive of the two. The 

soluble uranium intake, if any, during HTF closure will be controlled to 2.4 milligrams per week, which is 

below the NRC limit in 10 CFR 20.1201(e). [4Q1.1, Procedure 101A] 

7.3.7 Dose Equivalent to an Embryo/Fetus (10 CFR 20.1208(a)) 

The NRC regulation at 10 CFR 20.1208(a), concerning the dose equivalent to an embryo/fetus, provides 

in relevant part: 

(a) [E]nsure that the dose equivalent to the embryo/fetus during the entire pregnancy, due to the 

occupational exposure of a declared pregnant woman, does not exceed 0.5 rem (5 mSv). 

The DOE regulation at 10 CFR 835.206 (a) has the same dose limit. For the embryo/fetus occupational 

dose during HTF closure, doses will be controlled so the dose equivalent to the embryo/fetus during the 

entire pregnancy for a declared pregnant worker will not exceed 0.5 rem. Furthermore, after pregnancy 

declaration, DOE provides a mutually agreeable assignment option of work tasks, without loss of pay or 

promotional opportunity, such that further occupational radiation exposure during the remainder of the 
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gestation period is unlikely. In addition, personnel dosimetry is provided and used to carefully track 

exposure as controlled by the 5Q Manual, Chapter 2. 

7.3.8 Total Effective Dose Equivalent Limit for Individual Members of the Public (10 CFR 

20.1301(a)(1)) 

The NRC regulation at 10 CFR 20.1301(a), concerning dose limits for individual members of the public, 

provides in relevant part: 

(a) [C]onduct operations so that – 

(1) The total effective dose equivalent to individual members of the public ...does not exceed 

0.1 rem (1 mSv) in a year, exclusive of the dose contributions from background radiation,

from any medical administration the individual has received, from exposure to individuals

administered radioactive material and released..., from voluntary participation in medical 

research programs, and from the ...disposal of radioactive material into sanitary 

sewerage[.] 

Provisions in DOE Order 458.1 similarly limit public doses to less than 100 mrem/yr. However, the DOE 

application of the limit is more restrictive, in that it requires DOE to make a reasonable effort to ensure 

multiple sources (e.g., DOE sources and NRC regulated sources) do not combine to cause the limit to be 

exceeded. For individual members of the public during HTF closure, the TED limit to an individual 

member of the public will be controlled to less than 0.1 rem/yr. [5Q Manual, Chapter 2] The air pathway 

is the predominant pathway for doses to the public from SRS operations. The air pathway doses to 

members of the public have been, and are expected to continue to be, well below the 0.1 rem annual limit 

specified in 10 CFR 20.1301(a). [WSRC-TR-97-00322, WSRC-TR-98-00312, WSRC-TR-99-00299, 

WSRC-TR-2000-00328, WSRC-TR-2001-00474, WSRC-TR-2003-00026, WSRC-TR-2004-00015, 

WSRC-TR-2005-00005, WSRC-TR-2006-00007, WSRC-TR-2007-00008, WSRC-STI-2008-00057, 

SRNS-STI-2009-00190, SRNS-STI-2010-00175, SRNS-STI-2011-00059] 

7.3.9 Dose Limits for Individual Members of the Public in Unrestricted Areas (10 CFR 

20.1301(a)(2)) 

The NRC regulation at 10 CFR 20.1301(a), concerning dose limits for individual members of the public, 

provides in relevant part: 

(a) [C]onduct operations so that – 

(2) The dose in any unrestricted area from external sources, exclusive of the dose 

contributions from patients administered radioactive material and released ..., does not 

exceed 0.002 rem (0.02 millisievert) in any one hour. 

The DOE regulation at 10 CFR 835.602 establishes the expectation that TED in controlled areas will be 

less than 0.1 rem in a year. For individual members of the public during HTF closure, operations will be 

conducted such that the dose in any unrestricted area from external sources, exclusive of the dose 

contributions from patients administered radioactive material, will be less than 0.00005 rem per hour 

above background. The 5Q Manual, Chapter 2, also restricts the TED in controlled areas to less than 0.1 

rem in a year. To ensure these dose limits are met, the following measures have been instituted within 

controlled areas. Per 10 CFR 835.603, radioactive materials areas have been established for radioactive 

material accumulation possibly resulting in a radiation dose of 100 mrem in a year or greater. In addition, 

SRS has established Radiological Buffer Areas (RBAs) around posted radiological areas. Standard SRS 

practice is to assume a 2,000 hour per year continuous occupancy at the outer boundary of these areas; 

therefore, the dose rate at a RBA boundary is 0.00005 rem/hour (100 mrem/2,000 hours = 0.05 

mrem/hour or 0.00005 rem/hour). Since the controlled area encompasses a RBA, it is ensured the dose 

in the controlled area (but outside of radioactive material areas and RBA) will be less than 0.1 rem in a 

year. [5Q Manual, Chapter 2] Therefore, SRS implementation of the provisions at 10 CFR 835.602 and 

10 CFR 835.603 provides limits protective of the dose limit specified in 10 CFR 20.1301(a)(2). Training is 

required for individual members of the public for unescorted entry into controlled areas. In addition, to 

ensure no member of the public exceeds radiation exposure limits, use of dosimetry is required if a 
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member of the public is expected to enter a controlled area and receive a dose that may exceed 0.05 

rem/yr.131 [5Q Manual, Chapter 5] 

7.3.10 Dose Limits for Individual Members of the Public in Controlled Areas (10 CFR 20.1301(b)) 

The NRC regulation at 10 CFR 20.1301(b), concerning dose limits for individual members of the public, 

provides in relevant part: 

(b) If ... members of the public [are permitted] to have access to controlled areas, the limits for 

members of the public continue to apply to those individuals. 

The DOE regulation at 10 CFR 835.208 has the same dose limit. The TED limit to an individual member 

of the public granted access to controlled areas during HTF closure will be controlled to 0.1 rem/yr. 

Furthermore, training is required for individual members of the public for entry into controlled areas. In 

addition, to ensure no member of the public exceeds radiation exposure limits, use of dosimetry is 

required if a member of the public is expected to enter a controlled area and receive a dose that may 

exceed 0.05 rem/yr.132 [5Q Manual, Chapter 5] 

7.3.11 As Low As Reasonably Achievable (10 CFR 20.1003) 

The NRC regulation at 10 CFR 20.1003 defines ALARA in relevant part: 

ALARA ... means making every reasonable effort to maintain exposures to radiation as far below 

the dose limits ... as is practical consistent with the purpose for which the ... activity is 

undertaken...[.] 

The DOE has a similar requirement, and the DOE regulation at 10 CFR 835.2 defines ALARA as “... the 

approach to radiation protection to manage and control exposures (both individual and collective) to the 

work force and to the general public to as low as reasonable....” For radiological work activities during 

HTF closure, every reasonable effort will be made to maintain exposures to radiation as far below the 

dose limits as is practical consistent with the purpose for which the activity is undertaken. Furthermore, 

the DOE regulation at 10 CFR 835.101(c) requires the contents of each Radiation Protection Program 

(RPP) to include formal plans and measure for applying the ALARA process to occupational exposure as 

further discussed in Section 7.3.12.1 of this HTF 3116 Basis Document. 

7.3.12 Reasonable Assurance 

Measures that provide reasonable assurance that HTF closure will comply with the applicable dose limits 

and with the ALARA provisions include the documented RPP, the Documented Safety Analysis (DSA), 

design, regulatory and contractual enforcement mechanisms and access controls, training and dosimetry. 

These measures are discussed in the following subsections. 

7.3.12.1 SRS Radiation Protection Program 

The DOE regulates occupational radiation exposure at its facilities through 10 CFR 835, which 

establishes exposure limits and other requirements to ensure DOE facilities are operated in a manner 

such that occupational exposure to workers is maintained within acceptable limits and as far below these 

limits as is reasonably achievable. The requirements in 10 CFR 835, if violated, provide a basis for the 

assessment of civil penalties under the Atomic Energy Act of 1954, Section 234A, as amended. [42 USC 

2282a] 

Pursuant to 10 CFR 835, activities at SRS, including HTF closure operations, must be conducted in 

compliance with the documented RPP for SRS as approved by DOE. The key RPP elements include 

monitoring of individuals and work areas, access control to areas containing radiation and radioactive 

materials, use of warning signs and labels, methods to control the spread of radioactive contamination, 

radiation safety training qualification, objectives for the design of facilities, criteria for radiation and 

 

131 10 CFR 20.1003 defines restricted areas as “an area, access to which is limited ... for the purpose of protecting individuals 

against undue risks from exposure to radiation and radioactive materials.” A very similar definition is located in 10 CFR 835.2 for a 

controlled area. 

132 10 CFR 20.1301(d) allows licensees to request NRC authorization to allow an individual member of the public to operate up to an 

annual dose limit of 0.5 rem (5 mSv). 10 CFR 835 is more restrictive for the dose to an individual member of the public with a limit 

of 0.1 rem maximum annual dose as discussed in Section 7.3.8. 
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radioactive material workplace levels, and continually updated records to document compliance with the 

provisions of 10 CFR 835. The RPP also includes formal plans and measures for applying the ALARA 

process. 

The 10 CFR 835 requirements, as contained in the RPP, are incorporated in the Standards/Requirement 

Identification Document (S/RID) system. The S/RID system links the requirements of 10 CFR 835 to the 

site-level and lower-level implementing policies and procedures that control radiological work activities 

conducted across the site. These procedures control the planning of radiological work, the use of 

radiation monitoring devices by employees, the bioassay program, the air monitoring program, the 

contamination control program, the ALARA program, the training of general employees, radiological 

workers, radiological control inspectors and health physics professionals and technicians and the other 

aspects of an occupational RPP as required by 10 CFR 835. 

7.3.12.2 Documented Safety Analysis 

The HTF operates under a DSA in accordance with 10 CFR Part 830. As the first step in the 

development of the DSA, a formal Hazard Analysis (HA) was performed to systematically present the 

results of potential process-related hazards, Natural Phenomena Hazards and external hazards that can 

affect the public, workers and environment through the occurrence of single or multiple failures. [DOESTD-3009-94] The HA was performed by subject matter experts including operations, engineering, 

industrial hygiene, radiological protection, environmental compliance and maintenance professionals. 

The HA consisted of three phases: 

1. hazard identification 

2. hazard classification 

3. hazard evaluation [DOE-STD-3009-94] 

The hazard identification phase identifies possible radiological and chemical hazardous materials 

associated with normal and abnormal operations as well as potential energy sources to disperse 

hazardous materials into the environment. 

The hazard classification phase evaluates for the maximum possible quantities of hazardous materials, 

which are then evaluated against DOE criterion to determine the overall hazard classification. [DOESTD-1027-92] 

The hazard evaluation phase identifies possible normal and abnormal operational events that could 

expose the public and workers to hazardous material and, therefore, are evaluated to establish the 

magnitude of the risk. Additionally, the consequence and frequency of each operational event must be 

determined and risk level identified. The purpose of identifying the risk level is to determine which 

operational events pose risk (and thus require additional evaluation) and those events which present 

negligible risk to the public and workers. 

As waste is removed from the waste tanks during the closure process, the DSA requires controls on the 

waste tanks commensurate with the risk of the material remaining in the waste tank. These controls 

include engineering controls (e.g., physical isolation requirements on transfer lines and motive forces) 

and administrative controls (e.g., limits on waste transfers and equipment operation). 

The DSA identifies hazards in the HA that could impact the public, facility workers and the environment 

during normal operations and accident conditions. The DSA also discusses summary descriptions of key 

SRS safety management programs. 

In part, these administrative controls require: a facility manager be assigned who is accountable for safe 

operation and in command of activities necessary to maintain safe operation, personnel who carry out 

radiological controls functions have sufficient organizational freedom to ensure independence from 

operating pressure, that personnel receive initial and continuing training including radiological control 

training and an RPP shall be prepared consistent with 10 CFR 835. In addition, the design requirements 

implement 10 CFR 835 and, in particular, implement ALARA principles. 
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7.3.12.3 Radiological Design for Protection of Occupational Workers and the Public 

The HTF radiological facilities and facility modifications are designed to meet the requirements of 10 CFR 

835 Subpart K. The SRS Engineering Standard Number 01064 provides the requirements necessary to 

ensure compliance with 10 CFR 835. [WSRC-TM-95-1] The standard refers to 10 CFR 835, DOE 

Orders, DOE Standards, DOE handbooks, national consensus standards, SRS manuals, SRS 

engineering standards, SRS engineering guides and site operating experience in order to meet the 10 

CFR 835 specific requirements and additional requirements to ensure the design provides for protection 

of the worker and the environment. 

The standard covers the full spectrum of radiological design requirements and not just radiation exposure 

limits. The following are the specific areas addressed in the standard: radiation exposure limits; facility 

and equipment layout; area radiation levels; radiation shielding; internal radiation exposure; radiological 

monitoring; confinement; and ventilation. 

The facility design also incorporates radiation zoning criteria to ensure exposure limits are met by 

providing adequate radiation shielding. Areas in which non-radiological workers are present are assumed 

to have continuous occupancy (2,000 hours/year) and are designed to a dose rate less than 0.05 mrem 

per hour to ensure the annual dose is less than 100 mrem. [WSRC-TM-95-1] Other zoning criteria are 

established to ensure radiological worker doses are ALARA and less than 1,000 mrem/yr to meet the 10 

CFR 835.1002 design requirements. 

The design is also required to provide necessary radiological monitoring or sampling for airborne and 

surface contamination to ensure the engineered controls are performing their function and, in the event of 

a failure or upset condition, workers are warned and exposures avoided. 

Radiological protection personnel ensure applicable requirements of the standard are addressed and 

presented in design summary documentation as needed. The incorporation of radiological design criteria 

in the engineering standard ensures the requirements of 10 CFR 835 are met and the design provides for 

the radiological safety of the workers and environment. 

7.3.12.4 Regulatory and Contractual Enforcement 

Any violation of the 10 CFR 835 requirements is subject to civil penalties pursuant to the Atomic Energy 

Act of 1954, Section 234A, as amended, 42 USC 2011 et seq., as implemented by DOE regulations in 10 

CFR Part 820. In addition, the requirements in 10 CFR 835 and all applicable DOE Orders are 

incorporated into all contracts with DOE contractors. The DOE enforces these contractual requirements 

through contract enforcement measures, including the reduction of contract fees. [48 CFR 970] 

7.3.12.5 Access Controls, Training, Dosimetry and Monitoring 

Training or an escort is required for individual members of the public for entry into controlled areas. In 

addition, use of dosimetry is required if a member of the public is expected to enter a controlled area and 

exceed 0.05 rem/yr to ensure no member of the public exceeds radiation exposure limits. [5Q Manual, 

Chapter 5, 5Q Manual, Chapter 6] 

In addition, worker radiation exposure monitoring is performed for all workers expected to receive 100 

mrem/yr from internal and external sources of radiation to provide assurance no worker exceeds radiation 

exposure limits and all radiation doses are maintained as far below the limits as is reasonably achievable. 

[5Q Manual, Chapter 5] 

7.3.12.6 Occupational Radiation Exposure History for Savannah River Site 

The effectiveness of the RPPs, including the effectiveness of oversight programs to ensure they are 

implemented properly, is demonstrated by the occupational radiation exposure results. The highest 

annual dose received by an SRS worker from 1995-2010 was 1,808 mrem TED and the highest total 

dose received by an HTF worker from 1995 – 2011 was 687 mrem compared to the DOE Administrative 

Control Limit of 2,000 mrem/yr and the 10 CFR 835 limit of 5,000 mrem/yr. [PIT-MISC-0062, SRRCWDA-2012-00097] 

In addition, for all work activities, the average TED exposure for workers receiving a TED dose at SRS 

has been 127 mrem/yr over the last five years, 2006-2010. [DOE_HSS_ORE_2010] 
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7.4 10 CFR 61.44 

10 CFR 61.44 states: 

The disposal facility must be sited, designed, used, operated, and closed to achieve long-term 

stability of the disposal site and to eliminate to the extent practicable the need for ongoing active 

maintenance of the disposal site following closure so that only surveillance, monitoring, or minor 

custodial care are required. 

This section outlines the relevant factors of HTF siting, design, use, operation and closure, which ensure 

compliance with 10 CFR 61.44 for the purpose of the HTF 3116 Basis Document. 133

7.4.1 Siting 

A site characteristics review of demography, geography, meteorology, climatology, ecology, geology, 

seismology, and hydrogeology is presented in Section 3.0 of the HTF PA [SRR-CWDA-2010-00128] and 

Section 2.0 of this HTF 3116 Basis Document, and is briefly summarized below. The SRS is located in 

south-central South Carolina, approximately 100 miles from the Atlantic Coast. The major physical 

feature at SRS is the Savannah River, approximately 20 miles along the southwestern boundary of the 

site. The HTF is an active waste storage facility located within the GSA of SRS, approximately 5 miles 

from the boundary of SRS. The nearest towns are New Ellenton, South Carolina (5 miles), Jackson, 

South Carolina (5 miles), Snelling, South Carolina (5 miles) and the more populated areas include Aiken, 

South Carolina (12 miles) and Augusta, Georgia (15 miles). As of 2010, the region of influence 

population was over 570,000. 

The general climate for the SRS region is a humid subtropical climate characterized by short mild winters 

and long warm and humid summers. The site experiences an average of 49 inches of rainfall each year 

and the average monthly temperature has a low of 46 degrees in the winter to 81 degrees in the summer. 

SRS supports an abundant terrestrial and semi-aquatic wildlife. The areas around HTF have grasses, 

forests, and swamps. An abundance of terrestrial, avian, wetlands, and aquatic wildlife live within SRS. 

HTF itself is a heavy industrial complex surrounded by fencing and covered in asphalt and, therefore, few 

animals are seen near the tanks; however burrowing animals in the surrounding areas are common. At 

closure, a cover is expected to be designed incorporating the latest barrier technologies to limit burrowing 

animals and growth of plants with taproot system; however, those are expected to infiltrate, in time, after 

a loss of institutional controls. 

The principal geology of the region is characterized by unconsolidated soils. The vadose zone is 

comprised of cross-bedded, poorly sorted sands with clay lenses indicating fluvial deposition with 

occasional transitional marine influence. It is represented by wide differences in lithology and presents a 

very complex system of transmissive and confining beds. 

The GSA is bounded by two surface waters: UTR and Fourmile Branch. These waters eventually feed 

into the Savannah River. The aquifers of primary interest for HTF are the UTRA and Gordon Aquifer. 

Other aquifers do not contribute to the potential dose to the workers, public or intruder and were not 

included in the models. 

 

133 On page xx of its TER, under Key Recommendation ES-3.6, NRC states, “DOE should perform a closure cap settlement and 

stability analysis during the monitoring period (Medium Risk Significance, Intermediate Term).” On pages 4-164 of the TER the 

NRC states, “DOE has provided sufficient information for the NRC staff to perform a preliminary review of site-stability.” The NRC 

goes on to state, “The NRC staff thinks that additional information is needed with respect to the site-stability analysis as noted 

below. Additional analyses can be conducted during the monitoring period.” The NRC then notes the following areas that NRC 

recommends DOE continue to evaluate during the monitoring period: [ML14094A496] 

• Site-specific settlement analysis for HTF that includes the increased overburden from tank grout and the closure cap 

• Evaluation of vault and grout integrity that is consistent with observations and reasonable expectations of future degradation of 

cementitious materials 

• Assessment of the potential subsidence due to ongoing dissolution of calcareous sediment in the Santee formation 

The NRC TER contains additional recommendations related to closure cap design which the NRC considers to be of lower risk 

significance. DOE has considered these recommendations as well in preparing this HTF 3116 Basis Document. [SRR-CWDA2014-00080] DOE will evaluate, and incorporate, as appropriate, the recommendations as part of performance assessment 

maintenance under DOE Manual 435.1-1, and will consider the recommendations in the design and development of the final closure 

cap. 
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Because SRS is not located within a region of active plate tectonics characterized by volcanism, 

volcanology is not an issue of concern for SRS. The seismic history of the southeastern U.S. is 

dominated by the Charleston earthquake of 1886 with an estimated magnitude of 7.0. The two most 

recent seismic events within a 50-mile radius of SRS were in February 2014 with magnitudes of 4.1 and 

3.0. [SRR-LWDL-2014-00008, SRR-LWDL-2014-00009] In the past approximately 30 years there have 

been four earthquakes with epicenter locations within SRS boundaries, all with a magnitude less than 3.0. 

[SRR-CWDA-2010-00128] 

7.4.2 Design 

The closure design of the HTF waste tanks and ancillary structures provide long-term stability, which is 

consistent with the performance objective. 

There are multiple elements of the HTF design that will serve to minimize infiltration of water through the 

waste tanks and ancillary structures. The waste tank concrete vaults, steel tanks and secondary tanks or 

annular pans, where applicable, serve to significantly retard water flow through the waste tanks. The 

concrete structures, steel wall liners, if applicable, and transfer line encasements or outer jackets will 

serve to significantly retard water flow into ancillary structures. The HTF design features are described in 

detail in the HTF PA. In addition, the waste tanks and ancillary structures are expected to be covered 

with a closure cap,134 which further limits the water infiltration into the waste tanks and ancillary 

structures. 

Because the waste tanks will be filled with grout at closure, significant structural failure (i.e., collapse) is 

not likely. For tank types with an annulus, the annulus will also be grouted for stability and to minimize 

void spaces. The impact of potential waste tank degradation (e.g., cracking or corrosion leading to 

increased water infiltration) is considered in the HTF PA analysis. Ancillary structures such as diversion 

boxes, pump pits and pump tanks will be filled, as necessary, to prevent subsidence. 

Multiple waste tank design elements will serve as inadvertent intruder barriers. The HTF closure cap, 

concrete tops on Type I, II, III and IIIA tanks, grout-filled domed roof of the Type IV waste tanks135 and 

waste tank reducing grout are considered sufficient barriers to prevent drilling into the waste tanks, given 

regional well drilling practices and the presence of nearby land without underground rock or concrete 

obstructions. [SRR-CWDA-2010-00128] 

7.4.3 Use/Operation 

The use/operation of HTF waste tanks and ancillary structures will support long-term stability consistent 

with the performance objective. During operations, corrosion control and structural integrity programs are 

implemented to maintain design features utilized for waste containment (e.g., waste tanks and ancillary 

structures). These programs ensure that tanks are monitored for structural integrity via mechanisms such 

as a tank inspection program and a tank leak detection system. Programs such as these will be 

maintained throughout HTF use and operation. The HTF waste tanks and ancillary structures monitoring 

continues after closure via the site Groundwater Protection Program. [SRNS-TR-2009-00076] 

7.4.4 Closure 

Final HTF closure will support long-term stability consistent with this performance objective. In this 

context, long-term stability of the closed HTF site means that the stabilized residuals in the waste tanks 

and ancillary structures, the waste tanks, and the ancillary structures (including integral equipment) 

maintains structural integrity under the closure conditions for hundreds to thousands of years following 

closure. A stable closure system prevents subsidence of, and minimizes water intrusion into, the closed 

site and mitigates migration of residual material into the environment. In addition, a carefully designed 

 

134 The closure cap design described in the HTF PA is based on the best information available at the time the HTF PA was developed. 

[SRR-CWDA-2010-00128] The design information utilized is for planning purposes sufficient to support evaluation of the closure cap as part 

of the integrated site conceptual model evaluated in the HTF PA. Any actual closure cap design will be finalized closer to the time of HTF 

closure in accordance to the FFA for SRS (e.g., Section IX.E.(2).) [WSRC-OS-94-42], to take advantage of possible advances in materials 

and closure cap technology that could be used to improve the design. The final closure cap design will minimize water infiltration into the 

waste tanks and ancillary structures, and the likelihood of intrusion into the waste.

135 Grout used to fill the domed roof of the Type IV waste tanks will have a minimum 2,000 psi nominal compressive strength at 28 

days. 
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closure site minimizes the likelihood of inadvertent intrusion into the system and disturbance of the 

stabilized residuals. 

The waste tank systems (i.e., primary tank, secondary tank or annular pan, and concrete roofs and vaults, 

as applicable) and the ancillary structures themselves will provide the primary stability for the closed HTF 

site. Grouting of the waste tanks and backfilling of ancillary structures with an appropriate fill material, as 

necessary, will prevent subsidence and will minimize the migration of radioactive material into the 

environment over time, and will support long-term stability of both the tank structures and the residual 

waste.136 Grout used to fill the domed roof of the Type IV waste tanks will have a minimum 2,000 psi 

nominal compressive strength at 28 days to deter intrusion. Type I, II, III and IIIA tanks have sufficient 

thicknesses of reinforced concrete roofs to deter such intrusion. Grouting of the waste tanks, filling of the 

ancillary structures, and grout chemical and mechanical characteristics are discussed in detail in the HTF 

PA. [SRR-CWDA-2010-00128]

A closure cap is expected to be designed and constructed over the the HTF site following grouting of the 

HTF waste tanks and backfilling of the ancillary structures, as necessary.137 The closure cap design 

described in the HTF PA is based on the best information available at the time the HTF PA was 

developed. [SRR-CWDA-2010-00128] The design information utilized is for planning purposes sufficient 

to support evaluation of the closure cap as part of the integrated site conceptual model evaluated in the 

HTF PA. The actual closure cap design will be finalized closer to the time of HTF closure, to take 

advantage of possible advances in materials and closure cap technology that could be used to improve 

the design. The final closure cap design will minimize water infiltration into the waste tanks and ancillary 

structures, and the likelihood of intrusion into the waste. [SRR-CWDA-2010-00128] 

The DOE will maintain GSA ownership, which includes the HTF. The SRS Land Use Plan requires 

federal ownership and control of the site well beyond 100 years after tank closure. [PIT-MISC-0041] 

7.4.5 Conclusion 

As previously discussed, the site conditions do not present hazards that impact HTF stability. In addition, 

the HTF closure methods will result in a facility closure that does not require ongoing maintenance. 

Therefore, closure of the HTF complies with the 10 CFR 61.44 performance objective. 

 

136 It may be shown through calculation and other means that backfilling certain ancillary structures will not be required for the

purpose of long-term stability. For example, there are no current plans to grout or fill the HTF transfer lines. 137 Final closure of HTF will be performed per the requirements of the FFA for SRS (e.g., Section IX.E.(2). [WSRC-OS-94-42] 
Basis for Section 3116 Determination DOE/SRS-WD-2014-001 

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8.0 STATE-APPROVED CLOSURE PLAN 

Section Purpose 

The purpose of this section is to demonstrate removal from service and stabilizing of the 

HTF waste tanks and ancillary structures, as appropriate, will be performed pursuant to a 

State-approved closure plan. 

Section Contents 

This section discusses SCDHEC regulation of the waste tanks and ancillary structures and 

shows that removal from service of the HTF waste tanks and ancillary structures will be 

pursuant to State-approved Closure Modules, consistent with the HTF GCP. 

Key Points 

 The HTF waste storage and removal are governed, in part, by a SCDHEC industrial 

wastewater construction permit. 

 The overall plan for removing from service and stabilizing the HTF waste tanks and 

ancillary structures, referred to as the HTF GCP, requires approval by SCDHEC. 

 A specific Closure Module for each waste tank and ancillary structure, consistent with 

the requirements of the HTF GCP, will be developed and submitted to the SCDHEC for 

approval. The State must grant this approval before final stabilization activities may 

proceed. 

The NDAA Section 3116(a) provides in pertinent part: 

[T]he term “high-level radioactive waste” does not include radioactive waste resulting from the 

reprocessing of spent nuclear fuel that the Secretary of Energy..., in consultation with the Nuclear 

Regulatory Commission..., determines – 

(3)(A)(ii) [Will be disposed of] pursuant to a State-approved closure plan or State-issued 

permit, authority for the approval or issuance of which is conferred on the State outside of 

this section. 

8.1 State-Approved Closure Plan 

The HTF waste storage and removal operations are governed by an SCDHEC industrial wastewater 

construction permit issued January 25, 1993. [DHEC_01-25-1993] The permit was issued under the 

authority of the South Carolina Pollution Control Act, S.C. Code Ann., Section 48-1-10, et seq. (1985) and 

all applicable regulations implementing the Act. SCDHEC has authority for approval of wastewater 

treatment facility operational closure under Chapter 61, Articles 67 and 82 of the SCDHEC Regulations. 

[SCDHEC R.61-67, SCDHEC R.61-82] 

The HTF GCP addresses the State’s regulatory authority relevant to removing the HTF waste tanks and 

ancillary structures from service. The GCP sets forth the general protocol by which DOE intends to 

remove from service the HTF waste tanks and ancillary structures to protect human health and the 

environment. The HTF GCP was approved by SCDHEC August 2, 2012. Prior to approval by SCDHEC, 

the HTF GCP was made available to the public for review and comment. [SRR-CWDA-2011-00022, 

DHEC_08-02-2012] 

Before final stabilization activities commence,138 individual waste tank and ancillary structure closure 

plans, referred to as Closure Modules, describing closure details will be developed and submitted to 

SCDHEC for approval.139 Prior to approval, the Closure Modules will be made available to the public for 

review and comment as deemed appropriate by SCDHEC. The Closure Modules will describe the waste 

tank(s) or ancillary structure(s) being covered, waste removal activities performed and effectiveness, 

 

138 Final stabilization activities in this context refers to the addition of grout to the waste tanks, annulus, and cooling coils, or, in the 

case of ancillary structures, grout or other appropriate fill material, as necessary, for the purpose of stabilizing the structure. 139 Each individual waste tank and ancillary structure is required to be covered by a Closure Module approved by SCDHEC. Closure 

Modules may be written for individual waste tanks and ancillary structures or for groupings of waste tanks and ancillary structures. 
Basis for Section 3116 Determination DOE/SRS-WD-2014-001 

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Page 8-2 

justification that additional waste removal is not technically practicable from an engineering 

perspective,140 and characteristics of remaining residuals and the stabilization process. The Closure 

Modules will provide analysis for each waste tank or ancillary structure demonstrating conformance with 

the performance objectives set forth in the GCP. 

8.2 Conclusion 

As explained above, the HTF waste tanks and ancillary structures will be removed from service 

(operationally closed) and stabilized pursuant to State-approved Closure Modules, consistent with the 

HTF GCP. Thus, the HTF waste tanks, ancillary structures and the stabilized residuals “[will be disposed 

of] pursuant to a State-approved closure plan or State-issued permit, authority for the approval or 

issuance of which is conferred on the State outside of this section.”141

 

140 Per the FFA, the waste tanks will be cleaned until DOE-SR, SCDHEC and EPA agree that waste removal may cease. The 

Closure Module provides the basis for agency agreement. 141 On page 4-2 of the TER, NRC acknowledges DOE’s process for meeting this criterion, “An SCDHEC industrial wastewater 

construction permit, issued January 25, 1993, governs tank waste storage and removal operations for HTF closure (Sadler, 1993).

Stabilization of the HTF waste tanks and ancillary structures will be carried out pursuant to the HTF General Closure Plan (SRRCWDA-2011-00022, Rev. 0) approved by the State of South Carolina for all HTF tanks and ancillary structures. DOE will also 

develop a specific Closure Module for each tank or ancillary structure, or groupings of tanks and ancillary structures, and submit it to 

the State of South Carolina for approval. Final tank stabilization activities will not proceed until the state approves the closure 

documentation.” [ML14094A496] 
Basis for Section 3116 Determination DOE/SRS-WD-2014-001 

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9.0 CONCLUSION 

Based on consultation with the NRC and as demonstrated in the preceding sections of this HTF 3116 

Basis Document, the stabilized residuals within the waste tanks and ancillary structures, the waste tanks, 

and the ancillary structures (including integral equipment) located at HTF at the time of closure meet the 

criteria set forth in NDAA Section 3116(a). 

This document demonstrates that the HTF waste tanks, ancillary structures and their associated 

stabilized residuals will have had HRRs removed to the MEP at the time of closure. Removal of HRRs to 

the MEP in HTF waste tanks and ancillary structures occurs through a systematic progression of waste 

removal and cleaning activities using proven technologies to a point where further removal of HRRs is not 

sensible or useful in light of the overall benefit to human health, safety and the environment. 

The stabilized HTF wastes at closure are anticipated to meet concentration limits for Class C low-level 

waste as set out in 10 CFR 61.55. Nevertheless, DOE has consulted with the NRC on DOE’s disposal 

plans for HTF pursuant to the consultation process in NDAA Section 3116(a)(3)(B)(iii) to take full 

advantage of the consultation process established by NDAA Section 3116. 

This document demonstrates the stabilized residuals within the waste tanks and ancillary structures, the 

waste tanks, and the ancillary structures (including integral equipment) located at HTF at the time of 

closure will meet the 10 CFR 61, Subpart C performance objectives so as to provide for the protection of 

the public health and the environment. These performance objectives address protection of the general 

population from radioactivity releases, protection of individuals from inadvertent intrusion on the disposal 

site, protection of workers and the public during disposal facility operations and the stability of the 

disposal site after closure. 

Through use of the performance assessment process, DOE has analyzed the possible methods by which 

a future member of the public or inadvertent intruder could be exposed to the HTF residuals. The results 

of the HTF PA show that there is reasonable assurance the annual peak doses for a future hypothetical 

member of the public and a hypothetical inadvertent intruder will remain below 25 mrem and 500 mrem, 

respectively, in compliance with the performance objectives at 10 CFR 61.41 and 10 CFR 61.42. 

The DOE has programs in place to ensure protection of workers and the public during facility operations. 

As demonstrated in this document, the DOE requirements for occupational radiological protection and 

those for radiological protection of the public and the environment are equivalent to the requirements 

contained in the performance objectives at 10 CFR 61.43. 

This document demonstrates that the HTF at closure meets the performance objective at 10 CFR 61.44, 

concerning long-term site stability. DOE reviewed the site characteristics, including demography, 

geography, meteorology, climatology, ecology, geology, seismology and hydrogeology. As demonstrated 

in this HTF 3116 Basis Document, the site conditions do not present hazards that impact HTF stability. In 

addition, the HTF closure methods will result in a facility closure that does not require ongoing 

maintenance. 

The HTF waste tanks and ancillary structures will be removed from service (operationally closed) and 

stabilized pursuant to State-approved Closure Modules, consistent with the HTF GCP that has been 

approved by SCDHEC. Per the SRS FFA, the waste tanks will be cleaned until DOE-SR, SCDHEC and 

EPA agree that waste removal may cease. 

Furthermore, the stabilized residuals within the waste tanks and ancillary structures, the waste tanks, and 

the ancillary structures (including integral equipment) do not raise any unique considerations that, 

notwithstanding the demonstration that all other NDAA Section 3116(a) criteria have been met, require 

permanent isolation in a deep geologic repository. 

As summarized above and as discussed more fully in the preceding sections, this HTF 3116 Basis 

Document demonstrates that the HTF waste tanks, ancillary structures and residuals at closure meet the 

criteria in NDAA Section 3116(a). Moreover, DOE has engaged in extensive consultation with the NRC, 

as discussed previously. Accordingly, this document provides the basis for the Secretary of Energy, in 

consultation with the NRC, to determine that the NDAA Section 3116(a) criteria are met and, thus, the 

HTF waste is not high-level waste. 
Basis for Section 3116 Determination DOE/SRS-WD-2014-001 

for Closure of H-Tank Farm Revision 0 

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Page 10-1 

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Note 1: References identified as (Copyright) were used in the development of this document, but are 

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for Closure of H-Tank Farm Revision 0 

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Basis for Section 3116 Determination DOE/SRS-WD-2014-001 

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at the Savannah River Site December 2014 

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Page 10-4 

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W231206, Bldg. 241-H Additional Waste Storage Tanks, Riser & Plug Details Concrete & Instruments 

Process, Savannah River Site, Aiken, SC, Rev. 27, January 30, 1995. 

W449795, Bldg. 241-12H, Tanks 35, 36 &37, Additional Waste Storage Tanks Top Slab Plan – Tank #35 

Concrete, Savannah River Site, Aiken, SC, Rev. 39, July 24, 1996. 

W704010, Bldg. 241-54F, Tank 44, 45, 46 47, Additional Waste Storage Tanks Liner Plate Attachments 

Sh-2 Steel, Savannah River Site, Aiken, SC, Rev. 15, February 23, 1978. 

WDPD-13-40, Hennessey, B.T., Request for Concurrence to Proceed to Sample and Analysis Phase of 

the Tank Closure Process for Tank 16H (CERCLIS Number: 89), Savannah River Site, Aiken, SC, April 4, 

2013. 

WDPD-14-20, Hennessey, B.T., Request for Concurrence to Proceed to Sample and Analysis Phase of 

the Tank Closure Process for Tank 12H (CERCLIS Number: 89), Savannah River Site, Aiken, SC, 

January 21, 2014. 

WHJISB03.STGD000100, Liquid Waste Disposition Project Training, Savannah River Site, Aiken, SC, 

April 12, 2007. 

WSRC-IM-98-00033, Cercy, M.J., Systems Engineering Methodology Guidance Manual, Savannah River 

Site, Aiken, SC, Rev. 14, March 19, 2008. 

WSRC-IM-2004-00008, DSA Support Document – Site Characteristics and Program Descriptions,

Savannah River Site, Aiken, SC, Rev. 1, June 2007. 

WSRC-MS-2003-00617, Stevenson, D.A., et al., 2001-2002 Upper Three Runs Sequence of Earthquakes 

at the SRS, South Carolina, Savannah River Site, Aiken, SC, 2003. 

WSRC-OS-94-42, Federal Facility Agreement for the Savannah River Site, 

http://www.srs.gov/general/programs/soil/ffa/ffa.pdf, Savannah River Site, Aiken, SC, August 16, 1993. 

WSRC-RP-89-78-9B_p4, Waste Management Monthly Data Record, September 1989, Savannah River 

Site, Aiken, SC, December 7, 1989. 

WSRC-RP-91-78-8B, Waste Management Monthly Data Record, August 1991, Savannah River Site, 

Aiken, SC, September 30, 1991. 

WSRC-RP-92-78-7B, High Level Waste Engineering Data Record, July 1992, Savannah River Site, 

Aiken, SC, August 28, 1992. 

WSRC-RP-94-54, Thayer, P.A., et al., Petrographic Analysis of Mixed Carbonate-Clastic 

Hydrostratigraphic Units in the General Separations Area (GSA), Savannah River Site (SRS), Aiken, 

South Carolina, Savannah River Site, Aiken, SC, Rev. 0, December 17, 1993. 

WSRC-RP-94-383-2, High Level Waste Engineering Monthly Data Report, February 1994, Savannah 

River Site, Aiken, SC, February 1994. 

WSRC-RP-95-841-6, High Level Waste Engineering Monthly Data Report, June 1995, Savannah River 

Site, Aiken, SC, June 1995. 

WSRC-RP-2001-00024, HLW Tank 18 Waste Removal Systems Engineering Evaluation, Savannah River 

Site, Aiken, SC, Rev. 0, May 22, 2001. 
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WSRC-STI-2006-00198, Phifer, M. A., et al., Hydraulic Property Data Package for the E-Area and Z-Area 

Soils, Cementitious Materials, and Waste Zones, Savannah River Site, Aiken SC, Rev. 0, September 

2006. 

WSRC-STI-2007-00184, Denham, M., et al, FTF Closure Cap Concept and Infiltration Estimates, 

Savannah River Site, Aiken, SC, Rev. 2, October 15, 2007. 

WSRC-STI-2007-00192, Hay, M.S., et al., Characterization and Actual Waste Tests with Tank 5F 

Samples, Savannah River Site, Aiken, SC, Rev. 1, August 30, 2007. 

WSRC-STI-2007-00369, Dixon, K., et al., Hydraulic and Physical Properties of Tank Grouts and Base Mat 

Surrogate Concrete for FTF Closure, Savannah River Site, Aiken, SC, Rev. 0, October 23, 2007. 

WSRC-STI-2007-00613, Kabela, E.D., Weber, A. H., Summary of Data Processing for the 2002-2006 

SRS Meteorological Database, Savannah River Site, Aiken, SC, December 2007. 

WSRC-STI-2007-00641, Langton, C.A., et al., Grout Formulations and Properties for Tank Farm Closure,

Savannah River Site, Aiken, SC, Rev. 0, November 12, 2007. 

WSRC-STI-2008-00057, Savannah River Site Environmental Report for 2007, Savannah River Site, 

Aiken, SC, 2008. 

WSRC-STI-2008-00172, Hansen, E.K., et al., Closure of HLW Tanks – Formulation for a Cooling Coil 

Grout, Savannah River Site, Aiken, SC, Rev. 0, April 2008. 

WSRC-STI-2008-00298, Cozzi, A.D., et al., Closure of HLW Tanks - Phase 2 Full Scale Cooling Coils 

Grout Fill Demonstrations, Savannah River Site, Aiken, SC, Rev. 0, June 2008. 

WSRC-TM-95-1, SRS Engineering Standards Manual, Standard 01064, Radiological Design 

Requirements, Savannah River Site, Aiken, SC, Rev. 5, June 22, 2009. 

WSRC-TR-90-0284, Stephenson, D.E., Review of Seismicity and Ground Motion Studies Related to 

Development of Seismic Design Criteria at SRS, Savannah River Site, Aiken, SC, Rev. 0, August 1992. 

WSRC-TR-95-0046, Denham, M.E., SRS Geology & Hydrogeology Environmental Information Document, 

Savannah River Site, Aiken, SC, June 1999. 

WSRC-TR-96-0399-Vol. 1, Flach, G.P., et al., Integrated Hydrogeological Modeling of the General 

Separations Area, Vol. 1: Hydrogeologic Framework, Savannah River Site, Aiken, SC, Rev. 0, August 

1997. 

WSRC-TR-96-0399-Vol. 2, Flach, G.P., et al., Integrated Hydrogeological Modeling of the General 

Separations Area, Vol. 2: Groundwater Flow Model, Savannah River Site, Aiken, SC, Rev. 1, April 1999. 

WSRC-TR-97-00322, Savannah River Site Environmental Report for 1997, Savannah River Site, Aiken, 

SC, 1998. 

WSRC-TR-98-00312, Savannah River Site Environmental Report for 1998, Savannah River Site, Aiken, 

SC, 1999. 

WSRC-TR-99-00299, Savannah River Site Environmental Report for 1999, Savannah River Site, Aiken, 

SC, 2000. 

WSRC-TR-99-00369, Chen, K.F., Flood Hazard Recurrence Frequencies for C-, F-, E-,S-, H-, Y-, and ZAreas, Savannah River Site, Aiken, SC, September 30, 1999. 

WSRC-TR-99-4083, Aadland, R.K., et al., Significance of Soft Zone Sediments at the SRS, Savannah 

River Site, Aiken, SC, September 1999. 

WSRC-TR-2000-00310, Cumbest, R.J., et al., Comparison of Cenozoic Faulting at the Savannah River 

Site to Fault Characteristics of the Atlantic Coast Fault Province: Implications for Fault Capability, 

Savannah River Site, Aiken, SC, 2000. 

WSRC-TR-2000-00328, Savannah River Site Environmental Report for 2000, Savannah River Site, 

Aiken, SC, 2001. 

WSRC-TR-2001-00313, Stefanko, D., TNX/HLW Long Shaft Pumps 1995-2000, Savannah River Site, 

Aiken, SC, Rev. 0, June 30, 2001. 
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WSRC-TR-2001-00474, Savannah River Site Environmental Report for 2001, Savannah River Site, 

Aiken, SC, 2002. 

WSRC-TR-2002-00403, Pasala, N.R., Tank Farm Transfer Control Program & Pump Tank Transfer Jet 

Control Program, Savannah River Site, Aiken, SC, Rev. 19, August 2012. 

WSRC-TR-2003-00026, Savannah River Site Environmental Report for 2002, Savannah River Site, 

Aiken, SC, 2003. 

WSRC-TR-2003-00401, Barnes, M.J., Waste Tank Heel Chemical Cleaning Summary, Savannah River 

Site, Aiken, SC, Rev. 0, September 9, 2003. 

WSRC-TR-2004-00015, Savannah River Site Environmental Report for 2003, Savannah River Site, 

Aiken, SC, 2004. 

WSRC-TR-2004-00317, Pike, J., Flowsheet for SRS Waste Tank Heel Removal Using Oxalic Acid, 

Savannah River Site, Aiken, SC, Rev. 0, November 10, 2004. 

WSRC-TR-2005-00005, Savannah River Site Environmental Report for 2004, Savannah River Site, 

Aiken, SC, 2005. 

WSRC-TR-2005-00201, Wike, L.D., et al., SRS Ecology: Environmental Information Document, 

Savannah River Site, Aiken, SC, March 2006. 

WSRC-TR-2006-00007, Savannah River Site Environmental Report for 2005, Savannah River Site, 

Aiken, SC, 2006. 

WSRC-TR-2007-00008, Savannah River Site Environmental Report for 2006, Savannah River Site, 

Aiken, SC, 2007. 

WSRC-TR-2007-00118, Kabela, E.D., et al., Savannah River Site Annual Meteorology Report for 2006, 

Savannah River Site, Aiken, SC, April 2007. 

WSRC-TR-2007-00327, Digsby, T.L., Consolidated Hazard Analysis for Mechanical Cleaning of Waste 

Tanks 18 & 19, Savannah River Site, Aiken, SC, Rev. 0, November, 2007. 

X-CLC-H-00921, Sudduth, C., Effects of Low Temperature Aluminum Dissolution (LTAD) on Tank 12 

Sludge Heel, Savannah River Site, Aiken, SC, Rev. 0, June 6, 2012. 

X-SD-G-00001, Harrison, E.W., Waste Acceptance Criteria for Liquid Waste Transfers to the 241-F/H 

Tank Farms, Savannah River Site, Aiken, SC, Rev. 32, June 14, 2012. 
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11.0 GLOSSARY 

Actinide Group of elements of atomic number 89 through 103 including thorium, 

uranium, neptunium, plutonium, americium and curium. 

Alternate Cases In addition to the Base Case (i.e., Case A) which reflects the most probable 

and defensible waste tank configuration (i.e., expected conditions for the HTF 

closure system) other modeling cases which reflect alternate waste tank 

configurations were also considered in the HTF PA. The alternate cases 

reflect different modeling assumptions with respect to key modeling 

parameters so as to allow evaluation of sensitivities and uncertainties 

associated with Base Case modeling assumptions. The alternate waste tank 

configurations are described in detail in Section 4.4.2 of the HTF PA. 

Ancillary Structures Structures associated with the waste storage tanks, such equipment as 

transfer line piping, pump tanks or evaporators, which are used to distribute or 

control the transfer of waste from one storage point to another storage point, 

or are used to volume reduce the waste. 

Air Content Amount of air incorporated into the grout as the result of mixing and 

placement. 

Annulus Also referred to as the secondary containment of a waste tank, surrounds the 

primary tank of Types I, II, III and IIIA tanks, providing a location for collection 

of any leakage from the primary tank. 

Base Case The waste tank system configuration modeling case within the HTF PA that 

represents the most probable and defensible estimate of expected conditions 

for the HTF closure system based on currently available information. 

Basemat Concrete pad upon which the waste tank is constructed. The pad has close 

tolerances for tank leveling and the concrete is quality controlled to ensure the 

structural integrity to tank foundation. The basemat is also referred to as floor 

slab or foundation. 

Bleed Water Water that separates from the grout as the result of solids settling. 

Calcareous Zone Located within the Santee Formation and the lowermost part of the overlying 

Dry Branch Formation, zones consist of silty and clayey fine sands, finegrained clays, and calcareous shell fragments deposited in nearshore and 

inner shelf environments. Soft zones within the calcareous zones near the 

General Separations Area, which includes the HTF, are not cavernous voids, 

but are small, isolated, poorly connected, three-dimensional features filled with 

loose, fine-grained, water-saturated sediment. 

Chromate Cooling Water Coolant comprised of chromate-inhibited water that circulates through the 

cooling coils of waste tanks to remove radioactive decay heat and other 

sources of heat (e.g., steam heat loads, ventilation heat loads or mechanical 

heat loads from pumping/mixing operations). 

Closure Plan Plan that presents the environmental regulatory standards and guidelines 

pertinent to the closure of the tanks and describes that process for evaluating 

and selecting the closure configuration (i.e., residual inventory and form). 

Compressive Strength Force per unit area required to break an unconfined grout or concrete sample. 

A 

B 

C 
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Concentration Amount (e.g., in grams or moles) per volume of a substance. 

Conductivity Probes A simple electrical device that works on the principle that liquids conduct 

electricity more readily than air. If a liquid comes in contact with the probe it 

will complete an electrical circuit and send a signal for indication or alarm 

purposes. 

Cooling Coils Coils installed in the tanks to remove the decay heat that is generated by the 

waste in the tanks. Arrangements and designs of cooling coils differ, 

depending on the type of tank. Type I and II tanks, in addition to having 

vertical cooling coils, also have cooling coils across the bottom of the tank to 

provide a means for cooling the bottom of the tank. 

Contamination In the context of this HTF 3116 Basis Document, contamination refers to 

radioactive materials that have been projected to migrate from the closed HTF.

Core Pipe Internal pipe of transfer line that comes into contact with the waste materials. 

The core pipe is usually located within a jacket pipe. 

Curie A unit of radioactivity - the quantity of nuclear material that has 3.7E+10 

disintegrations per second.

Diffusion Movement of contaminants from an area of higher concentration to an area of 

lower concentration. 

Diversion Box A shielded reinforced concrete structure containing transfer line nozzles to 

which jumpers are connected in order to direct waste transfers to the desired 

location. 

Eh The symbol for reduction/oxidation (redox) potential in millivolts. 

Evaporator Steam-heated, water-cooled system installed in the tank farms to concentrate 

underground waste storage tank contents, in order to reduce the liquid waste 

volume. 

Exposure Being exposed to ionizing radiation or to radioactive material. 

Federal Facility 

Agreement 

Agreement between EPA, DOE and SCDHEC that directs the comprehensive 

remediation of the Savannah River Site. It contains requirements for (1) site 

investigation and remediation of releases and potential releases of hazardous 

substances and (2) interim status corrective action for releases of hazardous 

wastes or hazardous constituents. 

Fly Ash A mineral admixture used in grout to enhance finishing characteristics, make 

the mix more economical and to improve pumping. It is finer in consistency 

than cement and its particles are round. These fine particles make the mix 

finish and pump more easily. 

 

D 

E 

F 
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General Separations 

Area 

Centralized area of SRS including, E, F, H, J, S and Z Areas that are the 

heavily industrialized areas of SRS. 

Grout A cement mixture, sufficiently fluid, which can be pumped into waste tanks and 

equipment cavities creating a watertight bond and increasing the strength of 

the existing structural foundation. Capable of slowing the vertical movement 

or migration of water. 

H-Modified Process The modified PUREX process used in H-Canyon for separation and recovery 

of enriched uranium from used reactor fuel. 

High-Heat Waste Waste streams from the first cycle extractions from F-Canyon and H-Canyon 

Separations Facilities. 

Hydraulic Conductivity Velocity of water flow through saturated materials (e.g., concrete, grout, soil). 

Institutional Control A 100-year period in which DOE retains ownership and control of HTF such 

that HTF facility maintenance and controls will be performed to prevent 

inadvertent intrusion and protect public health and the environment. 

Leak Detection Boxes Structures that provide for the collection and detection of leakage from the 

transfer lines. 

Line Encasement 

(Sealed Concrete 

Trench) 

Enclosed core pipes in a covered reinforced concrete encasement below 

ground. Any core pipe leakage into the encasement and in-leakage of 

groundwater into the encasement will gravity drain to a catch tank. 

Low-Heat Waste Waste streams from the second cycle extractions from F-Canyon and 

H-Canyon Separations Facilities. 

NDAA Section 3116 The Ronald W. Reagan National Defense Authorization Act for Fiscal Year 

2005 Section 3116 was passed by Congress on October 9, 2004 and signed 

by the President on October 28, 2004. NDAA Section 3116 specifies that the 

term “high-level radioactive waste” does not include radioactive waste that 

results from reprocessing spent nuclear fuel if the Secretary of Energy 

determines, in consultation with the NRC, that the waste meets certain criteria. 

Oxalic Acid A relatively strong organic acid, about 10,000 times stronger than acetic acid. 

 

G 

H 

I 

L 

N 

O 
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Permeability Capability of a material to let pass other molecules or particles. 

pH A measure of the acidity or alkalinity of a solution, numerically equal to 7 for 

neutral solutions, increasing with increasing alkalinity and decreasing with 

increasing acidity. 

Pitting Localized corrosion of a metal surface, confined to a point or small area that 

takes the form of cavities. 

Porosity Grout porosity is generally defined as the percentage of total volume of cured 

grout that is not occupied by the starting cementitious materials and the 

products that result from reaction of these cementitious materials with water. 

Primary Tank The primary tank, sometimes referred to as the “shell” or “primary liner”, is the 

component of the waste tank that actually contains the liquid waste. The 

primary tank is contained within the secondary containment, if any. 

Progeny Decay products or descendants of specific radionuclides. 

Pump Pit Shielded reinforced concrete structures located below grade at the low points 

of transfer lines, contain pump tanks and are usually lined with stainless steel. 

Pump Tank All HTF pump pits (except HPP-1) house a pump tank with the pump pits 

providing secondary containment for pump tanks. The pump tanks have a 

nominal capacity of 7,200 gallons each. The pump tanks installed in HTF are 

all of the same basic size (8.5 feet tall, 12 feet in diameter) and are lined with 

stainless steel. 

PUREX Process The Plutonium Uranium Extraction process used in the F-Canyon and HCanyon to extract special nuclear material from aluminum-clad, depleted 

uranium targets which had been irradiated in the site’s nuclear production 

reactors. H-Canyon utilized the PUREX process from 1955 to 1959 prior to 

conversion to the HM process. 

Residuals For the purposes of this HTF 3116 Basis Document, the residual waste 

remaining in a waste tank or ancillary structure - following successful 

completion of waste removal activities and removal of highly radioactive 

radionuclides to the maximum extent practical - is referred to as “residuals”. 

Riser The risers are openings through the tank tops providing access to the tank and 

annulus interiors. Risers are used primarily to provide for the installation of 

equipment such as pumps and cooling equipment; instrumentation such as 

level probes and leak detection; ventilation; and to provide access to the tank 

interior for sampling, depth measurement and inspection. 

Saltcake Saltcake located in waste tanks consists of crystallized salts with interstitial 

void space and small quantities of entrained insoluble solids (assumed to be 

partially sludge solids). 

Saltstone Production A process in which low-activity salt solution is mixed with dry chemicals 

(cement, slag and fly ash) to form a homogeneous grout mixture. 

Sand Layer All Type II waste tanks have a one-inch thick primary sand layer between the 

primary tank and secondary liner and a one-inch thick secondary sand layer 

between the secondary liner and basemat. These sand layers are also 

commonly referred to as "Sand Pads". 

P 

R 

S 
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Secondary Containment Also referred to as an annulus of a waste tank. The secondary containment 

surrounds the primary tank of Types I, II, III and IIIA tanks, providing a location 

for collection of any leakage from the primary tank. 

Segregation Separation of sand from binder as the result of impact and separation of water 

from grout as the result of gravity settling of the solids from the grout slurry. 

Set Time Time after mixing at which the grout responds as a solid. 

Shotcrete Concrete conveyed through a hose and pneumatically projected at high 

velocity onto a surface. Shotcrete undergoes placement and compaction at 

the same time due to the force with which it is projected from the nozzle. 

Shotcrete was used in the construction of Type IV tanks. 

Slag Slag was introduced into the design mixes which in addition to its hydraulic 

activity, also provides chemical reducing power to the mix. Slag has been 

shown to possess chemically reducing properties that are favorable for 

technetium reduction and for plutonium and selenium. 

Solubility The ability of a substance to dissolve in a solvent. Solubility may also refer to 

the measure of this ability for a particular substance in a particular solvent, 

equal to the quantity of substance dissolving in a fixed quantity of solvent to 

form a saturated solution under specified temperature and pressure. The 

extent of the solubility of a substance in a specific solvent is measured as the 

saturation concentration, where adding more solute does not increase the 

concentration of the solution. The extent of solubility ranges widely, from 

infinitely soluble, such as ethanol in water, to poorly soluble, such as silver 

chloride in water. The term insoluble is often applied to poorly or very poorly 

soluble compounds. 

Source Term The amount and type of radioactive material released into the environment. 

Stabilized Contaminant Grouted waste remaining in the waste tanks or ancillary equipment after 

system closure. 

Stochastic A probabilistic distribution of parameters. 

Supernate Liquid salt solution found above the sludge layer after settling of solids in 

waste tanks has occurred as a result of a liquid waste transfer to one of the 

waste processing facilities or receipt tanks. Also referred to, generally, as 

supernatant. 

Underliner Sump An engineered feature that collects any leakage through the concrete or 

stainless steel liners beneath waste tanks. 

Unit Weight Weight of a unit volume, typically one cubic foot. 

Valve Boxes Transfer valve boxes facilitate specific waste transfers that are conducted 

frequently. The valves are generally manual ball valves in removable jumpers 

with flush water connections on the transfer piping. The valve boxes provide 

containment of, and access to, the valves. 

Vault Term used to describe the underground concrete floor, walls and roof that 

enclose the steel primary tank and secondary liner, if applicable, in the waste 

tank system. 

Viscosity Rheological quality of fluids describing the resistance to flow. 

 

U 

V 
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Waste Characterization 

System 

Computer-based system designed to integrate historical information, current 

sample data, and physical properties of constituents to develop predictions of 

concentrations and inventory. 

Working Slab Concrete surface usually placed to create a level construction surface. This 

concrete is normally lower quality without reinforcement and is either broken 

up after or cracked during construction activities between the tanks, thus is not 

considered a barrier to vertical water migration. 

W 
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APPENDIX A: LIQUID WASTE SYSTEM DESCRIPTION 

Appendix Purpose 

The purpose of this appendix is to provide a brief overall description of the SRS Liquid 

Waste System. 

Appendix Contents 

This appendix briefly describes the history of the SRS underground radioactive waste 

storage tanks and their contents, and describes the methods used to treat and dispose of 

this waste. 

Key Points 

 SRS has 51 waste tanks within two tanks farms, FTF and HTF, which entered service 

between 1954 and 1986. 

 There are four basic types of waste tanks, designated Types I, II, III/IIIA and IV. 

 The 27 Type III/IIIA tanks meet current EPA requirements for full secondary 

containment and leak detection; the other 24 do not meet these requirements. 

 As of July 1, 2014, approximately 36,800,000 gallons of radioactive waste were stored 

in the waste tanks, most of this from separation and recovery of special nuclear 

materials and enriched uranium in the two SRS nuclear materials processing facilities 

known as F-Canyon and H-Canyon. 

 The high-level waste fraction removed from the waste tanks (including the sludge 

waste) is being converted into borosilicate glass by the vitrification process that takes 

place in the DWPF, with the solidified glass contained in stainless steel canisters. 

 Salt waste removed from the waste tanks is pretreated with the resultant low-volume, 

high-activity fraction being sent to DWPF and the high-volume, low-activity fraction 

(referred to as decontaminated salt solution) being treated and disposed of as a nonhazardous, cementitious waste form (i.e., saltstone) in the SDF. 

A.1 Background 

The SRS is an approximately 310 square mile site located in the state of South Carolina and bordering 

the Savannah River. Since it became fully operational in 1954, it has produced nuclear material for 

national defense, research, medical, and space programs. The separation of nuclear material from 

irradiated targets and used fuels resulted in the generation of large quantities of radioactive liquid waste, 

which is currently stored on-site in large underground radioactive waste storage tanks. 

Most of the waste tank inventory currently stored at SRS is a complex mixture of chemical and radioactive 

waste generated during the separation of special nuclear materials and enriched uranium from irradiated 

targets and used fuel using the PUREX process in F-Canyon and the HM process in H-Canyon. Waste 

generated from the recovery of Pu-238 in H-Canyon for the production of heat sources for space missions 

is also included. The variability in both nuclide and chemical content in this liquid radioactive waste is 

due, in part, to the fact that waste streams from the first cycle (high-heat) and second cycle (low-heat) 

extractions from each canyon were typically stored in separate waste tanks to better manage waste heat 

generation. 

When these acidic streams were pH-adjusted with caustic to form a high alkaline solution and transferred 

to a waste tank in one of the two tank farms, the resultant precipitate settled into four characteristic sludge 

consistencies. Typically, this sludge waste can still be found in the waste tanks where it was originally 

deposited beginning in 1954. Historically, new waste receipts into the tank farms have been segregated 

into four general categories in the SRS tank farms: PUREX high activity waste (HAW), PUREX low 

activity waste (LAW), HM HAW and HM LAW. Because of this segregation, settled sludge solids 

contained in waste tanks that received new waste are readily identified as one of these four categories. 

Fission product concentrations are about three orders of magnitude higher in both PUREX and HM HAW 

sludge than the corresponding LAW sludge. 
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The soluble portions of the first and second cycle waste were similarly partitioned but have, and continue 

to undergo, blending in the course of waste transfer and staging of salt waste for evaporative 

concentration. Beginning in 1964, large evaporator systems were used to volume reduce the liquid salt 

waste to form two distinct waste types: concentrated supernate and saltcake. Combining and blending 

salt solutions has tended to reduce soluble waste into blended PUREX salt and concentrate and HM salt 

and concentrate, rather than maintaining four distinct salt compositions. Continued blending and 

evaporation of the salt solution deposits crystallized salts with overlying and interstitial concentrated salt 

solution in salt tanks located in both FTF and HTF. More recently, with transfers of sludge slurries to 

sludge washing tanks, removal of saltcake to support salt waste pretreatment, receipts of DWPF recycle, 

and space limitations restricting full evaporator operations, salt solutions have been transferred between 

the two tank farms. Intermingling of PUREX and HM salt waste will continue through the life of the 

program. [SRR-LWP-2009-00001]

A.2 Liquid Waste System 

The Liquid Waste System is the integrated series of facilities at SRS that safely manage the storage of 

existing waste inventory in the SRS waste tanks, support the transfer and waste reduction of this waste, 

perform the pretreatment of waste in preparation for eventual waste disposal, and provides for the 

permanent disposal of the high-volume, low-activity decontaminated salt solution, as well as other smaller 

low-level waste liquid streams, in the SDF. Key facets of the Liquid Waste System are briefly described in 

the text that follows. 

A.2.1 Waste Tank Storage 

The SRS has a total of 51 waste tanks located within two tank farms, FTF (22 waste tanks) and HTF (29 

waste tanks). These waste tanks were placed into operation between 1954 and 1986. There are four 

distinct waste tank designs, Types I through IV. Type III/IIIA tanks are the newest waste tanks and were 

placed into operation between 1969 and 1986. There are a total of 27 Type III/IIIA tanks. Figure A.2-1 

shows Type III tanks during construction. These waste tanks meet current EPA requirements for full 

secondary containment and leak detection. The remaining 24 waste tanks do not meet EPA 

requirements for secondary containment. The twelve Type I tanks are the oldest waste tanks and were 

constructed in the early 1950s. The four Type II waste tanks were constructed between 1955 and 1956. 

There are eight Type IV waste tanks that were constructed between 1958 and 1962. Two of these Type 

IV waste tanks, Tanks 17 and 20, were removed from service and filled with grout in 1997 under 

SCDHEC approved GCP. [PIT-MISC-0002, PIT-MISC-0004] Two additional Type IV tanks, Tanks 18 

and 19, were removed from service and 

filled with grout in 2012. Two FTF Type I 

tanks, Tanks 5 and 6, were removed from 

service and filled with grout in 2013. 

Tanks 5, 6, 18 and 19 were removed from 

service and grouted under the SCDHEC 

approved FTF GCP. [LWO-RIP-2009-

00009] In addition, two HTF tanks, Tanks 

12 and 16, have completed cleaning 

activities and are awaiting completion of 

residual sampling and analysis and 

subsequent final closure documentation. 

Thirteen SRS waste tanks have a history of 

leakage. These waste tanks include Type 

I, Type II and Type IV designs. No Type III 

or IIIA tanks have developed leak sites. 

Eight of the 13 waste tanks with a history 

of leakage are located in HTF (Tank 9, 

Tank 10, Tank 11, Tank 12, Tank 13, Tank 

14, Tank 15 and Tank 16). [SRR-STI2014-00243] Sufficient waste has been 

removed from these waste tanks such that 

Figure A.2-1: Waste Tanks Under Construction 
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there are currently no active leak sites. [SRR-LWP-2009-00001] Only once has waste leaking from a 

primary waste tank reached the environment. This event occurred in September 1960 and was 

associated with Tank 16 in HTF. Based on a review of historical waste levels in the primary tank and 

annulus, and subsequent soil and groundwater sampling that was performed, it was determined that “only 

a few tens of gallons of waste” reached the surrounding soils. [DP-1358] In all other cases, leakage from 

a primary waste tank has been confined within the secondary containment annuli and has not reached 

the surrounding soils. 

Although outside the scope of this determination, waste did reach the soil from a spill associated with 

Tank 8. In 1961, Tank 8 was inadvertently over-filled and waste escaped along the junction area where a 

transfer line entered Tank 8. The majority of the liquid was contained within the secondary encasement 

associated with the transfer lines for Tanks 1 through 8 in FTF, but an estimated 1,500 gallons spilled into 

the surrounding soil. [DPSPU 75-11-8] 

A.2.2 Waste Tank Space Management 

To make better use of available waste tank storage capacity, incoming liquid waste is evaporated to 

reduce its volume. Since 1951, the tank farms have received over 140,000,000 gallons of liquid waste, of 

which over 100,000,000 gallons have been evaporated. As of July 1, 2014, approximately 36,800,000 

gallons of liquid radioactive waste was stored in the waste tanks. Projected available waste tank space is 

carefully tracked to ensure that the tank farms do not become “water logged,” a term meaning that so 

much of the usable Type III/IIIA tank space, the waste tanks that have full secondary containment 

systems, has been filled that normal operations, waste removal, and waste processing operations cannot 

continue. A portion of the Type III/IIIA tank space must be reserved as contingency space should a new 

waste tank leak be realized. [SRR-LWP-2014-00030] 

Waste receipts and transfers are normal tank 

farm activities as the tank farms receive new 

waste from the H-Canyon legacy material 

stabilization program, liquid waste from DWPF 

processing (typically referred to as “DWPF 

recycle”), and wash water from sludge washing. 

The tank farms also make routine transfers to 

and from waste tanks and evaporators. 

Currently, there is very little waste that has not 

had the water evaporated from it to its 

maximum extent. The working capacity of the 

tank farms has steadily decreased and this 

trend will continue until sufficient salt waste has 

been treated and disposed of in SDF. Two 

evaporator systems are currently operating at 

SRS, the 2H and 3H systems. Evaporator 

operations are currently impacted due to limited 

salt waste storage space. [SRR-LWP-2009-

00001] 

As of July 1, 2014, approximately 272,000,000 

curies of radioactivity were stored in the SRS 

tank farms. This waste is a complex mixture of 

insoluble metal hydroxide solids, commonly 

referred to as sludge, and soluble salt 

supernate. The supernate volume is reduced 

by evaporation which also concentrates the 

soluble salts to their solubility limit. The resultant solution either crystallizes as salts or remains as a 

concentrated supernate solution (Figure A.2-2). The resulting crystalline solids are commonly referred to 

as saltcake. [SRR-LWP-2014-00030] 

Figure A.2-2: Supernate (Top) and 

Saltcake (Bottom) 
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The sludge component (Figure A.2-3) of the 

radioactive waste represents approximately 

2,700,000 gallons (7 % of total) of waste volume 

but contains approximately 135,000,000 curies 

(50 % of total) of the radioactivity. The salt 

waste makes up the remaining 34,200,000 

gallons (93 % of total) of waste and contains 

approximately 137,000,000 curies (50 % of total) 

of the radioactivity. Of that salt waste, the 

supernate accounts for approximately 

18,000,000 gallons and 125,000,000 curies of 

the 137,000,000 curies total salt waste related 

activity. The saltcake accounts for 

approximately 16,200,000 gallons and 

12,000,000 curies of the remaining salt waste. 

The sludge contains the majority of the longlived (half-life greater than 30 years) 

radionuclides (e.g., actinides) and strontium. 

Figure A.2-4 shows the breakdown of this 

waste. [SRR-LWP-2014-00030] 

Radioactive waste volumes and radioactivity 

inventories reported herein are based on the 

WCS database, which includes the chemical 

and radionuclide inventories on a tank-by-tank 

basis. The WCS is a dynamic database frequently updated with new data from ongoing operations, such 

as decanting and concentrating of free supernate via evaporators, preparation of sludge batches for 

DWPF feed, waste transfers between waste tanks, waste sample analyses, and influent receipts such as 

H-Canyon waste and DWPF recycle. Volumes and curies referenced in this appendix are current as of 

July 1, 2014. [SRR-LWP-2014-00030] 

Figure A.2-4: FTF and HTF Waste Tank Composite Inventory 

Figure A.2-3: Sludge Component of 

Radioactive Waste 

34.2 Mgal 

Volume

36.8 Mgal

Curies

137 MCi 

(50 %) 

135 MCi 

(50 %) 

272 MCi 

125 MCi 

(46 %) 

Sludge

(93 %) 

2.7 Mgal 

(7 %) 

18.0 Mgal

(49 %) Salt Supernate

12 MCi 

Saltcake (4 %) 

16.2 Mgal

(44 %) 

Inventory values as of July 1, 2014 

[SRR-LWP-2014-00030] 

Note: Due to rounding, inventory values may differ slightly 

from those contained in the 07/01/2014 ─ July 2014 Curie 

and Volume Inventory Report. [SRR-LWP-2014-00030] 
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Approximately 91 % of the salt waste radioactivity is short-lived (half-life 30 years or less) Cs-137 and its 

daughter product, Ba-137m, along with lower levels of actinide contamination. Depending on the 

particular waste stream (e.g., canyon waste, DWPF recycle waste), the cesium concentration may vary. 

The precipitation of salts following evaporation can also change the cesium concentration. The 

concentration of cesium is significantly lower than non-radioactive salts in the waste, such as sodium 

nitrate and nitrite; therefore, the cesium does not reach its solubility limit and only a small fraction 

precipitates. [SRR-LWP-2009-00001] As a result, the cesium concentration in the saltcake is much lower 

than that in the liquid supernate and interstitial liquid fraction of the salt waste. 

A.3 Safe Disposal of the Waste 

In the 1980s, waste removal operations were initiated in the SRS tank farms. Ultimately, the goal is 

remove and treat the waste, safely closing the waste tanks and disposing of the waste in one of two final 

waste forms: glass, which will contain greater than 99 % of the radioactivity, and saltstone, which will 

contain the vast amount of volume but less than one percent of the radioactivity. Both the salt waste and 

the sludge waste must be uniquely treated to prepare these two waste forms for disposal. The sludge 

must be washed to remove non-radioactive salts that would interfere with glass production. The washed 

sludge can then be sent to DWPF for vitrification. The salt must be treated to separate the bulk of the 

radionuclides from the non-radioactive salts in the waste. This separation will be accomplished in SWPF. 

However, until the startup of the SWPF, additional salt treatment processes, known as Interim Salt 

Processing, will be used to accomplish this separation on a limited amount of the lower activity salt waste. 

A.3.1 Salt Processing 

A final DOE technology selection for salt processing was completed and a Salt Processing EIS ROD was 

issued in October 2001. [DOE/EIS-0303 ROD] The ROD designated Caustic Side Solvent Extraction 

(CSSX) as the preferred alternative to be used to separate cesium from the salt waste. Based on the 

current Liquid Waste System Plan, DOE anticipates using a two-phase, three-part process to treat salt 

waste. [SRR-LWP-2009-00001] 

During the initial Interim Salt Processing phase, relatively small volumes of the lower activity will be 

treated using the following two salt waste treatment processes: 

 Deliquification, Dissolution and Adjustment (DDA): For salt waste in Tank 41 as of June 9, 

2003, the treatment of deliquification (i.e., extracting the interstitial liquid) was sufficient to 

produce a salt waste that met the SPF waste acceptance criteria. Deliquification is an effective 

decontamination process because the primary radionuclide in salt is Cs-137, which is highly 

soluble. To accomplish the process, the saltcake is first deliquified by draining and pumping. 

The deliquified saltcake is dissolved by adding water and pumping out the salt solution. The 

resulting salt solution is given time to allow additional insoluble solids to settle prior to being sent 

to the SPF feed tank. If necessary, the salt solution may be aggregated with other Tank Farm 

waste to adjust batch chemistry for processing at SPF. 

 Actinide Removal Process (ARP)/Modular CSSX Unit (MCU): For salt waste in selected 

waste tanks (e.g., Tank 25), further decontamination is performed. Salt waste from these waste 

tanks first will be sent to ARP. In ARP, monosodium titanate is added to the waste as a finely 

divided solid. Actinides are sorbed on the monosodium titanate and then filtered out of the liquid 

to produce a stream that is sent to MCU. The salt solution is further treated to reduce the 

concentration of Cs-137 using the CSSX process. 

Interim Salt Processing will be utilized pending the construction and start-up of the SWPF, the second 

and more robust phase of salt processing. The SWPF is a large, high-capacity facility that incorporates 

both the ARP and CSSX processes in a full-scale shielded facility capable of handling and effectively 

decontaminating salt waste with high levels of radioactivity. 
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A.3.2 Sludge Processing 

Sludge is “washed” to reduce the amount of 

soluble salts remaining in the sludge slurry. 

The processed sludge is called “washed 

sludge.” During sludge processing, large 

volumes of wash water are generated and 

must be volume-reduced by evaporation. 

Over the life of the waste removal program, 

the sludge currently stored in SRS waste 

tanks will be blended into separate sludge 

“batches” to be processed and fed to DWPF 

for vitrification. 

Final processing for the washed sludge and 

the high-activity fraction of the salt waste 

occurs at DWPF. This waste includes 

monosodium titanate sludge from ARP or 

SWPF, the cesium strip effluent from MCU or 

SWPF, and the washed sludge slurry. In a 

complex sequence of carefully controlled 

chemical reactions, this waste is blended with 

glass frit and melted to vitrify it into a 

borosilicate glass form. The resulting molten glass is poured into stainless steel canisters (Figure A.3-1). 

As the filled canisters cool, the molten glass solidifies (Figure A.3-2), immobilizing the radioactive waste 

within the glass structure. After the canisters have cooled, they are permanently sealed, and the external 

surfaces are decontaminated to meet United States Department of Transportation requirements. The 

canisters are then ready to be stored on an interim basis, on-site, in the Glass Waste Storage Building 

(GWSB). A low-level recycle waste stream from DWPF is returned to the tank farms. DWPF has been 

fully operational since 1996. [SRR-LWP-2009-00001] 

A.3.3 Saltstone: On-Site Disposal of Low-Level Waste 

The Saltstone Facility, located in Z Area, consists of two 

facility segments: SPF and SDF. The SPF is permitted as 

a wastewater treatment facility per SCDHEC Regulations. 

SPF receives and treats the decontaminated salt solution to 

produce saltstone by mixing the low-level waste liquid 

stream with cementitious materials (cement, flyash, and 

slag). A slurry of the components is pumped into the 

disposal cells located in SDF, where the saltstone grout 

solidifies into a monolithic, non-hazardous, solid low-level 

waste form called saltstone. SDF is permitted as an 

Industrial Solid Waste Landfill Facility (ISWLF). [SRR-LWP2009-00001] 

The SDF will be comprised of many large disposal units. 

Each of the disposal units will be filled with saltstone. The saltstone itself provides primary containment 

of the waste, and the walls, floor, and roof of the disposal units provide additional engineered barriers. 

The original SDF disposal unit (Vault 1) dimensions are approximately 100 feet wide, 600 feet long and 

27 feet high. The disposal unit is divided into two units which are 100 feet x 300 feet with a three-inch 

separation gap between units. Each unit is further divided into three cells, with each cell measuring 

approximately 100 feet x 100 feet. Thus, Vault 1 is comprised of six cells each approximately 100 feet x 

100 feet. 

The second disposal unit constructed (Vault 4) dimensions are approximately 200 feet wide, 600 feet long 

and 30 feet high. The disposal unit is divided into two units which are 200 feet wide and 300 feet long, 

with a three-inch separation gap between the units. Each unit is further divided into six cells, with each 

Figure A.3-1: Vitrification Canister Prior to Use 

Figure A.3-2: Sample of Vitrified 

Radioactive Glass 
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cell measuring approximately 100 feet x 100 feet. Thus, Vault 4 is comprised of 12 cells each 

approximately 100 feet x 100 feet. [SRR-CWDA-2009-00017] 

The current active disposal cells are nominally 150 feet diameter by 22 feet high each and designed in 

compliance with provisions contained in the Consent Order of Dismissal in Natural Resources Defense 

Council, et al. v. South Carolina Department of Health and Environmental Controls, et al. (South Carolina 

Administrative Law Court, August 7, 2007). [SRR-CWDA-2009-00017] Construction of three disposal 

units, comprised of Disposal Cells 2A, 2B, 3A, 3B, 5A and 5B, is complete in the SDF. 

An additional disposal unit (Saltstone Disposal Unit 6) is currently under construction. The new disposal 

unit is a larger design with dimensions, approximately 375 feet diameter by 43 feet high (Figure A.3-3). 

Closure operations will begin near the end of the active disposal period in the SDF, i.e., after most or all 

the disposal units have been constructed and filled. Backfill of native soil will be placed around the 

disposal units. The present closure concept includes an erosion barrier, two drainage layers along with 

backfill and a vegetative cover. [SRR-CWDA-2009-00017] 

Construction of the Saltstone Facility and the first two disposal units was completed between February 

1986 and July 1988. The Saltstone Facility started radioactive operations June 12, 1990. Future 

disposal cells will be constructed on an as-needed basis in coordination with salt processing production 

rates. [SRR-CWDA-2009-00017] 

 Figure A.3-3: Saltstone Facility (June, 2014) 
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APPENDIX B: APPROACH TO DOCUMENTING REMOVAL OF RADIONUCLIDES TO SUPPORT 

DOE CLOSURE AUTHORIZATION 

Appendix Purpose 

The purpose of this appendix is to outline the approach used by DOE for the SRS waste 

tanks and ancillary structures to document removal of radionuclides, with emphasis on 

HRRs. 

Appendix Contents 

This appendix briefly describes DOE’s closure approach leading to DOE Closure 

Authorization and provides details on each phase of DOE’s radionuclide removal process to 

support Closure Authorization. 

Key Points 

 In order to proceed with closure of the HTF, DOE Manual 435.1-1, DOE Guide 435.1-1 

and DOE practice requires that DOE issue Tier 1 Closure Authorization specifying the 

requirements necessary to achieve closure. 

 DOE Manual 435.1-1, DOE Guide 435.1-1 and DOE practice require that subsequent 

DOE Tier 2 Closure Authorization to proceed with permanent stabilization of a specific 

waste tank or ancillary structure be based on Tier 2 closure documentation. 

 DOE’s approach to radionuclide removal consists of the following phases: Initial 

Technology Selection, Technology Implementation, Technology Execution, Technology 

Effectiveness Evaluation and Additional Technology Evaluation. 

 For each waste tank or ancillary structure, documentation or information collected 

from each phase of the process eventually contributes to a final removal report 

supporting Tier 2 Closure Authorization. 

B.1.0 Executive Summary 

The discussions presented in this Appendix outline and describe the approach used by the DOE for each 

of the SRS waste tanks or ancillary structures to document removal of radionuclides, with emphasis on 

HRRs. This approach consists of the following phases: initial technology selection, technology 

implementation, technology execution, technology effectiveness evaluation and additional technology 

evaluation. For each waste tank or ancillary structure, documentation or information collected from each 

phase of the process eventually contributes to a removal report describing removal of radionuclides, with 

emphasis on HRRs, for that particular waste tank or ancillary structure. In some instances a report may 

be written to capture more than one tank or ancillary structure if several are removed from service at the 

same time. The removal report further integrates into documentation supporting removal from service 

and stabilization of waste tanks and ancillary structures in support of eventual closure of the SRS tank 

farms. The information documented in the removal report is utilized by the DOE-SR to support Tier 2 

Closure Authorization. 

B.2.0 Closure Approach 

In order to proceed with closure of the FTF and the HTF at the SRS, DOE Manual 435.1-1, DOE Guide 

435.1-1 and DOE practice requires that DOE issue Tier 1 Closure Authorization specifying the 

requirements necessary to achieve closure. The Tier 1 closure documentation defines the parameters, 

approach and plans by which tank farm closure activities will be accomplished. Once the specified 

documentation has been approved, an Authorization to Proceed will be issued. [DOE M 435.1-1, DOE G 

435.1-1] At SRS, separate Tier 1 closure documentation will be issued for FTF142 and HTF. The tank 

 

 

142 In March 2012, DOE issued Tier 1 closure documentation for FTF. 
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farm-specific Tier 1 closure documentation requirements are anticipated to include the following 

documentation: 

 Appropriate NEPA documentation 

 Tank Farm specific (FTF or HTF) performance assessment 

 SRS CA 

 Tank Farm specific (FTF or HTF) NDAA 3116 Determination by the Secretary, and its supporting 

3116 Basis Document 

 State-approved Tank Farm specific (FTF or HTF) Industrial Wastewater GCP 

At the completion of waste removal activities for a specific waste tank or ancillary structure, the individual 

tank or structure is removed from service. As required by DOE Manual 435.1-1, DOE Guide 435.1-1 and 

DOE practice, approval to proceed with permanent stabilization of the specific waste tank or ancillary 

structure will be based on Tier 2 closure documentation. [DOE M 435.1-1, DOE G 435.1-1] This Tier 2 

closure documentation provides the waste tank-specific or ancillary structure-specific information 

demonstrating that the process described in the Tier 1 closure documentation has been implemented and 

the criteria required by the Tier 1 closure documentation have been met. Figure B.2-1 shows the 

documentation pathway that leads to Tier 2 Closure Authorization. Tier 2 Authorization to proceed with 

closure (e.g., stabilization activities) will be issued by the Manager for DOE-SR. 

Figure B.2-1: Tier 2 Closure Authorization 

B.3.0 Radionuclide Removal Process to Support Tier 2 Closure Authorization 

B.3.1 General Approach 

The following section describes the progression of defined stages that must be utilized to support Tier 2 

Closure Authorization. The process presented begins following the bulk removal of the solids and liquid 

from a waste tank or ancillary structure. This final waste removal phase is typically referred to as “heel 

removal.” Waste removal will continue per the progression described in the following section. 

Proceeding with operational closure activities, e.g., stabilization, must be authorized through the Tier 2 

approval process. 

It should be noted that in some ancillary structures it may not be practical to undertake further removal of 

HRRs following bulk waste removal efforts. As a general matter, such a situation may arise if HRRs are 

present in such low quantities that they make an insignificant contribution to potential doses to workers, a 

hypothetical future member of the public, and the hypothetical future human intruder. 
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The contamination remaining in a waste tank or ancillary structure following successful completion of heel 

removal is referred to as “residuals.” 

B.3.2 Activities and Steps 

The approach is outlined in Figure B.3-1 and consists of the following phases, which are described in 

more detail in subsequent sections of this document: 

 Initial Technology Selection 

 Technology Implementation 

 Technology Execution 

 Technology Effectiveness Evaluation 

 Additional Technology Evaluation 

Throughout the waste tank or ancillary structure cleaning process various reports, evaluations, analyses, 

data, operational documents, and presentations are developed to support DOE’s final documentation 

supporting Tier 2 Closure Authorization. The level of specific data collection or documentation for each 

waste tank or ancillary structure will vary depending upon the attributes of the waste tank or ancillary 

structure and the technologies being implemented. For example, the documentation may be as simple as 

a memo to file when the planned technology is the same as the baseline (e.g., use of mixing pumps for 

mechanical heel removal) due to similar waste characteristics as a previously completed tank. However, 

a more formal report or documented systems engineering evaluation may be conducted if it is expected 

that the deployment of a new type of technology is needed. 

For each waste tank or ancillary structure, documentation or information collected from each phase of the 

process eventually contributes to a final removal report supporting Tier 2 Closure Authorization. In some 

instances a report may be written to capture more than one tank or ancillary structure if several are 

removed from service at the same time. 
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Figure B.3-1: Approach 
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B.3.2.1 Initial Technology Selection (Figure B.3-1, Steps 1-5) 

This section outlines the technology selection methodology that DOE will utilize 

to choose the optimal technology for radionuclide removal, with emphasis on 

HRRs, given the conditions at the time of evaluation. This process requires an 

initial systematic, documented evaluation of the options available, with a 

quantitative analysis arriving at the optimal technology choice. The specific 

methodology that DOE utilizes for choosing a removal technology and the 

formality of the associated documentation can vary (see Section B.3.2) 

depending on the waste properties, waste location and timing of the waste 

removal, but will include the following activities, which are described in 

subsequent sub-sections: 

1. Clear description of the structure, or structures, being evaluated for 

closure (i.e., waste tanks or ancillary structure) (see Section B.3.2.1.1) 

2. Characterization of waste remaining in the structure (see Section 

B.3.2.1.2) 

3. Characterization of associated Liquid Waste System status and the 

impact to this overall system posed by waste removal actions for the 

structure (see Section B.3.2.1.3) 

4. Systematic evaluation of removal technologies and selection of best 

available technology (see Section B.3.2.1.4) 

5. Assessment of whether to perform additional removal utilizing the 

selected technology (see Section B.3.2.1.5) 

B.3.2.1.1 Description of Structure to be Evaluated (Figure B.3-1, Step 1) 

The components associated with the specific structure (or structures) that are 

to be evaluated for radionuclide removal and eventual stabilization pending closure will be described. 

These structures could include either a waste tank or ancillary structure. In all cases, waste tank refers to 

the entire structure including both primary tank and annulus.143 The description of the structure could 

include assembly of engineering drawings, schematics, maps of obstructions to cleaning efforts, lists of 

components, operating status and location of integral equipment, boundaries between the tank or 

structure and tank farm system, photographs, and other pertinent documentation. 

B.3.2.1.2 Waste Characterization (Figure B.3-1, Step 2) 

With a clear description of the structure that is undergoing waste removal, DOE will prepare a 

characterization of the waste to be removed. The purpose of this characterization is to provide the 

baseline of radionuclide content, with an emphasis on HRRs, to ensure selection of optimal waste 

removal technologies. This characterization will take into consideration the specific HRRs within the tank 

or ancillary structure and may include: 

 Photographs and video of waste inside the waste tank or ancillary structure 

 Sample results from internal surface area and mounds 

 Estimated volumes 

 Historical information related to radionuclides present and associated concentrations 

 Analysis from similar waste tanks or ancillary structures 

 Physical properties (e.g., density, rheology, viscosity, particle size, yield strength) 

 Waste composition (e.g., HRRs present, non-radioactive materials present, concentrations) 

 Maps of waste locations within a waste tank or structure (e.g., layers, mounds) 

 Documentation of past removal activities and their relative success 

 Worker dose records for sampling activities 

 Waste removal history (e.g., relative success of cleaning efforts, progress photographs, 

equipment used) 

 Cooling coil sample analysis 

 

143 An exception to this is the Type IV tanks, which do not have an annulus. 
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 Annulus condition (e.g., leak history, photographs) 

 Leak detection system historical data 

Information on waste characterization will be collected from applicable lab analysis reports, drawings, 

mapping diagrams and other pertinent reports. This evaluation may be documented as a stand-alone 

document or may be included with other documentation within the Initial Technology Selection phase. 

B.3.2.1.3 Liquid Waste System Status Characterization (Figure B.3-1, Step 3) 

In addition to the characterization of the waste inside the structure, DOE will characterize the Liquid 

Waste System status associated with the further waste removal activities for the structure. The purpose 

of the system characterization is to understand and document the status and condition of the Liquid 

Waste System, in total, at the time of the proposed waste removal evolutions. These considerations of 

status, at a minimum, will include: 

 Available tank storage space capacity for applicable waste tanks required to support waste 

removal efforts 

 Compatibility of potential waste, waste removal streams, or agents added to the waste to aid in 

waste removal (e.g., oxalic acid) with other waste stored 

 Downstream processing impacts (e.g., impact of oxalates, waste volumes) 

 Status of salt waste and sludge batch processing and preparation 

 Impact on future waste removal activities in waste receipt tanks 

 Available equipment (e.g., explanation of resources used or not used) 

This status determination will either document these considerations or, in cases of similar time and 

circumstances, refer to previous documentation that remains valid for the current configuration and 

systems being evaluated. This evaluation may be documented as a stand-alone document or may be 

included with other documentation within the Initial Technology Selection phase. 

B.3.2.1.4 Radionuclide Removal Technology Evaluation (Figure B.3-1, Step 4) 

A technology selection evaluation will be performed based on the structure undergoing waste removal, 

the initial waste characterization, and the system characterization outlined in Sections B.3.2.1.1, B.3.2.1.2 

and B.3.2.1.3 respectively. This evaluation will focus on removal of radionuclides, with emphasis on 

HRRs, as well as other closure considerations related to DOE Manual 435.1-1 and other requirements. 

The formality of the associated documentation of the evaluation can vary (see Section B.3.2) depending 

on the waste properties, waste location and timing of the removal activities. 

When performing a technology evaluation, DOE will take one of two paths: rely on a previously performed 

evaluation where conditions for a waste tank or ancillary structure are similar to previously completed 

waste removal evolutions; or initiate a new technology evaluation, particularly where the tank structure or 

wastes are different than previously evaluated tanks. The “Alternative Studies” method is an example of 

a technology selection process that has been used successfully at the SRS. [WSRC-IM-98-00033] The 

“Alternatives Studies” method uses a formal analysis based on a set of weighted decision criteria. The 

depth of detail required by the technology selection will depend on specific conditions associated with the 

waste tank or ancillary structure under evaluation. A sensitivity analysis may be included in the analysis 

to aid in proper selection of a preferred technology. The technology selection process generally follows 

the “Alternative Studies” methodology and typically includes the assembly of a small but diverse group of 

knowledgeable individuals with experience that is relevant to the evaluation, and typically includes 

activities similar to the following steps: 

1. Identification of the communities of practice to be surveyed for viable technologies – In addition to 

the removal technologies that have previously been used at SRS, technologies from other DOE 

sites, DOE-sponsored technical exchanges, the industrial sector, the international sector and 

other relevant organizations may also be considered. 

2. Identification of removal technologies – A wide range of current technologies will be considered, 

at a minimum, including sluicing, mixing, chemical cleaning, vacuum retrieval techniques, 

mechanical manipulators and robotic vehicles. Any relevant future developments in removal 

technologies will also be considered at the time. The DOE will consider targeted HRR-specific 
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removal technologies as well as overall volume reduction technologies. Additionally, pertinent 

combinations of removal technologies will be taken into account. 

3. Identification of criteria that will be used to compare the various removal technologies – Criteria 

will include, at a minimum, the technologies’ expected radionuclide removal capability, with 

emphasis on HRRs, likelihood to meet the desired results effectively, costs, technical maturity, 

technical complexity and reusability. Furthermore, some examples of costs that will be 

considered are dose to workers, dose to public, financial costs, system-wide impacts (e.g., effects 

on downstream systems, generation of secondary waste streams), impacts to DOE’s mission and 

schedule and radiological control requirements. 

4. Evaluation of technologies against the selected criteria – Each technology is evaluated against 

each criterion and will be assigned a comparative ranking. 

5. Selection of a preferred technology – A scoring methodology will be used to select the optimum 

technology from among the set of technologies. 

The anticipated result of this process will be the identification of the optimal technology, or technologies, 

to remove radionuclides, with emphasis on HRRs, from the defined structure accounting for the specific 

characterization of the waste, surrounding Liquid Waste System status, schedule, and current 

technological maturity at the time the evaluation is performed. 

B.3.2.1.5 Assessment of Additional Removal (Figure B.3-1, Step 5) 

The activities described in Section B.3.2.1.4, will identify which available technology, or technologies, is 

the most viable option for additional radionuclide removal. The progression advances to the 

implementation of the technology as discussed in Section B.3.2.2 (Figure B.3-1, Steps 6 and 7). 

However, if it is not obvious that the technology can be and should be implemented to continue waste 

removal efforts (beyond bulk waste removal efforts previously completed), data for a cost-benefit analysis 

should be collected. The types of data supporting a cost-benefit analysis are described in Section B.3.2.6 

(Figure B.3-1, Steps 19-22). If, during the course of collecting this information, it appears probable that 

implementation and execution of any additional waste removal technology (beyond bulk waste removal 

efforts previously completed) is not practical, then a qualitative analysis will be performed and 

documented. As required by the HTF GCP (SRR-CWDA-2011-00022), DOE will review this information 

with the SCDHEC and the EPA and, if the three agencies (DOE, SCDHEC, EPA) concur, DOE would 

suspend waste removal activities and move into final sampling and analysis (Figure B.3-1, Step 19). 

The Technology Evaluation and Assessment steps will be documented or, if the current technology has 

not changed significantly, referenced to a previous report. For example, when two similar tanks (in 

construction and waste type) are undergoing similar waste removal processes at the same time, it is not 

necessary to undertake a selection and determination process for the second tank if the assumptions and 

parameters of the first still apply and no new technology has become available. This decision, however, 

will be documented. As discussed in Section B.3.2, the formality of the associated documentation can 

vary depending on the waste properties, waste location and timing of the removal activities. 

Examples of documentation/information that support this step are: 

 Operational history 

 Selection, operational performance and effectiveness of cleaning technologies used during each 

cleaning phase 

 Rationale for suspending use of each cleaning technology 

 Effectiveness in removing overall waste volume 

 Cost-benefit analysis of continuing waste removal efforts 

B.3.2.2 Technology Implementation (Figure B.3-1, Steps 6-7) 

An Operating Plan will be developed on how best to implement the selected technology safely for the 

particular waste tank or ancillary structure. An Operating Plan is a document that describes the cleaning 

process to be implemented, the methods of implementation, identification of anticipated end states and 

identification of specific metrics that ideally provide real-time indication of effectiveness. 
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These tailored metrics, which are necessary to track progress in waste removal evolutions, will be defined 

in the Operating Plan. Such metrics are dependent on the technology being implemented and structure 

undergoing waste removal but are expected to include such things as: 

 Monitoring radiation levels on 

transfer line 

 Waste removal equipment 

operating parameters (e.g., 

current drawn by a mixer 

pump, transfer rates) 

 Monitoring density readings 

for a solution 

 Monitoring solids 

concentration being removed 

 Waste volume reduction achieved by comparing pictures, video and mapping results 

 Effective cleaning radius of mixing devices 

The Operating Plan will reference estimated end states and metrics, and detail how the data will be 

obtained. The Operating Plan will also reflect any planned chemical cleaning flow sheets, include the 

projected mixing strategy (e.g., hours of operation, orientation, mode of operation, liquid level, mixer 

speed), and incorporate lessons learned from earlier waste removal efforts. If modifications to the 

equipment operation and/or the Operating Plan can result in greater technology effectiveness, then DOE 

will revise the Operating Plan to reflect this advantage. 

B.3.2.3 Technology Execution (Figure B.3-1, Steps 8-11) 

DOE will execute the technology until it is no longer considered an effective 

means of radionuclide removal, with an emphasis on HRR removal. 

Effectiveness will be assessed based on the technical data (i.e., metrics) 

outlined in the Operating Plan and captured throughout the execution of the 

removal technology. 

Data collection during this phase is expected to include such things as: 

 Photographs of any tank modification required for waste removal 

installation 

 Video mapping and high quality digital still photographs (before and 

after cleaning photographs from the same location) inside the primary 

tank and annulus 

 High quality photographs of obstructions to mixing 

 Records of daily operational decisions and their underlying reasons 

(e.g., logbooks, memos) 

 Solids volume reduction for each cleaning phase 

 Process sample analysis results 

 Volume and radionuclide concentration reductions 

 Weight percent solids in slurry 

 Tank temperature and pH 

 Impact of high efficiency particulate air (HEPA) filter loading 

 Liquid additions to the system 

 Secondary waste generation 

 Activities to address equipment issues 

 Costs of modifications, installation and operation 

 Mixer and transfer pump amps 

 Effective cleaning radius 

 Transfer line radiation dose rate data 

 Worker dose data 

 
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 Historical timeline of events 

 Documentation of each cleaning phase and reasons for proceeding to the next phase 

This data will be analyzed on an ongoing basis and used to determine whether the technology has 

reached the point of diminished effectiveness. Actual results will be compared with the expected results 

to support the evaluation of effectiveness. If the technology continues to be effective (Figure B.3-1, Step 

11), then DOE will continue to execute the technology (Figure B.3-1, Steps 8-10). If the technology is 

determined to be at the point of diminished effectiveness, the technology will be evaluated to determine 

whether to further deploy this technology. 

B.3.2.4 Technology Effectiveness Evaluation (Figure B.3-1, Steps 12-14) 

If a technology is no longer effective, the reason must be diagnosed and recorded. Examples for 

diminished effectiveness include: 

 Technology limitation (i.e., the inability of the current configuration to clean any further due to 

physical limitations of equipment) 

 Deterioration or failure of the equipment utilized by the technology 

 An outside factor that decrease effectiveness 

The diagnosed reason for 

diminished effectiveness 

determines what assessment will 

be done to establish whether to 

stop further execution of this 

technology or to modify the 

system or system parameters and 

continue execution (Figure B.3-1, 

Step 12). If the technology is no 

longer yielding effective results 

due to a technological limitation, 

DOE will assess whether it is 

practical to optimize the existing system to increase effectiveness. Optimization could include such things 

as adjusting pump indexing, altering flow rates, changing cleaning patterns or changing the concentration 

of a cleaning agent such as oxalic acid. Major modifications of equipment, such as identification and 

installation of an alternative transfer or mixing pump, are not considered optimization of the existing 

system. 

If effectiveness is reduced due to deterioration or failure of equipment, DOE will evaluate repairing or 

replacing the component. Likewise, if the technology is no longer effective due to an outside factor such 

as a constraint in the Liquid Waste System beyond the structure undergoing waste removal (see Section 

B.3.2.1.3), DOE will evaluate whether or not resolving that factor would be a means of increasing 

effectiveness. 

If DOE is making optimization adjustments, repairing/replacing equipment, or resolving an outside factor, 

(Figure B.3-1, Step 13), the appropriate changes in the Operating Plan as described in Section B.3.2.2 

(Figure B.3-1, Step 7) will be made and DOE will continue to execute the removal technology. If effecting 

these changes is not believed to be practical based on sound engineering judgment and the knowledge 

gained during the initial technology selection process, DOE will document (Figure B.3-1, Step 14) that the 

implemented technology will no longer be used based on earlier documented metrics and move into the 

next phase of the progression (i.e., Additional Technology Evaluation, Figure B.3-1, Steps 15-18). 

Although a formal cost-benefit analysis will not be performed at this stage, the underlying principles of 

such an evaluation will be included in the documentation supporting this decision. 

B.3.2.5 Additional Technology Evaluation (Figure B.3-1, Steps 15-18) 

This section outlines the technology evaluation methodology that DOE will employ to determine whether it 

is practical to continue removal operations with an additional technology. The specific methodology can 

vary depending on the waste properties, waste location and timing of the removal at this stage, and will 

include the following considerations: 
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 Characterization of remaining residuals to be removed 

(see Section B.3.2.5.1) 

 Characterization of potential impact to the Liquid Waste 

System during the scheduled time required for the 

evolution (see Section B.3.2.5.2) 

 Evaluation of alternative radionuclide removal 

technologies and selection of best available option (see 

Section B.3.2.5.3) 

 Assessment of whether to perform additional removal 

utilizing the selected technology (see Section B.3.2.5.4) 

The process is described in the following subsections. 

B.3.2.5.1 Residual Characterization (Figure B.3-1, Step 15) 

Because the removal operations may have altered the waste 

form, previously excluded alternative technologies may be viable 

at this point. The DOE will use the waste characterization 

methodology discussed in Section B.3.2.1.2 (Figure B.3-1, Step 

2) to re-evaluate the remaining waste and build a basis for the 

subsequent alternative selection. Once again, DOE will, at a 

minimum, consider the quantity, physical properties, composition, 

location of residual waste and the success of past removal 

activities. It may be necessary to collect actual samples of the 

residual material (typically referred to as “process samples”) and 

perform some sort of limited analysis suite to determine key 

characteristics. 

B.3.2.5.2 Liquid Waste System Status Characterization (Figure B.3-1, Step 16) 

Changes to the Liquid Waste System status could have occurred since earlier evaluations. DOE will reevaluate the status of the Liquid Waste System using the methodology discussed in Section B.3.2.1.3 

(Figure B.3-1, Step 3) to consider any changes that might affect the subsequent technology selection. 

These characterizations will consider the information collected in the previous characterization steps as 

well as the operational data collected in previous technology operation steps, discussed in Section 

B.3.2.3 (Figure B.3-1, Steps 8-11). At a minimum, DOE will consider waste tank storage space capacity, 

compatibility of waste, downstream processing impacts and the impacts on other risk-reducing evolutions 

within the Liquid Waste System. 

B.3.2.5.3 Additional Radionuclide Removal Technology Evaluation (Figure B.3-1, Step 17) 

DOE will perform an additional technology selection evaluation utilizing the residual and Liquid Waste 

System evaluations outlined in Section B.3.2.5.1 and Section B.3.2.5.2 (Figure B.3-1, Steps 15 and 16). 

This analysis will review available waste removal technologies to determine if a viable technology could 

be practically implemented to remove additional quantities of radionuclides, with emphasis on HRRs. The 

alternative removal technology selection methodology will resemble the methodology discussed in 

Section B.3.2.1 (Figure B.3-1, Steps 1-5). If the residual waste has not changed greatly from the previous 

characterization and the Liquid Waste System status characterization is similar, previous technology 

selection data may be used to inform the present technology selection. Technological advances since 

the previous technology selection will be considered. 

This methodology uses a structured approach for the identification and comparison of viable technologies 

to determine the most practical option for additional radionuclide removal, with an emphasis on HRR 

removal. This will include activities similar to those outlined in Section B.3.2.1.4 (Figure B.3-1, Step 4). 

The level of detail and formality will align with the extent the waste and the Liquid Waste System as a 

whole changed during the previous removal operations. The result of this selection methodology will be 

the identification of the best available alternative technology, or technologies, that could potentially be 

deployed to remove additional radionuclides, with emphasis on HRRs, from the defined structure 

accounting for current conditions. 
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B.3.2.5.4 Assessment of Additional Removal (Figure B.3-1, Step 18) 

As DOE evaluates the actions necessary to implement the potential new technology, or technologies, a 

qualitative evaluation will be performed to assess its implementation. This evaluation will consider: Liquid 

Waste System constraints; ratio of implementation costs per gallon of waste potentially removed or total 

curies, with emphasis on HRRs, potentially removed; potential worker exposure for technology 

installation; and execution. At this stage, if it is not obvious that the selected technology can be or should 

be implemented to continue waste removal efforts, data for a cost-benefit analysis should be collected. 

The types of data supporting a cost-benefit analysis are described in Section B.3.2.6 (Figure B.3-1, Steps 

19-22). If, during the course of collecting this information, it appears probable that implementation and 

execution of the technology is not practical, then a qualitative analysis will be performed and documented. 

As required by the HTF GCP (SRR-CWDA-2011-00022), DOE will review this information with SCDHEC 

and the EPA. If the three agencies (DOE, SCDHEC, EPA) concur, DOE will suspend waste removal and 

move into final sampling and analysis (Figure B.3-1, Step 19). 

B.3.2.6 Final Documentation of Radionuclide Removal (Figure B.3-1, Steps 19-22) 

The DOE will proceed 

to the sampling and 

analysis stage of the 

waste tank system 

operational closure 

process (Figure B.3-

1, Step 19) and 

perform a final 

characterization of 

the residuals with 

emphasis on the curies and locations of remaining HRRs. To support waste tank residual 

characterization, DOE will develop and document a sampling plan that minimizes uncertainty through 

representative sampling of the residuals. In some cases, process knowledge and historical sampling may 

be used to support final characterization of residuals. The process knowledge and historical sampling will 

be properly referenced. If process knowledge is used as a basis to support final characterization, the 

specific basis for the process knowledge will be identified and documented. Final characterization 

includes a volume determination as well as radionuclide concentrations. This information is used to 

develop a final radionuclide inventory. 

A cost-benefit analysis will be performed, informed, in part, by the qualitative dose impact results and 

conclusions of the associated performance assessment with the final radionuclide inventory 

considered.144 In this analysis, cost examples may include financial costs, increased risks to workers and 

members of the public, generation of secondary waste streams, schedule delays and associated impacts 

on other risk reduction activities, and downstream Liquid Waste System impacts. Typically, the costbenefit analyses will be relatively simple and will focus on the financial costs for implementation of new 

technologies versus the decrease in potential future doses resulting from the closure actions. [NUREG1854] If the development of the characterization data or the cost-benefit analysis demonstrates that it 

may be practical to remove additional radionuclides, with emphasis on HRRs, then additional removal 

technologies, or optimization of existing technologies, will be evaluated for possible additional removal. 

When final residual characterization and the cost-benefit analysis are complete, a removal report (Figure 

B.3-1, Step 22) will be prepared. This documentation will detail the design, construction and operational 

service histories of each tank. It will document all waste removal activities, including bases and 

 

144 To support Tier 2 Closure Authorization at the completion of waste removal activities for each waste tank and ancillary structure, 

DOE will prepare a special analysis utilizing the final residual characterization information. The special analysis process is

consistent with DOE Manual 435.1-1 and DOE Guide 435.1-1 and is a systematic process for determining the impact on the results 

and conclusions of a performance assessment when parameter values, such as radionuclide inventory, change. For each waste 

tank and ancillary structure, DOE will prepare a special analysis to evaluate the known actual inventories to date, for waste tanks 

and ancillary structures which have had final residual characterization completed, and projected future inventories, for waste tanks 

and ancillary structures yet to be cleaned, to determine the impacts of the final residual characterization values on the conclusions 

of the HTF PA. The results of these special analyses will be used to support the cost-benefit analyses and subsequent removal 

reports generated for each of the waste tanks and ancillary structures. 
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justifications for proceeding from one phase to another. The removal report will also document the 

selection, operational performance and effectiveness of technology used and will include the 

effectiveness of removing overall waste volume and specific HRRs. The report will combine 

documentation from the entire process to provide the complete demonstration of radionuclide removal to 

support Tier 2 Closure Authorization. 

B.3.2.7 DOE Tier 2 Closure Authorization 

The Tier 2 closure documentation, which 

includes, among other documentation (see 

Section B.2.0), the removal report discussed 

in the preceding subsection, must be 

approved by the Manager for DOE-SR 

authorizing the cessation of waste removal 

activities and removal from service and 

stabilization of the waste tank or ancillary 

structure. 
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APPENDIX C: H-TANK FARM PERFORMANCE ASSESSMENT DOSE SUMMARY 

Appendix Purpose 

The purpose of this appendix is to provide a brief discussion on dose results provided in 

the HTF PA. 

Appendix Contents 

This Appendix provides a summary of doses provided in the HTF PA. 

Key Points 

 The Base Case (i.e., Case A) represents the waste tank system configuration modeling 

case within the HTF PA that represents the most probable and defensible estimate of 

expected conditions for the HTF closure system based on currently available 

information. 

 In addition to the Base Case, other modeling cases which reflect alternate waste tank 

configurations were also considered in the HTF PA. The alternate cases reflect 

different modeling assumptions with respect to key modeling parameters so as to 

allow evaluation of sensitivities and uncertainties associated with Base Case modeling 

assumptions. 

 Table C.1-1 provides a summary of doses, for the Base Case and Alternate Cases, 

provided in the HTF PA. 

 Table C.1-2 is a partial re-creation of Table 4-3 in the NRC TER. DOE has added 

corresponding DOE notes alongside the NRC notes provided in the TER. This table is a 

summary of doses provided in the HTF PA. 

 A summary of the alternate waste tank configurations as compared to the Base Case 

is provided in Table C.1-3. A detailed description of the alternate waste tank 

configurations is provided in Section 4.4.2 of the HTF PA. 

 For the purposes of demonstrating reasonable assurance that the performance 

objective at 10 CFR 61.41 will be met, DOE utilizes the peak all-pathways dose at or 

outside of the 100-meter buffer zone.145

 For the purposes of demonstrating reasonable assurance that the performance 

objective at 10 CFR 61.42 will be met, DOE considers whether the peak intruder dose 

is less than 500 mrem/yr. 

Table C.1-1: HTF PA Dose Summary Table 

Protection of the 

General Population 

10 CFR 61.41 Notes 

10 CFR 61.41 

Conclusion 

Performance 

objective of 

25 mrem/yr is met 

 DOE’s conclusion that there is reasonable assurance 

this performance objective is met is based on doses 

associated with the Base Case (Case A), taking into 

account other cases (i.e., alternate configurations) 

and additional sensitivity analyses. In addition, 

additional sensitivity cases (described in Section 5.6 

of the HTF PA) further support DOE’s conclusions. 

 

145 See Section 7.0 of this HTF 3116 Basis Document for additional discussion. 
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Table C.1-1: HTF PA Dose Summary Table (Continued) 

Protection of the 

General Population 

10 CFR 61.41 Notes 

Deterministic 

(Case A) 

0.3 mrem/yr Within 1,000 

years 

Deterministic modeling of Case A in the HTF PA 

results in a peak dose well below the performance 

objective within 1,000 years and 10,000 years after 

HTF closure. The Case A dose results peak at 

approximately year 91,000 with a dose equivalent to 

approximately 20 percent of the 620 mrem average 

annual dose currently received by an individual 

member of the United States population. [NCRP160] 

4.0 mrem/yr Within 10,000 

years 

120 mrem/yr Peak dose 

within 100,000 

years 

Deterministic 

(Alternative or 

Sensitivity 

Cases)

12, 2.1 mrem/yr Case B, C 

(Within 1,000 

years) 

Deterministic modeling of Case B and Case C shows 

that the dose results remain below the performance 

objective within 1,000 years and 10,000 years after 

HTF closure even assuming alternative tank failure 

scenarios. 14, 16 mrem/yr Case B, C 

(Within 10,000 

years) 

12 mrem/yr Case D (Within 

1,000 years) 

Deterministic modeling of Case D shows that dose 

results remain below the performance objective within 

1,000 years and 10,000 years after HTF closure even 

assuming a fast flow path exists though the entire 

closed system for each waste tank. The fast flow 

configuration (Case D, described in Section 4.4.2.4 of 

the HTF PA) sensitivity analysis results were 

presented in Section 5.6.7.4 of the HTF PA to assess 

the impact of input variability on the groundwater 

pathways. 

18 mrem/yr Case D (Within 

10,000 years) 

3.7 mrem/yr Case E (Within 

1,000 years) 

Deterministic modeling of Case E shows that dose 

results remain below the performance objective within 

1,000 years after HTF closure even assuming 

multiple “failed” barriers. The Case E results are 

higher within 10,000 years after HTF closure due to 

Case E simultaneously and non-mechanistically 

simulating multiple “failed” barriers (e.g., a fast flow 

path exists though the entire closed system and the 

tank grout imparts no reducing capacity upon the tank 

waste). The Case E results within 10,000 years are 

less than the 620 mrem average annual dose 

currently received by an individual member of the 

United States population. [NCRP-160] 

240 mrem/yr Case E (Within 

10,000 years) 

0.7 mrem/yr No Closure 

Cap (Within 

1,000 years) 

Deterministic modeling of the Case A waste tank 

configuration shows that dose results remain within 

the performance objective for 1,000 years and 10,000 

years after HTF closure even assuming no closure 

cap is placed over the HTF. 4.7 mrem/yr No Closure 

Cap (Within 

10,000 years) 
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Table C.1-1: HTF PA Dose Summary Table (Continued) 

Protection of the 

General Population 

10 CFR 61.41 Notes 

Deterministic 

(Alternative or 

Sensitivity 

Cases) 

(Continued)

2.7 mrem/yr Synergistic 

Case (Within 

1,000 years) 

This non-mechanistic synergistic sensitivity analysis 

(Case F) is presented to address uncertainty related 

to three Base Case key modeling parameters. The 

three parameters analyzed further are gas transport 

impacts on reducing grout, liner failure times, and 

solubility controlling phases. The synergistic case 

evaluates the combined results of altering these three 

key modeling parameters. The relatively low 

resultant peak doses despite the synergistic case 

reflecting multiple modified assumptions further 

provides reasonable assurance that the performance 

objectives would be met. 

5.7 mrem/yr Synergistic 

Case (Within 

10,000 years) 

13 mrem/yr Synergistic 

Case (Within 

20,000 years) 

Probabilistic 

Modeling 

(peak of mean) 

Dose Statistics

 The “peak of the mean” analysis is based on the time 

at which the average annual dose (averaged over all 

samples) is maximum. The Uncertainty Analysis 

results are described in Section 5.6.4 of the HTF PA. 

The HTF probabilistic model is not intended to predict 

future potential doses, rather the goal is to 

characterize the context of uncertainty and sensitivity 

surrounding the PA calculations to further inform 

closure discussions. 

Probabilistic 

(peak of mean)

13 mrem/yr Case A (Within 

10,000 years) 

Probabilistic modeling of Case A results in a peak of 

the mean dose below the performance objective 

within 10,000 years after HTF closure. The fact that 

the peak of the means dose is higher than the 

deterministic peak dose is not unexpected, since 

many of the stochastic distributions used in the 

probabilistic modeling are, by design, reasonably 

conservative, driving the peak of the means higher. 

35 mrem/yr Case D (Within 

10,000 years) 

Probabilistic modeling of Case D results in a peak of 

the mean dose slightly above the performance 

objective within 10,000 years after HTF closure. In 

the Case D probabilistic modeling, a fast flow path is 

assumed to exist through the entire closed system for 

each waste tank. These modeling runs are intended 

to provide insight into the sensitivity of the HTF model 

to specific assumptions. The relatively low resultant 

peak doses for an alternative waste tank 

configuration with failed barriers provides further 

reasonable assurance that the performance objective 

would be met. The maximum results are less than 

the 620 mrem average annual dose currently 

received by an individual member of the United 

States population. [NCRP-160] 

 
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Table C.1-1: HTF PA Dose Summary Table (Continued) 

Protection of the 

General Population 

10 CFR 61.41 Notes 

Probabilistic 

(peak of mean) 

(Continued)

15 mrem/yr All Cases 

(Within 10,000 

years) 

The peak of the mean dose for all cases is below the 

performance objective within 10,000 years after HTF 

closure. While the probabilistic modeling that 

encompasses all of the alternative waste tank 

configurations results in a peak of the mean dose 

above the performance objective well beyond 10,000 

years, the resultant peak dose (at approximately 

67,000 years) is approximately an order magnitude 

higher than the performance objective value. The 

impact of individual modeling realizations can have a 

more pronounced impact on the peak of the mean 

dose as the modeling time period expands. The 

maximum results are less than the 620 mrem 

average annual dose currently received by an 

individual member of the United States population. 

[NCRP-160] 

205 mrem/yr All Cases 

(Within 

100,000 years) 

Protection of the 

Inadvertent 

Intruder 

10 CFR 61.42 Notes 

10 CFR 61.42 

Conclusion 

Performance 

objective of 

500 mrem/yr is met 

 DOE’s conclusion that there is reasonable assurance 

this performance objective is met is based on doses 

associated with the Base Case (Case A), taking into 

account other cases (i.e., alternate configurations) 

and additional sensitivity analyses. In addition, 

additional sensitivity cases (described in Section 5.6 

of the HTF PA) further support DOE’s conclusion. 

Deterministic 

(Case A) 

1.3 mrem Acute Dose 

(Within 1,000 

and 10,000 

years) 

Deterministic modeling of the Case A Intruder Dose 

in the HTF PA results in a peak acute dose well 

below a 500 mrem/yr dose within 1,000 years and 

10,000 years after HTF closure. 

40 mrem/yr Chronic Dose 

(Within 1,000 

years) 

Deterministic modeling of the Case A Intruder Dose 

in the HTF PA results in a peak chronic dose well 

below a 500 mrem/yr dose within 1,000, 10,000 and 

20,000 years after HTF closure. 50 mrem/yr Chronic Dose 

(Within 10,000 

years) 

260 mrem/yr Chronic Dose 

(Within 20,000 

years) 

Probabilistic 

(Case A peak 

of mean) 

760 mrem/yr Within 10,000 

years 

The peak of the mean dose of 760 mrem/yr is higher 

than the median peak dose of 500 mrem/yr, and 

illustrates that the mean is greater than the median 

indicating that a few realizations sampled at the tail of 

the parameter distributions can cause the mean to be 

high. As explained in Section 6.5.2 of the HTF PA, 

the GoldSim model intruder results were conservative 

(e.g., about three times higher than the deterministic 

PORFLOW results) due to differences in the models. 
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Table C.1-2: NRC TER Dose Summary Table 

Protection of the 

General Population 

10 CFR 61.41146 NRC Notes147 DOE Notes 

Compliance 

Conclusion 

Performance 

objective 

[25 mrem/yr] can be 

met within 

10,000 year 

compliance period. 

PA results show 

doses are expected 

to be significantly 

above performance 

objective at some 

uncertain time in the 

future. 

As noted by the NRC, the performance 

objective (less than 25 mrem/yr) can be met for 

a period of 10,000 years, ten times the period of 

compliance required by DOE requirements. 

Deterministic 

(Case A) 

4.0 mrem/yr Peak dose at 100-m 

boundary line within 

10,000 years 

Although NRC states that the “PA results show 

results to be significantly above performance 

objective” at some point in time beyond 10,000 

years (likely around 90,000 years following HTF 

closure), this dose (124 mrem/yr) is still only 

20% of the 620 mrem average annual dose 

currently received by an individual member of 

the United States population. [NCRP-160] 

124 mrem/yr Peak dose at 100-m 

boundary line within 

100,000 years 

Deterministic 

(Case B) 

14 mrem/yr Peak dose at 100-m 

boundary line within 

10,000 years 

Deterministic modeling of Case B shows that 

the dose results remain below the performance 

objective through 10,000 years even assuming 

alternative tank failure scenarios and remains at 

or below 20% of the 620 mrem average annual 

dose currently received by an individual 

member of the United States population. 

[NCRP-160]. 

125 mrem/yr Peak dose at 100-m 

boundary line within 

100,000 years 

Deterministic 

(Case C) 

16 mrem/yr Peak dose at 100-m 

boundary line within 

10,000 years 

Deterministic modeling of Case C shows that 

the dose results remain below the performance 

objective through 10,000 years even assuming 

alternative tank failure scenarios and remains at 

or below 20% of the 620 mrem average annual 

dose currently received by an individual 

member of the United States population. 

[NCRP-160]. 

124 mrem/yr Peak dose at 100-m 

boundary line within 

100,000 years 

Deterministic 

(Case D) 

18 mrem/yr Peak dose at 100-m 

boundary line within 

10,000 years 

Deterministic modeling of Case D shows that 

the dose results remain below the performance 

objective through 10,000 years even assuming 

a fast flow path exists through the entire closed 

system for each tank, and remains at or below 

20% of the 620 mrem average annual dose 

currently received by an individual member of 

the United States population. [NCRP-160]. 

123 mrem/yr Peak dose at 100-m 

boundary line within 

100,000 years 

Deterministic 

(Case E) 

239 mrem/yr Peak dose at 100-m 

boundary line within 

10,000 years 

Deterministic modeling of Case E is a sensitivity 

case that demonstrates the impact of failing 

multiple barriers on all waste tanks 

simultaneously. Even using these unrealistic 

assumptions, the resultant peak in 1,000 years 

is less than 5 mrem/yr and the peak dose in 

100,000 years is approximately 40% of the 620 

mrem average annual dose currently received 

by an individual member of the United States 

population. [NCRP-160]. 

239 mrem/yr Peak dose at 100-m 

boundary line within 

100,000 years 

 

 

146 Values/information from Table 4-3 of the NRC TER. [ML14094A496] 147 NRC notes from Table 4-3 of the NRC TER. [ML14094A496] 
Basis for Section 3116 Determination DOE/SRS-WD-2014-001 

for Closure of H-Tank Farm Revision 0 

at the Savannah River Site December 2014 

Page C-6 

Table C.1-2: NRC TER Dose Summary Table (Continued) 

Protection of the 

General Population 

10 CFR 61.41148 NRC Notes149 DOE Notes 

Probabilistic 

(Case A)

13 mrem/yr Within 10,000 years 

peak of the mean 

The “peak of the mean” analysis is based on 

the time at which the average annual dose 

(averaged over all probabilistic sampling cases) 

is maximum. As seen, even using this analysis 

technique, the probabilistic modeling of Case A 

and when all the Cases are considered (All 

Cases – Cases A, B, C, D and E) results in a 

peak of the mean doses below the performance 

objective within 10,000 years. The fact that the 

“mean of the peaks” are higher is expected 

since this is the average of all of the peaks seen 

from all realizations, and is skewed higher by a 

relatively small subset of results where 

essentially all barriers are at their most 

conservative state. All of these doses are 

below what an individual member of the United 

States population currently receives in a year 

(620 mrem). 

85 mrem/yr Within 10,000 years 

mean of the peak 

Probabilistic 

(Case D)

35 mrem/yr Within 10,000 years 

peak of the mean 

210 mrem/yr Within 10,000 years 

mean of the peak 

Probabilistic 

(All Cases)

15 mrem/yr Within 10,000 years 

peak of the mean 

220 mrem/yr Within 10,000 years 

mean of the peak 

205 mrem/yr Within 100,000 years 

peak of the mean 

530 mrem/yr Within 100,000 years 

mean of the peak 

Protection of the 

Inadvertent 

Intruder 

10 CFR 61.42150 NRC Notes151 DOE Notes 

Compliance 

Conclusion 

Performance 

objective [500 

mrem/yr] can be 

met. 

Robust compliance 

conclusion for drill 

cuttings. Uncertainty 

with dose from 

intruder well. 

As noted, the performance objective of 

500 mrem/yr is demonstrated for the 

inadvertent intruder. 

Deterministic 

(Case A, B, C, 

D) 

Chronic Exposure Peak dose at 1-m 

boundary line within 

10,000 years where 

intruder assumed to 

drill through: 

All doses are well below the 500 mrem/yr 

performance objective. 

51 mrem/yr  3-in transfer line 

74 mrem/yr  4-in transfer line 

100 mrem/yr  Waste Tank 13 

260 mrem/yr Peak dose at 1-m 

boundary line within 

20,000 yrs where 

intruder assumed to 

drill through a 3-in 

transfer line. 

 

148 Values/information from Table 4-3 of the NRC TER. [ML14094A496] 149 NRC notes from Table 4-3 of the NRC TER. [ML14094A496] 150 Values/information from Table 4-3 of the NRC TER. [ML14094A496] 151 NRC notes from Table 4-3 of the NRC TER. [ML14094A496] 
Basis for Section 3116 Determination DOE/SRS-WD-2014-001 

for Closure of H-Tank Farm Revision 0 

at the Savannah River Site December 2014 

Page C-7 

Table C.1-2: NRC TER Dose Summary Table (Continued) 

Protection of the 

Inadvertent 

Intruder 

10 CFR 61.42152 NRC Notes153 DOE Notes 

Deterministic 

(Case E) 

Acute Exposure Peak dose at 1-m 

boundary line within 

10,000 years where 

intruder assumed to 

drill through: 

All doses are well below the 500 mrem/year 

performance objective. 

1.4 mrem/yr  3-in transfer line 

2.6 mrem/yr  4-in transfer line 

14 mrem/yr  Waste Tank 13 

Probabilistic 

(peak of 

mean) 

Chronic Exposure Peak of the mean 

doses inside 1-m 

boundary at 

hypothetical drilling 

locations adjacent to 

specified waste tanks 

(adjacent to Tank 12): 

As discussed in Section 7.2.5 of this HTF 3116 

Basis document, the peak of the mean doses 

are higher than the median peak doses, which 

illustrates that the mean is greater than the 

median indicating that a few realizations 

sampled at the tail of the parameter 

distributions can cause the mean dose to be 

high. As explained in Section 6.5.2 of the HTF 

PA, the GoldSim model intruder results were 

conservative (e.g., about three times higher 

than the deterministic PORFLOW results) due 

to differences in the models. These dose 

results during the initial 10,000 years after HTF 

closure are comparable to the 620 mrem 

average annual dose currently received by an 

individual member of the United States 

population. [NCRP-160] During the initial 

20,000 years after HTF closure the peak of the 

mean dose (i.e., 961 mrem) remains below the 

average annual dose (i.e., above 1,000 mrem) 

received by a person living in Denver, Colorado. 

[NRC_01-01-2011] 

762 mrem/yr  Within 10,000 years 

961 mrem/yr  Within 20,000 years 

 

152 Values/information from Table 4-3 of the NRC TER. [ML14094A496] 153 NRC notes from Table 4-3 of the NRC TER. [ML14094A496] 
Basis for Section 3116 Determination 

for Closure of H-Tank Farm at the Savannah River Site 

Page C-8 

Table C.1-3: HTF Waste Tank Case Comparison Versus Base Case (Case A)

Alternate Cases 

B C D E 

No Closure Cap 

Barriers

Closure Cap NC NC NC NC No 

Closure Cap 

Waste Tank Top 

Tank Top Fast Flow 

Tank Top Fast Flow 

Tank Top Fast Flow 

Tank Top Fast Flow 

NC 

Waste Tank Liner 

Early liner degradation 

Early liner degradation 

Early liner degradation 

Early liner degradation 

NC 

Waste Tank Grout 

Early grout degradation 

NC Early grout 

degradation 

NC NC 

Contamination Zone Reducing Capacity 

NC Grout monolith 

reducing capacity impact minimized 

NC Grout monolith 

reducing capacity impact minimized 

NC 

Contamination Zone Solubility Limits 

NC NC NC NC NC 

Waste Tank Basemat NC NC Basemat 

Fast Flow 

Basemat Fast Flow 

NC 

Vadose Zone Beneath Waste Tanks 

NC NC NC NC NC 

NC - No Change from Base Case