DOE/SRS-WD-2010-001 

Revision 0 

Draft Basis for Section 3116 Determination 

for 

Closure of F-Tank Farm 

at the 

Savannah River Site 

September 30, 2010 
Draft Basis for Section 3116 Determination DOE/SRS-WD-2010-001 

for Closure of F-Tank Farm Revision 0 

at the Savannah River Site September 30, 2010 

Page ii 

REVISION SUMMARY

REV. # DESCRIPTION DATE OF ISSUE 

0 Initial Issue 09/30/2010 
Draft Basis for Section 3116 Determination DOE/SRS-WD-2010-001 

for Closure of F-Tank Farm Revision 0 

at the Savannah River Site September 30, 2010 

Page iii 

TABLE OF CONTENTS

Page 

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

LIST OF TABLES .........................................................................................................................................v

LIST OF FIGURES ......................................................................................................................................vi

APPENDICES............................................................................................................................................viii

ACRONYMS / ABREVIATIONS................................................................................................................viii

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

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

1.2 Purpose and Scope.......................................................................................................................1-2

1.3 Schedule and Plans for Closing Tanks.........................................................................................1-2

1.4 Outline of Draft FTF 3116 Basis Document..................................................................................1-3

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

2.1 Savannah River Site and F-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.................................................................................................................2-15

2.1.6 Geochemistry.................................................................................................................2-26

2.1.7 Natural Resources .........................................................................................................2-26

2.1.8 Natural and Background Radiation................................................................................2-28

2.1.9 Tank Farm Operations ...................................................................................................2-29

2.1.10 F-Tank Farm ..................................................................................................................2-30

2.1.11 Waste Tanks ..................................................................................................................2-31

2.1.12 Ancillary Structures ........................................................................................................2-41

2.2 F-Tank Farm Wastes ..................................................................................................................2-48

2.2.1 Sources of the Waste.....................................................................................................2-49

2.2.2 Waste Management.......................................................................................................2-50

2.3 Waste Removal Approach for Waste Tanks, Annuli and Ancillary Structures............................2-56

2.3.1 Tank Waste Removal History ........................................................................................2-57

2.3.2 Waste Removal Technologies .......................................................................................2-58

2.3.3 Chemical Cleaning.........................................................................................................2-59

2.3.4 Mantis - Type IV Waste Tanks 18 and 19......................................................................2-61

2.3.5 Removal of Residual Waste from Failed Cooling Coils .................................................2-63

2.3.6 Annulus Cleaning...........................................................................................................2-63

2.3.7 Potential Future Waste Tank Cleaning Technology ......................................................2-64

2.3.8 Ancillary Structures Cleaning.........................................................................................2-64

2.4 Radionuclide Inventory in F-Tank Farm Facility Systems, Structures and Components ...........2-65

2.4.1 Residual Inventory for Tank Annuli, Inside Failed Cooling Coils and Internal Tank 

Surfaces .........................................................................................................................2-69

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

Evaporator Systems.......................................................................................................2-70

2.5 Residual Waste Stabilization ......................................................................................................2-70

2.6 Closure Cap ................................................................................................................................2-72

2.6.1 Function of Closure Cap Layers ....................................................................................2-77
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TABLE OF CONTENTS (Continued)

Page 

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 ........................................................................5-2

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 ........................5-5

5.1.6 Highly Radioactive Radionuclides Based on Uncertainty and Sensitivity Analyses........5-6

5.1.7 Highly Radioactive Radionuclides Summary ...................................................................5-7

5.2 Removal of Highly Radioactive Radionuclides to the Maximum Extent Practical ........................5-9

5.2.1 Current Status of F-Tank Farm Waste Removal Activities ..............................................5-9

5.2.2 Waste Removal Technologies .......................................................................................5-10

5.2.3 Optimization of Existing Technologies...........................................................................5-16

5.3 Removal to the Maximum Extent Practical .................................................................................5-16

5.4 Conclusion ..................................................................................................................................5-19

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 F-Tank Farm Residuals ..................................................6-3

6.3 Methodology..................................................................................................................................6-3

6.3.1 Methodology Inputs..........................................................................................................6-4

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

6.4 Waste Concentration Calculations..............................................................................................6-10

6.4.1 Waste Tank Concentration Calculation..........................................................................6-10

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 Part 61, Subpart C Performance Objectives ...................................................................7-2

7.1.1 10 CFR 61.40...................................................................................................................7-2

7.1.2 F-Tank Farm Performance Assessment [SRS-REG-2007-00002]..................................7-2

7.1.3 10 CFR 61.41...................................................................................................................7-3

7.1.4 As Low As Reasonably Achievable .................................................................................7-6

7.1.5 10 CFR 61.42...................................................................................................................7-7

7.1.6 10 CFR 61.43.................................................................................................................7-10

7.1.7 10 CFR 61.44.................................................................................................................7-18
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TABLE OF CONTENTS (Continued)

Page 

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

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 Database ..........................................2-22

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

Year 2007).......................................................................................................................................2-27

Table 2.1-4: Water Quality in Upper Three Runs ...................................................................................2-27

Table 2.1-5: Summary of Maximum Groundwater Monitoring Results for Major Areas that Outcrop to 

Fourmile Branch, 2006–2007 ..........................................................................................................2-28

Table 2.1-6: Waste Tank Nominal Capacities, Nominal Dimensions and Other Features.....................2-31

Table 2.1-7: Type III/IIIA Waste Tanks Primary Tank and Secondary Liner Plate Data ........................2-36

Table 2.6-1: FTF Closure Cap Layers ....................................................................................................2-77

Table 2.6-2: Function of the FTF Closure Cap Layers ...........................................................................2-78

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

Table 5.1-2: Radionuclides Contained in FTF PA Initial Inventory ...........................................................5-7

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

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

Table 6.3-2: FTF Waste Tank Site-Specific Factors for 10 CFR 61.55 Table 2 Radionuclides ...............6-9

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

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

Table 6.4-1: Sum of the Fractions Calculation Using the Closure Inventory for Tank 18 

(10 CFR 61.55 Table 1 Radionuclides) ...........................................................................................6-11

Table 6.4-2: Sum of the Fractions Calculation Using the Closure Inventory for Tank 18 

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

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

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

Table 6.4-4: Sum of the Fractions Calculation Using the FTF PA, Revision 1 Inventory for a 3-inch 

Transfer Line (10 CFR 61.55 Table 2 Radionuclides).....................................................................6-13

Table 7.1-1: Crosswalk Between Applicable 10 CFR 20 Standards and DOE Requirements ...............7-11
Draft Basis for Section 3116 Determination DOE/SRS-WD-2010-001 

for Closure of F-Tank Farm Revision 0 

at the Savannah River Site September 30, 2010 

Page vi 

LIST OF FIGURES

Page 

Figure 2.1-1: Physical Location of Savannah River Site...........................................................................2-2

Figure 2.1-2: Location of Savannah River Site and Adjacent Areas ........................................................2-3

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

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

Figure 2.1-5: General Layout of FTF ........................................................................................................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 Geologic 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.....................................................................2-13

Figure 2.1-12: Seismic Events within a 50-Mile Radius of SRS .............................................................2-14

Figure 2.1-13: Regional Scale Faults for SRS and Vicinity ....................................................................2-15

Figure 2.1-14: Regional NW to SE Cross Section ..................................................................................2-16

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

in the SRS Region ...........................................................................................................................2-16

Figure 2.1-16: Potentiometric Surface of the Upper Three Runs Aquifer...............................................2-17

Figure 2.1-17: Potentiometric Surface of the Gordon Aquifer.................................................................2-18

Figure 2.1-18: Hydrostratigraphic Picks in GSAD Database ..................................................................2-19

Figure 2.1-19: Laboratory Determined Permeability Data in GSAD Database.......................................2-20

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

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

Figure 2.1-22: Slug Test Data in GSAD Database .................................................................................2-21

Figure 2.1-23: Conceptual Diagram of Groundwater Flow beneath the GSA ........................................2-22

Figure 2.1-24: Savannah River Basin Dams...........................................................................................2-23

Figure 2.1-25: SRS Watershed Boundaries and Major Tributaries ........................................................2-24

Figure 2.1-26: GSA Gauging Stations ....................................................................................................2-25

Figure 2.1-27: Major Sources of Radiation Exposure in the Vicinity of SRS ..........................................2-28

Figure 2.1-28: General Layout of FTF ....................................................................................................2-30

Figure 2.1-29: Typical Type I Waste Tank Cross Section ......................................................................2-32

Figure 2.1-30: Type I Waste Tank Basemat Construction......................................................................2-33

Figure 2.1-31: Type I Waste Tank Primary and Secondary Liner Construction .....................................2-33

Figure 2.1-32: Typical Type I Waste Tank Column Support...................................................................2-33

Figure 2.1-33: Typical Type I Waste Tank Cooling Coil Obstacles ........................................................2-34

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

Figure 2.1-35: Typical FTF Type III Waste Tank Cross Section.............................................................2-35

Figure 2.1-36: Typical Type IIIA Waste Tank Cross Section ..................................................................2-35
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LIST OF FIGURES (Continued)

Page 

Figure 2.1-37: Typical Type III/IIIA Waste Tank Basemat Construction..................................................2-35

Figure 2.1-38: Typical Type IIIA Waste Tank Primary Tank and Secondary Liner - Late Construction 

(Vault Wall Not Constructed)...........................................................................................................2-36

Figure 2.1-39: Typical Type III/IIIA Waste Tank Radial Air Grooves ......................................................2-37

Figure 2.1-40: Typical Type III Waste Tank Deployable Cooling Coils ..................................................2-37

Figure 2.1-41: Typical Type IIIA Waste Tank Cooling Coils During Construction ..................................2-38

Figure 2.1-42: Typical Type III Waste Tank Access Area for Waste Removal Equipment Diagram......2-38

Figure 2.1-43: Typical Type IIIA Waste Tank Access Area for Waste Removal Equipment Diagram ...2-38

Figure 2.1-44: Typical FTF Type IV Waste Tank Cross Section ............................................................2-39

Figure 2.1-45: Typical Type IV Waste Tank Stiffener Angles.................................................................2-39

Figure 2.1-46: Typical Type IV Waste Tank Wall Steel Bands During Construction..............................2-40

Figure 2.1-47: Typical Type IV Waste Tank Domes and Risers During Construction............................2-40

Figure 2.1-48: General Layout of FTF ....................................................................................................2-41

Figure 2.1-49: Typical Waste Tank Transfer Line During Construction..................................................2-42

Figure 2.1-50: Typical Diversion Box and Pump Pit Design...................................................................2-43

Figure 2.1-51: Typical Pump Tank Top View During Construction.........................................................2-43

Figure 2.1-52: Typical Catch Tank Cross Section ..................................................................................2-44

Figure 2.1-53: 242-F Evaporator System Schematic..............................................................................2-45

Figure 2.1-54: 242-16F Evaporator System Schematic..........................................................................2-46

Figure 2.2-1: Saltcake at Bottom of Waste Tank ....................................................................................2-49

Figure 2.3-1: Typical Standard Slurry Pump...........................................................................................2-57

Figure 2.3-2: Submersible Mixer Pump Waste Removal Diagram .........................................................2-58

Figure 2.3-3: Submersible Mixing Pump and Submersible Transfer Pump............................................2-59

Figure 2.3-4: Submersible Mixing Pump in a Test Tank.........................................................................2-59

Figure 2.3-5. ECC Process Diagram ......................................................................................................2-61

Figure 2.3-6: Mantis ................................................................................................................................2-62

Figure 2.3-7: Waste Deposits on Exterior Primary Tank Wall Before Cleaning (Tank 5).......................2-63

Figure 2.3-8: Waste Deposits on Annulus Floor Before Cleaning (Tank 6)............................................2-64

Figure 2.6-1: FTF Closure Cap General Concept...................................................................................2-72

Figure 2.6-2: FTF Closure Cap Conceptual Design, Cap Footprint .......................................................2-73

Figure 2.6-3: FTF Closure Cap Conceptual Design, Sections A-A, B-B, and C-C.................................2-74

Figure 2.6-4: FTF Closure Cap Conceptual Design, Sections D-D and E-E ..........................................2-75

Figure 2.6-5: FTF Closure Cap Conceptual Design, Finished Cap ........................................................2-76

Figure 2.6-6: FTF Closure Cap Conceptual Design, Grading Cap.........................................................2-77

Figure 5.2-1: FTF Waste Tank Composite Inventory..............................................................................5-10

Figure 5.2-2: Submersible Mixer Pump Waste Removal Diagram .........................................................5-11

Figure 5.2-3: Submersible Mixing Pump and Submersible Transfer Pump............................................5-12

Figure 5.2-4: Submersible Mixing Pump in a Test Tank.........................................................................5-12
Draft Basis for Section 3116 Determination DOE/SRS-WD-2010-001 

for Closure of F-Tank Farm Revision 0 

at the Savannah River Site September 30, 2010 

Page viii 

LIST OF FIGURES (Continued)

Page 

Figure 5.2-5 ECC Process Diagram ......................................................................................................5-14

Figure 5.2-6: Mantis ................................................................................................................................5-15

Figure 5.2-7: Robotic Sampler at Test Facility........................................................................................5-16

Figure 5.3-1: Tank 18 Waste Removal ...................................................................................................5-17

Figure 5.3-2: Tank 19 Waste Removal ...................................................................................................5-18

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

Figure 7.1-1: FTF Boundary and Buffer Zone...........................................................................................7-2

Figure 7.1-2: Scenario with Well Water as Primary Source......................................................................7-4

Figure 7.1-3: Acute Intruder-Drilling Scenario...........................................................................................7-7

Figure 7.1-4: Chronic Intruder-Agricultural (Post-Drilling) Scenario .........................................................7-8

APPENDICES

Page 

Appendix A Liquid Waste System Description A-1 

ACRONYMS / ABREVIATIONS

ADMP Advanced Design Mixer Pump

AEC Atomic Energy Commission 

ALARA As Low As Reasonably Achievable 

ARP Actinide Removal Process 

CFR Code of Federal Regulations 

Ci curie(s)

CLM Central Climatology 

cm centimeter(s) 

CSRA Central Savannah River Area 

CSSX Caustic Side Solvent Extraction

DDA Deliquification, Dissolution, and Adjustment

DOE United States Department of Energy

DSA Documented Safety Analysis 

DWPF Defense Waste Processing Facility

ECC Enhanced Chemical Cleaning 

Eh Measure of reduction (or oxidation) potential

EIS Environmental Impact Statement

EPA United States Environmental Protection Agency 

ETF Effluent Treatment Facility

FDB F-Tank Farm Diversion Box

FFA Federal Facility Agreement

FPP F-Tank Farm Pump Pit

FPT F-Tank Farm Pump Tank

FTF F-Tank Farm

FTF PA F-Tank Farm Performance Assessment
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ACRONYMS / ABREVIATIONS (Continued)

g gram(s) 

GCL Geosynthetic Clay Liner 

GCP General Closure Plan 

gpm gallons per minute

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 

HEU Highly Enriched Uranium

HM H-Modified

hr hour(s) 

HRR Highly Radioactive Radionuclide 

HTF H-Tank Farm

lb pound(s) 

ISWLF Industrial Solid Waste Landfill Facility 

Kd Distribution Coefficient 

LAW Low Activity Waste 

LDB Leak Detection Box 

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) 

NASA National Aeronautics and Space Administration 

NCRP National Council on Radiation Protection and Measurements 

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

NPDES National Pollutant Discharge Elimination System 

NRC United States Nuclear Regulatory Commission 

NRMP Natural Resources Management Plan 

OSHA Occupational Safety and Health Administration 

OUO Official Use Only 

PA Performance Assessment 

PEL Permissible Exposure Limit 

pH Measure of acidity or alkalinity of a solution 

psi pounds per square inch 

PUREX Plutonium Uranium Extraction 

RBA Radiological Buffer Area 

ROD Record of Decision 

rpm revolutions per minute 

RPP Radiation Protection Program 
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ACRONYMS / ABREVIATIONS (Continued)

ROI Region of Influence 

SA Special Analysis 

SCDHEC South Carolina Department of Health and Environmental Control 

SDF Saltstone Disposal Facility 

SMA Strong Motion Accelerometer 

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 

TNX Training and Experimental Test Facility 

UTR Upper Three Runs 

WAC Waste Acceptance Criteria 

WCS Waste Characterization System 

WMC Waste Mixing Chamber 

yr year(s) 
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Page 1-1 

1.0 INTRODUCTION AND PURPOSE 

Section Purpose 

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

Section 3116 Determination for Closure of F-Tank Farm at the Savannah River Site

(hereinafter referred to as: Draft FTF 3116 Basis Document). 

Section Contents 

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

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

Key Points 

 The Department of Energy (DOE) is issuing this Draft FTF 3116 Basis Document to 

provide a draft basis for a potential determination by the Secretary of Energy, in 

consultation with the 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 Draft FTF 3116 Basis Document applies to stabilized residuals in waste tanks1 and 

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

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

 The FTF is a 22-acre site consisting of underground radioactive waste storage tanks 

and supporting ancillary structures. 

 The FTF 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 FTF 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 FTF waste tanks will be stabilized by 

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

subsidence. 

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

take place after individual component cleaning is complete. 

 The Final FTF 3116 Basis Document will be issued by DOE following consultation with 

the NRC.

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 Draft FTF 3116 Basis Document shows that those criteria are satisfied, to 

support a potential determination by the Secretary pursuant Section 3116. This Draft FTF 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 FTF at the time of closure. 

The DOE is issuing this Draft FTF 3116 Basis Document for NRC consultative review, as part of DOE’s 

consultation with NRC. Although not required by NDAA Section 3116, DOE will issue this Draft FTF 3116 

Basis Document for public review and comment. This Draft FTF 3116 Basis Document will be finalized 

after DOE consults with the NRC and considers public comments. 

 

1

 The FTF has 22 waste tanks. Two of those waste tanks, Tanks 17 and 20, were cleaned and operationally closed in 1997, prior to 

enactment of NDAA Section 3116. Accordingly, Tanks 17 and 20 are not within the scope of this draft FTF 3116 Basis Document. 

Rather, the remaining 20 tanks in FTF (and their residuals) are addressed by this Draft FTF 3116 Basis Document, and, unless 

otherwise specified, references in this document to “the waste tanks” or “FTF waste tanks” refer to those remaining tanks. 

Nevertheless, the experience gained from Tanks 17 and 20 is discussed in certain sections of this document to provide insight and 

additional information. Also, the Performance Assessment for the F-Tank Farm at the Savannah River Site [SRS-REG-2007-

00002], cited in several sections of this Draft FTF 3116 Basis Document, takes into account radionuclide residuals from all FTF

waste tanks, including Tanks 17 and 20, for accuracy and completeness. 
Draft Basis for Section 3116 Determination DOE/SRS-WD-2010-001 

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at the Savannah River Site September 30, 2010 

Page 1-2 

1.2 Purpose and Scope 

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

stabilization activities are complete, the stabilized residuals in the FTF 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 Draft FTF 3116 Basis Document specifically addresses the stabilized residuals in the 

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

equipment) at the time of FTF closure2

. This Draft FTF 3116 Basis Document does not include other 

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

1.3 Schedule and Plans for Closing Tanks 

The FTF waste tanks3

 are closed in accordance with the SRS FFA, a formal agreement between DOE, 

Region 4 of the United States Environmental Protection Agency (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 (old style tanks, 

specifically the Type I and Type IV tanks in FTF) must be removed from service according to the FFA 

schedule. The current FFA calls for operational closure of Tanks 18 and 19 by December 2012, and 

staggered operational closure of the other eight FTF (Type I) waste tanks (tanks numbers not specified in 

the SRS FFA) by September 20224

. [WSRC-OS-94-42] DOE addresses the closure of the remaining 

FTF tanks (Type III and IIIA tanks) and ancillary structures in the SRS Liquid Waste System Plan5

. [SRRLWP-2009-00001] 

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 FTF 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 an SCDHEC industrial wastewater 

construction permit. [DHEC_03-03-1993] Removal from service and stabilization of the FTF waste tanks 

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

Closure Plan, which contains the overall plan for removing from service and stabilizing the FTF waste 

tanks and ancillary structures. [LWO-RIP-2009-00009] A specific Closure Module for each tank or 

ancillary structure or groupings of tanks and ancillary structures will be developed and submitted to the 

State of South Carolina for approval. Final tank stabilization activities shall not proceed until the State 

grants approval. Stabilization of individual FTF 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 

 

2

 For the purpose of this Draft FTF 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 FTF waste tanks will be carried out by filling 

the 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 FTF transfer lines. 

3

 The FTF contains four designs (types) of tanks, called Type I, Type IV, Type III and Type IIIA tanks. The FTF tanks are also 

numbered, although not sequentially. The tank types are discussed in further detail in Section 2.0 of this Draft FTF 3116 Basis 

Document. 

4

 The FFA includes procedures for revision of the FFA schedule. 

5

 Under the current SRS Liquid Waste System Plan, DOE anticipates closure of the remaining FTF tanks (Type III and IIIA tanks) by 

2029. The Liquid Waste System Plan is updated periodically as appropriate. 

According to the Performance Assessment for the F-Tank Farm at the Savannah River Site [SRS-REG-2007-00002], DOE’s 

planned sequence for closure of FTF is as follows: 

 Closure of Type IV tanks, the Type I 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 FTF 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 FTF waste tanks and ancillary structures, FTF will undergo final closure in accordance with the 

FFA.
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material. This method was chosen as the most preferred environmental alternative and the least 

hazardous for closure of the waste tanks and associated equipment. [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 free-standing liquids and void spaces in the 

waste tanks. 

1.4 Outline of Draft FTF 3116 Basis Document 

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

stabilized residuals within the FTF 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 Draft FTF 3116 Basis Document provides an overview of SRS and the FTF. In 

addition, extensive descriptions of the FTF and waste processing facilities are provided in the 

Performance Assessment for the F-Tank Farm at the Savannah River Site (hereinafter referred to as: FTF 

PA). [SRS-REG-2007-00002] Section 3.0 of this Draft FTF 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 for the Secretary of Energy, in consultation with the NRC, to make a determination that the 

NDAA Section 3116(a) criteria are met and, thus, the waste is not high-level waste. 
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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 FTF with descriptions of the different waste tank 

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

in FTF 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 FTF Closure Cap. 

Key Points 

 The FTF occupies 22 acres in the General Separations Area (GSA) near the center of 

the SRS. 

 The FTF contains 22 carbon steel waste tanks of three different basic designs, eight 

with a nominal capacity of 750,000 gallons per tank and fourteen with a nominal 

capacity of 1,300,000 gallons per tank. 

 Most of the waste in these waste tanks originated in the SRS F-Canyon Facility, which 

recovered nuclear material produced in the site’s nuclear production reactors. 

 Two of the FTF waste tanks, Tank 17 and Tank 20, have been operationally closed in 

place with the approval of the South Carolina Department of Health and 

Environmental Control prior to enactment of NDAA Section 3116, and are not within 

the scope of this Draft FTF 3116 Basis Document. 

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

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

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

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

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

assessed in the FTF PA. 

 After waste removal, FTF 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 F-Tank Farm Facility Overview 

This section provides brief descriptions of the site and the FTF 6. 

2.1.1 Geography and Demography 

2.1.1.1 SRS Site Description 

The SRS, one of the sites in the DOE complex, was constructed starting 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 198,344 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. [WSRC-STI-2008-

00057] 

 

6 Sections 1.0 and 2.0, as well as Appendix A, of this Draft FTF 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 Draft FTF 3116 Basis Document, as outside 

the scope of this Draft FTF 3116 Basis Document and not included as NDAA 3116 requirements or criteria for purposes of NRC 

consultation.
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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 is located 

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

Figure 2.1-1: Physical Location of Savannah River Site 

BAMBERG 

COUNTY 
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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, pg 1.4-11] 

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, R). 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 is the coal-fired powerhouse that provides steam to SRS. M Area and Training and Experimental 

Test Facility (TNX) have undergone Deactivation and Decommissioning. 

Figure 2.1-2: Location of Savannah River Site and Adjacent Areas 
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Figure 2.1-3: SRS Operational Area Location Map 
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2.1.1.2 Closure Site Description 

The FTF is in F Area which is 

located in the central region of SRS. 

Figure 2.1-4 presents the area 

known as the General Separations 

Area (GSA). The GSA is located 

atop a ridge running southwestnortheast that forms the drainage 

divide between Upper Three Runs 

(UTR) to the north and Fourmile 

Branch to the south. The GSA 

contains the F and H Area 

Separations Facilities, the S-Area 

Defense Waste Processing Facility 

(DWPF), the Z-Area Saltstone 

Facility and the E-Area Low-Level 

Waste Disposal Facilities. The FTF 

is a liquid waste storage facility 

consisting of 22 carbon steel waste 

tanks, shown in Figure 2.1-5, which 

store liquid radioactive waste 

generated primarily from the FCanyon PUREX process. Note that 

two tanks, 17 and 20, have been 

cleaned, removed from service and 

filled with grout. The FTF 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 U.S. Census Bureau 

data, the estimated 2008 population 

in the eight-county region of 

influence (ROI) was 550,675. 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 

84% of the ROI population live in 

the following three counties: Aiken 

(28%), Richmond (36.2%) and 

Columbia (20.1%). Only about 16% 

of the ROI population live in the 

remaining counties as shown in 

Table 2.1-1. [http://factfinder. 

census.gov] 

Figure 2.1-4: Layout of the GSA

Figure 2.1-5: General Layout of FTF 
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From 2000 to 2008 the population in the eight-county region grew an estimated 5.8%. Columbia County 

had the highest estimated growth at approximately 23.9% followed by Aiken County with an estimated 

growth of approximately 8.1% and Burke County 

with an estimated growth of 2.2%. Allendale, 

Bamberg, Barnwell and Screven counties 

experienced a net population loss. Calculations 

are based on information obtained from the U. S. 

Census Bureau website: 

[http://factfinder.census.gov] 

Population projections and further information 

regarding the region around SRS can be found in 

the High-Level Waste Tank Closure Final 

Environmental Impact Statement. [DOE/EIS0303] 

2.1.1.4 Land Use – Present and Planned 

Land within a five-mile radius of FTF is entirely 

within SRS boundaries and is currently used either 

for industrial purposes or as forested land. 

Current land use within the entire GSA is 

classified as heavy nuclear industrial. Plans for 

the future of SRS are addressed in two key 

planning documents identified below. 

 The SRS End State Vision [PIT-MISC0089]

 The Savannah River Site (SRS) Long 

Range Comprehensive Plan [PIT-MISC0041] 

The Long Range Comprehensive Plan assumes 

that the entire site will be owned and controlled by 

the Federal Government in perpetuity7

. 

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 long, 

warm and humid summers. Summer-like conditions typically last from May through September, when the 

area is frequently under the influence of a western extension in the semi-permanent Atlantic subtropical 

anticyclone (i.e., the ‘Bermuda’ high). Winds in summer are light and cold fronts generally remain well 

north of the area. Daily high temperatures during the summer months exceed 90°F on more than half of 

all days on average. Scattered afternoon and evening thunderstorms are common. 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 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 an average of once every two 

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

 

7

 For the purposes of the FTF 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 2020. 

Table 2.1-1: Population Distribution and 

Percent of Region of Influence (% ROI) for 

Counties and Selected Communities 

Jurisdiction 

2008 

Population 

Estimate1 2008 % ROI 

SOUTH CAROLINA 

Aiken County 154,071 28 

 Aiken, City 29,434 5.4 

 Jackson, Town 1,647 0.3 

 New Ellenton, Town 2,227 0.4 

 North Augusta, City 20,712 3.8 

Allendale County 10,447 1.9 

 Allendale, Town 3,659 0.7 

Bamberg County 15,307 2.8 

 Bamberg, Town 3,432 0.6 

Barnwell County 22,872 4.2 

 Barnwell, City 4,783 0.9 

GEORGIA 

Burke County 22,732 4.1 

Columbia County 110,627 20.1 

Richmond County 199,486 36.2 

Screven County 15,133 2.7 

Eight-County Total 550,675 

1 2008 Population estimates based on 2000 population census and 

are provided by the U.S. Census Bureau, Population Estimates 

Program; data for births, deaths, and domestic and international 

migration were used by the U.S. Census Bureau to update the 2000 

base counts. [http://factfinder.census.gov] 
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weather is somewhat windy, temperatures are usually mild and humidity is relatively low. [WSRC-TR2007-00118] 

2.1.2.2 Meteorological Data Collection 

Meteorological data are collected at SRS 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 (2 

meters) and 201.3 feet above 

ground. Temperature and dew 

point are also measured at 2 

meters. 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 1-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 then uploaded to a 

relational database for permanent 

archival. 

In addition, the Tall Tower facility 

near Beech Island, South 

Carolina, provides a set of high-quality meteorological measurements that is unique to the southeast 

United States. This facility utilizes fast-response sonic anemometers, water vapor sensors, barometric 

pressure sensors, slow-response temperature sensors and relative humidity sensors. 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. 

Precipitation measurements are collected from a network of 13 rain gauges across SRS (Figure 2.1-6). 

Twelve of these gauges are read manually by security or operations personnel once per day, usually 

around 6 am. 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 previously addressed above. [WSRC-TR-2007-00118] 

2.1.2.3 Data Pertinent to FTF PA Modeling 

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

temperature. Each is discussed below. 

Figure 2.1-6: SRS Meteorological Monitoring Network

[WSRC-TR-2007-00118] 
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2.1.2.3.1 Atmospheric Dispersion 

Since the mid-1970s, a five-year database of meteorological conditions at SRS is updated in order to 

support dose calculations for accident or routine release scenarios for onsite and offsite 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] This data is used to determine dose release factors 

in the evaluation for air pathway dose modeling described in Section 4.0 of the FTF PA, and reported in 

WSRC-STI-2007-00343. [SRS-REG-2007-00002] 

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 

provided in WSRC-STI-2007-00184. An average precipitation of 48.5 inches per year results from the 55 

year monitoring period for the site and 49 inches per year from the 46 year monitoring period for F Area. 

This data is used to determine 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 provided in WSRC-STI-2007-00184. For this 37-year period the annual 

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

approximately 46°F to a high of approximately 81°F. This data is used to determine 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 can be found in SRS Ecology 

Environmental Information Document and are 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 SRS 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 and 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 collectively referred to as “species of concern.” [WSRC-TR-2005-00201, page 3-1] 

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-EIS-0303, pg 3-26] 

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 welldrained are characterized by a mixture of oak species, as well as red maple, sweetgum and other 

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

cypress and water tupelo. [DOE-EIS-0303, page 3-26, 3-29] 
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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, page 3-2] 

Waterfowl and wading birds, as well as many upland species, use SRS aquatic habitats year round. 

Sixty-seven percent use Carolina bays and emergent marshes. Sixty-eight percent of the upland species 

use this habitat type. Edge or shoreline areas accounted for high numbers of upland birds at Carolina 

bays and emergent marshes, stream, and small drainage corridors and river swamp habitats. The 

aquatic birds are most common in large and small open water habitat. [WSRC-TR-2005-00201, pages 3-

10] 

More than 255 species of birds are found in the SRS. 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 and 

endangered plant and animal species known to occur or that might occur on the overall SRS include the 

smooth purple coneflower, wood stork, red-cockaded woodpecker and shortnose sturgeon. 

The FTF 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 F Area 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 sweetgum, red maple and red bay. Sweet bay is also 

common. The understory consists largely of saplings of these same species, as well as a herbaceous 

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

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

tag alder, willow, sweetgum and wax myrtle. The UTR seepline is located in a similar bottomland 

hardwood forest community. [DOE-EIS-0303, page 3-30] 

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, pg 3-30] 

According to summaries of studies on UTR 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 contain 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 to UTR are dominated by shiners and other small-bodied species (i.e., pirate perch, madtoms 

and darters) indicative of unimpacted streams in the Atlantic Coastal Plain. In the 1970s, the United 

States Geological Survey designated UTR as a National Hydrological Benchmark Stream due to its high 
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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, page 3-31] 

2.1.4 Geology, Seismology, and Volcanology 

Regional and local information on the geologic and seismic characteristics of the FTF 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 FTF PA, and thus further discussion of this topic 

is omitted from the following discussion. [WSRC-IM-2004-00008, page 1.5-7] 

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. Beneath the Coastal Plain 

sedimentary sequence are two geologic terranes: 1) the Dunbarton basin, a Triassic-Jurassic Rift basin, 

filled with lithified terrigenous and 

lacustrine sediments; and 2) a 

crystalline terrane of metamorphosed 

sedimentary and igneous rock that may 

range in age from Precambrian to late 

Paleozoic 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-TR95-0046] 

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 slopes to the southeast approximately 5 feet per mile. The Plateau is 

bounded by the Savannah and Congaree Rivers and 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. [WSRC-TR-95-0046] 

Figure 2.1-9 shows the topography and 10-foot contour lines of the GSA. [SRS-REG-2007-00002] 

Figure 2.1-7: Regional Geological Provinces of Eastern 

United States 

[WSRC-TR-95-0046, Figure 4-1] 
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Figure 2.1-8: Regional Geologic Provinces of South Carolina 

 [WSRC-TR-95-0046, Figure 2-3] 

Figure 2.1-9: GSA Topography
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Currently, FTF storm water drainage is directed to an outfall, which will be unaffected by FTF operations 

and tank closure activities. The installation of the FTF closure cap (Section 2.6) will necessitate changes 

to the FTF drainage system which will be designed later as part of the overall closure of FTF. 

2.1.4.2 Local Geology and Soils 

The vadose zone is comprised of the middle to late Miocene-age “Upland Unit,” that extends over much 

of SRS. 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 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. 

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 at F Area may 

contain as much as 20% to 

40% clay. [DOE-EIS-0303, 

pages 3-1 and 3-5] 

The SRS is comprised of 

seven major soil 

associations: ChastainTawcaw-Shellbluff; RembertHornsville; 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, U.S. 

Department of Agriculture, 

Soil Conservation Service. 

[http://soildatamart.nrcs.usda.

gov/Manuscripts/SC696/0/sa

vannah.pdf] 

Figure 2.1-10: General Soil Associations for SRS 

[http://soildatamart.nrcs.usda.gov/Manuscripts/SC696/0/savannah.pdf] 
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The overall general soil association for F Area is the Fuquay-Blanton-Dothan. The most predominant soil 

types within F 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. The soil material has been removed, mixed and moved. 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 

unconsolidation 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. 

A more detailed description of the geology and soils of the F 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 U.S. (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 (1698-1974), 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. [WSRCMS-2003-00617, http://neic.cr.usgs.gov/neis/last_event_states/] 

Figure 2.1-11: Historical Seismic Events in the Southeast 

[http://neic.cr.usgs.gov/neis/last_event_states/] 
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The most recent seismic event occurring within a 50-mile radius of SRS was on March 27, 2009, with a 

magnitude of 2.6. No damage to SRS was recorded. There have, however, been four earthquakes with 

epicenter locations within SRS. They occurred on June 9, 1985 (magnitude of 2.6); August 5, 1988 

(magnitude of 2.0); May 17, 1997 (magnitude of 2.3), and October 8, 2001 (magnitude of 2.6). No Strong 

Motion Accelerometers (SMAs) were triggered as a result of these earthquakes. Note that additional 

seismic events with epicenter locations within SRS occurred shortly after the October 2001 earthquake, 

however, these seismic events were attributed to aftershocks and not actual earthquakes. [WSRC-MS2003-00617] 

The regional faults within SRS and vicinity are shown in Figure 2.1-13, a study entititled Comparison of 

Cenozoic Fauling 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. The study concludes 

that these regional faults exhibit the same general characteristics, and 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. 

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

[WSRC-MS-2003-00617, http://neic.cr.usgs.gov/neis/last_event_states/] 
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In 1976, a short-period seismic network was established. This network continues to be upgraded and in 

1999 a 10-station SMA network was installed throughout the complex. Specific to F Area, one SMA is 

located near the tanks in the tank farm. Detailed information regarding seismic characteristics at SRS 

can be found in the Documented 

Safety Analysis document, 

WSRC-IM-2004-00008. 

Seismic considerations are 

included in the design of the FTF 

closure cap to ensure seismic 

induced degradation mechanisms 

are addressed. Section 2.6 

discusses the design of the FTF 

closure cap which will 

appropriately 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 FTF is 

required in order for an estimate 

of the fate and transport of the 

residual FTF contaminants to be 

modeled. Characterization and 

monitoring data in the SRS GSA 

is extensive and provides a clear 

understanding of hydrogeology 

containing the FTF, and permitted 

generation of the General 

Separations Area Database 

(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 1,219 meters (4,000 feet) at the Atlantic Coast. At SRS, coastal plain 

sediments thicken from approximately 213 meters (700 feet) at the northwestern boundary to 

approximately 430 meters (1,410 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, Dublin and Midville 

systems. [WSRC-TR-96-0399-Vol. 1] 

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 and their corresponding hydrostratigraphic units at 

SRS. This classification system is consistent with the established system, and is now widely used as 

SRS standard. [WSRC-TR-96-0399-Vol. 1, WSRC-TR-95-0046, WSRC-STI-2008-00057, Page 7-1] 

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

[WSRC-TR-2000-00310, Figure 10]
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Figure 2.1-14: Regional NW to SE Cross Section

[WSRC-TR-95-0046] 

Figure 2.1-15: Comparison of Chronostratigraphic, Lithostratigraphic, 

and Hydrostratigraphic Units in the SRS Region 

[WSRC-TR-95-0046] 
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Figure 2.1-16 and Figure 2.1-17 illustrate potentiometric maps of the UTR and Gordon Aquifers. 

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, such as Fourmile Branch, McQueen Branch and Crouch Branch in the GSA, 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, page 3-10] 

Figure 2.1-16: Potentiometric Surface of the Upper Three Runs Aquifer

[WSRC-STI-2008-00057-SRS Maps, Page 18] 
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Figure 2.1-17: Potentiometric Surface of the Gordon Aquifer

[WSRC-STI-2008-00057-SRS Maps, Page 19] 
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2.1.5.2 Characterization of Local Hydrogeology 

The GSA has been the focus of numerous geological and hydrogeologic 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 locations of laboratory permeability data, multiple 

well pump tests, single well pump test 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 FTF. 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 4.0 of the 

FTF PA. [SRS-REG-2007-00002] 

Figure 2.1-18: Hydrostratigraphic Picks in GSAD Database 

[WSRC-TR-96-0399-Vol. 2] 
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Figure 2.1-19: Laboratory Determined Permeability Data in GSAD 

Database 

[SRNL-ESB-2007-00035] 

Figure 2.1-20: Multiple Well Pump Test Data in GSAD Database 

[WSRC-TR-96-0399-Vol. 2] 
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Figure 2.1-21: Single Well Pump Test Data in GSAD Database 

[WSRC-TR-96-0399-Vol. 2] 

Figure 2.1-22: Slug Test Data in GSAD Database 

[WSRC-TR-96-0399-Vol. 2] 
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2.1.5.3 Groundwater Flow in the GSA 

The aquifers of primary interest for FTF modeling are the UTR and Gordon Aquifers. Plate 17 of the 

Hydrogeological Framework of West-Central South Carolina Report (PIT-MISC-0112) gives the leakance 

of the Crouch Branch confining unit (of the Meyers Branch confining system) as roughly 3E-06 day-1, 

which corresponds to 0.13 inch/year for every 10 feet of head difference. The head difference across the 

Crouch Branch confining unit 

ranges from 0 to 20 feet, 

causing an upward flow 

averaging 0.13 inch/year. [PITMISC-0112, Figure 30] Flow 

across the unit is therefore a 

small fraction of total recharge, 

and is negligible in the FTF 

modeling. Potential 

contamination from the FTF is 

not expected to enter the 

deeper Crouch Branch Aquifer 

because an upward gradient 

exists between the Crouch 

Branch and Gordon Aquifers 

near UTR. Figure 2.1-23 is a 

cross-sectional schematic 

representation of groundwater 

flow patterns in the UTR and 

Gordon Aquifers along a northsouth cross-section running 

through the center of FTF, 

shown with significant vertical 

exaggeration. Section 4.0 of 

the FTF PA provides the 

modeling inputs associated with 

groundwater flow characteristics 

obtained from the GSAD. 

[SRS-REG-2007-00002] 

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

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 

Gordon Aquitard 52 Locations 

Gordon Aquifer 146 Locations 

Green Clay Aquitard 161 Locations 

Upper Three Runs – Lower Zone (UTR-LZ) 222 Locations 

Tan Clay Aquitard 225 Locations 

WSRC-TR-96-0399-Vol. 1, App. A 

Permeability Measurements 

Pumping Tests 85 Values 

Slug Tests 481 Values 

Laboratory Permeability 258 Values 

WSRC-TR-96-0399-Vol. 2, App. B 

Water Levels 

Gordon Aquifer 79 Locations 

UTR-LZ 173 Locations 

Upper Three Runs – Upper Zone 387 Locations 

WSRC-TR-96-0399-Vol. 2, App. C 

Figure 2.1-23: Conceptual Diagram of Groundwater Flow 

beneath the GSA 
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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 river-miles 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 averages approximately 10,400 cubic feet per second at SRS. 

For 2007, the annual average measured flow rate was 6,090 cubic feet per second. [WSRC-TR-2005-

00201, Table 4-24, WSRC-STI-2008-00057] 

Figure 2.1-24: Savannah River Basin Dams 
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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 (Section 2.1.2), water pumped 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, Savannah River supplies domestic water and is used 

for commercial and sport fishing, boating and other recreational activities. [DOE-EIS-0303, page 3-7] 

Figure 2.1-25: SRS Watershed Boundaries and Major Tributaries 

 [WSRC-STI-2008-00057] 
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The natural flow of SRS streams range from 8 cubic feet per second in smaller streams to 245 cubic feet 

per second in UTR. [WSRC-IM-2004-00008, Table 1.4-14] 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

contains detailed information on flow rates and water quality of the Savannah River and SRS streams. 

[WSRC-STI-2008-00057] 

The SCDHEC regulates the 

physical properties and 

concentrations of chemicals and 

metals in SRS effluents under 

the National Pollutant Discharge 

Elimination System (NPDES) 

program. SCDHEC, which also 

regulates biological Water 

Quality Standards for SRS 

waters, have classified the 

Savannah River and SRS 

streams as “Freshwaters.” 

“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. [DOE-EIS-0303] 

The UTR, longest of the SRS streams, 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. It is the only major tributary on SRS that has not received thermal discharges. The UTR valley has 

meandering channels, 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 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 flows parallel 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 fairly 

steep to gently sloping. The floodplain is up to 1,000 feet wide. [DOE-EIS-0303, pages 3-8 and 3-9] 

Flood hazard recurrence frequencies have been calculated for the various SRS site areas. The 100-year, 

1,000-year and 10,000-year flood water levels for UTR basin near F Area is approximately 138, 140 and 

143 feet above mean sea level (MSL), respectively. [WSRC-TR-99-00369] As shown in Table 2.1-6, the 

lowest elevation of a waste tank within FTF is approximately 227 feet above MSL; thus the highest period 

flood water level of approximately 143 feet above MSL is approximately 84 feet below the lowest 

elevation of an FTF waste tank. In addition, the lowest elevation of the lower foundation layer, at the 

bottom of the side slope of the conceptual closure cap is approximately 270 feet above MSL, which is 

Figure 2.1-26: GSA Gauging Stations
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approximately 127 feet above the highest flood water level of 143 feet. [WSRC-STI-2007-00184] 

Therefore, flooding is not a concern for the FTF PA. 

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 Coefficients (Kds) and are identified below. The 

data used in the radionuclide transport model is presented in Section 4.0 of the FTF PA specific to the 

GSA and is not reproduced in this section. [SRS-REG-2007-00002] The following studies detail the 

information: 

 WSRC-TR-2006-00004, Geochemical Data Package for Performance Assessment Calculations 

Related to the Savannah River Site

 SRNL-ESB-2007-00008, F-Area Tank Farm Vadose Zone Material Property Recommendations

 WSRC-STI-2007-00544, Conceptual Model of Waste Release from the Contaminated Zone of 

Closed Radioactive Waste Tanks

 SRNL-SCS-2007-00011, Preliminary Guidance for the Distribution of Cs, SR, and U, 

Geochemical Input Terms to Stochastic Transport Models

2.1.7 Natural Resources 

Natural resources at SRS are managed under the Natural Resources Management Plan (NRMP) 

prepared for the DOE by the United States Department of Agriculture. [NRMP-2005] The NRMP was 

recently updated in May 2005, and fosters the following principles which govern SRS natural resource 

management: 

 All work will be done in accordance with Integrated Safety Management Procedures found in 

DOE P 450.4, Safety Management System 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 state-administered 

NPDES permit program and requirements, under the Federal Water Pollution Control Act (commonly 

called the Clean Water Act). [http://epw.senate.gov/water.pdf] As required by EPA and SCDHEC, SRS 

has NPDES permits in place for discharges to the waters of the United States. These permits establish 

the specific sites to be monitored, parameters to be tested, and monitoring frequency—as well as 

analytical, reporting, and collection methods. [WSRC-STI-2008-00057, page 4-7] Continuous 

surveillance monitoring of site streams occurs downstream of several process areas to detect and 

quantify levels of radioactivity in effluents transported to the Savannah River. [WSRC–STI–2008–00057, 

page 5-3] 

2.1.7.1.1 Surface Water 

The SRS streams and the Savannah River are classified by SCDHEC as “Freshwaters,” which are 

defined as surface water suitable for: 

 primary and secondary contact recreation and as a drinking water source after treatment in 

accordance with SCDHEC requirements and 
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 fishing and survival and propagation of a balanced indigenous aquatic community of fauna and flora, 

and Industrial and agricultural uses. [WSRC–STI–2008–00057] 

Table 2.1-3 characterizes Savannah River water quality both upstream and downstream of SRS. Table 

2.1-4 characterizes water quality in the UTR upstream and downstream of the GSA. 

2.1.7.1.2 Groundwater 

The SDWA was enacted in 1974 to protect public drinking water supplies. 

[www.epa.gov/safewater/sdwa/index.html] SRS domestic water is supplied by 17 separate systems, 

all of which utilize groundwater sources. The A-Area, D-Area, and K-Area systems are actively regulated 

by SCDHEC, while the remaining 14 site water systems receive a reduced level of regulatory oversight. 

[WSRC-STI-2008-00057]

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

Year 2007) 

Upstream Downstream 

Parameter Unit of Measure Minimum Maximuma

 Minimum Maximuma

Aluminum mg/L 0.062 0.97 0.0777 0.622 

Cadmium mg/L ND ND ND ND 

Chromium mg/L ND ND ND ND 

Copper mg/L ND 0.0388 ND 0.0083 

Dissolved Oxygen mg/L 6.1 11.1 6.2 10.1 

Gross Alpha 

Radioactivity pCi/L ND 1.26 ND 1.55 

Lead mg/L ND ND ND 0.0023 

Mercury mg/L ND 0.000023 ND 0.000024 

Nickel mg/L ND 0.003 ND 0.0066 

Nitrate (as N) mg/L 0.18 0.57 0.19 0.41 

pH pH units 5.9 7.1 6.6 16.1b

Phosphate mg/L 0.032 0.18 0.084 0.17 

Suspended solids mg/L 1 47 4 15 

Temperature °F 52.5 78.6 53.2 82.4 

Tritium pCi/L ND 229 140 1,060 

Zinc mg/L ND 0.0087 ND 0.0173 

a

 The maximum listed concentration is the highest single result found during one sampling event. 

b

 Highest one month sample, next highest value reported was 7.2 pH unit. 

Note: Information extracted from WSRC-STI-2008-00057 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 Upper Three Runs

Temperature (o

F) pH 

Dissolved 

Oxygen (mg/L) 

Total Suspended 

Solids (mg/L) 

Sampling Location: Upper Three Runs (Upstream of GSA) 

Mean 61 5.9 9.1 4.6 

Range 51 - 70 5.4 - 6.7 4.4 - 14 3 - 9 

Sampling Location: Upper Three Runs (Downstream of GSA) 

Mean 65 6.5 9.6 6.2 

Range 52 - 74 5.2 - 7.6 6.4 - 16 4 - 11 

Note: All data extracted from WSRC-STI-2008-00057 accompanying data files 
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Page 2-28 

Industrical

(<0.1%)

Occupational

(<0.1%)

Consumer

(2%)

Conventional

Radiography/

Fluoroscopy

(medical)

(5%)

Interventional

Fluoroscopy

(medical)

(7%)

Terrestrial

(background)

(3%)

Internal

(background)

(5%) Space

(background)

(5%)

Radon & Thoron

(background)

(37%)

Nuclear Medicine

(medical)

(12%)

Computed

Tomography

(medical)

(24%)

Table 2.1-5 provides the summary of maximum groundwater monitoring results for those areas that most 

likely outcrop to Fourmile Branch obtained from the 2007 Environmental Report. [WSRC-STI-2008-

00057] The groundwater in these areas is not being consumed and active remediation projects are in 

progress to address the groundwater conditions. 

2.1.8 Natural and Background 

Radiation 

All human-beings are exposed to 

sources of ionizing radiation that 

include naturally occurring and manmade sources. An individual’s 

average dose contribution estimates 

from various sources were obtained 

from the review information 

presented in NCRP Report 160 and 

are shown in Figure 2.1-27. [NCRP160] 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/year. The dose from SRS 

operations to the maximally exposed 

offsite individual during calendar 

year 2007 was estimated to be 0.1 

mrem. [WSRC-STI-2008-00057, 

page 6-7] 

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 

Figure 2.1-27: Major Sources of Radiation Exposure in the 

Vicinity of SRS

 [NCRP-160] 

Table 2.1-5: Summary of Maximum Groundwater Monitoring Results for Major Areas that 

Outcrop to Fourmile Branch, 2006–2007 

Location 

Major 

Contaminants Units 

2006 

Maximum MCL 

2007 

Maximum Likely Outcrop Point 

E-Area Tritium 

TCE 

pCi/L 

ppb 

33,600,000 

750 

20,000 

5 

30,800,000 

370 

UTR/Crouch Branch in 

North; Fourmile Branch in 

South 

F-Area 

TCE 

Tritium 

Gross alpha 

Beta 

ppb 

pCi/L 

pCi/L 

pCi/L 

78.9 

91,500 

2,030 

1,620 

5 

20,000 

15 

4 mrem/yra

52.2 

73,000 

2,120 

380 

UTR/Crouch Branch in 

North; Fourmile Branch in 

South 

F-Area 

Seepage 

Basin 

Tritium 

Gross alpha 

Beta 

pCi/L 

pCi/L 

pCi/L 

7,140,000 

627 

2,360 

20,000 

15 

4 mrem/hra

5,710,000 

523 

1,870 

Fourmile Branch 

H-Area 

Tritium 

Gross alpha 

Beta 

pCi/L 

pCi/L 

pCi/L 

80,400 

98 

116 

20,000 

15 

4 mrem/yra

67,200 

25.5 

55.6 

UTR/Crouch Branch in 

North; Fourmile Branch in 

South 

H-Area 

Seepage 

Basins 

Tritium 

Gross alpha 

Beta 

pCi/L 

pCi/L 

pCi/L 

3,690,000 

103 

2,840 

20,000 

15 

4 mrem/yra

3,020,000 

88.4 

2,970 

Fourmile Branch 

a

 The activity (pCi/L) equivalent to 4 mrem/yr varies according to which specific beta emitters are present in the sample. [WSRC-STI-2008-00057, 

Page 7-9] 
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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 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 (<1 mrem/yr) and occupational exposure 

(<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 NASA space missions. [SRR-LWP2009-00001] 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 separation 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 such as C-14, Sr-90 and Tc-99. 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-2010-00040] 

As of March 2010, 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 37,400,000 

gallons. [SRR-LWP-2009-00001, SRR-LWP-2010-00040] Operations have been extremely effective in 

minimizing waste volume stored in SRS waste tanks but, because the majority of the waste has been fully 

concentrated using the available SRS equipment, further reductions via evaporation of the total stored 

waste volume is not practical. The SRS no longer conducts weapons or NASA-related material 

production 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 dissolving non-irradiated fuel and recovering and blending HEU for nondefense related use. 

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 wash water from sludge washing. 

2.1.10 F-Tank Farm 

The F Area occupies approximately 364 acres. It includes FTF and the F-Canyon chemical separations 

facility8

, where nuclear materials were recovered using the PUREX process and the FTF. The FTF is 

located in the southern region of F Area and occupies approximately 22 acres. The FTF is an active 

waste storage facility used to store liquid radioactive waste generated primarily during prior operations of 

F Canyon. Figure 2.1-28 shows the FTF layout. 

 

8

 The final transfer of waste from F Canyon to the FTF occurred in August 2005. 

Figure 2.1-28: General Layout of FTF
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The FTF consists of: 

 22 carbon steel waste tanks9

 (i.e., Tanks 1 through 8, Tanks 17 through 20, Tanks 25 through 28, 

Tank 33, Tank 34 and Tanks 44 through 47), 

 ancillary structures: 

− the FTF transfer line system including approximately 45,000 linear feet of underground waste 

transfer lines, 

− three pump pits (FPP-1, FPP-2 and FPP-3), 

− three pump tanks (FPT-1, FPT-2 and FPT-3), 

− one 11,700 gallon catch tank, 

− two evaporator systems used to reduce waste volume (242-F and 242-16F), 

− the 242-F concentrate transfer system, 

− six diversion boxes (FDB-1, FDB-2, FDB-3, FDB-4, FDB-5 and FDB-6), and 

− eight valve boxes (valve boxes 1 through 5, valve boxes 28A and 28B and LDB-17 valve 

box). 

This equipment is discussed separately as follows: 

 waste tanks (Type I, III/IIIA and IV) 

 ancillary structures 

2.1.11 Waste Tanks 

All 22 of the FTF waste tanks were built of carbon steel and reinforced concrete in three principal designs, 

Types I, III/IIIA and IV. The FTF does not contain any Type II waste tanks10. 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. Table 2.1-6 summarizes the important design features. 

The general design features of the waste tanks are summarized below, followed by brief descriptions and 

illustrations of the different waste tanks types. 

2.1.11.1 General Tank Design Features 

Each Type I, III and IIIA waste 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 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. 

 

9

 Two tanks, Tank 17 and Tank 20, were operationally closed in 1997 and are not part of this Draft FTF 3116 Basis Document. 

10 Four Type II tanks were constructed in HTF; no Type II tanks were constructed in FTF. 

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

Type 

Waste Tank 

Numbers 

Diameter 

(ft) 

Height 

(ft) 

Capacity 

(gal) 

Cooling 

Coils 

Secondary 

Containment 

Elevation 

(ft)a

I 1, 2, 3, 4, 5, 6, 

7, 8 75.0 24.5 750,000 Yes Yes 237 - 241 

III/IIIAb

25, 26, 27, 28, 

33, 34, 44, 45, 

46, 47 

85.0 33.0 1,300,000 Yes Yes 244 - 246 

IV 17, 18, 19, 20 85.0 34.5 1,300,000 No No 227 - 228 

a Approximate feet above MSL for the waste tank basemat. b Tanks 33 and 34 are Type III waste tanks. The remaining tanks are Type IIIA.
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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, 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 and 

Type III/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 Waste Tanks 

Type I waste tanks (Tanks 1 through 8) 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 waste tanks have similarly configured 

vertical and horizontal cooling coils. A typical11 Type I waste 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 FTF PA. [SRS-REG-2007-00002] 

 

11 The word “typical” as used throughout this section refers to representative design features of the system being described. 

Figure 2.1-29: Typical Type I Waste Tank Cross Section 
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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 3-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 4-inch diameter carbon steel jacket pipe where it bridges the waste 

tank annulus. [SRS-REG-2007-00002] 

The waste tank vault is constructed of 22-inch thick reinforced concrete with an inner diameter of 80 feet. 

Approximately nine feet of soil covers the vault roof as shown in Figure 2.1-29. [SRS-REG-2007-00002] 

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. [SRS-REG2007-00002] 

Cooling coils in Type I waste 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 waste 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 section 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. 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 2-inch 

carbon steel pipe cooling coils in a Type I waste 

tank. [D116048, C-CLC-G-00364, D116001] 

Figure 2.1-30: Type I Waste Tank Basemat 

Construction Figure 2.1-31: Type I Waste Tank Primary and 

Secondary Liner Construction 

Figure 2.1-32: Typical Type I Waste Tank Column 

Support 
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Visual and equipment manipulation access within the Type I waste 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] 

2.1.11.3 Type III/IIIA Waste Tanks 

The Type III waste tanks were constructed in 1969 (Tank 33) and 1972 (Tank 34). The Type IIIA waste 

tanks were complete in 1978 (Tanks 25 through 28) and 1980 (Tanks 44 through 47). All Type III and IIIA 

waste 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). [SRS-REG-2007-00002] 

Each Type III waste tank has a minimum 3.5-foot thick reinforced concrete basemat placed on top of a 

4-inch concrete working slab. The Type IIIA waste tanks have a minimum 3-foot-7-inch thick reinforced 

concrete basemat placed on top of a 4-inch concrete working slab. Both Type III and IIIA waste tanks 

have a center drop panel that varies the basemat thickness for each tank type. Additional Type III and 

IIIA waste tank details are available in Section 3.0 of the FTF PA. [SRS-REG-2007-00002] 

A typical FTF Type III waste tank cross section is shown in Figure 2.1-35, typical Type IIIA waste tank 

cross section is shown in Figure 2.1-36 and waste tank basemat construction is shown in Figure 2.1-37. 

Figure 2.1-33: Typical Type I Waste 

Tank Cooling Coil Obstacles Figure 2.1-34: Type I Waste Tank Access Area 

for Waste Removal Equipment Diagram 
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Figure 2.1-35: Typical FTF Type III Waste Tank Cross Section 

 Figure 2.1-36: Typical Type IIIA Waste Tank Cross Section 

Figure 2.1-37: Typical Type III/IIIA Waste Tank 

Basemat Construction 
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Upper Knuckle

Basemat 

Secondary Liner

Primary Tank Wall

The Type III/IIIA waste 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 2.1-38. The Type III/IIIA waste 

tank primary tanks were fully stress-relieved by heating after fabrication. The waste tank primary tank 

and secondary liner plate data are summarized in Table 2.1-7. [SRS-REG-2007-00002] 

Both Type III and IIIA waste tanks contain multiple penetrations through a carbon steel primary tank (e.g., 

2-inch diameter pipe in 3-inch diameter sleeves) near the top of the waste tank for transfer lines into and 

out of the tank. [SRS-REG-2007-00002] 

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. The Type III/IIIA tanks were 

constructed below grade with native soil 

backfilling around and between the tanks. 

Type III/IIIA waste 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 

Figure 2.1-38: Typical Type IIIA Waste Tank Primary Tank and Secondary Liner - Late 

Construction (Vault Wall Not Constructed) 

Table 2.1-7: Type III/IIIA Waste Tanks Primary 

Tank and Secondary Liner Plate Data 

Location Thickness (in) 

Primary tank roof 0.5 

Primary tank floor 0.5 

Upper knuckle 0.5 

Secondary liner - upper band 0.5 

Secondary liner - middle band 0.625 

Secondary liner - lower band 0.750 

Primary tank wall - upper band 0.5 

Primary tank wall - lower band 0.625 

Lower knuckle - secondary 0.875 

Lower knuckle - primary 0.625 
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the primary tank floor and then out to the outer annulus through the radial air grooves, as shown in Figure 

2.1-39. 

Type IIIA waste 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 steellined center support column that is integrated into the concrete basemat (Figure 2.1-35 and Figure 

2.1-36). 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. [SRS-REG-2007-

00002] 

Type III waste tanks cooling coil 

piping consists of deployable 

coolers (conical coolers and 

cylindrical coolers as shown in 

Figure 2.1-40). These coolers 

were inserted through the risers in 

the closed position and deployed 

(opened) once inside the 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 have 

permanently installed cooling coils 

as shown in Figure 2.1-41. These 

are made of 2-inch diameter 

carbon steel pipe. [M-CLC-H02820, SRS-REG-2007-00002] 

Figure 2.1-39: Typical Type III/IIIA Waste Tank Radial Air 

Grooves 

Figure 2.1-40: Typical Type III Waste Tank Deployable Cooling Coils 

Conical Cylindrical 
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The tank riser design configuration restricts visual and equipment access into Type III/IIIA waste 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 2.1-42 and Figure 2.1-43. [W704010, W449795] 

Figure 2.1-41: Typical Type IIIA Waste 

Tank Cooling Coils During Construction 

Figure 2.1-42: Typical Type III Waste Tank 

Access Area for Waste Removal Equipment 

Diagram 

Figure 2.1-43: Typical Type IIIA Waste Tank 

Access Area for Waste Removal Equipment 

Diagram 
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Stiffener Angle 

Stiffener Angle 

2.1.11.4 Type IV Waste Tanks 

The Type IV waste tanks were constructed in the late 1950s. 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 FTF Type IV waste tank cross 

section is shown in Figure 2.1-44. Additional Type IV waste tank details are available in Section 3.0 of 

the FTF PA. [SRS-REG-2007-00002] 

The Type IV waste 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 2.1-44. The primary waste tank 

liner is internally reinforced with three, 

circumferential, 4-inch carbon steel 

stiffener angles as shown in Figure 

2.1-45. The primary liner is anchored 

to the concrete. Each waste tank has 

side wall penetrations through the 

vault and waste tank liner located just 

below the dome for transfer lines into 

and out of the tank. [SRS-REG-2007-

00002] 

Each Type IV waste tank is enclosed 

in a concrete vault. Each vault was 

constructed in layers using a 

“shotcrete” technique. No secondary 

containment structure or annulus 

exists with this design. 

The concrete vault walls are cylindrical with an inside diameter of 85 feet and surmounted by a dome ring 

33 feet high. The core wall is 7 inches at the top and 11 inches at the bottom. This core wall was 

prestressed with 163 steel bands that remained in place and were covered with sprayed-on concrete 

(Figure 2.1-46). The wall thickness, including the bands and concrete cover, is 10 inches at the top and 

15 inches at the bottom. [SRS-REG-2007-00002] 

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 spring line. 

The Type IV waste tank dome and risers are shown, 

near the end of the concrete construction phase, in 

Figure 2.1-47. [SRS-REG-2007-00002] 

Figure 2.1-44: Typical FTF Type IV Waste Tank Cross 

Section 

Figure 2.1-45: Typical Type IV Waste Tank 

Stiffener Angles 
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The concrete roof of a Type IV waste tank is not lined with carbon steel on the inside. The center riser 

has an inner diameter of 10 feet and is approximately 4 feet in length. The six smaller risers each have a 

inner diameter of 2 feet and are approximately 5 feet in length. The tank riser design configuration 

provides limited access to the tank interior. New tank top openings have been constructed in Type IV 

tanks to support waste removal and sampling activities. [SRS-REG-2007-00002] 

Figure 2.1-46: Typical Type IV Waste Tank Wall Steel Bands 

During Construction 

Figure 2.1-47: Typical Type IV Waste Tank Domes and 

Risers During Construction 
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2.1.12 Ancillary Structures 

In addition to the waste tanks, FTF contains ancillary structures with a residual inventory that must be 

accounted for as part of FTF 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 FTF 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 2.1-48 identifies locations of FTF-specific ancillary structures. 

A description of the FTF ancillary structures is 

provided as follows: 

 the FTF transfer line system (approximately 

45,000 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 FTF pump tanks (FPT-1, FPT-2 and FPT3), FTF pump pits (FPP-1, FPP-2 and FPP-3) 

and the FTF catch tank, 

 the 242-F Evaporator System, including the 

evaporator cell and support tanks (e.g., 

mercury collection tank, cesium removal 

column feed tank, concentrate transfer system 

tank and two overheads tanks), 

 the 242-16F Evaporator System, including the 

evaporator cell and support tanks (e.g., 

mercury collection tank, cesium removal 

column feed tank and two overheads tanks), 

 the FTF diversion boxes (FDB-1, FDB-2, FDB3, FDB-4, FDB-5 and FDB-6), and 

 the FTF valve boxes (valve boxes 1 through 5, 

valve boxes 28A and 28B and LDB-17 valve 

box). 

[SRS-REG-2007-00002] 

2.1.12.1 Waste Transfer Lines 

There are over 45,000 linear feet of waste transfer line 

in FTF, with the line segments ranging from a few feet 

to over 4,000 feet in length. The FTF 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 cement-asbestos. 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. [SRS-REG-2007-00002] 

Figure 2.1-48: General Layout of FTF
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Transfer Lines 

During 

Construction 

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 2.1-49 shows waste transfer lines during construction. [SRS-REG-2007-00002] 

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 3 feet 

thick. All FTF pump pits house a pump tank with the pump pits providing secondary containment for 

pump tanks. Figure 2.1-48 shows the location of the three pump pits in FTF. [SRS-REG-2007-00002] 

The pump pits are often constructed in conjunction with a diversion box (diversion box details are 

discussed later in this section). A typical diversion box/pump pit configuration is depicted in Figure 

2.1-50. Figure 2.1-51 is a photograph of a pump tank. 

Figure 2.1-49: Typical Waste Tank Transfer Line During Construction 
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Figure 2.1-50: Typical Diversion Box and Pump Pit Design 

Figure 2.1-51: Typical Pump Tank Top View During 

Construction 
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The following is a description of pump pit and pump tank features. 

 FPP-1/FPT-1: The FPP-1 has a volume of approximately 33,000 gallons. The FPT-1 has a 

volume of approximately 7,200 gallons and currently serves as the inter-area pump tank. 

 FPP-2/FPT-2 and FPP-3/FPT-3: These pump pits have volumes of approximately 

37,400 gallons. The FPP-3 is connected through a pipe chase to the adjacent FPP-2. The FPP2 and FPP-3 underliners drain to the FPP-2 underliner sump. The FPT-2 and FPT-3 are 7,200 

gallon tanks that previously received waste transfers from the F-Canyon Facility. 

 [SRS-REG-2007-00002] 

There is a single catch tank in FTF designed to collect drainage from FDB-1 and the Type I waste 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 over 3 feet 

thick and is built on a 4-inch thick concrete pad (Figure 2.1-52). [SRS-REG-2007-00002] 

2.1.12.3 Evaporator Systems 

There are two evaporator systems in the FTF, the 242-F Evaporator System and the 242-16F Evaporator 

System. Evaporators are used to reduce the volume of liquid radioactive waste within FTF 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. The 242-F Evaporator System also included the 242-3F 

concentrate transfer system, which was used to distribute evaporator concentrated material throughout 

FTF. [SRS-REG-2007-00002] 

2.1.12.3.1 242-F Evaporator System 

The 242-F Evaporator Facility was constructed and placed into service in 1960 and was removed from 

service in 1988. The 242-F evaporator cell is a cuboid with a 16 feet x 15 feet base and a height of 25 

feet. The cell includes a floor sump measuring 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. [SRS-REG-2007-00002] Figure 2.1-53 provides a schematic of the 242-F 

Evaporator System. 

Figure 2.1-52: Typical Catch Tank Cross Section 
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 242-F Evaporator Vessel: The evaporator vessel, located inside the 242-F evaporator cell, is a 

stainless steel, cylindrical vessel with a cone bottom. The cylindrical portion is 8 feet in diameter 

with a height of 8-foot-9.75-inches. The cone has a maximum diameter of 8 feet and a height of 

5-feet-11-inches. The evaporator was used to concentrate liquid by evaporating water from the 

waste to reduce waste volumes. 

 242-F Overheads System: The receiver cell is cuboid with a 15 feet x 9 feet x 8-inch base and a 

height of 6-foot-4-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 North and South overheads tanks, located inside the 242-F receiver 

cell, are each cylindrical, stainless steel vessels having a diameter of 6 feet and a height of 6 feet. 

The overheads tanks functioned as receipt tanks for liquids condensed from evaporator vapors. 

The 242-F condenser is a stainless steel cylindrical vessel with an outer diameter of 18 inches 

and a height of 9-foot-10.25-inches. The condenser functioned to condense evaporator vapors 

into liquid, which was drained to the North and South overheads tanks. 

 242-F Concentrate Transfer System: The concentrate from the 242-F evaporator was steam 

lifted to the concentrate transfer system. The concentrate transfer system draw-off pump 

circulated the concentrate continuously through a loop line to the concentrate receipt tanks. 

Various tanks have served as concentrate receipt tanks over the service life of the 242-F 

Evaporator System. When concentrate reached a predetermined level in the concentrate transfer 

system waste tank, a drop valve opened adding the concentrate to a receipt tank. 

The concentrate transfer system pit is cuboid with a 12 feet x 12 feet base and a height of 21 feet. 

The pit includes a floor sump with a 1.5 feet x 1.5 feet base and a depth of 1.5 feet. The 

concentrate transfer system pit provided containment for the concentrate transfer system waste 

tank and featured a stainless steel liner. Cell covers provide personnel protection. The stainless 

steel concentrate transfer system tank, located inside the 242-3F concentrate transfer system pit, 

is a cylindrical vessel with a diameter of 8 feet and a height of 8-foot-4-inches. 

[SRS-REG-2007-00002]

Figure 2.1-53: 242-F Evaporator System Schematic 
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2.1.12.3.2 242-16F Evaporator System 

The 242-16F Evaporator Facility was constructed in 1980 and continues to operate. The 242-16F 

evaporator facilities are arranged into three cells and a gang valve house. The evaporator cell contains 

the evaporator, the condenser cell contains the condenser and a diked overheads cell contains 

overheads system components other than the condenser. Figure 2.1-54 provides a schematic of the 242-

16F Evaporator System. [SRS-REG-2007-00002] 

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 2 feet 

thick, composed of 

concrete slab sections 

and a sloped, galvanized 

steel rain cover. The 

concrete slab sections 

and the rain cover have 

access ports for valves 

and viewing. The 

evaporator cell is 

stainless steel-lined for 

collecting leakage 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. [SRS-REG-2007-00002] 

The condenser cell is 9 feet x 10 feet x 14 feet high with walls constructed of concrete and has a roof 

composed of concrete slab sections, and a sloped, galvanized steel rain cover. The concrete slab 

sections and the rain cover have access ports for viewing. The condenser cell contains a stainless steel 

liner pan on a sloped floor. The condenser cell has an opening to the evaporator cell. The deentrainment column piping enters the condenser cell through this opening, which also permits airflow to 

the evaporator cell. [SRS-REG-2007-00002] 

The overheads cell (which is open to the environment) contains the mercury removal tank, cesium 

removal column feed tank, two cesium removal column pumps and two overheads tanks, an overheads 

tank sample system, filters for removing zeolite from condensate, cesium removal column gamma 

monitors, cesium removal column charging jet, and two overheads pumps. [SRS-REG-2007-00002] 

 242-16F Evaporator Vessel: The evaporator vessel has a capacity of approximately 4,400 

gallons. The insulated vessel is 8 feet in diameter with a height of 16.5 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. 

 242-16F 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 

Figure 2.1-54: 242-16F Evaporator System Schematic 
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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. The overheads tanks are constructed of stainless steel and are 6 feet 

in diameter with a height of 6 feet. A 2-inch diameter overheads tanks overflow line is routed to 

the cell sump. 

[SRS-REG-2007-00002] 

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. 

Diversion boxes are located below grade and are either stainless steel-lined or sealed with waterproofing 

compounds, as described below, to prevent ground contamination. The walls are approximately 2 to 3 

feet thick and sloped floors are approximately 3 feet thick. The diversion boxes have concrete slab-type 

cell covers, approximately 2 to 3 feet thick that must be removed for changing jumper alignment. [SRSREG-2007-00002] 

The following is a description the features associated with the FTF diversion boxes: [SRS-REG-2007-

00002] 

 FDB-1 has a total volume of approximately 84,000 gallons. The FDB-1 drains to the F-Area 

catch tank. The interior of the diversion box is painted with chemical-resistant paint and sealer. 

FDB-1 has no sump. 

 FDB-2 has a total volume of approximately 8,300 gallons. The FDB-2 is ventilated through a vent 

duct to the adjacent FTF pump pit (FPP-1). Leakage into the FDB-2 sump drains into FPP-1 

through an opening in the sump. The FDB-2 is the connection point from FTF to the Inter-Area 

Line. 

 FDB-3 has a total volume of approximately 1,900 gallons. The diversion box is stainless steellined and has a sump that gravity drains to FDB-2. 

 FDB-4 has a total volume of approximately 21,700 gallons with an underliner. The FDB-4 is 

ventilated through openings (vent and drain slots and through the pipe chase) to the adjacent 

FPP-2 and FPP-3. The diversion box is stainless steel-lined and has a sump that gravity drains 

into FPP-2. 

 FDB-5 has a total volume of approximately 18,600 gallons. The FDB-5 was installed to connect 

the concentrate transfer system with Tanks 25 through 28 or Tanks 33 and 34. The diversion box 

is stainless steel-lined and has a sump that gravity drains to FPP-2. 

 FDB-6 has a total volume of approximately 22,000 gallons. The FDB-6 was installed so that the 

242-F evaporator could be fed from Tank 26 as well as from Tank 7. The diversion box is 

stainless steel-lined and has a sump that gravity drains to FPP-3. 

[SRS-REG-2007-00002] 

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 

FTF 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. [SRS-REG-2007-00002] 

 Valve boxes 1 through 5 facilitate transfers between FDB-2, FPP-1 and Tanks 4 through 6 and 

Tank 8. Valve boxes 1 through 4 have approximate volumes of 330 gallons, while valve box 5 is 

slightly larger at approximately 380 gallons. 

 LDB-17 valve box has an approximate volume of 140 gallons and facilitates Tank 8 transfers. 

Leakage in LDB-17 drains back to a riser on Tank 8. 
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 Valve boxes 28A and 28B - Valve box 28A is located on the top of Tank 28 and has an 

approximate volume of 450 gallons. Valve box 28B is adjacent to Tank 28 and has an 

approximate volume of 365 gallons. Leakage to valve box 28B drains to leak detection boxes 

and the leak detection boxes drain header. Valve box 28A drains back to Tank 28. Valve boxes 

28A and 28B, together, are commonly referred to as valve box 28. 

[SRS-REG-2007-00002] 

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 leaks are 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. [SRS-REG-2007-00002] 

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. [SRS-REG-2007-00002] 

2.2 F-Tank Farm Wastes 

Most of the waste managed in FTF originated in the F-Canyon Separations Facility12, which made use of 

the PUREX separations process. Typically in this process, uranium and plutonium were extracted from 

aluminum-clad, depleted uranium slugs (known as targets) that were irradiated in the site’s nuclear 

production reactors. 

Before transfer of the waste from the F Canyon to the tank farms, sodium hydroxide was 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. 

Providing additional space in the waste tanks allows additional receipt of waste from the processing 

facilities. [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 radioactive material such as C-14, Sr-90 and Tc-99. 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 FTF waste tanks is comprised of a combination of sludge, supernate and saltcake. 

 

12 The final transfer of waste from F Canyon to the FTF occurred in August 2005. 
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2.2.1 Sources of the Waste 

This section describes the origin of the wastes managed in FTF and explains how these wastes are 

managed. 

2.2.1.1 The PUREX Process 

This subsection briefly describes the PUREX process that generated most of the FTF waste. [DPSPU 

77-11-1] 

As discussed previously, most of the waste managed in FTF originated in the F-Canyon chemical 

separations facility, which made use of the PUREX separations process. The first major step in the FCanyon PUREX process involved dissolving the aluminum cladding. After this was done, the targets 

were dissolved in nitric acid. The resulting solution contained uranium (principally U-238) and plutonium, 

along with fission products such as Cs-137 and Sr-90. 

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 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 and plutonium. 

The solvent product stream underwent additional processing cycles to separate the uranium and 

plutonium and to purify these materials. The aqueous waste stream containing 95% of the fission 

products was evaporated to reduce the waste volume and recover the nitric acid. It was then processed 

through a primary recovery column where plutonium and neptunium were recovered. The waste stream 

was then neutralized and transferred to one of the waste tanks. 

This neutralized waste stream is known as high-heat waste. Another waste stream from the PUREX 

process resulted from stripping and recovery of nitric acid from first and second uranium purification 

cycles. This waste stream is known as low-heat waste. These two waste streams made up the bulk of 

the waste managed in FTF. [DPSPU 77-11-1] 

2.2.1.2 Evaporator Products 

The two FTF evaporators 

concentrated the liquid waste 

generated in the PUREX 

process to remove excess water, 

a process which generated 

concentrated supernate and 

evaporator overheads. As 

discussed previously, the 

saltcake is precipitated salt 

waste from the concentrated 

supernate and is comprised 

principally of inert salts such as 

nitrites and nitrates. Figure 2.2-

1 shows the saltcake at the 

bottom of a waste tank. The 

concentrated supernate is in an 

alkaline solution that is relatively 

high in radioactivity. Overheads 

are the excess water removed from the waste stream by the evaporation process. Overheads are 

sampled and, depending on the sample results, are sent to the Effluent Treatment Facility (ETF) or, if 

cesium levels are high, returned to a waste tank to repeat the evaporation process. Additional cesium 

removal is also part of the process at ETF. 

Figure 2.2-1: Saltcake at Bottom of Waste Tank
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2.2.1.3 Zeolite Resin 

Some waste tanks contain zeolite resin from ion exchange columns, which are also known as cesium 

removal columns. Tanks 19, 25 and 27 were equipped with these columns to remove cesium from the 

242-F and 242-16F evaporator overheads waste streams, with the 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. As discussed in the individual 

waste tank histories in Section 2.2.2.2, Tanks 7, 18, 19, 25 and 27 contain zeolite resin. [CBU-PIT-2005-

00099] 

2.2.1.4 Miscellaneous Waste Streams 

As described below, some FTF waste tanks have also received limited amounts of radioactive waste from 

other sources such as HTF, which receives waste from the H-Modified (HM) process in H Canyon, and 

SRNL, which analyzes process samples and tritiated water originating from the K-Area Reactor. 

2.2.2 Waste Management 

This section addresses processes used for waste storage and volume reduction and provides a capsule 

history of each FTF waste tank. 

2.2.2.1 Waste Storage and Volume Reduction 

As discussed previously, the primary function of FTF has been to support the F-Canyon operation by 

storing waste produced from the PUREX process. Management of waste stored in FTF has been 

complex due to the large volumes of waste produced by F-Canyon operations, efforts to reduce the 

volume of waste and waste removal from waste tanks. Because the PUREX process was based on a 

nitric acid flowsheet and because the waste tanks in FTF were constructed of carbon steel, the pH of the 

waste originating from F Canyon was chemically adjusted from an acidic solution to an alkaline solution 

through the addition of sodium hydroxide prior to transfer to FTF to protect the integrity of carbon steel 

waste tanks. 

The volume of waste managed in FTF has been significantly reduced using the FTF evaporator systems. 

The 242-F Evaporator System operated from 1960 through 1988. The 242-16F Evaporator Facility 

entered service in 1980 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 F 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. 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 1 was constructed during 1952 and 1953 and entered service in 1954 as an F-Canyon waste 

receipt tank. This waste tank remained active and operational until 1973. Since then, Tank 1 has not 

received a waste transfer. The largest volume of waste stored in Tank 1 has been approximately 748,000 

gallons. On March 31, 2010, Tank 1 contained approximately 480,000 gallons of saltcake, 2,700 gallons 

of supernate and 7,000 gallons of sludge. [DPSPU 78-11-8, SRR-LWP-2010-00040] Tank 1 received 

only high-heat waste initially; however, during 1962, it received low-heat waste. After 1962, the 

remainder of total waste was high-heat waste, which was last received in 1967. To prepare the waste 
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tank for salt waste storage service, sludge removal was performed in 1969. Tank 1 became an 

evaporator concentrate receipt tank receiving concentrated supernate from the 242-F evaporator in 1969. 

This service concluded in 1973. Since then, Tank 1 has served as a salt waste storage tank. [DPSPU 

78-11-8, WSRC-TR-93-425] Tank 1 is known to have leaked and a small quantity of waste was observed 

on the annulus floor. However, the exact location of the leak site 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-2010-00283] 

Tank 2 was constructed during 1952 and 1953 and entered service in 1955 as an F-Canyon waste 

receipt tank. This waste tank remained active and operational until 1973. Since then, Tank 2 has not 

received a waste transfer. The largest volume of waste stored in Tank 2 has been approximately 730,000 

gallons. On March 31, 2010, Tank 2 contained approximately 536,000 gallons of saltcake and 4,000 

gallons of sludge. [DPSPU 83-11-5, SRR-LWP-2010-00040] Tank 2 received high-heat waste for less 

than a year with the final receipt in 1956. To prepare the waste tank for salt waste storage service, 

sludge removal was performed in 1966. In 1967, Tank 2 became an evaporator concentrate receipt tank, 

receiving concentrated supernate from the 242-F evaporator. This service concluded in 1973. Since 

then, Tank 2 has served as a salt waste storage tank. [DPSPU 83-11-5, WSRC-TR-93-425] Tank 2 has 

no known leak sites. [SRR-STI-2010-00283] 

Tank 3 was constructed during 1952 and 1953 and entered service in 1956 as an F-Canyon waste 

receipt tank. This waste tank remained active and operational until 1974. Since then, Tank 3 has not 

received additional waste. The largest volume of waste stored in Tank 3 at any time has been 

approximately 718,000 gallons. On March 31, 2010, Tank 3 contained approximately 536,000 gallons of 

saltcake and 4,000 gallons of sludge. [DPSPU 83-11-9, SRR-LWP-2010-00040] Tank 3 served as an FCanyon waste receipt tank for high-heat waste through 1961 when it received its last direct receipt from F 

Canyon. To prepare the waste tank for salt waste storage service, sludge removal was performed in 

1968. In 1968, Tank 3 became an evaporator concentrate receipt tank, receiving concentrated supernate 

from the 242-F evaporator. This service concluded in 1974. A brief interstitial liquid removal campaign 

was performed in 2003. [CBU-SPT-2004-00099] Since then, Tank 3 has served as a salt waste storage 

tank. [DPSPU 83-11-9, WSRC-TR-93-425] Tank 3 has no known leak sites. [SRR-STI-2010-00283] 

Tank 4 was constructed during 1952 and 1953 and entered service in 1961 as an F-Canyon waste 

receipt tank. This waste tank remained active and operational until 1981. Since then, Tank 4 has not 

received waste. The largest volume of waste stored in Tank 4 has been approximately 728,000 gallons. 

[DPSPU 80-11-9] On March 31, 2010, Tank 4 contained approximately 32,000 gallons of sludge and 

353,000 gallons of supernate. [SRR-LWP-2010-00040] Tank 4 received high-heat waste during two 

periods. The first began in 1961 and concluded in 1967. The second began in 1974 and concluded in 

1980. Between the periods, Tank 4 was used as an evaporator concentrate receipt tank, receiving 

concentrated supernate from the 242-F evaporator. This service concluded in 1981. In 2007, a hard 

mineral layer called burkeite that existed above the sludge in the waste tank was dissolved and 

transferred out of the waste tank to prepare for sludge removal activities. [DPSPU 80-11-9, LWO-LWE2007-00190, WSRC-TR-93-425] In 2009, approximately 136,000 gallons of sludge waste was transferred 

to HTF Tank 51. [SRR-LWP-2010-00007] Tank 4 has no known leak sites. [SRR-STI-2010-00283] 

Tank 5 was constructed during 1952 and 1953 and entered service in 1959 as an F-Canyon waste 

receipt tank. This waste tank remained active and operational until 1973. The largest volume of waste 

stored in Tank 5 has been approximately 730,000 gallons. [DPSPU 80-11-9] On March 31, 2010, Tank 5 

contained approximately 3,300 gallons of sludge and approximately 2,100 gallons of supernate. [SRRLWP-2010-00040] Tank 5 received high-heat waste through 1969 when it last received direct receipt 

from F Canyon. Tank 5 also received approximately 106,000 gallons of waste described as “high 

chloride” from SRNL in 1968. From 1971 to 1973, Tank 5 supported 242-F evaporator operations by 

receiving and transferring concentrated supernate. In 2001, approximately 266,000 gallons of liquid was 

transferred from H Area to Tank 5 to create space in support of H Canyon and DWPF operations. Leak 

sites were identified and waste was removed below the lowest known leak site after the external primary 

waste tank wall was inspected in 2001. [DPSPU 81-11-10, WSRC-TR-93-425, HLW-2002-00025, ESHEPG-2006-00005] Bulk waste removal and mechanical cleaning using submersible mixer pumps (SMPs) 

performed from 2005 to 2008 and left approximately 3,500 gallons of residual solids. [M-ESR-F-00147] 

A chemical cleaning campaign using oxalic acid, completed in 2008, reduced the residual solids volume 
Draft Basis for Section 3116 Determination DOE/SRS-WD-2010-001 

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at the Savannah River Site September 30, 2010 

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to approximately 3,300 gallons. [M-ESR-F-00160] Additional cleaning in Tank 5 is planned for 2010. 

Inspections of Tank 5 through 2009 identified 34 leak sites and a small quantity of waste (less than 10 

gallons) has been observed on the annulus floor. 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-2010-00283] 

Tank 6 was constructed in 1952 and entered service in 1964 as an F-Canyon waste receipt tank. This 

waste tank remained active and operational until 1974. The largest volume of waste stored in Tank 6 at 

any one time has been approximately 732,000 gallons. [DPSPU 81-11-4] On March 31, 2010, Tank 6 

contained approximately 3,500 gallons of sludge and approximately 400,000 gallons of supernate. [SRRLWP-2010-00040] Tank 6 received high-heat waste through 1972 when it received its last direct receipt 

from F Canyon. Tank 6 also received approximately 194,000 gallons of high-heat waste from SRNL in 

1970. From 1972 through 1973, Tank 6 received and transferred over 1,000,000 gallons of high-heat 

waste supernate to support 242-F evaporator operations. In 1997, Tank 6 received approximately 

280,000 gallons of material to support the Tank 17 closure activities. In 2001, approximately 315,000 

gallons of liquid were transferred from H Area to Tank 6 to create space in support of DWPF operations. 

[DPSPU 81-11-4, WSRC-TR-93-425, HLW-2002-00025, ESH-EPG-2007-00016] Bulk waste removal and 

mechanical cleaning using SMPs, performed from 2006 to 2007, left approximately 6,000 gallons of 

residual solids. [M-ESR-F-00132] A chemical cleaning campaign using oxalic acid, completed in 2009, 

reduced the residual solids volume to approximately 3,500 gallons. [M-ESR-F-00165] Additional 

cleaning is planned for 2010. Tank 6 is known to have leaked at 11 identified leak sites and 

approximately 90 gallons of waste has been observed on the annulus floor. 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-2010-00283] 

Tank 7 was constructed during 1952 and 1953 and entered service in 1954 as an F-Canyon waste 

receipt tank until 1968. This waste tank remains active and operational. The largest volume of waste 

stored in Tank 7 has been approximately 733,000 gallons. On March 31, 2010, Tank 7 contained 

approximately 133,000 gallons of sludge and 263,000 gallons of supernate. [WSRC-TR-93-425, SRRLWP-2010-00040] From 1960 and intermittently through 1973, Tank 7 also received approximately 

402,000 gallons of non-canyon waste, which consisted of mostly heat exchanger and sand filter flushes 

from other SRS areas. Because the transfer jets for Tanks 1 through 6 and Tank 8 discharged into Tank 

7, many transfers were made into and out of this waste tank. From 1960 to 1976, Tank 7 supported 242-

F evaporator operations by receiving and transferring concentrated supernate from other Type I tanks to 

the Tank 18 evaporator feed tank. In 1974, the first inter-area waste transfer took place in which highheat waste from HTF was transferred to FTF Tanks 7 and 18. Tank 7 directly fed the 242-F evaporator 

between 1977 and 1984. Tank 7 material was transferred to Tank 26, from 1985 to 1987, to support 242-

16F evaporator operations. In 2003, Tank 7 supported Tank 18 waste removal and, in 2004, sludge 

waste was transferred from Tank 7 to HTF Tank 51. In 2003, a transfer of Tank 18 material introduced 

zeolite into Tank 7. The transfer system design places Tank 7 along the conventional transfer route to 

exit the Tank 1 through Tank 8 area. From 2005 to 2010, Tank 7 supported the waste removal 

operations for Tanks 4, 5, 6, 8, 18 and 19. [WSRC-TR-93-425, HLW-2002-00025, LWO-PIT-2007-00083, 

LWO-LWP-2008-00007, SRR-LWP-2010-00007] Tank 7 has no known leak sites. [SRR-STI-2010-

00283] 

Tank 8 was constructed during 1952 and 1953 and entered service in 1956 as an F-Canyon waste 

receipt tank. This waste tank remained active and operational until 1981. The largest volume of waste 

stored in Tank 8 has been approximately 732,000 gallons. On March 31, 2010, Tank 8 contained 

approximately 20,000 gallons of sludge and 431,000 gallons of supernate. [WSRC-TR-93-425, SRRLWP-2010-00040] Tank 8 received only low-heat waste from 1956 through 1959. In 1960, Tank 8 began 

receiving high-heat waste, which was last received in 1974. In 1974, the remainder of the total waste 

received was low-heat waste, which was last received in 1981. During 1968 to 1970, Tank 8 also 

received approximately 436,000 gallons of waste from SRNL. Bulk waste removal operations were 

performed in 2000 through 2001. Sludge from Tank 8 went into a sludge batch for feed to DWPF [UESR-F-00009] In 2004, Tank 8 underwent sludge heel removal operations. [WSRC-TR-93-426, CBUENG-2005-00005] Supernate was transferred into and out of Tank 8 to support waste removal in Tank 4, 

5, 6, 18 and 19 from 2005 to 2009. Tank 8 stored liquid from Tank 4 in 2007 to support burkeite removal 
Draft Basis for Section 3116 Determination DOE/SRS-WD-2010-001 

for Closure of F-Tank Farm Revision 0 

at the Savannah River Site September 30, 2010 

Page 2-53 

from Tank 4. Aluminum-rich leachate was transferred to Tank 8 from Tank 51 in support of lowtemperature aluminum dissolution for sludge processing preparation in 2009. [LWO-PIT-2006-

00063_Redacted, LWO-PIT-2007-00083, LWO-LWP-2008-00007, SRR-LWP-2010-00007] Tank 8 has 

no known leak sites. [SRR-STI-2010-00283] 

Type IV Tanks 

Tank 1713 was constructed in 1958 and entered service in 1961 as an F-Canyon waste receipt tank. This 

waste tank remained active and in operational service until 1982; it is now closed and filled with grout. 

The largest volume of waste stored in Tank 17 was approximately 1,300,000 gallons. [WSRC-TR-2004-

00284, ESH-EPG-2008-00012, WSRC-TR-93-425] Tank 17 was an F-Canyon low-heat waste receipt 

tank during two periods. The first period began in 1961 and concluded in 1962. The second period, as 

an F-Canyon waste receipt tank, began in 1969 and concluded in 1977. In between the two periods, 

Tank 17 was used as an evaporator concentrate receipt tank, receiving concentrated supernate from the 

242-F evaporator from 1964 to 1966. In 1977, at the end of its operations as a canyon waste receipt 

tank, Tank 17 received approximately 198,000 gallons of concentrated salt solution from the 242-F 

evaporator. From 1974 and to 1982, Tank 17 received approximately 177,000 gallons of non-canyon 

waste from various other SRS areas. In 1983, bulk waste removal operations began and then concluded 

early in 1986 with a spray washing campaign. In 1992, approximately 100,000 gallons of tritiated water 

from the K-Reactor Area was placed in Tank 17. Final waste removal in support of waste tank closure 

was completed in 1997 when Tank 17 was filled with grout and closed. The SCDHEC approved closure 

of Tank 17 on December 15, 1997. [PIT-MISC-0004, WSRC-TR-93-425, PIT-MISC-0016] Tank 17 had 

no history of leaks. 

Tank 18 was constructed in 1958 and entered service in 1959 as an F-Canyon waste receipt tank. This 

waste tank remained active and operational until 1986 when waste removal activities were initiated. The 

largest volume of waste stored in Tank 18 has been approximately 1,300,000 gallons. On March 31, 

2010, Tank 18 contained approximately 4,000 gallons of supernate and 4,000 gallons of sludge. [WSRCTR-2004-00284, CBU-PIT-2004-00024, SRR-LWP-2010-00040] Between 1959 and 1977, Tank 18 

received low-heat waste directly from F Canyon during multiple periods. Tank 18 also supported the 242-

F evaporator operations, as both a receiver of concentrated supernate and overheads and as a feed tank 

for the evaporator. From 1962 to 1981, Tank 18 received concentrated supernate and, from 1966 to early 

1983, overheads from the 242-F evaporator. From 1960 through 1976, Tank 18 was used as a feed tank 

for the 242-F evaporator. In 1973, Tank 18 also received approximately 12,000 gallons of waste from 

HTF evaporator overheads and, in 1974, approximately 719,000 gallons of high-heat waste from HTF 

were received. Tank 18 was designed as the sole conventional transfer route to exit the Tank 17 through 

Tank 20 area. In 1980 and 1981, Tank 18 received salt and/or sludge removal waste from Tanks 17, 19 

and 20 waste removal activities. Throughout this period, Tank 18 served as a hub for Tanks 17 through 

20 activities. During this operation, some of the zeolite resins, which were confined to Tank 19, were 

transferred to, and settled in, Tank 18. [CBU-PIT-2005-00124] Mechanical heel removal occurred in 

Tank 18 in 2003 using a large mixer pump, the Advanced Design Mixer Pump (ADMP). Additional heel 

removal was performed in 2009 using a mechanical crawler-based ultra-high-pressure eductor retrieval 

system (Mantis) that reduced the volume of residual solids to approximately 4,000 gallons. [U-ESR-F00041] Tank 18 has no known leak sites. [SRR-STI-2010-00283] 

Tank 19 was constructed in 1958 and entered service in 1961 as an F-Canyon cascade waste receipt 

tank. This waste tank remained active and in operational service until 1980 when waste removal 

activities were initiated. The largest volume of waste stored in Tank 19 has been approximately 

1,300,000 gallons. On March 31, 2010, Tank 19 contained approximately 2,000 gallons of sludge and 

less than 830 gallons of supernate. [WSRC-TR-93-425, CBU-PIT-2005-00124, SRR-LWP-2010-00040] 

Tank 19 initially received fresh canyon waste from Tank 17, the primary F-Canyon waste receipt tank, to 

create space for Tank 17 to directly receive additional canyon waste. Therefore, Tank 19 was considered 

to be a cascade waste receipt tank. Tank 19 served this function by receiving low-heat waste in the first 

six months of operation and again from 1969 to 1972. In 1974, Tank 19 received a relatively small 

volume of low-heat waste directly from F Canyon. From 1962 to 1976, Tank 19 served as an evaporator 

 

13 Tank 17 and Tank 20 were grouted and removed from service in 1997 and are not part of this Draft FTF 3116 Basis Document. 
Draft Basis for Section 3116 Determination DOE/SRS-WD-2010-001 

for Closure of F-Tank Farm Revision 0 

at the Savannah River Site September 30, 2010 

Page 2-54 

concentrate receipt tank for the 242-F evaporator. Tank 19 also received approximately 13,000 gallons of 

spent zeolite resin from a cesium removal column that was mounted on a Tank 19 riser to remove cesium 

from the evaporator overheads. In 1980, Tank 19 salt removal began and concluded in 1981. In 2000 

and 2001, mechanical heel removal was performed using SMPs. [CBU-PIT-2005-00124] Subsequently, 

additional heel removal was performed in 2008 and 2009 using a Mantis that reduced the volume of 

residual solids to approximately 2,000 gallons. [U-ESR-F-00042] Tank 19 has a history of in-leakage. 

Visual inspections have revealed two sites where in-leakage occurred prior to 1994. Two additional inleakage sites where documented in 2009. [SRR-STI-2010-00283] 

Tank 20 was constructed in 1958 and entered service in 1960 as a concentrate receipt tank for the 242-F 

evaporator. This waste tank remained active and operational until 1977; it is now closed and filled with 

grout. The largest volume of waste stored in Tank 20 was approximately 1,300,000 gallons. [WSRC-TR93-425, ESH-EPG-2008-00012, DPSPU 82-11-10] Tank 20 began its operational life as the 242-F 

evaporator concentrate receipt tank from 1960 and to 1976. During this time, Tank 20 received a 

combination of evaporator concentrate from both low-heat and high-heat waste processing. Tank 20 also 

received approximately 264,000 gallons of condensation water from the concentrate transfer system and 

flushes of the concentrate transfer system pump pit. The first saltcake removal campaign began in 1980 

and concluded in 1982. In 1986, the second salt removal campaign occurred. In 1988, Tank 20 was 

spray washed. In 1996, Tank 20 was filled with grout and closure operations concluded. The SCDHEC 

approved closure of Tank 20 on July 31, 1997. [DPSPU 82-11-10, PIT-MISC-0002, PIT-MISC-0018] 

Tank 20 had a history of in-leakage only. [SRR-STI-2010-00283, SRR-LWP-2010-00007] 

Type III Tanks 

Tank 33 was constructed in 1969 and entered service in 1974 as a concentrate receipt tank used, 

primarily for the 242-F evaporator, but also with limited use for the 242-16F evaporator. The largest 

volume of waste stored in Tank 33 has been approximately 1,270,000 gallons. As of March 31, 2010, 

Tank 33 contained approximately 294,000 gallons of saltcake, 659,000 gallons of supernate and 80,000 

gallons of sludge. [SRR-LWP-2010-00040] Through 1984, Tank 33 received mainly concentrated salt 

solution from the 242-F evaporator and a small volume from the 242-16F evaporator. From 1983 to 

2003, Tank 33 received both high-heat and low-heat waste directly from F Canyon. In 2006 and 2007, 

supernate was removed from Tank 33 to create space to receive dissolved burkeite salt solution from 

Tank 4. In 2007, Tank 33 received approximately 200,000 gallons of dissolved burkeite salt solution. In 

2008 and 2009, approximately 400,000 gallons of supernate were transferred from Tank 33 to Tank 7 to 

neutralize oxalic acid used during chemical cleaning of Tanks 5 and 6. Tank 33 was fitted with 

deployable cooling coils in 1976. [WSRC-TR-93-425, LWO-PIT-2007-00083, SRR-LWP-2010-00007] 

Tank 34 was constructed in 1972 and entered service in 1973 as a concentrate receipt tank for the 242-F 

evaporator. The largest volume of waste stored in Tank 34 has been approximately 1,298,000 gallons. 

As of March 31, 2010, Tank 34 contained approximately 191,000 gallons of saltcake, 902,000 gallons of 

supernate and 12,600 gallons of sludge. [WSRC-TR-93-425, SRR-LWP-2010-00040] Through 1985, 

Tank 34 received concentrated salt solution from the 242-F evaporator. From 1980 to 1995, Tank 34 

received high-heat and low-heat waste directly from F Canyon. Tank 34 also received waste solution 

from the HTF in 2001 and 2006. In 2006 and 2007, Tank 34 received interstitial liquid from Tank 25. In 

2008, approximately 270,000 gallons of supernate were transferred from Tank 34 to Tank 7 to neutralize 

oxalic acid used during chemical cleaning of Tanks 5 and 6. Tank 34 was fitted with deployable cooling 

coils in 1976. [LWO-PIT-2007-00088, WSRC-TR-93-425, SRR-LWP-2010-00007] 

Type IIIA Tanks 

Tank 25 was constructed in 1978 and entered service in 1980 as a concentrate receipt tank for the 242-F 

evaporator. This tank continues to operate. The largest volume of waste stored in Tank 25 has been 

approximately 1,280,000 gallons. On March 31, 2010, Tank 25 contained approximately 506,000 gallons 

of saltcake and 306,000 gallons of supernate. [CBU-SPT-2004-00081, SRR-LWP-2010-00040] Tank 25 

began its operational life of as the 242-F evaporator concentrate receipt tank from 1980 to 1988. The 

tank was equipped with a cesium removal column used to remove cesium from the evaporator 

overheads. Through 1988 approximately 1,050 gallons of zeolite resin used in the column were 

discharged into the tank. From 1988 through 1991, only transfers out of Tank 25 occurred. From 1991 to 

2005, Tank 25 was inactive. [CBU-SPT-2004-00081, CBU-PIT-2005-00099, WSRC-TR-93-425] In 2005, 
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for Closure of F-Tank Farm Revision 0 

at the Savannah River Site September 30, 2010 

Page 2-55 

saltcake removal operations were initiated to allow Tank 25 to serve as a concentrate receipt tank for the 

242-16F evaporator and the supernate above the salt layer was removed. In 2006 and 2007, interstitial 

liquid was removed and transferred to Tank 34. Salt dissolution activities were performed in 2008 and 

2009, and Tank 25 was converted to the 242-16F evaporator concentrate receipt tank in 2010. [LWOPIT-2007-00039, CBU-SPT-2004-00081, SRR-LWE-2010-00027] 

Tank 26 was constructed in 1978 and entered service in 1980 as a feed tank for the 242-16F evaporator. 

The largest volume of waste stored in Tank 26 has been approximately 1,274,000 gallons. On March 31, 

2010, Tank 26 contained approximately 876,000 gallons of supernate and 267,000 gallons of sludge. 

[SRR-LWP-2010-00040, WSRC-TR-93-425] Tank 26 has served as a feed tank for the 242-F and 

242-16F. Because the 242-F evaporator is no longer in service, this waste tank currently feeds only the 

242-16F evaporator. Tank 26 received concentrated supernate, low-heat waste and overheads from the 

evaporators. Tank 26 also received supernate from the HTF, some of which originated from the 

Receiving Basin for Off-site Fuel, which was fed to the 242-16F evaporator. [WSRC-TR-93-425, SRRLWP-2010-00007] 

Tank 27 was constructed in 1978 and entered service in 1980 as a receipt tank for salt wastes and 

concentrated supernate. The largest volume of waste stored in Tank 27 has been approximately 

1,274,000 gallons. As of March 31, 2010, Tank 27 contained approximately 1,160,000 gallons of 

saltcake, 87,000 gallons of supernate and 3,900 gallons of sludge. [SRR-LWP-2010-00040, WSRC-TR93-425] From 1980 to 1984, Tank 27 received low-heat waste supernate, salt waste and dissolved salt 

solution from Tanks 7 and 18. From 1983 through 1988, Tank 27 served as an evaporator concentrate 

receipt tank for the 242-F evaporator. It was equipped with a cesium removal column and served as an 

evaporator concentrate receipt tank for the 242-16F evaporator from 2004 to 2009. Since then, Tank 27 

has served as a salt waste storage tank. Approximately 5,000 gallons of zeolite resin used in the cesium 

removal column was discharged into the tank through 1989. During 2004 and 2006, Tank 27 also 

received waste solutions from the HTF to feed the 242-16F evaporator. [WSRC-TR-93-425, CBU-PIT2005-00099, SRR-LWP-2010-00007] 

Tank 28 was constructed in 1978 and entered service in 1980 as an evaporator concentrate receipt tank 

for the 242-16F evaporator. The largest volume of waste stored in Tank 28 has been approximately 

1,266,000 gallons. As of March 31, 2010, Tank 28 contained approximately 1,030,000 gallons of saltcake 

and 195,000 gallons of supernate. [CBU-PIT-2005-00051, SRR-LWP-2010-00040] From 1980 until 

1985, Tank 28 served as an evaporator concentrate receipt tank for the 242-16F evaporator and, in 1980, 

also received concentrated supernate from Tank 18. Transfers out of Tank 28 were made until 1988. 

Since then, Tank 28 has not received a waste transfer. [CBU-PIT-2005-00051, WSRC-TR-93-425] 

Tank 44 was constructed in 1980 and entered service in 1982 as an evaporator concentrate receipt tank 

for the 242-16F evaporator. The largest volume of waste stored in Tank 44 has been approximately 

1,273,000 gallons. As of March 31, 2010, Tank 44 contained approximately 1,010,000 gallons of saltcake 

and 244,000 gallons of supernate. [CBU-PIT-2006-00101, SRR-LWP-2010-00040] From 1982 through 

1992, Tank 44 received concentrated salt solution from the 242-16F evaporator. Supernate was 

transferred from Tank 44 to Tank 26 from 1982 to 1992 and again in 2004. Since then, Tank 44 has not 

received a waste transfer. [CBU-PIT-2006-00101, SRR-LWP-2010-00007] 

Tank 45 was constructed in 1980 and entered service in 1982 as an evaporator concentrate receipt tank 

for the 242-16F evaporator. The largest volume of waste stored in Tank 45 has been approximately 

1,269,000 gallons. As of March 31, 2010, Tank 45 contained approximately 1,100,000 gallons of saltcake 

and 144,000 gallons of supernate. [CBU-PIT-2006-00044, SRR-LWP-2010-00040] From 1982 to 1993, 

Tank 45 received concentrated salt solution from the 242-16F evaporator and supernate was transferred 

from Tank 45 to Tank 26 from 1983 to 1993. Since then, Tank 45 has not received a waste transfer. 

[WSCR-TR-93-425, CBU-PIT-2006-00044, SRR-LWP-2010-00007] 

Tank 46 was constructed in 1980 as a spare tank and entered service in 1994 to support the 242-16F 

evaporator. The largest volume of waste stored in Tank 46 has been approximately 1,249,000 gallons. 

As of March 31, 2010, Tank 46 contained approximately 1,160,000 gallons of saltcake and 96,700 gallons 

of supernate. [SRR-LWP-2010-00040] Tank 46 received concentrated salt solution from the 242-16F 

feed tank in 1994 and supernate was transferred from Tank 46 to Tank 26 from 1994 to 1995, and from 
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at the Savannah River Site September 30, 2010 

Page 2-56 

2003 to 2005. Since then, Tank 46 has not received a waste transfer. [WSRC-TR-93-425, SRR-LWP2010-00007] 

Tank 47 was constructed in 1980 and entered service in 1981 as a waste receipt tank for fresh F-Canyon 

waste and concentrated salt solution from evaporator operations. The largest volume of waste stored in 

Tank 47 has been approximately 1,272,000 gallons. As of March 31, 2010, Tank 47 contained 

approximately 773,000 gallons of saltcake, 209,000 gallons of supernate and 248,000 gallons of sludge. 

[WSRC-TR-93-425, SRR-LWP-2010-00040] From 1981 to 1986, Tank 47 received low-heat waste from 

F Canyon. From 1983 to 1988, the tank also received concentrated salt solution from the 242-16F 

evaporator. In 1981, Tank 47 received salt removal waste from Tank 18, in 1987 concentrated supernate 

from Tank 26, and from 1982 to 1991 Tank 47 received approximately 280,000 gallons of non-canyon 

low-level waste from other SRS areas. Tank 47 served as the 242-16F evaporator concentrate receipt 

tank in 2008 and 2009. [SRR-LWP-2010-00007] 

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, 

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 chemically treated water 

must be added. Type III/IIIA tank space must be available for receiving and adjusting these solutions. 

[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 that could not be removed by mechanical methods and water addition alone. 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). 

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 will be removed 

as appropriate from the annulus as described in Section 2.3.6. 

Waste tank cleaning is subject to a variety of operating constraints including: 

 maintaining emergency tank space, 

 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, 

 preparing waste for downstream waste treatment facilities, and 

 meeting safety basis requirements. 

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-LWP-2009-00001] See Section 1.3 for discussion on the closure schedule. 
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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 

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 submerged in 

the 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 spinning the jets in the horizontal 

plane. 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 so the entire waste tank 

contents are slurried. 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. 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 FTF tanks in which this 

technique has been successfully used for bulk sludge 

removal include Tanks 8, 17 and 18. [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 FTF, this technique has been successfully used for bulk salt 

removal in Tank 19. [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 water, which is water chemically treated to prevent 

corrosion of the carbon steel waste tanks, until the saltcake surface is flooded. The water dissolves a 

Figure 2.3-1: Typical Standard Slurry 

Pump 

[WSRC-TR-2001-00313] 
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Page 2-58 

portion of the saltcake forming a salt solution. The salt solution is then transferred out of the tank. In 

FTF, this process has been used for saltcake removal in Tank 20. [HLW-2002-00025] 

Processing the salt waste requires waste tank space. To make the salt mobile for transferring and to 

adjust the salt concentrations (molarity) for processing and disposal, significant quantities of chemically 

treated water 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 FTF waste tanks may continue to be used in waste removal activities. 

In addition, a new generation of waste removal equipment has been developed as described in the 

following subsections. 

2.3.2.1 Submersible Mixer Pumps - Type I and III/IIIA Waste 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. [GESR-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 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)14. [GESR-G-00051] 

 

14 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 tank configuration and waste characteristics. 

Figure 2.3-2: Submersible Mixer Pump Waste Removal Diagram 
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To date, the SMPs have been used to perform bulk waste removal on Tanks 4, 5 and 6 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 hp motor that can operate up to 1,600 rpm. 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. [M-CLC-G-00344] 

Waste transfers during Tanks 4, 5 and 6 bulk waste removal and Tanks 5 and 6 mechanical cleaning 

were achieved by using a STP. The STP is a 15 hp, 3,450 rpm, 250 gpm 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. 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. 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 

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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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 optimal 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 Salt Waste 

Processing Facility [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 planned for heel removal in FTF waste tanks. 

However, other cleaning solutions or methods may be used if they have comparable or greater 

effectiveness. The planned chemical cleaning methods are similar to bulk oxalic acid cleaning methods 

used in Tanks 5, 6 and HTF Tank 1615, consisting of several oxalic acid strikes, use of agitation to 

facilitate particle-acid contact and a final clean water spray wash and mixing operation. If needed, oxalic 

acid will 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 Saltstone Disposal Facility16. Therefore, the oxalic acid 

flowsheet evaluations have considered the effects of oxalates on DWPF and salt processing to determine 

the quantities of oxalic acid that can be tolerated in 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. 

Enhanced Chemical Cleaning (ECC) will also be used to minimize downstream impacts of bulk oxalic acid 

cleaning. The ECC system consists of three distinct process steps: 1) dissolution/removal, 2) oxalic acid 

decomposition and 3) solid/liquid separation (Figure 2.3-5)17. In this system, waste is dissolved using 

dilute oxalic acid and transferred in batches to the oxalate decomposition process. In the decomposition 

process ozone is used to decompose the oxalate. As the oxalate is decomposed, the solids reform 

primarily as oxides, creating a dilute slurry. This dilute slurry is oxidized until the oxalate concentration 

falls below the target. When the oxalate concentration is low enough, the decomposition loop is stopped 

and the dilute slurry is fed to a settling tank (i.e., FTF waste tank) with the liquid fraction undergoing 

 

15 Discussion on Tank 16 cleaning is included for information regarding use of the oxalic acid cleaning process; however, Tank 16 is 

located in the HTF at SRS and is not within the scope of this Draft FTF 3116 Basis Document. 

16 See Appendix A for a brief description of DWPF, SWPF and Saltstone Disposal Facility operations. 17 The ECC process is still in the stage of being tested and designed by DOE. The figure depicted represents a conceptual diagram 

of the ECC process and is subject to change as development of the process advances. 
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dissolved 

sludge

Waste Tank Undergoing Cleaning Type III/IIIA Waste Tank

ozone

Oxalate

Decomposition Dissolution

dilute 

slurry

ECC Module

Ozone 

Generation 

Module

Solid-Liquid 

Separation

Oxalic Acid 

6.5 wt%

Primary 

Ventilation 

Module

Cooling 

Water 

Module

Secondary 

Ventilation 

Module Mixer(s) 

De-ionized 

Water Unit 

subsequent evaporation. This process is repeated until the tank sludge heel has been removed. [LWO2008-0056] 

2.3.4 Mantis - Type IV Waste Tanks 18 and 19 

Tanks 18 and 19 have each undergone bulk waste removal using slurry pumps and mechanical heel 

removal. Bulk waste removal was performed in Tank 18 during 1986 and 1987; approximately 515,000 

gallons of sludge was removed. [CBU-PIT-2005-00233] Mechanical heel removal occurred in Tank 18 in 

2003 using a large mixer pump, the ADMP. The Tank 18 interior tank walls were spray washed following 

mechanical heel removal. Bulk waste removal was performed in Tank 19 during 1980 and 1981; 

approximately 1,033,000 gallons of saltcake was removed. [DPSP-84-17-7] Mechanical heel removal 

occurred during 2000 and 2001 using rotating submersible jet mixer pumps. The principal difficulty of the 

Tank 19 heel removal efforts was the characteristics of zeolite. Zeolite has proven to be resistant to 

removal because of its significantly higher settling velocity when compared to that of metal hydroxides 

and insoluble salts (i.e., traditional sludge waste). Zeolite is also resistant to dissolving with acids. 

[WSRC-TR-2002-00288] 

As a result of the March 2006 DOE-sponsored Tank Cleaning Technical Exchange, a new waste tank 

tethered mechanical crawler-based cleaning technology was identified. The DOE has adapted and 

successfully used this new technology in the unobstructed Type IV tanks, Tanks 18 and 19. [CBU-PIT2006-00067] The cooling coils in Type I, Type III and Type IIIA tanks severely restrict the mobility of 

tethered mechanical crawlers and make deployment of this device impractical in obstructed tanks. This 

new cleaning device allowed additional removal of waste from Tanks 18 and 19 to a greater extent than 

the technologies available when waste removal was initially discontinued in 2003 and 2001, respectively. 

Figure 2.3-5. ECC Process Diagram 
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The cleaning device, called a Mantis, consisted of a mechanical crawler and an eductor assembly that 

made up a retrieval system utilizing an ultra-high-pressure water eductor (Figure 2.3-6). The process 

system consisted 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, aboveground hose-in-hose transfer lines, Waste Mixing Chamber (WMC) and support equipment. The device 

entered the tank through a 24-inch riser in a folded position. Once inside the tank, the device unfolded 

into its operational configuration. 

The Mantis was remotely driven around the waste tank floor by an operator located in the Control Center. 

A high pressure (maximum 10 thousand pounds per square inch) hydro-lance at its front was used to 

break up waste mounds and an eductor was used to aspirate waste from the floor of the waste tank. 

Operators monitored the operation of the Mantis using in-tank lighting, cameras and video surveillance 

equipment. The waste traveled through the eductor in-tank waste retrieval flexible hose and up into a tee 

spool piece located on top of the riser. 

The waste then traveled through above-ground transfer lines (one from Tank 18 and one from Tank 19) 

routed to the receipt tank. The above-ground transfer lines consisted of a “hose within a hose” design 

with appropriate shielding. The above-ground waste transfer lines terminated inside a WMC installed 

inside a riser on the receipt tank. The WMC contained particle size reduction equipment (an immersion 

mill). The immersion mill, located near the bottom of the WMC, size-reduced solid waste particles so that 

the particles could be more easily re-suspended in future waste removal activities. [WSRC-TR-2007-

00327] Heel removal in 2008 and 2009 using the Mantis in Tank 19 reduced the volume of residual 

solids to approximately 2,000 gallons. [U-ESR-F-00042] Heel removal operations with the Mantis in 

Tank 18 in 2009 reduced the volume of residual solids to approximately 4,000 gallons. [U-ESR-F-00041] 

Figure 2.3-6: Mantis 

In Tank 19 Prior to Cleaning

In Test Facility

In Tank 19 After Cleaning
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Leak Site

Ventilation

Duct

Secondary 

Tank Wall

Dried Salt 

Waste

Primary 

Tank Wall

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 needed. 

2.3.6 Annulus Cleaning 

Currently, three waste tanks have leak sites on the primary tank walls. None of the leaks from the 

primary tanks have resulted in waste escaping to the surrounding soils. Tank 1 is known to have leaked 

and a small quantity of waste has been observed on the annulus floor. The location of the leak site has 

not been identified. Inspections performed through 2009 have identified 34 leak sites in Tank 5 with a 

small quantity of waste (less than 10 gallons) on the annulus floor (Figure 2.3-7). Dried salt nodules and 

deposits have formed on the tank exterior wall. Tank 6 has 11 known leak sites with approximately 92 

gallons of dried waste on the annulus floor (Figure 2.3-8). Waste material in the annulus is expected to 

be readily dissolved and removed. In 2009, 75% of the Tank 5 exterior wall and 100% of the Tank 6 

exterior wall were inspected and dried deposits were cleaned using a magnetically mounted wall crawler 

with washing capabilities. [SRR-STI-2010-00283] 

Figure 2.3-7: Waste Deposits on Exterior Primary Tank 

Wall Before Cleaning (Tank 5) 
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Dried Salt Waste on 

Annulus Floor

Secondary 

Tank Wall

Primary 

Tank Wall

Ventilation 

Duct

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. 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. 

2.3.8 Ancillary Structures Cleaning 

Ancillary structures are described in Section 2.1.12. 

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. 

[WSRC-TR-2002-00403] 

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, which make up approximately 57% of the total linear 

footage of lines, are tested as part of the Structural Integrity Program. [S-TSR-G-00001] Most of the 

remaining lines that are no longer in service are part of the 242-F Evaporator System and the Tank 17 

Figure 2.3-8: Waste Deposits on Annulus Floor Before Cleaning 

(Tank 6) 
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through 20 system. 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 FTF 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 two evaporator systems in FTF, the 242-F Evaporator System 

and the 242-16F Evaporator System. The 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 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] 

The 242-16F evaporator cell and vessel remain operational. The final residual levels in the 242-16F 

Evaporator System 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. 

There are three stainless steel pump tanks in FTF: The FPT-1 serves as the inter-area pump tank, and 

both FPT-2 and FPT-3 previously received waste transfers from F Canyon. There is a single catch tank 

in FTF designed to collect drainage from FDB-1 and the Type I tank transfer line encasements. These 

stainless steel tanks are accessible for waste removal with existing technologies. 

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 minimal radionuclide 

contamination at the time of closure. 

2.4 Radionuclide Inventory in F-Tank Farm Facility Systems, Structures and Components 

Minimal quantities of residual material are expected to remain in FTF at the time of closure following 

waste tank and ancillary structures cleaning activities. Data developed for the FTF PA, Revision 118, 

provided the estimated FTF residual radionuclide inventory used in this Draft FTF 3116 Basis Document. 

[SRS-REG-2007-00002] 

Estimated radionuclide concentrations for residual material in FTF waste tanks at closure are determined 

by three methods: 

 sample analysis 

 process knowledge data maintained in a controlled database (WCS) 

 special analysis (SA) 

The inventory of each waste tank was used to establish the characterization of the residual material in the 

ancillary structures and was decayed to year 202019. The following subsections describe the process for 

estimating residual material inventory at closure for the FTF waste tanks and ancillary structures, as 

provided in the FTF PA, Revision 1. [SRS-REG-2007-00002] 

 

18 Unless otherwise noted, the information in Section 2.4 and subsequent Sections 2.5 and 2.6 is based on information developed in 

the FTF PA, Revision 1. As required by DOE Manual 435.1-1, maintenance of the FTF PA will include future updates to incorporate 

new information, update model codes, analysis of actual residual inventories, etc., as appropriate. [SRS-REG-2007-00002] 

19 The year 2020 corresponds to the year the FTF PA assumes, for the purposes of analysis, that a 100-year period of institutional 

control will begin. 
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A methodical approach was used to construct reasonably conservative estimates of FTF waste tank 

closure inventories to be used in PA modeling. Independent steps were developed to systematically 

construct the FTF tank inventories, with each step adjusting inventory either by tank or by radionuclide. 

The steps used in inventory development were as follows: [SRS-REG-2007-00002] 

1. The initial list of radionuclides to be included in the FTF 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] 

2. The contaminant screening process consisted of several steps to arrive at an appropriate list of 

54 isotopes to be included in the FTF waste tank closure inventory estimates to be used in the 

FTF 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 FTF inventory for the PA. 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 FTF waste production history information for the potential for a specific 

radionuclide. This criterion screened out radionuclides not present within the FTF 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. 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 FTF PA modeling initial inventory radionuclide number to 54. The screening 

process is described in detail in Section 3.3 of the FTF PA. [SRS-REG-2007-00002] 

3. Additional radionuclides were added to the list of radionuclides of concern based on the potential 

for activation products being present. These additions were based on the possibility that these 

radionuclides may have been created at various points of production in SRS history. These 

radionuclides are Cl-36, K-40, Nb-93m, Pd-107, Pt-193 and Zr-93. These six radionuclides were 

added to the list of radionuclides of concern at inventories equivalent to the sample analysis 

detection capability. Because minimal amounts of these radionuclides were created, setting the 

inventory for this radionuclides at the detection capability was reasonably conservative 

(approximated at 1.0E-03 Ci). 

4. The list of radionuclides to be included was then reduced based on individual isotope half-life. 

Radionuclides with a half-life of less than five years were removed from inventory estimates. This 

was based on the consideration that active institutional control over the closed waste tanks will be 

maintained for 100 years. Therefore, any radionuclides with less than a five-year half-life would 

decrease to insignificant levels due to radioactive decay during the institutional control period. 

This does not apply to radionuclides that are part of a decay chain with a parent with a greater 

than five-year half-life. The twelve radionuclides removed from further inventory estimates based 
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on the five-year half-life criteria are as follows: Bk-249, Ce-144, Cm-242, Cs-134, Eu-155, Na-22, 

Pm-147, Pr-144, Rh-106, Ru-106, Sb-126 and Te-125m. 

5. The radionuclides inventories developed for Revision 0 of FTF PA [SRS-REG-2007-00002] were 

used as a starting point for the inventory of each individual waste tank. The WCS database 

[DOE-WD-2005-001] was utilized in developing the FTF PA Revision 0 inventories. 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 (WAC). 

6. The waste tanks were binned according to waste tank type. 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. In addition, the 

tanks built within a specific time frame were built in close proximity to each other. 

7. Within each bin, the inventories were adjusted as applicable within that bin. 

Type I: An adjustment to the inventory estimates was made by multiplying by a factor of 10 to 

provide for greater uncertainties in the residual inventories from the planned waste removal 

operation. Based on recent waste removal activities in Tanks 5 and 6, there is uncertainty 

around the projected inventories for the waste tanks remaining to be cleaned. To account for 

this uncertainty and ensure that the PA would provide a reasonably conservative inventory 

projection, the existing inventories were multiplied by a factor of 10. 

For a majority of the radionuclides with an adjusted inventory less than one curie, the 

inventories were adjusted to either one curie or the analytical detection limit (1.0E-03 Ci) to 

allow for more efficient and cost effective means of confirming concentrations within residual 

materials for radionuclides with a limited potential impact to dose. For those radionuclides that 

have been observed (through previous analyses or scoping studies) to have greater potential 

impact on the overall dose, the inventory was adjusted to the analytical detection limit. The 

following radionuclides fit into this category: Ac-227, Cm-247, Cm-248, I-129, Nb-94, Pa-231 

and Pu-244. Those radionuclides with adjusted inventories greater than one curie were not 

adjusted in this step. 

The inventory estimates for both Ra-226 and Th-230 were revised from the overly conservative 

estimates included in Revision 0 of the FTF PA. The initial Revision 0 estimates assumed that 

these radionuclides were in full transient equilibrium with the parent U-234. This assumption is 

extremely conservative; therefore, a more reasonable estimate was made assuming a decay 

ingrowth period of 70 years. Although this estimate was more reasonable, the resulting 

inventory value fell below the detection level capability. Therefore, the inventory estimate for 

both Ra-226 and Th-230 were adjusted to the minimum detection level. 

Within the Type I waste tank group, the maximum waste tank inventory for any one tank was 

used to estimate inventory for the other waste tanks within the grouping. 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 FTF. 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 for all tanks within the Type I waste tank 

group. 

Type IV: Tanks 17 and 20 have been removed from service and grouted. Following waste 

removal activities, the residual material within Tanks 17 and 20 was sampled and analyzed. 
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This analysis provided the basis for estimates of the residual material. The Tanks 17 and 20 

residual estimates were used exclusively; no adjustments were made. The Tanks 18 and 19 

inventory estimates are based on concentrations documented in residual characterization 

reports with the total final inventories adjusted to reflect the bounding waste residual estimates 

(i.e., 5,000 gallon for Tank 18 and 3,000 gallon for Tank 19) at the time. The final inventory 

estimates for Tanks 18 and 19 apply the bounding waste residual floor estimates while allowing 

the corrosion product estimates to remain unchanged from Revision 0 of the FTF PA. 

Calculating the inventory contained within the corrosion products (excluding contaminant 

diffusion into the steel) required an estimate of the Kd value used for estimating the uranium 

content in the tank walls’ corrosion products. This value was 600 L/kg. An evaluation was 

performed to determine the appropriate value to use for the SRS tanks not cleaned with oxalic 

acid. Due to the high pH environment present in the waste tanks, high ionic strength and 

moderately high carbonate concentrations, the adsorption of the uranium ion to the rust 

surfaces was expected to be two or three orders of magnitude less than the 6,000 L/kg value 

previously used. In the absence of additional data, a decrease of only one order of magnitude 

was used to provide a conservative uranium inventory in the corrosion products. 

Diffusion of contaminates into the steel was considered negligible. A comparison of the amount 

of material diffused into the transfer lines to the amount of material estimated in the corrosion 

products was performed. This comparison demonstrated that the estimated quantity of material 

in the corrosion products exceeded that which diffused into the steel. Therefore there was no 

estimate of material diffused into the steel. 

For a majority of the radionuclides with an adjusted inventory less than one curie, their 

inventories were adjusted to either one curie or the analytical detection limit (1.0E-03 Ci). This 

allows for more efficient and cost effective means of confirming concentrations within residual 

materials for radionuclides with a limited potential impact to dose. For those radionuclides that 

have been observed (through previous analyses or scoping studies) to have greater potential 

impact on the overall dose, the inventory was adjusted to the analytical detection limit. The 

following radionuclides fit into this category: Ac-227, Cm-247, Cm-248, I-129, Nb-94, Pa-231 

and Pu-244. Those radionuclides with adjusted inventories greater than one curie were not 

adjusted in this step. 

Based on recent preliminary sample results, the estimates for Am-243, Cs-135 and Np-237 

(Tank 18 only) were further adjusted higher to ensure the estimates used are reasonably 

conservative. 

The inventory estimates for both Ra-226 and Th-230 were revised from the overly conservative 

estimates included in Revision 0 of the FTF PA. The initial Revision 0 estimates assumed that 

these radionuclides were in full transient equilibrium with the parent U-234. This assumption is 

extremely conservative; therefore, a more reasonable estimate was made assuming a decay 

ingrowth period of 70 years. Although this estimate was more reasonable, the resulting 

inventory value fell below the detection level capability. Therefore, the inventory estimate for 

both Ra-226 and Th-230 were adjusted to the minimum detection level. 

Type IIIA: The Type IIIA tank grouping consists of Tanks 25 through 28 and Tanks 44 through 

47. An adjustment to the inventory estimates was made by multiplying by a factor of 10 to 

provide for greater uncertainties in the residual inventories from the planned waste removal 

operation. Based on recent waste removal activities in Tanks 5 and 6, there is uncertainty 

around the projected inventories for the waste tanks remaining to be cleaned. To account for 

this uncertainty and ensure that the PA would provide a reasonable bounding inventory 

projection, the existing inventories were multiplied by a factor of 10. 

For a majority of the radionuclides with an adjusted inventory less than one curie, the 

inventories were adjusted to either one curie or the analytical detection limit (1.0E-03 Ci). This 

allows more efficient and cost effective means of confirming concentrations within residual 

materials for radionuclides with a limited potential impact to dose. For those radionuclides that 

have been observed (through previous analyses or scoping studies) to have greater potential 
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impact on the overall dose, the inventory was adjusted to the analytical detection limit. The 

following radionuclides fit into this category: Ac-227, Cm-247, Cm-248, I-129, Nb-94, Pa-231 

and Pu-244. Those radionuclides with adjusted inventories greater than one curie were not 

adjusted in this step. 

The inventory estimates for both Ra-226 and Th-230 were revised from the overly conservative 

estimates included in Revision 0 of the FTF PA. The initial Revision 0 estimates assumed that 

these radionuclides were in full transient equilibrium with the parent U-234. This assumption is 

extremely conservative; therefore a more reasonable estimate was made assuming a decay 

ingrowth period of 70 years. Although this estimate was more reasonable, the resulting 

inventory value fell below the detection level capability. Therefore, the inventory estimate for 

both Ra-226 and Th-230 were adjusted to the minimum detection level. 

Within the Type IIIA waste tank group, the maximum waste tank inventory for any one tank was 

used to estimate inventory for the other waste tanks within the grouping. As each waste tank is 

cleaned, the waste removed from within the tank during waste removal will be transferred to 

other tanks. In most case, this transfer will be to another tank of the same type. So the 

material in one tank will typically pass through other tanks of the same type prior to exiting the 

FTF. 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 for all 

tanks within the Type IIIA waste tank group. 

Type III: The Type III tank grouping consists of Tanks 33 and 34. For a majority of the 

radionuclides with an adjusted inventory less than one curie, their inventories were adjusted to 

either one curie or the analytical detection limit (1.0E-03 Ci). This allows more efficient and 

cost effective means of confirming concentrations within residual materials for radionuclides 

with a limited potential impact to dose. For those radionuclides that have been observed 

(through previous analyses or scoping studies) to have greater potential impact on the overall 

dose, the inventory was adjusted to the analytical detection limit. The following radionuclides fit 

into this category: Ac-227, Cm-247, Cm-248, I-129, Nb-94, Pa-231 and Pu-244. Those 

radionuclides with adjusted inventories greater than one curie were not adjusted in this step. 

The inventory estimates for both Ra-226 and Th-230 were revised from the overly conservative 

estimates included in Revision 0 of the FTF PA. The initial Revision 0 estimates assumed that 

these radionuclides were in full transient equilibrium with the parent U-234. This assumption is 

extremely conservative; therefore a more reasonable estimate was made assuming a decay 

ingrowth period of 70 years. Although this estimate was more reasonable, the resulting 

inventory value fell below the detection level capability. Therefore, the inventory estimate for 

both Ra-226 and Th-230 were adjusted to the minimum detection. [SRS-REG-2007-00002] 

At the completion of waste removal for each of the tanks, the estimated inventory identified for that tank in 

the FTF PA will be compared and evaluated against the actual inventory measured after the tank has 

been cleaned. [SRS-REG-2007-00002] The 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. 

2.4.1 Residual Inventory for Tank Annuli, Inside Failed Cooling Coils and Internal Tank Surfaces 

The inventories in the annulus, inside failed cooling coils and on the surface of the waste tank walls, 

cooling coils and columns are assumed to be covered by the estimated total tank inventories. The 

material in the annulus is expected to be dried salt that can be readily dissolved and effectively removed 

when the annular regions are cleaned during closure preparation. At the time of closure, radioactive 

contamination will remain on the wall and floor of the annulus. Cooling coils with the potential for residual 

waste holdup will be evaluated and flushed 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. [SRS-REG-2007-00002] 

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 
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small quantities of radioactive material could be held on the corrosion products. [SRT-WPT-2005-00049] 

The surfaces of waste tank internal walls, cooling coils and support columns are not expected to contain 

significant deposits based on inspections performed to date to support mechanical heel removal activities. 

2.4.2 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 FTF catch tank, diversion boxes, valve boxes 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. [SRS-REG-2007-00002] 

Ancillary structures inventories are estimated for the following three categories: 

 transfer lines, 

 pump tanks and catch tank, and 

 evaporators and concentrate transfer system tank. 

The inventory of each waste tank is used to establish the characterization of the residual material in the 

ancillary structures, as described in the FTF PA. The results of a review of waste transfers within FTF 

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 FTF waste tank. The representative concentration was then 

determined by applying a weighted average to each isotopic distribution in the FTF 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 as 

significant to an inadvertent intruder as the long-lived isotopes and actinides. [SRS-REG-2007-00002] 

2.5 Residual Waste Stabilization 

In May 2002, DOE issued an Environmental Impact Statement (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 Record of Decision 

(ROD) selecting the Fill with Grout alternative for SRS waste tank closure. [DOE-EIS-0303 ROD, SRSREG-2007-00002] 

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,

 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,
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 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 be reducing grout, which has low reduction potential (Eh) minimizing the 

mobility of the radionuclides after closure. All grout formulas are alkaline because grout is a cementbased material that naturally has a high pH, which is compatible with the tank carbon steel. The tank fill 

grout will have 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. 

[SRS-REG-2007-00002] 

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). [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 

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 be used to fill the Type I, III and IIIA waste tanks. In the Type IV tanks, reducing grout 

will 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). 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. 

For waste tank types with cooling coils and annulus, the cooling coils and annulus 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 annulus, 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. 

With the exception of the transfer lines, ancillary structures (e.g., diversion boxes, pump tanks, pump pits) 

associated with FTF will be filled, as necessary, at final closure to prevent subsidence. The waste 

transfer lines were modeled in the FTF PA with no grout and the result was in compliance with the 

required performance objectives. 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 

benefit to grouting these small diameter transfer lines. [SRS-REG-2007-00002] 
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2.6 Closure Cap 

An engineered closure cap will be installed over the FTF following the closure of the waste tanks and 

ancillary equipment. The closure cap description is based on a recent SRNL report on the FTF closure 

cap concept and estimated initial infiltration presented in WSRC-STI-2007-00184. 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 FTF PA. The closure cap design will be finalized 

closer to the time of FTF closure, to take advantage of possible advances in materials and closure cap 

technology that could be used to improve the design. 

A detailed discussion on closure cap design is provided in Section 3.0 of the FTF PA. An overview of the 

general design features and figures from the FTF PA are provided below. [SRS-REG-2007-00002] 

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. Figure 2.6-3 and Figure 2.6-4 present cross sections of the closure 

cap conceptual design. Figure 2.6-5 depicts the finished closure cap. Figure 2.6-6 presents the 

elevations of the closure cap surfaces and the grading plan. 

Figure 2.6-1: FTF Closure Cap General Concept 
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Figure 2.6-2: FTF Closure Cap Conceptual Design, Cap Footprint 
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Figure 2.6-3: FTF Closure Cap Conceptual Design, Sections A-A, B-B, and C-C 

NOTE: Vertical scale of sections has been exaggerated 5 times in order to show all closure cap layers. 
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Figure 2.6-4: FTF Closure Cap Conceptual Design, Sections D-D and E-E 

NOTE: Vertical scale of sections has been exaggerated 5 times in order to show all closure cap layers. 
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Figure 2.6-5: FTF Closure Cap Conceptual Design, Finished Cap 
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2.6.1 Function of Closure Cap Layers 

It is anticipated that the FTF closure cap will 

consist of the layers illustrated in Figure 2.6-1 

and identified in Table 2.6-1 from top to 

bottom. Table 2.6-2 summarizes the function 

of each of these layers. Detailed discussion of 

each layer in the FTF closure cap concept is 

provided in WSRC-STI-2007-00184. The 

concepts for the side slopes and toes of the 

closure cap based upon the results of physical 

stability calculations referred to above are also 

detailed in WSRC-STI-2007-00184. 

Figure 2.6-6: FTF Closure Cap Conceptual Design, Grading Cap 

Table 2.6-1: FTF Closure Cap Layers 

Layer 

Layer 

Thickness (in) 

Vegetative Cover NA 

Topsoil 6 

Upper Backfill 30 

Erosion Barrier 12 

Geotextile Fabric - 

Middle Backfill 12

Geotextile Filter Fabric - 

Lateral Drainage Layer 12 

Geotextile Fabric - 

HDPE Geomembrane 0.06 (60 mil) 

Geosynthetic Clay Liner (GCL) 0.2 

Upper Foundation Layer 12 

Lower Foundation Layer 72 (minimum) 

- Negligible 

NA - Not Applicable 

[WSRC-STI-2007-00184, Table 11] 
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Table 2.6-2: Function of the FTF 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 PMP 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 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 State-approved closure plan 

or permit. Criteria (2) and (3) will be discussed in subsequent sections of this Draft FTF 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 plans20. 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 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 nonproliferation risks that the Secretary concludes cannot be adequately guarded against absent deep 

geologic disposal21. 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 

 

20 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.” 

21 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 FTF 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). 
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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 highly radioactive radionuclides (HRRs) 

removed “to the maximum extent practical” (MEP). This section demonstrates that the 

FTF 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 Draft FTF 3116 Basis Document. 

Section 5.2 describes the removal processes used to remove HRRs to the maximum 

extent practical. Section 5.3 demonstrates that, at closure, the HRRs will have been 

removed to the MEP. 

Key Points 

 The list of HRRs for FTF describes the radionuclides that could reasonably be expected 

to exist in the FTF 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 FTF 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. 

 Greater than 95% of the waste volume and radioactivity will have been removed from 

the FTF tanks and ancillary structures collectively prior to closure activities using 

mechanical, chemical or vacuum technologies that have previously been successfully 

demonstrated in similar waste removal activities at the SRS.

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-234, Np-237, Pu-238, 

Pu-239, Pu-240 and Am-241 are the HRRs in the FTF stabilized residuals, FTF waste tanks and FTF 

ancillary structures at the closure of FTF that DOE has determined contribute significantly to radiological 

risk to workers, the public and the environment, taking into account scientific principles, knowledge and 

expertise22. 

 

22 Some of the radionuclides listed as HRRs in this Draft FTF 3116 Basis Document may not be listed in other 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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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 (such as National Council on Radiation Protection and 

Measurements information), and included radionuclides not necessarily present in the projected FTF 

inventory at closure. [CBU-PIT-2005-00228] The initial listing of 849 radionuclides included those 

radionuclides from Tables 1 and 2 of 10 

CFR 61.5523. 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 FTF PA base 

(expected) 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 long-lived radionuclides that 

persist well into the future, may be mobile 

in the environment or may pose a risk to 

humans if inhaled or ingested. 

The list of FTF HRRs in Table 5.1-1 

account for approximately 99% of the 

current radioactivity in the FTF 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 FTF 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 Draft FTF 

3116 Basis Document focus on Cs-137 or Sr-90 since approaches that are effective in removing Cs-137 

and Sr-90 also remove Ba-137m and Y-90, respectively. Approximately 98% of the current radioactivity 

in the FTF waste is associated with these two radionuclides and their short-lived daughters. [SRR-LWP2010-00058] Moreover, Cs-137, Sr-90 and their daughters are present in sufficient concentrations in the 

FTF waste that, without shielding and controls, they produce radiation emissions that would present risk 

to humans simply due to their proximity 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 radionuclides combined account for less than 1% of the current radioactivity 

in the FTF waste today24. [SRR-LWP-2010-00058] 

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 FTF PA, which includes sensitivity and 

 

23 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. 

24 The remaining radionuclides in the FTF waste in combination contribute approximately 1% of the current radioactivity. 

Table 5.1-1: FTF 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 c, e 2.11E+05 X 

I-129 a 1.57E+07 X 

Cs-137 b, c, d, f 3.00E+01 X 

U-234 a, c 2.46E+05 X 

Np-237 a, b, c, e 2.14E+06 X 

Pu-238 a, e 8.77E+01 X 

Pu-239 a, c, e 2.41E+04 X 

Pu-240 a, c, e 6.56E+03 X 

Am-241a, b, c, e 4.32E+02 X 

a

 HRRs based on groundwater analyses results from the FTF PA. 

[SRS-REG-2007-00002] b

 HRRs based on intruder analysis results from the FTF PA. 

[SRS-REG-2007-00002] c

 HRRs based on uncertainty and sensitivity run results from the FTF PA. 

[SRS-REG-2007-00002] 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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uncertainty analyses. The FTF PA applied a rigorous, documented, multi-step, multi-factor screening 

methodology (including consideration of National Council on Radiation Protection and Measurements 

[NCRP] information) to 849 radionuclides compiled from radionuclides present in the FTF (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 FTF PA to focus on those radionuclides that contribute to the dose for various pathways. [SRSREG-2007-00002] 

For the purpose of determining which radionuclides should be evaluated in the FTF 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. [http://www.epa.gov/radiation/heast/] 

 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 FTF inventory for the PA. The screening and adjustment 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. 

 The FTF waste production history information was evaluated for the potential for a specific 

radionuclide to be present in FTF. This step screened out radionuclides not present within the 

FTF 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.” 

 Several radionuclides were added back to the list based on the possibility that these 

radionuclides may have theoretically been created at various points of production in SRS history. 
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 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 FTF 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 FTF PA modeling. The results of the FTF PA analyses were then evaluated using the 

methodology described below to determine which of the 54 radionuclides would be considered HRRs for 

FTF. [SRR-CWDA-2009-00045, SRS-REG-2007-00002] 

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 FTF 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 consultation to date25. While the approach 

followed by DOE in this Draft FTF 3116 Basis Document differs in some aspects from the approaches 

followed in prior 3116 Basis Documents, 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. 

Several radionuclides have been included on the HRR list based on an evaluation of the FTF 100-meter 

groundwater dose, using the groundwater dose results calculated in the FTF PA. For the FTF PA, the 

100-meter point is the point of maximum exposure at, or outside of, the FTF 100-meter buffer zone. The 

groundwater analysis in the FTF PA utilized the initial inventory of 54 radionuclides resulting from the 

screening analysis described in the previous section as input for the FTF PA model. The model used to 

perform the groundwater analysis accounted for radioactive decay and ingrowth throughout the 

assessment period. [SRS-REG-2007-00002] 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 FTF PA, from 

all individual radionuclides in the FTF 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 FTF PA groundwater 

analysis at any time within 20,000 years26. 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 1.25 mrem/year27. These radionuclides were 

screened from the HRR list. Based on this evaluation, the remaining radionuclides - I-129, Cs-135, Ra226, Th-229, U-233, U-234, Np-237, Pu-239 and Pu-240 - were initially identified for consideration. For 

those radionuclides, the projected inventories (as shown in the FTF PA) at the time of FTF closure were 

reviewed. For those radionuclides with a relatively insignificant initial inventory (i.e., Ra-226 and Th-229) 

the associated decay chains were examined. Reduction of these radionuclides is accomplished through 

 

25 NRC suggested during a public Draft FTF NDAA 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] 

26 The FTF PA evaluates the consequences of closure activities during the performance period (i.e., 10,000 years) after closure. 

The DOE also has evaluated for periods beyond 10,000 years in the FTF PA to further inform closure decisions. 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 tank degradation for Type I and Type III/IIIA waste tanks beyond 10,000 years. DOE is using a 20,000-year evaluation period for 

determination of HRRs due to the expected tank liner failures beyond the 10,000-year performance period and resultant peak 

doses. This tank liner failure analysis is specific to the SRS FTF Type I and Type III/IIIA waste tanks. 

27 This approach is consistent with the guidance and general approach in Volume 2 of NUREG-1757, Consolidated NMSS 

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 Section 3116 of the NDAA 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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the removal of the respective parent radionuclides, i.e., U-233 (for Th-229), U-234 (for Ra-226), Pu-238 

(for Ra-226). An additional parent radionuclide, Am-241, was included based on progeny ingrowth of Np237. 

Since the completion of the FTF PA, DOE has performed waste characterization of the residuals in both 

Tank 18 and Tank 19 which showed that Cs-135 and U-233 were significantly lower than projected in the 

FTF PA. [SRR-CWDA-2010-00117, SRR-CWDA-2010-00118] Moreover, based on the actual Tanks 18 

and 19 source terms and the results of the FTF PA analysis, the associated doses from the Cs-135, U233 and its daughter Th-229 inventories in aggregate with the previously screened radionuclides are not 

expected to contribute greater than 1.25 mrem/yr to the public groundwater dose. 

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 Draft FTF 3116 Basis 

Document, I-129, U-234, Np-237, Pu-238, Pu-239, Pu-240 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 FTF 100-

meter dose from airborne radionuclides, using the air pathway dose results calculated in the FTF PA. In 

the FTF 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 FTF PA, the 100-meter point is the point of maximum exposure at or 

outside of the FTF 100-meter buffer zone. The FTF PA shows that the air pathway is not a significant 

contributor to dose, and contributes, in aggregate approximately 0.2 mrem/yr peak dose. [SRS-REG2007-00002] Therefore, for the purpose of this Draft FTF 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 FTF inadvertent 

intruder dose, using the intruder dose results calculated in the FTF PA. The intruder analysis in the FTF 

PA utilized the initial inventory of 54 radionuclides resulting from the screening analysis described 

previously as input for the FTF PA model. The model used to perform the intruder analysis did account 

for radioactive decay and ingrowth throughout the assessment period. [SRS-REG-2007-00002] 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 FTF PA, 

from all individual radionuclides in the FTF 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 FTF PA intruder analysis at any 

time within 20,000 years28. The individual doses from each radionuclide were then listed in order of 

contribution. DOE conducted a quantitative analysis to determined those radionuclides with an aggregate 

contribution to dose of less than or equal to 25 mrem/yr29. These radionuclides were screened from the 

HRR list. Based on this evaluation the remaining radionuclides - Sr-90, Y-90, Cs-137, Ba-137m, Th-229, 

U-233 and Np-237 - were initially identified for consideration. For those radionuclides, the projected 

 

28 The FTF PA evaluates the consequences of closure activities during the performance period (i.e., 10,000 years) after closure. 

The DOE has evaluated for periods beyond 10,000 years in the FTF PA to further inform closure decisions. 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 tank degradation for Type I and Type III/IIIA waste tanks beyond 10,000 years. DOE is using a 20,000-year evaluation 

period for determination of HRRs due to the expected tank liner failures beyond the 10,000-year performance period and resultant 

peak doses. This tank liner failure analysis is specific to the SRS FTF Type I and Type III/IIIA waste tanks. 

29 This approach is consistent with the guidance and general approach in Volume 2 of NUREG-1757, Consolidated NMSS 

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 Section 3116 of the NDAA 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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inventories (as shown in the FTF PA) at the time of FTF closure were reviewed. For the radionuclide with 

a relatively insignificant initial inventory (i.e., Th-229), the associated decay chain was examined. 

Reduction of this radionuclide is accomplished through the removal of the respective parent radionuclide 

(i.e., U-233). An additional parent radionuclide, Am-241, was included based on progeny ingrowth of Np237. 

Since the completion of the FTF PA, DOE has performed waste characterization of the residuals in both 

Tank 18 and Tank 19 which showed that U-233 was significantly lower than projected in the FTF PA. 

[SRR-CWDA-2010-00117, SRR-CWDA-2010-00118] Moreover, based on the actual Tanks 18 and 19 

source terms and the results of the FTF PA analysis, the associated doses from U-233 and its daughter, 

Th-229, inventories, in aggregate with the previously screened radionuclides, are not expected to 

contribute greater than 25 mrem/yr to the public groundwater dose. 

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 contribution 

to the inadvertent intruder dose. 

DOE believes the screening of radionuclides whose dose contribution, in aggregate, is less than or equal 

to 25.0 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 Draft FTF 3116 Basis Document, Sr-90, Cs-137, Np-237 and Am-241 

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 FTF PA. [SRS-REG-2007-00002] The FTF 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 FTF PA model. In addition, as part of the 

sensitivity analyses, modeling was performed for a period of 100,000 years. [SRS-REG-2007-00002] 

The FTF PA considers the uncertainties and sensitivities associated with the projected dose results to a 

member of the public through uncertainty analysis of the FTF probabilistic model, through sensitivity 

analysis using the FTF probabilistic model and through sensitivity analysis using the FTF deterministic 

model. A review of the uncertainty analysis realizations with the highest peak doses show Tc-99, Np-237, 

Pu-239 and Pu-240 to be significant. A review of the sensitivity analysis performed using the FTF 

probabilistic model shows Tc-99 and Pu-239 to be significant. A review of the sensitivity analysis 

performed using the FTF deterministic model, including a barrier analysis, show Tc-99, U-234, Np-237, 

Pu-239 and Am-241 to be significant. Based on an evaluation of the FTF PA results for these different 

analyses, Tc-99, U-234, Np-237, Pu-239, Pu-240 and Am-241 were included in the HRR list. The FTF 

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. Probabilistic intruder sensitivity 

analyses performed identified Tc-99 and Pu-239 as potentially significant to the inadvertent intruder 

results. A review of the FTF probabilistic sensitivity analyses identified Sr-90, Y-90, Cs-137, Ba-137m 

and Am-241 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 
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Page 5-7 

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, Pu-239 and Am-241 were included in the HRR list based on an evaluation 

of the results for the FTF PA inadvertent intruder uncertainty and sensitivity analyses.

Based on the FTF PA uncertainty and sensitivity analyses, Sr-90, Tc-99, Cs-137, U-234, Np-237, Pu-239, 

Pu-240 and Am-241 were included in the HRR list. 

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 FTF 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-234, Np-237, Pu-238, Pu-239, Pu-240 

and Am-241 are considered the HRRs in the FTF stabilized residuals, FTF waste tanks and FTF ancillary 

structures at the closure of FTF. 

Table 5.1-2: Radionuclides Contained in FTF PA Initial Inventory

Evaluation Resultsa

Groundwater 

Analysis 

Air Pathway 

Analysis 

Intruder 

Analysis 

Radionuclide with Initial Inventory Half-Life (yr) 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 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 

 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 Draft FTF 3116 Basis Document are highlighted. 

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Table 5.1-2: Radionuclides Contained in FTF PA Initial Inventory (Continued) 

Evaluation Resultsa

Groundwater 

Analysis 

Air Pathway 

Analysis 

Intruder 

Analysis 

Radionuclide with Initial Inventory Half-Life (yr) 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 

I-129b 1.57E+07  x x 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  

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 

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 x x 

U-234 2.46E+05   x x x x  x 

U-235 7.04E+08 x 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 

Pu-238b 8.77E+01 x  x x x x x x 

Pu-239b 2.41E+04  x x x x x  x 

Pu-240b 6.56E+03  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 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 Draft FTF 3116 Basis Document are highlighted. 

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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 practical30.” Section 5.2 and Section 5.3 demonstrates 

that the FTF 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”; “useful” 

(http://infoplease.com/ipd/A0598638.html); selected “mindful of the results, usefulness, advantages or 

disadvantages, etc., of [the] action or procedure” (http://dictionary.cambridge.org/

define.asp?key=practical*2+0&dict=A); fitted to “the needs of a particular situation in a helpful way”; 

“effective or suitable.” 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 benefits31 of deploying a technology with respect to the conditions in a specific FTF 

waste tank or ancillary structure. The MEP standard contemplates room for exercising expert judgment in 

weighing several factors. Such factors may include 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 contribution32 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 FTF 

waste tanks and ancillary structures to the MEP for the purpose of removal from service33 and eventual 

closure of the waste tanks and ancillary structures. Removal of HRRs to the MEP in FTF 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. 

5.2.1 Current Status of F-Tank Farm Waste Removal Activities 

The FTF includes 22 waste storage tanks. Two of these waste tanks, Tanks 17 and 20, previously 

underwent final waste removal activities, were removed from service in 1997, were filled with a 

 

30 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. 

31 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. 

32 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 NMSS 

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 Section 3116 of the NDAA or that either the NUREG or 10 CFR 20, Subpart E is applicable in 

the 3116 context. 

33 The term “removal from service” refers to the protocols set forth in the State-approved General Closure Plan (GCP), as described 

in Section 8.0 of this Draft FTF 3116 Basis Document, to stabilize the waste tank or ancillary structure and prepare it for final 

closure. 
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combination of grouting materials, and are not encompassed within this Draft FTF 3116 Basis Document. 

[PIT-MISC-0002, PIT-MISC-0004] All of the remaining Type I or Type IV waste tanks in FTF are no 

longer actively receiving fresh canyon waste and many are currently undergoing waste removal activities. 

Tank 18 and Tank 19 have undergone extensive waste removal campaigns and tank residual samples 

have been collected for the purposes of characterizing the remaining tank residuals. It is anticipated that 

Tank 18 and Tank 19 will be the next two waste tanks removed from service in FTF. Tank 5 and Tank 6 

have completed bulk waste removal and have undergone extensive heel removal activities. Tank 5 and 

Tank 6 are anticipated to be the next set of tanks removed from service following Tank 18 and Tank 19. 

Tank 5 and Tank 6 will represent the first waste tanks removed from service that are Type I tanks and 

have extensive interior structures (i.e., support columns and cooling coils). Tank 3, Tank 4, Tank 7 and 

Tank 8 have all undergone initial waste removal activities. Waste removal activities have not been 

initiated in Tank 1 or Tank 2, or any of the ten Type III/IIIA tanks located in FTF. 

As of March 31, 2010, FTF stored approximately 121,000,000 curies in approximately 15,000,000 gallons 

of waste. The sludge portion of this waste represents approximately six percent of the volume but 

contains approximately 45 percent of the radioactivity. Of the approximate 14,000,000 gallons of salt 

waste, approximately 8,800,000 gallons is in the form of saltcake with the remaining approximately 

5,200,000 gallons being concentrated supernate. The concentrated supernate accounts for 

approximately 50 percent of the total radioactivity in the FTF. Figure 5.2-1 graphically presents the 

breakdown of the waste in FTF in terms of both volume and curies for each of the three primary waste 

types. [SRR-LWP-2010-00040] 

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 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). 

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 

Figure 5.2-1: FTF Waste Tank Composite Inventory 

Volume

14.8 Mgal

Curies

67 MCi

(55%)

54 MCi

(45%)

121 MCi

60 MCi

(50%)

Sludge

14 Mgal

(94%)

0.8 Mgal

(6%)

5.2 Mgal

(35%)

FTF Inventory values as of March 31, 2010

Salt Supernate

7 MCi Saltcake (5%)

8.8 Mgal

(59%)
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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] 

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. 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. 

The following discusses the current technologies and how they are being utilized at SRS. 

5.2.2.1 Mechanical Heel Removal Cleaning 

As documented in the Systems Engineering Evaluation, 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 (chemically-treated water34 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. 

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 

5.2-2)35. 

 

34 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. 

35 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 tank configuration and waste characteristics. 

Figure 5.2-2: Submersible Mixer Pump Waste Removal Diagram 
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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-HP motor that can 

operate up to 1,600 rpm. The SMPs utilize the tank liquid waste to cool the 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-00344] While obstructions such as cooling coils and 

support columns affect the effective cleaning radius of SMPs, the use of multiple SMPs in an obstructed 

tank can make a significant contribution to removal of waste. Experience in Tank 5 and Tank 6 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 the most 

challenging tank 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 and IIIA tanks. Pump configurations are shown in Figure 5.2-3 and Figure 5.2-4. 

The first deployments and operations of SMPs in FTF 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 

removal in Tank 5 and Tank 6. [M-ESR-F-00107, M-ESR-F-00147, M-ESR-F-00132] 

Figure 5.2-3: Submersible Mixing Pump and Submersible 

Transfer Pump 

Figure 5.2-4: Submersible 

Mixing Pump in a Test Tank 
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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. The oxalic acid may also 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 1636, located in 

HTF, in the early 1980s. Tanks 5 and 6 were cleaned using three large batches of 8 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.). [LWO-SPT-2008-00033] 

Oxalic acid cleaning in Tank 16 (part of an overall waste removal program which also employed 

mechanical cleaning) resulted in the removal of over 99% of the activity. [DPST-81-441, DPSP-80-17-23] 

The oxalic acid was again used as the chemical heel removal method in Tank 5 and Tank 6 in 2008 and 

2009. Oxalic acid was applied multiple times in each tank using various methods (e.g., downcomer, 

spray 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. [M-ESR-F-00160, 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 percent 

increase in waste volume between the first and second chemical cleaning cycles and the second and 

 

36 Discussion on Tank 16 cleaning is included to demonstrate successful deployment of the oxalic acid cleaning process; however, 

Tank 16 is located in the HTF at SRS and is not within the scope of this Draft FTF 3116 Basis Document. 
Draft Basis for Section 3116 Determination DOE/SRS-WD-2010-001 

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Page 5-14 

dissolved 

sludge

Waste Tank Undergoing Cleaning Type III/IIIA Waste Tank

ozone

Oxalate

Decomposition Dissolution

dilute 

slurry

ECC Module

Ozone 

Generation 

Module

Solid-Liquid 

Separation

Oxalic Acid 

6.5 wt%

Primary 

Ventilation 

Module

Cooling 

Water 

Module

Secondary 

Ventilation 

Module Mixer(s) 

De-ionized 

Water Unit 

third chemical cleaning cycles for Tank 6 and between the first and second chemical cleaning cycles in 

Tank 5. [M-ESR-F-00158, M-ESR-F-00165] 

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 

Saltstone Disposal Facility37. 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. Modeling shows that for every tank that undergoes 

chemical cleaning, about 51,000 kg of new sodium oxalates (a non-radioactive compound) solids will be 

created for feed to DWPF. 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-STI-2010-

00015] The oxalates are also anticipated to create evaporator foaming and scaling problems. [LWOSPT-2008-00033] Due to these downstream impacts, the amount of Bulk Oxalic Acid Cleaning in FTF will 

be carefully controlled to optimize the cleaning effectiveness and the downstream waste treatment 

processes. 

ECC is a new oxalic acid chemical cleaning process being tested and designed by DOE for use in FTF 

tanks. ECC is based on adapting proven techniques from the commercial reactor and steam generating 

industries to the cleaning of carbon steel waste tanks. The ECC process uses dilute oxalic acid to 

dissolve residual waste solids and clean the tank internals. The oxalates in the resulting acid stream are 

then destroyed using ozone. [SRR-STI-2010-00015] A dilute slurry containing the remaining dissolved 

metals and associated radionuclides is fed to a settling tank (i.e., FTF waste tank) with the liquid fraction 

undergoing subsequent evaporation. Since this process can be utilized with minimal impacts on the 

Liquid Waste System, ECC can continue until the process is no longer effective in removing the residuals. 

This chemical cleaning process minimizes the impacts of the use of oxalic acid for residual heel cleaning 

and potentially permits additional cleaning opportunities. [LWO-SPT-2008-00033] Figure 5.2-5 provides 

a conceptual ECC Process diagram38. 

 

37 See Appendix A for a brief description of DWPF, SWPF and Saltstone Disposal Facility operations. 38 The ECC process is still in the stage of being tested and designed by DOE. The figure depicted represents a conceptual diagram 

of the ECC process and is subject to change as development of the process advances. 

Figure 5.2-5 ECC Process Diagram 
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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 the 

unobstructed Type IV tanks, Tanks 18 and 19. This technology 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, 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. 

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. 

Figure 5.2-6: Mantis 

In Tank 19 Prior to Cleaning

In Test Facility

In Tank 19 After Cleaning
Draft Basis for Section 3116 Determination DOE/SRS-WD-2010-001 

for Closure of F-Tank Farm Revision 0 

at the Savannah River Site September 30, 2010 

Page 5-16 

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 III and 

Type IIIA tanks severely restrict the mobility 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. 

5.2.3 Optimization of Existing Technologies 

DOE continues to optimize its existing technologies that have been successfully deployed for waste 

removal. 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 to sample Tanks 18 

and 19 (Figure 5.2-7) and is being adapted 

for applications in future tanks. Such tactical 

applications of tailored robotic platforms will 

continue to be used in future waste removal 

activities in FTF. 

5.3 Removal to the Maximum Extent 

Practical 

As described above, extensive waste 

removal operations have occurred at SRS 

and specifically within FTF. Based on waste 

removal experience to date and anticipated 

new technologies, FTF waste removal 

activities will result in significant collective 

removal of all waste including HRRs39. 

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. Following bulk waste removal, heel removal is performed using a mix of 

technologies described above, as appropriate, accounting for the physical configuration of the tank and 

the chemical characteristics of the waste. 

Throughout the 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. 

The FTF Type IV tanks, Tanks 17, 18, 19 and 20, 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 that was used for the selection 

of the heel removal technology for Tank 19 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 

 

39 Experience to date with waste removal for Tanks 18 and 19 resulted in removal of approximately 99% of the radioactivity based 

on a starting point of the maximum historical radionuclide inventory in those tanks. See SRR-CWDA-2009-00030. In this regard, 

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 99% of HRRs, based on a 

starting point of the maximum historical radionuclide inventory in the overall FTF, individual 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 

FTF PA is not sole justification to conclude cleaning activities on an individual tank or ancillary structure. 

Figure 5.2-7: Robotic Sampler at Test Facility
Draft Basis for Section 3116 Determination DOE/SRS-WD-2010-001 

for Closure of F-Tank Farm Revision 0 

at the Savannah River Site September 30, 2010 

Page 5-17 

Waste Solids 

addition from

Tank 17 Heel

Removal 

Campaign

Waste Solids 

addition from

Tank 19 Heel

Removal 

Campaign

Completion of

Phase 3 — 

Heel Removal

with ADMP

Completion of

Phase 4 — 

Heel Removal

with Mantis

Start of Waste

Removal

Completion of

Phase 1 — 

Supernate 

Removal

Completion of

Phase 2 — 

Bulk Solids 

Removal

0

200,000

400,000

600,000

800,000

1,000,000

1,200,000

1,400,000

Tank 18 Volume (gallons)

0%

20%

40%

60%

80%

100%

Percent of Starting Inventory

99.7 % of the waste 

volume removed from 

Tank 18 in four specific 

waste removal phases

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. Figures 5.3-1 and 5.3-2 depict 

the overall waste removal effectiveness for Tanks 18 and 19 respectively. 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. This alternative 

technology analysis considered the current tank conditions as well as the impact to the overall Liquid 

Waste System from the operation of such technologies. [SRR-CWDA-2009-00030] Based on this 

technology review, which included consideration of costs, worker dose, and other potential impacts, DOE 

determined that deployment of an alternate technology in Tank 18 and Tank 19 was not practical. 

Figure 5.3-1: Tank 18 Waste Removal
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for Closure of F-Tank Farm Revision 0 

at the Savannah River Site September 30, 2010 

Page 5-18 

Completion of

Phase 2 — 

Bulk Salt 

Removal

Completion of

Phase 4 — 

Heel Removal

with Mantis

Completion of 

Phase 3 — 

Heel Removal 

with Mixer Pumps

Start of Waste

Removal

Completion of

Phase 1 — 

Supernate 

Removal

0

200,000

400,000

600,000

800,000

1,000,000

1,200,000

1,400,000

Tank 19 Volume (gallons)

0%

20%

40%

60%

80%

100%

Percent of Starting Inventory

99.8 % of the waste 

volume removed from 

Tank 19 in four specific 

waste removal phases

Two of the eight Type I tanks in FTF have also undergone extensive heel removal campaigns. These two 

Type I tanks, Tank 5 and Tank 6, 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 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 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. 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 

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 

Figure 5.3-2: Tank 19 Waste Removal
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for Closure of F-Tank Farm Revision 0 

at the Savannah River Site September 30, 2010 

Page 5-19 

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. 

Effective radionuclide removal is expected to be achieved during cleaning of the remaining tanks, and the 

cleaning and/or flushing of ancillary equipment, for a number of reasons. As discussed above, DOE has 

successfully removed waste from 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 FTF. The cleaning process 

employed is thorough, and the process is reviewed and documented during cleaning to maximize 

effectiveness. DOE will continue to use such measures as visual (remote) observation of remaining tank 

residuals against benchmarks in the tanks (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 FTF 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 

technologies40. 

5.4 Conclusion 

Removal of HRRs to the MEP in FTF 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 FTF waste tanks, ancillary structures and 

their associated stabilized residuals will have had HRRs removed to the MEP at the time of closure. 

 

40 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]. 
Draft Basis for Section 3116 Determination DOE/SRS-WD-2010-001 

for Closure of F-Tank Farm Revision 0 

at the Savannah River Site September 30, 2010 

Page 6-1 

6.0 RADIONUCLIDE CONCENTRATIONS OF STABILIZED RESIDUALS, TANKS AND ANCILLARY 

STRUCTURES 

Section Purpose 

The purpose of this section is to demonstrate whether the FTF 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 FTF 

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 FTF PA. 

 In addition, DOE, in consultation with the NRC41, has derived site-specific 

concentration averaging expressions for FTF 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 will exceed the Class C concentration limits in 10 CFR 61.55, DOE 

nevertheless is also consulting with the NRC on DOE’s disposal plans, as described in 

the Draft FTF 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. 

 

41 NRC suggested during a public Draft FTF NDAA 3116 Basis Document scoping meeting held on July 13-14, 2010 that DOE 

consider this approach. [SRR-CWDA-2010-00091] 
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at the Savannah River Site September 30, 2010 

Page 6-2 

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 Draft FTF 

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 Draft FTF 

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 

FTF are not expected to exceed concentration limits for Class C low-level waste. Nevertheless, DOE is 

also consulting 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 

is specifically requesting in this Draft FTF 3116 Basis Document that NRC identify what changes, if any, 

NRC would recommend to DOE’s disposal plans as described in the Draft FTF 3116 Basis Document, 

and DOE intends to consider the NRC recommendations, as appropriate, in the development of DOE’s 

plans. 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. 

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]

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]
Draft Basis for Section 3116 Determination DOE/SRS-WD-2010-001 

for Closure of F-Tank Farm Revision 0 

at the Savannah River Site September 30, 2010 

Page 6-3 

6.2 Approach to Waste Concentrations for F-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 FTF 

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 FTF 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 Draft FTF 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 risk-informed approach that takes into consideration 

such things as the specific conditions of the FTF 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 FTF at closure to the Class C concentration limits. 

6.3 Methodology 

The Category 3—Site-Specific Averaging approach to concentration averaging reflects site-specific 

conditions of FTF 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 

Figure 6.2-1: Intruder-Driller Scenarios for Concentration Calculations 
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for Closure of F-Tank Farm Revision 0 

at the Savannah River Site September 30, 2010 

Page 6-4 

FTF, DOE has developed, consistent with the Category 3—Site-Specific Averaging methodology, 

averaging expressions for FTF based on the results of the inadvertent intruder analysis performed within 

the FTF PA. [SRS-REG-2007-00002] 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 Tank 18, this 

comparison was based on the actual residual inventory in the tank following cleaning. For the ancillary 

structures, this comparison was based on the projected inventories at closure in the FTF PA. 

For purposes of comparison to the Class C concentration limits, and to align with the inputs used in 

developing the averaging expressions for FTF, the residual inventory used for these calculations are 

decayed to the inventory that will be present at the time of closure (assumed to be 2020 for the purposes 

of analysis in the FTF PA). [SRS-REG-2007-00002] As discussed below, the radionuclide 

concentrations of the stabilized residuals are compared, using the sum of fractions methodology and the 

FTF 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 FTF. [SRS-REG-2007-00002] To demonstrate compliance 

with 10 CFR 61.42 the FTF 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 FTF PA to develop the FTF 

averaging expressions used for waste classification. 

The FTF PA models used to simulate the performance of the FTF closure system take into account the 

release of radiological contaminants from the waste tanks and the associated ancillary structures in the 

FTF and simulates transport of the radiological contaminants through soil and groundwater to the 

assessment point. The models use numerous FTF-specific input parameters to represent the FTF 

closure system behavior over time. Many of the input parameters are based on site-specific data (e.g., 

soil and cementitious materials distribution coefficients) used in transport modeling. In addition, sitespecific information is used to model the behavior of individual barriers within the FTF, 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 FTF averaging expressions. A detailed discussion of the FTF 

PA Intruder analyses is provided in Section 7.1.5 of this Draft FTF 3116 Basis Document and the FTF PA. 

[SRS-REG-2007-00002] 

The stabilized contaminant materials after FTF 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 3-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 FTF footprint. However, for the purposes of developing averaging expressions for FTF, 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 FTF waste tanks and 

ancillary structures. 

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. 
Draft Basis for Section 3116 Determination DOE/SRS-WD-2010-001 

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Page 6-5 

 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, 

piping, cooling coils and any structure that will be abandoned 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 FTF, 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 FTF 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. [SRS-REG-2007-00002] 

6.3.2 Site-Specific FTF 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 FTF 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 body42. At closure, the depth of the stabilized residuals within the FTF waste tanks and 

ancillary structures will be well below (i.e., greater than 10 feet) the FTF closure cap and a robust intruder 

barrier (e.g., grouted waste tanks, diversion box cell covers and valve box shielding), as described in the 

FTF PA, will be in place. [SRS-REG-2007-00002] 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 FTF, site-specific averaging 

expressions for FTF, based on the results of the Inadvertent Intruder Analyses performed within the FTF 

PA, have been developed. 

The FTF PA provides the estimated dose to an intruder who resides within the boundary of the FTF after 

the period of institutional control (100 years). The intruder is assumed to be exposed via various 

pathways from water collected from a 1-meter well and from drill cuttings. The groundwater associated 

with the 1-meter well is contaminated from all the sources within the FTF (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 FTF PA assumes that a 3-inch transfer line is 

penetrated by a driller and the cuttings are spread among the garden – thus an additional source is added 

to the contaminated well source. The impact of drilling into a 4-inch transfer line has been presented in 

the FTF PA with respect to the chronic intruder. The impact of drilling into a waste tank was also 

considered in the FTF 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 

 

42 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] 
Draft Basis for Section 3116 Determination DOE/SRS-WD-2010-001 

for Closure of F-Tank Farm Revision 0 

at the Savannah River Site September 30, 2010 

Page 6-6 

drilling once significant resistance is encountered, a chronic intruder was not assessed. [SRS-REG2007-00002] 

Because the stabilized residuals in FTF 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 xSiteFactor

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 FTF after closure 

The FTF averaging expressions, based on the above equation, that DOE is utilizing are shown below. 

6.3.2.1 FTF Waste Tank Waste Concentration Calculation Averaging Expressions 

For FTF 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 xSiteFactor

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., 2020), 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) 
Draft Basis for Section 3116 Determination DOE/SRS-WD-2010-001 

for Closure of F-Tank Farm Revision 0 

at the Savannah River Site September 30, 2010 

Page 6-7 

For mass-based concentrations: 

i

R G

R

i

i xSiteFactor

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., 2020), 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 FTF Transfer Line Waste Concentration Calculation Averaging Expressions 

For FTF 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 xSiteFactor

V

I SurfaceArea

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., 2020), 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) 

For mass-based concentrations: 

i

TL

R

i

i xSiteFactor

M

I SurfaceArea

x

Table Value

SOF ( ) ( )

_

1  

where: 

SOFi

 = Radionuclide “i” contribution to the sum of the fractions 
Draft Basis for Section 3116 Determination DOE/SRS-WD-2010-001 

for Closure of F-Tank Farm Revision 0 

at the Savannah River Site September 30, 2010 

Page 6-8 

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., 2020), 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 FTF Averaging Expressions 

The site-specific factors in the FTF 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 FTF PA inadvertent intruder analyses along with the FTF PA inventory at closure were utilized. 

The FTF PA probabilistic model was utilized to determine the dose to the chronic intruder assuming the 1-

meter well contaminated source and one of three drill cutting sources including a 3-inch diameter transfer 

line, a 4-inch diameter transfer line or a waste tank. Because it is the primary contributor to the peak 

dose in the FTF, Tank 18 was used for the waste tank calculations. 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 time 

period evaluated started at 500 years after closure and for transfer lines the time period started at 100 

years after closure. Although the FTF 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 FTF 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 FTF PA, the individual 

radionuclide site-specific factors that would result in an inadvertent intruder under the FTF 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 SiteFactor i

PA

i

i

500 /

_ 

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 FTF PA inventory at closure, of the 

drilled source for radionuclide “i” 

Dosei = Peak dose, based on results of the FTF 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 
Draft Basis for Section 3116 Determination DOE/SRS-WD-2010-001 

for Closure of F-Tank Farm Revision 0 

at the Savannah River Site September 30, 2010 

Page 6-9 

Using values for the FTF 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., 3-inch diameter transfer line, a 4-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 3-inch diameter transfer line or the 4-inch diameter transfer line. 

Table 6.3-1 and Table 6.3-2 provide the FTF waste tank site-specific factors for 10 CFR Table 1 and 

Table 2 radionuclides, respectively, which are used in the FTF averaging expressions. Table 6.3-3 and 

Table 6.3-4 provide the FTF transfer line site-specific factors for 10 CFR Table 1 and Table 2 

radionuclides, respectively, which are used in the FTF averaging expressions. [SRR-CWDA-2010-00122] 

Table 6.3-1: FTF Waste Tank Site-Specific Factors for 10 CFR 61.55 Table 1 

Radionuclides 

Table 1 

Radionuclide 

Site-Specific 

Factor 

C-14 4.4E-01 

Ni-59 3.0E-01 

Nb-94 8.8E-01 

Tc-99 1.8E-01 

I-129 1.9E+00 

Np-237 1.1E+00 

Pu-238 1.8E-05 

Pu-239 9.0E-04 

Pu-240 1.5E-03 

Pu-241 5.9E-04 

Pu-242 2.1E-03 

Pu-244 4.5E-03 

Am-241 8.9E-04 

Am-242m 1.9E-03 

Am-243 1.7E-02 

Cm-243 1.5E-04 

Cm-244 2.7E-05 

Cm-245 2.3E-02 

Cm-247 1.0E-01 

Cm-248 3.9E-01 

Cf-249 5.2E-04 

Note: Radionuclides listed are from SRR-CWDA-2009-00045. Only the 10 CFR 61.55 Table 1 radionuclides in the 

tank inventory are listed. 

Table 6.3-2: FTF 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 7.1E-04 

Sr-90 3.3E-03 

Cs-137 1.3E-02 

(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-2009-00045. Only the 10 CFR 61.55 Table 2 radionuclides in the 

tank inventory are listed.
Draft Basis for Section 3116 Determination DOE/SRS-WD-2010-001 

for Closure of F-Tank Farm Revision 0 

at the Savannah River Site September 30, 2010 

Page 6-10 

Table 6.3-3: FTF Transfer Line Site-Specific Factors for 10 CFR 61.55 

Table 1 Radionuclides 

Table 1 

Radionuclide 

Site-Specific 

Factor 

C-14 3.0E+02 

Ni-59 5.2E+00 

Nb-94 6.9E-01 

Tc-99 3.6E-01 

I-129 7.9E+02 

Np-237 3.6E+02 

Pu-238 5.1E-04 

Pu-239 8.6E-02 

Pu-240 2.6E-01 

Pu-241 6.7E-02 

Pu-242 6.5E-01 

Pu-244 3.5E+00 

Am-241 4.4E-03 

Am-242m 1.6E-02 

Am-243 2.0E+00 

Cm-243 7.0E-05 

Cm-244 1.2E-05 

Cm-245 7.4E+03 

Cm-247 8.8E+13 

Cm-248 1.5E+15 

Cf-249 1.7E+09 

Note: Radionuclides listed are from SRS-REG-2007-00002. Only the 10 CFR 61.55 Table 1 radionuclides in the 

transfer line inventory are listed. 

Table 6.3-4: FTF 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 4.5E-03 

Sr-90 1.4E+01 

Cs-137 3.9E+01 

(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 SRS-REG-2007-00002. 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 18 

based on actual final characterization results is used. [SRR-CWDA-2010-00117] This inventory was 

selected on the basis that Tank 18 is the primary contributor to the peak dose in the FTF. The 

contribution of each radionuclide to the sum of the fractions was calculated using the FTF averaging 

expressions presented in Section 6.3.2.1 for a mass or volume basis as necessary. For example, using 
Draft Basis for Section 3116 Determination DOE/SRS-WD-2010-001 

for Closure of F-Tank Farm Revision 0 

at the Savannah River Site September 30, 2010 

Page 6-11 

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: 

4.4E 01

15.2

(9.0E 01 )

8.0E 00 /

1

[ 14] 3 3  

   x

m

Ci

x

Ci m

SOF C = 3.3E-03 

For the mass-based Pu-241 fraction of Class C concentration limit : 

5.9E 04

(15.2 ) (2.06 / ) (1.0E 06 / )

(2.7E 11 )

3.5 03 /

1

[ 241] 3 3 3 3    



   x

m g cm cm m

nCi

x

E nCi g

SOF Pu = 1.5E-03 

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. 

Table 6.4-1: Sum of the Fractions Calculation Using the Closure Inventory 

for Tank 18 (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 9.0E-01 8.0E+00 Ci/m3 3.3E-03 

Ni-59 3.3E-01 2.2E+02 Ci/m3 3.0E-05 

Nb-94 5.5E-04 2.0E-01 Ci/m3 1.6E-04 

Tc-99 9.0E-01 3.0E+00 Ci/m3 3.6E-03 

I-129 2.7E-04 8.0E-02 Ci/m3 4.2E-04 

Np-237 1.9E-01 1.0E+02 nCi/g 6.7E-02 

Pu-238 1.3E+03 1.0E+02 nCi/g 7.5E-03 

Pu-239 2.8E+02 1.0E+02 nCi/g 8.1E-02 

Pu-240 6.5E+01 1.0E+02 nCi/g 3.1E-02 

Pu-241 2.7E+02 3.5E+03 nCi/g 1.5E-03 

Pu-242 2.7E-02 1.0E+02 nCi/g 1.8E-05 

Pu-244 6.2E-06 1.0E+02 nCi/g 8.9E-09 

Am-241 1.6E+02 1.0E+02 nCi/g 4.6E-02 

Am-242m 3.8E-02 1.0E+02 nCi/g 2.3E-05 

Am-243 2.3E+00 1.0E+02 nCi/g 1.3E-02 

Cm-243 1.8E-02 1.0E+02 nCi/g 8.6E-07 

Cm-244 9.8E+01 1.0E+02 nCi/g 8.5E-04 

Cm-245 1.2E-02 1.0E+02 nCi/g 8.8E-05 

Cm-247 2.1E-06 1.0E+02 nCi/g 6.7E-08 

Cm-248 9.5E-05 1.0E+02 nCi/g 1.2E-05 

Cf-249 2.3E-03 1.0E+02 nCi/g 3.8E-07 

 

Sum of the Fractions 2.5E-01 

a

 Values from 10 CFR 61.55 Table 1. 

Note: Inventory values are from SRR-CWDA-2010-00117. Only the 10 CFR 61.55 Table 1 radionuclides in the tank inventory are listed. 

 
Draft Basis for Section 3116 Determination DOE/SRS-WD-2010-001 

for Closure of F-Tank Farm Revision 0 

at the Savannah River Site September 30, 2010 

Page 6-12 

Table 6.4-2: Sum of the Fractions Calculation Using the Closure Inventory for Tank 18 

(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 1.4E-02 (1) NA 

Co-60 1.2E+00 (1) NA 

Ni-63 1.7E+01 7.0E+02 1.1E-06 

Sr-90 3.2E+03 7.0E+03 7.8E-05 

Cs-137 1.2E+04 4.6E+03 1.7E-03 

Sum of the Fractions 1.8E-03 

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-2010-00117. 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 2.5E-01 and 1.8E-03, 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 FTF PA estimated residual inventory for a 3-inch transfer line is used43. [SRSREG-2007-00002] This inventory was selected on the basis that over 99% of the transfer lines located in 

FTF are 3-inch transfer lines or smaller. The contribution of each radionuclide to the sum 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 

3-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: 

3.0E 02

4.38E 04

(4.22E 08 / ) (8.03E 01 )

8.0E 00 /

1

3

2 2

[ 14] 3  

  

   xm

Ci ft ft x

Ci m

SOF C = 2.9E-03 

For the mass-based Pu-241 fraction of Class C concentration limit: 

6.7E 02

3,438

(1.23E 05 / ) (8.03E 01 )

3.5 03 /

1 2 2

[ 241]    

   x

g

nCi ft ft x

E nCi g

SOF Pu = 5.5E-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. 

 

43 The inventory used for this calculation is based on the 3-inch transfer line inventory developed to support the FTF PA. [SRSREG-2007-00002] 
Draft Basis for Section 3116 Determination DOE/SRS-WD-2010-001 

for Closure of F-Tank Farm Revision 0 

at the Savannah River Site September 30, 2010 

Page 6-13 

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

3-inch Transfer Line (10 CFR 61.55 Table 1 Radionuclides) 

Table 1 

Radionuclide 

3-inch Transfer Line 

Inventory44 (Ci/ft2

) 

Class C 

Concentration Limita Fraction of Class C Concentration Limit 

C-14 4.22E-08 8.0E+00 Ci/m3 2.9E-03 

Ni-59 1.66E-06 2.2E+02 Ci/m3 7.2E-05 

Nb-94 3.73E-08 2.0E-01 Ci/m3 2.4E-04 

Tc-99 1.48E-05 3.0E+00 Ci/m3 3.3E-03 

I-129 7.03E-11 8.0E-02 Ci/m3 1.3E-03 

Np-237 7.81E-08 1.0E+02 nCi/g 6.6E-02 

Pu-238 1.82E-04 1.0E+02 nCi/g 2.2E-04 

Pu-239 7.13E-05 1.0E+02 nCi/g 1.4E-02 

Pu-240 2.62E-05 1.0E+02 nCi/g 1.6E-02 

Pu-241 1.23E-04 3.5E+03 nCi/g 5.5E-04 

Pu-242 2.16E-07 1.0E+02 nCi/g 3.3E-04 

Pu-244 1.02E-10 1.0E+02 nCi/g 8.3E-07 

Am-241 5.79E-04 1.0E+02 nCi/g 6.0E-03 

Am-242m 8.32E-07 1.0E+02 nCi/g 3.1E-05 

Am-243 9.45E-08 1.0E+02 nCi/g 4.4E-04 

Cm-243 1.39E-08 1.0E+02 nCi/g 2.3E-09 

Cm-244 2.67E-05 1.0E+02 nCi/g 7.5E-07 

Cm-245 3.43E-10 1.0E+02 nCi/g 5.9E-03 

Cm-247 1.28E-22 1.0E+02 nCi/g 2.6E-05 

Cm-248 2.95E-23 1.0E+02 nCi/g 1.0E-04 

Cf-249 8.00E-25 1.0E+02 nCi/g 3.2E-12 

 

Sum of the Fractions 1.2E-01 

a

 Values from 10 CFR 61.55 Table 1. 

Note: Inventory values are from SRS-REG-2007-00002. Only the 10 CFR 61.55 Table 1 radionuclides in the transfer line inventory are listed. 

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

3-inch Transfer Line (10 CFR 61.55 Table 2 Radionuclides) 

Table 2 

Radionuclide 

3-inch Transfer Line 

Inventory45 (Ci/ft2) 

Class C Concentration 

Limit(Ci/m3

)

 a Fraction of Class C 

Concentration Limit 

H-3 8.08E-07 (1) NA 

Co-60 2.31E-05 (1) NA 

Ni-63 1.38E-04 7.0E+02 1.6E-06 

Sr-90 3.10E-02 7.0E+03 1.1E-01 

Cs-137 5.79E-03 4.6E+03 9.0E-02 

Sum of the Fractions 2.0E-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 SRS-REG-2007-00002. 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 1.2E-01 and 2.0E-01, respectively. Using 

the inventory values assumed for this calculation, the stabilized residuals in this 3-inch transfer line would 

not exceed Class C concentration limits. 

 

44 See footnote 43. 45 See footnote 43. 
Draft Basis for Section 3116 Determination DOE/SRS-WD-2010-001 

for Closure of F-Tank Farm Revision 0 

at the Savannah River Site September 30, 2010 

Page 6-14 

6.5 Conclusion 

As demonstrated by the above discussion, the stabilized FTF waste at closure meets concentration limits 

for Class C low-level waste as set out in 10 CFR 61.55. Nevertheless, DOE is also consulting 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 is specifically requesting in 

this Draft FTF 3116 Basis Document that NRC identify what changes, if any, NRC would recommend to 

DOE’s disposal plans as described in the Draft FTF 3116 Basis Document, and DOE intends to consider 

the NRC recommendations, as appropriate, in the development of DOE’s plans. 
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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 FTF 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 FTF 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 FTF siting, design, use, operation and closure, which ensure compliance 

with 10 CFR 61.44. 

Key Points 

 Based on the FTF 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 FTF boundary is assumed. 

 The FTF PA evaluates a performance period of 10,000 years, and provides additional 

data beyond this period of time, for the purpose of making risk-informed decisions 

related to the closure of FTF. 

 The FTF 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 FTF PA analysis demonstrates compliance with the performance objective in 10 

CFR 61.42 based on compliance with a dose limit of 500 mrem/yr to a future 

hypothetical inadvertent intruder of the closed FTF.

 The DOE regulatory and contractual requirements for FTF 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 FTF waste tanks will be filled with grout to provide long-term stability. 

 The FTF 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 FTF 

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. 
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Legend: 

 Boundary Buffer Zone (100 meters from FTF 

7.1 10 CFR Part 61, Subpart C Performance Objectives 

The 10 CFR 61, Subpart C, Sections 61.40 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, 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. The following subsections discuss the 10 CFR 61.40 through 10 CFR 61.44 

performance objectives. 

7.1.1 10 CFR 61.40 

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. 

The 10 CFR 61.40 general provisions require “reasonable assurance” that exposures are within the limits 

of the subsequent performance objectives for 10 CFR 61.41 through 10 CFR 61.44 and are discussed 

below. 

7.1.2 F-Tank Farm Performance Assessment [SRS-REG-2007-00002] 

The DOE has developed an FTF PA which provides the technical basis and results demonstrating that 

the 10 CFR 61.41 and 10 CFR 61.42 performance objectives will be met after FTF 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 FTF PA 

will include future revisions to incorporate new information, update model codes, etc., as appropriate. 

The FTF PA details the analysis 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 FTF. Individual FTF system 

behaviors are evaluated within the FTF PA for 

various waste tank and ancillary structure 

configurations, including a base (expected) case, 

which provides results reflecting the closure system 

behavior. The FTF 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. 

These calculations were performed to provide 

information over a minimum of 10,000 years. 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 FTF. 

The following general definitions and assumptions 

are used in the FTF PA and will serve as the basis 

for future FTF PA revisions. 

FTF Boundary: The FTF boundary is the line of 

demarcation enclosing the FTF waste tanks (Figure 

7.1-1). 

Figure 7.1-1: FTF Boundary and Buffer Zone 
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Buffer Zone: The buffer zone is the radial area that encompasses the FTF 100 meters from its boundary 

(Figure 7.1-1). 

Institutional Control: Institutional control is a 100-year period in which DOE retains ownership and 

control of FTF such that FTF facility maintenance and controls will be performed to prevent inadvertent 

intrusion and protect public health and the environment46. 

Performance Period: The performance period is the 10,000 years following final closure activities. The 

FTF PA evaluates impacts of closure activities during the performance period (i.e., 10,000 years) after 

closure. The DOE has evaluated for periods beyond 10,000 years in the FTF PA to further inform closure 

decisions47

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 FTF PA includes uncertainty and sensitivity analyses. 

7.1.3 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. 

7.1.3.1 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. 

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 FTF 

100-meter buffer zone. For the purposes of demonstrating compliance with 10 CFR 61.41, the peak allpathways dose at or outside of the 100-meter buffer zone during the 10,000-year performance period will 

be used. 

The pathways for release to a member of the public considered in the FTF PA analyses are discussed 

below. The scenarios are not assumed to occur until after the assumed 100-year institutional control 

period ends. [SRS-REG-2007-00002] 

 

46 To ensure a conservative analysis relative to potential public risk, DOE is using an institutional control period of 100 years. As 

described in Section 7.1.7.4 of this Draft FTF 3116 Basis Document, the SRS Land Use Plan requires Federal ownership and 

control of the site well beyond 100 years after closure. 

47 The FTF PA evaluates a performance period of 10,000 years, and provides additional data beyond this period of time, for the 

purpose of making risk-informed decisions related to the closure of FTF. 
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7.1.3.2 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 

documented in the FTF PA. The following exposure pathways involving the use of contaminated48 well 

water were considered (Figure 7.1-2): 

 direct ingestion of well water, 

 ingestion of milk and meat from livestock (e.g., dairy and beef cattle) that drink well water, 

 ingestion of 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 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) 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 

FTF 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 soil that was irrigated with well water, and 

 direct radiation exposure from radionuclides deposited on the soil that was irrigated with well 

water. 

 

48 Contaminated in this context refers to radioactive materials that have been projected to migrate from the closed FTF. 

Figure 7.1-2: Scenario with Well Water as Primary Source 

1. Direct ingestion of well water 

2. Ingestion of milk and meat from livestock (e.g., 

dairy and beef cattle) that drink well water 

3. Ingestion of vegetables grown in garden soil 

irrigated with well water 

4. Ingestion of milk and meat from livestock (e.g., 

dairy and beef cattle) that eat fodder from 

pasture irrigated with well water 

5. Ingestion and inhalation of well water while 

showering 

6. Direct irradiation during recreational activities 

(e.g., swimming, fishing) from stream water 

7. Dermal contact with stream water during 

recreational activities (e.g., swimming, fishing) 

8. Incidental ingestion and inhalation of stream 

water during recreational activities 

9. Ingestion of fish from the stream water 

10. Direct plume shine 

11. Inhalation 
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The point of assessment for the groundwater wells used in the member of the public scenario is located 

100 meters from the FTF, as shown in Figure 7.1-1. The peak concentrations used to determine the peak 

doses for the member of the public exposure pathways are calculated and documented in the FTF 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 FTF PA conceptual model. The 

conceptual model is used to simulate the performance of the FTF closure system during the 10,000-year 

period and beyond following FTF closure and is comprised of both near-field and far-field models that 

represent the FTF 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. 

The conceptual model used numerous FTF-specific input parameters to represent the FTF closure 

system behavior over time. Many of the input parameters are based on site-specific data (e.g., soil and 

cementitious materials distribution coefficients) used in transport modeling. In addition, site-specific 

information is used to model the behavior of individual barriers within the FTF conceptual model, such as 

the waste tank carbon steel primary tanks and secondary tanks or annular pans, if applicable, and 

cementitious barriers. The models and model inputs used in the FTF conceptual model to calculate 

groundwater concentrations are described in detail in the FTF PA. 

The groundwater peak dose for the member of the public is calculated in the FTF 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 FTF 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 air-pathway 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 FTF. An air-pathway dose was then calculated based on the specific curies of each 

radionuclide assumed to be transported to the surface of FTF. 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 all-pathways peak dose analyses, additional analyses are provided in the FTF PA to 

characterize the context of uncertainty and sensitivity surrounding the FTF 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., inventory sensitivity analysis, distribution coefficient 

sensitivity analysis, alternate configuration sensitivity analysis) performed through deterministic modeling. 

The probabilistic model allows for variability of multiple parameters simultaneously, 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 FTF 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 concerning which parameters are of 

most importance to the FTF PA model. [SRS-REG-2007-00002] 
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7.1.3.3 Results of the Analysis 

The FTF 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 FTF PA, Revision 1, projected the peak 

all-pathways dose to the FTF public receptor (i.e., individual greater than or equal to 100 meters from the 

FTF) to be 2.5 mrem/year, which is less than the 25 mrem/yr performance objective during the 10,000-

year performance period49. The 25 mrem/yr peak all-pathways dose includes the groundwater pathways 

and air pathways associated with all 22 FTF waste tanks and associated ancillary structures with the 

groundwater pathway being the most significant contributor. [SRS-REG-2007-00002] 

The FTF PA modeling performed for the uncertainty and sensitivity analyses was used to determine the 

projected dose to the FTF public receptor for the base case, as well as other tank configurations, over a 

wide range of variability in input parameters. The FTF PA, Revision 1, uncertainty analysis projected a 

4.8 mrem/yr peak of the mean dose to a member of the public within the 10,000-year performance period 

for 1,000 base case realizations. In addition, the FTF PA, Revision 1, uncertainty and sensitivity analysis 

projected a 14 mrem/yr peak of the mean dose to a member of the public within the 10,000-year 

performance period when all modeled tank and ancillary structure configurations are considered. In both 

cases, the peak of the mean dose is less than the 25 mrem/yr performance objective within the 10,000-

year performance period50. [SRS-REG-2007-00002] 

Since there are 29 unique and independent inventory sources modeled in the FTF model, there is 

significant temporal and spatial complexity inherent in the modeling system. The uncertainty and 

sensitivity analyses (in particular the barrier 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 involved51. [SRS-REG-2007-00002] Additional discussion regarding the 

radionuclides most impacting the dose results can be found in Section 5.1 of this Draft FTF 3116 Basis 

Document. 

Compliance with the 25 mrem/yr peak all-pathways dose limit is demonstrated by the peak FTF allpathways base (expected) case dose of 2.5 mrem/yr calculated in the FTF PA for the 10,000-year 

performance period. In addition, the uncertainty and sensitivity analyses included in the FTF 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 during the 10,000-year performance 

period. [SRS-REG-2007-00002] 

7.1.3.4 Conclusion 

As demonstrated in the preceding discussion, reasonable assurance is provided that the performance 

objective at 10 CFR 61.41 will not be exceeded. 

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 as low as reasonable achievable 

(ALARA). The FTF 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).

 

49 The 2.5 mrem/year dose reflects the current results of the base (expected) case through deterministic (PORFLOW) modeling for 

FTF PA, Revision 1, and is estimated to occur at year 10,000 following closure of FTF. The peak FTF all-pathways base case dose 

provided in the FTF PA is not to be considered a limit. As required by DOE Manual 435.1-1, maintenance of the FTF PA will include 

future updates to incorporate new information, update model codes, analysis of actual residual inventories, etc., as appropriate. 50 The 4.8 mrem/yr and 14 mrem/yr doses reflect the current results of the uncertainty analyses projected through probablilistic 

modeling for FTF PA, Revision 1. The peak of the means doses provided in the FTF PA are not to be considered limits. As 

required by DOE Manual 435.1-1, maintenance of the FTF PA will include future updates to incorporate new information, update 

model codes, analysis of actual residual inventories, etc., as appropriate. 

51 Detailed discussion of the uncertainty and sensitivity analyses can be found in Section 5.6 (pages 568 to 749) of the FTF PA 

Revision 1. The impact of the various integrated conceptual model segments on the dose results is detailed in Section 7.1 of the 

FTF PA Revision 1. [SRS-REG-2007-00002] 
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As discussed previously, the FTF PA provides the information to demonstrate compliance with the 25 

mrem all-pathways dose performance objective, including stabilization52 of the residual waste using grout 

to minimize releases to the environment. [SRS-REG-2007-00002] Section 5.2 of this Draft FTF 3116 

Basis Document provides the information to show that the residual waste inventory in the waste tanks will 

be removed to the maximum extent practical. These factors demonstrate reasonable effort to maintain 

releases of radioactivity ALARA. 

7.1.5 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. 

7.1.5.1 General Approach 

Demonstration of compliance with a 500 mrem/yr peak intruder dose will be used to demonstrate 

compliance with the 10 CFR 61.42 performance objective. The requirement of 10 CFR 61.42 exhibits the 

NRC’s intent to protect persons who inadvertently intrude on the waste. The performance objective does 

not place quantitative limits on exposure. However, the 10 CFR 61 Final EIS suggests a dose limit of 

500 mrem/yr for the waste classification scheme in 10 CFR 61.55. Consequently, the NRC uses 500 

mrem/yr dose limit for evaluating impacts to an inadvertent intruder for purposes of 10 CFR 61.4253. 

[NUREG-0945, NUREG-1854] 

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 

FTF 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. 

7.1.5.2 Acute Intruder-Drilling Scenario 

The bounding Acute Intruder scenario analyzed in 

the FTF PA is an Acute Intruder-Drilling Scenario. 

This scenario assumes that after the end of active 

institutional controls a well is drilled within the FTF 

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 

FTF, 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. 

The exposure pathways for this acute drilling 

scenario include (Figure 7.1-3): 

 

52 Stabilization of the FTF waste tanks will be carried out by filling the 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 FTF transfer lines. 

53 For additional information, DOE Manual 435.1-1 also establishes a 100 mrem/yr chronic dose limit for evaluating impacts of an

inadvertent intruder. 

Figure 7.1-3: Acute Intruder-Drilling Scenario 

 1. Inhalation of resuspended drill cuttings 

 2. External exposure to the drill cuttings 

 3. Inadvertent drill cuttings ingestion 
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 inhalation of resuspended drill cuttings, 

 external exposure to the drill cuttings, and 

 inadvertent drill cuttings ingestion. 

7.1.5.3 Chronic Intruder-Agricultural (Post-Drilling) Scenario 

The bounding chronic intruder scenario analyzed in the FTF PA is Chronic Intruder-Agricultural (PostDrilling) Scenario. This scenario assumes that after the end of active institutional controls, a farmer lives 

within the FTF buffer zone and consumes food crops grown, and meat and milk from animals raised 

there, using water from a well drilled within the FTF 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 FTF 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 7.1-4): 

 direct ingestion of well water, 

 ingestion and inhalation of well water while showering, 

 ingestion of milk and meat from livestock (e.g., dairy and beef cattle) 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 pasture 

irrigated with well water, 

 inhalation of well water used for irrigation, 

 inhalation of dust from the soil that was irrigated with well water, 

 ingestion of soil that was irrigated with well water, and 

 direct radiation exposure from radionuclides deposited on the soil that was irrigated with well 

water. 

Figure 7.1-4: Chronic Intruder-Agricultural (Post-Drilling) Scenario 

 1. Direct ingestion of well water 

 2. Ingestion and inhalation of well water while 

showering 

 3. Ingestion of milk and meat from livestock (e.g., dairy 

and beef cattle) that drink well water 

 4. Ingestion of vegetables grown in garden soil irrigated 

with well water and containing contaminated drill 

cuttings 

 5. Ingestion of milk and meat from livestock (e.g., dairy 

and beef cattle) that eat fodder from pasture irrigated 

with well water 

 6. Inhalation of well water used for irrigation 

 7. Inhalation of dust from the soil that was irrigated with 

well water 

 8. Ingestion of soil that was irrigated with well water 

 9. Direct radiation exposure from radionuclides 

deposited on the soil that was irrigated with well 

water 

 10. Direct plume shine 

 11. Inhalation 
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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. 

7.1.5.4 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 FTF 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. 

The point of assessment for the groundwater wells used in the intruder scenario is located one meter 

from the FTF boundary (Figure 7.1-1). The peak concentrations used to determine the peak doses for 

the intruder release exposure pathways are calculated and documented in the FTF 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 FTF PA conceptual model. The 

conceptual model is used to simulate the performance of the FTF closure system during the 10,000-year 

period following FTF closure. The conceptual model is comprised of models that represent the FTF 

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 FTF-specific input parameters to represent the FTF closure 

system behavior over time. Many of the input parameters are based on site-specific data (e.g., soil and 

cementitious materials distribution coefficients) used in transport modeling. In addition, site-specific 

information is used to model the behavior of individual barriers within the FTF 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 FTF conceptual model to calculate 

groundwater concentrations and the waste transfer line drill cutting inventory are described in detail in the 

FTF PA. 

The peak intruder dose is calculated in the FTF 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 FTF PA. 

In addition to the intruder peak dose analyses, additional analyses are provided in the FTF PA to 

characterize the context of uncertainty and sensitivity surrounding the FTF 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., inventory sensitivity analysis, distribution coefficient 

sensitivity analysis, alternate configuration sensitivity analysis) performed through deterministic modeling. 

The probabilistic model allows for variability of multiple parameters simultaneously, 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 FTF 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 
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analyses versus a single approach provides additional information concerning which parameters are of 

most importance to the FTF PA model. [SRS-REG-2007-00002] 

7.1.5.5 Results of the Analysis 

The FTF PA modeling was used to determine an inadvertent intruder dose for comparison with the 10 

CFR 61.42 performance objective. The FTF PA, Revision 1, projected the peak inadvertent intruder (i.e., 

individual within the FTF boundary) dose to be 73 mrem/yr chronic dose and 1.6 mrem/yr acute dose, 

which is less than the 500 mrem/yr performance objective during the 10,000-year performance period54. 

The 500 mrem/yr inadvertent intruder dose considers releases associated with the closure of all 22 waste 

tanks and related ancillary structures within FTF. 

The FTF 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 FTF PA, Revision 1, uncertainty analysis projected a 

639 mrem/yr peak of the mean chronic dose to an inadvertent intruder within the 10,000-year 

performance period for 1,000 base case realizations. This peak of the mean dose is greater than the 500 

mrem/yr performance objective 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 distributions55. 

Since there are 29 unique and independent inventory sources modeled in the FTF model, there is 

significant temporal and spatial complexity inherent in the modeling system. The uncertainty and 

sensitivity analyses (in particular the barrier 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 involved56. Additional discussion regarding the radionuclides most 

impacting the dose results can be found in Section 5.1 of this Draft FTF 3116 Basis Document. 

Demonstration of compliance with the 500 mrem/yr peak inadvertent intruder dose is provided by the fact 

that peak FTF base case inadvertent intruder dose calculated in the FTF PA is less than or equal to 500 

mrem/yr during the 10,000-year performance period. In addition, the uncertainty and sensitivity analyses 

included in the FTF PA provide sufficient information on parameter sensitivities and modeling 

uncertainties during the 10,000-year performance period. [SRS-REG-2007-00002] 

7.1.5.6 Conclusion 

The preceding discussion demonstrates that there is reasonable assurance that the 10 CFR 61.42 

performance objective will not be exceeded after FTF closure. 

7.1.6 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 

 

54 The 73 mrem/yr chronic dose and 1.6 mrem/yr acute dose reflect the current results of the base (expected) case through 

deterministic (PORFLOW) modeling for FTF PA, Revision 1, and is estimated to occur at year 10,000 and 100 respectively, 

following closure of FTF. The peak FTF inadvertent intruder dose values provided in the FTF PA are not to be considered as limits. 

As required by DOE Manual 435.1-1, maintenance of the FTF PA will include future updates to incorporate new information, update

model codes, analysis of actual residual inventories, etc., as appropriate. 

55 The 639 mrem/yr dose reflects the current results of the uncertainty analyses results projected through probablilistic modeling for 

FTF PA, Revision 1. The peak of the means dose provided in the FTF PA is not to be considered a limit. As required by DOE 

Manual 435.1-1, maintenance of the FTF PA will include future updates to incorporate new information, update model codes, 

analysis of actual residual inventories, etc., as appropriate. 

56 Detailed discussion of the uncertainty and sensitivity analyses can be found in Section 5.6 (pages 568 to 749) of the FTF PA 

Revision 1. The impact of the various integrated conceptual model segments on the dose results is detailed in Section 7.1 of the 

FTF PA Revision 1. [SRS-REG-2007-00002] 
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and those for radiological protection of the public and the environment are provided in DOE Order 5400.5. 

[DOE O 5400.5] 

Consistent with NDAA Section 3116(a), the cross-referenced “standards for radiation protection” in 10 

CFR 20 that are considered in detail in this Draft FTF 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)57. 

[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 requirements for DOE facilities and activities. The following subsections show the FTF 

closure meets these dose limits and that doses will be maintained ALARA58. Table 7.1-1 provides a 

crosswalk between the standards set forth in 10 CFR 20 and the applicable DOE requirements. 

 

57 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 Section 3116 of the NDAA, 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 Draft FTF 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 FTF. Therefore, this limit is not discussed further in this Draft FTF 3116 

Basis Document. Likewise, 10 CFR 20.1207 specifies occupational dose limits for minors. However, there will not be minors working at FTF 

who will receive an occupational dose. Therefore, this limit is not discussed further in this Draft FTF 3116 Basis Document. 

58 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.

Table 7.1-1: 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 5400.5 7.1.6.1 Air Emissions Limit for Individual Member of the 

Public 

10 CFR 20.1201(a)(1)(i) 10 CFR 835.202 (a)(1) 7.1.6.2 Total Effective Dose Equivalent Limit for Adult 

Workers 

10 CFR 20.1201(a)(1)(ii) 10 CFR 835.202 (a)(2) 7.1.6.3 Any Individual Organ or Tissue Dose Limit for 

Adult Workers 

10 CFR 20.1201(a)(2)(i) 10 CFR 835.202 (a)(3) 7.1.6.4 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) 7.1.6.5 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 440.1A 7.1.6.6 Limit on Soluble Uranium Intake 

10 CFR 20.1208(a) 10 CFR 835.206 (a) 7.1.6.7 Dose Equivalent to an Embryo/Fetus 

10 CFR 20.1301(a)(1) DOE Order 5400.5 (II.1.a) 7.1.6.8 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 

7.1.6.9 Dose Limits for Individual Members of the Public 

in Unrestricted Areas 

10 CFR 20.1301(b) 10 CFR 835.208 7.1.6.10 Dose Limits for Individual Members of the Public 

in Controlled Areas 
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7.1.6.1 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 5400.5. The DOE is also subject to and complies with the EPA requirement in 40 CFR 61.92, 

which has the same limit59. 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 2008. [WSRC-TR-97-00322, WSRC-TR98-00312, 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] These values (0.04 mrem to 0.11 mrem from 

1997 to 2008) for the SRS operations, not only FTF closure operations, are well below the dose limit 

specified in 10 CFR 20.1101(d) of 10 mrem, 0.1 mSv per year. 

7.1.6.2 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 employees60. For the occupational dose to adults during FTF 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 

dose received by an SRS worker is 1,808 mrem/yr. [SRR-CWDA-2010-00025] The highest total dose 

received by an FTF worker from 1995 - 2009 was 545 mrem. [PIT-MISC-0062, SRR-CWDA-2010-00025] 

The TED to workers from FTF closure is expected to remain well below the DOE/NRC limit since the 

highest dose received by an FTF worker in 1997 was 215 mrem. Since Tank 17 and Tank 20 were 

operationally closed in 1997, this dose is of particular interest in addressing the dose potential for a 

worker during closure operations. Given that the entire FTF highest dose was 215 mrem in 1997, 

reasonable assurance is provided that doses received by a worker during closure activities will be below 

5 rem. [PIT-MISC-0062, SRR-CWDA-2010-00025] 

7.1.6.3 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 – 

 

59 40 CFR 61.92 provides in relevant part as follows: Emissions of radionuclides to the ambient air from DOE 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. 

60 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. 
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(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 similar61 to the dose limit specified in 10 CFR 

835.202 (a)(2). For the occupational dose to adults during FTF 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 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] 

7.1.6.4 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. 

(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 FTF closure, the annual dose limit to 

the eye lens will be controlled using the ALARA principles, and will be below 15 rem/yr. The SRS 

Standard Number 01064 provides the design basis annual occupational exposure limits for the eye lens 

cannot exceed 3 rem/yr, which is well below the NRC limit of 15 rem/yr. [5Q Manual, Chapter 2, WSRCTM-95-1] 

7.1.6.5 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 FTF 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] 

7.1.6.6 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[.] 

 

61 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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In addition to the adult annual dose limits during FTF closure, the soluble uranium intake by an individual 

is controlled to less than 10 milligrams per week. The DOE Order 440.1A 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 FTF 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] 

7.1.6.7 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 FTF 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 

gestation period is unlikely. In addition, personnel dosimetry is provided and used to carefully track 

exposure as controlled by the 5Q Manual, Chapter 2. 

7.1.6.8 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 5400.5 (II.1.a) 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 FTF 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] 

7.1.6.9 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 – 
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(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 FTF 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 mrem/hour (hr) (100 mrem/2,000 hrs = 0.05 

mrem/hr or 0.00005 rem/hr). 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 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/yr62. [5Q 

Manual, Chapter 5] 

7.1.6.10 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 FTF 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/yr63. [5Q Manual, Chapter 5] 

7.1.6.11 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 

FTF 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 

 

62 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.” This is the same as the definition in 10 CFR 835.2 for a 

controlled area. 

63 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 7.1.6.8. 
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(RPP) to include formal plans and measure for applying the ALARA process to occupational exposure as 

further discussed in Section 7.1.6.12.1 of this Draft FTF 3116 Basis Document. 

7.1.6.12 Reasonable Assurance 

Measures that provide reasonable assurance that FTF 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. 

7.1.6.12.1 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 FTF 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 

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 system (S/RIDs). 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. 

7.1.6.12.2 Documented Safety Analysis 

The FTF 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] 
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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. 

7.1.6.12.3 Radiological Design for Protection of Occupational Workers and the Public 

The FTF radiological facilities and facility modifications are designed to meet the requirements of 10 CFR 

835 Subpart K. The SRS 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 per 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. 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. 

7.1.6.12.4 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 
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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] 

7.1.6.12.5 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 dose are maintained as far below the limits as is reasonably achievable. 

[5Q Manual, Chapter 5] 

7.1.6.12.6 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 1995-2009 was 1,808 mrem TED and the highest total 

dose received by an FTF worker from 1995 – 2009 was 545 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-2010-00025] 

In addition, for all work activities, the average TED exposure for workers receiving a TED dose at SRS 

has been 134 mrem/yr over the last five years, 2004-2008. [10-ORAU-0098] 

7.1.7 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 FTF siting, design, use, operation and closure, which ensure 

compliance with 10 CFR 61.44 for the purpose of the Draft FTF 3116 Basis Document. 

7.1.7.1 Siting 

A site characteristics review of demography, geography, meteorology, climatology, ecology, geology, 

seismology, and hydrogeology is presented in Section 3.0 of the FTF PA [SRS-REG-2007-00002] and 

Section 2.0 of this Draft FTF 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 FTF 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 2008, the region of influence 

population was over 550,000. 

The general climate for the SRS region is a humid subtropical climate characterized by short mild winters 

and long warm and humid summers. 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 FTF have grasses, forests, and 

swamps. An abundance of terrestrial, avian, wetlands, and aquatic wildlife live within SRS. FTF 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 
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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: Upper Three Runs and Fourmile Branch. These waters 

eventually feed into the Savannah River. The aquifers of primary interest for FTF are the Upper Three 

Run and Gordon Aquifers. Other aquifers do not contribute to the potential dose to the workers, public or 

intruder and were not included in the models. 

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 most recent 

seismic event within a 50-mile radius of SRS was in March 2009 with a magnitude of 2.6. 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. [SRS-REG-2007-00002] 

7.1.7.2 Design 

The closure design of the FTF waste tanks and ancillary structures provide long-term stability, which is 

consistent with the performance objective. 

There are multiple elements of the FTF 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 FTF design features are described in 

detail in the FTF PA. In addition, the waste tanks and ancillary structures are expected to be covered 

with a closure cap64, 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 FTF 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 FTF closure cap, 

concrete tops on Type I, III and IIIA waste tanks, grout-filled domed roof of the Type IV waste tanks65 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. [SRS-REG-2007-00002] 

7.1.7.3 Use/Operation 

The use/operation of FTF 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 

 

64 The closure cap design described in the FTF PA is based on the best information available at the time the FTF PA was 

developed. [SRS-REG-2007-00002] 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 FTF PA. Any actual closure cap design will be finalized 

closer to the time of FTF 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.

65 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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maintained throughout FTF use and operation. The FTF waste tanks and ancillary structures monitoring 

continues after closure via the site Groundwater Protection Program. [SRNS-TR-2009-00076] 

7.1.7.4 Closure 

Final FTF closure will support long-term stability consistent with this performance objective. In this 

context, long-term stability of the closed FTF 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 

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 FTF 

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 waste 

form66. 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 and Type III/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 FTF 

PA. [SRS-REG-2007-00002]

A closure cap is expected to be designed and constructed over the the FTF site following grouting of the 

FTF waste tanks and backfilling of the ancillary structures, as necessary67. The closure cap design 

described in the FTF PA is based on the best information available at the time the FTF PA was 

developed. [SRS-REG-2007-00002] 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 FTF 

PA. The actual closure cap design will be finalized closer to the time of FTF 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. [SRS-REG-2007-00002] 

The DOE will maintain GSA ownership, which includes the FTF. The SRS Land Use Plan requires 

Federal ownership and control of the site well beyond 100 years after tank closure. [PIT-MISC-0041] 

7.1.7.5 Conclusion 

As previously discussed, the site conditions do not present hazards that impact FTF stability. In addition, 

the FTF closure methods will result in a facility closure that does not require ongoing maintenance. 

Therefore, closure of the FTF complies with 10 CFR 61.44 performance objective. 

 

66 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 FTF transfer lines. 

67 Final closure of FTF will performed per the requirements of the FFA for SRS (e.g., Section IX.E.(2). [WSRC-OS-94-42] 
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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 

FTF waste tanks and ancillary structures, as appropriate, will be performed pursuant to a 

State-approved closure plan. 

Section Contents 

This section discusses the State of South Carolina regulation of the waste tanks and 

ancillary structures and shows that removal from service of the FTF waste tanks and 

ancillary structures will be pursuant to State-approved Closure Modules, consistent with 

the FTF General Closure Plan (GCP). 

Key Points 

 The FTF 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 FTF waste tanks and 

ancillary structures, referred to as the FTF General Closure Plan, requires approval by 

SCDHEC. 

 A specific Closure Module for each waste tank and ancillary structure, consistent with 

the requirements of the FTF 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 FTF waste storage and removal operations are governed by an SCDHEC industrial wastewater 

construction permit issued March 3, 1993. [DHEC_03-03-1993] 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 regulations implementing the Act. The State of South Carolina 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 FTF GCP addresses the State’s regulatory authority relevant to removing the FTF waste tanks and 

ancillary structures from service. The GCP sets forth the general protocol by which DOE intends to 

remove from service the FTF waste tanks and ancillary structures to protect human health and the 

environment. The FTF GCP requires approval by SCDHEC. Prior to approval by SCDHEC, the FTF 

GCP will be made available to the public for review and comment. [LWO-RIP-2009-00009] 
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Before final stabilization activities commence68, individual waste tank and ancillary structure closure 

plans, referred to as Closure Modules, describing closure details will be developed and submitted to 

SCDHEC for approval69. 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, 

justification that additional waste removal is not technically practical from an engineering perspective70,

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 FTF waste tanks and ancillary structures will be removed from service, stabilized 

and operationally closed pursuant to State-approved Closure Modules, consistent with the FTF General 

Closure Plan. Thus, the FTF 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.” 

 

68 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. 69 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. 70 Per the FFA, the waste tanks will be cleaned until Department of Energy-Savannah River Operations Office, SCDHEC and EPA 

agree that waste removal may cease. The Closure Module provides the basis for agency agreement. 
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9.0 CONCLUSION 

As demonstrated in the preceding sections of this Draft FTF 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 FTF at the time of closure are not high-level waste pursuant to 

the criteria set forth in NDAA Section 3116(a). This Draft FTF 3116 Basis Document will be finalized after 

DOE has completed consultation with NRC and, although not required by NDAA Section 3116, after 

public review and comment. 
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10.0 REFERENCES 

Note 1: Reference numbers with a REDACTED designator were classified: “Official Use Only, 

Exemption 2 — Not Releasable to the Public or Foreign Nationals without prior approval from 

DOE-SR.” However, the restricted information does not pertain to this document, thus a redacted copy 

with the Restricted Information Removed has been provided. 

Note 2: References identified as (Copyright) were used in the development of this document, but are 

protected by copyright laws. No part of the publication may be reproduced in any form or by any means, 

including photocopying or electronic transmittal in any form by any means, without permission in writing 

from the copyright owner. 

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for Closure of F-Tank Farm Revision 0 

at the Savannah River Site September 30, 2010 

Page 10-3 

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for Closure of F-Tank Farm Revision 0 

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for Closure of F-Tank Farm Revision 0 

at the Savannah River Site September 30, 2010 

Page 10-6 

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Savannah River Remediation, Savannah River Site, Aiken, SC, June 2010. 
Draft Basis for Section 3116 Determination DOE/SRS-WD-2010-001 

for Closure of F-Tank Farm Revision 0 

at the Savannah River Site September 30, 2010 

Page 10-7 

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WSRC-STI-2007-00184, Denham, M., et. al, FTF Closure Cap Concept and Infiltration Estimates, 

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SRS Meteorological Database, Savannah River Site, Aiken, SC, December 2007. 

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Aiken, SC, 2008. 

WSRC-STI-2008-00244, Jones, W. E., and Phifer, M. A., Saltstone Disposal Facility Closure Cap 

Concept and Infiltration Estimates, Savannah River Site, Aiken, SC, May 2008. 
Draft Basis for Section 3116 Determination DOE/SRS-WD-2010-001 

for Closure of F-Tank Farm Revision 0 

at the Savannah River Site September 30, 2010 

Page 10-8 

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1997. 

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SC, 1998. 

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SC, 1999. 

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SC, 2000. 

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Aiken, SC, 2002. 

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Aiken, SC, 2004. 

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Savannah River Site, Aiken, SC, Rev. 0, November 10, 2004. 
Draft Basis for Section 3116 Determination DOE/SRS-WD-2010-001 

for Closure of F-Tank Farm Revision 0 

at the Savannah River Site September 30, 2010 

Page 10-9 

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Aiken, SC, 2005. 

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Aiken, SC, 2007. 

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Tanks 18 & 19, Savannah River Site, Aiken, SC, Rev. 0, November, 2007. 
Draft Basis for Section 3116 Determination DOE/SRS-WD-2010-001 

for Closure of F-Tank Farm Revision 0 

at the Savannah River Site September 30, 2010 

Page 11-1 

11.0 GLOSSARY 

Actinide Group of elements of atomic number 89 through 103 including thorium, 

uranium, neptunium, plutonium, americium and curium. 

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 waste tanks, providing a location for 

collection of any leakage from the primary tank. 

Base Case The waste tank system configuration modeling case within the FTF PA that 

represents the best estimate of expected conditions for the FTF 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. 

Chromated Cooling 

Water 

Coolant comprised of choromate-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. 

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 Draft FTF 3116 Basis Document, contamination refers to 

radioactive materials that have been projected to migrate from the closed FTF. 

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. 

C 

B 

A 
Draft Basis for Section 3116 Determination DOE/SRS-WD-2010-001 

for Closure of F-Tank Farm Revision 0 

at the Savannah River Site September 30, 2010 

Page 11-2 

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 oxidation/reduction reaction (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 easier and pump easier. 

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 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 in H Canyon for separation of special nuclear 

materials and enriched uranium from irradiated targets. 

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 FTF such 

that FTF facility maintenance and controls will be performed to prevent 

inadvertent intrusion and protect public health and the environment. 

I 

H 

G 

F 

E 

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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. 

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. Section 3116 of the NDAA 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, being about 10,000 times stronger than acetic 

acid. 

Oxidation Potential (Eh) The measure of a material to oxidize or lose electrons. 

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,” 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 FTF pump pits house a pump tank with the pump pits providing secondary 

containment for pump tanks. The pump tanks have a nominal capacity of 

8,000 gallons each. The pump tanks installed in FTF are all of the same basic 

size (8.5 feet tall, 12 feet in diameter). 

PUREX Process The Plutonium Uranium Extraction process used in the F-Canyon 

reprocessing facility to extract uranium and plutonium from aluminum-clad, 

depleted uranium targets which had been irradiated in the site’s nuclear 

production reactors. 

P 

O 

N 

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Residual Radioactivity Radioactivity in structures, materials, soils, groundwater and other media at a 

site remaining after closure. 

Riser The risers through the tank tops provide for 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 entrained soluble 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. 

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 waste 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. 

Silica fume Also known as microsilica, is a byproduct of the reduction of high-purity quartz 

with coke in electric arc furnaces in the production of silicon and ferrosilicon 

alloys. Silica fume is used as an addition in Portland cement concretes to 

improve properties. It has been found that silica fume improves compressive 

strength, bond strength and abrasion resistance. Addition of silica fume also 

reduces the permeability of concrete to chloride ions, which protects 

concrete’s reinforcing steel from corrosion. 

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 Mixture of at least two liquid components or of at least one solid and a liquid 

component. 

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. 

S 

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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 liner of the waste tank. 

Viscosity Rheological quality of fluids describing the resistance to flow. 

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 

V 

U 
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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 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 March 2010, approximately 37,400,000 gallons of radioactive waste were stored 

in the waste tanks, most of this from separation 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 

being treated and disposed of as a non-hazardous, cementitious waste form (i.e., 

saltstone) in the Saltstone Disposal Facility (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 fuels resulted in the generation of large quantities of radioactive liquid waste, which 

is currently stored onsite 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 spent 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. Because of differences in the PUREX and HM processes, 
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the chemical compositions of principal sludge components (iron, aluminum, uranium, manganese, nickel, 

and mercury) also vary over a broad range between these sludge components. 

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 49 active 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 waste tanks are the newest waste tanks and 

were placed into operation between 1969 and 1986. There are a total of 27 Type III/IIIA waste tanks. 

Figure A.2-1 shows Type III waste 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 waste 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 

General Closure Plan. [PIT-MISC-0002, 

PIT-MISC-0004] 

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 waste tanks have 

developed leak sites. Five of the 13 

waste tanks with a history of leakage are 

located in FTF (Tank 1, Tank 5, Tank 6, 

Tank 19 and Tank 20). [SRR-STI-2010-

00283] Sufficient waste has been 

removed from these waste tanks such 

that 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 

Figure A.2-1: Waste Tanks Under Construction 
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and was associated with Tank 16 in HTF. [DPSOX-5954] 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 March 31, 2010, approximately 37,400,000 

gallons of liquid radioactive waste was stored in 49 active 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 waste 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 waste tank space must be reserved as 

contingency space should a new waste tank leak be realized. [SRR-LWP-2010-00040] 

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. Three 

evaporator systems are currently operating at 

SRS, the 2H, 3H, and 2F systems. Evaporator 

operations are currently impacted due to limited 

salt waste storage space. [SRR-LWP-2009-

00001] 

As of March 31, 2010, approximately 

358,000,000 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-2010-00040] 

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 

3,000,000 gallons (8% of total) of waste volume 

but contains approximately 175,000,000 curies 

(49% of total) of the radioactivity. The salt 

waste makes up the remaining 34,400,000 

gallons (92% of total) of waste and contains 

approximately 183,000,000 curies (51% of total) 

of the radioactivity. Of that salt waste, the 

supernate accounts for approximately 

18,500,000 gallons and 171,000,000 curies of 

the 183,000,000 curies total salt waste related 

activity. The saltcake accounts for 

approximately 15,900,000 gallons and 

12,000,000 curies of the remaining salt waste. 

The sludge contains the majority of the longlived (half-life > 30-years) radionuclides (e.g., 

actinides) and strontium. Figure A.2-4 shows 

the breakdown of this waste. [SRR-LWP-2010-

00040] 

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 

March 2010. [SRR-LWP-2010-00040] 

Figure A.2-3: Sludge Component of Radioactive 

Waste 

Figure A.2-4: FTF and HTF Waste Tank Composite Inventory 

34.4 Mgal 

Volume 

37.4 Mgal 

Curies 

183 MCi 

(51%) 

175 MCi 

(49%) 

358 MCi 

171 MCi 

(48%) 

Sludge

(92%) 

3 Mgal 

(8%) 

18.5 Mgal 

(49%) Salt Supernate

12 MCi 

Saltcake (3%) 

15.9 Mgal 

(43%) 

Inventory values as of March 31, 2010 

[SRR-LWP-2010-00040] 
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Approximately 95% 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 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) is sufficient to produce a 

salt waste that meets the Saltstone Production Facility (SPF) WAC. 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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In addition, DOE is exploring the viability of 

augmenting salt processing capabilities 

using rotary microfilters and ion exchange 

columns installed in Type IIIA waste tank 

risers. The salt solution would be struck 

with monosodium titanate similar to the 

actinide removal processes associated 

with Interim Salt Processing and SWPF, 

and the insoluble solids and monosodium 

titanate would be filtered using rotary 

microfiltratrion technology. This clarified 

salt solution would then pass through ion 

exchange column(s) designed to target the 

removal of cesium. 

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 volumereduced 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) site. [SRR-LWP-2009-00001] 

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. 

Figure A.3-1: Vitrification Canister Prior to Use

Figure A.3-2: Sample of Vitrified 

Radioactive Glass 
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Disposal Cells 2A and 2B 

SPF Vault 4 

Vault 1 

The current active disposal unit (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 

3-inch separation gap between the units. Each unit is further divided into six cells, with each cell 

measuring approximately 100 feet x 100 feet. Thus, Vault 4 is comprised of 12 cells each approximately 

100 feet x 100 feet. 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 3-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. [SRS-REG-2007-00002] 

Future disposal cells are planned to be nominally 150 feet diameter by 22 feet high each and will be 

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). The first of the future disposal units, comprised of 

Disposal Cells 2A and 2B, is currently under construction in the SDF (Figure A.3-3). [C-CG-Z-00030] 

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. [WSRC-STI-2008-00244] 

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-LWP-2009-00001] 

Figure A.3-3: Saltstone Facility