DOE/NE-ID-11226

Revision 0

Basis for Section 3116 

Determination 

for the 

Idaho Nuclear Technology and 

Engineering Center Tank Farm 

Facility

November 2006 
DOE/NE-ID-11226

Revision 0

Basis for Section 3116 Determination 

for the 

Idaho Nuclear Technology and Engineering Center 

Tank Farm Facility 

November 2006 
ii 

CONTENTS 

ACRONYMS..............................................................................................................................................vii

1. INTRODUCTION AND PURPOSE.................................................................................................. 1

2. BACKGROUND................................................................................................................................ 5

2.1 Tank Farm Facility Description............................................................................................. 9

2.2 Origin and Management of TFF Wastes ............................................................................. 15

2.3 Tank Farm Facility Closure Activities and Status............................................................... 23

2.3.1 Cleaning Approach for Stainless Steel Tanks and Vaults................................. 25

2.3.2 Results of Tank and Ancillary Equipment Cleaning from 2002 to 2005.......... 28

2.4 Residual Inventory in Tank Farm Facility Structures, Systems, and Components ............. 30

2.4.1 Models and Calculations Used to Develop Residual Inventory at Closure....... 31

2.4.2 Residual Inventory at Closure in 300,000-gal Tanks........................................ 33

2.4.3 Residual Inventory at Closure for 30,000-gal Tanks ........................................ 36

2.4.4 Residual Inventory at Closure for Sandpads ..................................................... 37

2.4.5 Residual Inventory at Closure for Contaminated Piping, Encasements, and 

Valve Boxes ...................................................................................................... 41

3. SECTION 3116 OF THE RONALD W. REAGAN NATIONAL DEFENSE AUTHORIZATION 

ACT FOR FISCAL YEAR 2005...................................................................................................... 43

4. WASTE DOES NOT REQUIRE PERMANENT ISOLATION IN A DEEP GEOLOGIC 

REPOSITORY FOR SPENT FUEL OR HIGH-LEVEL RADIOACTIVE WASTE ...................... 45

5. WASTE HAS HAD HIGHLY RADIOACTIVE RADIONUCLIDES REMOVED TO THE 

MAXIMUM EXTENT PRACTICAL.............................................................................................. 47

5.1 Highly Radioactive Radionuclides ...................................................................................... 47

5.1.1 Approach........................................................................................................... 47

5.1.2 Radionuclides from Performance Assessment.................................................. 48

5.2 Removal to the Maximum Extent Practical......................................................................... 50

5.2.1 Percentage of Radionuclides Removed............................................................. 54

5.2.2 Cost of Developing a More Efficient Cleaning Technology............................. 56

5.2.3 Conclusions....................................................................................................... 57

6. RADIONUCLIDE CONCENTRATIONS OF STABILIZED WASTE.......................................... 59

6.1 300,000-gal Tanks ............................................................................................................... 61

6.1.1 300,000-gal Tank Waste Concentration............................................................ 63
iii 

6.1.2 Summary and Conclusions................................................................................ 65

6.2 300,000-gal Tank Vaults ..................................................................................................... 65

6.2.1 Summary and Conclusions................................................................................ 69

6.3 30,000-gal Tanks ................................................................................................................. 69

6.3.1 Summary and Conclusions................................................................................ 71

6.4 Grouted Piping, Encasements, and Valve Boxes ................................................................ 71

6.4.1 Summary and Conclusions................................................................................ 73

6.5 Sensitivity Analyses ............................................................................................................ 73

6.6 Conclusions ......................................................................................................................... 73

7. WASTE WILL BE DISPOSED OF IN ACCORDANCE WITH PERFORMANCE 

OBJECTIVES IN 10 CFR 61, SUBPART C ................................................................................... 74

7.1 Analysis of Performance ..................................................................................................... 76

7.1.1 Institutional Control Period............................................................................... 77

7.1.2 Post-Institutional Control Period....................................................................... 77

7.1.3 Exposure Pathways ........................................................................................... 77

7.2 Models and Assumptions .................................................................................................... 78

7.2.1 Source Term Model........................................................................................... 78

7.2.2 Groundwater Flow and Transport Modeling..................................................... 79

7.2.3 Chronic Intruder Exposure Scenarios ............................................................... 82

7.3 Acute Intruder-Drilling Scenario Definition ....................................................................... 82

7.4 Acute Intruder-Construction Scenario................................................................................. 84

7.5 Chronic Intruder Post-Drilling Scenario Definition ............................................................ 84

7.6 Chronic Intruder Post-Construction Scenario Definition .................................................... 85

7.7 Estimated Radiation Dose Using the Current Estimate of the Residual Inventory at 

Closure................................................................................................................................. 86

7.8 10 CFR 61.40, “General Requirement”............................................................................... 87

7.9 10 CFR 61.41, “Protection of the General Population from Releases of Radioactivity” .... 88

7.10 10 CFR 61.42, “Protection of Individuals from Inadvertent Intrusion” .............................. 90

7.11 10 CFR 61.43, “Protection of Individuals during Operations” ........................................... 94
iv 

7.11.1 Air Emissions Limit for Individual Member of the Public 

[10 CFR 20.1101(d)]......................................................................................... 95

7.11.2 Total Effective Dose Equivalent Limit for Adult Workers 

[10 CFR 20.1201(a)(1)(i)]................................................................................. 96

7.11.3 Any Individual Organ or Tissue Dose Limit for Adult Workers 

[10 CFR 20.1201(a)(1)(ii)]................................................................................ 96

7.11.4 Annual Dose Limit to the Lens of the Eye for Adult Workers 

[10 CFR 20.1201(a)(2)(i)]................................................................................. 97

7.11.5 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)].............................. 97

7.11.6 Limit on Soluble Uranium Intake [10 CFR 20.1201(e)] ................................... 98

7.11.7 Dose Equivalent to an Embryo/Fetus [10 CFR 20.1208(a)] ............................. 98

7.11.8 Dose Limits for Individual Members of the Public [10 CFR 20.1301(a)(1)] ... 98

7.11.9 Dose Limits for Individual Members of the Public [10 CFR 20.1301(a)(2)] ... 99

7.11.10 Dose Limits for Individual Members of the Public [10 CFR 20.1301(b)]........ 99

7.11.11 As Low As Reasonably Achievable (10 CFR 20.1003).................................... 99

7.11.12 Reasonable Assurance..................................................................................... 100

7.12 10 CFR 61.44, “Long-Term Stability of Disposal Site”.................................................... 102

7.12.1 Siting ............................................................................................................... 102

7.12.2 Design ............................................................................................................. 103

7.12.3 Use/Operation ................................................................................................. 104

7.12.4 Closure ............................................................................................................ 104

7.13 Conclusions ....................................................................................................................... 105

8. STATE-APPROVED CLOSURE PLANS .................................................................................... 106

9. CONCLUSION .............................................................................................................................. 110

10. REFERENCES............................................................................................................................... 111

11. DRAWING..................................................................................................................................... 121

Appendix A—INTEC Tank Farm Facility Closure Supporting Tables and Photographs........................A-1 

Appendix B—Summary of Analysis for Selecting Tank Cleaning Technologies.................................... B-1 

Appendix C—TFF Grout Design Mix and Quality Assurance/Quality Control Testing ......................... C-1 

Appendix D—Sensitivity Evaluations......................................................................................................D-1 

Appendix E—Residual Inventory and Grout Volume used in Final 300,000-gal Tank Concentration 

Calculations .................................................................................................................................... E-1 

Appendix F—Tables for Sum of the Fractions Calculations Discussed in Sensitivity Evaluations..........F-1 

Appendix G—Comparison of Final Waste Form Calculation Methodology to Fundamental 

Principles on Concentration Averaging in the NRC Draft Interim Guidance ................................G-1 
v 

Appendix H—Past NRC Review Recommendations ...............................................................................H-1 

Appendix I—Management Control Systems ..............................................................................................I-1 

FIGURES 

1. Idaho National Laboratory site map ...................................................................................................7

2. Tank Farm Facility location at INTEC...............................................................................................8

3. Plan view of the Tank Farm Facility ..................................................................................................9

4. Cross-sectional view of a typical tank with cooling coils ................................................................11

5. Vault designs used at the INTEC TFF..............................................................................................12

6. Photograph showing the internal structure and cooling coils of Tank WM-185 during 

construction ......................................................................................................................................13

7. Artist’s depiction of tanks to be included in the INTEC TFF closure..............................................14

8. Spent fuel reprocessing at INTEC....................................................................................................16

9. Volumes in TFF 300,000-gal tanks as of April 30, 2005 .................................................................22

10. Typical tank cleaning system ...........................................................................................................26

11. Photographs of the interior of TFF Tank WM-183 before tank cleaning.........................................29

12. Photograph of the interior of TFF Tank WM-183 after tank cleaning .............................................30

13. Radioactivity (counts per minute per gallon) measured during cleaning operations........................53

14. Schematic of the grout placement sequence.....................................................................................63

15. The DUST-MS conceptual model for release from the TFF vaults and tanks .................................79

16. Illustration of the two-dimensional modeling slice used in PORFLOW..........................................80

17. Graphical representation of the acute intruder-drilling scenario ......................................................83

18. Chronic intruder post-drilling scenario.............................................................................................85

TABLES 

1. Residual inventory at closure for Tank WM-182.............................................................................34

2. Residual inventory at closure for Tank WM-106.............................................................................37

3. Sandpad residual inventory at closure (Ci per sandpad) ..................................................................40
vi 

4. Residual inventory at closure for piping...........................................................................................42

5. List of highly radioactive radionuclides ...........................................................................................49

6. Percentage of highly radioactive radionuclides removed from all tanks and ancillary equipment ..55

7. Cost estimate of TFF cleaning system..............................................................................................56

8. All-pathways dose reduction and cost per mrem..............................................................................57

9. Radionuclide concentrations in the final Tank WM-182 grouted waste form (Table 1 of 

10 CFR 61.55) ..................................................................................................................................64

10. Radionuclide concentrations in the final Tank WM-182 grouted waste form (Table 2 of 

10 CFR 61.55) ..................................................................................................................................64

11. Radionuclide concentrations in the grouted vaults (Table 1 of 10 CFR 61.55) ...............................67

12. Radionuclide concentrations in the (Table 2 of 10 CFR 61.55) grouted vaults. ..............................67

13. Sum of the fractions for 32.5 m3 of grout in the tank vault (Table 1 of 10 CFR 61.55)...................68

14. Radionuclide concentrations in a final 30,000-gal tank grouted waste form (Table 1 of 

10 CFR 61.55) ..................................................................................................................................70

15. Radionuclide concentrations in a final 30,000-gal tank grouted waste form (Table 2 of 

10 CFR 61.55) ..................................................................................................................................70

16. Sum of the fractions for 30,000-gal tanks using only the mass of steel (Table 1 of 

10 CFR 61.55) ..................................................................................................................................71

17. Radionuclide concentrations in the piping (with grout) (Table 1 of 10 CFR 61.55)........................72

18. Radionuclide concentrations in the piping (with grout) (Table 2 of 10 CFR 61.55)........................73

19. Comparison of performance assessment results to predicted dose at closure ..................................86

20. Groundwater all-pathways dose (less than 25 mrem/yr) ..................................................................89

21. Acute intruder-construction dose (less than 500 mrem)...................................................................91

22. Acute intruder-drilling dose (less than 500 mrem)...........................................................................92

23. Chronic intruder-construction dose ..................................................................................................93

24. Chronic intruder-drilling dose comparison.......................................................................................93
vii 

ACRONYMS 

ALARA as low as reasonably achievable 

ASME American Society of Mechanical Engineers 

ASTM ASTM International (formerly American Society for Testing and Materials) 

CERCLA Comprehensive Environmental Response, Compensation, and Liability Act 

CFR Code of Federal Regulations 

CWI CH2M-WG Idaho, LLC 

DCF dose conversion factor 

DEQ State of Idaho Department of Environmental Quality 

DOE U.S. Department of Energy 

DOE Idaho U.S. Department of Energy Idaho Operations Office 

DQA data quality assessment 

DQO data quality objective 

DUST-MS Disposal Unit Source Term-Multiple Species 

EIS Environmental Impact Statement 

EPA U.S. Environmental Protection Agency 

ESRP Eastern Snake River Plain 

FR Federal Register 

HLW high-level waste 

HWMA Hazardous Waste Management Act 

ICRP International Commission on Radiological Protection 

IDAPA Idaho Administrative Procedures Act 

INEEL Idaho National Engineering and Environmental Laboratory 

INEL Idaho National Engineering Laboratory 

INL Idaho National Laboratory 

INTEC Idaho Nuclear Technology and Engineering Center 
viii 

LLW low-level waste 

MCP management control procedure 

M&O management and operating 

NA not applicable 

ND not detected 

NDAA National Defense Authorization Act 

NRC Nuclear Regulatory Commission 

PA performance assessment 

PEWE Process Equipment Waste Evaporator 

QA quality assurance 

RAI request for additional information 

RCRA Resource Conservation and Recovery Act 

ROD Record of Decision 

RPP radiation protection program 

SAP sampling and analysis plan 

SBW sodium-bearing waste 

SNF spent nuclear fuel 

TEDE total effective dose equivalent 

TER technical evaluation report 

TFA Tanks Focus Area 

TFF Tank Farm Facility 

USC United States Code 

WAC waste acceptance criteria 
1 

Basis for Section 3116 Determination 

for the 

Idaho Nuclear Technology and Engineering Center 

Tank Farm Facility 

1. INTRODUCTION AND PURPOSE 

Key Points 

• This 3116 Basis Document addresses the disposal of stabilized residuals in the TFF, and the TFF 

tank system, and disposal of the tanks, vaults, and associated piping and ancillary equipment at 

INTEC. The TFF tank system comprises the eleven 300,000-gal tanks, four 30,000-gal tanks, 

and the vaults, piping, structures, and ancillary equipment associated with these tanks. 

• The scope of this 3116 Basis Document does not include any other facilities or systems at 

INTEC or the INL Site such as the evaporators, the New Waste Calcining Facility, the calcine 

storage bin sets, the 900,000 gal of stored sodium-bearing waste, contaminated soils, or the 

treated sodium-bearing waste for off-Site disposal. 

• The planned TFF closure date of 2012 is assumed throughout this 3116 Basis Document.a

• The TFF tank system closure process includes cleaning activities and stabilization activities. 

After the sodium-bearing waste is removed, the tanks, vaults, piping, structures, and ancillary 

equipment will be cleaned. Then the system will be stabilized by filling the components with 

grout. This distinction between cleaning activities and stabilization activities (grouting) applies 

throughout this 3116 Basis Document. The term “final closure” is used to describe the 

completion of both cleaning and stabilization activities. 

• Stabilization of individual tanks, vaults, and other TFF components may take place once 

cleaning of those individual components is completed and need not be delayed until the cleaning 

of all tanks, vaults, and other TFF components is completed. 

• This 3116 Basis Document addresses only radioactive constituents of the TFF residuals; 

hazardous constituents of this waste are addressed in State-approved closure plans. 

The Secretary of Energy is making a 3116 Determination pursuant to Section 3116 of the Ronald 

W. Reagan National Defense Authorization Act for Fiscal Year 2005 (NDAA) (Public Law 108-375, 

2004). The Secretary of Energy’s 3116 Determination and this 3116 Basis Document concern the disposal 

of grouted residual waste in the tank systems at the Idaho Nuclear Technology and Engineering Center 

(INTEC) Tank Farm Facility (TFF) that are being closed in accordance with U.S. Department of Energy 

(DOE) requirements for managing a radioactive waste storage facility (DOE O 435.1, 2001) and 

Hazardous Waste Management Act (HWMA) (State of Idaho 1983)/Resource Conservation and Recovery 

Act (RCRA) (42 United States Code [USC] 6901 et seq. 1976) requirements for an interim status tank 

system. The DOE is closing the TFF tank system in response to a January 1990 Notice of Noncompliance 

and subsequent Consent Order (State of Idaho et al. 1992). The TFF consists of eleven 300,000-gal 

 

a. The numbers and percentages in this 3116 Basis Document are either rounded numbers and percentages and/or are DOE’s best 

estimates at this time. The numbers, percentages, and dates in this 3116 Basis Document should be viewed as approximate 

numbers, percentages, and dates. 
2 

belowgrade stainless steel tanks in unlined concrete vaults, four 30,000-gal belowgrade stainless steel 

tanks, and associated ancillary equipment and piping. Historically, the TFF tanks were used to store 

various INTEC wastes, including those from spent nuclear fuel (SNF) reprocessing (first-, second-, and 

third-cycle reprocessing wastes), decontamination waste, laboratory waste, and contaminated liquids from 

other INTEC operations. In general, because of significantly higher radioactivity levels, first-cycle 

reprocessing wastes were segregated from the other types of liquid waste. These other tank wastes, 

referred to as sodium-bearing waste (SBW) because of their high sodium levels, were made up of wastes 

other than first-cycle reprocessing wastes, were generally much lower in radioactivity, and had a 

significantly different chemical composition than first-cycle reprocessing wastes. 

Spent nuclear fuel reprocessing was curtailed in 1992, and the first-cycle extraction process wastes 

stored in the TFF were removed and solidified (by calcination) by February 1998 (DOE 2002).b

 Since that 

time, the tanks used for storing the first-cycle wastes have been reused to store SBW. Because of the 

design of the tank’s originally installed waste removal system, a tank emptied using this system still 

contains several thousand gallons of waste on the bottom. Thus, the reuse of the TFF tanks that were used 

to store first-cycle extraction process wastes to store SBW, along with the various historical transfers of 

wastes among TFF tanks to manage TFF volumes, has resulted in the current volume of SBW containing 

a small percentage (about 1% by volume) of first-cycle extraction process wastes (Loos 2004). 

The TFF is being closed in phases to support continued INTEC operations. The closure process 

comprises tank system cleaning and stabilization activities. Over the last several years, TFF operations 

have included consolidating the remaining SBW tank wastes into the minimum number of tanks 

necessary and commencing cleaning activities in the emptied tanks. Tank cleaning began in late 2002, 

and the project has progressed on schedule. Final closure is planned for 2012. As of July 2005, seven of 

the 300,000-gal tanks, the four 30,000-gal tanks, and associated ancillary equipment had been cleaned. 

None of the remaining tanks have been cleaned as of November 2006. Approximately 900,000-gal of 

SBW remain stored at the TFF awaiting future treatment. This SBW is stored in three 300,000-gal tanks; 

one 300,000-gal tank is maintained as a spare.c

 After completing tank and ancillary equipment cleaning 

operations, a small amount of residual radioactive waste that cannot be removed remains. As each tank is 

cleaned, these residuals are sampled and analyzed to confirm that radionuclide and hazardous constituent 

concentrations meet performance objectives to ensure protection of the public and the environment. After 

cleaning activities are completed for individual tanks, vaults, and other TFF components, DOE plans to 

stabilize each of these individual TFF components by filling them with grout. Upon completion of the 

phased cleaning and grouting of all the tanks and ancillary equipment in the TFF tank system, final 

closure of the TFF would be accomplished. The first grouting activity may begin as early as November 

2006. Throughout this 3116 Basis Document, grouting (stabilization) activities are distinct from cleaning 

activities and not considered part of the cleaning process. The term “final closure” is used to describe the 

completion of both cleaning and stabilization activities. 

Section 3116(a) of the NDAA specifies that the term “high-level radioactive waste” does not 

include radioactive waste that results from reprocessing SNF if the Secretary of Energy determines, in 

consultation with the Nuclear Regulatory Commission (NRC), that the waste meets certain criteria, which 

specify that the waste must not require disposal in a deep geologic repository; must have had highly 

 

b. This calcined waste is planned to be disposed of in the geologic repository at Yucca Mountain, Nevada, after the repository is 

licensed and operational. The calcined waste is not addressed in this 3116 Basis Document because, among other things, Section 

3116(c), in conjunction with 3116(d), provides that Section 3116 does not apply to waste transported from the State of Idaho. 

c. The SBW, which will be or has been removed from the tanks, and treated for off-Site disposal is not part of this 3116 Basis 

Document because, among other things, Section 3116(c), in conjunction with 3116(d), provides that Section 3116 does not apply 

to waste transported from the State of Idaho. 
3 

radioactive radionuclides removed to the maximum extent practical; must meet performance objectives 

for low-level waste (LLW) and either meet certain concentration limits or consult with the NRC regarding 

the disposal plan; and must be disposed of pursuant to a State-approved closure plan or State-issued 

permit. This 3116 Basis Document provides the basis for the Secretary of Energy’s determination that the 

stabilized residuals in the TFF, and in the TFF tank system, after final closure activities are completed 

will meet all of these criteria. Accordingly, these wastes are not high-level waste (HLW) and may be 

grouted and disposed of in place. For the purposes of this 3116 Basis Document, the TFF tank system 

comprises the eleven 300,000-gal tanks, four 30,000-gal tanks, and the vaults, piping, structures, and 

ancillary equipment associated with these tanks. The scope of this 3116 Basis Document does not include 

any other facilities or systems at INTEC or the Idaho National Laboratory (INL) Site such as the 

evaporators or the New Waste Calcining Facility. 

To demonstrate that the stabilized residuals in the TFF, and in the TFF tank system, at final closure 

will meet the Section 3116 criteria, historical waste management information, performance assessment 

(PA) results, and sampling and analysis results from the recent tank cleaning activities were reviewed and 

analyzed.d

 In addition, the residual inventory at closure was updated to reflect the results of TFF cleaning 

activities. Prior to cleaning and receiving sampling and analysis results of the first tank cleaning, planning 

documents such as the first HWMA/RCRA closure plan (DOE-ID 2003a) and the TFF PA 

(DOE-ID 2003b) were prepared. These planning documents determined the baseline inventory and 

consequently established the goals for the cleaning activities to meet or exceed. The total 

post-decontamination inventory for each of the cleaned TFF tanks is significantly less than the total 

conservative post-decontamination inventory of a single tank estimated in the PA (24,103 Ci), indicating 

that the tank cleaning operations performed since this baseline inventory was established performed better 

than planned, as presented in Appendix A. The greatest inventory for a cleaned tank is Tank WM-182, 

which contains 2,393 Ci. 

The purpose of this 3116 Basis Document is to demonstrate and document that the stabilized 

residuals in the TFF, and TFF tank system, meet the 3116 criteria, and therefore by law, are not HLW and 

may be disposed of at INTEC in accordance with Section 3116(a) of the NDAA (Public Law 108-375, 

2004). This 3116 Basis Document addresses only radioactive constituents in the INTEC TFF after final 

closure. Characterization of the hazardous constituents of the TFF residuals in accordance with RCRA is 

not addressed in this 3116 Basis Document. The State of Idaho, which has primacy for RCRA compliance 

but does not have regulatory oversight of the radioactive materials contained in the TFF tanks and 

ancillary equipment, approves the HWMA/RCRA closure plans.e

 

Pursuant to Section 3116(a) of the NDAA, DOE has consulted with the NRC. To facilitate the 

consultative process, DOE provided NRC with a draft 3116 Determination on September 7, 2005. During 

its review of the draft 3116 Determination, NRC requested additional information from DOE 

(NRC 2006a). The DOE responded to the NRC request for additional information (RAI) in three separate 

submittals (DOE 2006a, 2006b, 2006c). The DOE also provided other additional information as requested 

 

d. DOE also took into account the results of its consultations with NRC in 2002. Appendix H discusses the NRC conclusions and 

recommendations from these consultations. 

e. Section 3116 is not dependent on the independent process under the Comprehensive Environmental Response, Compensation, 

and Liability Act (CERCLA) (42 USC 9601 et seq. 1980) and does not provide a basis for any new authority, responsibility, or 

obligation for DOE or any other entity with respect to the CERCLA process or otherwise affect the CERCLA process. Decisions 

regarding past releases of contaminants and the impacts of contaminated soils associated with the TFF will be addressed under 

the CERCLA process as specified in the Federal Facility Agreement and Consent Order among DOE, the State of Idaho, and the 

U.S. Environmental Protection Agency (EPA) (State of Idaho et al. 1991). 
4 

by the NRC, and DOE and NRC held public consultation meetings as well as several consultation 

telephone calls. 

On October 20, 2006, the NRC issued its “Technical Evaluation Report for the U.S. Department of 

Energy Idaho National Laboratory Site Draft Section 3116 Waste Determination for Idaho Nuclear 

Technology and Engineering Center Tank Farm Facility” (NRC 2006b). The TER presents NRC’s views 

and conclusions with respect to whether the stabilized residuals under DOE’s disposal and closure 

approach can meet the applicable requirements of Section 3116(a) of the NDAA such that the Secretary 

may determine that the waste is not HLW. As noted in the NRC’s executive summary (page x), “Based on 

the information provided by DOE, NRC staff has concluded in this TER that there is reasonable assurance 

that the applicable criteria of the NDAA can be met for residual waste associated with the TFF.” 

As discussed in the TER, the NRC staff concluded that the NDAA criteria in Section 3116(a)(1), 

(a)(2), and (a)(3)(B) can be met for the stabilized residuals in the TFF, and the TFF tank system, to be 

disposed of at the TFF (see TER Section 5, Overall Conclusions). Although NRC identified certain 

uncertainties or concerns, and in some instances adopted alternative approaches or differing analyses or 

rationale than DOE, NRC similarly concluded that the applicable criteria of the NDAA would be met by 

DOE’s disposal strategy. DOE has considered the NRC observations in preparing this 3116 Basis 

Document, and has addressed the NRC observations as appropriate throughout this document. 

Section 3116(b) requires that the NRC, in coordination with the State of Idaho, monitor disposal 

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

10 CFR 61, Subpart C. In Subsection 4.4 of its TER, NRC staff identified five “key monitoring areas”, 

and stated that NRC will coordinate with the State of Idaho to develop a monitoring plan/approach to 

address those areas. DOE will work with the NRC and the State of Idaho to facilitate their efforts to fulfill 

their independent responsibilities.f

Although not required by Section 3116, the DOE also made available the draft 3116 Determination 

for a 30-day public review and comment period in September 2005 (70 Federal Register [FR] 54374). No 

public comments were received during the comment period; however, comments were subsequently 

received after the comment period from the State of Idaho and the INL Site Environmental Management 

Citizens Advisory Board (Hardesty 2006; Kipping 2006; Trever 2006). These comments were considered 

in preparing this 3116 Basis Document (Provencher 2006; Van Camp 2006a, 2006b). 

Although not part of the Secretary’s 3116 Determination or this 3116 Basis Document, DOE is also 

issuing an Amended Record of Decision for the Idaho High-Level Waste and Facilities Disposition Final 

Environmental Impact Statement concerning closure of the TFF in conjunction with the issuance of the 

3116 Determination. 

 

f. DOE also notes that Appendix A (page 115) of the NRC’s TER (NRC 2006b) contained nine recommendations for DOE’s 

consideration, “The purpose of the recommendations is to communicate actions that DOE might consider to further enhance its 

approach for management of the TFF waste at INL, as well as the approach for future waste determinations. As stated in this 

Technical Evaluation Report (TER), the NRC staff has concluded that DOE has adequately demonstrated that it can meet 

applicable criteria of the Ronald W. Reagan National Defense Authorization Act for Fiscal Year 2005 (NDAA). Thus, it is the 

NRC staff’s view that implementation of these recommendations is not necessary to meet the criteria in the NDAA.” Although 

most of the recommendations are outside the scope of Section 3116, the DOE, nevertheless, intends to consider these 

recommendations in its disposal of the TFF residuals and development of future 3116 Determinations, as appropriate and 

consistent with law. 
5 

2. BACKGROUND 

Key Points 

• The residual inventory at closure is an estimate of the amount of radioactivity remaining in the 

residual in the TFF at closure in 2012, the planned closure date. All radionuclide concentrations 

are decayed to 2012. 

• The residual inventory for the eleven 300,000-gal tanks is based on analytical data from samples 

obtained after cleaning operations from the seven tanks that have been cleaned. 

- The inventory for the 300,000-gal tanks assumes (for the reasons explained in Subsection 

2.4.2) that the same degree of radionuclide removal will be achieved in the four tanks 

remaining to be cleaned as that achieved in the seven tanks that have been cleaned. 

- Residual solids are estimated from a sample of solids collected from Tank WM-183. 

Concentrations for radionuclides not detected in this sample are estimated using the 

ORIGEN2 numerical model and 137Cs concentrations in Tank WM-188 samples collected 

prior to cleaning. 

- The inventory is not reduced by the amount of additional residual that may be removed 

from the cleaned tanks during grouting operations. 

• A 5-mil (0.005-in.) thick film of residual solids is assumed for the 30,000-gal tank inventory, for 

the reasons described in Subsection 2.4.3. Radionuclide concentrations are assumed (for the 

reasons described in this 3116 Basis Document) to be that found in the Tank WM-183 solid 

samples. 

• The residual inventory for the sandpads underlying Tanks WM-185 and WM-187 is a 

conservative inventory representing the sandpad under these tanks in the PA inventory. The 

inventory is based on the concentrations of radionuclides in liquid residuals that were siphoned 

into the Tank WM-185 vault in 1962. 

• The piping residual inventory is based on analytical results from sampling pipe sections 

removed from the Tank WM-182 process waste lines after cleaning. The inventory for valves is 

included with the piping inventory. No inventory is calculated for valve boxes, pipe 

encasements, and off-gas piping because these components did not normally contain process 

solutions. 

• This 3116 Basis Document addresses disposal of the stabilized residuals in the TFF, and the 

TFF tank system, at closure when the entire TFF has been cleaned and grouted as described in 

this 3116 Basis Document. The closure is expected to be completed by 2012 and decay is 

calculated based on that date. But, as described in more detail in this 3116 Basis Document, 

differences in decay between the present time and 2012 are not significant (and do not affect 

compliance with the performance objectives of 10 CFR 61, Subpart C as considered in this 3116 

Basis Document). 

• Modeling is used to estimate concentrations for radionuclides for which analytical data are not 

available or for which radionuclides were not detected in samples. Modeling and calculations 

are based on the planned 2012 closure date. 

• The sandpads are structural components of the TFF and are located inside the vault and under 

the 300,000-gal tanks. The vaults of Tanks WM-182 through WM-190 contain sandpads. 
6 

The INTEC TFF is located on the INL Site. The INL is an approximately 2,305-km2 (890-mi2

) 

reservation owned by the United States government and located in southeastern Idaho (see Figure 1). 

Established over 50 years ago as the National Reactor Testing Station, INL’s initial mission was to 

develop civilian and defense nuclear reactor technologies. Over the years, the INL mission evolved 

beyond the original reactor technology focus. The INL is now involved in various defense, energy supply, 

industrial technology, and environmental programs. In recognition of this evolution to a multi-program 

installation, the site was designated the Idaho National Engineering Laboratory (INEL) in 1974. In 

January 1997, the name was changed to the Idaho National Engineering and Environmental Laboratory 

(INEEL) to reflect greater emphasis on the laboratory’s environmental missions. In February 2005, the 

INEEL became two separate organizations: INL, which performs its continuing research mission, and 

Idaho Cleanup Project, which carries out the site’s cleanup responsibilities. 

In 1953, the Idaho Chemical Processing Plant, now INTEC, was chartered to recover fissile 

uranium by reprocessing SNF. In 1992, the DOE officially discontinued reprocessing SNF at INTEC. 

This decision changed INTEC’s mission to managing, storing, and treating reprocessing wastes generated 

from past and current operations and activities. The INTEC facility is located approximately 29 km 

(18 mi) from the closest eastern boundary, approximately 23 km (14 mi) from the closest western 

boundary, approximately 16 km (10 mi) from the closest southern boundary, and approximately 29 km 

(18 mi) from the closest northern boundary. The TFF, located within the northern portion of INTEC 

(see Figures 2 and 3), comprises eleven 300,000-gal belowgrade stainless steel tanksg

 in unlined concrete 

vaults of various construction, four inactive 30,000-gal stainless steel tanks, interconnecting waste 

transfer lines, and associated support instrumentation and valves. Structures located above ground level in 

the TFF include the TFF Control House, the Computer Interface Building, and the tank and vault sump 

riser covers. A perimeter fence encloses the TFF. Subsection 2.1 provides a detailed description of the 

TFF. Subsection 2.2 presents a discussion of the wastes stored in the TFF and waste management 

practices. 

Recent tank cleaning operations have resulted in the removal of the remaining SBW and tank heels 

from seven 300,000-gal tanks and four 30,000-gal tanks to the maximum extent practical for the cleaned 

tanks. Four 300,000-gal tanks remain to be cleaned, and these four tanks are anticipated to be cleaned as 

efficiently as the other 300,000-gal tanks that have been cleaned. Sampling and analysis of tank and 

ancillary equipment residuals indicate that the residual inventory at closure is less than that established in 

the 2003 PA (DOE-ID 2003b). The residual inventories in the stabilized residuals in the TFF, and the TFF 

tank system at final closure will meet Section 3116 criteria. Subsection 2.3 describes the TFF closure 

activities and status, including the tank cleaning equipment. (A summary of the tank cleaning technology 

evaluation is presented in Appendix B.) The residual inventory at closure and the characterization of the 

TFF residuals after tank cleaning are presented in Subsection 2.4. 

 

g. Tanks WM-180 and WM-181 are actually 318,000-gal tanks; however, for ease of reference, all 11 tanks will be referred to as 

300,000-gal tanks. 
7 

Figure 1. Idaho National Laboratory site map. 
8 

Figure 2. Tank Farm Facility location at INTEC. 
9 

Figure 3. Plan view of the Tank Farm Facility. 

2.1 Tank Farm Facility Description 

Placed into service between 1953 and 1966, the eleven 300,000-gal tanks (WM-180 through 

WM-190) are approximately 15.2 m (50 ft) in diameter and 6.4–7.0 m (21–23 ft) in height. Nine of the 

eleven 300,000-gal tanks are constructed of Type 304L stainless steel; two tanks (WM-180 and WM-181) 

use Type 347 stainless steel. Each tank has four or five 30-cm (12-in.) diameter risers to provide access to 

the tank. Tanks WM-184 through WM-190 also have one or two 46-cm (18-in.) diameter risers. Most 

risers have installed equipment and instrumentation, including radio frequency probes for level 

measurement, corrosion coupons,h

 and steam jets and airlifts for waste transfer operations. Two steam jets 

are located inside each tank, except for Tanks WM-189 and WM-190, which have one steam jet and one 

 

h. A corrosion coupon is a piece of material, in this case metal, of known weight used to monitor corrosion in a specific 

environment. The coupon weight before and after exposure is measured to determine loss from corrosion. The condition of the 

coupon may also be observed. 
10 

airlift pump. High-pressure steam is forced through a steam jet to create a suction pumping action to 

move liquid from the tank to Tank WM-187 until final treatment and disposition are identified. 

A single steam-jet pump can transfer waste out of a tank at approximately 50 gpm. The original 

design of the suction pumps restricts the ability of the pumps to clear the tanks of liquid completely. 

Using original steam jet placement and prior to any tank cleaning activities, an 8–30-cm (3–12-in.) deep 

residual containing both liquids and solids remains on the floor of the tanks. Therefore, steam-jet pumps 

are lowered to the bottom of the tank for final tank cleaning. Figure 4 presents the dimensions and 

structure of a typical 300,000-gal tank. The 300,000-gal tanks are housed in concrete vaults 

approximately 13.7 m (45 ft) below grade level. The vaults have one of three different designs: 

(1) octagonal pillar-and-panel vaults (five tanks), (2) cast-in-place square vaults (four tanks), and 

(3) cast-in-place octagonal vaults (two tanks). Figure 5 depicts the different vault designs. Cooling coils 

on the floor and walls of eight of the tanks (WM-180, WM-182, WM-183, WM-185, and WM-187 

through WM-190) provide heat transfer capabilities. Figure 6 is a photograph showing the cooling coils in 

Tank WM-185 during construction. The 30,000-gal tanks and 15-cm (6-in.) thick concrete vault roofs are 

covered with approximately 3 m (10 ft) of soil to provide radiological protection to workers and the 

public. 

When the TFF was constructed, the 300,000-gal tanks were designed with flat bottoms. The 

concrete floors of the tank vaults were designed with sloped floors to promote drainage of any liquids 

toward the perimeter of the vaults for efficient removal. In order for the tank to rest on the sloped floor 

without causing unacceptable stresses in the tank, a leveling pad of sand was installed to support the 

tanks. The sandpad is 15 cm (6 in.) thick at the perimeter and 5 cm (2 in.) thick at the center. The exterior 

of the sandpad extends 15 cm (6 in.) beyond the outer edge of the tanks and is confined by a curb 

measuring 15 by 15 cm (6 by 6 in.). The volume of the sandpad is 23.39 m3

. 

Constructed in 1954, the four inactive 30,000-gal stainless steel belowgrade storage tanks 

(WM-103 through WM-106) sit on reinforced concrete pads. The tanks are horizontal cylinders 

approximately 3.5 m (11.5 ft) in diameter and 11.6 m (38 ft) in length. The 30,000-gal tanks do not have 

vaults. All four tanks contain stainless steel cooling coils to provide heat transfer capabilities. Three 

15-cm (6-in.) diameter risers and one 8-cm (3-in.) diameter riser that reach to grade level provide tank 

access. Tank steam jets are provided for liquid waste removal (nominal flow rate of 50 gpm). These tanks 

were removed from service in February 1983. 

Figure 7 is an artist’s representation of the eleven 300,000-gal tanks (WM-180 through WM-190) 

and the four 30,000-gal tanks (WM-103 through WM-106) to be included in the INTEC TFF closure. 

Liquid waste transfers to, from, and among the tanks are managed through a system of lines, valves, and 

diversion boxes. Waste transfer lines are contained within one of the following types of secondary 

encasements: split tile, carbon steel, stainless steel-lined concrete troughs, double-walled stainless steel 

pipes, or buried directly in concrete. Because of upgrades over the years, the pipe sections encased with 

the carbon steel and split tile have either been replaced or abandoned in place, except for two small (about 

1.5-m [5-ft]) sections that are no longer used. The majority of the upgraded piping is contained in 

stainless steel-lined concrete trenches with the remainder in double-walled stainless steel piping. 

Double-walled stainless steel piping was used primarily in areas where single pipe runs were upgraded or 

where access did not allow lined trench installation. Liquid waste is routed through waste transfer valves 

located in belowgrade, stainless steel-lined concrete boxes (referred to as valve boxes). The waste transfer 

valves are operated manually using reach rods. The valve boxes are designed to provide access to the 

valves for inspection and maintenance. 
11 

Figure 4. Cross-sectional view of a typical tank with cooling coils.i

 

 

i. This schematic is labeled as typical because it shows a configuration for the 300,000-gal tanks in the TFF. Not all features of 

each type of tank are shown but features such as the cooling coils, risers, and sandpad are shown in this schematic. 
12 

OCTAGONAL VAULT

PLAN VIEW

WM-180 & WM-181

PILLAR AND PANEL VAULT

PLAN VIEW

WM-182 THROUGH WM-186

SQUARE VAULT

PLAN VIEW

WM-187 THROUGH W

SECTION A-A

SECTION B-B SECTION C-C

Figure 5. Vault designs used at the INTEC TFF. 
13

Figure 6. Photograph showing the internal structure and cooling coils of Tank WM-185 during 

construction. 
14 

INTEC TANK FARM CLOSURE INTEC TANK FARM CLOSURE

VES-WM-103

VES-WM-104 VES-WM-105

VES-WM-106

182 183

185 186

187 189190

188

184

181

180

Tank Farm Facility

Octagon Vaults: WM-180, WM-181

Pillar and Panel Vaults: WM-182, WM-183, WM-184, WM-185, WM-186Square Vaults: WM-187, WM-188, WM-189, WM-190

G

Figure 7. Artist’s depiction of tanks to be included in the INTEC TFF closure. 
15

2.2 Origin and Management of TFF Wastes 

During reprocessing operations between 1953 and 1992, spent fuel was dissolved at INTEC using 

various processes, depending on the fuel type (see Figure 8).j

 Each dissolution process produced an acidic 

aqueous solution (any caustic dissolution processes were adjusted to be acidic before further treatment). 

The aqueous solution was processed through the first-cycle extraction system (Phase I in Figure 8) with 

an organic solvent (usually tributyl phosphate in kerosene). The extraction system used several 

contactors, including pulsed-plate columns and packed columns. The uranium was partitioned from the 

bulk of the fission products and placed in intercycle storage to await purification. The aqueous waste 

phase that contained greater than 99% of the radioactive material was the first-cycle extraction waste 

(Loos 2004), which was generally stored in the TFF in the belowgrade tanks equipped with cooling coils 

because of significant heat generation from the high radioactivity levels in the waste. 

Typical of liquid-liquid extractions, the separated uranium contained some carryover radioactive 

material. After sufficient product accumulated in the intercycle storage, the uranium was processed 

through the second- and third-cycle extractions (Phase II in Figure 8), where the excess radioactive 

material was removed to produce a clean uranium product.k

 

The second and third fuel reprocessing steps (called the second and third cycles) generated the 

smallest volumes of waste of the five major waste sources. The second- and third-cycle processes were 

uranium purification steps and were very similar to each other. The second- and third-cycle processes 

operated together (in series) and were closely coupled. The second cycle purified the uranium product 

from the first-cycle process in a liquid extraction system. The second-cycle process produced a purified 

aqueous uranium product and a waste stream (second-cycle raffinate) containing radioactive 

contamination. The third cycle was an additional purification step that provided further purification of the 

second-cycle uranium product in a liquid extraction system, similar to that of the second cycle. The 

third-cycle process produced a further purified uranium product and a waste stream (third-cycle raffinate) 

containing radioactive contamination. Originally, the purified aqueous uranium product from the 

third-cycle process was shipped to Oak Ridge National Laboratory. Later, a plant modification provided a 

solidification process, the denitrator, to convert the aqueous uranium product into a solid granular form 

for shipment to Oak Ridge National Laboratory (Loos 2004). 

The original SNF reprocessing system combined the second- and third-cycle wastes into a single 

waste stream for storage in the TFF, due to their similarity in chemical and radionuclide content. The 

combined second/third-cycle raffinate was stored separately from first-cycle waste in the TFF. The 

activity of the second/third-cycle waste was low enough that it did not require cooling. Consequently, 

second/third-cycle waste was originally stored in tanks without cooling capability (Loos 2004). 

Unlike the first-cycle waste, the composition of the second/third-cycle waste varied little with the 

type of fuel being processed. Chemicals unique to various first-cycle wastes that came from the fuel 

dissolution process, such as Zr, F, Cd, and Hg, were separated from the uranium product and went with 

 

j. Four fuel types were reprocessed during INTEC operations: Al-clad fuel, Zr-clad fuel, stainless steel-clad fuel, and graphite 

matrix fuel. The dissolution processes varied by fuel type: nitric acid was used for Al-clad fuel, hydrofluoric acid was used for 

Zr-clad fuel, sulfuric acid/nitric acid and electrolytic/nitric acid were used for stainless steel-clad fuel, and graphite matrix fuel 

was burned and the ash dissolved in hydrofluoric acid (Loos 2004). 

k. The second- and third-cycle extractions were liquid extraction systems that operated in series to provide uranium purification 

of the first-cycle extraction uranium product. Because they operated on concentrated feed, solutions compared to first-, second-, 

and third-cycle extractions produced less waste per gallon of feed. The radioactivity level of the second- and third-cycle waste 

stream was approximately 1,000 times less than that of the first-cycle waste (Loos 2004). 
16

the first-cycle raffinate. The second/third cycle processed the first-cycle uranium product. The 

second/third-cycle waste was primarily acidified aluminum nitrate, regardless of the type of fuel that was 

processed. The chemical composition (not radioactivity) of the second/third-cycle waste was similar to 

the first-cycle waste from Al-clad fuel (Loos 2004). 

Unlike other wastes, the second/third-cycle raffinate did not contain any chemicals of concern to 

waste storage or treatment, such as F, Cl, and Na. It was not so highly radioactive (compared to first-cycle 

raffinate) that it required cooling or other special considerations beyond that of first-cycle raffinate. 

Consequently, names reflecting the terms second- or third-cycle waste faded from use. Instead, the names 

of the waste came from the dominant waste type in terms of waste volume or chemical/radionuclide 

significance with which it was blended. For example, if 3,000 gal of third-cycle waste was added to 

100,000 gal of SBW, the resulting mixture was commonly called SBW, with no reference to the 

third-cycle portion of the waste (Loos 2004). 

The aqueous intermediate liquid waste in Figure 8 from second- and third-cycle extraction was 

approximately 1% of the initial reprocessing inventory of radioactive material. 

Figure 8. Spent fuel reprocessing at INTEC.
17

Much of the INTEC reprocessing equipment was designed for contact maintenance rather than 

remote maintenance. Because INTEC personnel had to access equipment for maintenance purposes 

frequently, efforts to decontaminate plant areas and equipment to lower radiation fields and allow safe 

working conditions generated large amounts of decontamination wastes, which were concentrated by 

evaporation and sent to the TFF. As of November 2003, the remaining TFF inventory contained 

(by volume) 1% first-cycle waste, 2% second-cycle waste, and 4% third-cycle waste. The remaining 

waste is composed of decontamination wastes and bottoms from the evaporatorl

 (Loos 2004). 

In general, TFF wastes were managed by segregating the first-cycle solvent extraction waste from 

other reprocessing wastes because of significantly higher radioactivity levels in the first-cycle waste. The 

DOE decided not to neutralize waste or combine the first-cycle extraction wastes with other reprocessing 

wastes, as was the standard practice at other DOE facilities (Knecht et al. 1997). Instead, INTEC 

maintained the waste in its original acidic form using the 300,000- and 30,000-gal stainless steel storage 

tanks and physically segregated first-cycle extraction wastes. This decision reduced the volume of liquid 

waste requiring storage. 

Waste volume was reduced further by stabilizing the first-cycle solvent extraction waste and most 

of the second- and third-cycle extraction wastes through calcination. Calcine results from heating a 

substance to a high temperature that is below its melting or fusing point. At the INL Site, calcination was 

carried out in the calciner in the New Waste Calcining Facility where liquid HLW and mixed transuranic 

waste/SBW are converted into the granular solid known as calcine. The liquid waste is drawn from TFF 

and sprayed into a vessel containing an air-fluidized bed of granular solids. The bed is heated by 

combustion of a mixture of kerosene and oxygen. All of the liquid evaporates, while radioactive fission 

products adhere to the granular bed material in the vessel. The gases from the reaction vessel (called 

off-gases) are processed in the off-gas cleanup system before they are released to the environment. 

Calcination reduces the volume of the radioactive liquid waste (usually two to ten times), so less storage 

space is needed. The final waste form is a dense powder similar in consistency to powdered detergent. 

These calcined solids are transferred to the Calcined Solids Storage Facilities, commonly referred to as 

bin sets. The bin sets are a series of concrete vaults, each containing three to seven stainless steel storage 

bins (DOE 2002). 

Calcined waste is not part of this 3116 Basis Document. The New Waste Calcining Facility, the 

evaporators, the reprocessing facility, or other nearby facilities are also not included in this 3116 Basis 

Document. 

By evaporating and calcining, and not neutralizing, the liquid radioactive waste, INTEC avoided 

the construction of up to 195 additional 300,000-gal storage tanks (Knecht et al. 1997). The solidified 

(calcined) reprocessing waste was transferred to stainless steel bins for storage. By February 1998, the 

liquid first-cycle extraction waste was removed from the TFF. Only small (1,000–15,000 gal) heels in 

eight of the eleven 300,000-gal storage tanks remained, which could not be removed with existing 

equipment. Reuse of the first-cycle waste storage tanks to store SBW has resulted in the mingling of the 

first-cycle waste heels with SBW. 

 

l. Evaporators have been used at INTEC to reduce the volumes of radioactive waste. Bottoms are the concentrated material 

remaining after the liquid portion of the waste is evaporated. 
18

Between 2002 and 2005, TFF tank contents were evaporated to less than one million gallons and 

consolidated into three 300,000-gal tanks. This SBW remains in storage in Tanks WM-187, WM-188, 

and WM-189. The current SBW inventory is a mixture of wastes from various sources, including: 

• Decontamination solutions from past spent fuel reprocessing maintenance activitiesm

• Solids and liquids from the bottom of the TFF tanks, which include some first-cycle wasten

• Liquid wastes from ongoing maintenance and closure activities at INTEC 

• Second- and third-cycle spent fuel reprocessing extraction wastes. 

The DOE is pursuing methods for treating and disposing the remaining SBW, and the three tanks 

storing SBW will be emptied during the treatment process. Tank heels will be removed as part of closure 

activities, and they will be treated with the bulk SBW. It is planned to have the treatment of the SBW 

completed by 2012. As explained previously, the treated SBW is not part of this 3116 Determination. 

Each of the eleven 300,000-gal tanks in the INTEC TFF has a slightly different configuration, 

waste storage, and processing history (INEEL 1999a, 2000a). The following is a history of the TFF tanks, 

including the sources of the solid and liquid residuals in the tank heel, and a description of the tank 

contents and level as of April 30, 2005. Liquid levels in the cleaned tanks (WM-180 through WM-186) 

are not the levels that will remain at closure. The liquid after sampling is left at a depth of approximately 

13 cm (5 in.) to avoid the tank drying prior to grouting and creating potential airborne contamination 

problems during the grouting operations. The liquid will be removed to approximately 3 cm (1 in.) prior 

to the addition of grout during stabilization operations. The liquid that will be removed will be sent to the 

SBW treatment system or to operational tanks. 

The steam jets have been lowered to 0.97 cm (0.38 in.); therefore, based on steam jet performance, 

it is conservative to assume that the tank liquid will be lowered to a 3-cm (1-in.) depth (Portage 2005a, 

2005b, 2005c, 2005d, 2005e, 2005f, 2006a). The levels shown in the tank descriptions below are the tank 

levels at the time the samples were collected after tank cleaning. After the liquid level is reduced using 

the steam jets, additional liquid will be removed by the grout placements. No credit is taken in the PA and 

in this 3116 Basis Document for liquid removed during grout placements,o

 which constitutes 

approximately 3 Ci (per tank), as identified in Table 1. Details of the grout composition are shown in 

 

m. Typical decontamination cycles from reprocessing fuels consisted of a strong caustic solution to break down the resistant 

oxide layer, followed by corrosive and/or chelating agents to expose base metal and contaminants. The cycle was completed with 

a 6-M nitric acid flush to restore the oxide layer (INEEL 2002a). 

n. Particle size distribution analysis and settling rate testing of the WM-182 and WM-183 TFF heel slurry samples (samples were

collected prior to tank cleaning) were performed. The conclusions and recommendations based on the results of this work follow 

that provide insight into the properties of the solid residual, which include the following (INEEL 2000b): 

• The overall results for the standards testing were satisfactory and demonstrated excellent repeatability and acceptable 

accuracy during the actual TFF sample testing. 

• Particle size distribution analysis showed that for both the WM-182 and WM-183 samples, the particles range 

approximately from a minimum of 0.5 μm to a maximum of 230 μm, with about 90% volume between approximately 2 and 

133 μm. The results of the particle size determination for samples from WM-182 and WM-183 are quite similar considering 

the minute quantities used from the two separate vessels to obtain these results. 

• Settling rate testing results were fairly consistent in that it appears that most of the mass of solids settle to an easily 

redispersed layer. 

o. As discussed later in this 3116 Basis Document, DOE has considered the NRC and State of Idaho’s observations in this regard. 
19

Appendix C. Information about the grout sequence and the engineering grout pour are included in 

Section 6. The solid residuals exist as fine particles. (All tanks contain cooling coils unless otherwise 

noted.) Figure 9 summarizes the tanks’ volumes as of April 30, 2005. 

• Tank WM-180—This tank was placed into service in 1954 and was used to store Al fuel 

reprocessing wastes. The majority of the first-cycle extraction process waste was calcined in 

1966–1967. Since 1972, the tank has been used solely to store SBW. The tank is housed in an 

octagonal, poured-in-place reinforced concrete vault. This tank does not have an underlying 

sandpad. On November 30, 2003, WM-180 was in service and contained 276,000 gal of waste. The 

waste was primarily Process Equipment Waste Evaporator (PEWE) bottoms with small amounts of 

second- and third-cycle raffinate. The tank was emptied to its heel in 2004 and the waste 

transferred to the evaporator or to WM-187 (Loos 2004). Tank cleaning operations were completed 

in October 2004. As of April 30, 2005, Tank WM-180 contains 7,600 gal of liquid after tank 

cleaning operations and an estimated 540 kg of solid residual (Portage 2005a). The remaining 

liquid is water from tank cleaning. Prior to grouting, the remaining liquid volume will be removed 

to a 3-cm (1-in.) depth or approximately 1,300 gal. The estimated inventory at closure for 

Tank WM-180 is approximately 1,047 Ci (Portage 2005a). 

• Tank WM-181—This tank was placed into service in 1953 and was used only to store SBW. The 

tank is housed in an octagonal, poured-in-place reinforced concrete vault. This tank does not 

contain cooling coils or have an underlying sandpad. Tank WM-181 was “emptied” of waste 

(the volume in the tank was reduced to its heel, using the installed waste transfer equipment) in 

November 2001. On November 30, 2003, Tank WM-181 contained 22,800 gal of waste. The waste 

was primarily PEWE bottoms from Tank WL-101 and “other” waste with a small amount of 

third-cycle raffinate. The tank was never used for storage of first-cycle raffinate (Loos 2004). Tank 

cleaning operations were completed in May 2004. As of April 30, 2005, Tank WM-181 contains 

7,300 gal of liquid after tank cleaning operations and an estimated 250 kg of solid residual 

(Portage 2006a). The remaining liquid is water from tank cleaning. Prior to grouting, the remaining 

liquid volume will be removed to a 3-cm (1-in.) depth or approximately 1,300 gal. The estimated 

inventory at closure for Tank WM-181 is approximately 475 Ci (Portage 2006a). 

• Tank WM-182—This tank was placed into service in 1956 and was used to store Al and Zr fuel 

reprocessing wastes. The tank was emptied to heel level in 1993 and has since been flushed with 

small quantities (about 10,000 gal) of SBW. The tank is housed in an octagonal, pillar-and-panel 

concrete vault. Between December 1996 and June 1999, two transfers of waste were made 

(15,100 gal total) out of the tank. These transfers removed the liquid to heel level. Tank inspection 

began in 1999. Tank cleaning was completed in September 2002. As of April 30, 2005, Tank 

WM-182 contains 6,500 gal of liquid after tank cleaning operations and an estimated 1,240 kg of 

solid residual (Portage 2005b). The remaining liquid is water from tank cleaning. Prior to grouting, 

the remaining liquid volume will be removed to a 3-cm (1-in.) depth or approximately 1,300 gal. 

The estimated inventory at closure for Tank WM-182 is approximately 2,393 Ci (Portage 2005b). 

• Tank WM-183—This tank was placed into service in 1958 and was used to store Al and stainless 

steel fuel reprocessing wastes. The first-cycle extraction process waste was transferred from the 

tank in 1981, and the tank was refilled with various types of SBW. The tank is housed in an 

octagonal, pillar-and-panel concrete vault. Between February 1997 and June 1999, 16,000 gal of 

low-activity waste was sent to WM-183. The accumulated waste in WM-183 was transferred to 

WM-187 in December 1999, leaving a heel of about 13,000 gal. Tank cleaning was completed in 

March 2003. As of April 30, 2005, Tank WM-183 contains 8,000 gal of liquid after tank cleaning 

operations and an estimated 700 kg of solid residual (Portage 2005c). The remaining liquid is water 

from tank cleaning. Prior to grouting, the remaining liquid volume will be removed to a 3-cm 
20

(1-in.) depth or approximately 1,300 gal. The estimated inventory at closure for Tank WM-183 is 

approximately 1,360 Ci (Portage 2005c). 

In 2004, while consolidating SBW to Tank WM-187 from Tank WM-180, it was noticed that the 

liquid level of Tank WM-183 had increased. Upon further investigation it was determined that 

approximately 200 gal of SBW and water had leaked into the tank during transfers of WM-180 

waste to the evaporator tank system, then to WM-187 during the period between July and October 

2004. The leakage was determined to have been through an improper valve setting. The isolation of 

the tank has been confirmed. The tank has been re-cleaned and sampled. Visual inspection of the 

tank indicates that the 200 gal of waste that leaked into the tank has been removed and no 

additional solids are present in the tank. Data collected were consistent with prior sampling of this 

tank (ICP 2006a). These new data did not change the results or conclusions presented in this 3116 

Basis Document. 

• Tank WM-184—This tank was placed into service in 1958 and was used only to store SBW. The 

tank is housed in an octagonal, pillar-and-panel concrete vault. This tank does not contain cooling 

coils. Between May and December 2001, WM-184 was emptied to a heel volume of 5,100 gal. 

After this time, no new wastes were added to the tank. Tank cleaning was completed in December 

2003. As of April 30, 2005, Tank WM-184 contains 3,100 gal of liquid after tank cleaning 

operations and an estimated 560 kg of solid residual (Portage 2005d). The remaining liquid is water 

from tank cleaning. Prior to grouting, the remaining liquid volume will be removed to a 3-cm 

(1-in.) depth or approximately 1,300 gal. The estimated inventory at closure for Tank WM-184 is 

approximately 1,077 Ci (Portage 2005d). 

• Tank WM-185—This tank was placed into service in 1959 and was used to store Al and Zr fuel 

reprocessing wastes. The first-cycle extraction process waste was calcined in 1983, and the tank 

was refilled with SBW. The tank is housed in an octagonal, pillar-and-panel concrete vault. The 

waste was removed from WM-185 in January 2002, leaving a heel of about 13,000 gal of waste in 

the tank. Tank cleaning operations were completed in October 2003. As of April 30, 2005, Tank 

WM-185 contains 5,800 gal of liquid after tank cleaning operations and an estimated 720 kg of 

solid residual (Portage 2005e). The remaining liquid is water from tank cleaning. Prior to grouting, 

the remaining liquid volume will be removed to a 3-cm (1-in.) depth or approximately 1,300 gal. 

The estimated inventory at closure for Tank WM-185 is approximately 1,391 Ci (Portage 2005e). 

• Tank WM-186—This tank was placed into service in 1962 and was used to store Al fuel 

reprocessing waste. The first-cycle extraction process waste was transferred from the tank in 1967, 

after which the tank was used solely to store SBW. The tank is housed in an octagonal, 

pillar-and-panel concrete vault. This tank does not contain cooling coils. In May 2001, the tank was 

emptied to a heel of 20,300 gal. Tank cleaning operations were completed in November 2003. As 

of April 30, 2005, Tank WM-186 contains 6,600 gal of liquid after tank cleaning operations and an 

estimated 340 kg of solid residual (Portage 2005f). The remaining liquid is water from tank 

cleaning. Prior to grouting, the remaining liquid volume will be removed to a 3-cm (1-in.) depth or 

approximately 1,300 gal. The estimated inventory at closure for Tank WM-186 is approximately 

646 Ci (Portage 2005f). 

• Tank WM-187—Placed into service in 1959, Tank WM-187 was used to store Al and Zr fuel 

reprocessing wastes. The first-cycle extraction process waste was calcined between 1990 and 1993, 

and then SBW was added to the heel. The tank is housed in a square, reinforced-concrete vault. 

This tank is being used to collect SBW solutions from other TFF tanks undergoing cleaning 

activities prior to final closure (Loos 2004). As of April 30, 2005, Tank WM-187 contains 

271,500 gal of liquid, which was transferred from cleaned tanks (ICP 2005a). This tank will be 
21

cleaned to remove highly radioactive radionuclides to the maximum extent practical and is 

expected to meet the other criteria in Section 3116(a) of the NDAA. 

• Tank WM-188—This tank was placed into service in 1959 and was used to store Al and Zr fuel 

reprocessing wastes. The tank was emptied to heel level when the waste was calcined in 1998. The 

tank is housed in a square, reinforced-concrete vault (Loos 2004). As of April 30, 2005, Tank 

WM-188 contains 282,600 gal of liquid that is primarily SBW (ICP 2005a). This tank will be 

cleaned to remove highly radioactive radionuclides to the maximum extent practical and is 

expected to meet the other criteria in Section 3116(a) of the NDAA. 

• Tank WM-189—Placed into service in 1966, Tank WM-189 was used to store Zr fuel reprocessing 

wastes. The first-cycle extraction process waste was emptied to heel level in 1996. Since then, the 

tank has been used to store a variety of SBW solutions from continuing operations (high-fluoride 

decontamination wastes, bottoms from the evaporator tank system, and other SBW). The tank is 

housed in a square reinforced concrete vault (Loos 2004). As of April 30, 2005, Tank WM-189 

contains 282,300 gal of liquid that is primarily SBW (ICP 2005a). This tank will be cleaned to 

remove highly radioactive radionuclides to the maximum extent practical and is expected to meet 

the other criteria in Section 3116(a) of the NDAA. 

• Tank WM-190—Tank WM-190 was designated as a spare tank for use in emergencies and was 

never placed into service. This tank was never used to store any waste, including first-cycle 

extraction process waste, although the tank has been contaminated with this waste. By 1980, 

approximately 7,000 gal of liquid had accumulated in the tank from two sources: (1) rainwater that 

collected in the vault sump was jetted into the tank and (2) a small quantity (less than 50 gal total) 

of first-cycle extraction process waste was passed inadvertently through the transfer valve. The 

transfer valve was opened slightly and closed (to confirm valve closure) prior to starting waste 

transfers to tanks that shared a common transfer line (Loos 2004). (Waste from the last transfer 

would collect above the valve.) In 1982, the tank contents were transferred using a specially 

designed sump pump, which left approximately 500 gal of liquid in the tank as of April 30, 2005 

(ICP 2005a). This tank will be cleaned to remove highly radioactive radionuclides to the maximum 

extent practical and is expected to meet the other criteria in Section 3116(a) of the NDAA. 
22

Figure 9. Volumes in TFF 300,000-gal tanks as of April 30, 2005 (ICP 2005a).p

The four 30,000-gal tanks (WM-103 through WM-106) were placed into service in 1956 and were 

used primarily (the tanks held injection well waste and process evaporator condensate for a short period in 

1983 and all of the 30,000-gal tanks held steam valve condensate after 1990) for first-cycle extraction 

process waste from stainless steel and Zr fuel reprocessing. Each of the tanks were emptied to heel level 

(8–13 cm [3–5 in.] in the tank bottom, approximately 1,000 gal per tank)q

 in 1974 and flushed with water. 

They remained at heel level until 1982, when they were used to store condensate liquid (not first-cycle 

process wastes) from the PEWE for 5 months, after which they were emptied again to heel level and 

taken out of service in February 1983.r

 The waste inlets to these tanks were subsequently cut and capped; 

 

p. Tanks WM-187, WM-188, and WM-189 continue to be used for interim storage of INTEC wastes. As such, these volumes 

fluctuate over time as wastes are transferred into or out of these tanks. 

q. Jacobson, V. L., INEEL, Memo to File: “Discussion on WM-103 through -106 with Frank Ward, BBWI Operations 

Engineering,” September 26, 2001. 

r. The PEWE treats diluted, low-activity waste by boiling the waste, and condensing and collecting the vapors. The PEWE splits 

the waste into two streams. One stream is a small volume of concentrated liquid that contains most of the chemical and 

radioactive constituents that are originally in the PEWE feed. The concentrated waste stream is called the “bottoms.” Typically, 

the PEWE generates 1–2 gal of bottoms from every 100 gal of feed, concentrating the waste feed by a factor of 50–100. The 

second waste stream is the condensed vapor that left the PEWE. This stream is called “process condensate,” which is a large 

volume of waste that contained 98–99 gal of every 100 gal of PEWE feed solution. Most of the chemical and radioactive 

constituents in the PEWE feed (such as Al, Zr, 137Cs, and 90Sr) are nonvolatile and concentrated in the evaporator bottoms. As a 

result, the PEWE condensate is relatively clean water, containing only trace quantities of most chemicals and radionuclides 

(Loos 2004). 
23

however, the outlets remained operational to permit tank decontamination activities to occur. In 1990, the 

30,000-gal tank heels were sampled and analyzed. The results indicated a small amount of radioactivity 

(the 137Cs concentrations ranged from 1.20E–07 to 1.10E–05 Ci/L) (DOE-ID 2003b). Then the tanks were 

each flushed with 12,000 gal of water and emptied to their heels (DOE Idaho 2004a). 

In April 2004, the 30,000-gal tanks were cleaned by flushing each tank with 3,000–5,000 gal of 

water and flushing the piping three times. (The length of pipe being cleaned is flushed with three volumes 

of water. The volume of water used is equal to the volume of the length of pipe.) The tanks were 

inspected after flushing and the samples were analyzed to determine the concentrations of residual 

radionuclides. A conservative estimate of the solid residual was made based on the presence of a film 

layer in the lower half of the tank. Approximately 19 kg of waste residual remain in each tank is estimated 

based on video inspection and sampling and analysis (Portage 2006b). Each 30,000-gal tank has an 

estimated inventory at closure of approximately 36 Ci. 

2.3 Tank Farm Facility Closure Activities and Status 

The DOE is closing the TFF tanks in response to a January 1990 Notice of Noncompliance and 

subsequent Consent Order (State of Idaho et al. 1992). The Idaho Department of Health and Welfares

 and 

EPA issued the Notice of Noncompliance to the DOE because the tanks in the TFF did not meet the 

secondary containment requirements as set forth by Idaho Administrative Procedures Act (IDAPA) 

58.01.05.009 (40 CFR 265.193). The resulting 1992 Consent Order (and subsequent modifications) (State 

of Idaho et al. 1992, 1994, 1998, 1999) required the DOE to permanently cease uset

 of the five 300,000-

gal tanks that are contained in five pillar-and-panel vaults by June 30, 2003. The Consent Order also 

required the DOE to permanently cease use of the remaining 300,000-gal tanks by December 31, 2012, or 

bring the tanks into compliance with secondary containment requirements. The DOE decided to close the 

TFF tanks because radiation fields would make compliance with secondary containment requirements 

impractical, and because the DOE did not anticipate a need for such storage after 2012. (Compliance is 

impractical because the radiation fields in the tank vault would prevent practical entrance by personnel to 

add equipment or upgrades to the vaults to meet the secondary containment requirements [DOE 2002].) 

In addition, in October 1995, the State of Idaho, the Department of the Navy, and the DOE settled 

the cases of Public Service Co. of Colorado v. Batt, No. CV 91-0035-S-EJL (D.Id.) and United States v. 

Batt, No. CV-91-0065-S-EJL (D.Id.) (State of Idaho et al. 1995). Among other things, the resultant 

settlement agreement required the DOE to complete the process of calcining all remaining non-sodiumbearing liquid HLW by June 30, 1998, and to remove SBW from the INTEC TFF tanks by December 31, 

2012. In accordance with this settlement agreement, the DOE completed the removal of first-cycle 

extraction waste from the TFF tanks to heelu

 level in February 1998. The tanks were then used to store 

additional SBW. The DOE has evaluated options for waste treatment and INTEC facility disposition. This 

evaluation is discussed in the Idaho High-Level Waste and Facilities Disposition Final Environmental 

Impact Statement (EIS) (DOE 2002). DOE issued the Record of Decision for the Idaho High-Level Waste 

 

s. On July 1, 2000, the Division of Environmental Quality, a division within the Idaho Department of Health and Welfare, was 

elevated to the State of Idaho Department of Environmental Quality (DEQ). This department now oversees the implementation of 

the Consent Order. 

t. “Cease use” means to empty the tanks down to their heels (i.e., the liquid level remaining in each tank after lowering to the 

greatest extent possible by use of existing transfer equipment). Closure plans developed for these tanks will address the 

remaining heel and vaults and the use of these tanks and equipment for closure, including any flushing or other cleaning of the

tanks (State of Idaho 1998). 

u. “Tank heel” means the liquid level remaining in each tank after lowering the level to the greatest extent possible by using 

existing transfer equipment, such as steam jets. 
24

and Facilities Disposition Final Environmental Impact Statement on December 19, 2005 (70 FR 75165), 

in which DOE decided, among other things, to select steam reforming as a treatment technology for the 

SBW and pursue a phased decision-making process by issuing an amended Record of Decision (ROD) in 

2006. The initial 2005 ROD stated that this amended ROD would specifically address closure of the TFF 

in coordination with the Secretary of Energy’s Determination pursuant to Section 3116. Therefore, in a 

separate decision document, the DOE is amending its initial ROD and is addressing closure of the TFF in 

coordination with the Secretary’s Determination under Section 3116. 

The TFF is being closed in accordance with DOE requirements as a radioactive waste storage 

facility, and with HWMA/RCRA requirements for closure of an interim status HWMA/RCRA tank 

system. These requirements include preparing several documents. The documents describe DOE’s actions 

to close tanks and meet closure objectives. The DOE requires closure plans, a PA, and a composite 

analysis to address radioactive constituents. To meet HWMA/RCRA requirements, closure plans and 

sampling and analysis plans (SAPs) addressing the hazardous waste constituents are required. 

The following major documents have been prepared for the TFF closurev,w: 

• Deactivated HLW facility DOE Tier 1 closure plan (DOE-ID 2006a)—Developed in accordance 

with DOE Order 435.1 (2001) requirements and guidance, this plan defines the approach and plans 

for TFF closure and includes a PA (DOE-ID 2003b) and a composite analysis (DOE-ID 2006b). 

The PA documents the projected radiological dose impacts associated with TFF closure to meet the 

requirements of 10 CFR 61, Subpart C, “Performance Objectives.” The composite analysis is a 

planning tool that provides a reasonable expectation that the proposed closure activities will not 

result in the need for future corrective or remedial actions to ensure protection of the public and 

environment. This analysis shows that radiation doses to members of the public through the 

groundwater pathway at the INL Site boundary are protective of human health. Prior to grouting 

activities, DOE plans to finalize the Tier I plan in accordance with DOE Order 435.1. 

• HWMA/RCRA closure plans (DOE-ID 2003a, 2003c, 2003d, 2004a; DOE Idaho 2004a, 2004b) 

and associated SAPs (ICP 2004a, 2004b; INEEL 2001, 2002b, 2003). Section 8 addresses the 

approval status of these HWMA/RCRA closure plans. 

In general, the closure process includes removing the SBW for treatment then closing the tanks to 

meet RCRA and Section 3116 criteria. The TFF tank system’s closure process includes waste removal; 

cleaning of the tanks, piping, and ancillary equipment (to remove waste to the maximum extent practical); 

 

v. Following closure to these HWMA/RCRA requirements and DOE Orders, decisions and actions regarding final capping 

(including capping of the TFF tanks), monitoring, and long-term maintenance of the site will be conducted as part of the 

CERCLA program. Although not part of this document or affected by Section 3116, DOE’s Proposed Plan for Tank Farm Soil 

and Groundwater at the Idaho Nuclear Technology and Engineering Center, Operable Unit 3-14 (ICP 2006k), which includes 

capping the surface of the TFF, has been issued for public comment pursuant to CERCLA. The CERCLA ROD and decision is 

planned for 2007. Capping will reduce water infiltration and provide worker protection where appropriate. 

w. In December 2002, the DOE prepared a draft waste determination for the TFF (“Idaho Nuclear Technology and Engineering 

Center Tank Farm Facility Residuals–Waste-Incidental-to-Reprocessing Determination Report,” DOE-ID-10777, Draft Rev. B, 

December 2002) to demonstrate that the residuals in the TFF tanks met the DOE’s requirements (DOE M 435.1-1, 2001) for 

managing the waste as LLW and submitted the draft document and the PA (DOE-ID 2003b) to the NRC for review. In general, 

the NRC found that the DOE demonstrated that the waste meets DOE’s and NRC’s waste determination guidance to manage the 

waste as LLW. The NRC also concluded that the DOE’s determination that the residuals from tank closure activities can be 

managed as LLW “has sound technical assumptions, analysis, and conclusions with regard to protecting public health and safety, 

and the environment” (NRC 2003a). However, with the enactment of Section 3116 of the NDAA, the DOE did not issue a final 

waste incidental to reprocessing determination. 
25

and stabilizationx

 of the tank configuration and ancillary equipment. To complete SBW removal, 

remaining liquid and solid waste residue will be removed to the maximum extent practical from the tanks 

and ancillary equipment. Following waste removal from the tanks and TFF cleaning activities, 

confirmatory sampling and analysis will be performed to assess the decontamination effectiveness and for 

waste characterization. (As discussed below, this sampling and analysis has already been performed for 

the tanks and ancillary equipment that have been cleaned as of May 2005.) 

Some residuals that cannot be removed by the cleaning process or other technically practical means 

will remain. The residuals will be sampled and analyzed to determine the concentrations of radionuclide 

and hazardous constituents remaining in the tanks. As each TFF component is cleaned and analytical data 

show that performance objectives are expected to be met, the DOE plans to stabilize each of the 

applicable TFF components by filling them with grout. Process lines will be decontaminated and capped, 

and all lines (including process lines) that provide a pathway to the tanks will be grouted and capped 

(Appendix C; DOE-ID 2006a). 

This 3116 Basis Document includes the evaluation of all TFF equipment and structures that are 

potentially contaminated with reprocessing wastes as a result of past INTEC reprocessing operations. 

These structures, systems, and components will be isolated and grouted as a part of the INTEC TFF final 

closure and include the stainless steel tanks, concrete vaults, sandpads, piping, encasements, valve boxes, 

and instrumentation lines. The TFF tank cleaning operations began in 2002 for those tanks emptied of 

SBW. Closure of the INTEC TFF is being conducted in phases to support continued INTEC operations.y

As of May 2005, Tanks WM-180 through WM-186 and WM-103 through WM-106 have been cleaned 

along with the ancillary equipment of each tank. No additional cleaning has been performed since that 

time. Samples have been collected from the tanks and ancillary equipment, and the results have been 

reported in data quality assessment (DQA) reports (ICP 2006b, 2006c, 2006d, 2006e, 2006f, 2006g, 

2006h, 2006i 2006j; INEEL 2006a, 2006b). Waste liquids and solids removed and generated during tank 

closure activities are being collected for final treatment and disposal. 

2.3.1 Cleaning Approach for Stainless Steel Tanks and Vaults 

A tank cleaning system comprising a washball, directional nozzle, and modified steam-jet pumping 

system has been developed and used successfully thus far in the TFF tank cleaning operations. Figure 10 

illustrates this cleaning system. During washball and directional nozzle operations, the steam-jet ejectors 

are operated to remove the waste-containing slurry from the tank. The goal of tank cleaning is to remove 

as much waste as practical. During this operation, radiation levels are monitored on the steam-jet 

transport line as an indication of cleaning effectiveness. Monitoring the radiation levels near the transport 

line provides the cleaning system operators and project manager an indication of when continued tank 

cleaning ceases to be effective. When radiation levels decrease to the lowest value and remain constant, 

cleaning is stopped and the tanks are inspected. If visual inspection via a remote-controlled camera 

confirms that the tank has been cleaned to the extent practical, then samples are collected and analyzed to 

 

x. Waste stabilization is intended to minimize radiation exposures to the public caused by leaching of radionuclides and 

chemicals out of the waste form and by intrusion into the waste. Stability limits exposures to an inadvertent intruder by providing 

greater assurance that the waste form will remain recognizable and nondispersible. 

y. INTEC operations include management of SBW and newly generated liquid waste. Also included are the evaporator tank 

system and the PEWE (the PEWE is not part of this 3116 Basis Document). Some of the newly generated liquid waste is the 

water from tank cleaning operations. This liquid is sent to Tank WM-187 through existing process piping and valve boxes. 

Additionally, these lines and valve boxes are used to transfer the liquid to the PEWE. A phased approach to cleaning was used at 

TFF because the transfer of liquid waste was necessary, closure of tanks and associated piping had to be scheduled to allow 

liquids to be transferred from a tank being cleaned yet not restrict operations such that another tank cannot be cleaned, or valves 

and piping that allow waste to be transferred to the PEWE are not closed prematurely. 
26

verify performance objectives are met. During the visual inspection, residual solid waste depths are 

estimated by comparing the solids depths to benchmarks within the tanks such as cooling coil support 

brackets and associated welds. For example, in tanks with cooling coils, the bottom weld and stainless 

steel bracket thickness measures 0.97 cm (0.38 in.). Knowing this thickness, the depth of waste next to 

these brackets can be estimated. A reflection from the stainless steel at the tank bottom indicates that no 

solids are present (Portage 2005a, 2005b, 2005c, 2005d, 2005e, 2005f, 2006a). The radiation monitor 

allows tank cleaning to proceed without repeated visual inspection or sample collection and aids in 

ensuring that as much waste as practical is removed from the tanks (Kimmitt 2002). Samples of the 

residual are collected with small positive-displacement pumps. Submersible pumps are lowered into the 

bottom of tanks or vaults through risers. The pump is activated and liquid and solids are pumped to 

sample containers on the surface. The submersible pump can only reach the residual directly beneath the 

riser through which it is lowered. However, because the residual will be agitated before sampling, it is 

reasonable to assume that the liquid in the tank is homogeneous and the flocculent solids may be in 

suspension. 

Figure 10. Typical tank cleaning system. 

Prior to its implementation, the TFF tank cleaning system was tested in a full-scale mockup tank 

(INEEL 1999b) using simulated waste (EDF-015722-041, 2000). The system demonstrated effectiveness 

in subsequent tank cleaning. The washball/directional nozzle tank cleaning system and the modified 

steam-jet pumping system were used to slurry the solid and liquid wastes and remove them from the 

tanks. Steam jets were modified by cutting the steam supply and discharge lines and installing a new 

steam jet lower in the tank. During cleaning system development, the INL Site and the Tanks Focus Area 

(TFA) performed a review of tank cleaning technologies. The DOE formed the TFA to address all aspects 

of remediating radioactive wastes from underground storage tanks DOE-wide, including tank cleaning 

technology. The TFA review is described in a report prepared by the Pacific Northwest National 

Laboratory (PNNL 2001). This review focused not only on the technical feasibility and appropriateness of 

the approach selected by the INL Site but also on technology gaps that could be addressed by using 

technologies or performance data available at other DOE sites and in the private sector. The review 

supported the design and implementation of the INL Site cleaning system. As a result of this development 

and testing work, the cleaning system (washball, directional nozzle, and modified steam-jet pumping 
27

system) has performed beyond expectations as demonstrated by a comparison of the total 

post-decontamination inventories estimated in the PA and calculated for WM-182, as presented in 

Appendix A. 

The tank vaults are cleaned by iterative flushing with water. The water is removed using the 

existing steam jets. Access to the vaults is limited and initial radionuclide inventories are considerably 

less than the tanks (DOE-ID 2003b); therefore, flushing is a practical and effective cleaning method. 

Process piping in the TFF is cleaned by triple flushing, which consists of flushing three piping system 

volumes through the system with a pressure equal to previous waste transfers to ensure that the pipe area 

contacted by waste has been contacted by water and rinsed during the flushing operations; this method 

has been shown to be effective based on analytical data.z

 The acidic nature of the waste and the 

procedures that required flushing with water after waste transfers during operations limited accumulation 

of residual in piping that needed to be removed. 

The vaults provide secondary containment for tank leakage. The tanks have not leaked during the 

life of the TFF. The contamination in the vaults for Tanks WM-185 and WM-187 (which resulted from 

back-siphoning events as discussed in Subsection 2.4.4) is considered in the PA. In all tank vaults, 

rainwater/snowmelt in-leakage through the vault roof has been pumped periodically from the vault sumps 

to waste tanks. 

Tank residuals remaining after cleaning and before grouting consist of a relatively small amount of 

solids (see Subsection 2.2) and contaminated flush water. Extensive mockup testing shows that most of 

the remaining flush water and some solids will be removed during the grouting process for stabilizing the 

residuals. This action will be accomplished by using the grout to push and corral the residuals toward the 

removal equipment (jet pumps). Any remaining residual liquid will be stabilized with a grout material. 

 

z. Analysis of residual metal contamination in process waste piping was performed in 2002. The estimate of residual radionuclide 

inventory in piping is based on this sampling and analysis. The Sampling and Analysis Plan for the Post-Decontamination 

Characterization of the Process Waste Lines from INTEC Tank Farm Facility Tanks WM-182 and WM-183 (INEEL 2001) was 

prepared to define the steps to collect data that represent residual in piping. The process waste lines in the WM-182 and WM-183 

tank systems have carried acidic waste in solution and have routinely been flushed after waste transfer with either acid or an acid 

and water flush. During closure of the tank systems, the piping was triple rinsed with water to remove loose residuals. Sections of 

horizontal and vertical process waste line have been removed from Tank WM-182. Samples from the decontaminated process 

waste lines will be collected and the data will be used to represent the effectiveness of triple rinsing all of the lines remaining in 

the WM-182 and WM-183 tank systems. The pipe to be sampled was removed from the system. The piping was filled with water, 

allowed to equilibrate for a minimum of 30 minutes, and sampled and analyzed for total metals. 

The concentration of metals in liquid (mg/L) SBW is proportional to the concentrations of radionuclides (mg/L). Therefore, the 

residual metal concentrations in piping can be used to calculate the residual radionuclide concentrations. The maximum 

concentration of each metal result in rinsate samples of piping was summed with the other 23 metals analyses. This yielded a 

concentration of 2.9 mg/L. The rinsate was collected from 46-cm (18-in.) lengths of the 6-cm (2-in.) diameter pipe. It is not 

assumed that the chemical properties of radionuclides and metals are the same. Therefore, the conservative assumptions shown 

below were used to ensure the decontamination factor for metals was not overstated, the fixed radioactive contamination would 

be included in the estimate, and to provide for the possibility of greater concentrations being found in other piping. This estimate 

of radionuclide concentrations in the piping was reviewed previously by the NRC and found to be acceptable (NRC 2003a), and 

was also found to be reasonable in the NRC TER (NRC 2006b). Several conservative assumptions were made to ensure this 

estimate is conservative: 

• The mass of metals equals the mass contained in 0.3 m (1 ft) of piping. Therefore, the starting value was one-third greater 

than indicated by the analysis. 

• The sample volumes were less than 1 L. The data were not adjusted downward to correspond with the actual sample 

volumes. 

• A safety factor of 500 was applied to the data to ensure a reasonably conservative estimate. 

The use of these data results in a mass of 15.5 kg of solids remaining in the piping (DOE-ID 2003b). 
28

Prior to grouting, the small amount of liquid waste in the vault sumps will be emptied using the existing 

steam-jet pumps. The lines connecting the vault sumps to the tanks will be grouted, followed by grouting 

of the vaults. 

2.3.2 Results of Tank and Ancillary Equipment Cleaning from 2002 to 2005 

The results of visual inspection and sampling and analysis performed after tank cleaning operations 

were completed provide evidence that the cleaning technology used for the 300,000-gal tanks, 30,000-gal 

tanks, and ancillary equipment effectively removes the majority of highly radioactive radionuclides from 

the tanks while keeping occupational exposure to radiation as low as reasonably achievable (ALARA). 

For the seven 300,000-gal tanks that have been cleaned as of May 2005 (Tanks WM-180 through 

WM-186), the washball and directional nozzle tank cleaning system was used to wash the tank walls and 

ceiling. The high-pressure water from the washball and directional nozzle also agitated the tank heel to 

suspend the solids and facilitate heel removal. The washball and directional nozzle were lowered into the 

tank through one of the tank risers. The water from the washball and nozzles hit the tank walls, roof, and 

heel, and dislodged the bulk of the contamination on the walls and ceiling of the tank to allow subsequent 

removal using the steam jets. A camera and lighting system was used to monitor the decontamination and 

heel removal effectiveness. Existing tank equipment and new equipment used for waste removal and 

decontamination operations were left inside that tank after these operations were completed for permanent 

disposal when practical. However, the washball and directional nozzle were decontaminated when moved 

to a new tank. The four remaining 300,000-gal tanks (WM-187 through WM-190) will be cleaned using 

the same methods after the SBW is removed for future treatment. 

Visual examination indicated that the water spray removed solids from the tank walls easily. After 

tank cleaning operations were completed, post-decontamination samples of the residuals were obtained 

from the tanks and ancillary equipment. Samples from Tanks WM-180 through WM-186 were collected. 

A minimum of five samples were collected from each tank. The remaining tank contents were agitated 

between sample collections to ensure random selection of the samples. Because few solids remained on 

the tank bottoms, only a few grams of solids from one tank (WM-183) were retrieved during the sampling 

activities. Attempts to sample solid material in other tanks failed because of a lack of solid material 

available to collect. Figures 11 and 12 are photographs of the interior of WM-183 before tank cleaning 

and after tank cleaning, respectively. Figure 11 shows the side of the tank and the cooling coil on the tank 

wall, which is approximately 20 cm (8 in.) above the tank floor. Figure 12 shows the cleaned tank floor 

and cooling coils, which are obscured by waste in Figure 11. After cleaning, the tank floor has only 

isolated areas of solid residual as shown in the photograph (see Figure 12). The cooling coil support 

brackets and welds on the floor and on the cooling coil base plate (0.97 cm [0.38 in.]) are clearly visible. 

Similar results were achieved for the other cleaned tanks. Post-decontamination photographs of the 

cleaned tanks are presented in Appendix A. 

The four 30,000-gal tanks (WM-103 through WM-106) were flushed with water to remove any 

residual waste. A tank cleaning system was not used for these tanks because of the low radionuclide 

concentrations remaining after previous tank flushing (see Subsection 2.2). Post-decontamination samples 

(at least five from each tank) from Tanks WM-103 through WM-106 were also collected (Portage 2006b). 
29

Figure 11. Photographs of the interior of TFF Tank WM-183 before tank cleaning. 
30

Figure 12. Photograph of the interior of TFF Tank WM-183 after tank cleaning. 

Sampling and analysis were performed in accordance with SAPs (ICP 2004a, 2004b; INEEL 2001, 

2002b, 2003) prepared in conjunction with the tank and ancillary equipment closure plans. Each SAP 

used the data quality objective (DQO) process to determine the sampling strategy, number of samples, 

and analytical methods to be used. Issues such as the representativeness of the samples obtained and the 

homogeneity of the population sampled are also addressed. The DQO process is a planning approach 

developed by the EPA for use in preparing sampling designs for data collection activities that support 

decision-making. The process is used to ensure that the type, quantity, and quality of data used in 

decision-making are appropriate for the intended application (EPA 2006a). The analytical results have 

been reported in a series of DQA reports (ICP 2006b, 2006c, 2006d, 2006e, 2006f, 2006g, 2006h, 2006i, 

2006j; INEEL 2006a, 2006b). A DQA is a scientific and statistical evaluation of the quality of the 

collected characterization data to determine whether the data can be used to meet the DQOs that were 

established (EPA 2006b). 

In this case, the results of post-decontamination sampling and analysis were used to determine the 

concentrations of the radioactive and hazardous constituents remaining in the tanks and ancillary 

equipment. Analysis results were used to confirm that the radionuclide concentrations met the closure 

requirements, and that they were bounded by the concentrations assumed in the conservative inventory in 

the 2003 PA (DOE-ID 2003b). The results were also used to estimate the residual TFF radionuclide 

inventories. Results of sampling and analysis of the residuals indicate that the radionuclide inventory in 

the tanks after cleaning is an order of magnitude lower than that estimated in the 2003 PA. A comparison 

of the PA conservative inventory and the Tank-WM-182 inventory at closure is presented in Appendix A. 

2.4 Residual Inventory in Tank Farm Facility 

Structures, Systems, and Components 

Although the tank and ancillary equipment cleaning activities have been shown to be successful, 

some residuals will remain in the TFF. These residuals contain radionuclides that will potentially create a 

radiation dose to members of the public. The residual inventory at closure, an estimate of the amount of 
31

residuals remaining after the TFF tank system is closed, is used in this 3116 Basis Document to evaluate 

this potential radiation dose and to support demonstration of compliance with Section 3116 criteria. 

The residual inventory at closure is based on validated analytical data from tanks sampled between 

2002 and 2005, after cleaning operations. (See Subsection 2.3.2 for a description of and references to the 

SAPs and DQA reports.) This inventory describes the radionuclide concentrations in the TFF tank system 

residuals, assuming that the residuals in the four 300,000-gal tanks and ancillary equipment remaining to 

be cleaned have a radionuclide inventory similar to that of the residuals in the seven tanks and ancillary 

equipment that have already been cleaned.aa Radionuclide concentrations are decayed to 2012, the year of 

final TFF closure. This date is used because performance objectives will not be met until all tanks and 

ancillary equipment are cleaned and stabilized with grout. (However, this inventory does not include the 

grout, nor is the inventory reduced by the amount of residuals that may be further removed by grouting 

operations.) The following subsections describe the residual inventory at closure for the 300,000-gal 

tanks, 30,000-gal tanks, sandpads, and piping; how each inventory was developed; and how the inventory 

will be used to support the demonstration of compliance with Section 3116 criteria. 

2.4.1 Models and Calculations Used to Develop Residual Inventory at Closure 

Historically, the INTEC uses 137Cs analytical data to aid in process control and estimate 

radionuclide inventories. This radionuclide is a major component in INTEC waste streams and is a strong 

gamma-emitting radionuclide, which makes it easy to detect with confidence in radioactive waste 

streams. Because no or limited analytical data are available for many other radionuclides, 137Cs data are 

also used to estimate values for other radionuclides. Radionuclide inventories for all tanks are presented 

in Appendix A. The 137Cs sample data and ORIGEN2 (Croff 1980) and analytical ratios, as explained 

below, are used to estimate radionuclide inventories when analytical data for specific radionuclides are 

not available. 

The gamma-emitting radionuclide 137Ba contributes heavily to the overall activity in the inventory. 

This activity level has a tendency to “overshadow” the activity of lower energy or lower activity 

radionuclides (e.g., 14C). This occurs because the intensity and high concentration of the gamma-emitting 

radionuclides create overlap, scatter, and other interferences during analysis, which in effect, result in a 

loss in instrument sensitivity. Depending on the levels, this results in either difficulty or the inability of 

the laboratory to measure low-energy/low-activity nuclides. 

The ORIGEN2 model is used to predict radionuclide inventories based on nuclear fuel types 

(e.g., Al, Zr, and stainless steel) that have been processed at INTEC. Input parameters to the model are 

adjusted to allow the model to predict SBW radionuclide inventories based on weighted averages of the 

different fuel types that have been processed. The INL Site uses radionuclide and 137Cs values from the 

ORIGEN2 model to calculate a radionuclide-to-137Cs ratio called the ORIGEN2 ratio for each 

radionuclide. Analytical results from SBW samples that contain reliable data for a specific radionuclide 

and 137Cs are used to calculate an analytical ratio. All reliable analytical data are used to generate average 

analytical ratios for radionuclides in SBW. These two ratios, the ORIGEN2 and analytical ratios, have 

been modeled and published and are informally called the Wenzel Tables (EDF-FDO-006, 1997). These 

tables have been integrated into a FORTRAN program, which is used to calculate radionuclide data 

(when analytical data are lacking) for SBW based on 137Cs sample results (see the following example). 

 

aa. As explained previously in this 3116 Basis Document, DOE will sample the remaining tanks after cleaning pursuant to DOE’s 

Atomic Energy Act (42 USC 2011 et seq. 1954) responsibilities and will share this information with the NRC, as a matter of 

policy, to facilitate NRC monitoring, in coordination with the State of Idaho, pursuant to Section 3116(b). 
32

Example of using ORIGEN2 and analytical ratios to calculate radionuclide data for SBW: 

• Data from Wenzel tables 

- The concentration of 238U is 2.20E–08 Ci/L (radioactive decay to 2012) as calculated by the 

ORIGEN2 model for SBW 

- The 137Cs concentration in 2012 for SBW is 2.80E−02 Ci/L (based on analytical results from 

historical analysis of SBW) 

• ORIGEN2 and analytical ratios 

- The ratio for 238U is calculated by dividing the concentration of 238U by the concentration of 

137Cs or 2.20E−08/2.80E−02, yielding an analytical ratio = 8.20E−07 for 238U since it was 

based on ORIGEN2 results 

• Concentrations based on 137Cs analytical results and ORIGEN2 and analytical ratios: 

- Based on the analytical data gained from gamma spectroscopy, the concentration of 137Cs 

detected in Tank WM-182 is 2.20E−04 Ci/L (radioactive decay to 2012). Therefore, the 

estimated concentration in 2012 for 238U based on the 137Cs analytical data and the analytical 

ratio would be 2.20E−04 Ci/L x 8.20E−07 = 1.80E−10 Ci/L. 

(Note: The values used in the calculations have been rounded.) 

The analytical ratio is used preferentially over the ORIGEN2 ratio to calculate TFF inventories 

when available, because the analytical ratio is modified to the measured 137Cs concentrations 

(DOE-ID 2003b). These analytical ratios are used to provide an inventory that has a basis in analytical 

data. The ORIGEN2 and analytical ratios are also used to validate analytical results of 

difficult-to-measure isotopes when they are found in waste streams with moderate to high radionuclide 

concentrations. The results of laboratory analyses are used to modify the ORIGEN2 results to better 

reflect the concentrations of radionuclides in the inventory. 

Expensive and time-consuming separation techniques are available that increase the likelihood of 

obtaining more reliable laboratory results. However, even when these techniques are applied, a detection 

of a radionuclide may not be made or may be at concentrations that have a high degree of uncertainty. In 

lieu of these techniques, the industry has developed and adopted the use of tools such as the ORIGEN2 

model to obtain reasonable estimates of low-energy/low-activity components when evaluating large 

inventories of radionuclides. This practice is widely recognized throughout the industry as a means to 

obtain reasonable estimates of radionuclides that are difficult or impossible to measure in the presence of 

high-energy/high-activity components (DOE-ID 2003b).bb

 

bb. During consultation, NRC expressed concerns with DOE’s use of ORIGEN2 ratios to estimate the concentrations of 

radionuclides not detected in the post-cleaning sampling. On pages 31–32 of its TER (NRC 2006b), the NRC noted that DOE had 

addressed NRC’s concerns regarding the use of the ORIGEN2 model (see DOE’s response to NRC RAI Comment 2 

[ICP 2006l]), and the NRC concluded that the methodology used to calculate the post-cleaning inventories is reasonable. 
33

2.4.2 Residual Inventory at Closure in 300,000-gal Tanks 

The residual inventory at closure for the 300,000-gal tanks is based on analytical data from samples 

collected after tank cleaning from 2002 through early 2005. Of the 300,000-gal tanks, Tanks WM-180, 

WM-181, WM-182, WM-183, WM-184, WM-185, and WM-186, and the associated ancillary equipment 

have been cleaned. Analytical data from the cleaned tanks are the primary source of information for 

creating the residual inventory at closure. The radionuclide concentrations used to calculate inventories 

for each tank are based on analytical data decayed to 2012. Concentrations for radionuclides for which 

samples were not analyzed or for which radionuclides were not detected are replaced with the estimates 

calculated from the ORIGEN2 and analytical ratios, and analytical results for 137Cs as described in 

Subsection 2.4.1. Samples were analyzed for all radionuclides that were important dose contributors. Due 

to the nature of the SBW, some radionuclides were hard to detect without specialized analyses or the 

development of new analytical methods. These radionuclides were not analyzed if they were not 

considered to be significant dose contributors. Therefore, these radionuclide concentrations were 

estimated as described in Subsection 2.4.1. 

The residual solids data are based on a sample of solids collected from Tank WM-183. Although 

seven 300,000-gal tanks have been cleaned, only one sample of solid material was able to be collected 

because of the lack of residual solids in the cleaned tanks. 

The residual solids inventory for those radionuclides modeled using the ORIGEN2 analytical ratios 

is based on the mean of the 137Cs concentrations from Tank WM-188 collected prior to cleaning (decayed 

to 2012). Tank WM-188 137Cs data are used because this tank had the highest measured 137Cs 

concentration of any TFF tank before or after cleaning and because only one solid sample could be 

collected from the tanks after cleaning operations. Therefore, using the higher concentration data from 

Tank WM-188 is conservative. Concentrations estimated from the analytical ratios are used for 

radionuclides not able to be detected. 

The single solid sample from Tank WM-183 introduces uncertainty to the radionuclide inventory 

for the TFF tanks. While it is difficult to quantify the uncertainty, a qualitative uncertainty discussion is 

included in the following paragraphs to show that the inventory for the entire TFF is reasonably 

conservative. The inventory that is shown in this 3116 Basis Document is based on the data collected 

from the cleaned tanks. The PA inventory is compared to the residual inventory in Appendix A, Table 

A-12. 

In all of the cleaned 300,000-gal tanks, five liquid samples were collected from three locations, 

with two of the locations sampled twice. Limited access to the tank through existing risers requires this 

type of sample collection effort. While all tanks were sampled, only a few grams of solids were collected 

from Tank WM-183. Many factors contribute to the overall reasonably conservative inventory based on a 

single sample of solids, which are: 

• Tank WM-190 has never been used and has been only slightly contaminated. The solid sample is 

conservative for this tank. To ensure a conservative estimate, residual mass from Tank WM-182 

was used for tanks that have not yet been decontaminated, including WM-190. The estimated mass 

of solid residuals by tank are as follows: Tank WM-180 (542 kg), Tank WM-181 (246 kg), 

Tank WM-182 (1,238 kg), Tank WM-183 (702 kg), Tank WM-184 (558 kg), Tank WM-185 

(720 kg), and Tank WM-186 (334 kg). 

• Analytical data are the primary source used to develop the radionuclide tank inventory. When these 

analytical data are not available, the ORIGEN2 model is used to predict the inventories. 
34

• A 137Cs ratio of 1.8 was established for solids. This value is the mean of all solid samples collected 

from Tank WM-188. These samples were collected in 1999, prior to tank cleaning. Use of this ratio 

increases the total radioactivity of all the radionuclides that were not detected. 

• Although additional tank residuals may be removed during the grouting process (as discussed in 

Subsection 2.3.1), the residual inventory does not take credit for this additional residual removal. 

• Actual analytical values from each tank were used for the 137Cs ratio for liquids. 

• Based on mass, WM-182 has the highest activity of any tank cleaned to date. 

• An estimated density of 1.4 g/mL was applied for tank solids (EDF-TST-001, 2000). 

Table 1 presents an example of the residual inventory at closure for a 300,000-gal tank. The 

inventory for Tank WM-182 is shown because this inventory contains the largest amount of residual 

radioactivity of the cleaned tanks. Inventories at closure for all tanks are presented in Appendix A. The 

analytical data indicate the decontamination efficiency is greater than planned, as shown in Subsection 

A-2 of Appendix A. The residual inventory at closure indicates that approximately 2,400 Ci remain in 

Tank WM-182. The total amount of radioactivity is associated almost entirely with residual solids. 

Approximately 2 Ci remain in the liquid in Tank WM-182. 

Table 1. Residual inventory at closure for Tank WM-182.

Radionuclide

Residual Liquids 

(Ci)a,b

Residual Solids 

(Ci)a,b

Total Residuals 

(Ci)c

241Am 5.30E−04 5.43E+00 5.43E+00 

137mBad 1.11E+00 1.14E+03 1.14E+03 

242Cm 6.58E−07 1.32E−03 1.32E−03 

137Cs 1.11E+00 1.14E+03 1.14E+03 

14C 5.39E−08 4.90E−06 4.96E−06 

129I 1.12E−06 7.73E−04 7.74E−04 

3

H 1.66E−05 7.17E−01 7.17E−01 

94Nb 4.03E−05 2.06E−01 2.06E−01 

59Ni 1.25E−05 2.51E−02 2.51E−02 

63Ni 1.43E−03 2.86E+00 2.86E+00 

237Np 2.71E−07 4.70E−02 4.70E−02 

238Pu 2.47E−03 1.14E+01 1.14E+01 

239Pu 2.44E−04 3.40E+00 3.40E+00 

240Pu 6.74E−04 1.35E+00 1.35E+00 

241Pu 5.89E−04 1.95E+01 1.95E+01 

242Pu 4.93E−07 9.87E−04 9.88E−04 

90Sr 2.41E−01 2.32E+01 2.34E+01 

99Tc 4.54E−05 7.64E−01 7.64E−01 
Table 1. (continued). 

35

Radionuclide

Residual Liquids 

(Ci)a,b

Residual Solids 

(Ci)a,b

Total Residuals 

(Ci)c

90Y 2.41E−01 2.32E+01 2.34E+01 

Totale Ci all radionuclides 3 2,391 2,394 

a. Radionuclide inventories are based on (1) heel residuals that are estimated using remote video inspection of cleaned tank 

internals to map out estimates of depth of remaining residual solids and liquids across tank bottoms using tank internal reference 

points of known height, (2) best estimated radionuclide concentrations from past and recent samples as calculated in the 

associated engineering design file (Portage 2005b), and (3) radioactive decay to 2012. 

b. Analytical results for Tank WM-182 were used to calculate the liquids inventory at closure for 241Am, 14C, 137Cs, 154Eu, 3

H, 129I, 237Np, 238Pu, 239Pu, 241Pu, 125Sb, 90Sr, 99Tc, 234U, and 90Y. Analytical results for Tank WM-183 were used to calculate the 

solids inventory at closure for 241Am, 137mBa, 60Co, 137Cs, 129I, 94Nb, 238Pu, 239Pu, 125Sb, 90Sr, 99Tc, 234U, and 90Y. Inventories at 

closure for all remaining nuclides were calculated using the methodology presented in Subsection 2.4.1. 

c. Radionuclides decayed to 2012. 

d. A 1:1 ratio is assumed for 137Cs to 137mBa as a conservative estimate of radionuclide inventory; based on decay probability, 137mBa is approximately 95% of the 137Cs inventory. 

e. Radionuclides shown are contributors in PA dose calculations or regulated by concentration limits in 10 CFR 61.55. The totals 

are based on the entire inventory of radionuclides. This inventory is based on the 95% upper confidence limit of the radionuclide 

concentrations from liquid samples and an average of the estimate of remaining mass. Since only one sample of solid residual 

was able to be obtained, that concentration was used with the estimated remaining mass to determine the total solid residual. 

Tank volume estimates in the 300,000-gal tanks for solid residual and interstitial water used in the 

residual inventory at closure are based on viewing videotapes of the tanks taken before, during, and after 

the final cleaning and sampling events. From the videotapes, residual solid depths were estimated by 

comparing the solids depths to the cooling coil support brackets and associated welds. Depth assumptions 

were based on the bottom weld and stainless steel bracket thickness measuring 0.97 cm (0.38 in.). 

Close-ups from the video were critical in determining the depths of residuals next to the brackets and 

areas where no apparent solid residuals were observed. Depths of solid residuals ranged from 0 to 

0.97 cm (0 to 0.38 in.). Areas where no solids were present on the bottom of the tank were apparent by 

the reflection from the stainless steel bottom. The volume of solids and the density of the solid material 

were used to determine the mass of residual in each tank (Portage 2005a, 2005b, 2005c, 2005d, 2005e, 

2005f, 2006a). Videos and photographs of the tank walls show staining and discoloration, and no 

discernible buildup of residuals. Therefore, no source term for the tank walls was included in the tank 

inventory. Inventories at closure for all tanks are provided in Appendix A. 

Residual solids depth measurements were recorded and entered into an AutoCAD 2004 drawing in 

association with the cooling coil support locations. Cartesian coordinates (x and y) were determined for 

each cooling coil support and associated residual depth measurement. The residual solids depth data and 

associated Cartesian coordinates for the cooling coil supports were then exported to Surfer 8. The residual 

solids volume was estimated using Kreiging methods in Surfer 8, with point-Kreiging being the preferred 

gridding method. A simple variogram model was used with linear interpretation using a slope equal to 1.0 

and anisotropy equal to 1.cc

Contour, post, and surface plot maps are provided for the residual solids volume estimates in the 

engineering design files prepared for each tank volume (Portage 2005a, 2005b, 2005c, 2005d, 2005e, 

2005f, 2006a). Residual solid volume estimates were determined using Surfer 8 with 0 cm (0 in.) as a 

 

cc. Kreiging is an interpolation technique that uses variograms as weighting functions. Variograms are measures of the continuity 

of spatial phenomena expressed as an average squared difference between measured quantities at different locations. Anisotropy 

is a state in which a physical characteristic varies when measured in different directions. 
36

lower surface boundary and the Trapezoid Ruledd as the preferred volume estimate method. The depth of 

liquid above the interstitial waste was estimated to be approximately 3 cm (1 in.) by comparison to the 

height of the steam-jet lines in the post-decontamination video. 

The residual inventory at closure assumes that the same degree of radionuclide removal will be 

achieved in the tanks remaining to be cleaned (WM-187, WM-188, WM-189, and WM-190) as was 

achieved in the tanks that have been cleaned. As described in Subsection 2.2, Tank WM-190 is an 

emergency spare tank and has never been used to store waste. However, this tank was contaminated with 

a small volume of first-cycle extraction process waste when the waste passed inadvertently through a 

transfer valve. As noted previously, Tank WM-182 contains the largest amount of residual radioactivity 

of the cleaned tanks. 

Although there may be variability between the tank inventories, as the NRC noted in its TER, for 

the reasons explained below, these assumptions are conservative. Solids and liquids from the cleaned 

tanks have been transferred to Tank WM-187 via existing piping. The decontamination fluids in Tank 

WM-187 were evaporated to provide space for additional decontamination fluids as additional tanks were 

decontaminated and emptied. This tank will contain all of the solids removed from the cleaned tanks and 

some of the liquid. Most of the liquid has been evaporated using the evaporator tank system. Although 

combining all of the solids in one tank may make cleaning this tank slightly more difficult, the solids in 

the tanks have been found to be made up of small particles that are removed easily using the tank cleaning 

system. Particle size distribution studies performed on samples collected from Tanks WM-182 and 

WM-183 conclude the particles that comprise the solids are small and easily redispersed after settling 

(INEEL 2000b). The basic findings of the particle size study were substantiated during the course of tank 

cleaning. The solids were dispersed easily using high-pressure water sprays, and the particles remained 

suspended in the liquid for a significant period of time before they settled to the bottom. The 

characteristics of the solid particles are not known to have been altered by transferring to Tank WM-187. 

2.4.3 Residual Inventory at Closure for 30,000-gal Tanks 

The residual inventory at closure for the 30,000-gal tanks is based on analytical data from postcleaning sampling. Solid residual samples were not collected because an adequate volume of material was 

not present in the tanks (Portage 2006b). A film layer was observed on the lower half of all four tanks that 

appeared to be algae or another form of biological growth. This film layer was clearly not residual solids 

as found in the 300,000-gal tanks and is not likely to contain any significant radioactivity. However, to 

establish a conservative estimate, the film is assumed to be 5 mil (0.005 in.) thick and contains the 

concentrations of radionuclides found in Tank WM-183 solid samples (Portage 2006b). The inventories 

for each 30,000-gal tank vary from 36.2 to 36.7 Ci. Table 2 presents the inventory for Tank WM-106, 

which has the highest remaining Ci content of the 30,000-gal tanks. 

 

dd. The Trapezoid Rule is a way to approximate definite integrals. 
37

Table 2. Residual inventory at closure for Tank WM-106 (Portage 2006b). 

Radionuclidea

Residual Liquids 

(Ci)b

Residual Solids 

(Ci)b

Total Residual 

(Ci)c

241Am 1.85E–06 6.36E–03 6.36E–03 

137mBa 2.96E–02 1.72E+01 1.72E+01 

242Cm 1.75E–08 1.99E–05 1.99E–05 

137Cs 2.96E–02 1.72E+01 1.72E+01 

14C 5.28E–10 7.41E–08 7.46E–08 

129I 2.98E–08 1.17E–05 1.17E–05 

3

H 3.90E–06 1.08E–02 1.08E–02 

94Nb 1.87E–05 3.10E–03 3.12E–03 

59Ni 3.34E–07 3.79E–04 3.80E–04 

63Ni 8.75E–06 4.32E–02 4.32E–02 

237Np 3.51E–08 7.10E–04 7.10E–04 

238Pu 1.65E–05 1.73E–01 1.73E–01 

239Pu 2.66E–06 5.14E–02 5.14E–02 

240Pu 1.80E–05 2.04E–02 2.04E–02 

241Pu 1.22E–05 2.95E–01 2.95E–01 

242Pu 1.31E–08 1.49E–05 1.49E–05 

90Sr 2.84E–01 3.50E–01 6.33E–01 

99Tc 8.32E–05 1.15E–02 1.16E–02 

90Y 2.63E–02 3.50E–01 3.76E–01 

Totald

 Ci all radionuclides 0.6 36.1 36.7 

a. Radionuclides based on results from the 2003 TFF PA (DOE-ID 2003b). 

b. Analytical results from sampling of Tank WM-106 were used to calculate the liquids inventory at closure for 241Am, 14C, 244Cm, 60Co, 137Cs, 154Eu, 3

H, 129I, 94Nb, 63Ni, 237Np, 238Pu, 239Pu, 241Pu, 90Sr, 99Tc, 234U, and 235U. Analytical results for Tank 

WM-183 were used to calculate the solids inventory at closure for 241Am, 137mBa, 60Co, 137Cs, 129I, 94Nb, 238Pu, 239Pu, 125Sb, 90Sr, 99Tc, 234U, and 90Y. Inventories at closure for all remaining nuclides were calculated using the methodology presented in 

Subsection 2.4.1. 

c. Radioactive decay to 2012. 

d. Radionuclides shown are contributors in PA dose calculations or regulated by concentration limits in 10 CFR 61.55. The 

totals are based on the entire inventory of radionuclides.

2.4.4 Residual Inventory at Closure for Sandpads 

Tanks WM-182 through WM-190 rest on a 15-cm (6-in.) layer of commercial-grade sand 

overlaying a concrete slab approximately 0.76 m (2.5 ft) thick. A 15- by 15-cm (6- by 6-in.) concrete curb 

encloses the sandpad area. The commercial-grade sand occupies approximately 23.39 m3

 per tank and 

provides a cushion for the stainless steel tanks to prevent shifting. A drain-sump system removes liquid 

(the liquid that accumulates is mostly rainwater; a small amount of liquid from secondary containments 

does enter vault sumps if a valve leaks or during some maintenance activities) that accumulates in the 

vaults. Typically, the vault area receives an accumulation of rainwater and snowmelt in the spring each 

year. 
38

The sandpads at the TFF are between the tanks and vault floors. The sandpad at the bottom of the 

concrete vault was designed to cushion the tank bottom. Additionally, the curb has drain holes to allow 

liquid to drain from the sandpad beneath the tanks to the sumps outside the curb. The sumps were 

periodically used to remove liquid that had accumulated in the vaults. Liquid was present in the vault 

from sources such as precipitation and irrigation activities. The sandpads beneath Tanks WM-185 and 

WM-187 are contaminated due to accidental releases into the vaults in March 1962. These are the only 

tanks where such an incident occurred. A description of the back-siphoning event into the tanks is 

available in Latchum et al. (1962). Before and after these releases, water from precipitation, spring runoff, 

and irrigation was pumped out of the tank vaults at least semiannually. The residual inventory predicted 

for 2012 is based on 38 “flushing” eventsee when water infiltrated to the vault from leaks in the tank/vault 

roof and was then jetted out of tank vaults. The amount of water collected in the vaults is documented in 

the INEL internal memorandum, “HLLW Tank Sump Transfer History.”ff The report documents over 100 

transfers from tank vaults associated with WM-185 and WM-187. Reports are not available before 1962 

but water transfers from the vaults are known to have occurred. Water transfers continue from vault 

sumps. Thirty-eight events correspond with each year from the back-siphoning events in 1962–2000. 

Water was jetted from the tank vault sumps at least twice yearly and will continue until each tank is 

closed. The actual number of water transfers from the tank vaults and associated leaching of radionuclides 

from the sandpad likely exceeds 130 for each vault to date. 

The results from the analysis are provided for each radionuclide in the year 2012, which is the 

expected closure date. These activities have been estimated by modeling diffusion of radionuclides from 

the liquid into the sandpad for the contaminating event in 1962. 

The approach uses diffusion of radionuclides to the sand. The sandpad was subsequently flushed by 

precipitation infiltration through the roof of the tank and radionuclides were decayed to 2012. The 

approach assumes radionuclides enter the sandpad through diffusion at the curb/sandpad interface. This 

assumption is reasonable since the contaminated liquid, which entered the vault in 1962, appears to have 

filled the concrete vault from the bottom toward the vault top (i.e., there was no horizontal pressure 

gradient driving the liquid through the sandpad). The radionuclides entered the sandpad by diffusion. 

Once contained in the sandpad, radioactive decay and periodic (annual) flushing are modeled to 

determine the present-day activities. Flushing is assumed to have occurred through 2000 (38 flushings) 

and then the inventory is decayed over 12 years to estimate 2012 activities of all radionuclides. The 

diffusion model is used because before the vault partially filled, the vault was filled with water from a 

spring runoff event. The sandpad was saturated with water because water had filled a portion of the vault 

recently and in the past. The back-siphoning event that caused the waste to enter the vault was a result of 

jetting the water to the tank. 

After the accident, the mass of each radionuclide in the sandpad is assumed to be impacted by 

radioactive decay and flushing. These processes are modeled by assuming the mass of radionuclides in 

the sandpad undergoes radioactive decay for a period of time representing the length of time between 

flushing events. Then the system is flushed. During flushing, the sandpad is assumed to be saturated and 

the radionuclides are partitioned at equilibrium between the liquid and solid phases according to their 

sorption coefficients (Kd values). The flushing is assumed to remove all liquid and radionuclides that have 

partitioned into the liquid phase from the sandpad except for the residual liquid. Thus, the radionuclides 

remaining in the sandpad are contained in the residual liquid and sorbed onto the sand. The mass of 

radionuclides in the sandpad is again calculated assuming radioactive decay occurs over a period of time 

 

ee. The number and impacts of these events are further explained in DOE’s response to NRC RAI Comment 3 (ICP 2006l). 

ff. Rebish, K. J., INEL, to J. M. Roberts and B. H. O’Brien, INEL, October 15, 1993, “HLLW Tank Sump History,” KJR-10-93. 
39

that represents the time between flushing events. The cycle of modeling radioactive decay and flushing is 

repeated until 2000. In the absence of data, it was assumed that the flushing events occurred once a year 

for 38 years (i.e., once a year the seasonal effects of precipitation and irrigation are purged from the 

vaults). After activities in 2000 are computed, the results are decayed over 12 years to 2012, which 

corresponds to the time of TFF closure. 

To determine the mass of radionuclides in the sandpad following the initial event in 1962, the 

radioactive decay and flushing of the sandpad is modeled. A FORTRAN computer code was developed to 

model radioactive decay and then flushing of the sandpad based on partition coefficients and the volume 

of sand, void space, and residual saturation. The mathematical formulation of this approach is provided 

below; the FORTRAN source code is presented in Appendix A of the PA (DOE-ID 2003b). 

The accidental spill in 1962 is modeled assuming the liquid is present in the vault for 24 hours. The 

24-hour time is consistent with the estimated time from the beginning of the spill to the time when the 

spilled liquid was removed by pumping the vault (Latchum et al. 1962). The area of the flux surface 

(i.e., the liquid/sandpad interface) is assumed to be an annular region having a radius equal to the tank 

radius and a thickness equal to the maximum sandpad height. That is, this area is approximately equal to 

the area of a rectangle with a length of 2 πr, where r is the radius of the tank (7.6 m [25 ft]), and a width 

of 15 cm (6 in.) (i.e., the maximum estimated height of the sandpad). 

The initial amount of each radionuclide in the tanks at the time of the accidental spills was 

evaluated with limited sampling of Tanks WM-185 and WM-187 on February 14, 1962. Due to the 

limited number of radionuclides provided by this sampling analysis, an alternative method using 

information from a leak from Tank WM-181 was used to determine the inventory (EDF-CPP-058, 1997). 

Using a pseudo Al-clad fuel with an initial 235U enrichment of 93% and a burnup of the processed fuel of 

18%, EDF-CPP-058 evaluated the expected radionuclide content of the tank. The fuel from MTR Cycle 

No. 198 (Dykes 1963) was taken as typical for the fuel processed. The reactor contained 4,842 g of 235U 

and had 684 MWd of operation over a 417-day period. For calculation purposes, inventories were 

normalized to the activity in a typical 200-g element. The ORIGEN2 data were corrected to the 

concentration of 137Cs in Tank WM-185 one month before the incident. Data for Tanks WM-185 and 

WM-187 were shown in the record of the incident (Latchum et al. 1962). Tank WM-185 was used 

because of the slightly higher 137Cs concentration of 1.71 Ci/L. 

These assumptions and the data used in the analysis are considered reasonably conservative. That 

is, available data that represent the modeled system are used. When data are not available, conservative 

assumptions are made. For example, the surface available for diffusive transport in this analysis is 

assumed to remove all radionuclides that reach the surface instantaneously (i.e., the liquid/sandpad 

interface). Thus, mass transfer is greater under these conditions than if a concentration gradient is 

assumed to exist in the sandpad. With this and other assumptions and the available data, the analysis is 

expected to provide a reasonable and conservative estimate of the mass of each radionuclide contained in 

the sandpad. That is, the actual sandpad inventories are not expected to be larger than these estimates. 

The amount of radionuclides that have been flushed from the sand has not been measured directly 

and cannot be determined from the analysis of the liquid in the vaults because other sources (such as vault 

sumps) contribute to vault contamination. However, because of the numerous flushing events that have 

occurred, any additional cleaning by flushing would not provide a significant benefit. 
40

Table 3 presents a conservative residual inventory at closure for the sandpadsgg based on the 

analysis of the two contaminated sandpads. This inventory represents the sandpad assumed under Tanks 

WM-185 and WM-187 in the PA. The residual inventory at closure was established using various 

numerical models to predict the initial radionuclide concentrations, the subsequent removal by the 

periodic flushing of the sandpads, and radioactive decay to estimate concentrations in 2012 

(DOE-ID 2003b; Portage 2005a, 2005b, 2005c, 2005d, 2005e, 2005f, 2006a). 

 

gg. Although the NRC noted uncertainties in the sandpad inventory due, in part, to the inability to directly sample the sandpads 

under the tanks, on page 36 of its TER (NRC 2006b) the NRC concluded, “... DOE Idaho made a good faith effort to evaluate 

the uncertainty in the sand pad inventory through additional sensitivity and uncertainty analysis, modeling, and analysis of 

indirect sampling data. The sand pad inventory is adequate for the purpose of assessing DOE Idaho’s demonstration of 

compliance with NDAA Criteria 2 and 3.” 

Table 3. Sandpad residual inventory at closure (Ci per sandpad). 

Radionuclides 

Sandpad Inventory 

(Ci)a

241Amb

 1.89E+00 

137mBa 1.65E+03 

242Cm 1.38E–05 

137Csb

 1.65E+03 

14Cb

 3.90E–07 

129I

b

 1.08E–06 

3

H 3.10E–22 

94Nb 2.29E–02 

63Ni 1.69E–10 

237Np 3.72E–04 

238Pub

 2.06E+00 

239Pub

 1.57E+00 

240Pub

 3.54E–01 

241Pu 2.28E+00 

242Pu 5.69E–05 

90Srb

 2.49E+02 

99Tcb

 2.02E–12 

90Y 2.49E+02 

Totalc

 Ci all radionuclides 3,850 

a. Radioactive decay to 2012. 

b. Radionuclides based on results from the 2003 TFF PA (DOE-ID 2003b). 

c. Radionuclides shown are contributors in PA dose calculations or regulated by concentration limits in 10 CFR 61.55. The total

is based on the entire inventory of radionuclides, not just the key radionuclides presented in this table. 
41

2.4.5 Residual Inventory at Closure for Contaminated Piping, Encasements, and 

Valve Boxes 

Liquid waste transfers to, from, and between the tanks are managed through a series of lines, 

valves, and diversion boxes. Piping that penetrates tanks, vaults, or valve boxes is part of this 3116 

Determination. For each phase of the TFF closure, contaminated piping, encasements,hh and valve boxes 

will be isolated and decontaminated. In general, pipelines entering or leaving the INTEC TFF will be 

isolated at the first valve box or manhole inside the TFF fence. Lines that begin and end within the TFF 

will be closed entirely; however, they may be closed in stages, depending on the TFF waste handling 

needs. Lines that enter the TFF boundary from outside will be closed at a convenient junction, such as a 

valve box. The remaining line section will be closed with the tank or facility from which it originates. 

Waste transfer piping will be decontaminated by rinsing each line with three volumes of water. 

Valve boxes are underground, stainless steel-lined concrete boxes that house waste transfer valves 

through which liquid waste is routed. During each TFF closure phase, the inside surfaces of each valve 

box will be washed, and the valve box will be decontaminated with demineralized water. Based on past 

maintenance decontamination practices, this protocol will be adequate to clean the interior pipe surfaces 

that have contacted process solutions. Piping that penetrates tanks and encasements that penetrate vaults 

will be filled with grout to the extent practical (INEEL 2000c). 

Valve boxes and piping encasements that penetrate tanks or vaults will be drained of liquids 

resulting from decontamination efforts or leakage (DOE-ID 2003a) and then grouted along with their 

drain lines. A residual inventory at closure is not calculated for valve boxes because they do not normally 

contain process solutions; they are designed so that any leakage into the valve box would drain to sumps, 

which is then jetted to tanks; and the surface area of the valve boxes is small compared with the piping 

and tanks. Valve boxes have been (or will be) decontaminated and samples collected of the remaining 

water in the sump, but an inventory was not prepared. The valve box sumps are approximately 0.3 m (1 

ft) square and have an outlet that is approximately 5–8 cm (2–3 in.) above the bottom of the sump. The 

liquid that is sampled is taken from the 5–8 cm (2–3 in.) of volume in the sump bottom. Almost all of the 

liquid is removed by sampling. 

The TFF process lines are stainless steel pipe with secondary containment or encasements 

(see Subsection 2.1). Process waste lines and associated encasements are covered with 1.5–4.6 m 

(5–15 ft) of soil with about 30% being within 3 m (10 ft) of the surface (INEEL 2000c). During TFF 

closure, the active encasements will be decontaminated using a pressure-washing system. Most pipe 

encasements open into the associated valve boxes or vaults or have drains that drain into the associated 

vault. Each encasement is sloped to allow drainage and is flushed during tank isolation activities. The 

grout will be pumped into an encasement and will run or be pushed to the end of the encasement 

(INEEL 2000c). A residual inventory at closure is not calculated for pipe encasements. The pipe 

encasements are similar to valve boxes in that they do not normally contain process solutions; they are 

designed so that any leakage into the pipe encasements would drain to sumps, which is then jetted to 

tanks; and the contamination in the secondary containment is addressed by using the safety factor of 500 

established for the piping inventory. The safety factor of 500 is described in Subsection 2.3.1. 

Tank vessel off-gas piping will not be flushed because the piping did not transport process 

solutions (INEEL 2000c). The vessel off-gas lines will be grouted and permanently capped. A residual 

 

hh. Piping encasements are enclosures (secondary containment) surrounding the waste transfer piping to prevent escape of 

radioactive waste. Any fluid that leaked from a process line drained into an encasement and then into a valve box or vault sump. 
42

inventory at closure is not calculated for tank vessel off-gas piping because the piping has never contained 

process solutions. 

The residual inventory at closure for TFF piping is based on analytical results from sampling pipe 

sections that had been removed from the Tank WM-182 process waste lines after decontamination. The 

samples were obtained and analyzed as described in the associated SAP (INEEL 2001). From these 

results, the amount of residuals remaining in 3,231 linear m (10,600 linear ft) of process waste piping is 

calculated to be 15.5 kg of SBW solid residuals in the piping. The radionuclide concentration from the 

tank residual inventory is then apportioned to the mass to estimate a conservative residual inventory at 

closure for the piping. Table 4 shows the total radioactivity by radionuclide for the piping. 

Table 4. Residual inventory at closure for piping. 

Radionuclides Piping Inventory (Ci)a,b

241Amc 5.28E−03 

137mBa 1.43E+01 

242Cm 1.65E−05 

137Csc

 1.43E+01 

14Cc 6.21E−08 

129I

c 9.70E−06 

3

H 1.08E–02 

94Nb 2.58E−03 

59Ni 3.15E−04 

63Ni 3.58E−02 

237Np 5.89E−04 

238Puc 1.43E−01 

239Puc 4.27E−02 

240Puc 1.69E−02 

241Pu 2.44E−01 

242Pu 1.24E−05 

90Src 2.93E−01 

99Tcc 9.57E−03 

90Y 2.93E−01 

Totald

 Ci all radionuclides 30 

a. Radionuclide inventories are based on (1) residual solids in piping at 0.305 g/m and 3,231 linear m (10,600 linear ft) of 

piping, (2) WM-182 radionuclide concentrations in the SBW solids, and (3) radioactive decay to 2012. 

b. Analytical results for Tank WM-183 were used to calculate the solids inventory at closure for 241Am, 137mBa, 60Co, 137Cs, 129I, 94Nb, 238Pu, 239Pu, 125Sb, 90Sr, 99Tc, 234U, and 90Y. Inventories at closure for all remaining nuclides were calculated using 

the methodology presented in Subsection 2.4.1. 

c. Radionuclides based on the 2003 TFF PA (DOE-ID 2003b). 

d. Radionuclides shown are contributors in PA dose calculations or regulated by concentration limits in 10 CFR 61.55. The 

totals are based on the entire inventory of radionuclides. 
43

3. SECTION 3116 OF THE RONALD W. REAGAN NATIONAL 

DEFENSE AUTHORIZATION ACT FOR FISCAL YEAR 2005 

Section 3116(a) of the NDAA provides that radioactive waste that results from reprocessing SNF is 

not “high-level radioactive waste” if the Secretary of Energy determines, in consultation with the NRC, 

that the waste meets certain specified criteria. 

Section 3116(a) of the NDAA states: 

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 high-level 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, 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 
44

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

with the Commission. 
45

4. WASTE DOES NOT REQUIRE PERMANENT ISOLATION IN A 

DEEP GEOLOGIC REPOSITORY FOR SPENT FUEL OR 

HIGH-LEVEL RADIOACTIVE WASTE 

Section 3116(a) of the NDAA 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 high-level radioactive waste. 

Under Section 3116(a), certain wastes from reprocessing are not “high-level radioactive waste” if 

the Secretary, in consultation with the NRC, determines that certain criteria are met. Section 3116(a) sets 

out two specific criteria in Clauses (2) and (3). Clause (2) requires the DOE to remove highly radioactive 

radionuclides to the maximum extent practical. Clause (3) generally mirrors the criteria that the NRC has 

established for determining whether waste qualifies for land disposal as LLW (see 10 CFR 61.55 and 

61.58). This clause provides that disposal of the waste must meet the NRC performance objectives of 

10 CFR 61, Subpart C, and that the waste must not exceed the concentration levels for Class C waste in 

10 CFR 61.55 or the Secretary must consult with NRC concerning DOE’s disposal plans. 

Clause (1), noted above, is a broader criterion for the Secretary, in consultation with the NRC, to 

consider whether, notwithstanding that waste from reprocessing meets the other two criteria, there are 

other considerations that, in the Secretary’s judgment, require its disposal 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 has consulted 

with NRC concerning DOE’s disposal plans. As the NRC explained in In the Matter of Louisiana Energy 

Services, L.P. (National Enrichment Services) (CLI-05-05, 2005), the 10 CFR Part 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.” 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 stream does not “require disposal in a deep 

geologic repository” because non-repository disposal will be protective of public health and safety. 

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) is an acknowledgement 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 near- or intermediate-surface, nevertheless presents non-proliferation risks that the 

Secretary concludes cannot be adequately guarded against absent deep geologic disposal. Clause (1) gives 

the Secretary, in consultation with NRC, the authority to consider such factors in determining whether 

waste that meets the other two criteria needs disposal in a deep geologic repository in light of such 

considerations. 
46

That is not the case here. As is demonstrated later in this 3116 Basis Document, disposal of the 

stabilized residuals in the TFF, and the TFF tank system, at 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 risk to public health 

and safety. Furthermore, disposal of the stabilized residuals in the TFF, and the TFF tank system, does not 

raise any unique considerations that, notwithstanding these demonstrations, require its permanent 

isolation in a deep geologic repository. Accordingly, the stabilized residuals in the TFF, and the TFF tank 

system, meet the criterion of Clause (1) of Section 3116(a).ii

 

ii. On page 21 of its TER (NRC 2006b), the NRC also concluded that this criterion can be met. 
47

5. WASTE HAS HAD HIGHLY RADIOACTIVE RADIONUCLIDES 

REMOVED TO THE MAXIMUM EXTENT PRACTICAL 

Key Points 

• The list of highly radioactive radionuclides for the stabilized residuals in the TFF, and the TFF 

tank system, describes the radionuclides that could reasonably be expected to exist in the waste 

and that, using a risk-informed approach, contribute significantly to the radiological risk to 

workers, the public, and the environment, taking into account scientific and health physics 

principles, knowledge, and expertise. 

• In demonstrating that radionuclides will be removed to the maximum extent practical as 

required by Section 3116, the same degree of radionuclide removal is expected for the four 

300,000-gal tanks remaining to be cleaned as that achieved in the seven tanks that have been 

cleaned, as described in Subsection 5.2. 

• The demonstration that radionuclides will be removed to the maximum extent practical does not 

include the amount of additional residuals that may be removed from the cleaned 300,000-gal 

tanks during grouting operations. 

Section 3116(a) of the NDAA 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 Approach 

Based on consultations with 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. Table 5 lists the radionuclides for the grouted (stabilized) TFF 

residual waste and TFF tank system at closure and disposal that DOE has determined, on the basis of a 

risk-informed approach, contribute significantly to radiological risk to workers, the public, and the 

environment, taking into account scientific and health physics principles, knowledge, and expertise. 
48

The list of highly radioactive radionuclides was developed, beginning with the entire inventory of 

145 radionuclides stored at the TFF at INTEC (ICP 2005b). The DOE reviewed this inventory of 

radionuclides and identified those radionuclides in Tables 1 and 2 of 10 CFR 61.55,jj as well as any 

additional radionuclides that may be important to meeting the performance objectives in 10 CFR 61, 

Subpart C because they contribute to the dose to workers, the public, and/or the inadvertent intruder (for 

one or more reasonable intruder scenarios) in the expected and degraded cases, using sensitivity analyses. 

In DOE’s view, this approach results in a risk-informed list of highly radioactive radionuclides 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. This list includes the short-lived fission 

products 137Cs and 90Sr and their equilibrium daughters, 137mBa and 90Y, that are responsible for the bulk 

of activity in the TFF tank system inventory. Based on process knowledge, 137Cs and 90Sr and their 

equilibrium daughters, 137mBa and 90Y, account for approximately 99% of the activity in the TFF tank 

waste (ICP 2005b). This list also includes the long-lived actinides (241Am, 242Cm, 237Np, and the 

plutonium isotopes) that are responsible for approximately 0.08% of the activity. The other radionuclides 

included in the list are 14C, 3

H, 129I, 94Nb, 59Ni, 63Ni, and 99Tc, which contribute approximately 0.05% of 

the activity. The list of highly radioactive radionuclides (shown in Table 5) account for approximately 

99% (see Appendix A) of the radioactivity in the TFF tank waste as of April 2005, based on process and 

sampling knowledge. 

5.1.2 Radionuclides from Performance Assessment 

As explained above, DOE has included in the list of highly radioactive radionuclides (shown in 

Table 5) 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 TFF PA, which includes sensitivity and uncertainty analyses (DOE-ID 2003b). The DOE used the 

following approach in the PA to focus on those radionuclides that contribute to the dose for various 

pathways. The results of the PA show that 137mBa and 90Y should also be included in the list of highly 

radioactive radionuclides shown in Table 5. 

5.1.2.1 Groundwater Radionuclide Screening. Screening of radionuclides and numerical 

modeling and dose assessment was performed to form the list of highly radioactive radionuclides. The 

TFF inventory includes several radionuclides that are significant in terms of projected receptor doses from 

the groundwater pathway. Therefore, screening analysis methods were investigated for use in the TFF PA 

to focus the groundwater analysis on highly radioactive radionuclides. The groundwater screening 

analysis considered that active institutional control will be maintained over the disposal site for 100 years 

after facility closure and, furthermore, that the performance of the facility will be monitored to detect 

and/or prevent significant releases of radionuclides to the environment throughout the period of active 

institutional control. This consideration eliminates from concern any radionuclide with a half-life of less 

than 5 years because the inventory in the waste at 100 years after facility closure will be reduced to 

innocuous levels by radioactive decay. However, radionuclides with half-lives of less than 5 years cannot 

 

jj. Although Tables 1 and 2 in 10 CFR 61.55 specify concentration limits for certain radionuclides in the form of activated metal, 

DOE has included such radionuclides, if present in the waste, in the list of “highly radioactive radionuclides” as it exists in the 

waste, without regard to whether such radionuclides are in the form of activated metal. Consistent with Table 1, DOE has 

excluded alpha-emitting transuranic nuclides with half-lives of 5 years or less from the list of highly radioactive radionuclides. 

Some of the radionuclides listed as highly radioactive radionuclides in this 3116 Basis Document may not be listed in other 3116 

Determinations if such radionuclides are not present in the waste or do not contribute to dose to the worker, the public, or the 

inadvertent intruder. 
49

be neglected if the radionuclide appears in a decay chain. This is because its activity may increase with 

time due to decay of a parent radionuclide, unless the parent is also short-lived.

Table 5. List of highly radioactive radionuclides. 

Radionuclide 

Radionuclide 

Half-Life 

(yr) 

Long-Term 

Radiation Hazards 

Short-Term 

Radiation Hazards 

241Ama,b 4.30E+02 X 

14Ca,b 5.70E+03 X 

242Cmb,c 4.50E–01 X 

60Cod 5.30E+00 X 

137Csa,d 3.00E+01 X 

137mBaa 4.90E–06 X 

3

Ha,d 1.20E+01 X 

129I

a,b 1.60E+07 X 

94Nbb 2.00E+04 X 

59Nib 7.50E+04 X 

63Nib 1.00E+02 X 

237Npa,b 2.10E+06 X 

238Pua,b 8.80E+01 X 

239Pua,b 2.40E+04 X 

240Pua,b 6.60E+03 X 

241Pub,c 1.40E+01 X 

242Pub 3.80E+05 X 

90Sra,d 2.90E+01 X 

90Ya 7.30E–03 X 

99Tca,b 2.10E+05 X 

a. Highly radioactive radionuclides based on dose assessment results from the 2003 PA (DOE-ID 2003b). 

b. Taken from Table 1 of 10 CFR 61.55. 

c. Daughter product of long-lived radionuclide. 

d. Taken from Table 2 of 10 CFR 61.55. 

The use of the 5-year half-life screening criteria results in the elimination of the following 

radionuclides from further consideration: 102Rh, 119mSn, 134Cs, 150Eu, 153Gd, 155Eu, 55Fe, and 171Tm. In 

addition, the following radionuclides in short decay chains were eliminated from further analysis since the 

parent and progeny each have half-lives of less than 5 years: 106Ru/106Rh, 109Cd/109mAg, 110mAg/110Ag, 

125Sb/125mTe, and 144Ce/144mPr/144Pr. Additional radionuclides were screened from further consideration 

since their half-lives indicate that they are either stable or have such long half-lives that their specific 

activity, and thus contribution, to dose would be insignificant. These radionuclides include 142Ce 

(5.00E+16 yr), 149Sm (1.00E+15 yr), 144Nd (2.40E+15 yr), and 148Sm (2.00E+14 yr). 

The screening procedure used for the TFF inventory was based first on the concentrations of 

radionuclides in the waste pore water that would give an annual effective dose equivalent of 4 mrem/yr 

from consumption of 70 oz/d (i.e., 200 gal/yr) of contaminated water. The 4-mrem/yr standard was used 
50

for screening each nuclide because this portion of the screening process does not consider transport from 

the waste form to the groundwater. A total of 29 radionuclides were retained after screening using the 

pore water concentrations. 

The next step in the screening process involved evaluation of the release of radionuclides from the 

waste form and the resulting groundwater concentrations. Releases and groundwater concentrations were 

previously analyzed by numerical modeling. The results are presented in Appendix F of the PA (DOE-ID 

2003b). Based on these analyses, only 129I, 99Tc, and 90Sr were found to result in concentrations in 

groundwater that contribute significantly to doses (at least 99% of the all-pathways dose is attributable to 

these radionuclides). 

The HLW and Facilities Disposition Final EIS (DOE 2002) also conducted screening and 

groundwater transport analyses for the TFF. Their screening and analyses found that only 99Tc and 129I 

resulted in significant doses from the groundwater pathway. Therefore, based on previous modeling of the 

TFF presented in Appendix F of the PA, radionuclides (i.e., 129I, 99Tc and 90Sr) were determined to result 

in appreciable groundwater concentrations and doses. Therefore, the groundwater analysis was focused on 

the three radionuclides and they were included in the list of highly radioactive radionuclides. 

5.1.2.2 Intruder Pathway Radionuclide Screening. The intruder analysis considers that 

active institutional control will be maintained over the disposal site for at least 100 years after facility 

closure. This consideration eliminates from concern any radionuclide with a half-life of less than 5 years.

Using the 5-year half-life screening criteria results in the elimination of the following radionuclides 

from further consideration: 102Rh, 119mSn, 134Cs, 150Eu, 153Gd, 155Eu, 55Fe, and 171Tm. In addition, the 

following radionuclides in short decay chains were eliminated from further analysis since the parent and 

progeny each have half-lives of less than 5 years: 106Ru/106Rh, 109Cd/109mAg, 110mAg/110Ag, 125Sb/125mTe, 

and 144Ce/144mPr/144Pr. Additional radionuclides were screened from further consideration because their 

half-lives indicate that they are either stable or have such long half-lives that their specific activity and 

thus, contribution to dose, would be insignificant. These radionuclides include 142Ce (5.00E+16 yr), 

149Sm (1.00E+15 yr), 144Nd (2.40E+15 yr), and 148Sm (2.00E+14 yr). 

Dose assessment was then used to define the radionuclides that would be included in the intruder 

scenarios. The inclusion of radionuclides is explained in Section 5 of the PA (DOE-ID 2003b). A review 

of the PA shows the dose contribution by highly radioactive radionuclides was a factor of ten greater than 

the dose from the radionuclides excluded from further intruder scenario analysis. The air pathway 

analysis included 3

H and 14C because these radionuclides are volatile or gaseous. 

5.2 Removal to the Maximum Extent Practical 

Section 3116(a) of the NDAA provides that certain waste resulting from reprocessing is not HLW 

if the Secretary, in consultation with the NRC, determines, among other things, that the waste has had 

highly radioactive radionuclides removed “to the maximum extent practical.”kk This subsection discusses 

the basis for the Secretary to conclude that DOE’s approach for removing the highly radioactive 

radionuclides listed above meets this criterion. 

 

kk. Section 3116 does not specify “remedial goals” or other numerical objectives and does not require DOE to develop any such 

removal goals or objectives. Nevertheless, DOE notes, as discussed previously in this 3116 Basis Document, that the radioactive

waste removal expectations (which were used in the TFF PA [DOE-ID 2003b]) were exceeded for each of the tanks cleaned to 

date. 
51

Removal to the maximum extent “practical” is not removal to the extent “practical” or theoretically 

“possible.” Rather, a “practical” approach to removal is one that is “adapted to actual conditions” 

(Fowler 1930); “adapted or designed for actual use” (Random House 1997); “useful” (Random House 

1997); selected “mindful of the results, usefulness, advantages or disadvantages, etc., of [the] action or 

procedure” (Random House 1997); fitted to “the needs of a particular situation in a helpful way” 

(Cambridge 2004); “effective or suitable” (Cambridge 2004). Therefore, the determination as to whether 

a particular highly radioactive radionuclide will be removed to the maximum extent practical will vary 

from situation to situation, based not only on the available technologies but also on the overall costs and 

benefitsll of deploying a technology with respect to a particular waste stream. The “maximum extent 

practical” 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 radionuclide 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 the maximum extent practical 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. 

Although not an issue here, DOE notes that that it may not be practical to undertake any further 

removal of certain radionuclides because further removal is not sensible or useful in light of the overall 

benefit to human health or 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 contribution to 

potential doses to workers, the public, and the hypothetical human intruder.mm

Highly radioactive radionuclides have been removed from the TFF tank system to the maximum 

extent practical by removing liquid waste from the TFF tanks for calcining operations (described in 

Subsection 2.2) and by SBW removal and tank cleaning activities. The same degree of radionuclide 

removal is expected for the four 300,000-gal tanks and ancillary equipment remaining to be cleaned.nn To 

demonstrate compliance with Section 3116(a)(2), the list of highly radioactive radionuclides identified in 

Subsection 5.1, the total inventory of radionuclides in the TFF waste stream, and the residual inventory at 

closure and disposal are used to determine the effectiveness of radionuclide removal. These radionuclides 

in the list of highly radioactive radionuclides in Table 5 are listed or referred to in 10 CFR 61.55, “Waste 

 

ll. While prior NRC and DOE requirements for waste determinations called for removal “to the maximum extent technically and 

economically practical” (Bernero 1993; DOE M 435.1-1, 2001), 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. 

mm. The DOE normally would view radionuclides as making an insignificant contribution if the contribution to dose from those 

radionuclides, in both the expected case and fully degraded case using sensitivity analysis, 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. For perspective, DOE may 

also consider the extent to which such radionuclides contribute to an annual dose of 4 mrem to the receptor from the groundwater 

pathway. 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 (NRC 2003b), 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. 

nn. The same degree of removal is expected because the waste has not been and will not be radically altered. The particle size 

and density of the solids are not affected by tank-to-tank transfers during cleaning. Moreover, transfers between tanks have been 

a common practice in the past. 
52

Classification,” are significant contributors to the PA dose calculations and contribute to the radionuclide 

inventory (DOE-ID 2006a). Further removal of the residual inventory is not practical, as discussed below. 

The TFF tank system cleaning activities result in removal of radionuclides to the maximum extent 

practical. During tank cleaning activities for the 300,000-gal tanks, waste is removed from each tank in 

two phases, as described in Subsection 2.3.1.oo First, waste is removed using existing waste transfer 

equipment. Then, additional waste is removed using the tank cleaning system and modified waste transfer 

equipment. This second phase of waste removal continues until radiation levels near the steam-jet 

transport line decrease to the lowest constant value, indicating that the cleaning and waste transfer 

operations are no longer effective. Any remaining tank waste is regarded as waste residual. The inventory 

in the piping is insignificant (30 Ci) when compared to that of other system components 

(DOE-ID 2003b). The sandpad inventory is conservative, and no practical means exist to remove 

additional Ci from the sandpads that are located under the 300,000-gal tanks. A review of the tank 

construction and design, which includes access risers and the dimensions of the tank and sandpad, do not 

allow for removal of sand. The amount of radionuclides that have been flushed from the sand has not 

been measured directly and cannot be determined from the analysis of the liquid in the vaults because 

other sources (such as vault sumps) contribute to vault contamination. However, because of the numerous 

flushing events that have occurred, any additional cleaning by flushing would not provide a significant 

benefit. Therefore, tank cleaning is the only practical source for inventory reduction. Approximately 57% 

of the total system residual waste inventory at closure is expected to be contained in the fine layer of 

solids that is distributed unevenly over the bottom of the 300,000-gal tanks (see photographs in Figures 

11 and 12) (Portage 2005a, 2005b, 2005c, 2005d, 2005e, 2005f, 2006a). The remaining 43% is attributed 

to the conservative estimated inventory (as discussed below) for the contaminated sandpads (WM-185 

and WM-187). 

The tank cleaning process is thorough, and the process is reviewed during cleaning to maximize 

effectiveness. Videotapes are reviewed by observing benchmarks in the bottom of the tanks. The support 

brackets for cooling coils (0.97 cm [0.38 in.]) or the welds in the bottom (0.32 cm [0.13 in.]) of the tank 

are compared to the depth (if any) of solid residual. Operators and supervisors work together to review 

videos of cleaning and to view real-time cleaning on cameras to maximize performance of the operators 

during the cleaning cycle. Cleaning activities are monitored using cameras and video recording. In 

addition, radiation levels are monitored on the steam-jet transport line to ascertain cleaning effectiveness. 

Monitoring the radiation levels near the transport line provides the cleaning system operators an 

indication of when tank cleaning should cease. As cleaning operations progress, the level of radiation 

observed near the transport line decreases asymptotically over time to a minimum radiation level that 

cannot be significantly reduced further with continued cleaning operations. When radiation levels 

decrease to this lowest constant value, cleaning is stopped, and the tanks are visually examined using 

remote video inspection. Use of the radiation monitor allows tank cleaning to proceed without repeated 

visual inspection or sample collection and aids in ensuring that as much waste as practical is removed 

from the tanks (Kimmitt 2002). Personnel review videotape of cleaning activities and compare the visual 

results to the radiation monitoring results to optimize solid residual removal in each tank. Based on 

observations made during tank cleaning activities performed to date, tank agitation by the directional 

nozzles is most effective when directional nozzles are used with approximately 5,000–7,000 gal of water 

in the tank for a short time (100–140 minutes). This technique ensures that solid particles are suspended 

in the liquid for a period comparable to the time required to remove the liquid and suspended solids by 

pumping (EDF-TST-001, 2000). 

 

oo. As noted previously, removal of radionuclides occurs from a number of processes, including bulk removal of wastes and tank 

cleaning activities. 
53

The optimization techniques developed during tank cleaning increase radionuclide removal 

efficiency. Monitoring the process waste lines is proving to be an excellent tool to determine when 

cleaning techniques are reaching the maximum effectiveness as practical. To illustrate, Figure 13 shows 

the radioactivity levels during Tank WM-182 cleaning operations. The graph shows that significant 

radioactivity was detected in the outlet process piping when cleaning first began. Peaks in radioactivity 

removed indicate when the steam jets were removing residuals. The troughs indicate when the steam jets 

were not in use and washball or directional nozzles were being used to clean tank internal surfaces or 

suspend solids by agitation. Cleaning and pumping were often performed in unison. When the steam jet 

lost suction because liquid had been removed to a negligible level, pumping stopped until flush water 

from cleaning accumulated to a level that suction could resume. Near the end of the cleaning (after 

approximately 20,000 gal of flush water through directional nozzles), the radioactivity detected remained 

at or near the baseline level while cleaning and waste removal progressed. This condition was a clear 

indication that waste removal had been maximized. Subsequent video inspection and sampling and 

analysis confirmed the tank had been cleaned to the maximum extent practical. There is no evidence of 

any buildup of residual on the side walls after cleaning. Large areas of the tank floor are bare and some 

areas of the tank floor have a fine layer (approximately 0.97 cm [0.38 in.]). Additional spray cleaning was 

not able to remove this small quantity of residual. 

CPM/Gallon vs. Cumulative Gallons Pumped

(5)

-

5

10

15

20

25

30

35

40

- 20,000 40,000 60,000 80,000 100,000 120,000

Cumulative Gallons Pumped from WM-182

CPM/gallon

Figure 13. Radioactivity (counts per minute per gallon) measured during cleaning operations. 

As discussed in the following subsections, further removal of this residual waste is not practical 

and any efforts to remove more of the small quantity of waste remaining would not significantly reduce 

the potential risk to the public. The INL Site and DOE have successfully evaluated, selected, and 

deployed a tank cleaning system so that radionuclides are removed to the maximum extent practical. 

Various cleaning technologies were considered for removing residuals effectively from the TFF with the 

goal of protecting public and occupational health and safety at the tank closure site. Technologies were 

evaluated by comparing information on effectiveness, maturity, and cost balanced against minimizing 

occupational radiation dose. As discussed in Section 2, the TFF tank cleaning technology was selected in 

coordination with the DOE TFA. The DOE commissioned this team to coordinate technology needs for 

DOE complex-wide radioactive tank waste remediation problems. The team considered both chemical 

and mechanical processes for cleaning tanks. The INL Site determined that chemical processes were not 
54

practical because the TFF waste is acidic and adding other chemicals would be unlikely to increase waste 

removal efficiency. Some mechanical cleaning systems were found to be applicable, and high-pressure 

water systems (which slurry the solids) and mechanical arms would perform well in the TFF tanks. The 

washball and directional spray nozzle high-pressure water systems were determined practical for 

removing the small quantity of solids from the tank walls and for slurrying the solids on the tank bottoms. 

(See Appendix B for a discussion of the tank cleaning technology selection.) 

5.2.1 Percentage of Radionuclides Removed 

To quantitatively address how effective the waste removal techniques have been, a mass balance 

approach has been developed. Historically, radionuclides were removed from the total TFF waste stream 

by removing the liquid waste for the calcination process. The total inventory of radionuclides in the TFF 

waste stream has been identified and is provided in Generation, Disposition, and Current Inventory of 

Radionuclides in the INTEC Tank Farm (ICP 2005b). This inventory was identified by preparing a mass 

balance of radionuclides received from SNF reprocessing and sent to the TFF for storage and disposition. 

The mass balance relied on analytical data, numerical modeling, and data extrapolation to arrive at the 

total radioactivity present at INTEC from SNF reprocessing. The mass balance, which also considered 

nuclear material shipped to other DOE facilities for further processing, was used to determine the total 

radioactivity of waste generated from spent fuel reprocessing, which was 36 million Ci. 

Table 6 shows the initial inventory of all waste at INTEC (decayed to 2012) from reprocessing by 

radionuclide; the estimated residual waste in the 300,000-gal tanks, 30,000-gal tanks, the two 

contaminated sandpads, and piping after cleaning at closure; and the percentage of radioactivity removed 

by treatment (calcination and tank cleaning), shipment of nuclear material, and radioactive decay to the 

date of tank closure (2012). This percentage is determined by comparing the initial INTEC inventory of 

the radioactivity (Ci) that has been sent to the TFF with the estimated residual inventory at closure. As 

discussed in Section 2, the residual inventory at closure relies on analytical data from samples obtained 

during recent tank cleaning operations. Cleaning in the remaining 300,000-gal tanks is assumed to be as 

effective as in those that have been cleaned already for the reasons explained earlier in this 3116 Basis 

Document. The physical properties (INEEL 2000b) of the solid residual are well known. Solids were 

effectively cleaned from the tanks. There is not any information that indicates the potential for the tank 

cleaning system to be ineffective in the remaining tanks (Portage 2005a, 2005b, 2005c, 2005d, 2005e, 

2005f, 2006a, 2006b). More volume of water or a longer cleaning time may be necessary because of the 

amount of solids now stored in remaining tanks (with the exception of Tank WM-190). The residual 

inventory for the four tanks that have not been cleaned is represented by the estimated inventory in Tank 

WM-182 since it contains the highest residual inventory of any tank cleaned to date. Individual estimated 

tank inventories at closure are presented in Appendix A. Removal of residuals during tank cleaning 

activities has been demonstrated to be technically practical and efficient. The tank cleaning system has 

performed as designed and has achieved higher removal efficiencies than initially projected in the PA, as 

presented in Section A-2 of Appendix A. 
55

Table 6. Percentage of highly radioactive radionuclides removed from all tanks and ancillary equipment.

Radionuclides 

Total Ci Generated 

at INTEC

Residual Ci in 

Tanks at Closurea

Percent Removed at 

Closure 

241Amb

 9.28E+03 6.97E+00 99.92% 

137mBab

 8.95E+06 1.19E+04 99.87% 

14C 2.91E–02 3.85E–05 99.87% 

242Cm 1.51E+01 1.00E–02 99.93% 

60Co 1.67E+03 4.79E–01 99.97% 

137Csb

 9.46E+06 1.19E+04 99.87% 

129I

b 6.01E+00 5.87E–03 99.90% 

3

H 7.13E+03 5.43E+00 99.92% 

94Nb 1.54E+03 1.60E+00 99.90% 

59Ni 3.71E+03 1.90E–01 99.99% 

63Ni 4.36E+05 2.17E+01 99.99% 

237Npb

 7.53E+01 3.57E–01 99.53% 

238Pub

 1.07E+05 9.08E+01 99.92% 

239Pub

 2.83E+03 2.90E+01 98.98% 

240Pu 1.46E+03 1.09E+01 99.25% 

241Pu 4.73E+04 1.52E+02 99.68% 

242Pu 3.94E+00 7.60E–03 99.81% 

90Srb

 8.42E+06 6.78E+02 99.99% 

99Tcb

 3.67E+03 5.79E+00 99.84% 

90Yb

 8.42E+06 6.75E+02 99.99% 

Total (Ci)c 3.59E+07 2.58E+04 99.93% 

a. Total Ci at closure includes 137mBa and 90Y and radionuclide decay to 2012. Based on: (1) heel residuals that are estimated 

using remote video inspection of cleaned tank internals to map out estimates of depth of remaining residual solids and liquids 

across tank bottoms using tank internal reference points of known height, (2) best estimated radionuclide concentrations from 

past and recent samples as calculated in the associated engineering design files (Portage 2005a, 2005b, 2005c, 2005d, 2005e, 

2005f, 2006a, 2006b), and (3) radioactive decay to 2012. 

b. Radionuclides that are significant contributors to dose calculations in the 2003 TFF PA (DOE-ID 2003b). 

c. Radionuclides shown are contributors to the dose calculations or regulated by concentration limits in 10 CFR 61.55. The totals 

are based on the entire inventory of radionuclides.

Of the approximately 36 million Ci in 9.4 million gal of waste generated during spent fuel 

reprocessing operations, approximately 25,800 Ci are estimated to remain in the tank system following 

closure, representing about 0.07% of the initial spent fuel waste inventory. Approximately 92% of the 

total tank waste residual inventory at closure is represented by 137Cs/137mBa. The half-life of 137Cs is 30.17 

years. 

As discussed above, these values have been calculated by decaying the radionuclides to 2012, 

which is the projected end of closure activities. However, for perspective, if the inventory were to be 

calculated by decaying the radionuclides to the present, the total Ci remaining at closure would be 

somewhat higher, on the order of 30,000 Ci. The change is caused by 137Cs and 90S and their daughter 

products since these radionuclides account for the majority of the radioactivity, and they decay relatively 

quickly. Calculating these values assuming a decay date to the present results in a slightly higher estimate 

of total Ci remaining. However, if a table similar to Table 6 was constructed to show percent removed 
56

values for current decay dates, the percent removed values would remain the same, because the relatively 

short-lived radionuclides would also increase in the INTEC inventory at the same rate as the TFF 

inventory based on the half-life of each radionuclide. 

5.2.2 Cost of Developing a More Efficient Cleaning Technology 

In an effort to analyze the cost and benefits to remove the relatively small amount of residual, a 

cost estimate is provided below. While cleaning began in 2002, development of technology and preparing 

supporting analysis began earlier. In 1998, a report was prepared that estimated the costs of various 

options for closure of the TFF (INEEL 1998). The estimate to develop a tank cleaning system without 

deployment was $46M (not escalated). The estimate was based on a tank cleaning system similar to the 

system deployed at the TFF. This estimate included the following items and associated costs shown in 

Table 7. 

The expenditure for TFF closure activities from 1999 to 2005 is $35M. The $46M in expenditures 

to develop a new technology is presented for comparison to the $35M used in this analysis. Seven 

300,000-gal tanks and four 30,000-gal tanks have been cleaned from 2002 to 2005. 

Table 7. Cost estimate of TFF cleaning system (INEEL 1998). 

Activity 

Estimate Cost 

($M) 

Design of System 19.0 

Proof of Process 5.8 

Site Preparation 9.7 

Characterization of Waste 4.6 

Tank Isolation 7.0 

Total 46.1 

Development and deployment of a new technology is estimated to cost slightly more than 

development and deployment of the existing tank cleaning system. The estimated cost is projected by 

simply using an escalation rate of 10%. Therefore, the estimate for development and deployment of a new 

technology is $38.5M. The cost would likely be much higher because the tank cleaning system that has 

been deployed at the TFF has used much of the existing equipment in the TFF, including the transfer 

piping (underground shielded piping and valve boxes), the transfer system (steam jets and steam lines), 

tank access (used existing risers), monitoring system (level detection, alarms, and radiation measurement 

meters), and the vessel off-gas system. A new system may not use all of the existing equipment. 

Modification to the TFF, if made for a new technology, would increase costs to deploy the system. 

The access to tanks through existing risers is a limiting factor for deployment of a new technology. 

The risers are a maximum of 46 cm (18 in.) in diameter, with a maximum of five risers on any tank. 

Existing equipment currently in the risers decreases the available space in risers. The equipment includes 

instrumentation that cannot be readily removed. If new access must be designed that penetrates the soils, 

vault roof, and tank it would increase the cost considerably. Additional worker exposure from installation 

of new risers from tank contents and soils, plus contamination control, would add to costs. Additionally, 

excavation of soils was not necessary for the tanks already cleaned and it is not anticipated for the 

remaining four tanks. Weight restrictions on the surface of the TFF are also a concern. Multiple pieces of 

large equipment cannot be placed on the TFF if necessary to deploy a new cleaning technology. 
57

The cleaning technology for the TFF has been designed to use existing systems and commercially 

available systems or parts. During the review of technologies for cleaning, all known general categories 

of cleaning technologies were examined. Some types of specific technologies have been designed for 

tanks at other DOE facilities. Technologies such as a robotic crawler are not viable because of the cooling 

coils on the bottom of many of the TFF tanks. Therefore, a new technology or a refinement of an existing 

technology would be developed, designed, tested (found to be more effective), and deployed to the TFF. 

Deploying other tank cleaning technologies may delay closure of the TFF past 2012 with 

associated additional maintenance costs (this is not included in the estimated cost of $38.5M). A factor 

that also must be considered is that approximately 43% of the inventory in the TFF is not in the tanks but 

in the sandpads. Therefore, use of a new tank cleaning system can only potentially decrease the total TFF 

tank system inventory by approximately 57%. 

Table 8 shows example efficiencies of a hypothetical tank cleaning technology up to a maximum of 

60%. Also shown are the projected dose in mrem/yr of the all-pathways assessment, and the associated 

reduction in dose in mrem/yr. Table 8 also shows the cost per mrem of deploying a new technology.pp

Even using the most efficient system (which cleans the tanks completely), the dose reduction for a 50-

year period (50 years of exposure) is approximately 0.27 mrem/yr, and the cost per mrem reduction is 

approximately $2.8M. 

Table 8. All-pathways dose reduction and cost per mrem. 

Efficiency of Technology 

(%) 

20 40 50 60 

Dose (mrem/yr) 0.368 0.276 0.23 0.184 

Dose reduction (mrem/yr) 0.092 0.184 0.23 0.276 

Cost/mrem reduction $8,369,565.22 $4,184,782.61 $3,347,826.09 $2,789,855.07 

 

Note: See Table 20, which presents the scaled groundwater all-pathways dose (0.46 mrem/yr). 

The system design and testing that was used in this comparison was estimated at $38.5M. Based on 

other estimates (INEEL 1998), the cost could be approximately $50M if a 10% escalation was added. 

5.2.3 Conclusions 

Based on specific conditions, the INL Site determined that a limited number of technologies would 

effectively remove residuals from the tanks. The benefits of developing additional technologies, as well as 

complete tank removal, were evaluated for reducing residuals in the TFF. It is likely a system could be 

developed and deployed to remove the remaining residual by other remote means (INEEL 1998). 

However, initiating a long-term development project to develop other technologies would take many 

years to complete, and the project would be very costly (INEEL 1998). Although the cleaning to date has 

been successful, the small volume of residual certainly poses a direct radiation concern for anyone who 

would be in close proximity to the residual. No “hands-on” cleaning is practical and any new technology 

may require workers to contact the residual waste so worker exposure would increase while the projected 

 

pp. On pages 51–53 of its TER (NRC 2006b), NRC noted that there were uncertainties in the metric that DOE used to quantify 

the cost of additional radionuclide removal in the tanks. However, the NRC agreed with DOE that additional radionuclide 

removal appears to have minimal benefit. 
58

dose to the public from the residual waste, which is already low, would be lowered only slightly by using 

the new technology. In practice, during tank cleaning operations, even when the radiation monitor 

indicates that radioactivity levels are no longer decreasing and the videotape shows that the tank is clean, 

tank washing is continued for another day by flushing with several thousand gallons of water. Any 

appreciable decrease in the projected radiation dose to the public would be gained only if the TFF tank 

system is removed entirely. 

Tank cleaning worker exposure is estimated to total approximately 650 mrem per tank for 

23 workers (30 mrem per worker), which results in a total exposure of about 7.15 rem for eleven 

300,000-gal tanks (Appendix B-5; Jacobson 2002). Tank removal worker doses are significant because 

the action would require a large project to excavate the tanks, cut them up, and package them for disposal. 

Worker exposure is estimated to be 1,070 mrem/yr/worker for an average of 326 workers/yr for an 

estimated 26 years for a total exposure of over 9,000 rem (Appendix B-5; INEEL 1998). Therefore, this 

option would be several billion dollars (INEEL 1998), cause more exposure, and result in radioactive 

waste for which disposal would be difficult or not possible. 

As shown in Section 7, the remaining residuals, if left in place, pose a potential radiation dose to a 

member of the public on the order of 0.5 mrem/yr, and approximately half of this radiation dose is due to 

residuals in the tanks. Therefore, any new tank cleaning system could only achieve a reduction of 

approximately 0.25 mrem/yr, regardless of cost. With typical average doses to the public from natural 

sources and medical treatment in the range of 300–400 mrem/yr, it is not judged practical or cost effective 

to reduce the estimated dose from TFF by such a small amount. 

Thus, for the reasons discussed previously, DOE has determined that highly radioactive 

radionuclides have been or will be removed from the TFF tank system to the maximum extent practical.qq

 

qq. On page 55 of its TER (NRC 2006b) NRC concluded, “There is reasonable assurance that Criterion Two of the NDAA can be 

met because: 

• The estimated inventory developed for the tanks and sand pad in the performance assessment and validated through 

sampling for the tanks in the waste determination is reasonable for the purpose of evaluating compliance with NDAA 

criteria. 

• NRC staff has confidence that HRRs [highly radioactive radionuclides] have been or will be removed to the maximum 

extent practical based on the evaluation of DOE’s selection of HRRs; DOE’s selection, implementation, and effectiveness 

demonstration for its preferred cleaning technology; and NRC staff’s evaluation of the costs and benefits of additional 

removal.” 
59

6. RADIONUCLIDE CONCENTRATIONS OF STABILIZED WASTE 

Key Points 

• Concentrations of waste stabilized in grout were calculated for the closed tanks, vaults, and 

piping, and compared to the Class C concentration limits in 10 CFR 61.55.rr

• While DOE believes there is a reasonable basis to determine that none of the stabilized waste at 

the TFF will exceed the Class C concentration limits in 10 CFR 61.55, DOE, nevertheless, has 

consulted with NRC on its disposal plans in light of the calculated concentrations set forth in the 

draft 3116 Determination. 

Section 3116(a) of the NDAA 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 

(3)(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. 

Section 3116(a)(3) provides that, regardless of whether the waste exceeds or does not exceed the 

concentration limits for Class C LLW as set out in 10 CFR 61.55, the Secretary must determine that the 

waste will be disposed of in compliance with the performance objectives of 10 CFR 61, Subpart C and 

 

rr. The third criterion in Section 3116(a) references the Class C concentration limits in 10 CFR 61.55. Since the NRC regulations 

specifically allow averaging of the radionuclide inventory over the volume or weight of the waste [see 10 CFR 61.55(a)(8)], it is 

appropriate to consider average concentrations in the 3116 context. NRC agreed in its TER that concentration averaging is 

appropriate (see TER page 58) and this concept was also reflected in the State of Idaho comments. 
60

that the waste will be disposed of in accordance with State-approved closure plans. In Section 7 of this 

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 of this 3116 Basis 

Document, information is presented that demonstrates that the waste will be disposed of in accordance 

with State-approved closure plans. In situations where the waste is determined to exceed the 

concentration limits for Class C LLW, Section 3116(a)(3)(B)(iii) provides for consultation with NRC 

about the disposal plans for the waste. 

As discussed in this section, the DOE has determined that under DOE’s disposal strategy the 

stabilized residuals in the TFF, and the TFF tank system, will not exceed the concentration limits for 

Class C LLW. Nevertheless, DOE has also completed the consultation process established by Section 

3116(a)(3)(B)(iii) to take full advantage of the consultation process established by Section 3116. In this 

regard, DOE specifically requested in its draft 3116 Determination that NRC identify what changes, if 

any, NRC would recommend to DOE’s disposal plans (i.e., the actions DOE intends to undertake as 

described in the draft 3116 Determination and this 3116 Basis Document). In its TER, NRC explained 

that the extensive consultation process and review of DOE’s disposal plans satisfied Section 3116 

(a)(3)(B)(iii) and that consequently no additional NRC consultation is needed. Therefore, this 3116 Basis 

Document demonstrates NDAA criteria in Section 3116 (a)(3)(A) and (a)(3)(B) will be met for the 

stabilized residuals in the TFF, and the TFF tank system. NRC and the State of Idaho have also stated 

their views that DOE’s extensive consultation with NRC meets the additional requirements for waste that 

could be greater than Class C waste.ss 

Following submittal of the draft 3116 Determination to NRC in September 2005, the NRC released 

an FR notice, “Draft Interim Concentration Averaging Guidance for Waste Determinations” 

(70 FR 74846), for public comment. 

The draft interim guidance introduces a concept of ratios of unstabilized-to-stabilized waste. The 

factor of ten is an approximation derived from a consideration that most stabilization techniques 

commonly envision use of cementitious materials, and most cementitious waste forms can readily achieve 

a 10% waste loading. For illustrative purposes, an analysis was performed using a ten-to-one ratio of 

unstabilized-to-stabilized waste mass. Such an analysis results in each of the 300,000-gal tanks exceeding 

Class C concentration limits. However, the ten-to-one ratio is a general goal that should take into account 

other considerations. Moreover, DOE notes that NRC found that it would be reasonable for DOE to 

determine that the residual waste would not exceed Class C concentration limits on the basis of a number 

of factors, including consideration of a less conservative inventory and consideration of additional volume 

or mass from the tank floors and walls, interstitial liquids, and liquid wastes.tt For other components of the 

 

ss. On page 62 of its TER (NRC 2006b) NRC concluded, “Based on this analysis NRC staff concludes that it is reasonable for 

DOE to find that the residual waste in the tanks, sand pads, and piping that is the subject of this waste determination does not 

exceed Class C concentration limits. However, NRC cannot confirm that all large tanks contain waste that is within Class C 

limits due to the uncertainty in the residual waste inventory. Section 3116(a)(3)(B)(iii) requires DOE to consult with the NRC in 

its development of disposal plans in cases where the waste exceed Class C concentration limits. In the draft waste determination 

for the INL TFF, DOE stated its intent to ‘take full advantage of the consultation process established by Section 3116’ and 

requested NRC to ‘identify what changes, if any, it would recommend to DOE’s disposal plans’ as described in the draft waste 

determination. NRC has reviewed DOE’s disposal plans for TFF waste as part of the extensive consultation process that is 

documented in this TER, thereby satisfying the requirements of Section 3116(a)(3)(B)(iii). Consequently, no additional DOE 

consultation with the NRC is required for tanks containing residual waste that could exceed Class C concentrations. Idaho DEQ 

has also stated its view that DOE’s extensive consultation with NRC meets the additional requirements for waste that could be 

greater than Class C [Trever 2006].” 

tt. On page 59 of its TER (NRC 2006b) the NRC noted, “DOE Idaho did not exhaust all acceptable methods provided in the draft 

guidance for averaging the waste inventory over a large volume of grout...” Specifically, NRC noted that DOE Idaho did not 

take credit for the volume of liquid residuals that may remain in the tanks, for portions of the tank walls and floor, or for the 
61

TFF tank system, using examples or extrapolation of examples from the guidance yields concentrations 

that are within Class C concentration limits as discussed in Subsections 6.2 through 6.5. (See Section 

D-1.2.2 in Appendix D for further discussion. See Appendix G for a comparison of the TFF final waste 

form calculation methodology to fundamental principles presented in the NRC draft interim guidance on 

concentration averaging.) 

The rest of this section demonstrates that, by using assumptions and scenarios as appropriate and 

reasonable for the particular characteristics of the TFF, calculations of the sum of the fractions show that 

the TFF is within Class C concentration limits at closure and disposal.uu This demonstration takes into 

account consultations with NRC, including a revised evaluation of final waste form concentrations 

developed in response to an NRC RAI. 

Several sensitivity evaluations are included in Appendix D to assess the impacts of variability of 

various parameters on concentration results. The sensitivity analyses vary grout volumes, inventories, and 

averaging methodologies to provide additional insight into these calculations and support the conclusion 

of a Class C waste determination. 

6.1 300,000-gal Tanks 

In calculating the radionuclide concentration, 10 CFR 61.55(a)(8) states that the radioactivity in the 

waste may be divided by the volume of the final waste form. Therefore, to calculate a waste concentration 

in these tanks, the estimate of final residual inventory discussed in Section 2 is divided by the volume of 

the final waste form. Videos and photographs of the tank walls show staining and discoloration, and no 

discernible buildup of residuals. Therefore, no source term for the tank walls was included in the tank 

inventory discussed in Section 2. 

After cleaning is completed on each tank, the radionuclide inventory in the 300,000-gal tanks is in 

a dispersed layer across the tank bottom. The layer varies from 0.32 cm (0.13 in.) in thickness to near zero 

and is not concentrated in one specific location in the tank. The addition of grout for final closure is not 

expected to significantly mix with the residual waste or concentrate the radionuclide inventory. Mockup 

testing of grout placements (INEEL 1999b) in the tanks has shown that the resulting waste form will not 

be completely homogeneous; rather, the waste form will be a grouted monolith with some of the 

remaining residuals mixed in with the grout, some trapped between grout layers, and some encapsulated 

between the tank structure and the grout.vv 

 volume of interstitial liquids within the solid residuals. Although NRC noted uncertainties, NRC concluded that the waste from 

all of the 300,000-gal tanks could reasonably be found to not exceed Class C concentration limits (as noted on page 62 of the 

TER). 

uu. The nature of the acidic waste and stainless steel tanks are unique to the DOE complex. Because the waste remained acidic, 

significant amounts of stiff, recalcitrant sludge do not remain in cleaned tanks. Methods used in this document to evaluate waste 

concentrations are not likely to be used in the same manner by other DOE sites because the characteristics of the waste and tanks, 

and the general characteristics of individual DOE sites, are different, which may lead to a variety of approaches for determination 

of waste concentrations within the concepts of the NRC draft interim guidance. 

vv. Nevertheless, the resulting waste form may be considered “reasonably homogeneous,” as discussed in the Branch Technical 

Position on Concentration Averaging and Encapsulation (NRC 1995). The Branch Technical Position describes a reasonably 

homogeneous waste type as “one in which the radionuclide concentrations are likely to approach uniformity in the context of the

intruder scenarios.” The TFF PA intruder scenarios include activities such as excavation into TFF piping and drilling into grouted 

tanks. The remaining residuals in the closed tanks would not be concentrated in any specific location in the tank. If such activities 

were to occur, they would, by their nature, result in a significant amount of additional mixing of grout, surrounding soils, and a 

relatively small volume of residuals. Therefore, the resulting inventory and dose would be similar regardless of where the drilling 

or excavations occurred within a specific tank. In this context, the TFF tanks will meet this definition of reasonably homogeneous 
62

As described below, the engineered grout placement sequence to move remaining solids and 

liquids toward the removal pumps is estimated to use 85 m3

 of grout. The function of the engineered grout 

placements is to provide additional assurance that the waste is being removed. The engineered grout 

placements also aid in the mixing of residuals with the grout. The engineered grout placements also 

provide a reducing environment. A final encapsulation pour of 140 m3

 of grout is planned to level the 

engineering grout placement. A portion of the 140 m3 (33 m3) is used to stabilize the residual waste on the 

grout surface. The total volume of grout for this operation is 225 m3

, which results in a layer of grout 

about 1.2 m (4 ft) thick from the tank floor; however, the total volume used for calculating whether the 

waste will exceed Class C concentration limits through concentration averaging is 118 m3

 (see discussion 

in Section E-2 in Appendix E).ww The engineered grout placement sequence is described below. 

The grout will be introduced through two available risers with specially designed grout masts. The 

first two placements go in directly below the available risers to a height of 0.9–1.2 m (3–4 ft). The 

purpose of the first two placements is to begin moving residuals toward the steam jet for removal and to 

provide troughs to direct placements 3 through 5 to the other areas of the tank. In the grout mockup 

(INEEL 1999b), these placements were successful in moving solid and liquid surrogate materials. 

Placements 3 and 4 use the same riser access as placements 1 and 2, and displace the residuals 

between the tank wall and the steam jet. The purpose of placement 5 is to displace the residuals on the 

opposite side of the tank from the steam jet. These placements flow through the trough to sweep residuals 

toward the steam jet. Placement 5 may be replaced with two separate placements to allow better residual 

removal, but the purpose of the placement is the same. Figure 14 shows a schematic of the grout 

placements. 

Based upon mockup results, some portion of remaining residuals is likely to be pushed to the steam 

jet and removed during grouting. Some of the residuals will likely remain trapped between the tank 

surfaces and the grout, some will be mixed within the grout, and some will remain on top of the 

engineered grout placements. A final encapsulation pour will then be used to ensure adequate 

immobilization of any remaining residuals. 

To calculate a waste concentration in these tanks, the estimate of final residual inventory at closure 

is divided by the volume or mass of the final waste form.xx 

 when the residuals are stabilized with grout and the radionuclide concentrations of such a mixture would likely approach 

“uniformity.” 

ww. On page 59 of the TER (NRC 2006b) the NRC stated, “...the estimated 85 m3

 of grout for the engineered pour is just 

slightly higher than the volume of grout needed to meet Class C concentrations...” and that the NRC staff was “...concerned that 

DOE Idaho used an approach that can support the use of an excessive amount of grout for the purpose of waste classification 

without sufficient justification.” As discussed in detail in Appendix E of this document, the volume of grout necessary for the

engineered pour was determined by extensive mockup demonstrations in 1999, prior to development of final tank inventory 

estimates in 2005 or possible final waste form concentrations. The quantity of grout was not designed to meet Class C 

concentrations, but to provide for effective tank closures. 

xx. This approach is consistent with 10 CFR 61.55(a)(8), which states that the radioactivity in the waste may be divided by the

volume or mass of the final waste form, and was not developed to justify Class C waste concentrations. 
63

Figure 14. Schematic of the grout placement sequence. 

6.1.1 300,000-gal Tank Waste Concentration 

To determine the radionuclide concentration of the final waste form in the 300,000-gal tanks, a 

residual inventory of 2,394 Ci and a grout volume of 118 m3

 were used as described in Appendix E. 

Tables 9 and 10 show the calculations of the residual inventory at closure for Tank WM-182 averaged 

over 118 m3 of grout.yy The mass of grout is equal to 2.48E+08 g (density of 2.1 g/cc). The Class C 

concentration limits for long- and short-lived radionuclides (Tables 1 and 2 of 10 CFR 61.55) are shown 

in Tables 9 and 10. To ensure conservatism in the analysis, the mass of tank steel is not used in these 

calculations.zz The sum of the fractions is shown for the Tank WM-182 grouted waste form. The other 

300,000-gal tanks that have been cleaned have a lower residual waste inventory at closure than Tank 

WM-182 and are, therefore, bounded by these calculations. The residual inventory at closure for the 

300,000-gal tanks does not take credit for any additional residuals that may be removed during grouting 

operations. 

Long-lived radionuclides shown in Table 1 of 10 CFR 61.55 are the significant concentration limits 

for the TFF tanks (as described in Sections 3–5) and ancillary equipment. Concentrations of 238Pu and 

 

yy. On page 61 of its TER (NRC 2006b), the NRC noted that DOE did not list 244Cm in the tables of this section, although other 

alpha-emitting TRU isotopes with half-lives greater than 5 years are addressed. DOE did not list 244Cm since this isotope was not 

detected during analysis of the cleaned tanks and is not a significant contributor to risk (as NRC acknowledged on page 39 of its 

TER). 

zz. For perspective, the results of calculations that include the mass of the tank steel are shown in Table D-3 (Appendix D) as part 

of a sensitivity evaluation. 
64

239Pu contribute significantly to the sum of the fractions. A review of Table 10 shows that short-lived 

radionuclides do not affect the analysis for Class C waste concentration limits as the sum of the fractions 

is two orders of magnitude lower than the sum of the fractions for long-lived radionuclides. 

The sum of the fractions for short-lived radionuclides (Table 2 of 10 CFR 61.55) in the vault is also 

two orders of magnitude lower than the sum of the fractions for long-lived radionuclides and the sum of 

the fractions for 30,000-gal tanks and piping are three to four orders of magnitude lower. Tables for 

short-lived radionuclides are shown in Appendix F. 

Table 9. Radionuclide concentrations in the final Tank WM-182 grouted waste form (Table 1 of 

10 CFR 61.55). 

Tank Inventory 

Class C 

Concentration Limit 

Radionuclidea

Half-Life 

(yr) (Ci)b

 Ci/m3 nCi/g (Ci/m3 or nCi/g) 

Fraction of 

Class C 

Concentration 

Limit 

241Am 4.30E+02 4.20E–01 1.70E+00 100 0.0170 

14C 5.70E+03 5.00E–06 4.20E–08 8 0.0000000053 

242Cm 4.50E–01 1.30E–03 5.30E-03 20,000 0.00000027 129I 1.60E+07 7.70E–04 6.60E–06 0.08 0.000082 59Ni 7.50E+04 2.50E–02 2.10E–04 220 0.00000097 

94Nb 2.00E+04 2.10E–01 1.70E–03 0.2 0.0087 237Np 2.10E+06 4.70E–02 1.90E–01 100 0.00190 238Pu 8.80E+01 1.10E+01 4.60E+01 100 0.46 

239Pu 2.40E+04 3.40E+00 1.40E+01 100 0.137 240Pu 7.00E+03 1.40E+00 5.50E+00 100 0.055 241Pu 1.40E+01 1.90E+01 7.90E+01 3,500 0.022 

242Pu 3.80E+05 9.90E–04 4.00E–03 100 0.000040 99Tc 2.10E+05 7.60E–01 6.50E–03 3 0.0022 

Sum of the Fractions 0.71 

a. Radionuclides shown in italics are compared to Class C concentration limits in units of Ci/m3 ; remaining nuclides are compared to limits in 

units of nCi/g. 

b. Radioactive decay to 2012; the sum of the fractions will not significantly change if decayed to the present.

Table 10. Radionuclide concentrations in the final Tank WM-182 grouted waste form (Table 2 of 

10 CFR 61.55). 

Half-Life Tank Inventory Class C Concentration Limit 

Radionuclide (yr) (Ci)a

 Ci/m3 (Ci/m3) 

Fraction of Class C 

Concentration Limit

137Cs 3.00E+01 1.10E+03 9.70E+00 4,600 0.0021 

63Ni 1.00E+02 2.90E+00 2.40E–02 700 0.000035 

90Sr 2.90E+01 2.30E+01 2.00E–01 7,000 0.000028 

Sum of the Fractions 0.0022 

 

a. Radioactive decay to 2012; the sum of the fractions will not significantly change if decayed to the present.

The sum of the fractions from Tables 9 and 10 is 0.71 and 0.0022, respectively. Based on the 

results of these tables and a review of the data and assumptions, the residuals in the 300,000-gal tanks 

would not exceed Class C concentration limits. 
65

6.1.2 Summary and Conclusions 

At closure and disposal, the stabilized residuals in the 300,000-gal tanks would not exceed Class C 

concentration limits. The basis for this is the completion of waste concentration tables as described in 

10 CFR 61.55. 

Additional sensitivity evaluations and operational constraints described in Appendix D provide 

reasonable scenarios for waste concentration calculations. For perspective, an analysis was also 

performed using a ten-to-one ratio of unstabilized-to-stabilized waste mass, which showed that each of 

the 300,000-gal tanks exceeds Class C concentration limits. However, as discussed in Section D-2.1.2 of 

Appendix D, such an approach is not appropriate or reasonable for calculating final waste concentrations 

at the TFF. The scenarios that have been developed, as appropriate and reasonable for the particular 

characteristics of the TFF tank system, all have a sum of the fractions of less than 1. 

6.2 300,000-gal Tank Vaults 

The tank secondary containment system consists of the tank vaults and sandpads, and is discussed 

together in this subsection. An inventory has not been developed for the tank vaults because solids are not 

present in the vaults and liquid is confined to the volume of the sump. Generally, the sumps are 0.3- by 

0.3-m (1- by 1-ft) concrete enclosures in the vault bottom. Some of the sumps are slightly larger. Any 

residual inventory at closure for the vaults is considered negligible when compared to the sandpad 

inventory. 

For the vaults around the 300,000-gal tanks, contamination of the vault surfaces is from liquid 

releases from valves boxes or directly from the tanks. As discussed in Section 2, the tanks have never 

leaked into the vaults. Waste entered the vaults directly from the tanks only twice from Tanks WM-185 

and WM-187 during back-siphoning events. Typically, vaults have contained contaminants from only two 

sources, secondary containment of process piping and valve boxes. Process piping and valve leaks are 

infrequent occurrences. In the event of a leak, the system of liquid level indicators and radiation detection 

instruments would alert operators of a leak and the contamination would be removed well before the 

volume of the vault sump (volume is approximately 0.3 m3

) was exceeded. Therefore, such sources 

contributed only minor amounts of contamination. Additionally, vaults collected water from precipitation 

events or snowmelt. This water entered the vault from the vault top, which is made of pre-stressed 

concrete slabs that were never designed to prevent water inflow. Vaults associated with the cleaned tanks 

have been flushed and sampled. Solids were not expected in the tank vaults and solid samples were not 

collected. Solids were not expected because waste was not contained in any vault for a long period of 

time. Vaults for Tanks WM-185 and WM-187 did hold several thousand gallons of waste from a 

back-siphoning event in each vault in 1962. The waste was removed from each vault within 24 hours after 

each event began. 

During recent cleaning operations, the tank vaults were flushed with water during the cleaning 

process, but access to the vaults did not allow for the aggressive cleaning and agitation that was 

performed in the tanks. Sampling from the vault sumps may have contributed to elevated radionuclide 

concentrations in liquid samples, as discussed below. The vault sumps are the low points in the vaults 

designed to capture leaks from valve boxes and to allow removal of infiltration from precipitation and 

waste as necessary. The vault sumps were never emptied completely because the steam jets are 

approximately 5 cm (2 in.) from the sump floor. Therefore, residual contamination in the sump may have 

biased the sample concentrations. No solids have been observed in the tank vaults other than debris. 

Liquids do not collect in the vault because the floors are sloped toward vault sumps. After cleaning, the 

water used to flush the vaults was pumped out using the steam jets in vault sumps. The sumps are 

generally 0.3 m (1 ft) square (some vaults have larger sumps). Review of the sampling data shows that the 
66

concentrations of radionuclides in liquid vault samples do not exceed the liquid concentration of the 

cleaned tanks. 

A realistically conservative engineering analysis is used to estimate the contamination remaining in 

the sandpads because sampling the sandpads is not technically feasible and no sandpad sampling data 

exist. Since the sandpads are located directly beneath the tanks, any useful sampling would require a task 

equal to complete tank removal given the limited access.aaa The engineering analysis indicates 

radionuclides from the early contamination events remain absorbed to the sand despite multiple washings 

(DOE-ID 2003b). Cleaning the sandpad is not practical given the limited access and location of this sand. 

The commercial-grade sand in the sandpads is in solid physical form and will be surrounded by the vault, 

the grout, and the tank (filled with grout).bbb 

The vault will be filled completely with grout. Incremental steps in the grout fill will accomplish 

covering the sandpad by making grout pours through the two available risers, filling to level the initial 

grout pour, and finally filling the entire vault. A minimum grout height of 1 m (3 ft) is necessary to allow 

the grout to flow around the vault and rise to a level of at least 46 cm (18 in.) above the grout floor. The 

46-cm (18-in.) height is desirable because the sandpad and associated dike are 15 cm (6 in.) above the 

floor and approximately 46 cm (18 in.) level, which will ensure encapsulation of the sandpad. Grout 

poured into the vault will fill the area surrounding the sandpads, leaving no voids and providing structural 

integrity to the sandpads. Liquid entering the vault system will be drained from the vault sump prior to 

closure. Grouting of the vault area surrounding the sandpad will incorporate any remaining liquids. The 

inventories of the vaults are insignificant when compared to the sandpad because very little liquid remains 

in the vault sumps and no solid residual is present. 

The residual inventory at closure for the sandpads was presented in Section 2. The calculations in 

Tables 11 and 12 are based on a total sandpad volume of 23.39 m3

 (Staiger 1999) and the mass of the 

sandpad and vault grout (65 m3

 or 1 m [3 ft] high in a typical vault) of 136,500 g. Tables 11 and 12 show 

the radionuclide concentrations in the grouted vaults. For reference, the Class C concentration limits for 

long- and short-lived radionuclides (Tables 1 and 2 of 10 CFR 61.55) are also shown. 

The radionuclide 59Ni is not included in the sandpad inventory. The initial amount of each 

radionuclide in the tanks that was back-siphoned to the vault and contacted the sandpad at the time of the 

accidental spills was evaluated using limited analytical data and ORIGEN2. The ORIGEN2 data were 

corrected to the 137Cs concentration in Tank WM-185 1 month before the incident. Tank WM-185 was 

used because of its slightly higher 137Cs concentration than in Tank WM-187. However, this ORIGEN2 

modeling did not include the activation product 59Ni. The concentration of 59Ni in the sandpads will not 

affect the comparisons made in Tables 11 and 12 because 59Ni is typically found in very low 

concentrations as evidenced by the concentrations in the tanks and piping. 

 

aaa. In DOE’s response to NRC’s RAI Comment 1 (ICP 2006l), the difficulty in direct sampling of the sandpads was discussed. 

On page 36 of its TER (NRC 2006b), NRC concluded, “...DOE Idaho made a good faith effort to evaluate the uncertainty in the 

sand pad inventory through additional sensitivity and uncertainty analysis, modeling, and analysis of indirect sampling data.” 

bbb. The sandpads are part of the tank ancillary equipment and are not considered to be contaminated media outside of the tank 

system. 
67

Table 11. Radionuclide concentrations in the grouted vaults (Table 1 of 10 CFR 61.55). 

Radionuclidea

Half-Life 

(yr) 

Sandpad 

Inventory 

(Ci)b

Sandpad 

Inventory 

in Ci/m3

Sandpad 

Inventory 

in nCi/g 

Class C 

Concentration 

Limit 

(Ci/m3 or nCi/g)c

Fraction of Class C 

Concentration 

Limit 

241Am 4.30E+02 1.90E+00 1.40E+01 100 0.14 

14C 5.70E+03 3.90E–07 6.00E–09 8 0.00000000075 242Cm 4.50E–01 1.40E–05 1.00E–04 20,000 0.0000000051 

129I 1.60E+07 1.10E–06 1.70E–08 0.08 0.00000021 

94Nb 2.00E+04 2.30E–02 3.50E–04 0.2 0.0018 237Np 2.10E+06 3.70E–04 2.70E–03 100 0.000027 238Pu 8.80E+01 2.10E+00 1.50E+01 100 0.15 

239Pu 2.40E+04 1.60E+00 1.20E+01 100 0.12 240Pu 7.00E+03 3.50E–01 2.60E+00 100 0.026 241Pu 1.40E+01 2.30E+00 1.70E+01 3,500 0.0048 

242Pu 3.80E+05 5.70E–05 4.20E–04 100 0.0000042 

99Tc 2.10E+05 2.00E–12 2.10E–15 3 0.00000000000000071 

Sum of the Fractions 0.44 

 

a. Radionuclides shown in bold italics are concentration limits in units of Ci/m3; remaining nuclides are concentration limits in units of nCi/g. 

b. Radioactive decay to 2012; the sum of the fractions will not significantly change if decayed to the present. 

c. Table 1 of 10 CFR 61.55.

Table 12. Radionuclide concentrations in the (Table 2 of 10 CFR 61.55) grouted vaults (DOE-ID 2003b). 

Radionuclide 

Half-Life 

(yr) 

Sandpad Inventory 

(Ci)a

Sandpad 

Inventory in 

Ci/m3

Class C 

Concentration Limit 

(Ci/m3)

b

Fraction of 

Class C 

Concentration 

Limit 

137Cs 3.00E+01 1.60E+03 2.50E+01 4,600 0.00543 

63Ni 1.00E+02 1.70E–10 2.60E–12 700 3.71E–15 

90Sr 2.90E+01 2.50E+02 3.80E+00 7,000 0.000543 

Sum of the Fractions 0.00597

 

a. Radioactive decay to 2012; the sum of the fractions will not significantly change if decayed to the present. 

b. Table 2 of 10 CFR 61.55.

Further uncertainty analysis of the sandpad inventory is provided in DOE’s responses to NRC RAI 

Comments 1, 3, and 4 (ICP 2006l). The sandpad and small amount of residuals in the vault will be 

stabilized using grout, which can be introduced into the vault in only two places, the risers over the north 

and south vault sumps. Sufficient grout must be added to fill the vault floor and contain the sandpad.ccc

 

ccc. The data or assumptions used in the calculations include the following: 

1. The volume of grout equals 32.5 m3

 (INEEL 2000a). 

2. The volume of the sandpad equals 23.6 m3

 (DOE-ID 2003b). 

3. The mass of grout equals 6.83E+07 g (DOE-ID 2003b). 
68

The calculations for the sum of the fractions shown above assumed a 1-m grout pour and an 

inventory of 3,850 Ci. The ratio of long- to short-lived radionuclides in the sandpad inventory is different 

than in the tank inventory. That scenario assumed the vault has been leveled to a 1-m (3-ft) depth. For 

purposes of vault waste concentration calculations, an alternative is to use a 1-m (3-ft) deep pour to 

ensure the grout has been able to flow around the vault and achieve a minimum depth of 0.5 m (1.5 ft) at 

the opposite side of the vault. The volume of grout in this scenario has been calculated to be 32.5 m3

. 

The sum of the fractions for this scenario is 0.51 as shown in Table 13. This alternative scenario 

reduces the amount of grout used in encapsulation and provides an alternative to a 1-m (3-ft) deep grout 

pour in the vault. This alternative reflects the fact that grout is not self-leveling in the tank vaults, which 

have a circumference of 48 m (157 ft). Additional alternatives to describe the performance of grout in 

smaller quantities are not shown because 32.5 m3

 is the lowest volume of grout that can be used to ensure 

the sandpad is encapsulated. Because only two risers are available and the access through these risers is 

limited, video confirmation of sandpad encapsulation is not possible. Calculated volumes of grout with 

expected flow characteristics are used rather than video confirmation. 

Table 13. Sum of the fractions for 32.5 m3

 of grout in the tank vault (Table 1 of 10 CFR 61.55). 

Conservative Tank 

Inventory 

Class C 

Concentration Limit 

Radionuclidea

Half-Life 

(yr) (Ci)b

 Ci/m3 nCi/g (Ci/m3 or nCi/g) 

Fraction of Class C 

Concentration Limit 

241Am 4.30E+02 1.90E+00 1.60E+01 1.00E+02 1.60E–01 

14C 5.70E+03 3.90E–07 6.96E–09 8.00E+00 8.70E–10 242Cm 4.50E–01 1.40E–05 1.17E–04 2.00E+04 5.87E–09 129I 1.60E+07 1.10E–06 1.93E–08 8.00E–02 2.41E–07 

94Nb 2.00E+04 2.30E–02 4.09E–04 2.00E–01 2.04E–03 237Np 2.10E+06 3.70E–04 3.15E–03 1.00E+02 3.15E–05 238Pu 8.80E+01 2.10E+00 1.75E+01 1.00E+02 1.75E–01 

239Pu 2.40E+04 1.60E+00 1.33E+01 1.00E+02 1.33E–01 

240Pu 7.00E+03 3.50E–01 3.00E+00 1.00E+02 3.00E–02 

241Pu 1.40E+01 2.30E+00 1.93E+01 3.50E+03 5.53E–03 

242Pu 3.80E+05 5.70E–05 4.82E–04 1.00E+02 4.82E–06 99Tc 2.10E+05 2.00E–12 3.60E–14 3.00E+00 1.20E–14 

Sum of the Fractions 0.51 

 

a. Radionuclides shown in italics are compared to Class C concentration limits in units of Ci/m3 ; remaining nuclides are compared to limits in 

units of nCi/g. 

b. Radioactive decay to 2012; the sum of the fractions will not significantly change if decayed to the present.

 

4. The mass of the sandpad equals 4.14E+07 g (DOE-ID 2003b).

5. The inventory was estimated using the ORIGEN2 numerical model. A total of 3,850 Ci is in the WM-185 sandpad 

inventory (DOE-ID 2003b). 

6. The 137Cs data collected from Tank WM-185 just prior to the back-siphoning events were used in the ORIGEN2/Wenzel 

ratios. A 137Cs concentration of 1.7 Ci/L was measured in 1962 (Latchum et al. 1962). 

7. A grout with a density of 2.1 g/cc was used in the calculation (INEEL 2000a). 
69

6.2.1 Summary and Conclusions 

At closure and disposal, the 300,000-gal tank vaults would not exceed Class C concentration limits. 

The basis for this is the completion of the waste concentration tables as described in 10 CFR 61.55. 

Encapsulation of the sandpad material and any other vault contamination can be performed only by 

making grout pours in two available risers. The access and operational constraints allow for grout to be 

introduced, mound in the location directly under the access risers, and flow in the annular space until a 

level of 0.3–0.46 m (0.9–1.5 ft) is reached in all locations of the vault. This ensures encapsulation of the 

sandpad contamination. This volume of grout is necessary because video inspection to ensure the grout 

has reached a minimum depth is not practical because of operational and physical limitations. A 

calculated volume must be used to ensure the sandpad is encapsulated. Additional grout (not used in the 

sum of the fractions calculations) will be used to continue to fill the vaults until the level is approximately 

1 m deep, and ultimately, completely filled. 

6.3 30,000-gal Tanks 

The residuals remaining in each 30,000-gal tank contains approximately 36 Ci, which is 

approximately 1.5% of the residual inventory in a typical 300,000-gal tank. As discussed in Section 2, the 

residual inventory at closure for the 30,000-gal tanks assumes that each tank has a 5-mil (0.005-in.) thick 

film on the lower half of the tanks. The 5-mil (0.005-in.) thickness was used as a conservative estimate 

since solid samples were not collected because of a lack of material to sample. The tanks had not 

contained acidic waste for at least 20 years, and the heat from the steam valve condensate allowed the 

development of what appeared to be a biological film on the tanks. However, this film could not be 

sampled to determine the radionuclide concentrations. Therefore, conservative thickness and radionuclide 

concentrations assumptions are used. For the radionuclide concentrations in this film layer, the analytical 

results from the solid samples from Tank WM-183 in Ci/kg are applied to this mass of solid material. The 

liquid residuals were sampled and analyzed as discussed in Section 2. Liquid sampling results averaged 

are approximately 0.23 Ci in liquids with 36.1 Ci of solid residual. 

During grouting operations, the tanks will be filled to the midway point initially because the thin 

film observed in the tanks does not extend beyond the bottom half of the tank. This grout pour will be 

performed to contain the residual on the tank wall. These tanks will be filled completely with grout prior 

to closure. Since the bottom area of a horizontal, cylindrical tank is relatively small compared to a 

flat-bottomed tank, and minimal residuals remain in the tank bottom, the multi-point grout placement 

technique planned for the 300,000-gal tanks will not be necessary for these tanks. 

The volume of grout needed to fill the tanks to the half of the volume is 58 m3

 with a density of 

2.1 g/cm3

. Since the tank is considered to be part of the final waste form, the mass of the stainless steel 

tank is included in the concentration averaging calculation. Only stainless steel mass that encloses the 

engineering grout pour (half of the tank) is applied to the calculation. Tables 14 and 15 show the 

calculations of the residual inventory at closure for a 30,000-gal tank averaged over 58 m3

 of grout (with 

a density of 2.1 g/cm3

). For reference, the Class C concentration limits for long- and short-lived 

radionuclides (Tables 1 and 2 of 10 CFR 61.55, respectively) are also shown. 
70

Table 14. Radionuclide concentrations in a final 30,000-gal tank grouted waste form (Table 1 of 

10 CFR 61.55). 

Radionuclidea

Half-Life 

(yr) 

Average 

30,000-gal 

Tank 

Inventory (Ci)b

Average 

30,000-gal 

Tank 

Inventory in 

Ci/m3

Average 

30,000-gal 

Tank 

Inventory in 

nCi/g 

Class C 

Concentration 

Limit 

(Ci/m3 or 

nCi/g)c

Fraction of 

Class C 

Concentration

Limit 

241Am 4.30E+02 6.40E–03 4.90E–02 100 0.00049 

14C 5.70E+03 1.10E–07 1.90E–09 8 0.00000000024 242Cm 4.50E–01 2.00E–05 1.50E–04 20,000 0.0000000077 

129I 1.60E+07 1.20E–05 2.00E–07 0.08 0.0000025 94Nb 2.00E+04 3.10E–03 5.30E–05 0.2 0.00027 

59Ni 7.50E+04 3.80E–04 6.50E–06 220 0.000000030 237Np 2.10E+06 7.10E–04 5.50E–03 100 0.000055 

238Pu 8.80E+01 1.70E–01 1.30E+00 100 0.013 

239Pu 2.40E+04 5.10E–02 4.00E–01 100 0.0040 240Pu 7.00E+03 2.00E–02 1.60E–01 100 0.0016 

241Pu 1.40E+01 2.90E–01 2.30E+00 3,500 0.00065 

242Pu 3.80E+05 1.50E–05 1.10E–04 100 0.0000011 

99Tc 2.10E+05 1.20E–02 2.00E–04 3 0.000066 

Sum of the Fractions 0.020 

a. Radionuclides in bold italics are concentration limits in units of Ci/m3; remaining nuclides are concentration limits in units of nCi/g. 

b. Radioactive decay to 2012; the sum of the fractions will not significantly change if decayed to the present. 

c. Table 1 of 10 CFR 61.55.

Table 15. Radionuclide concentrations in a final 30,000-gal tank grouted waste form (Table 2 of 

10 CFR 61.55). 

Radionuclidea

Half-Life 

(yr) 

Average 30,000-gal 

Tank Inventory 

(Ci)a

Average 

30,000-gal 

Tank Inventory 

in Ci/m3

Class C 

Concentration 

Limit 

(Ci/m3)

b

Fraction of 

Class C 

Concentration 

Limit 

137Cs 3.00E+01 1.70E+01c

 3.00E–01 4,600 0.00006 

63Ni 1.00E+02 4.30E–02 7.40E–04 700 0.0000011 

90Sr 2.90E+01 4.50E–01c

 7.80E–03 7,000 0.0000009 

Sum of the Fractions 0.00007 

a. Radioactive decay to 2012; the sum of the fractions will not significantly change if decayed to the present. 

b. Table 2 of 10 CFR 61.55. 

c. Radioactivity for 137Cs (137Ba) and 90Sr (90Y) daughters are not shown in this table.

An alternative to this is evaluated that would simply use the mass of steel for half of the tank. Since 

the residual contamination is almost entirely associated with the film on the lower half of the tank wall, 

the mass of steel for half of the tank is the appropriate mass and volume of steel. The stainless steel walls 

are 11/16 in. thick and the mass of this amount of steel is 2.07E+07 g. Using the mass of steel and no 

added grout, the sum of the fractions is well below unity. Table 16 shows the sum of the fractions for 

long-lived radionuclides using the mass of steel for half of a 30,000-gal tank. 
71

No analysis was performed for the 30,000-gal tanks using a ten-to-one ratio of unstabilized-tostabilized waste mass, since the calculations of final waste form concentrations did not use any stabilizing 

material. Only the mass of the tank steel was used in these calculations. 

Table 16. Sum of the fractions for 30,000-gal tanks using only the mass of steel (Table 1 of 

10 CFR 61.55). 

Conservative Vault 

Inventory 

Class C 

Concentration Limit 

Radionuclidea

Half-Life 

(yr) (Ci)b

 Ci/m3 nCi/g (Ci/m3 or nCi/g) 

Fraction of 

Class C 

Concentration 

Limit 

241Am 4.30E+02 6.40E–03 3.07E–01 1.00E+02 3.07E–03 

14C 5.70E+03 1.10E–07 4.27E–08 8.00E+00 5.34E–09 242Cm 4.50E–01 2.00E–05 9.62E–04 2.00E+04 4.81E–08 129I 1.60E+07 1.20E–05 4.54E–06 8.00E–02 5.68E–05 59Ni 7.50E+04 3.80E–04 1.47E–04 2.20E+02 6.68E–07 94Nb 2.00E+04 3.10E–03 1.21E–03 2.00E–01 6.03E–03 237Np 2.10E+06 7.10E–04 3.43E–02 1.00E+02 3.43E–04 238Pu 8.80E+01 1.70E–01 8.34E+00 1.00E+02 8.34E–02 239Pu 2.40E+04 5.10E–02 2.49E+00 1.00E+02 2.49E–02 240Pu 7.00E+03 2.00E–02 9.86E–01 1.00E+02 9.86E–03 

241Pu 1.40E+01 2.90E–01 1.42E+01 3.50E+03 4.07E–03 242Pu 3.80E+05 1.50E–05 7.21E–04 1.00E+02 7.21E–06 99Tc 2.10E+05 1.20E–02 4.48E–03 3.00E+00 1.49E–03 

Sum of the Fractions 0.13 

a. Radionuclides shown in italics are compared to Class C concentration limits in units of Ci/m3; remaining nuclides are compared to limits in 

units of nCi/g. 

b. Radioactive decay to 2012; the sum of the fractions will not significantly change if decayed to the present.

6.3.1 Summary and Conclusions 

At closure and disposal, the radioactive waste concentration for the 30,000-gal tank waste meets 

Class C concentration limits.ddd The basis for this is the completion of waste concentration tables as 

described in 10 CFR 61.55. As shown in the alternative concentration for the 30,000-gal tanks, the tank 

steel alone provides adequate mass and volume to produce concentrations of radionuclides that are well 

below the Class C concentration limits. Encapsulation grout to fill half of the tank is desirable to contain 

the residual waste that remains. 

6.4 Grouted Piping, Encasements, and Valve Boxes 

Like the other TFF system components, the TFF ancillary components and equipment (e.g., piping, 

encasements, valves, and valve boxes) will be grouted at final closure (i.e., disposal). The grout will 

provide a solid physical form to incorporate any residuals and to prevent their dispersal into the 

surrounding environment. The piping and ancillary equipment grout must necessarily be able to flow 

more easily during grouting operations, and consequently, this grout will have a higher water content than 

the grout mixture used for other TFF system components. However, the grout will still be able to harden 

into a solid physical form to provide waste stability. The pour process will be adjusted so that effective 

 

ddd. On page 62 of its TER (NRC 2006b), NRC concluded that the waste in the smaller (30,000-gal) tanks, sandpads, and piping 

would not exceed the Class C concentration limits. 
72

grout distribution through the piping, encasements, and other equipment is achieved. Characteristics of 

the grout are presented in Appendix C, along with the required QA tests and results from grout testing. 

To develop waste concentrations in the grouted piping, the residual inventory at closure for the 

piping presented in Section 2 (15.5 kg of SBW solids) is averaged over the 3,231 linear m (10,600 linear 

ft) of 5-cm (2-in.) piping (EDF-2973, 2001). Tables 17 and 18 show the radionuclide concentrations of 

the grouted piping. For reference, the Class C concentration limits for the long- and short-lived 

radionuclides are also shown. For perspective, Tables A-13 and A-14 in Appendix A show the 

radionuclide concentrations of the piping without adding grout to the pipe. In this case, the waste form is 

assumed to be Schedule 40 stainless steel pipe and the associated radioactive contamination. Tables 17 

and 18 include the mass and volume of the stainless steel piping and the mass and volume of the grout in 

the calculations while Tables A-13 and A-14 include only the mass and volume of the piping. The entire 

mass of stainless steel piping is included in the calculations. 

The valve boxes and encasements contain significantly less contamination after cleaning than the 

process piping (as described in Subsection 2.4.5). Therefore, the radionuclide concentrations in the valve 

boxes and encasements after closure would be less than those shown for the piping. 

Table 17. Radionuclide concentrations in the piping (with grout) (Table 1 of 10 CFR 61.55). 

Radionuclidea

Half-Life 

(yr) 

Piping 

Inventory

(Ci)b

Piping 

Inventory 

in Ci/m3

Piping 

Inventory 

in nCi/g 

Class C 

Concentration 

Limit 

(Ci/m3 or nCi/g)c

Fraction of 

Class C 

Concentration 

Limit 

241Am 4.30E+02 5.30E–03 1.70E–01 100 0.0017 

14C 5.70E+03 6.20E–08 7.10E–09 8 0.00000000089 242Cm 4.50E–01 1.70E–05 5.30E–04 20,000 0.000000026 

129I 1.60E+07 9.70E–06 1.10E–06 0.08 0.000014 

94Nb 2.00E+04 2.60E–03 2.90E–04 0.2 0.0015 

59Ni 7.50E+04 3.10E–04 3.60E–05 220 0.00000016 237Np 2.10E+06 5.90E–04 1.90E–02 100 0.00019 238Pu 8.80E+01 1.40E–01 4.60E+00 100 0.046 

239Pu 2.40E+04 4.30E–02 1.40E+00 100 0.014 240Pu 7.00E+03 1.70E–02 5.40E–01 100 0.0054 241Pu 1.40E+01 2.40E–01 7.80E+00 3,500 0.0022 244Pu 3.80E+05 1.20E–05 4.00E–04 100 0.0000040 

99Tc 2.10E+05 9.60E–03 1.10E–03 3 0.00036 

Sum of the Fractions 0.071 

 

a.Radionuclides shown in bold italics are concentration limits in units of Ci/m3; remaining nuclides are concentration limits in units of nCi/g. 

b. Radioactive decay to 2012; the sum of the fractions will not significantly change if decayed to the present. 

c. Table 1 of 10 CFR 61.55.
73

Table 18. Radionuclide concentrations in the piping (with grout) (Table 2 of 10 CFR 61.55). 

Radionuclide 

Half-Life 

(yr) 

Piping Inventory 

(Ci)a

Piping 

Inventory 

in Ci/m3

Class C 

Concentration Limit 

(Ci/m3)

b

Fraction of 

Class C 

Concentration 

Limit 

137Cs 3.00E+01 1.40E+01 1.60E+00 4,600 0.00036 

63Ni 1.00E+02 3.60E–02 4.10E–03 700 0.0000059 

90Sr 2.90E+01 2.90E–01 3.40E–02 7,000 0.0000048 

Sum of the Fractions 0.00037

 

a. Radioactive decay to 2012; the sum of the fractions will not significantly change if decayed to the present. 

b. Table 2 of 10 CFR 61.55.

No analysis was performed for the TFF piping using a ten-to-one ratio of unstabilized-to-stabilized 

waste mass. As described above, final waste form concentration calculations were performed with and 

without considerations for stabilizing material. 

6.4.1 Summary and Conclusions 

At closure and disposal, the radioactive waste concentrations for the piping, encasements, and 

valve boxes meet Class C concentration limits.eee The basis is the completion of waste concentration 

tables as described in 10 CFR 61.55. 

6.5 Sensitivity Analyses 

In addition to the above analyses, several additional evaluations have been performed to aid in 

understanding the sensitivity of residual inventory estimates, grout volumes, waste density, and other 

parameters. These additional evaluations are included in Appendix D and support the conclusion that the 

TFF tanks, vaults (including sandpads), and piping will meet Class C concentration limits at closure and 

disposal. 

6.6 Conclusions 

DOE has concluded that the radioactive waste concentrations for the TFF at closure and disposal 

would not exceed Class C concentration limits. In addition, DOE has extensively consulted with NRC on 

its disposal plans for TFF waste during the NRC’s review of the draft 3116 Determination. This extensive 

consultation would satisfy the requirements of Section 3116(a)(3)(B)(iii) if they were applicable. 

Therefore, DOE concludes that NDAA criteria can be met under either Section 3116(a)(3)(A) or 

(a)(3)(B). 

 

eee. On page 62 of its TER (NRC 2006b), NRC concluded that the waste in the smaller (30,000-gal) tanks, sandpads, and piping 

would not exceed the Class C concentration limits. 
74

7. WASTE WILL BE DISPOSED OF IN ACCORDANCE WITH 

PERFORMANCE OBJECTIVES IN 10 CFR 61, SUBPART C 

Key Points 

• The TFF PA (DOE-ID 2003b) was prepared in accordance with DOE guidance. Moreover, the 

information in the PA demonstrates that the performance objectives in 10 CFR 61, Subpart C are 

met. The PA is used to demonstrate that the TFF residual waste will be disposed of in 

accordance with these performance objectives as required by Section 3116 of the NDAA. 

• Grout in the tanks and tank vaults provides structural stability to the waste form and prevents 

collapse. 

• The institutional control period is 100 years following closure (2012–2112). The 

post-institutional control period is 2112–3012. 

• Members of the public are assumed to reside 100 m (328 ft) from the TFF during the 

post-institutional control period. The all-pathways dose assessment used the well location where 

the highest concentration of radionuclides was predicted in the aquifer (600 m [1,969 ft] from 

TFF). 

• Grouted tanks and piping are assumed to fail (release contaminants) at 500 years post-closure. 

Grouted vaults are assumed to fail (release contaminants) at 100 years post-closure. 

• Inadvertent intruder scenario assumptions: 

- Institutional controls are effective for at least 100 years following closure. 

- Only the intruder-drilling and post-drilling scenarios are applicable for the tank and 

sandpad contamination as discussed in Subsection 7.2. The intruder-construction and 

intruder post-construction scenarios are considered for the piping inventory located less 

than 3 m (10 ft) below the surface. 

Section 3116(a) of the NDAA 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. 

Title 10 CFR 61, Subpart C, Sections 61.40 through 61.44, detail performance objectives the NRC 

established for land disposal of radioactive waste. For LLW disposal, 10 CFR 61.40, “General 

Requirement,” 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 §§61.41 through 61.44.” These sections address protection of the general 

population from radioactivity releases, protection of individuals from inadvertent intrusion onto the 

disposal site, protection of individuals during disposal facility operations, and disposal site stability after 

closure. Compliance with requirements in 10 CFR 61.40–61.44 is discussed in Subsections 7.8–7.12, 

respectively. 
75

This section documents that the TFF PA shows that disposal of the residuals and TFF tank system 

will meet 10 CFR 61, Subpart C performance objectives. The PA was developed originally as part of a 

DOE Tier 1 closure plan (DOE-ID 2006a) and as the support for a DOE draft waste determination prior to 

enactment of Section 3116 of the NDAA. As such, the PA was prepared in accordance with DOE 

guidance. However, the information in the PA similarly demonstrates that the LLW land disposal 

performance objectives of 10 CFR 61, Subpart C are met. 

The site-specific Performance Assessment for the Tank Farm Facility at the Idaho National 

Engineering and Environmental Laboratory (DOE-ID 2003b) considers all the sources of radioactive 

waste that may remain in the closed TFF, including the tank residuals, tanks, vaults, sandpads, piping, and 

other ancillary components for the eleven 300,000-gal tanks and vaults and the four 30,000-gal stainless 

steel tanks. The assessment calculates potential releases to the public, including those that result from 

inadvertent intrusion, to provide a reasonable expectation that the performance objectives are not 

exceeded for a 1,000-year period after closure, which is planned to be in 2012. The PA uses the current or 

planned land use as a basis for scenario development. The receptor is placed in all scenarios where the 

estimated radiation dose from the TFF is greatest. Therefore, the exposure date and location are placed at 

the predicted maximum radiation dose. Assumptions in the PA include grouting all voids such that the 

tanks and tank vaults (for the 300,000-gal tanks) are filled with grout to provide structural stability to the 

waste form and prevent collapse. The PA process uses conceptual models that link radionuclide 

inventory, release (or source term), environmental transfer, and impact assessment. The final step is 

determining radiological doses to receptors. These analyses are used in this 3116 Basis Document to 

demonstrate the TFF closure (i.e., disposal) is in compliance with Section 3116 criteria. 

The potential dose posed by the TFF closure (i.e., disposal) is assessed by conservatively 

estimating the release and transport of radionuclides from (1) the release of radionuclides from the tanks, 

sandpads, and piping, and subsequent transport through the environment; and (2) exposures to members 

of the public via air, groundwater, and food chain pathways. Two receptor types are assessed: a member 

of the public and an intruder. During the operational and institutional control periods, a member of the 

public is assumed to reside at the INL Site boundary. During the post-institutional control period 

(100 years after closure), a member of the public is assumed to reside at the INTEC facility boundary. 

This receptor is used to evaluate the drinking water pathway, all-pathways (including food chains), and 

air pathway doses for comparison to the performance objectives. The hypothetical intruder receptor type 

is assumed to intrude inadvertently onto the TFF during the post-institutional control period under four 

scenarios. 

Calculations in sensitivity and uncertainty analyses for all analyses other than the inadvertent 

intruder analysis were carried out beyond 1,000 years to determine the maximum impact regardless of the 

time at which the maximum occurs. Thus, the timeframe for the analysis should be long enough to 

capture the maximum impact even if the maximum does not occur for several thousand years. For the 

inadvertent intruder analysis, the sensitivity and uncertainty analysis is limited to qualitative arguments. 

These analyses, together with the sensitivity and uncertainty analyses discussed below, were used in this 

3116 Basis Document and provide significant information on the reasonably expected performance up to 

10,000 years and beyond. 

The NRC’s LLW PA guidance suggests a two-part approach to establishing the time of analysis 

(NRC 2000). The first part is a 10,000-year analysis period and the second part is a qualitative evaluation 

of the analysis beyond 10,000 years to identify any significant deficiencies in disposal system 

performance. In the TFF PA, groundwater contaminant transport simulations (i.e., the dominant dose 

pathway) are conducted for three radionuclides (90Sr, 99Tc, and 129I) that provide the majority of the 

groundwater pathway doses during the 1,000-year analysis period. 
76

The PA’s performance period of 1,000 years after INTEC TFF closure includes an institutional 

control period and a post-institutional control period, while the sensitivity analysis for the groundwater 

pathway provides simulations out past 10,000 years. The use of the groundwater modeling past 10,000 

years, coupled with the intruder scenarios, which peak immediately after the institutional control period 

(100 years following TFF closure).fff 

The PA assumes an institutional control period at the INL Site of 100 years following TFF closure. 

Developing an assumption for institutional controls in performance assessments includes consideration of 

site-specific land use planning and planning for engineered controls such as active barriers (e.g., caps, 

fences, and security) and passive barriers (e.g., signs, monuments, and markers). The INEEL 

Comprehensive Facility and Land Use Plan (DOE-ID 1997) describes government control of the site for 

at least the next 100 years. The INL has not prepared specific plans or designs for active or passive 

controls following this timeframe. Consistent with guidance in NRC’s LLW PA (NRC 2000), a 100-year 

post-closure institutional control period has been selected. 

During this institutional control period, periodic maintenance and monitoring activities are 

assumed to be conducted, and active controls such as guards and fences are expected to be used. During 

the institutional control period, members of the public are assumed to reside at the INL Site boundary and 

intruders are assumed to be restricted from entering the site (DOE-ID 1995). During the post-institutional 

control period (2112–3012), the INL Site will not be maintaining the facility, so the facility could be 

accessed by the public. During this period, members of the public are assumed to reside as close as 100 m 

(328 ft) from the TFF. The PA evaluates the degradation of the various barriers (e.g., grout, tanks, vaults, 

and piping) to radionuclide release. Although models predict that these barriers will last 40,000 years, the 

TFF PA assumes a tank and tank grout life of only 500 years (NRC 2000), and a life of only 100 years for 

the vaults and grout system between the tank and the vault outer walls. The groundwater model analysis 

shows that the contamination plume center (where the highest concentrations enter the regional aquifer) 

would be 600 m (1,969 ft) southward in the downgradient direction from the center of the southernmost 

TFF tank. The contamination plume center is taken as the source of drinking water after the institutional 

control period of 100 years has expired. This center is the distance from the TFF for the highest calculated 

radiation dose. The intruder scenario considers contamination spread from two activities: a 3-m (10-ft) 

excavation for a basement and a well drilled through the top of the tank. The maximum dose would occur 

immediately after institutional control ended at 100 years after closure (2112). 

The dose posed by the residual inventory at closure, the general requirements for disposal in a near 

surface facility described in 10 CFR 61.40, and the performance objectives for LLW disposal described in 

10 CFR 61.41–61.44 are discussed in the following subsections. 

7.1 Analysis of Performance 

The conceptual models developed and the computational approach used to assess the performance 

of the TFF are also described fully in the PA. The conceptual models are derived from technical 

information presented in Section 2 of the PA (DOE-ID 2003b). These models embody a number of 

simplifying assumptions to facilitate the computational analysis required to assess the long-term 

performance of the TFF. The dose calculations are presented in Subsection 3.3.5 of the PA. 

 

fff. The contaminant transport simulations are evaluated out to 1 million years for these radionuclides. Prior modeling 

simulations for the TFF assessed the remaining radionuclides for a period of 1 million years. The results of this prior modeling 

indicate that the remaining radionuclides are not of concern for the TFF PA (DOE-ID 2003b). 
77

7.1.1 Institutional Control Period 

The institutional control period is the 100-year time interval specified in the INEEL Comprehensive 

Facility and Land Use Plan (DOE-ID 1997) following closure of the TFF. The assumed closure date of 

the TFF is currently 2012. Periodic maintenance and monitoring activities are conducted during the 

institutional control period. The TFF site is assumed to be stabilized and no longer operational during this 

period but will remain part of the INL, and therefore, fenced and patrolled to eliminate the possibility of 

inadvertent intruders. 

7.1.2 Post-Institutional Control Period 

The final time period of concern is when the facility is no longer maintained by the INL Site and 

could be accessed by the public. The total duration of this period is 1,000 years from the time of closure; 

however, to determine groundwater dose, which may peak at a later time, simulations were carried out to 

1 million years. Therefore, the 10,000-year period of performance is addressed by the PA. Projections of 

conditions and activities after 1,000 years are uncertain and difficult to assess. In addition, peak doses 

may not occur from TFF releases until several thousand years after closure because of the presence of 

long-lived radionuclides. The potential peak doses from these long-lived radionuclides are in the 

sensitivity analysis. 

7.1.3 Exposure Pathways 

Numerous exposure pathways were analyzed in the TFF PA to demonstrate conformance with a 

variety of performance objectives. These exposure pathways include: (1) drinking water dose from 

groundwater,ggg (2) all-pathways dose from groundwater, and (3) air dispersion pathways.hhh In addition, 

the intruder pathway and radon flux analyses were also considered and are described in Subsections 3.3.4 

and 5 of the TFF PA (DOE-ID 2003b). Only the all-pathways dose assessment and intruder analysis are 

applicable to demonstrate compliance with the NDAA criteria and will be further discussed in this 3116 

Basis Document. Intruder scenarios are discussed in later in Section 7. 

The TFF PA includes a comprehensive all-pathways exposure scenario. The primary mechanism 

for transport of radionuclides from the TFF is expected to be leaching to the groundwater and subsequent 

human consumption. Thus, in the all-pathways dose analysis, an off-Site member of the public is assumed 

to use water from a well for domestic purposes. The well is assumed to be located where the maximum 

concentrations of radionuclides in groundwater are predicted to occur. 

 

ggg. The groundwater protection standard chosen for evaluation in the PA is a drinking water dose of 4 mrem/yr. The nearest 

location from the disposal site for off-Site members of the public depends on the time period after disposal. During the period of 

active institutional control for the first 100 years after facility closure, off-Site members of the public are assumed to be located 

no closer to the disposal site than the present boundary of the INL. However, after active institutional control ceases, off-Site 

members of the public could be located as close as 100 m (328 ft) from the TFF. However, for the groundwater pathway, the 

member of the public is assumed to have a well located in the groundwater at the point of maximum concentration, which for the 

analyses presented in the PA is located 600 m (1,969 ft) from the TFF. 

The primary mechanism for transport of radionuclides from the TFF is expected to be leaching of radionuclides to the 

groundwater and subsequent human consumption. Thus, in the dose analysis for groundwater protection, an off-Site member of 

the public is assumed to use water from a well for domestic purposes. The well is assumed to be located where the maximum 

concentrations of radionuclides in groundwater are predicted to occur. 

hhh. Volatile radionuclides 3

H and 14C were evaluated in the air dispersion scenario. The depth to the waste and the waste 

physical characteristics (i.e., grouted waste) limits the analysis to volatile radionuclides in the waste form. Additional details on 

the air dispersion scenario are provided in Subsection 3.3.5 of the PA (DOE-ID 2003b). 
78

The following exposure pathways involving the use of contaminated well water are assumed to 

occur: 

• Direct ingestion of contaminated water 

• Ingestion of milk and meat from dairy and beef cattle that drink contaminated water 

• Ingestion of vegetables grown in garden soil irrigated with contaminated water 

• Ingestion of milk and meat from dairy and beef cattle that eat fodder from pasture irrigated with 

contaminated water. 

Additional exposure pathways for off-Site members of the public could involve releases of 

radionuclides into the air (i.e., volatile radionuclides). Exposures from the air pathway also are considered 

in the PA. 

7.2 Models and Assumptionsiii

7.2.1 Source Term Model 

For the TFF concrete vaults, release rates of radionuclides were estimated using the Disposal Unit 

Source Term-Multiple Species (DUST-MS) computer code (Sullivan 1993, 2001). One-dimensional 

DUST-MS transport simulations were conducted for radionuclide sources in the grouted tank, piping, and 

the sandpad beneath the tank. 

7.2.1.1 DUST-MS Model Description. DUST-MS considers four major processes: fluid flow, 

container degradation, waste form leaching, and contaminant transport. The DUST-MS model permits 

selection of a unique failure time and waste form type for each container. To simulate different waste 

forms, DUST-MS has four different models to estimate release rates: rinse with partitioning, diffusion, 

uniform degradation, and solubility-limited release. For these four processes, DUST-MS computes release 

rates with an analytical model or a finite difference model. After calculating waste form releases, the 

movement of the contaminants through subsurface system materials is determined using a 

one-dimensional finite difference procedure with material- and contaminant-specific distribution 

coefficients, diffusion coefficients, dispersion coefficients, and bulk densities. The governing transport 

equation simulates the distribution and movement of contaminants from advection, dispersion, diffusion, 

radioactive production and decay, sorption, and sources and sinks external to the containers. The 

DUST-MS code has received extensive testing and verification. The DUST-MS code predictions compare 

favorably to known analytical solutions as well as other code predictions (DOE-ID 2003b).

7.2.1.2 Conceptual Model. The conceptual model for computing release rates of radionuclides 

out of the TFF vault is illustrated in Figure 15. Infiltrating water from the ground surface contacts the 

waste inside the vaults. 

 

iii. Although NRC expressed several concerns in its TER (NRC 2006b) regarding the DOE’s conceptual model with respect to 

hydrology, NRC concluded for a number of reasons that “NRC has confidence that overall system performance of both the 

engineered and natural barriers will be sufficient to meet performance objectives” (as noted on page 101 of the TER). DOE 

acknowledges differences in professional opinion regarding scientifically complex modeling, including approaches to account for

uncertainties. For the reasons explained in the ensuing subsections and in DOE’s PA (DOE-ID 2003b), DOE has confidence in its 

modeling approach. Importantly, despite certain differences in professional opinion and approach, both NRC and DOE reached 

the same conclusion, that is, that the performance objectives in 10 CFR 61, Subpart C will be met. 
79

Figure 15. The DUST-MS conceptual model for release from the TFF vaults and tanks (DOE-ID 2003b). 

In Figure 15, radionuclides are assumed to be located in the grouted tank heel and in the sandpad 

beneath the tank. The release of radionuclides from the grouted tank heel and from the sandpad is 

modeled in DUST-MS assuming surface rinsing. The vault is assumed to remain intact for 100 years. The 

tanks are assumed to remain intact for 500 years. Upon vault and tank failure, radionuclides are released 

from the two sources. 

The DUST-MS code calculates release rates out of the vault and tanks into the unsaturated zone. 

Release rates are computed for transport through the grout inside the tank (15 cm [6 in.]), the sandpad 

(15 cm [6 in.]), and the degraded concrete (0.76 m [2.5 ft]). Retardation is assumed to occur in the waste 

release model (surface rinsing), in the grout inside the tank, in the sandpad, and in the degraded concrete. 

7.2.2 Groundwater Flow and Transport Modeling 

A two-dimensional unsaturated/saturated PORFLOW model was used to simulate water and 

contaminant transport in the subsurface at the INTEC facility (see Figure 16) (ACRi 2000). This approach 

allowed a detailed approximation of the complex geology underlying the facility. The two-dimension 

problem domain represented a vertical slice in a north-south direction beginning at the Big Lost River, 

through the center of two tank vaults, and southward in the downgradient direction for a total distance of 

2,500 m (8,202 ft). The top of the model is located at land surface and extends to a depth of 200 m 

(656 ft)—well below the top of the water table located at 134–139 m (440–456 ft) below ground surface. 

By default, the unit thickness of the two-dimensional model in an east-west direction is 1 m (3 ft). 
80

Figure 16. Illustration of the two-dimensional modeling slice used in PORFLOW (DOE-ID 2003b). 

As noted earlier, DUST-MS was used to predict the release rates from the various waste forms. 

Radionuclide releases coincided with degradation rates of the waste forms discussed in Subsection 3.1.2.2 

of the PA (DOE-ID 2003b). Release rates calculated by DUST-MS were incorporated as input in 

PORFLOW, with the initial radionuclide releases beginning at the time of degradation. For example, the 

outer vault began to release radionuclides after 100 years and the tanks/piping after 500 years. These 

assumptions are conservative, based on the best estimates provided from the degradation analysis. 

An important component of the conceptual model for the transport of contaminants is the 

relationship of dispersion in the two-dimensional model. Since the model uses a unit thickness of 1 m 

(3 ft), there is no lateral dispersion of contaminants beyond this thickness. It is assumed that contaminants 

disperse or diffuse out of the lateral boundary at the same rate as contaminants move into the unit 

thickness of the model domain. Essentially, all dispersion of contaminants occurs in the longitudinal 

direction, with zero transverse dispersion. This is a conservative approach to predicting downgradient 

contaminant concentrations. 

Since the model domain is located in the center of the tanks, the highest contaminant 

concentrations are located in the source area of the model. Although it is assumed that the same amount 

of contaminants disperse in a transverse direction out of, as into the model domain, in reality there would 

be the same loss of contaminants, albeit small, in the transverse direction. Consequently, downgradient 

contaminant concentrations will be slightly higher for the two-dimensional simulation compared to a fully 

three-dimensional simulation. 
81

Radionuclide concentrations were observed downgradient in the regional aquifer at the location 

where the highest concentrations occur. During unsaturated flow simulations, it was observed that the 

perched water zones deflected the contamination. Consequently, the usual 100-m (328-ft) downgradient 

location was inappropriate to quantify the maximum impacts of radionuclide releases from the tanks and 

vaults. Based on the radionuclide concentrations distribution, the maximum observed concentrations 

occur approximately 600 m (1,969 ft) downgradient from the tanks. 

The intruder scenarios were screened based on the physical aspects of the facility (i.e., depth to the 

waste). Intruder scenarios considered for use in the PA were those previously described for low-level 

radioactive waste PAs (Kennedy and Peloquin 1988; NRC 1982).jjj These intruder scenarios include both 

acute and chronic exposure scenarios. Acute exposure scenarios involve exposures of short duration, and 

include an intruder-construction scenario, a discovery scenario, and a drilling scenario. 

7.2.2.1 Intruder-Construction Scenario. The intruder-construction scenario involves an 

inadvertent intruder who chooses to excavate or construct a building on the disposal site. In this scenario, 

the intruder is assumed to dig a basement excavation to a depth of approximately 3 m (10 ft) (Oztunali 

and Roles 1986). It is assumed that the intruder does not recognize the hazardous nature of the material 

excavated. He or she is exposed to radioactive constituents in the waste during the excavation of the 

basement. The intruder also is exposed to the exhumed waste by inhalation of resuspended contaminated 

soil and external irradiation from contaminated soil. 

The depth to the residual waste at the bottom of the tanks and/or sandpads (i.e., 3 m [10 ft] of soil 

plus 9 m [30 ft] of concrete) would preclude direct contact with the waste from the 3-m (10-ft) 

excavation. However, approximately 30% of the contaminated piping associated with the tanks is located 

within 3 m (10 ft) of the surface. 

Due to the disposal depth of the waste in the tanks at the TFF (i.e., greater than 10 m [33 ft]) and 

the depth of the sandpad contamination, the intruder-construction scenario was not considered applicable 

to these waste. However, the intruder-construction scenario was considered for the radionuclide inventory 

located in the associated piping less than 3 m (10 ft) from the surface. 

7.2.2.2 Intruder-Drilling Scenario. The intruder-drilling scenario assumes the short-term 

exposure of a hypothetical intruder to drill cuttings from a borehole penetrating the waste disposal site. 

This scenario involves wastes buried below the depth of typical construction excavations.

Oztunali and Roles (1986) indicate that for waste below 10 m (33 ft), the only applicable intrusion 

scenario is the intruder-drilling scenario. They also note that for grouted waste or waste disposed of in 

reinforced concrete structures, a time period of 500 years after site closure is assumed as the effectiveness 

limit for this waste form. Therefore, this scenario is not considered applicable for drilling through 

reinforced concrete until 500 years post-closure. However, they also note that the scenario is assumed to 

be fully applicable for grouted waste at any time. Because the waste could be contacted and moved to the 

surface, thereby exposing the intruder, the intruder-drilling scenario was retained for analysis in the TFF 

PA (DOE-ID 2003b). 

 

jjj. The Idaho High-Level Waste and Facilities Disposition Final Environmental Impact Statement (DOE 2002) considered 

similar intruder scenarios. While not part of this 3116 Basis Document, DOE believes the analysis in the environmental impact 

statement is consistent with the analysis of intruder scenarios in this document. 
82

7.2.3 Chronic Intruder Exposure Scenarios 

These scenarios are described below. Those scenarios not considered applicable to the TFF were 

screened out from further consideration. 

7.2.3.1 Intruder Post-Construction Scenario. The chronic intruder post-construction 

(i.e., agriculture) scenario is an extension of the acute intruder-construction scenario. It is assumed in this 

scenario that an intruder lives in the building constructed as part of the intruder-construction scenario and 

engages in agricultural activities on the contaminated site. The intruder is exposed to contamination by 

inhalation of resuspended contaminated soil, inhalation of gaseous radionuclides released from the waste, 

external irradiation, ingestion of contaminated soil, ingestion of contaminated beef and milk, and 

ingestion of contaminated vegetables. 

As stated earlier, the intruder-construction scenario was only considered applicable to the piping 

associated with the TFF that is less than 3 m (10 ft) from the surface. The depth of the residual waste in 

the tanks and the sandpads prevents its excavation to the surface. Therefore, the intruder-agriculture 

scenario was retained for the piping inventory. 

7.2.3.2 Post-Drilling Scenario. The chronic post-drilling scenario is an extension of the acute 

drilling scenario. It assumes that the intruder occupies the site after drilling a water well and grows crops 

on a mixture of clean soil and contaminated drill cuttings. After exhumation of the waste, the exposure 

pathways are the same as for the intruder-agriculture scenario. This intruder scenario was retained for 

further analysis.

7.3 Acute Intruder-Drilling Scenario Definition 

The acute intruder-drilling scenario assumes that an inadvertent intruder drills a well into the 

contents of the tank and vault system. The intruder is exposed to contaminated drill cuttings spread over 

the ground and contaminated airborne dust. In the standard drilling scenario used in many PAs, the 

intruder is assumed to be exposed to contaminated drill cuttings in a mud pit. However, site-specific 

information developed through interviews with local well drilling contractors in the Idaho Falls area 

indicates that drillers spread the cuttings over the ground and do not use mud pits (Seitz 1991). The 

authors of the Radioactive Waste Management Complex PA (Maheras et al. 1997) used this site-specific 

deviation of the standard drilling scenario; it also was incorporated into the TFF intruder-drilling scenario. 

The assumption that the drill cuttings are spread over the ground will result in higher dose estimates than 

if the cuttings were assumed to be in a mud pit because of the decrease in the shielding factor. 

The drill cuttings are assumed to be spread over a 2,200-m2

 (23,681-ft2) lot, which corresponds to 

about 0.5 acre. Typical lot sizes located outside of the Idaho Falls city limits are typically 1–3 acres. 

Therefore, a 2,200-m2

 (23,681-ft2) lot size was considered conservative for use in the Radioactive Waste 

Management Complex PA (Maheras et al. 1997), and also was incorporated into the TFF intruder-drilling 

scenario. 

Well drilling contractors in the Idaho Falls area have reported that two types of wells are typically 

drilled: small-diameter residential wells and large irrigation wells. The small residential wells are 

typically 15–20 cm (6–8 in.) in diameter, serve a single residence, and may provide enough water for a 

family garden and small quantities of livestock. The large-diameter irrigation wells are drilled to serve 

systems that irrigate hundreds of acres; the wells are located in the middle of farm fields, not near the 

farmer’s residence. Therefore, a farmer would not drill an irrigation well to acquire water for his 

residence. Large-diameter irrigation wells are currently drilled in 46-cm (18-in.) diameters but drilling 

contractors indicated that 56-cm (22-in.) diameter irrigation wells would be drilled in the future 
83

(Seitz 1991). An acute drilling exposure could result from either drilling a 15-cm (6-in.) diameter 

residential well or a 56-cm (22-in.) diameter irrigation well. The larger 56-cm (22-in.) diameter irrigation 

well was assessed for the acute intruder-drilling scenario. 

The intruder is assumed to reside by the contaminated cuttings for 160 hours, the time local Idaho 

Falls well drilling contractors say it would take to drill and develop a 56-cm (22-in.) diameter irrigation 

well (Seitz 1991). The exposure pathways for this acute drilling scenario include inhalation of 

resuspended drill cuttings, external exposure to the ground source, and inadvertent soil ingestion. Figure 

17 illustrates the acute intruder-drilling scenario. 

The TFF concrete vaults are reinforced concrete but the grout between the tank and vault and also 

inside of the tank is not reinforced concrete. However, considering that the depth to the waste in the tank 

is 10 m (33 ft), that the waste is contained within a stainless steel tank, and that the overlying concrete 

vault roof is reinforced, credit beyond 100 years could be taken for a barrier to intrusion. However, the 

intruder-drilling scenario was assumed to begin 100 years after closure. 

Vault 

Water well 

22-in. diameter 

2,200 m 

contaminated 

area 

2 

Figure 17. Graphical representation of the acute intruder-drilling scenario (DOE-ID 2003b). 
84

7.4 Acute Intruder-Construction Scenario 

The acute intruder-construction scenario assumes that an inadvertent intruder moves onto the TFF 

and excavates a 20- by 10- by 3-m (66- by 33- by 10-ft) excavation for a basement (Oztunali and Roles 

1986). 

The intruder is assumed to spend 160 hours working on the excavation of the basement. The 

exposure pathways for this acute scenario include inhalation of resuspended soil, external exposure to the 

ground surface, and inadvertent soil ingestion. 

The determination of the unit concentration dose factors for each exposure pathway was 

determined using the equations in Subsection 5.2 of the PA (DOE-ID 2003b) for the acute 

intruder-drilling scenario with the exception that the external dose factors for contaminated soil were 

based on infinite thickness and infinite lateral extent from Federal Guidance Report 12 (EPA 1993). 

The activity concentration of radionuclides in the excavation were determined by assuming that the 

radionuclide inventory in the 30% of the piping located within 3 m (10 ft) of the surface was located over 

the 5-acre (20,234-m2

 [217,797-ft2]) area of the TFF. 

The results of the intruder-construction analyses are given in Subsection 5.6 of the PA 

(DOE-ID 2003b). 

7.5 Chronic Intruder Post-Drilling Scenario Definition 

The chronic intruder post-drilling scenario assumes that an inadvertent intruder moves onto the 

TFF and drills a residential well into the waste. The drilling portion of the scenario evaluates a 15-cm 

(6-in.) diameter residential well. This type of well serves a single residence and provides sufficient water 

for a family garden and small quantities of livestock. As described in the acute intruder-drilling scenario, 

large-diameter wells are drilled to serve irrigation systems (i.e., hundreds of acres) that are located in the 

middle of farm fields, not near a farmer’s residence. Therefore, in the chronic post-drilling scenario, the 

residence/home garden is evaluated using the traditional drinking water well diameter of 15 cm (6 in.). 

The drill cuttings that are brought to the surface are assumed to be spread over 2,200 m2

(23,681-ft2

) (approximately 0.5 acre) of land surface. The waste is assumed to be mixed to a depth of 

0.6 m (2 ft). The mixing depth of 0.6 m (2 ft) is based on using a deep tilling plow to increase the depth of 

the root zone and to break up soil compaction. To minimize erosion these plows are used in areas of 

southeastern Idaho with highly erodible soils. Deep tilling plows have shanks that till to a depth of 0.6 m 

(2 ft) and are sold at Idaho Falls implement dealers (Maheras et al. 1997). 

The chronic post-drilling scenario assumes that the intruder is exposed to the drill cuttings during 

plowing and cultivation (i.e., dust inhalation). In addition, the intruder is assumed to ingest contaminated 

food products from the garden and from beef and milk cattle consuming contaminated forage. The intake 

of contaminated forage by cattle was adjusted according to the fraction of feed grown on contaminated 

cuttings and the necessary remaining feed obtained from uncontaminated ground. Figure 18 illustrates the 

chronic intruder post-drilling scenario. 
85

Figure 18. Chronic intruder post-drilling scenario (DOE-ID 2003b). 

7.6 Chronic Intruder Post-Construction Scenario Definition 

The chronic intruder post-construction scenario assumes that an inadvertent intruder moves onto 

the TFF, excavates a basement, spreads the excavated material over the land, and subsequently farms the 

area. 

The excavation soil is assumed to be spread over 2,200 m2

 (23,681 ft2) (approximately 0.5 acre) of 

land surface. The waste is assumed to be mixed to a depth of 0.6 m (2 ft), which is based on using a deep 

tilling plow to increase the depth of the root zone and to break up soil compaction. These plows are used 

in areas of southeastern Idaho with highly erodible soils to minimize erosion. Deep tilling plows have 

shanks that till to a depth of 0.6 m (2 ft) and are sold at Idaho Falls implement dealers (Maheras et al. 

1997). 

The chronic intruder post-construction scenario assumes that the intruder is exposed to the 

excavated soil during plowing and cultivation (i.e., dust inhalation). In addition, the intruder is assumed to 

ingest contaminated food products from the garden and from beef and milk cattle that consume 

contaminated forage. The intake of contaminated forage by cattle was adjusted according to the fraction 

of feed grown on contaminated cuttings and the necessary remaining feed obtained from uncontaminated 

ground. 
86

7.7 Estimated Radiation Dose Using the Current Estimate of the 

Residual Inventory at Closure 

Performance objectives for radioactive waste land disposal are provided in 10 CFR 61, Subpart C. 

Table 19 compares the dose based on the conservative residual inventory used in the PA and the dose 

posed by the current estimate of the residual inventory at closure (described in Section 2) with the listed 

performance objectives. Results of the PA analyses indicate compliance with the performance objectives 

shown in Table 19. The dose for the current estimate of the residual inventory at closure for the TFF tank 

system is less than the dose calculated for the inventory used in the PA. The doses by radionuclide for the 

groundwater all-pathways, acute intruder-drilling scenario, and chronic intruder-drilling scenario are less 

than the doses predicted in the PA. The dose comparison shows a decline in dose to various receptors.kkk 

Table 19. Comparison of performance assessment results to predicted dose at closure (DOE-ID 2003b). 

Performance Objectives PA Results 

Current Estimate of 

Residual at Closure 

All-pathways dose to the public (not exceeding 25 mrem/yr) 1.86 mrem/yr 0.46 mrem/yr 

Acute-drilling scenario (less than 500 mrem) 232 mrem 152 mrem 

Acute-construction scenario (less than 500 mrem) 0.80 mrem 0.23 mrem 

Chronic post-drilling scenario (less than 500 mrem/yr) 91.1 mrem/yr 25 mrem/yr 

Chronic post-construction scenario (less than 500 mrem/yr) 26.1 mrem/yr 3.15 mrem/yr 

Notes: 

1. The groundwater pathway contributed 1.35 mrem/yr. 

2. The peak annual dose to the thyroid is approximately 6 mrem/yr compared to the 10 CFR 61.41 limit of 75 mrem/yr. 

3. The peak annual dose to any other organ is approximately 0.15 mrem/yr compared to the 10 CFR 61.41 limit of 

25 mrem/yr. 

For the residual inventory at closure, the Tanks WM-180 and WM-182 inventories are used as a 

basis for the estimated doses shown in the last column of Table 19. Tanks WM-180 and WM-182 are 

used because this tank combination has the largest residual inventory of the cleaned 300,000-gal tanks. 

The modeling in the PA was designed to use the inventory from a set of tanks that are oriented north to 

south to match the regional groundwater flow. The radiation doses are estimated using a dose conversion 

factor (DCF), which is the ratio between the radiological inventory (in Ci) and the radiological dose (in 

mrem/yr), as determined in the PA. Although the DCF is a ratio determined from data used in the PA, the 

same ratio is applicable to the residual material present in a cleaned tank at the completion of closure 

activities. Therefore, since the post-closure inventory of the cleaned tanks is known, the exposure 

scenario doses may be determined using the DCF. A description of the methodology and results are 

 

kkk. The Idaho High-Level Waste and Facilities Disposition Final Environmental Impact Statement (DOE 2002) shows higher 

doses for the action alternatives concerning tank closure than shown in this 3116 Basis Document. The EIS analysis is based on 

an inventory that does not reflect DOE’s current refined and enhanced methodology for cleaning the tanks. The TFF PA 

(DOE-ID 2003b) reflects the potential removal of radionuclides using DOE’s current cleaning methodology and shows lower 

inventory and calculated doses than the EIS. The inventory and calculated doses in this 3116 Basis Document (because Section 

3116 calls for removal of radionuclides to the maximum extent practical) are based on actual and projected tank cleaning results. 

The inventory and calculated doses are less than either the TFF PA or the EIS. (See Staiger, M. D., and C. B. Millet, 2000, 

“Inventory Estimates for the Tank Farm and CSSF’s, MDS-01-00,” Idaho National Engineering and Environmental Laboratory, 

interdepartmental communication to J.T. Beck, February 18. [cited at page C.9-50, DOE 2002]; TtNUS [Tetra Tech NUS], 2001, 

Calculation Package for Appendix C.9 to the Idaho High-Level Waste and Facilities Disposition EIS, Tetra Tech NUS, Aiken, 

South Carolina. Id;. EIS Appendix C, pages C.9-24 to 25.) 
87

shown in “Radiological Dose Calculations for the Tank Farm Facility at the Idaho Nuclear Technology 

and Engineering Center” (Portage 2005g). The ratio for DCFs is shown as an equation below. 

DCF

PA

PA

i

d =

DCFxCRi = CRd

where: 

PAi = performance assessment inventory 

PAd = performance assessment radiation dose 

DCF = dose conversion factor 

CRi = current residual inventory 

CRd = current inventory radiation dose. 

Since the PA inventory, the PA radiation dose, and the current estimate of residual inventory are 

known, the current residual inventory radiation dose can be calculated using the simple ratio shown 

above. 

7.8 10 CFR 61.40, “General Requirement” 

The general requirement of 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 TFF PA was developed under similar guidance and requirements from DOE, which required 

the analysis to provide a “reasonable expectation” that the performance objectives are not exceeded as a 

result of operation and closure of the facility. The PA meets the performance objective of “reasonable 

assurance” by using conservative estimates of parameter values, concrete degradation rates, and human 

activity models. In addition, a sensitivity and uncertainty analysis is used to ensure that a conservative 

analysis was conducted for the TFF closure that bounds the potential final closure status of the facility. 

The following subsections compare each of the 10 CFR 61 performance objectives with those used 

in the TFF PA analyses.lll 

 

lll. The NRC issued the requirement for LLW disposal, 10 CFR 61, on December 27, 1982. The requirement was developed to 

provide specific regulations for the disposal of LLW in near-surface disposal facilities. 
88

7.9 10 CFR 61.41, “Protection of the General Population from 

Releases of Radioactivity” 

The measure for protecting the public from radioactive material that may be released from the 

disposal facility by any pathway is established in 10 CFR 61.41. The NRC requirement states: 

Concentrations of radioactive material which may be released to the 

general environment in groundwater, surface water, air, soil, plants, or 

animals must not result in an annual dose exceeding an equivalent of 25 

mrems to the whole body, 75 mrems to the thyroid, and 25 mrems 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. 

The standard for radiation dose to the general population reflects the standard dose methodology in 

use at the time 10 CFR 61 was promulgated, which was International Commission on Radiological 

Protection (ICRP)-2 (ICRP 1959). The TFF PA performance objective is stated as “doses to the 

representative members of the public shall not exceed 25 mrem (0.25 mSv) in a year total effective dose 

equivalent from all exposure pathways, excluding the dose from radon and progeny in air.” The 

performance measure of 25 mrem in a year total effective dose equivalent (TEDE) reflects the standard 

dose methodology currently in use, ICRP-30 (ICRP 1979). The measure of protection, 25 mrem, is the 

same. 

The NRC Performance Assessment Working Group (NRC 2000) recommends using ICRP-30 dose 

methodology in computing potential dose from a LLW disposal facility to compare to this performance 

objective. The group stated, “that as a matter of policy, the Commission considers 0.25 mSv/yr 

(25 mrem/yr) TEDE to be an appropriate dose limit to compare with the range of potential doses 

represented by the older whole body dose limits.” Therefore, the performance objective for the general 

population used in the TFF PA is in agreement with 10 CFR 61.41. 

Because 10 CFR 61.41 also has limits for doses to the thyroid and other key organs, these doses 

are calculated for this 3116 Basis Document. Using conversion factors specified by the EPA in Limiting 

Values of Radionuclides Intake and Air Concentrations and Dose Conversion Factors for Inhalation, 

Submersion, and Ingestion (EPA 1988), annual doses to the thyroid and other critical organs are calculated 

for each relevant radionuclide on an organ-by-organ basis. The peak annual dose to the thyroid is 

approximately 6 mrem/yr compared to the 10 CFR 61.41 limit of 75 mrem/yr. The peak annual dose to 

any other organ is approximately 0.15 mrem/yr compared to the 10 CFR 61.41 limit of 25 mrem/yr. 

The NRC performance objectives of 10 CFR 61.41 also state that reasonable efforts should be 

made to maintain releases of radioactivity in effluents to the general environment to ALARA. The TFF 

PA was conducted in accordance with a similar DOE performance objective, which, as quoted in the PA, 

states “performance assessments shall include a determination that projected releases of radionuclides to 

the environment shall be maintained as low as reasonably achievable.” As described in Section 5 of this 

3116 Basis Document and the TFF PA (DOE-ID 2003b), tank closure (i.e., disposal) activities are being 

conducted in a manner that will maintain releases of radioactivity ALARA. 

The TFF PA presents a detailed description of the modeling approach and computer codes used in 

the assessment (DOE-ID 2003b) and includes an uncertainty analysis that considers best, realistic, 

conservative, and worst-case exposure scenarios. The results show that the general population is protected 

in scenarios used in the PA (DOE-ID 2003b). 
89

Radionuclides released from the INTEC TFF to the environment have the potential to impact 

humans through a number of different pathways. Radionuclide transport from TFF waste to the aquifer is 

considered in the groundwater exposure pathway.mmm Other pathways that may contribute to human 

exposure are those associated with groundwater concentrations of contaminants. Irrigation with 

contaminated groundwater may lead to contamination of agricultural crops and animals. Ingestion of 

contaminated groundwater by terrestrial animals may lead to human exposure. In addition, human 

exposure may occur from ingestion of contaminated groundwater and from consumption of contaminated 

food supplies (DOE-ID 2003b). Additional pathways, such as the atmospheric transport pathway, were 

assessed; they provide approximately 0.5 mrem/yr dose to members of the public. As a result, allpathways dose calculations are based solely on the groundwater pathways. 

As Table 20 demonstrates, the PA results for the all-pathways dose to a future member of the 

public as a function of time after closure indicate a peak dose of 0.46 mrem/yr at approximately 

14,000 years after closure. The majority of the total dose is from 99Tc. This projected dose satisfies the 

10 CFR 61.41 requirements. 

Table 20. Groundwater all-pathways dose (less than 25 mrem/yr). 

Nuclide 

Decontaminated Tank Residual 

Inventory 

(mrem/yr) 

Time of Peak Dose 

(yr) 

90Sr/90Y 2.10E−08 551 

99Tc 0.46 14,590 

129I 0.29 890 

Peak Dose 0.46 

To understand parameter sensitivity and to analyze uncertainty in the TFF PA (DOE-ID 2003b) 

groundwater model,nnn the INL Site developed worst, conservative, expected, and best-case scenarios to 

include radionuclide inventory and transport parameters. Since the radionuclide inventory in a 30,000-gal 

tank is small when compared with a 300,000-gal tank, the impact from the 30,000-gal tanks is small. Data 

from these four scenarios are used to analyze the probability of radionuclide releases to the public and the 

environment. This sensitivity/uncertainty analysis shows that radionuclide releases are not underpredicted. The conservative-case scenario is used as the compliance-case scenario in the TFF PA. These 

evaluations are documented in the TFF PA (DOE-ID 2003b). 

 

mmm. Three radionuclides contribute to the drinking water pathway dose: 129I, 99Tc, and 90Sr. The maximum dose for 

groundwater is 0.16 mrem/yr, which occurs at 890 years from 129I. (At 890 years, dose contribution from the remaining 

radionuclides is negligible.) 

nnn. Page 100 of the TER (NRC 2006b) summarizes the NRC staff’s concerns with respect to the groundwater conceptual model 

presented in the PA, which include assumptions about the principal direction of vadose zone flow, the Big Lost River boundary 

condition, and hydrological features that are thought to be present in the INTEC area but that were not explicitly accounted for in 

the hydrology model. NRC acknowledges in its TER that there is a paucity of site-specific data, including well characterization

data, which are directly relevant to the known principal direction of vadose zone flow. On page 101 of the TER (NRC 2006b), 

NRC staff conclude that the uncertainties in the hydrologic modeling are offset through consideration of multiple lines of 

evidence, including sampling data from the tank vaults that suggest the 90Sr concentrations in the sandpad are likely bounding, 

consideration of monitoring data from historical TFF releases that provide a basis from which predictive model results can be 

evaluated for all groundwater highly radioactive radionuclides, and consideration of varying levels of performance of the 

concrete vaults that have the potential to essentially eliminate any concern with short-lived radionuclide releases from the TFF. 

NRC states, “Therefore, NRC staff has confidence that overall system performance of both the engineered and natural barriers 

will be sufficient to meet performance objectives.” 
90

7.10 10 CFR 61.42, “Protection of Individuals from 

Inadvertent Intrusion” 

Provisions in 10 CFR 61.42 require: 

[D]esign, 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. 

This requirement reflects the NRC’s intent that persons inadvertently intruding into the waste be 

protected. The performance objective does not place quantitative limits on exposure. However, the EIS 

for 10 CFR 61 (NRC 1982) suggests a dose limit of 500 mrem/yr for the waste classification scheme in 

10 CFR 61.55. Consequently, NRC uses a 500-mrem dose limit for evaluating impacts to an inadvertent 

intruder for purposes of 10 CFR 61.42.ooo

The PA provides a performance objective for assessing impacts on a person inadvertently intruding 

into the disposal facility that is more stringent than the NRC requirement. The PA performance objective 

for inadvertent intruders is: 

[F]or purposes of establishing limits on the concentration of 

radionuclides that may be disposed of near-surface, the PA shall include 

an assessment of impacts calculated for a hypothetical person assumed to 

inadvertently intrude for a temporary period into the [LLW] disposal 

facility. For intruder analyses, institutional controls shall be assumed to 

be effective for at least 100 years following closure. The intruder 

analyses shall use performance measures for chronic and acute-exposure 

scenarios, 500 mrem (5 mSv).ppp,qqq 

The intruder scenarios and associated doses for the INTEC TFF are evaluated in the PA 

(DOE-ID 2003b). A site-specific radiological PA is used to verify that requirements for protecting the 

inadvertent intruder were satisfied. The environmental setting and the logic for using site-specific 

scenarios is explained in detail in the PA. As described in the PA, several intruder scenarios were 

evaluated for applicability to the INTEC TFF closure. However, the TFF is a unique environment for 

intruder scenario evaluations (DOE-ID 2003b). Many of the standard scenarios are not considered 

applicable to the TFF because the depth of the waste in the tanks is greater than 10 m (33 ft). Therefore, 

only the intruder-drilling and post-drilling scenarios are considered applicable for the tank and sandpad 

contamination. Additional scenarios were evaluated for the piping inventory located less than 3 m (10 ft) 

below the surface. These scenarios include the intruder-construction scenario and the intruder 

post-construction scenario. 

 

ooo. NRC guidance does not provide for including the assessment of impacts to inadvertent intruders as a part of the PA to 

provide reasonable assurance that the dose limit in 10 CFR 61.41 will not be exceeded. Rather, the site must show that it is 

following the precepts of the waste classification scheme (i.e., stabilizing Class B waste and providing an intruder barrier for 

Class C waste) (NRC 1987a, 1987b). 

ppp. The NRC waste classification system is based on intruder calculations using a 500-mrem dose limit. The TFF PA meets the 

more stringent DOE guidance for protection of the inadvertent intruder. 

qqq. The PA uses active institutional controls. Passive controls are not considered, although such controls may be adopted in the 

future as site closure designs mature. This plan to use active controls without passive controls applies to the TFF only. Both 

active and passive controls may be used at other sites. 
91

The groundwater pathway was not included in the intruder scenarios because the groundwater 

pathways dose assessment was performed as part of compliance for protection of the general population 

from releases of radioactivity. However, if the groundwater all-pathways dose is included with the 

chronic intruder-drilling scenario it adds 1.35 mrem/yr to the 91.1 mrem/yr if the PA dose assessment is 

used. The total dose would increase to 92.5 mrem/yr or an increase of 1.5%. If the inventory after 

cleaning is used, the all-pathways dose is estimated at 0.46 mrem per year and the chronic 

intruder-drilling scenario adds 25 mrem/yr. The total dose would increase to 25.46 mrem/yr or an increase 

of 1.8%. These doses remain well below 500 mrem/yr. The groundwater pathway was prepared to 

estimate the radiation dose at the location and time that provided the greatest dose to the receptor. The 

time and location of the greatest drinking water dose (890 years) and location (600 m [1,969 ft]) south of 

the TFF do not correspond to the time or location for intruder scenarios. The intruder scenario was located 

directly over the TFF at 100 years after closure, or 2112. The time period for the intruder scenario was 

chosen because it predicts the greatest dose to intruders from those radionuclides that contribute the 

majority of the dose. The radionuclide 137Cs and its daughter 137Ba contribute over 80% of the dose. 137Cs 

has a half-life of 30.17 years; therefore, the radiation dose from 137Cs approaches 0 at approximately 250 

years. In contrast to the intruder scenario, the groundwater pathway increases to the maximum dose at a 

time 890 years in the future (Portage 2005g). 

For acute exposures, three potential intruder-exposure scenarios are considered: 

intruder-construction, intruder-discovery, and intruder-drilling scenarios. The intruder-discovery scenario 

is evaluated to analyze exposures from unearthing equipment and piping nearer than 3 m (10 ft) to the 

surface. The intruder is assumed to recognize that he or she is digging into very unusual soil immediately 

upon encountering the vault/tank/piping system and leaves the site. The intruder-discovery scenario is not 

considered for further analysis because the exposure time is low compared with exposures from the 

intruder-construction scenario. In the intruder-construction scenario, the intruder is assumed to dig a 

basement. This scenario is retained for evaluation because of contaminated TFF piping that is less than 

3 m (10 ft) below the surface. The maximum dose after institutional control (i.e., 100 years) for the acute 

intruder-construction scenario is 0.23 mrem. The tank inventory used for the intruder construction 

scenario is Tank WM-182 because it has the greatest inventory of the cleaned tanks. This total dose is 

well within the performance objective of 500 mrem for all times after closure. Table 21 shows the major 

radionuclide contributors and associated dose. 

Table 21. Acute intruder-construction dose (less than 500 mrem). 

Nuclide 

Dose Based on Residual Inventory at Closure

(mrem) 

90Sr/90Y 0.00005 

94Nb 0.001 

137Cs/137mBa 0.20 

237Np 0.0001 

238Pu 0.009 

239Pu 0.006 

240Pu 0.003 

241Am 0.001 

Total Dose 0.23 

The credible acute intruder-drilling scenario assumes short-term exposure of a hypothetical intruder 

to drill-cuttings from a borehole that has penetrated the closed facility and associated residuals. The 

intruder-drilling scenario takes no credit for the thick grout encapsulating the waste but assumes that 
92

concentrated residuals are spread over a 0.5-acre lot. The exposure pathways analyzed are inhalation of 

resuspended drill cuttings, external exposure to the ground source, and inadvertent soil ingestion. The 

acute scenario dose for the limiting tank (represented by the conservative tank residual inventory for Tank 

WM-185) is approximately 152 mrem at 100 years after closure. Tank WM-185 is used for the chronic 

drilling scenario because the tank system includes a contaminated sandpad. The drilling scenario assumes 

the intruder will drill directly through the tank residual and sandpad. 

This total dose is well within the limit of 500-mrem TEDE.rrr Table 22 lists the major radionuclides 

contributing to this dose. 

Table 22. Acute intruder-drilling dose (less than 500 mrem) (Portage 2005g). 

Nuclide 

Dose Based on Residual Inventory at Closure

(mrem) 

90Sr/90Y 0.3 

94Nb 0.14 

137Cs/137mBa 108 

237Np 0.07 

238Pu 7.71 

239Pu 15.1 

240Pu 2.82 

241Am 17.7 

Total Dose 152 

For chronic exposures, five intruder scenarios are investigated: intruder post-construction, 

intruder-resident, intruder-radon, biointrusion, and post-drilling. Again, because of the depth of the final 

waste form in the tank bottom and piping, intruder post-construction and post-drilling exposure are the 

only credible scenarios. The chronic intruder post-construction scenario is an extension of the acute 

intruder-construction scenario. This scenario assumes that an intruder lives in the building constructed as 

part of the intruder-construction scenario and engages in agricultural activities on the contaminated site. 

The intruder is exposed to contamination by inhalation of resuspended contaminated soil, inhalation of 

 

rrr. In its RAI Comment 16 (NRC 2006a), NRC noted that intruder analysis assumptions regarding the depth and area of 

contamination significantly impact the resulting dose for the acute or chronic well intruder scenarios and requested that DOE 

conduct additional sensitivity analysis. In DOE’s response to NRC RAI Comment 16 (ICP 2006l), an intruder sensitivity 

analysis was conducted for the acute intruder-drilling scenario. The acute intruder-drilling scenario assumed that the cuttings

from a 0.6-m (22-in.) irrigation well were placed in a 3- by 3-m (10- by 10-ft) pit during drilling operations. The intruder was 

assumed to stand next to the pit for the entire drilling duration of 160 hours. The residuals from the tanks and sandpads would

initially be exposed at the surface of the pit, and then increasing thicknesses of clean material would be placed on top of the

contaminated material. On page 107 of its TER (NRC 2006b), NRC notes that the results show the external dose from 137Cs 

could be a factor of two higher with this alternative model and a factor of three higher considering the uncertainty in the sandpad 

inventory. NRC asserts that considering both of these factors, the intruder dose could be higher than the performance objective of 

5 mSv/yr (500 mrem/yr). Although NRC notes that this would only occur under very pessimistic conditions, it did not provide its

calculations to confirm its assertion. DOE has confirmed that results of its sensitivity analysis for the acute-intruder drilling 

scenario presented in NRC RAI Comment 16 (NRC 2006a) demonstrate that the alternative model along with the highest 

predicted sandpad inventory provide a maximum dose of 307 mrem. This 307 mrem value includes both the contribution from 

the highest sandpad inventory and the alternative pit model. On page 108 of the TER (NRC 2006b), the NRC staff notes “With 

the additional sensitivity analyses, DOE Idaho developed reasonable intruder scenarios to evaluate protection of inadvertent 

intruders and demonstrated that performance objectives found in 10 CFR 61.42, ‘Protection of Individuals from Inadvertent 

Intrusion,’ could be achieved...All intruder doses are calculated to be less than 5 mSv/yr (500 mrem/yr), except under very 

pessimistic conditions.” 
93

gaseous radionuclides released from the waste, external irradiation, ingestion of contaminated soil, 

ingestion of contaminated beef and milk, and ingestion of contaminated vegetables. The maximum dose is 

3.15 mrem/yr for the chronic intruder post-construction scenario at 100 years post-closure. This result is 

well below the performance objective of 500 mrem/yr. Table 23 shows the major radionuclide 

contributors to the total chronic intruder dose (Portage 2005g). 

Table 23. Chronic intruder-construction dose (Portage 2005g). 

Nuclide 

Dose Based on Residual Inventory at Closure

(mrem/yr) 

90Sr/90Y 4.44E–02 

94Nb 1.55E–02 

137Cs/137mBa 2.88E+00 

237Np 2.58E–03 

238Pu 1.02E–01 

239Pu 6.30E–02 

240Pu 2.97E–02 

241Am 1.37E–02 

Total Dose 3.15E+00 

The chronic post-drilling scenario is a logical extension of the drilling scenario considered for 

acute exposures. This scenario assumes that the intruder occupies the site after drilling a water well and 

that the intruder grows crops on a mixture of clean soil and contaminated drill cuttings. Exposures from 

inhaling dust from farming activities, ingesting vegetables grown in the soil, and ingesting products from 

animals that have eaten contaminated feed are considered. The chronic scenario dose for Tank WM-185 is 

approximately 25 mrem/yr at 100 years after closure, as shown in Table 24. Tank WM-185 is used for the 

chronic drilling scenario because the tank system includes a contaminated sandpad. This scenario includes 

drilling through the tank residual and sandpad. The short-lived 137Cs/137mBa provides 85% of the total 

dose. Chronic intruder doses are reduced dramatically with time because of the decay of 137Cs 

(Portage 2005g). 

Table 24. Chronic intruder-drilling dose comparison (Portage 2005g). 

Nuclide 

Dose Based on Residual Inventory at Closure

(mrem/yr) 

90Sr/90Y 1.69 

94Nb 0.03 

137Cs/137mBa 21 

237Np 0.007 

238Pu 0.33 

239Pu 0.61 

240Pu 0.12 

241Am 0.75 

Total Dose 25 

The uncertainty in the intruder analyses is largely based on the assumptions and parameters 

selected for the mathematical modeling. However, every attempt was made while running the models to 

consider the site-specific environment and habits of the people currently in the region. Therefore, the 
94

analyses are considered to be representative of the INL Site region. The predicted doses are considered to 

be conservative, and there is a reasonable assurance that the performance objectives will not be exceeded 

for the 1,000-year post-closure period.sss

7.11 10 CFR 61.43, “Protection of Individuals during Operations” 

Requirements in 10 CFR 61.43 state: 

[O]perations at the land disposal facility must be conducted in 

compliance with the standards for radiation protection set out in part 20 

of this chapter [10 CFR], 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, “Standards for Protection Against Radiation,” which 

contains radiological protection standards for workers and the public. The DOE requirements for 

occupational radiological protection are provided in 10 CFR 835, “Occupational Radiation Protection,” 

and those for radiological protection of the public and the environment are provided in DOE Order 

5400.5, “Radiation Protection of the Public and the Environment” (1993). 

The closure activities at the TFF are maintained under the DOE dose limits for the worker, public, 

and environment. The TFF PA does not address the operational (i.e., closure and disposal activities) of the 

facility, only the post-closure (i.e., post-disposal) aspects of the facility. However, during Tank WM-182 

cleaning operations, the total radiation exposure to the 23 workers was about 650 mrem for an average 

exposure of about 30 mrem per individual. This information is based upon a review of dosimetry results 

from TFF radiation work permits for the January 2002 to June 2005 timeframe (Martin 2005), and the 

following is summarized as further explained in Appendix B-4: 

• The average radiation exposure per tank that will be experienced for cleaning each TFF tank is 

expected to total about 650 mrem for all occupational exposure 

• The average exposure per person for cleaning a TFF tank will be about 650 mrem divided by 

23 people, which is about 30 mrem per person 

 

sss. In its RAI Comment 16 (NRC 2006a), NRC noted that intruder analysis assumptions regarding the depth and area of 

contamination significantly impact the resulting dose for the acute or chronic well intruder scenarios and requested that DOE 

conduct additional sensitivity analysis. In DOE’s response to NRC RAI Comment 16 (ICP 2006l), an intruder sensitivity 

analysis was conducted for the chronic intruder-drilling scenario. Three parameters were varied: well diameter (25, 20, and 15 cm 

[10, 8, and 6 in.]), contaminant spreading area (1,100, 1,600, and 2,200 m2 [0.3, 0.4, and 0.5 acres]), and the tilling depth (30, 46, 

and 61 cm [12, 18 and 24 in.]). On page 107 of its TER (NRC 2006b), NRC notes the results of the sensitivity analysis show that

the dose of 0.21 mSv/yr (21 mrem/yr) reported in the PA could be a factor of three higher with a large diameter well (25 cm 

[10 in.] versus 15 cm [6 in.]), a factor of two higher with a smaller area of contamination (1,100 m2 [0.3 acres]), a factor of two 

higher considering the highest inventory in the sandpad, and a factor of two higher considering a factor of two reduction in the 

tilling depth. NRC states that all of these factors considered would lead to a maximum dose just over the 500-mrem/yr 

performance objective. NRC concludes that only under very pessimistic circumstances could the dose to an inadvertent intruder 

approach the performance objective of 500 mrem/yr for the chronic well drilling scenario. On page 108 of the TER 

(NRC 2006b), the NRC staff notes “With the additional sensitivity analyses, DOE Idaho developed reasonable intruder scenarios 

to evaluate protection of inadvertent intruders and demonstrated that performance objectives found in 10 CFR 61.42, ‘Protection

of Individuals from Inadvertent Intrusion,’ could be achieved...All intruder doses are calculated to be less than 5 mSv/yr 

(500 mrem/yr), except under very pessimistic conditions.” 
95

• The maximum radiation exposure for an individual worker is estimated to be 120 mrem for 

cleaning a single TFF tank. 

Worker dose for tank cleaning is minimal because all cleaning is accomplished remotely. Worker 

exposures would be limited to equipment installation and operation and maintenance activities on 

contaminated equipment. Worker exposure per tank is estimated to total approximately 650 mrem for 

23 workers (30 mrem per worker), which results in a total exposure of about 7.15 rem for cleaning eleven 

300,000-gal tanks. Therefore, based on estimates that future tank cleaning will probably be accomplished 

in 1 year and that minimal exposure is expected during grouting (INEEL 1998), an estimated worker dose 

of 30 mrem/yr is reasonable. 

Consistent with Section 3116(a), the cross-referenced “standards for radiation protection” in 

10 CFR 20 that are considered in detail in this 3116 Basis Document are the dose limits for the public and 

the workers during disposal operations set forth in 10 CFR 20.1101(d), 20.1201(a)(1)(i), 

20.1201(a)(1)(ii), 20.1201(a)(2)(i), 20.1201(a)(2)(ii), 20.1201(e), 20.1208(a), 20.1301(a)(1), 

20.1301(a)(2), and 20.1301(b).ttt These dose limits correspond to the dose limits in 10 CFR 835 and 

relevant DOE orders that establish DOE regulatory and contractual requirements for DOE facilities and 

activities. The following subsections show that closure operations of the TFF meet these dose limits and 

that doses will be maintained ALARA.uuu

7.11.1 Air Emissions Limit for Individual Member of the Public [10 CFR 20.1101(d)] 

The NRC regulation in 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)/yr from these emissions. 

 

ttt. The introductory “notwithstanding” phrase to Section 3116 makes it clear that the provisions of 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, 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 Subsections (a)(1)–(3) of Section 3116 are 

those that 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 (NDAA), also 

confirms that 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 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 Section 3116. As such, the “standards for radiation protection”

in 10 CFR 20 (as cross-referenced in the performance objective in 10 CFR 61.43), which are relevant in the context of Section 

3116, are the dose limits for radiation protection of the public and the workers during disposal operations, and not those which 

address general licensing, administrative, programmatic, or enforcement matters administered by NRC for NRC licensees. 

Accordingly, this 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 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 TFF. Therefore, this limit is not discussed further in this 3116 Basis Document. Likewise, 10 CFR 20.1207 specifies 

occupational dose limits for minors. However, there will not be minors working at TFF who will receive an occupational dose. 

Therefore, this limit is not discussed further in this 3116 Basis Document. 

uuu. In addition, 10 CFR 835, like 10 CFR 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, and 

training. 
96

The DOE similarly limits doses from air emissions to the public to 10 mrem/yr in DOE Order 

5400.5 (1993). The DOE is also subject to and complies with the EPA’s requirement in 40 CFR 61.92, 

which has the same limit.vvv The estimated dose per year from airborne emissions to the maximally 

exposed individual member of the public resulting from closure/disposal activities is 0.51 mrem/yr 

(DOE-ID 2003b). 

7.11.2 Total Effective Dose Equivalent Limit for Adult Workers 

[10 CFR 20.1201(a)(1)(i)] 

The NRC regulation in 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’s regulation in 10 CFR 835.202(a)(1) has the same annual dose limit for the annual 

occupational dose to general employees.www For the occupational dose to adults during closure operations 

at the TFF, the TEDE per year will be controlled ALARA below 5 rem, as shown in Appendix B-4. 

7.11.3 Any Individual Organ or Tissue Dose Limit for Adult Workers 

[10 CFR 20.1201(a)(1)(ii)] 

The NRC regulation in 10 CFR 20.1201(a), concerning occupational dose limits for adults, 

provides in relevant part: 

(a) [C]ontrol the occupational dose to individual adults, except for 

planned special exposures...to the following dose limits. 

(1) An annual limit, which is the more limiting of –... 

(ii) The sum of the deep-dose equivalent and the committed dose 

equivalent to any individual organ or tissue other than the lens of 

the eye being equal to 50 rems (0.5 Sv). 

 

vvv. 40 CFR 61.92 provides 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. It is 

assumed that the individual is an adult living at the site perimeter that is exposed to the maximum yearly radioactive atmospheric 

release and maximum radiation concentration in food for 365 days per year. For the airborne pathway, the dose is developed by 

the input of atmospheric release data, vegetation consumption data, milk consumption data, and beef consumption data.” 

www. The DOE’s regulation requires that the occupational dose per year for general employees shall not exceed both a TEDE of 

5 rem and the sum of the deep-dose equivalent for external exposures and the committed dose equivalent to any other organ or 

tissue other than the lens of the eye of 50 rem. The NRC’s regulation specifies that either of these two limits shall be met by NRC 

licensees, whichever is more limiting. This 3116 Basis Document shows that DOE will meet the more stringent of the dose limits 

in 10 CFR 835 and the relevant dose limits in 10 CFR 20. Because DOE imposes stricter, separate requirements, the provisions of

10 CFR 20.1201(a)(1) and (a)(2), which correlate to 10 CFR 835.202(a)(1) and (a)(2), are discussed in separate subsections in 

this 3116 Basis Document. 
97

The dose limit specified in 10 CFR 20.1201(a)(1)(ii) is the same as that specified in 

10 CFR 835.202(a)(2). For the occupational dose to adults during disposal/closure operations at TFF, 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 INL 

“Radiological Control Manual” (PRD-183, 2002) provides that the design basis annual occupational 

exposure limits for any organ or tissue other than the eye cannot exceed 50 rem/yr, which is the same as 

the NRC limit. Furthermore, TFF disposal/closure operations will consist predominantly of remote 

cleaning technologies to minimize hands-on work and associated exposure. 

7.11.4 Annual Dose Limit to the Lens of the Eye for Adult Workers 

[10 CFR 20.1201(a)(2)(i)] 

The NRC regulation in 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 exposure the following dose limits. ... 

(2) The annual limits to the lens of the eye, to the skin of the whole 

body or 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 DOE’s 

regulation in 10 CFR 835.202(a)(3). For the occupational dose to adults during closure operations at the 

TFF, the annual dose limit to the lens of the eye will be controlled to ALARA below a maximum of 

15 rem/yr. Furthermore, disposal/closure operations will consist predominantly of remote cleaning 

technologies to minimize hands-on work and associated exposure. 

7.11.5 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 in 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 in 10 CFR 835.202(a)(4). For the occupational dose to adults during closure operations at the 

TFF, which involve little hands-on activity, the annual dose limit to the skin of the whole body or to the 

skin of any extremity will be controlled to ALARA below a shallow-dose equivalent of 50 rem/yr. 

Furthermore, disposal/closure operations will consist predominantly of remote cleaning technologies to 

minimize hands-on work and associated exposure. 
98

7.11.6 Limit on Soluble Uranium Intake [10 CFR 20.1201(e)] 

The NRC regulation in 10 CFR 20.1201(e), concerning occupational dose limits for adults, 

provides in relevant part: “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.” 

Requirements in DOE Order 440.1A (1998) for soluble uranium intake are the more restrictive of 

the concentrations in the American Conference of Governmental Industrial Hygienists Threshold Limit 

Values (0.2 mg/m3

, which is the same as noted in 10 CFR 20, Appendix B) or the Occupational Safety 

and Health Administration permissible exposure limit (0.05 mg/m3

) (DOE O 440.1A, 1998). The 

permissible exposure limit for soluble uranium, which equates to a soluble uranium intake of 

2.4 mg/week, is the more restrictive of the two. Therefore, this limit is imposed for TFF closure 

operations. Furthermore, disposal/closure operations will consist predominantly of remote cleaning 

technologies to minimize hands-on work and associated exposure. 

7.11.7 Dose Equivalent to an Embryo/Fetus [10 CFR 20.1208(a)] 

The NRC regulation in 10 CFR 20.1208(a), concerning the dose equivalent to an embryo/fetus, 

provides in relevant part: “ensure 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’s regulation in 10 CFR 835.206(a) has the same dose limit. For the occupational dose to 

an embryo/fetus during closure operations at the TFF, doses will be controlled so that the dose equivalent 

to the embryo/fetus during the entire pregnancy for a declared pregnant worker will not exceed 0.5 rem. 

Furthermore, after declaration of pregnancy, DOE provides the option of a mutually agreeable assignment 

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 track exposure carefully. 

7.11.8 Dose Limits for Individual Members of the Public [10 CFR 20.1301(a)(1)] 

The NRC regulation in 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) (1993) similarly limit public doses to less than 

100 mrem/yr. However, DOE’s application of the limit is more restrictive in that DOE is required to make 

a reasonable effort to ensure that 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 closure 

operations at the TFF, the TEDE limit to an individual member of the public will be controlled to less 

than 0.1 rem/yr. The dose to the maximally exposed individual outside the INL Site boundary is 
99

calculated each year and shown less than 100 mrem/yr. The latest annual report shows the dose to be 

0.035 mrem/yr for Calendar Year 2003 (DOE-ID 2004b). 

7.11.9 Dose Limits for Individual Members of the Public [10 CFR 20.1301(a)(2)] 

The NRC regulation in 10 CFR 20.1301(a), concerning dose limits for individual members of the 

public, provides in relevant part: 

(a) [C]onduct operations so that – ... 

(2) The dose in any unrestricted area from external sources, 

exclusive of the dose contributions from patients administered 

radioactive material and released ..., does not exceed 0.002 rem 

(0.02 millisievert) in any one hour. 

The DOE’s regulation in 10 CFR 835.602 establishes the expectation that the TEDE in controlled 

areas will be less than 0.1 rem/yr. To ensure that these dose limits are met, the following measures have 

been instituted within controlled areas. Per 10 CFR 835.603(g), radioactive materials areas have been 

established for accumulations of radioactive material that could result in a radiation dose of 100 mrem/yr 

or greater. Averaged over a work year, this yields a constant dose rate of 0.00005 rem/hr as compared to 

the NRC limit of 0.002 rem in any one hour. In addition, training and dosimetry are required for 

individual members of the public for entry into controlled areas. 

7.11.10 Dose Limits for Individual Members of the Public [10 CFR 20.1301(b)] 

The NRC regulation in 10 CFR 20.1301(b), concerning dose limits for individual members of the 

public, provides in relevant part: “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’s regulation in 10 CFR 835.208 has the same dose limit. The TEDE limit to an individual 

member of the public granted access to controlled areasxxx during disposal operations at TFF will be 

controlled to 0.1 rem/yr. Furthermore, training is required for individual members of the public for entry 

into controlled areas. In addition, to ensure no member of the public exceeds radiation exposure limits, 

use of dosimetry is required if a member of the public is expected to enter a controlled area and receive a 

dose that may exceed 0.05 rem/yr.yyy

7.11.11 As Low As Reasonably Achievable (10 CFR 20.1003) 

The NRC regulation in 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.” 

 

xxx. 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 definition is the same as that in 10 CFR 835.2 for 

a controlled area. 

yyy. 10 CFR 20.1301(d) allows licensees to request NRC authorization to allow an individual member of the public to receive up 

to an annual dose limit of 0.5 rem. 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. 
100

The DOE has a similar requirement, and DOE’s regulation in 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 is reasonable.” For radiological work activities 

during closure operations at the TFF, 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, DOE’s regulation in 10 CFR 835.101(c) requires the contents of each radiation 

protection program (RPP) to include formal plans and measures for applying the ALARA process to 

occupational exposure. 

The ALARA process provides for rigorous reviews. Worker dose reduction is achieved through 

reviews of potential worker exposures against established criteria and then applying appropriate 

radiological engineering measures, including: 

• Inclusion of radiological control points 

• Use of work processes and special tools to reduce exposure time 

• Use of engineered controls, including fixatives or covering work areas 

• Specification of special radiological training or unique monitoring requirements 

• Use of mockup training for first of a kind, high exposure, or complex tasks 

• Engineering design and use of temporary shielding to reduce radiation levels and time in radiation 

fields 

• Walk-down or dry run of activity using applicable procedures 

• Staging and prefabrication of necessary materials and special tools 

• Maximization of prefabrication and shop work 

• Provisions for waste minimization. 

The institutionalized application of the ALARA program for TFF activities contributed to the low 

doses achieved for previous tank work as described in Appendix B-4. 

7.11.12 Reasonable Assurance 

Measures that provide reasonable assurance that closure operations at the TFF will comply with the 

applicable dose limits and with the ALARA provisions include: (1) the documented RPP; (2) the safety 

analysis; (3) design; (4) regulatory and contractual enforcement mechanisms; and (5) access controls, 

training, and dosimetry. These measures are discussed in the following subsections. 

7.11.12.1 INL Radiation Protection Program. The DOE regulates occupational radiation 

exposure at its facilities through 10 CFR 835, “Occupational Radiation Protection,” which establishes 

exposure limits and other requirements to ensure that 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 are nuclear safety requirements, which if 

violated, provide a basis for the assessment of civil penalties under the Section 234A of the Atomic 

Energy Act (42 USC 2011 et seq. 1954).
101

Pursuant to 10 CFR 835, activities at the INL Site, including closure operations at the TFF, must be 

conducted in compliance with the documented RPP for the INL Site as approved by DOE (PLN-260, 

2003). The key elements of the RPP include monitoring of individuals and work areas, control of access 

to areas containing radiation and radioactive materials, use of warning signs and labels, methods to 

control the spread of radioactive contamination, radiation safety training, objectives for the design of 

facilities, criteria for levels of radiation and radioactive material in the workplace, 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 RPP 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. 

7.11.12.2 Documented Safety Analysis. The DOE has approved a safety analysis report for 

operating the TFF in accordance with 10 CFR 830. The “Safety Analysis Report for the Tank Farm 

Facilities” (SAR-107, 2004) identifies, classifies, and evaluates hazards associated with the closure 

project. The first step in developing the safety analysis report was a formal hazards analysis to evaluate 

the potential risk of operations to the workers and the public. A group of subject-matter experts with 

expertise in operations, engineering, industrial hygiene, radiological protection, environmental 

compliance, and maintenance performed the analysis.

The hazards analysis consisted of three basis phases: hazard identification, hazard classification, 

and hazard evaluation. During the hazard identification phase, all possible radiological and chemical 

hazardous materials associated with the normal and abnormal operations of the facility were identified, 

along with all potential energy sources available to disperse the hazardous materials to the environment. 

During the hazard classification phase, the maximum quantities of hazardous materials possible in 

the TFF were evaluated against the criteria listed in DOE-STD-1027-92, “Hazard Categorization and 

Accident Analysis Techniques for Compliance with DOE Order 5480.23, Nuclear Safety Analysis 

Reports” (1997), to determine the overall hazard classification of the facility. The team determined that 

the hazard classification of the TFF was Hazard Category 2. This classification denotes a potential for 

significant onsite consequences. This facility has the potential for release of sufficient quantities of 

hazardous material, which would require onsite emergency planning activities. 

During the third and final phase of the hazard analysis, all possible normal and abnormal 

operational events that could result in exposing facility workers or the public to hazardous material were 

evaluated to determine the magnitude of the risk. During this hazard evaluation phase, the consequence 

and frequency of each operational event were determined qualitatively, and the resulting level of risk was 

identified. The purpose of identifying the level of risk was to determine which operational events posed 

some level of risk (and thus required additional evaluation) and which events presented negligible risk to 

the facility workers and public. 

The safety analysis report analyzed the hazards that were identified in the hazards analysis that 

could impact facility workers during normal operations and accident conditions, and specifically included 

radiation exposure hazards. The report identified the basis for derivation of the TFF technical safety 

requirements and also provided summary descriptions of the key features of safety management programs 

at the INL Site as they pertain to the TFF (see Appendix I). 
102

7.11.12.3 Regulatory and Contractual Enforcement. Any violation of the requirements in 

10 CFR 835 is subject to civil penalties pursuant to Section 234A of the Atomic Energy Act of 1954 

(42 USC 2011 et seq. 1954), as implemented by DOE regulations in 10 CFR 820. In addition, the 

requirements in 10 CFR 835 and all applicable DOE orders are incorporated into all contracts with DOE 

contractors, the DOE contractor for closure operations at the TFF, as well as other operations at the INL 

Site. The DOE enforces these contractual requirements through contract enforcement measures, including 

the reduction of contract fees.

7.11.12.4 Access Controls, Training, Dosimetry, and Monitoring. Training or an escort is 

required for individual members of the public to enter controlled areas. In addition, dosimetry is required 

if a member of the public is expected to enter a controlled area and exceed a dose of 0.05 rem/yr to ensure 

radiation exposure limits are not exceeded for any member of the public.

Also, worker radiation exposure monitoring is performed for all workers expected to receive 

100 mrem/yr from internal and external sources of radiation to provide that no worker exceeds radiation 

exposure limits and that all radiation doses are maintained as far below the limits as is reasonably 

achievable.zzz

7.12 10 CFR 61.44, “Long-Term Stability of Disposal Site” 

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. 

As discussed previously, several of the tanks in the TFF have been removed from service and 

cleaned in preparation for grouting and final closure of the TFF. Once the remaining SBW has been 

transferred from the tanks for treatment and disposal outside the State of Idaho, the remainder of the tanks 

will be cleaned and closed. Cleaning and closure of the TFF is planned to be completed by 2012. As 

shown below, disposal of the stabilized residuals in the TFF, and TFF tank system, in situ will meet the 

performance objectives of 10 CFR 61.44 for long-term stability. 

7.12.1 Siting 

A comprehensive review of site geology, seismology, hydrology, demography, meteorology, and 

environmental setting are presented in Section 2 of the TFF PA (DOE-ID 2003b), which is briefly 

summarized as follows. The INL is located in southeastern Idaho, on the west-central portion of the 

Eastern Snake River Plain (ESRP). The nearest INL boundaries are 35 km (22 mi) west of Idaho Falls, 

37 km (23 mi) northwest of Blackfoot, 71 km (44 mi) northwest of Pocatello, and 11 km (7 mi) east of 

Arco. There are no permanent residents within an 18-km (11-mi) radius of INTEC. The Snake River Plain 

Aquifer is approximately 122 m (400 ft) below ground surface at INTEC. 

The principal surface materials at the INL are basalt, alluvium, lakebed sediments, volcanic rocks, 

and sedimentary rocks. The natural plant life consists mainly of sagebrush and various grasses. The most 

 

zzz. On page 112 of its TER (NRC 2006b) NRC concluded, “...there is reasonable assurance that DOE Idaho can meet 

10 CFR 61.43 requirements for protection of individuals during operations.” 
103

prominent ground cover is a mixture of sagebrush and a variety of grasses. The soil at the INL TFF site is 

previously disturbed sandy gravel; the flat terrain precludes erosion. 

The seismically active intermountain seismic belt surrounds the ESRP. The intermountain seismic 

belt is a zone of concentrated seismicity, which extends from northwestern Montana to eastern Idaho. The 

ESRP is not a part of the intermountain seismic belt, and therefore, is not seismically active. The primary 

seismic hazards from earthquakes to INL Site facilities consist of the effects from ground shaking and 

surface deformation. Based on the seismic history and geologic conditions, earthquakes greater than a 

magnitude of 5.5 and associated strong ground shaking and surface fault ruptures are not likely to occur 

within the ESRP (DOE-ID 2003b). 

A large earthquake, in the vicinity of the INL but outside the ESRP, occurred in 1983 and had a 

surface-wave magnitude of 7.3. The epicenter for this event was located approximately 89–97 km 

(55–60 mi) from the INL Site. Although the earthquake ground motions were felt at the INL Site, only 

minor non-nuclear building damage occurred in the form of hairline cracks and settlement. The INTEC 

facility did not experience structural failures or waste spills as a result of the earthquake and waste storage 

facilities did not show evidence of permanent movement or resulting damage. Peak ground accelerations 

ranging from 0.022 to 0.078 g were recorded at several INL Site facility areas. The largest earthquake in 

the region occurred in, 1959, at Hebgen Lake, Montana, located approximately 193 km (120 mi) northeast 

of the INL. The event had a surface-wave magnitude of 7.5; it was felt at the INL Site, but caused no 

damage (DOE-ID 2003b). 

Volcanic hazards include the effects of lava flows, fissures, uplift, subsidence, volcanic 

earthquakes, and ash flows or airborne ash deposits. Most of the volcanic activity occurred from 4 million 

to 2,100 years ago in the INL Site area. The most recent and closest volcanic eruption occurred at the 

Craters of the Moon National Monument, 46 km (27 mi) southwest of INTEC. Volcanic activity is a 

result of large volumes of magma or hotspots relatively near the Earth’s surface. Because the hotspot is 

now situated beneath Yellowstone National Park and continues to move away from the INL Site, 

recurrence of volcanic activity in the INL Site area is unlikely (DOE-ID 2003b). 

The Big Lost River enters the INL Site near the southern end from the west. During exceptionally 

wet years, the Big Lost River flows in a large arc north to the foot of the Lemhi Mountain Range, where it 

ends in a series of playas (sinks). The northwest boundary of INTEC is closest to the Big Lost River 

channel, approximately 61 m (200 ft) away. The Big Lost River is the principle natural surface water 

feature on the INL and the only stream with potential impacts to the TFF. 

A record snow pack occurred in the Big Lost River Basin in the winter of 1964/1965. This flood is 

significant because it exhibited the largest crest and largest water volume to be discharged onto the INL 

Site in 65 years of record, yet caused no damage to INL Site facilities. In a review of the historical 

information, no flooding or inundation from storms or runoff has caused flooding of the INTEC site. To 

evaluate the potentially higher infiltration rates at the TFF, modeling was conducted to simulate a 

500-year flood event of the Big Lost River. The analysis performed as part of the PA sensitivity analysis 

indicated there was no significant increase to the groundwater receptor based on these high transient 

infiltration rates. 

7.12.2 Design 

As described in Section 2 of this 3116 Basis Document, the TFF design provides for eleven 

underground 300,000-gal stainless tanks surrounded by concrete vault structures. These tanks have a 

right-cylinder shape, and are approximately 15.2 m (50 ft) in diameter and approximately 6.4–7.0 m 

(21–23 ft) in height. The bottom and lower half of the tank walls are constructed of 0.8-cm (0.3-in.) thick 
104

Type 304L stainless steel with two tanks constructed of Type 307 stainless steel. The upper half to the 

tank wall and domes are constructed of 0.63-cm (0.25-in.) thick stainless steel. Stainless steel piping of 

approximately 5 cm (2 in.) in diameter is in place between the tanks to allow transfers to and from 

individual tanks and process facilities. 

The concrete vaults are of three different designs. The vaults surrounding Tanks WM-180 and 

WM-181 are approximately 0.3 m (1 ft) thick, including a 0.6-m (2-ft) thick vault floor. An 

approximately 15-cm (6-in.) thick layer of sand was placed between the bottom of the tanks and the vault 

floor to provide cushioning for the tank. The vaults that surround Tanks WM-182 through WM-186 are of 

different construction. The concrete walls are approximately 0.3 m (1 ft) in thickness and the floor is 

approximately 0.6 m (2 ft) in thickness. The vaults surrounding Tanks WM-187 through WM-190 have 

walls that are 1 m (3 ft) thick and a floor that is 0.8 m (2.5 ft) thick. A minimum space of approximately 

0.5–1.2 m (1.5–4 ft) between the outer tank walls and the inside of the vault wall is referred to as the tank 

annulus. The space and volume of the annulus surrounding the tanks varies because of varied vault 

construction. 

The four 30,000-gal tanks are constructed much differently. These tanks are horizontal cylinders 

(“cigar-shaped”) approximately 3.5 m (11.5 ft) in diameter and 11.6 m (38 ft) in length. The tanks were 

made of Type 316L stainless steel (thicker plate than the 300,000-gal tanks (1.7 cm [0.68-in.] cylinder 

and 0.27-cm [0.5-in.] heads), and had extra plate installed over the weld seams to protect them from 

corrosion. These tanks are not surrounded by a vault structure. Each of the 30,000-gal tanks is buried 

approximately 3 m (10 ft) underground and rests on a curbed concrete pad. Stainless steel piping of 

approximately 5 cm (2 in.) diameter is in place between the tanks allow transfers from individual tanks 

and to process facilities. 

7.12.3 Use/Operation 

The TFF is used to store a variety of wastes as described in Section 2 of this 3116 Basis Document. 

These wastes include both radioactive constituents and constituents that are considered hazardous under 

RCRA (42 USC 6901 et seq. 1976). As such, the TFF is operated regulated by the State of Idaho as an 

interim status unit under HWMA (State of Idaho 1983) regulations. As described in Section 2, seven of 

the 300,000-gal tanks and the four 30,000-gal tanks have been removed from operational service and 

cleaned in preparation for grouting and future closure of the TFF. Four other 300,000-gal tanks remain in 

an active, operational status to store the remaining SBW. In the future, as the SBW is removed from the 

tanks for treatment and disposal, these tanks will be cleaned, grouted, and closed. 

7.12.4 Closure 

As described in Section 2, closure of the tanks includes cleaning of the tanks, vaults, and associated 

interconnecting piping and ancillary equipment, followed by completely filling these components with a 

grout material to immobilize any remaining residual waste and provide for a stable structure.aaaa 

Long-term stability of the disposal site after closure means that the waste maintains structural 

integrity under the expected disposal conditions. As such, the long-term stability of the closed facility is 

an important element of meeting the performance objectives. Stability prevents trench subsidence, water 

 

aaaa. In a separate process under CERCLA, DOE also plans to cap the surface of the TFF to meet the remedial action objectives 

agreed to by DOE, DEQ, and EPA pursuant to the 1991 Federal Facility Agreement and Consent Order. The DOE’s Proposed 

Plan for Tank Farm Soil and INTEC Groundwater, Operable Unit 3-14 (ICP 2006k), which includes capping the surface of the 

TFF, has been issued for public comment. The CERCLA ROD is planned for 2007. Capping will reduce water infiltration and 

provide worker protection, where appropriate. 
105

infiltration, and radionuclide release because of disintegration of the waste form, and minimizes the 

likelihood of intrusion into the waste. The waste form itself can provide structural stability by processing 

the waste to a stable form or by placing the waste in a disposal container or structure that provides 

stability after disposal. The primary barrier relied upon at the TFF to provide structural stability for the 

tank residuals is the stainless steel tank. The primary barrier for the residuals in the vault and sandpads is 

the concrete vault walls and the grouted vault, which provide structural stability and reduce migration of 

contamination. The primary barrier for the residuals in the transfer piping is the stainless steel piping 

(DOE-ID 2003b). 

The grouted tanks, vaults, piping, and valve boxes provide a long-term, stable waste form. For 

guidance, DOE considered the Standard Format and Content of a License Application for a Low-Level 

Radioactive Waste Disposal Facility (NRC 1987a), which notes that site stability is focused on reducing 

the contact of water with the waste and providing assurance that active maintenance will not be needed 

following closure. The long-term stability of the facility is evaluated in the TFF PA (DOE-ID 2003b), 

which provides a degradation analysis of the grouted tanks, vaults, and piping, and shows the grout will 

likely remain intact for at least 2,000 years, but much more likely to remain intact for much longer. As 

shown above, the site conditions do not present hazards that impact the stability of the closed TFF. In 

addition, the methods used to close the TFF will result in a closed facility that does not require ongoing 

active maintenance of the facility following closure. As such, the performance of the closed TFF complies 

with the objectives of 10 CFR 61.44.bbbb

7.13 Conclusions 

The previous analysis demonstrates that disposal of the stabilized residuals in the TFF, and in the 

TFF tank system, will meet the performance objectives set out in 10 CFR 61, Subpart C.cccc 

 

bbbb. On page 112 of its TER (NRC 2006b) NRC concluded, “...there is reasonable assurance that DOE Idaho can meet 

10 CFR 61.44 stability requirements.” 

cccc. On page 111 of its TER (NRC 2006b) NRC concluded, “There is reasonable assurance that DOE Idaho can meet Criterion 

Three of the NDAA because: 

• Based on information provided by DOE Idaho, NRC staff expects the maximum public dose from all pathways to be below 

the 0.25-mSv/yr [25-mrem/yr] dose limit. Reasonable effort will be made by DOE Idaho to maintain releases in radioactive 

activity in effluents to the general environment as low as reasonable achievable. Therefore NRC concludes that there is 

reasonable assurance that DOE Idaho can meet 10 CFR 61.41 requirements. 

• Based on analysis provided by DOE Idaho, NRC staff concludes there is reasonable assurance that DOE Idaho can meet 

10 CFR 61.42 requirements for protection of individuals from inadvertent intrusion. 

• Workers are protected by DOE regulations that are comparable to 10 CFR Part 20. DOE Idaho controls are also in place to 

protect members of the public during operations. Therefore, the NRC staff concludes that there is reasonable assurance that 

DOE Idaho can meet 10 CFR 61.43 requirements for protection of individuals during operations. 

• DOE Idaho plans to fill the tanks, vaults, and ancillary equipment with grout which will provide structural stability and 

limit waste dispersal. Therefore, NRC staff concludes that there is reasonable assurance that DOE Idaho can meet 10 CFR 

61.44 stability requirements.” 
106

8. STATE-APPROVED CLOSURE PLANS 

Key Points 

• DOE is conducting performance-based closure of the TFF pursuant to closure plans approved by 

the State of Idaho. DOE intends for the TFF closure to meet clean closure standards for 

hazardous constituents. Clean closure is the removal or decontamination of all waste residues to 

meet State-approved levels for hazardous constituents. If these clean closure standards are not 

met, the TFF may be closed in accordance with landfill standards established by the State of 

Idaho in a State-approved closure plan. However, waste removal and decontamination activities 

needed to meet performance objectives for radioactive constituents will likely result in the 

removal of hazardous constituents to the clean closure standard. 

• Prior to closure, the closure plan for the four 300,000-gal tanks remaining to be cleaned will be 

submitted to the State of Idaho. 

• A final closure plan for all of the TFF tanks will be submitted to the State of Idaho after each 

partial closure phase of the TFF is completed. 

Section 3116(a) of the NDAA 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. 

As discussed in Subsection 2.3, the TFF closure is being performed under both DOE and State of 

Idaho Department of Environmental Quality (DEQ) requirements, which includes preparing closure 

plans. The wastes stored in the TFF are mixed wastes.dddd As such, the State of Idaho regulates the 

hazardous constituents and the DOE regulates the radioactive constituents. Because the TFF stores mixed 

wastes, TFF closure must comply with closure requirements for hazardous waste as well as radioactive 

waste. For hazardous waste, closure must comply with HWMA (State of Idaho 1983) and RCRA 

(42 USC 6901 et seq. 1976) as implemented by IDAPA 58.01.05.009 and 40 CFR 265, “Interim Status 

Standards for Owners and Operators of Hazardous Waste Treatment, Storage, and Disposal Facilities.” 

These regulations require the preparation of a closure plan that includes specific elements. Because the 

TFF tanks do not meet secondary containment standards, the regulations also specify that both a closure 

plan for clean closure and a contingent closure plan for closure of the TFF to meet landfill standards must 

be prepared. Clean closure is the removal or decontamination of waste residues to meet State-approved 

performance objectives for the hazardous constituents. Landfill closure is necessary if the clean closure 

standards cannot be met. The clean closure standards are included as part of the HWMA/RCRA closure 

plans. For all of the tanks and ancillary equipment cleaned from 2002 to 2005, these clean closure 

requirements have been met. 

 

dddd. A mixture of hazardous wastes regulated under Subtitle C of the Resource Conservation and Recovery Act of 1976 

(42 USC 6901 et seq. 1976) and radioactive wastes regulated under the Atomic Energy Act of 1954 (42 USC 2011 et seq. 1954). 
107

In accordance with these requirements, the DOE has prepared a series of closure plans to support 

the TFF closure. These closure plans present the strategy for clean closure of the TFF tanks to 

site-specific action levels. The closure plan for TFF Tanks WM-187, WM-188, WM-189, and WM-190 

will be submitted prior to closure. In addition, a final closure plan for all of the TFF tanks will be 

prepared and submitted to the state for approval. Closure actions, such as grout placement, will not be 

initiated until the DEQ approves the applicable closure plan. 

The following closure plans have been submitted to DEQ for approval: 

• Tanks WM-182 and WM-183—Idaho Hazardous Waste Management Act/Resource Conservation 

and Recovery Act Closure Plan for Idaho Nuclear Technology and Engineering Center Tanks 

WM-182 and WM-183 (DOE-ID 2003a) 

• Tanks WM-184, WM-185, and WM-186—Idaho Hazardous Waste Management Act/Resource 

Conservation and Recovery Act Closure Plan for Idaho Nuclear Technology and Engineering 

Center Tanks WM-184, WM-185, and WM-186 (DOE-ID 2004a) 

• Tanks WM-103, WM-104, WM-105, WM-106, and WM-181—Idaho Hazardous Waste 

Management Act/Resource Conservation and Recovery Act Closure Plan for Idaho Nuclear 

Technology and Engineering Center Tanks WM-103, WM-104, WM-105, WM-106, and WM-181

(DOE Idaho 2004a) 

• Tank WM-180—Idaho Hazardous Waste Management Act/Resource Conservation and Recovery 

Act Closure Plan for Idaho Nuclear Technology and Engineering Center Tank WM-180

(DOE Idaho 2004b). 

In addition, the DOE has prepared and submitted to DEQ two contingent landfill closure plans as 

required by the hazardous waste management regulations. The Contingent Landfill Closure and 

Post-Closure Plan for Idaho Nuclear Technology and Engineering Center Tanks WM-182 and WM-183

(DOE-ID 2003c) applies to TFF Tanks WM-182 and WM-183. The Contingent Landfill Closure and 

Post-Closure Plan for Idaho Nuclear Technology and Engineering Center Tanks in the Tank Farm 

Facility (DOE-ID 2003d) applies to all of the tanks in the TFF. Previously, contingent landfill closure 

plans were planned to be developed for each closure phase, in a manner similar to the clean closure plans. 

However, DEQ and DOE agreed that one contingent landfill closure plan should be developed for the 

entire TFF. 

The closure activities presented in these plans are designed to satisfy both DOE radioactive waste 

requirements and HWMA/RCRA hazardous waste requirements. Although contingent landfill closure 

plans were prepared as required, the TFF is being closed to HWMA/RCRA clean closure standards. 

Waste removal and decontamination activities needed to remove the required radionuclides to meet 

performance objectives for the radioactive waste will result in the removal of the hazardous constituents 

to the HWMA/RCRA clean closure standard. 

In addition to the closure plans, a SAP for each closure phase is submitted to DEQ for approval. 

These plans describe the sampling and analysis to be performed to confirm that the residuals remaining 

the tanks and ancillary equipment: (1) are not hazardous waste, (2) meet the clean closure action levels 

specified in the associated HWMA/RCRA closure plan, and (3) meet the DOE performance objectives. 

As the sampling and analysis activities for a closure phase are completed, a series of DQA reports are 

prepared. The purpose of the DQAs is to evaluate the quality of the characterization data collected to 

determine whether the data could be used to meet the DQOs established for the sampling and analysis 

activities. The following SAPs have been submitted to DEQ: 
108

• Tanks WM-182 and WM-183 

- Sampling and Analysis Plan for the Post-Decontamination Characterization of the WM-182 

and WM-183 Tank Residuals (INEEL 2002b) 

- Sampling and Analysis Plan for the Post-Decontamination Characterization of the Process 

Waste Lines from INTEC Tank Farm Facility Tanks WM-182 and WM-183 (INEEL 2001) 

• Tanks WM-184, WM-185, and WM-186 

- Sampling and Analysis Plan for the Post-Decontamination Characterization of the WM-184, 

WM-185, and WM-186 Tank Residuals (INEEL 2003)

• Tanks WM-103, WM-104, WM-105, WM-106, and WM-181 

- Sampling and Analysis Plan for the Post-Decontamination Characterization of the WM-103, 

WM-104, WM-105, WM-106, and WM-181 Tank Residuals (ICP 2004a) 

• Tank WM-180 

- Sampling and Analysis Plan for the Post-Decontamination Characterization of the WM-180 

Tank Residuals (ICP 2004b). 

The following DQA reports have been provided to DEQ: 

• Tanks WM-182 and WM-183 

- Data Quality Assessment Report for the Post-Decontamination Characterization of the 

Contents of Tank WM-182 at the Idaho Nuclear Technology and Engineering Center Tank 

Farm Facility (INEEL 2006a)

- Data Quality Assessment Report for the Post-Decontamination Characterization of the 

Contents of Tank WM-183 at the Idaho Nuclear Technology and Engineering Center Tank 

Farm Facility (INEEL 2006b)

- Data Quality Assessment Report for the Post-Decontamination Characterization of the 

Ancillary Equipment Associated with Tanks WM-182 and WM-183 at the Idaho Nuclear 

Technology and Engineering Center Tank Farm Facility (ICP 2006b)

- Data Quality Assessment Report for the Post-Decontamination Characterization of the 

Contents of Rewashed Tank WM-183 at the Idaho Nuclear Technology and Engineering 

Center Tank Farm Facility (ICP 2006a)

• Tanks WM-184, WM-185, and WM-186 

- Data Quality Assessment Report for the Post-Decontamination Characterization of the 

Contents of Tank WM-184 at the Idaho Nuclear Technology and Engineering Center Tank 

Farm Facility (ICP 2006c)

- Data Quality Assessment Report for the Post-Decontamination Characterization of the 

Contents of Tank WM-185 at the Idaho Nuclear Technology and Engineering Center Tank 

Farm Facility (ICP 2006d)
109

- Data Quality Assessment Report for the Post-Decontamination Characterization of the 

Contents of Tank WM-186 at the Idaho Nuclear Technology and Engineering Center Tank 

Farm Facility (ICP 2006e)

- Data Quality Assessment Report for the Post-Decontamination Characterization of the 

Ancillary Equipment Associated with Tanks WM-184, WM-185, and WM-186 at the Idaho 

Nuclear Technology and Engineering Center Tank Farm Facility (ICP 2006f)

• Tanks WM-103, WM-104, WM-105, WM-106, and WM-181 

- Data Quality Assessment Report for the Post-Decontamination Characterization of the 

Contents of Tanks WM-103, WM-104, WM-105, and WM-106 at the Idaho Nuclear 

Technology and Engineering Center Tank Farm Facility (ICP 2006g) 

- Data Quality Assessment Report for the Post-Decontamination Characterization of the 

Contents of Tank WM-181 at the Idaho Nuclear Technology and Engineering Center Tank 

Farm Facility (ICP 2006h)

- Data Quality Assessment Report for the Post-Decontamination Characterization of the 

Ancillary Equipment Associated with Tanks WM-103, WM-104, WM-105, WM-106, and 

WM-181 at the Idaho Nuclear Technology and Engineering Center Tank Farm Facility 

(ICP 2006i).

• Tank WM-180 

- Data Quality Assessment Report for the Post-Decontamination Characterization of the 

Contents of Tank WM-180 and Associated Ancillary Equipment at the Idaho Nuclear 

Technology and Engineering Center Tank Farm Facility (ICP 2006j). 

As of September 2006, DEQ has approved four of the closure plans for clean closure 

(DOE-ID 2003a, 2004a; DOE Idaho 2004a, 2004b) and their associated SAPs (ICP 2004a, 2004b; INEEL 

2001, 2002b, 2003). Both of the contingent landfill closure plans (DOE-ID 2003c, 2003d) have also been 

approved.eeee 

Results of the confirmatory sampling and analysis of the TFF tank residuals for the TFF closure 

activities performed as of May 2005 indicate that HWMA/RCRA clean closure requirements and DOE 

performance objectives as set forth in the closure plans have been met. The final closure of the TFF will 

be certified by an independent, Idaho-registered professional engineer, the facility contractor, and/or the 

DOE Idaho Operations Office, in accordance with IDAPA 58.01.05.009 (40 CFR 265.115).ffff 

 

eeee. On page xii of its TER (NRC 2006b) NRC states, “DOE Idaho is seeking clean closure status with the Idaho Department of 

Environment Quality (DEQ) for the tanks at INTEC under the Hazardous Waste Management Act (HWMA) and Resource 

Conservation and Recovery Act (RCRA) which would allow DOE Idaho to close the TFF without a state permit.” It should be 

noted that the TFF will be closed under state-approved closure plans as described in this section. Grouting will not be performed 

on tanks until they have State-approved closure plans. 

ffff. Certification of partial closures is not required (EPA 1998). 
110

9. CONCLUSION 

Based on consultation with the NRC and as shown in the preceding sections of this 3116 Basis 

Document,gggg the stabilized residuals in the TFF, and the TFF tank system, are not HLW pursuant to the 

criteria set forth in Section 3116(a) of the NDAA and may be disposed of at the INL Site in accordance 

with Section 3116 of the NDAA. This conclusion takes into account NRC comments during consultation, 

and although not required by Section 3116, State of Idaho and other public comments. 

 

gggg. On page 114 of its TER (NRC 2006b) NRC concludes, “DOE Idaho has adequately demonstrated that NDAA criteria in 

Section 3116(a)(1), (a)(2), and (a)(3)(A)(i) or (a)(3)(B)(i) can be met for residual waste disposed of at TFF.” 
111

10. REFERENCES 

10 CFR 20, 2004, “Standards for Protection Against Radiation,” Code of Federal Regulations, Office of 

the Federal Register, January 1, 2004. 

10 CFR 60, 2004, “Disposal of High-Level Radioactive Wastes in Geological Repositories,” Code of 

Federal Regulations, Office of the Federal Register, January 1, 2004. 

10 CFR 61, 2004, “Licensing Requirements for Land Disposal of Radioactive Waste,” Code of Federal 

Regulations, Office of the Federal Register, January 1, 2004. 

10 CFR 820, 2004, “Procedural Rules for DOE Nuclear Activities,” Code of Federal Regulations, Office 

of the Federal Register, January 1, 2004. 

10 CFR 830, Subpart A, 2004, “Quality Assurance Requirements,” Code of Federal Regulations, Office 

of the Federal Register, January 1, 2004. 

10 CFR 835, 2004, “Occupational Radiation Protection,” Code of Federal Regulations, Office of the 

Federal Register, January 1, 2004. 

40 CFR 61, 2004, “National Emissions Standards for Hazardous Air Pollutants,” Code of Federal 

Regulations, Office of the Federal Register, July 1, 2004. 

40 CFR 265, 2004, “Interim Status Standards for Owners and Operators of Hazardous Waste Treatment, 

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11. DRAWING 

055705, INL Reference Drawing, INTEC Plot Plan, Rev. 41, February 15, 2005. (This drawing has been 

classified as “Official Use Only.”)