DOE-WD-2005-001 

February 28, 2005 

Draft Section 3116 Determination 

Salt Waste Disposal 

Savannah River Site

 APPROVED for Release for

Unlimited (Release to Public)

3/24/2005
 

February 28, 2005 

Page 2 of 94 

TABLE OF CONTENTS

1.0 INTRODUCTION AND PURPOSE ............................................................................................. 6

2.0 BACKGROUND....................................................................................................................... 9

2.1 SALT PROCESSING STRATEGY ............................................................................................. 10 

3.0 SECTION 3116 (A) OF THE RONALD W. REAGAN NATIONAL DEFENSE AUTHORIZATION 

ACT FOR FISCAL YEAR 2005 (NDAA) .............................................................................. 25

4.0 THE WASTE DOES NOT REQUIRE PERMANENT ISOLATION IN A DEEP GEOLOGIC 

REPOSITORY FOR SPENT FUEL OR HIGH-LEVEL RADIOACTIVE WASTE ........................ 27

5.0 THE WASTE HAS HAD HIGHLY RADIOACTIVE RADIONUCLIDES REMOVED TO THE 

MAXIMUM EXTENT PRACTICAL........................................................................................ 29

5.1 RADIONUCLIDES CONSIDERED ............................................................................................ 29 

5.2 REMOVAL TO THE MAXIMUM EXTENT PRACTICAL ............................................................. 30 

6.0 THE WASTE DOES NOT EXCEED CONCENTRATION LIMITS IN 10 CFR 61.55................. 43

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

OBJECTIVES SET OUT IN 10 CFR PART 61, SUBPART C .................................................. 50

7.1 10 CFR PART 61 PERFORMANCE OBJECTIVES..................................................................... 50 

7.2 GENERAL REQUIREMENT (10 CFR 61.40)........................................................................... 51 

8.0 STATE-ISSUED PERMIT....................................................................................................... 76

9.0 CONCLUSION ...................................................................................................................... 77

10.0 REFERENCES....................................................................................................................... 78

APPENDIX 1: SYSTEM DESCRIPTION ............................................................................................ 83

APPENDIX 2: WASTE CHARACTERIZATION SYSTEM ................................................................... 93
 

February 28, 2005 

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LIST OF TABLES

TABLE 6.1: 10 CFR 61.55 TABLE 1................................................................................................ 43 

TABLE 6.2: 10 CFR 61.55 TABLE 2................................................................................................ 44 

TABLE 6.3: DDA WASTE STREAM COMPARISON TO 10 CFR 61.55 TABLE 1................................. 44 

TABLE 6.4: DDA WASTE STREAM COMPARISON TO 10 CFR 61.55 TABLE 2................................. 45 

TABLE 6.5: ARP/MCU WASTE STREAM COMPARISON TO 10 CFR 61.55 TABLE 1 ....................... 45 

TABLE 6.6: ARP/MCU WASTE STREAM COMPARISON TO 10 CFR 61.55 TABLE 2 ....................... 46 

TABLE 6.7: SWPF WASTE STREAM COMPARISON TO 10 CFR 61.55 TABLE 1............................... 46 

TABLE 6.8: SWPF WASTE STREAM COMPARISON TO 10 CFR 61.55 TABLE 2............................... 47 

LIST OF FIGURES

FIGURE 2.1: SALT PROCESSING PATHWAYS [2,4]........................................................................... 11 

FIGURE 2.2: SALT PROCESSING VOLUMES [3]................................................................................ 12 

FIGURE 2.3: DDA PROCESS (INITIAL CONDITIONS, STEPS 1 AND 2)............................................... 13 

FIGURE 2.4: DDA PROCESS (STEPS 3 THROUGH 5)......................................................................... 14 

FIGURE 2.5: DDA PROCESS (STEPS 6 THROUGH 8)......................................................................... 15 

FIGURE 2.6: ACTINIDE REMOVAL PROCESS.................................................................................... 18 

FIGURE 2.7: MODULAR CSSX UNIT PROCESS................................................................................ 19 

FIGURE 2.8: SWPF ALPHA STRIKE PROCESS ................................................................................. 20 

FIGURE 2.9: SWPF CSSX PROCESS............................................................................................... 21 

FIGURE 5.1: WASTE TANK COMPOSITE INVENTORY (AS OF 12/01/04) [2] ..................................... 34 
 

February 28, 2005 

Page 4 of 94 

LIST OF ACRONYMS

AEA Atomic Energy Act 

AFP Alpha Finishing Process 

ALARA As Low As Reasonably Achievable 

ARP Actinide Removal Process 

ASP Alpha Strike Process 

CEDE Committed Effective Dose Equivalent 

CFR Code of Federal Regulations 

CHA Consolidated Hazards Analysis 

CSS Clarified Salt Solution 

CSSX Caustic Side Solvent Extraction 

DDA Deliquification, Dissolution, and Adjustment 

DF Decontamination Factor 

DNFSB Defense Nuclear Facilities Safety Board 

DOA Department of Energy Organization Act 

DOE Department of Energy 

DSA Documented Safety Analysis 

DSS Decontaminated Salt Solution 

DWPF Defense Waste Processing Facility 

EIS Environmental Impact Statement 

EM Environmental Management 

EPA Environmental Protection Agency 

FFT Filter Feed Tank 

FHT Filtrate Hold Tank 

GWSB Glass Waste Storage Building 

HAW High Activity Waste 

HLW High Level Waste 

HM H Modified 

ICRP International Commission on Radiological Protection 

ISWLF Industrial Solid Waste Landfill 

ITP In-Tank Precipitation 

IW Inhibited Water 

kgal Thousand Gallons 

LAW Low Activity Waste 

LLW Low Level Waste 

MCi Million Curies 

MCL Maximum Contaminant Level 

MCU Modular CSSX Unit 

Mgal Million Gallons 

MST Monosodium Titanate 

NASA National Aeronautics and Space Administration 

NCRP National Council on Radiation Protection and Measurements 

NDAA National Defense Authorization Act 

NRC Nuclear Regulatory Commission 
 

February 28, 2005 

Page 5 of 94 

OSHA Occupational Safety and Health Administration 

PA Performance Assessment 

PEL Permissible Exposure Limit 

PUREX Plutonium Recovery and Extraction 

ROD Record of Decision 

RPP Radiation Protection Program 

SA Special Analysis 

SCDHEC South Carolina Department of Health and Environmental Control 

SDF Saltstone Disposal Facility 

SEIS Supplemental Environmental Impact Statement 

SNF Spent Nuclear Fuel 

SPF Saltstone Production Facility 

SRS Savannah River Site 

SWPF Salt Waste Processing Facility 

TEDE Total Effective Dose Equivalent 

TPB Tetraphenylborate 

TRU Transuranic 

TSR Technical Safety Requirements 

USC United States Code 

WAC Waste Acceptance Criteria 

WCS Waste Characterization System 

WCP Waste Compliance Plan 

WSRC Westinghouse Savannah River Company
February 28, 2005 

Page 6 of 94 

1.0 INTRODUCTION AND PURPOSE

This draft 3116 Determination is being issued pursuant to Section 3116 of the Ronald W. Reagan 

National Defense Authorization Act for Fiscal Year 2005 (NDAA) [1]. It concerns the disposal 

of separated, solidified low-activity radioactive salt waste at the Savannah River Site (SRS) near 

Aiken, South Carolina. 

As a matter of background, there are presently 36.4 million gallons [2] (Mgal) of liquid 

radioactive waste stored in underground waste storage tanks at SRS. Much of this waste resulted 

from the reprocessing of spent nuclear fuel for defense purposes by the Department of Energy 

(DOE) and its predecessor agencies, which has been commingled with non-reprocessing waste 

that resulted from the production of targets for nuclear weapons and the production of material 

for the National Aeronautics and Space Administration (NASA) space missions. Over 

approximately the next 14 years, DOE anticipates that the tanks will receive an additional 

approximately 41.3 Mgal [3] of waste1

. The Department of Energy’s top cleanup priority at SRS 

is to remove, stabilize, and dispose of the waste in these tanks and close the tanks. 

The tank waste consists of two distinct kinds of material: approximately 2.6 Mgal [2] of sludge 

consisting primarily of metals that settled at the bottom of the tanks; and approximately 33.8 

Mgal [2] of salt waste which is comprised of concentrated salt solution (supernate) and 

crystallized saltcake. 

DOE’s plans call for stabilizing and disposing of the sludge in a deep geologic repository for 

spent nuclear fuel (SNF) and high-level radioactive waste. This will be done by stabilizing the 

high-level waste (HLW) in a borosilicate glass matrix through vitrification in a facility known as 

the Defense Waste Processing Facility (DWPF.) This process has been on-going since 1996. 

As for the salt waste: DOE contemplates removing fission products and actinides from these 

materials using a variety of technologies, combining the removed fission products and actinides 

with the metals being vitrified in DWPF, and solidifying the remaining low-activity salt stream 

into a grout matrix known as saltstone grout suitable for disposal in vaults at the Saltstone 

Disposal Facility (SDF) at SRS. 

It is the disposal of this latter solidified low-activity salt stream that is the subject of the proposed 

determination set forth in this document. Section 3116 of the NDAA authorizes the Secretary of 

Energy, in consultation with the Nuclear Regulatory Commission (NRC), to determine that 

 

1

 The majority of this future waste is comprised of unconcentrated liquid waste resulting from the operation of the 

Defense Waste Processing Facility (DWPF). The DWPF recycle waste stream is a generally very low-activity waste 

stream that consists of condensate from chemical processing and melter operation, waste from decontamination 

activities, and waste from miscellaneous drains and sumps in DWPF. Before the recycle stream is transferred to the 

Tank Farm, it is chemically adjusted with sodium nitrite and sodium hydroxide to ensure that Tank Farm Waste 

Acceptance Criteria (WAC) is met. The remainder of the future waste results from the stabilization of legacy nuclear 

materials in the H-Canyon Facility. The volumes discussed above assume that H-Canyon will cease operations in 

2010. 
February 28, 2005 

Page 7 of 94 

certain waste from reprocessing is not HLW and that it may instead be disposed of as low-level 

waste (LLW) if it meets the criteria set forth in that Section: that it does not require disposal in a 

deep geologic repository, that it has had highly radioactive radionuclides removed to the 

maximum extent practical, that it meets concentration limits and/or dose-based performance 

objectives for near-surface disposal of radioactive waste, and that it will be disposed pursuant to a 

State-issued permit or State-approved closure plan. In this document, the Secretary proposes to 

determine that the treated, solidified low-activity salt waste from the tanks will meet all of these 

criteria. Accordingly, this material may be determined not to be HLW and may be disposed of 

instead as LLW in SDF. 

Although this draft 3116 Determination addresses only the disposal of solidified low-activity salt 

waste streams, it is important to understand that successful and timely salt waste removal and 

disposal is integral to DOE’s efforts to proceed with all aspects of tank cleanup and closure, 

extending well beyond disposal of the solidified low-activity salt waste streams themselves. This 

is not only for the obvious reason that the salt waste must be removed and treated before the 

tanks may be closed. Less obviously, but more importantly, the disposal pursuant to this draft 

3116 Determination will enable DOE to continue without interruption to remove and stabilize the 

high-activity fraction of the waste. This is because DOE uses the tanks to prepare the highactivity waste so that it may be fed into DWPF. The issue is that the salt waste is filling up tank 

space needed to allow this preparation activity to continue. Thus, removal and disposal of lowactivity salt waste is critical in order to relieve this tank space shortage and assure that 

vitrification of the high-activity fraction will be able to continue uninterrupted. 

The Department of Energy is tailoring the waste treatment for the salt waste based on the 

radiological properties of those wastes as shown by tank waste characterization data. DOE is 

separating the salt waste to segregate the low-activity fraction at SRS using a two-phase, threepart process. The first phase will involve two parts to treat the lower activity salt waste: a) 

beginning in 2005, processing of a minimal amount of the lowest activity salt waste through a 

process involving deliquification, dissolution, and adjustment (DDA) of the waste; and b) 

beginning in approximately 2007, processing of a minimal amount of additional salt waste with 

slightly higher activity levels using an Actinide Removal Process (ARP) and a Modular Caustic 

Side Solvent Extraction (CSSX) Unit (MCU), along with deliquification and dissolution of 

saltcake. The second, and longer term phase, beginning in approximately 2009, involves the 

separation and processing of the remaining (and by far the majority) of the salt waste using a Salt 

Waste Processing Facility (SWPF) (augmented as necessary by ARP). This second phase will 

begin as soon as SWPF is constructed, permitted by the State of South Carolina, and operational 

[3,4]. 

This document refers to the first phase as Interim Salt Processing because it will be used to 

initiate treatment and disposal of solidified low-activity salt waste streams using the technologies 

available during the interim period between 2005 and approximately 2009 when DOE anticipates 

that SWPF will become operational. As explained in more detail later in this draft 3116 

Determination, DOE believes it should proceed with this interim approach because doing so will 

enable DOE to continue uninterrupted use of DWPF as well as to use SWPF at higher capacity as 

soon as it comes on line. This will allow DOE to complete cleanup and closure of the tanks years 
February 28, 2005 

Page 8 of 94 

earlier than would otherwise be the case. That, in turn, will reduce the time during which the 

tanks – including some that do not have full secondary containment and have a known history of 

leak sites – continue to store liquid radioactive waste. Finally, Interim Salt Processing will make 

more tank space available for routine operations, thereby reducing the number of transfers among 

tanks and increasing the safety of operations. Therefore, Interim Salt Processing will accelerate 

the reduction of potential risk to the environment, the public, and SRS workers. 

SWPF will be a high-capacity salt processing facility that will process the remaining salt waste 

retrieved from aging tanks that do not have full secondary containment, as well as tanks with full 

secondary containment. Thus, SWPF will also support associated reduction of potential risk to 

the environment, the public, and SRS workers. 

After undergoing removal of radionuclides through DDA, ARP/MCU, and SWPF, as the case 

may be, the low-activity salt solution will be solidified into a grout matrix and disposed in the 

SDF vaults at SRS. The low-activity salt solution will be mixed with dry chemicals (cement, 

slag, and flyash) to form a homogeneous grout mixture in the Saltstone Production Facility (SPF) 

and the slurry is transferred to SDF where it solidifies. SDF and SPF are commonly collectively 

referred to as the Saltstone Facility. 

All told, these various technologies are expected to result in removal and vitrification through 

DWPF of approximately 97.8 to 98.7% (approximately 218 to 220 million curies (MCi) of the 

223 MCi total curies in salt waste) of the total curies originally contained in the salt waste. The 

total curies being disposed of in SDF will be approximately 1.3 to 2.2% (3 to 5 MCi) of the 

approximately 223 MCi. DOE’s current best estimate is that 3 MCi will be disposed of in SDF. 

The purpose of this draft 3116 Determination is to demonstrate and document that the resulting 

solidified low-activity salt streams meet the 3116 criteria, and therefore may be determined not to 

be HLW and may be disposed as LLW at SRS in accordance with Section 3116 of the NDAA. 

This draft 3116 Determination addresses disposal of the solidified low-activity salt streams from 

both Interim Salt Processing and SWPF through the end of salt processing at SRS. 

This draft 3116 Determination will be finalized after DOE has completed consultation with the 

NRC. Although not required by Section 3116, DOE is also issuing this draft 3116 Determination 

for public review and comment. 
February 28, 2005 

Page 9 of 94 

2.0 BACKGROUND

Since it became operational in 1951, SRS has produced nuclear material for national defense, 

research, medical, and space programs resulting in the generation of large quantities of 

radioactive waste which are currently stored onsite in 492

 underground carbon steel waste storage 

tanks. 

Since initiation of operations at SRS, the Tank Farms have received over 140 Mgal [5] of liquid 

waste from the chemical separation processes in F- and H-Canyons associated with the 

reprocessing of spent nuclear fuel, the production of nuclear materials for weapons, and the 

production of material for NASA space missions. Prior to transfer of the waste material from the 

F- and H-Canyons, chemicals (sodium hydroxide) are added to adjust the waste to an alkaline 

state to prevent corrosion of the carbon steel waste tanks. This chemical adjustment results in the 

precipitation of metals including strontium (Sr) and actinides (e.g., plutonium (Pu)). These solids 

settle to the bottom of the waste tanks forming a layer that is commonly referred to as sludge. 

After settling of the solids has occurred, the salt solution (supernate) above this sludge layer is 

decanted off. In order to maximize the space available in the tanks for storing additional waste, 

DOE’s practice at SRS has been to use the Tank Farm evaporator systems to reduce the volume 

of the decanted supernate and concentrate the waste. 

During the evaporation process, the salt waste is concentrated and forms two distinct phases – 

concentrated supernate solution and solid saltcake (collectively called salt waste). The solid 

saltcake is composed predominantly of nitrate and nitrite salts and contains relatively small 

quantities of radioactive material. Because of the relative high solubility of cesium (Cs), the 

predominant radionuclide present in salt waste, approximately 95% [2] of the Cs-137 in the salt 

waste is found in the concentrated supernate solution. As the result of the evaporation process, 

over 140 Mgal of liquid waste originally received have been reduced to the present volume of 

approximately 36 Mgal [2]. Evaporator operations have been extremely effective in minimizing 

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

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

the total waste volume stored are not possible. 

SRS no longer conducts the weapons or NASA-related materials production activities or the 

weapons-related spent nuclear fuel reprocessing that generated the original waste, and DOE is 

now in the process of deactivating and decommissioning F-Canyon and cleaning out and closing 

the Tank Farms. Nevertheless, DOE continues to have a need for tank space for a number of 

purposes. Perhaps most significantly, substantial amounts of tank space are required in order to 

safely and effectively remove and prepare the tank waste for disposal. Tank space is also needed 

for the waste generated by other on-going cleanup activities as well as stabilization of legacy 

nuclear materials in the DOE Complex. 

 

2

 SRS has a total of 51 underground waste storage tanks. In 1997, following consultation with the NRC and the State 

of South Carolina approval of closure modules, DOE operationally closed Tanks 17 and 20. 
February 28, 2005 

Page 10 of 94 

Accordingly, an integral element of DOE’s on-going efforts to clean up and close the tanks 

without significant inefficiency and delay is prompt initiation of salt retrieval and processing so 

as to create sufficient tank space to allow DOE to maintain sustained risk-reduction activities, 

including the vitrification of high-activity sludge waste through DWPF and subsequent disposal 

in a federal repository for HLW and spent nuclear fuel. 

In March 2001, DOE issued a draft environmental impact statement on salt waste processing 

alternatives. The SRS Salt Processing Alternatives Final Supplemental Environmental Impact 

Statement (SEIS) [6] was published in June 2001, and in October 2001, DOE issued a Record of 

Decision (ROD)3

 [7] identifying CSSX and monosodium titanate (MST) as the preferred 

treatment technologies for salt waste. The ROD also stated that: 

“... DOE will evaluate implementation of any of the other salt processing alternatives for 

specific waste portions for which processing could be accelerated or that could not be 

processed in the Caustic Side Solvent Extraction facility. These evaluations and potential 

operations would be undertaken to maintain operational capacity and flexibility in the 

HLW system, and to meet commitments for closure of high-level waste tanks.” 

 

Since issuing the SEIS and the ROD, DOE has evaluated alternative technologies for the 

purposes contemplated in the ROD. In particular, to separate, treat, and dispose of the salt waste, 

DOE has now developed a strategy using a two-phase, three-part approach: (a) “Interim Salt 

Processing” [5], which will rely on one of the alternative technologies, DDA along with smallscale implementation of the technologies originally selected for the SWPF (ARP and CSSX), 

beginning in approximately 2005 and ending in approximately 2009 when the full-scale SWPF 

becomes operational and (b) “High Capacity Salt Processing” using SWPF from approximately 

2009 through approximately 2019. The quantities and percentages shown in Figure 2.1 reflect 

the radioactivity (i.e., curies) in the approximately 33.8 Mgal of salt waste currently stored in 

underground storage tanks at SRS [2,4]. Although as explained above, DOE also anticipates 

processing an additional 41.3 Mgal of future waste which are not included in these estimates. 

DOE believes the inclusion of this waste will not appreciably affect the total curies to be disposed 

of in the SDF. This is because the majority of this future waste is compromised of the lowactivity DWPF Recycle waste stream and because future H-Canyon receipts will be processed 

through the SWPF, the final curies disposed of in the SDF should be as shown in this figure. 

DOE is currently preparing a Supplement Analysis and updated ROD to reflect specific use of 

DDA and ARP/MCU. 

2.1 SALT PROCESSING STRATEGY

As described above, DOE has developed a two-phase, three-part approach for salt processing. 

The salt processing pathways and associated curie distribution are shown below in Figure 2.1. 

 

3

 Additional discussion of the alternatives evaluated, their environmental impacts, and DOE’s decision to implement 

the CSSX and MST technologies to treat salt waste can be found in the SEIS and the ROD. 
February 28, 2005 

Page 11 of 94 

Figure 2.1: Salt Processing Pathways [2,4] 

 

SPF / SDF Salt 

Waste

MCU

Federal

Repository

ARP

SWPF 

DWPF

DDA

Interim Salt 

Processing

(through ~ 2009)

High-Capacity

Salt Processing

(beginning ~ 2009)

ARPb

~223 MCi

100%

~2.5 MCi

~0.3 MCi

~0.2 MCi

~0.1%

~0.1%

~1.1%

~220 MCi

~98.7%

~3.0 MCia

~1.3%

a 

Due to the uncertainty associated with the current characterization of the saltcake waste, the actual curie content of 

this material may be as high as 5 MCi and the percentages would change accordingly. Curie numbers include 

daughter products of Cs-137 and Sr-90.

b

ARP Facilities will have the capability to supplement the actinide removal capacity of SWPF if required.

~217.1 MCi

~3.1 MCi

~6.3 MCi

~3.4

MCi

~216.9 MCi

~8.8 MCi

The 3.0 MCi could 

be as high as 5.0 

MCi as noted below. 

Values in the chart 

reflect the current 

best estimate.

As stated earlier, approximately 33.8 Mgal of salt waste are currently stored in underground 

waste storage tanks at SRS. This waste, along with future salt waste receipts forecasted to be sent 

to the Tank Farms, will be processed through DDA, ARP/MCU, and SWPF. It is estimated that 

an additional 41.3 Mgal [3] of unconcentrated salt waste will be received by the Tank Farms 

between December 1, 2004 and the completion of salt waste processing in approximately 2019. 

After both liquid removal by processing through the Tank Farm evaporator systems and later 

additions of liquids for saltcake dissolution and chemistry adjustments required for processing, 

approximately 84 Mgal (5.9 Mgal existing salt waste through the DDA process, 1.0 Mgal future 

salt waste through the DDA process, 2.1 Mgal through ARP/MCU, 69.1 Mgal existing salt waste 

through SWPF, and 5.9 Mgal future salt waste through SWPF) of salt solution will be processed 

by Interim Salt Processing and SWPF resulting in approximately 168 Mgal of grout output from 

SPF to be disposed in SDF [3] (see Figure 2.2 below). 
February 28, 2005 

Page 12 of 94 

Figure 2.2: Salt Processing Volumes4

 [3] 

Salt Waste

33.8 Mgala

SWPF

ARP/MCU

DDA

Dissolution and/or 

Adjustment to 6.44 M Nad

SDF

Dissolution and/or 

Adjustment to 6.44 M Na

~5.9 Mgal 

~1.3 Mgal ~2.1 Mgal

~26.6 Mgal ~69.1Mgal

~8.6 Mgal

~95.8 Mgal

~2.8 Mgal

~13.4 Mgal 

Grout

~150.4 Mgal 

Grout

~4.4 Mgal 

Grout

H2Oc

Process H20

SPF

Future Waste 

Tank Receipts

Adjustment to 6.44 M Na

Salt Waste

41.3 Mgalb

Waste Tank Inventory

~1.0 Mgal 

~0.3 Mgal 

~38.1 Mgal 

~5.9 Mgal

a

Waste Tank Volumes are as of 12/01/2004

b Future Salt Waste Volumes are based on current projected unconcentrated waste tank receipts 

after 12/01/2004 through the remainder of Salt Processing. Volumes do not include dissolution 

liquid, adjustment liquid, or water added as part of normal operations (e.g., flushes, transfer jet 

dilution).

c Water will be removed through operation of Tank Farm Evaporators

d Salt solutions are adjusted to sodium (Na) molarity (M) of 6.44 

Process H20

Dissolution/

Adjustment Liquid

Dissolution/

Adjustment Liquid

Dissolution/

Adjustment Liquid

Each salt waste treatment process is described below. 

DDA PROCESS

The DDA process will be the first interim process used and will be used to process some of the 

lowest activity salt waste from 2005 until approximately 2009 when SWPF begins operation. In 

approximately 2007, ARP and MCU operations will be initiated to further process slightly higher 

activity salt waste. The DDA process will continue in parallel with the ARP and MCU facilities, 

handling the lowest activity salt waste. 

The DDA process involves the following: 1) removing the supernate from above the saltcake; 2) 

extracting interstitial liquid within the saltcake matrix; 3) dissolving the saltcake and transferring 

the resulting salt solution to a settling tank; and 4) transferring the salt solution to the Saltstone 

Facility feed tank where, if required, the salt solution is aggregated with other Tank Farm waste 

to adjust batch chemistry. Chemistry adjustment may be required to ensure the salt solution feed 

stream meets processing parameters (e.g., sodium concentration, organic content, facility 

shielding limitations) for processing at SPF. 

 

4

 Volumes shown are based on current knowledge of waste tank characterization data and process parameters. As 

processing continues and additional knowledge is gained, these volumes may be adjusted. 
February 28, 2005 

Page 13 of 94 

As shown in Figure 2.3, each waste tank undergoing the DDA process will initially contain salt 

waste consisting of both saltcake and supernate. Based on process knowledge to date, it is 

estimated that 30 percent of the saltcake volume is interstitial liquid [8]. 

Because of the relatively high solubility of Cs, approximately 95% [2] of the Cs-137 in the salt 

waste is found in the concentrated supernate solution and interstitial liquid. Cs-137 is the 

predominant radionuclide found in salt waste. Therefore, removal of supernate and interstitial 

liquid from the waste tank prior to dissolution of the saltcake will significantly reduce the total Cs 

inventory sent to SDF. The supernate and interstitial liquid removed during the DDA process 

will be transferred to another waste tank and stored for future processing through SWPF. The 

removal of the supernate and interstitial liquid is the “deliquification” phase of the DDA process. 

Figure 2.3: DDA Process (Initial Conditions, Steps 1 and 2) 

DDA Step #1

(Supernate Removal)

DDA Step #2

(Interstitial Pump Installation)

Initial Tank Conditions

Well 

Screen

Drain

Pump

Saltcake Saltcake Saltcake

Supernate

Transfer Jet

The first step in the deliquification phase (Figure 2.3, Step 1) is to remove the free supernate 

stored above the saltcake surface. Removal of the free supernate will generally be done using the 

existing tank transfer jet in each tank undergoing the DDA process. The free supernate will be 

transferred to another Tank Farm waste tank where it will be held for future treatment at SWPF. 

After completion of the free supernate removal, the transfer jet will be removed from the waste 

tank to allow equipment to be installed for deliquification of the tank interstitial liquid. The 

transfer jets typically installed in waste tanks operate at flow rates that are too high for interstitial 

liquid removal. To remove interstitial liquid, a pumping system must be installed as close to the 

tank bottom as possible and must operate at very low flow rates (less than one gallon per minute). 

In order to place a pump near the bottom of the tank, a well will be formed in the saltcake from 

the top of the saltcake surface to the bottom of the tank. The well is formed by dissolving a 

column of the solid saltcake, typically by using a high pressure water spray nozzle. As the well is 

formed, a stainless steel well screen, approximately 14 inches in diameter, will be lowered into 

the well. The well screen is the full height of the saltcake, which may be as high as 30 feet. The 

well screen protects the pump from collapse of the saltcake into the well. Once the well screen 

has been installed, a new transfer pump will be lowered to the bottom of the well screen (Figure 

2.3, Step 2). 

Removal of the interstitial liquid will be performed by operating the pump until the liquid level in 

the well is pulled down to pump suction level and flow can no longer be maintained. Then, as 
February 28, 2005 

Page 14 of 94 

interstitial liquid is removed from the saltcake, the well will refill with liquid and the pump will 

be turned back on to draw down the liquid level. This process will be repeated until there is no 

longer appreciable flow into the well (Figure 2.4, Step 3). The interstitial liquid that is removed 

from the tank will be sent to another Tank Farm waste tank and stored for future processing at 

SWPF. 

 

Figure 2.4: DDA Process (Steps 3 through 5) 

Dissolution

Pump

DDA Step #4

(Dissolution Pump Installation)

DDA Step #5

(Dissolution Water Addition)

Drained

Saltcake

Salt Solution

Saltcake

Dissolution

Water

DDA Step #3

(Interstitial Liquid Removal)

Saltcake 

Being Drained Dissolution

Pump

DDA Step #4

(Dissolution Pump Installation)

DDA Step #5

(Dissolution Water Addition)

Drained

Saltcake

Salt Solution

Saltcake

Dissolution

Water

DDA Step #3

(Interstitial Liquid Removal)

Saltcake 

Being Drained Drain

Pump

After interstitial liquid has been extracted to the point that the well recharge rate has become too 

slow to support the efficient removal of interstitial liquid, it is expected that approximately the 

bottom five feet of saltcake will remain fully saturated and approximately 50% of the interstitial 

liquid will have been extracted from the remainder of the saltcake, based on Tank 41 operational 

experience [8]. 

The next step in the DDA process is to dissolve the saltcake. This dissolution makes the saltcake 

into a mobile solution for processing, and allows the entrained solids to settle before the salt 

solution, minus the entrained solids that have settled, is transferred to the SDF. Since the 

interstitial liquid removal pump is designed to operate at only low flow rates, the pump must be 

replaced prior with a separate dissolution pump (Figure 2.4, Step 4). 

In order to dissolve the saltcake and create the salt solution batches, the tank will be filled with 

dissolution water, that is water chemically treated to prevent corrosion of the carbon steel waste 

tanks, (Figure 2.4, Step 5) until the saltcake surface has been flooded. This water dissolves a 

portion of the saltcake and a salt solution is formed. 

The next step in the process is transfer of the salt solution from the tank where it was dissolved to 

a second waste tank that acts as a settling tank (Figure 2.5, Step 6). 
February 28, 2005 

Page 15 of 94 

Figure 2.5: DDA Process (Steps 6 through 8) 

Insoluble

Solids

Salt Solution Salt Solution

DDA Step #7

(Salt Solution Settling)

DDA Step #8

(Adjustment)

DDA Step #6

(Salt Solution Removal)

Settling Tank SPF Feed Tank

Adjustment

Solution

Saltcake

After the salt solution has been transferred, the dissolution process will be repeated to continue to 

dissolve the remaining saltcake (Steps 5 and 6) until the desired batch volume has accumulated in 

the settling tank (typically approximately one million gallons). At that point, the insoluble solids 

that were transported with the salt solution will be allowed to settle in the tank (Figure 2.5, Step 

7). The solids include insoluble Sr and actinides so the effect of allowing them to settle is to 

remove some of these radionuclides from the salt solution before it is treated and stabilized for 

disposal. In the chemistry of the salt waste, actinides and Sr are predominately insoluble so 

removal of insoluble actinides and Sr through settling is expected to remove a significant portion 

of these radionuclides. 

After the salt solution has been held in the settling tank, the salt solution will then be decanted 

and transferred from the settling tank to Tank 50, the Saltstone Facility feed tank (Figure 2.5, 

Step 8). The settled solids will ultimately be transferred to DWPF for vitrification and disposal in 

the geological repository. Once the salt solution is in Tank 50, the salt solution may be 

aggregated with other Tank Farm waste to adjust batch chemistry. Chemistry adjustment may be 

required to ensure the salt solution feed stream meets SPF processing parameters (e.g., sodium 

concentration, organic content, facility shielding limitations) for processing at SPF. After 

confirmation through sample analysis, processing of the material through SPF is initiated. 

The DDA process, as described above, has been implemented up to the addition of the first batch 

of dissolution water (Figure 2.4, Step 5) on Tank 41 at SRS [8]. Removal of supernate from 

Tank 41 was undertaken in early 2002. Deliquification of Tank 41 occurred over a nine-month 

period beginning in September 2002. A total of approximately 150 thousand gallons (kgal) of 

interstitial liquid and 25 kgal of supernate were transferred to other SRS waste tanks for future 

processing at SWPF. Flow rates during the deliquification process ranged from an initial rate of 

approximately 10 gpm to a final rate of less than 0.5 gpm. At the completion of interstitial liquid 

removal, approximately 240 kgal of dissolution water were added to the tank, and an estimated 

350 kgal of new salt solution was formed. Dissolution water was added to Tank 41 in early July 

2003, and the tank has since been sitting idle pending resumption of salt disposition activities5

. 

 

5 Salt waste disposition activities were suspended in 2003 after a federal District Court in Idaho issued a decision 

striking down as exceeding the Secretary of Energy’s authority certain portions of DOE’s Order 435.1 and 
February 28, 2005 

Page 16 of 94 

In addition to the DDA activities already completed on Tank 41, two additional tanks will 

undergo deliquification and dissolution in support of the Interim Salt Processing Strategy. The 

tanks currently contemplated for deliquification and dissolution, in addition to Tank 41, are 

Tanks 25 and 28. Based on characterization information available today, those tanks are believed 

to contain some of the lowest activity material in all of the waste tanks at SRS [5]. Activities to 

characterize tank waste continue, and as additional knowledge of the dissolution process is 

gained, other tanks may be substituted for Tanks 25 and 28 if tanks better suited for DDA are 

identified. 

In addition, during the DDA part of Interim Salt Processing, DOE also plans to process a unique 

salt waste stream currently stored in Tank 48. This waste consists of approximately 0.24 Mgal 

[2] of relatively low-activity salt solution6

 containing approximately 19,000 Kg [9] of potassium 

and cesium tetraphenylborate (TPB) salts generated during an earlier unsuccessful effort to 

prepare salt waste for disposal, known as the In-Tank Precipitation (ITP) process. This process 

was briefly operated from 1995 through 1996 and was subsequently abandoned because DOE 

determined that the ITP process could not be safely operated at the necessary production rates. 

The ITP process resulted in generation of the TPB salts described above. The organic nature of 

TPB salts requires them to be stored separately from other tank waste. This is because as TPB 

breaks down into benzene and other organic compounds, it can form a potentially explosive 

mixture in the vapor space of a waste tank if not carefully managed. Other tanks are not 

equipped with safety systems required to manage this flammable mixture. Accordingly, all of the 

space in the 1.3 Mgal Tank 48 is being entirely used to store the 0.24 Mgal of TPB salts. In 

addition, this waste cannot be processed through DWPF because the breakdown of TPB in 

sufficient quantities in the DWPF melter could pose safety concerns. Currently, there is no 

practically available or contemplated technology that could be used to remove additional 

radioactivity and dispose of that radioactivity using DWPF. Accordingly, the waste in Tank 48 

will be processed without further removal of radionuclides by combining the Tank 48 stream 

with another salt waste stream, currently planned to be the low-activity liquid recycle waste 

stream from DWPF7

. The two waste streams will be aggregated to ensure the processing limits 

for allowable organic content at SPF are not exceeded. The aggregated low activity waste stream 

will then be transferred to the Saltstone Facility feed tank. 

 accompanying Manual. Natural Resources Defense Council v. Abraham, 271 F. Supp. 2d 1260 (D. Idaho. 2003). 

The provisions of the Order the court invalidated were those that set out criteria for classifying waste from 

reprocessing as “waste incidental to reprocessing” rather than “high-level radioactive waste.” DOE resumed salt 

waste disposition planning after Congress enacted section 3116 of the NDAA on October 28, 2004. That section 

clarified the Secretary’s classification authority with respect to certain waste from reprocessing, including the salt 

waste at issue here, and spelled out the criteria for making these classification decisions. Shortly thereafter, on 

November 5, 2004, the U.S. Court of Appeals for the Ninth Circuit reversed the District Court’s decision and 

remanded it with instructions to dismiss the case as unripe. Natural Resources Defense Council v. Abraham,388 

F.3d 701 (9th Cir. 2004), rehearing and rehearing en banc denied (9th Cir. 1/20/2004). 6

 It does not contain any saltcake. 

7

 The DWPF recycle waste stream is a generally very low-activity stream that consists of condensate from chemical 

processing and melter operation, waste from decontamination activities, and waste from miscellaneous drains and 

sumps in DWPF. Before the recycle stream is transferred to the Tank Farm, it is chemically adjusted with sodium 

nitrite and sodium hydroxide to ensure that Tank Farm WAC are met. 
February 28, 2005 

Page 17 of 94 

ARP AND MCU

The second interim process used for slightly higher activity salt waste will entail use of interim 

facilities to remove additional radionuclides, as well as cesium, before the resultant low-activity 

salt waste stream is sent to SPF for combination with a grout matrix and transfer to the SDF, 

where the waste-grout slurry hardens and is disposed of. The radionuclides that will be removed 

are primarily actinides (uranium (U), Pu, americium (Am), neptunium (Np), and curium (Cm)), 

strontium, and cesium. The removal of these radionuclides will be accomplished through ARP 

and MCU. 

ARP 

ARP will use modified existing facilities at SRS to remove Sr and actinides from some additional 

saltcake waste after the waste has undergone DDA. 

The ARP facilities will provide the capability to treat salt solution from the Tank Farms to 

remove soluble radioactive Sr and actinides by addition of MST and will remove insoluble Sr and 

actinides by cross-flow filtration. The ARP facilities will be operated during Interim Salt 

Processing and may also be used to augment actinide removal capacity during SWPF operations 

if needed. 

If sample analyses indicate that salt waste requires removal of soluble Sr and actinides in order to 

meet Class C concentrations limits in 10 CFR 61.55 in the grouted waste form, the waste will be 

received into either of the two MST Strike Tanks. Waste received in MST Strike Tank #1 or #2 

will be adjusted with water to approximately 5.6 Molar sodium concentration to provide optimum 

conditions for sorption of Sr and actinides onto MST. Following the addition of MST to either 

Strike Tank, the contents will be agitated for a reaction period between 4 and 24 hours based on 

the curie concentration of the soluble actinides to be removed. The resulting slurry will be 

transferred from either of the strike tanks into the Filter Feed Tank (FFT). If sample analyses 

demonstrate that decontamination of the salt solution to meet Class C concentration limits in the 

grouted waste form can be achieved without removal of soluble actinides and Sr, then the waste 

will be transferred without MST treatment from the Tank Farm directly to the FFT for ARP 

filter-only processing. 

The batch received in the FFT will be circulated through a cross-flow filter to remove and 

concentrate the insoluble solids and the MST solids loaded with Sr and actinides, if MST 

treatment of the waste was performed. The cross-flow filtration process will yield a Clarified 

Salt Solution (CSS) which will be received in the Filtrate Hold Tank (FHT). The solids will be 

concentrated in the FFT over a number of batches depending on the entrained solids loading of 

the waste incoming to ARP. The concentrated solids heel in the FFT will then be washed and 

filtered to remove soluble sodium salts. The washed MST/sludge slurry will be transferred to 

DWPF for vitrification processing. 

The CSS and spent wash water from the solids washing operation will then be transferred from 

the FHT to the Saltstone Facility feed tank for transfer to the Saltstone Facility, if the waste has a 
February 28, 2005 

Page 18 of 94 

sufficiently low cesium concentration to meet the Saltstone Facility WAC. The Saltstone Facility 

WAC will always require Cs concentrations to be below Class C limits. If the waste is a higher 

Cs salt waste, the CSS will be sent directly to MCU during Interim Salt Processing or to SWPF 

once SWPF is operational. 

Figure 2.6: Actinide Removal Process 

MST

STRIKE 

TANKS

HIGHER ACTINIDE 

WASTE FEED

MST/SLUDGE TO

DWPF

CAUSTIC

ALPHA 

SORPTION FILTER

CLARIFIED

SALT SOLUTION (CSS) SALT SOLUTION (CSS)

To MCU

FILTER 

FEED

TANK

MST LOW ACTINIDE

WASTE FEED

MCU 

MCU will utilize the CSSX process to remove Cs from the salt waste. The solvent extraction 

process will use an organic solvent to complex with (or tie up) Cs atoms in the waste stream. The 

solvent and salt solution will then be fed to a bank of centrifugal contactors which will ensure 

complete mixing of the waste and solvent and extraction of the solvent and Cs from the salt waste 

stream. The Cs will then be stripped from the solvent in another bank of contactors and sent to 

DWPF for vitrification. The solvent will then be recycled. The Decontaminated Salt Solution 

(DSS) resulting from solvent extraction will be sent to the Saltstone Facility feed tank. SRS is 

constructing this small scale solvent extraction capability to be operational in approximately 2007 

and it will process approximately 2.1 Mgal of salt solution [5]. 

MCU will be a short-term Cs removal process that will be operated downstream of ARP before 

SWPF becomes operational. MCU will utilize the same technology as that to be used by SWPF 

but its decontamination capability will be less than that of SWPF since MCU will only be used to 

process lower curie material. Once SWPF becomes operational, the current plan is to shut down 

MCU. 

As shown in Figure 2.7, the CSSX process will use multiple centrifugal contactors to provide the 

necessary contact and separation of the aqueous and organic streams. Current plans are to use 

seven stages for extraction, two stages for scrubbing, seven stages for stripping, and two 

additional stages for solvent washing. 
February 28, 2005 

Page 19 of 94 

Figure 2.7: Modular CSSX Unit Process 

CAUSTIC WASH

CONTACTORS

CSS

from ARP from ARP Ba 

DECAY

& 

HOLD

TANKS

DSS to

Saltstone Saltstone

SOLVENT

HOLD 

TANK

EXTRACTION CONTACTORS

(Cs Removal)

SCRUB SOLUTION

SCRUB CONTACTORS

STRIP SOLUTION

(NITRIC ACID) (NITRIC ACID)

Cs Waste Stream Cs Waste Stream

STRIPPING CONTACTORS

(Cs / Organic Solvent Separation)

to DWPF

Cs in Organic Solvent

Recycled Organic Solvent

CAUSTIC WASH CAUSTIC WASH

The extraction section will use the organic solvent to extract Cs (as well as some sodium and 

potassium) from the salt solution. The scrub section will then return the sodium and potassium to 

the salt solution leaving the Cs in the solvent and will neutralize any hydroxide carried over in the 

solvent. The strip section will return the Cs to a separate aqueous phase that will be sent to 

DWPF for vitrification and will allow the solvent to be reused in the process. 

The solvent will be made up of four components: an extractant based on a modified calixarene 

molecule designed to complex with Cs, a modifier that will improve the solubility of the 

extractant, a suppressor that will counter the negative effects of anionic impurities in the waste, 

and a diluent. The solvent will be contacted with the alkaline waste stream in a series of 

countercurrent centrifugal contactors (the extraction stages). The resulting DSS will be 

transferred to the Saltstone Facility via the Saltstone Facility feed tank for disposal. Following 

Cs extraction, the solvent will be scrubbed with dilute acid to remove other soluble salts from the 

solvent stream (the scrub stages). The scrubbed solvent will then pass into the strip stages where 

it will contact a very dilute acid stream to transfer the Cs to the aqueous phase. Based on 

volumetric flow ratios of the solvent extraction streams, the Cs concentration will increase by a 

factor of 12 to 15 over the salt feed solution to minimize the aqueous flow to DWPF. The 

aqueous strip effluent containing the Cs removed will be transferred to DWPF. 
February 28, 2005 

Page 20 of 94 

SWPF

DOE is designing and will construct a large scale SWPF. SWPF will utilize the same 

technologies as ARP and MCU to remove Sr, actinides, and Cs from the salt waste. However, 

due to its large scale, SWPF will provide a much higher decontamination factor (DF)8

 for Cs than 

the MCU, about 40,000 and 12, respectively, and DFs for Sr and actinides is similar to or higher 

than the interim ARP process that will be deployed earlier. SWPF also will have a much higher 

capacity (6 to 9 Mgal per year [4]) than ARP and MCU and will be able to treat all of the 

remaining salt waste including the supernate and interstitial liquid which start out with 

significantly higher concentrations of Cs than the small amount of low-activity salt waste that 

DOE anticipates treating through the interim processes. SWPF is planned to be completed and 

become operational in approximately 2009, and will consist of three basic unit operations: the 

Alpha Strike Process (ASP), CSSX, and the Alpha Finishing Process (AFP). 

ASP will be the initial processing stage of SWPF, operating in a batch mode to remove soluble Sr 

and actinides from the waste feed by sorption onto MST. The resultant slurry will be filtered to 

concentrate the MST and insoluble solids. The resulting filtrate, or CSS, will be sent forward to 

CSSX and the concentrated MST/sludge will be washed and sent to DWPF as shown in Figure 

2.8. 

Figure 2.8: SWPF Alpha Strike Process 

ALPHA

SORPTION 

TANK

WASTE 

FEED

MST/SLUDGE TO

DWPF

MST

CAUSTIC

ALPHA 

SORPTION FILTERS

CLARIFIED CLARIFIED

SALT SOLUTION (CSS) SALT SOLUTION (CSS)

To CSSX

FILTER 

FEED

TANK

The second SWPF processing operation will use the CSSX process to remove Cs from the CSS. 

Based on the current design, this will be a continuous flow process, utilizing centrifugal contactor 

stages for extracting (16 stages), scrubbing (2 stages), stripping (16 stages), and washing (2 

stages) CSS streams as shown in Figure 2.9. Cs will be captured by contacting the aqueous salt 

solution with a specially engineered extractant (BOBCalixC6) which will be part of a four 

component organic solvent that will also include a modifier that will improve the solubility of the 

 

8

 DF is defined as the quotient of the influent concentration of a given radionuclide and the effluent concentration of 

that same radionuclide. 
February 28, 2005 

Page 21 of 94 

extractant, a suppressor that will counter the negative effects of anionic impurities in the waste, 

and a diluent. The Cs then will be stripped from the organic solvent by contacting with a dilute 

nitric acid strip solution and separating the phases. The scrub and wash stages will be used to 

condition and purify the solvent for recycling. The strip effluent (high Cs concentration) is sent 

to DWPF for vitrification. After allowing for barium (Ba)-137m decay (half-life of 

approximately two minutes), the DSS, which will be comprised of the aqueous effluent raffinate 

from the extraction stages and aqueous wash effluent from the wash stages, will be sampled to 

confirm that process requirements have been met. The DSS then will be sent to the Saltstone 

Facility feed tank or to AFP if additional Sr and actinide removal is required. 

Figure 2.9: SWPF CSSX Process 

 

To Saltstone Disposal 

Facility 

Clarified Solution

From Alpha/Sr 

Removal

Extraction

Organic 

Recovery

To DWPF Feed 

Tank

Organic 

Recovery

Solvent 

Hold 

Tank

Scrub Strip

Scrub 

Makeup

Strip 

Makeup

Caustic Wash

Caustic Wash Makeup

Strip 

Effluent 

Hold 

Tank

AFP will be the final SWPF unit operation in which an additional MST strike may be performed. 

This unit operation will be located downstream of CSSX and will remove Sr and actinides not 

removed in ASP. AFP will be similar to ASP in equipment/tank sizes, configuration, and 

operational sequence. AFP will provide the benefit of additional MST strike capability without 

any reduction in SWPF throughput and allows the equipment to be located in a contact-handled 

area of the plant. This will provide further assurance that the processed salt waste will meet the 
February 28, 2005 

Page 22 of 94 

Saltstone Facility WAC limits, even if the waste feed has higher than expected concentrations of 

actinides or Sr than previously anticipated. 

As mentioned previously, SWPF will be capable of processing 6 to 9 Mgal of salt solution per 

year after the first year of operation with a nominal annual processing rate of approximately 7 

Mgal. SWPF is scheduled to begin hot commissioning activities in 2009 followed by one year of 

hot operations with a planned processing rate of 5 Mgal [4]. SWPF is expected to continue to 

operate until the completion of salt processing, approximately 2019. 

SALTSTONE FACILITY

As discussed earlier, DOE intends to use a low-level waste disposal facility known as the 

Saltstone Facility for final treatment, stabilization and disposal of the salt waste. This facility, 

located in Z-Area of the SRS site, consists of the SPF and SDF facilities. 

Construction of SPF and the first two vaults of SDF were completed between February 1986 and 

July 1998. The Saltstone Facility started radioactive operations June 1990. Since that time, it has 

continued to be operated on an intermittent, as-needed basis to treat and dispose of waste 

concentrate from the F- and H- Area Effluent Treatment Facility as well as low-level waste from 

the Naval Fuel Material Facility. 

SPF will receive and mix the low-activity salt solution from the DDA, ARP/MCU, and SWPF 

processes with cement, flyash, and slag to form a homogeneous grout mixture or slurry. The 

slurry will be mechanically pumped into engineered disposal vaults, located in SDF, where the 

grout solidifies into a monolithic, non-hazardous, solid low-level waste form called “saltstone.” 

The salt waste from the tanks has higher activity levels than the very low-activity waste the 

Saltstone Facility has been used to treat and dispose of to date. The Saltstone Facility was 

originally designed to dispose of very low-activity waste with concentrations less than Class A 

limits. Therefore several modifications in the design and operation of the facilities are being 

undertaken to prepare them for receipt of the low-activity salt solutions produced by Interim Salt 

Processing. To minimize radiation exposure to SPF workers during normal operations, additional 

shielding will be added to the major pieces of SPF process equipment, including the salt feed line 

to SPF and the grout line running from the process to the SDF disposal vaults. In addition, 

several inches of uncontaminated grout will be poured on top of the saltstone grout in the inactive 

vault cells to reduce the amount of exposure to facility workers9

 due to sky shine (a phenomenon 

where gamma radiation is reflected off molecules in the air and back toward the earth). 

Equipment is also being redesigned or replaced to minimize the impact of an SPF grout delivery 

system failure. This includes redesign to facilitate removal of affected system components 

without significant hands-on requirements. 

As presently planned, SDF will contain several large concrete vaults divided into cells. Each of 

the cells will be filled with solid saltstone grout. The saltstone grout itself provides primary 

 

9

 The facility workers will be primarily in the SPF rather than the SDF during normal operations. 
February 28, 2005 

Page 23 of 94 

containment of the waste, and the walls, floor, and roof of the vaults provide secondary 

containment. Existing vaults will be used for the low-activity salt waste resulting from Interim 

Salt Processing. Additional vaults will be added as needed to receive the low-activity salt waste 

on a just-in-time basis. A more detailed discussion of these facilities may be found in Appendix 

1. 

The Saltstone Facility is located approximately one mile from the salt processing facilities and 

approximately 6.2 miles from the nearest SRS site boundary. Z-Area, where the Facility is 

located, lies on a local topographic high at approximately 290 feet above mean sea level [10]. 

This site was selected for low-level waste disposal because of its location on a well-drained 

topographic high, as evidenced by the lack of marshes or other bodies of standing water. Z-Area 

is bounded by two streams: Upper Three Runs and McQueen Branch, a tributary to Upper Three 

Runs. McQueen Branch is located approximately 1.0 miles to the East and 0.75 miles to the 

Northeast of Z-Area, and Upper Three Runs is located approximately 1.0 miles to the Northwest 

[10]. Under Z-Area, the minimum depth to the water table from the ground surface in any given 

year is estimated to be 43 feet (i.e., at 247 feet above mean sea level) on the basis of water table 

fluctuations [11]. Per South Carolina Department of Health and Environmental Control 

(SCDHEC) Regulation R.61-107.16, the bottom of all current and future disposal vaults must be 

located at least two feet above this historic high water table. 

There are no lakes or flow control structures on Upper Three Runs or its tributaries. The 

Probable Maximum Flood value for Upper Three Runs is 175 feet above mean sea level, which is 

substantially below the planned maximum depth of the SDF vaults [10]. The 500-year and 100-

year flood plains are located at 153.1 feet and 151.7 feet above mean sea level, respectively [12]. 

Two major earthquakes have occurred within 100 miles of SRS. The largest known earthquake 

to affect SRS was the Charleston earthquake of 1886, with an epicenter approximately 90 miles 

from SRS and a magnitude of 6.6 on the Richter Scale. It is estimated that an earthquake of this 

magnitude would result in a peak ground acceleration of 0.10g at SRS [13]. A seismic evaluation 

of Z-Area shows that the soils beneath Z-Area are not susceptible to significant liquefaction for 

earthquakes having a peak ground acceleration less than or equal to 0.17g [14]. The second 

earthquake occurred approximately 90 – 100 miles from SRS, with an estimated magnitude of 

4.5. 

RADIONUCLIDE REMOVAL TECHNOLOGIES

The DDA process was developed to treat the lowest activity salt waste streams because it utilizes 

readily available technologies that could be implemented in the near term. Technologies from the 

chemical, mining, geology, and other industries were adapted for the DDA process to provide the 

capability to remove the cesium-bearing interstitial liquid from the saltcake [8]. Other options to 

improve deliquification effectiveness were considered. The sheer size of the waste tanks 

(typically 85 feet in diameter and 35 feet tall) and the process cooling coils (typically over a mile 

of piping within each individual tank) adds great complexity and limits the options. The 

structural design of SRS waste tanks precludes positively pressurizing them to force liquids from 

the solid saltcake. In addition, such positive pressure could result in a significant release of 
February 28, 2005 

Page 24 of 94 

airborne radioactivity to the environment. Also, options to use additional fluid to improve Cs 

removal efficiencies were considered. Methods to displace the existing Cs-laden interstitial 

liquid were not viable due to tank configuration and chemistry constraints (fluids that would 

displace the liquid would also dissolve the saltcake). Flushing of the saltcake after initial 

deliquification also was not viable due to tank space limits and the same chemistry constraints. 

 

The ARP and MCU facilities are based on the same technologies that will be implemented in 

SWPF. They have been scaled down to accelerate their availability and to allow the processes to 

fit within existing facilities at SRS. In addition, these small scale facilities offer DOE the 

opportunity to demonstrate the technologies that will be implemented on a large scale in SWPF. 

The identification, evaluation, and selection of the technologies for Sr, actinide, and Cs removal 

for SWPF were the subject of numerous studies and reviews. Starting in 1998, more than 140 

technologies were evaluated and considered [6]. In 2000 and 2001, the National Academy of 

Sciences reviewed DOE’s technology evaluations and selections [15]. This review validated the 

DOE technology evaluation and selection process and provided input for continuing technology 

development. DOE continued to invest in technology development for the alternatives 

considered and down-selected to CSSX and MST in late 2001. The Salt Processing Alternatives 

SEIS [6] was the basis for DOE’s selection of MST and CSSX as the technologies to be 

implemented in SWPF. 

The solidified salt solutions arising from the salt waste treatments described above will be safely 

disposed of as LLW at SDF. The higher activity fractions, separated from the waste by the DDA 

process, ARP/MCU, and SWPF and containing nearly all radioactive inventories (close to 99%), 

will be immobilized at DWPF in a vitrified glass waste form for disposal at a Federal geologic 

repository. Appendix 1 contains a system description, which includes a brief history and 

background on waste generating activities and a description of facilities used at SRS for waste 

storage, evaporation, waste removal, treatment, vitrification, and disposal activities. 
February 28, 2005 

Page 25 of 94 

3.0 SECTION 3116 (A) OF THE RONALD W. REAGAN NATIONAL DEFENSE 

AUTHORIZATION ACT FOR FISCAL YEAR 2005 (NDAA) 

Section 3116 (a) of the NDAA provides: 

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 

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

Commission. 
February 28, 2005 

Page 26 of 94 

As demonstrated in the following sections, the solidified DSS will not exceed the concentration 

limits for Class C waste in Code of Federal Regulations (CFR) 10 CFR 61.55 [16]. Therefore, 

the provisions of NDAA Section 3116 (a)(3)(B), concerning waste which exceeds Class C 

concentration limits, are not addressed in this draft 3116 Determination. 
February 28, 2005 

Page 27 of 94 

4.0 THE 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 waste from 

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[.] 

The purpose of Section 3116 is to clarify the Secretary’s authority, in consultation with the NRC, 

to determine that certain waste from reprocessing that meets the criteria set out in the section 

should not be classified as “high-level radioactive waste.” It sets out two specific criteria for 

making this determination in clauses (2) and (3). Clause (2) requires 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 

low level waste(See 10 CFR. 61.55 and 61.58). This includes waste that falls within one of the 

classes set out in Section 61.55, as well as waste that will be disposed of so as to meet the 

performance objectives of subpart C of Part 61. 

Clause (1) is a broader criterion that requires 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 speaking, this would be an unusual case. This is because waste 

that meets the third criterion would be waste that the Secretary, in consultation with the NRC, has 

already determined either falls within one of the classes set out in section 61.55 that the NRC has 

specified are considered “generally acceptable for near-surface disposal” or that the Secretary, in 

consultation with the NRC, has already determined will be disposed of in a manner that meets the 

Part 61 subpart C performance objectives. These latter objectives, in turn, as the NRC recently 

explained, “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.” 

In the Matter of Louisiana Energy Services (National Enrichment Services), CLI-05-05, slip op. 

at 11 (Jan. 18, 2005). 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 the public health and safety. 

That said, 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 
February 28, 2005 

Page 28 of 94 

considerations that warrant its disposal in a deep geologic repository. Clause (1) is an 

acknowledgment by Congress of that possibility. For example, the waste stream could contain 

material that, while not presenting a health and safety danger if disposed of at 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 the NRC, the authority to consider such factors in determining 

whether waste that meets the other two criteria nonetheless requires disposal in a deep geologic 

repository in light of such considerations. 

That is not the case here. As is demonstrated later in this document, the solidified low-activity 

salt waste falls within Class C limits and hence falls within one of the categories that the NRC 

has concluded is generally appropriate for near-surface disposal. Moreover, as is also 

demonstrated later in this document, its disposal in SDF will meet the performance objectives of 

Part 61. Therefore, the disposal of this waste in SDF will provide for the protection of the public 

health and the environment. Accordingly, the waste does not require disposal in a deep geologic 

repository by reason of risk to the public health and safety. Furthermore, disposal of the 

solidified low-activity salt waste in SDF does not raise any unique considerations that, 

notwithstanding these demonstrations, nevertheless require its permanent isolation in a deep 

geologic repository. Accordingly, the solidified low-activity salt waste qualifies for classification 

as other than high-level waste under the first clause of subsection (a). 
February 28, 2005 

Page 29 of 94 

5.0 THE WASTE HAS HAD HIGHLY RADIOACTIVE RADIONUCLIDES REMOVED TO 

THE MAXIMUM EXTENT PRACTICAL

Section 3116 (a)(2) of the NDAA provides in pertinent part: 

 [T]he term “high level 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 RADIONUCLIDES CONSIDERED

For the purpose of this draft 3116 Determination, the radionuclides considered in detail are: Cs137 (including its daughter, Ba-137m), Sr-90 (including its daughter, yttrium (Y)-90), the 

actinides (isotopes of U, Pu, Am, Np, and Cm), selenium (Se)-79, technetium (Tc)-99, iodine (I)-

129, and tin (Sn)-12610. 

The short-lived fission products Cs-137 and Sr-90 and their equilibrium daughters, Ba-137m and 

Y-9011, are by far the predominant sources of radioactivity present in the SRS salt waste. Based 

on process and sampling knowledge as reflected in the current WCS database, more than 99% [2] 

of the current radioactivity in the SRS tank salt waste is associated with these two radionuclides 

and their daughters. Indeed, Cs-137 and its daughter Ba-137m alone account for over 95% of the 

 

10 DOE has reviewed the inventory of 41 radionuclides in the salt waste in the SRS waste tanks, as reflected in the 

current Waste Characterization System (WCS) database. WCS is discussed in further detail in Appendix 2. The 

radionuclides considered in detail for removal in this draft 3116 Determination take into account scientific expertise, 

knowledge, and health physics principles as applied to the SRS salt waste, and include those radionuclides in Tables 

1 and 2 in 10 CFR 61.55 that are in the SRS salt waste and are in quantities such that they may be important to 

meeting the performance objectives in 10 CFR 61, Subpart C. All radionuclides in Tables 1 and 2 are considered in 

Section 6.0 and , where relevant, Section 7 of this draft 3116 Determination. However, radionuclides with less than 

five-year half-lives, H-3, C-14, Co-60, and Ni-63 are present in concentrations so low (well below Class A 

concentration limits) that they are not discussed for the purposes of removal in Section 5.0. The radionuclides 

considered in detail for the purposes of removal also include other radionuclides that are not in the above referenced 

tables that may be important to meeting the performance objectives in 10 CFR 61, Subpart C. Some of the 

radionuclides considered in detail in this draft 3116 Determination may not be considered in detail in other 3116 

Determinations if the circumstances or waste forms do not warrant such consideration. 

11 Cs-137, and its daughter Ba-137m, are typically considered as a single radionuclide for human health protection 

purposes because the half-life of Ba-137m is so short that it only exists when Cs-137 is present. The same is true for 

Sr-90 and its daughter Y-90. Accordingly, the discussions that follow in this draft 3116 Determination focus on Cs137 or Sr-90 since approaches that are effective in removing Cs-137 and Sr-90 also remove Ba-137m and Y-90, 

respectively. 
February 28, 2005 

Page 30 of 94 

salt waste’s radioactivity [2]. Moreover, Cs-137, Sr-90, and their daughters are present in 

sufficient concentrations in the salt waste so that without shielding and controls they produce 

radiation emissions that would present risk to humans simply due to their proximity, without 

direct inhalation or ingestion. Accordingly, they are of potential acute hazard to occupational 

workers, the public, and the environment. 

Although the actinides in the concentrations present in the salt waste (e.g., U, Pu, Am, Np, and 

Cm) do not require distance and shielding to protect workers and the public, these radionuclides 

contain the majority of the long-lived isotopes that may pose the greatest risk in the future to 

human health because of their long life and because they present human health risk if directly 

inhaled or ingested. Other radionuclides considered in detail are those that persist in the waste 

form well into the future due to their long lives, are relatively mobile in the environment, and 

may present a health risk if ingested. These radionuclides are Se-79, Tc-99, I-129, and Sn-126. 

These two groups of radionuclides (i.e., the actinides, Se-79, Tc-99, I-129, and Sn-126) together 

account for less than 1% [2] of the radionuclide inventory in the tank salt waste. 

As explained above, the radionuclides considered in detail in this draft 3116 Determination are 

based on, and take into account, scientific and health physics principles, knowledge, and 

expertise. In addition, the radionuclides considered in detail account for over 99.9% [2] of the 

current radioactivity in the SRS tank salt waste, based on process and sampling knowledge. 

5.2 REMOVAL TO THE MAXIMUM EXTENT PRACTICAL

Section 3116 (a) of the NDAA provides that certain waste resulting from reprocessing is not 

high-level waste 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.” This section discusses the basis on which the Secretary may conclude that DOE’s 

two-phase, three-part strategy for removing the radionuclides considered above meets this 

criterion. 

Removal to the extent “practical” is not removal to the extent “practicable” or theoretically 

“possible.” Rather, a “practical” approach to removal is one that is “adapted to actual 

conditions,”12 “adapted or designed for actual use; useful,”13 a method that is selected “mindful 

of the results, usefulness, advantages or disadvantages, etc., of [the] action or procedure.”14 The 

considerations that bear on whether radionuclide removal will be accomplished to the maximum 

 

12 A Dictionary of Modern English Usage 453, H. Fowler (1930). 13 Random House Unabridged Dictionary, Copyright © 1997, by Random House, Inc., on Infoplease 

(www.infoplease.com/ipd/A0598638.html). 14 Id. definition 7. See also Cambridge Dictionary of American English, Cambridge University Press 2004 

(http://dictionary.cambridge.org/define.asp?key=practical*2+0&dict=A): “practical (EFFECTIVE): fitting the needs 

of a particular situation in a helpful way; helping to solve a problem or difficulty; effective or suitable”. 

It is notable that while prior NRC and Department criteria for waste incidental to reprocessing called for removal “to 

the maximum extent technically and economically practical,” the statute omits the 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. 
February 28, 2005 

Page 31 of 94 

extent practical will therefore vary depending not only on the theoretically possible or available 

technologies that may be deployed but also the overall costs and benefits, not only economic but 

more broadly considered, of deploying them with respect to a particular waste stream. The 

“maximum extent practical” standard contemplates room for the exercise of expert judgment in 

weighing, for example, 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 public health and considerations that may ensue from delay; the 

reasonable availability of proven technologies; the usefulness of such technologies; and the 

sensibleness of utilizing such technologies. What may be removal to the maximum extent 

practical at one point in time may not be that which, on balance, is practical, feasible, or sensible 

at a prior or later point in time. 

DOE’s two-phase, three-part plan for processing the salt waste will result in removal of highly 

radioactive radionuclides to the maximum extent practical, bearing in mind the considerations 

outlined above. In all, the plan will result in removal of approximately 218 to 220 MCi of the 

approximately 223 MCi [2] present in the SRS salt waste. An estimated 3 to 5 MCi, mostly the 

short-lived radionuclide Cs-137 and its daughter Ba-137m, will be solidified and disposed at 

SDF. DOE’s current best estimate is that 3 MCi will be disposed of in SDF [4]. Collectively, 

DDA, ARP, MCU, and SWPF will remove approximately 98.7% of the curies from salt waste for 

vitrification in DWPF and ultimately permanent disposal in a deep geologic repository. 

Because of the techniques used to adjust the chemistry of the tank waste to prevent corrosion, the 

salt waste starts out with low concentrations of most of the radionuclides from reprocessing, 

because approximately 48% of the radioactivity, including by far the vast majority of the longlived actinides and Sr-90, is in the metals that precipitated to the bottom of the tanks to form the 

sludge waste. To remove additional radionuclides from the salt waste, DOE will process by far 

the greatest volume of this waste using SWPF, which will remove over 99% of the Cs-137 and its 

daughter product, Ba-137m, the majority of the Sr and actinides, as well as other radionuclides. 

In addition, DOE will use the first-phase, interim technologies to process a limited amount of salt 

while SWPF is being designed and constructed. These interim technologies will also remove 

significant quantities of the radionuclides considered, although less than SWPF will remove. 

Proceeding with interim processing using these technologies is necessary, however, if DOE is to 

be in a position to continue removal and vitrification of the high-activity sludge between now and 

the start-up of SWPF, including removing sludge waste from the tanks that lack full secondary 

containment, and to operate SWPF effectively after its construction is complete. The only 

practical way DOE will be able to move forward with these activities without significant 

interruption and delay is by using the interim technologies to remove and dispose of a limited 

amount of the salt waste currently in the tanks during this interim period. The alternative would 

be either: (1) to shut down sludge removal and vitrification activities now, or (2) to continue with 

those activities knowing that the effect will be to force a DWPF shutdown some time around 

2008. At that time, DOE will no longer have the tank space needed to prepare the feed for 

DWPF and will require several years of SWPF operation (operating, at first, well below capacity, 

because it too requires significant tank space to run efficiently) before DOE will have enough 

space to be able to resume full-scale DWPF operations. Either way, a multi-year delay in 
February 28, 2005 

Page 32 of 94 

removing and vitrifying the sludge and salt waste – the prerequisite to closing the tanks, 

including those that lack full secondary containment – will result. That is not in the interest of 

the public health and safety or the environment. 

Given the significant disadvantages of waiting to begin processing salt waste until SWPF is 

available, the preferable course from the perspective of the public health, safety, and the 

environment is, during this interim period, to process a minimal amount of the lowest activity salt 

waste using technologies that are the best available now and in the near future to remove 

significant amounts of radionuclides from some of the lowest-activity salt waste. After removal 

of these radionuclides, the solidified low-activity salt waste will be below Class C concentration 

limits and its disposal in SDF will meet 10 CFR Part 61, Subpart C, performance objectives. In 

light of all these considerations, DOE’s overall salt strategy, including the use of the interim 

technologies to process only enough salt to create enough tank space to allow uninterrupted 

sludge removal and treatment and efficient operation of SWPF, results in removal of highly 

radioactive radionuclides from the salt waste to the maximum extent practical. 

5.2.1 Salt Waste Processing Facility 

SWPF is the cornerstone of the salt processing strategy. Of the approximately 84 Mgal of salt 

solution that must be processed from the salt waste, approximately 75 Mgal [3] will be processed 

through SWPF. This facility will be capable of processing 6 to 9 Mgal of salt solution per year 

after the first year of operation with a nominal annual processing rate of 7 Mgal. SWPF will 

become operational in approximately 2009 with a planned processing rate of 5 Mgal during the 

first twelve months of operation. 

DSS produced by SWPF is anticipated to be below Class A concentration limits for Cs-137, Sr90, and actinides, and will add less than 0.2 MCi [4] to the SDF inventory through approximately 

2019, the projected end of salt treatment and disposal. 

SWPF uses specific processes that have been developed and that will be the state-of-the-art 

methods to target the removal of Cs-137, Sr-90, and actinides from SRS salt wastes. SWPF will 

remove approximately 99.998% [17] of the Cs-137/Ba-137m activity while also removing 

insoluble strontium and actinides. In addition, SWPF is anticipated to reduce the soluble 

concentrations of Pu, Np, U, Am, and Sr by approximately 82%, 58%, 26%, 78%, and 95% [17], 

respectively15. 

Although the SWPF processes do not specifically target the preferential removal of Se-79, Tc-99, 

Sn-126, and I-129, SWPF’s filtration process will also remove a majority of the insoluble species 

of these radionuclides16. A review of technology options for removal of similarly low 

 

15 SWPF removal percentages shown are based on a single 12-hr MST strike. The operational goal of SWPF will be 

to achieve a Class A waste stream after solidification. At this time, a minimum of one strike is anticipated for all 

waste batches. Reduction of the soluble concentration of actinides and Sr is dependent on the concentration of the 

incoming waste stream, duration of the MST strike, and the number of strikes. Some waste batches will use more 

than one strike to achieve the goal of Class A concentrations. 

16 The majority of the insoluble species of Se-79, Tc-99, Sn-126, and I-129 will be removed via filtration during the 
February 28, 2005 

Page 33 of 94 

concentrations and inventories of these isotopes was performed at the DOE Hanford Site [18]. 

This review did not identify any non-bench-scale, technically practical technologies to isolate and 

further remove these isotopes. As discussed previously, the inventories of these four 

radionuclides are relatively small when compared to the overall curie inventory of the salt waste. 

Although these specific radionuclides are long-lived and relatively mobile in the environment, the 

future dose hazard from the disposal of the entire inventory of these radionuclides in SDF is 

below the doses set forth in the performance objectives in 10 CFR Part 61, Subpart C. 

Accordingly, further isolation and removal of Se-79, Tc-99, Sn-126, and I-129 is not necessary or 

practical. 

5.2.2 Interim Salt Processing 

In addition to SWPF, as explained earlier, DOE’s salt processing strategy also contemplates use 

of interim technologies (DDA, ARP, and MCU) to process just enough of some of the lowest 

activity salt waste to create enough tank space to allow DOE to continue removing and vitrifying 

sludge and to operate SWPF effectively immediately upon becoming operational. These interim 

technologies will not remove as much radioactivity from the salt waste as SWPF. However, they 

will remove significant amounts. Moreover, DDA is available now and ARP and MCU will be 

available in approximately 2007, sooner than SWPF, which will not be available until 

approximately 2009. They are the best removal technologies available to DOE during this 

interim period. 

Rather than wait until SWPF becomes available, DOE believes the better course is for it to 

proceed to use these technologies to remove radionuclides from and process a minimal amount of 

some of the lowest activity salt waste so that it can dispose of these salt wastes at SDF. This will 

allow DOE to create the tank space necessary to continue cleanup of the sludge and operate 

SWPF effectively once that facility has been constructed. In DOE’s view, use of these interim 

technologies is the proper course of action because the public health, safety, and environmental 

risks and disadvantages that would result from waiting until SWPF is available and significantly 

delaying further sludge removal and full-scale operation of SWPF once it has been constructed 

plainly outweigh the negligible public health, safety, and environmental disadvantages from 

proceeding with disposal of this waste after using the interim technologies to remove those 

radionuclides that can be removed now. Accordingly, use of these interim technologies 

accelerates risk reduction through processing the minimal amount of the lowest activity salt waste 

to create the necessary tank space for continued sludge removal and earlier full SWPF operation 

resulting in removal of highly radioactive radionuclides to the maximum extent practical during 

that timeframe. 

5.2.2.a The Need for Interim Salt Processing 

As explained briefly above, the reason DOE sought to develop an interim salt processing strategy 

is that if DOE were to wait to begin processing salt waste until approximately 2009, when SWPF 

 ARP process or the SWPF process and sent to DWPF where they will be vitrified in glass and eventually disposed of 

in a deep geologic repository. 
February 28, 2005 

Page 34 of 94 

is expected to come online, DOE would be forced to decrease, postpone, and eventually halt the 

on-going activities to remove and stabilize tank waste that currently are reducing risk to the 

occupational workers, the public, and the environment. Chief among these risk-reduction 

activities is the processing of sludge waste in DWPF, including its removal from the aging waste 

tanks that lack full secondary containment (old-style tanks) 17. 

As described in Appendix 1, the underground waste storage tanks have been receiving waste 

since 1954. The last of the 51 waste storage tanks was placed into operation in 198618. Since 

that time, waste receipts have continued and waste processing (sludge removal, sludge 

processing, and immobilization of sludge in glass in DWPF) has been on-going. 

As shown in Figure 5.1, while salt waste makes up approximately 93% of the volume of waste 

stored at SRS, it represents approximately 52% of the total waste radioactivity. Sludge makes up 

approximately 7% of the volume but comprises approximately 48% of the total waste 

radioactivity. The sludge in the waste tanks contains nearly all the long-lived actinides (e.g., U, 

Pu, Am, Np, and Cm) and Sr and poses the greatest risk to human health and the environment. 

Removal of the sludge from the waste tanks and immobilization of this waste stream in glass 

significantly decreases this risk. Sludge removal, processing, and vitrification in DWPF have 

been ongoing since 1996 and are described in Appendix 1. 

Figure 5.1: Waste Tank Composite Inventory (As of 12/01/04) [2] 

Saltcake

Sludge

Volume

36.4 Million

Gallons (Mgal)

Curies

223 MCi

(52%)

203 MCi

(48%)

426 Million

Curies (MCi)

211 MCi

(49%)

12 MCi

(3%)

Salt Supernate

33.8 Mgal

(93%)

2.6 Mgal

(7%)

17.4 Mgal

(48%)

16.4 Mgal

(45%)

 

17 “Old-style tanks” refers to 24 Type I, Type II, and Type IV waste tanks that were constructed and placed into 

service between 1952 and 1964. These tanks lack full secondary containment, i.e., they are not a tank inside of a 

tank. The Type I and Type II tanks have a five-foot high annulus pan; the Type IV tanks have no secondary 

containment vessel. 

18 Two of the 51 waste tanks have been operationally closed; 49 waste tanks continue to be in active service. 
February 28, 2005 

Page 35 of 94 

The problem is that usable space is required in the waste tanks to process the sludge waste. The 

only space that is usable for this purpose is in newer waste tanks (new-style tanks)19 with full 

secondary containment that have usable space, since DOE does not want to return liquid waste to 

the old-style tanks that have leaked and that DOE is trying to empty. Tank space is required to 

wash the sludge, to process the wastes generated during the sludge washing process, and to store 

the concentrated wastes generated from sludge washing and from the DWPF recycle stream. If 

there is insufficient usable tank space available, especially tank space associated with the 

evaporator systems20, sludge processing at DWPF will be slowed and eventually halted. 

Although sludge processing results in removal of sludge from the tanks, it actually generates 

more waste in volume than it removes from the Tank Farms. Therefore, it is not the case that 

processing sludge creates tank space. To the contrary, on average, for every one gallon of settled 

sludge removed and processed through DWPF, 1.3 gallons [4] of salt waste is generated in one of 

the newer waste tanks (new-style tanks). 

Further compounding this is the fact that the sludge that is currently being removed is from the 

old-style tanks, since these tanks, with their history of corrosion and leak sites (13 of the 24 oldstyle tanks have leakage history), represent the greatest risk of environmental insult. Over 1.1 

Mgal of sludge waste [2] remain in these old-style tanks. Continuing sludge removal activities 

for these old-style tanks continues to be the highest priority at SRS but will have to stop unless 

salt waste processing starts in a timely manner and recovers the new-style tank space necessary 

for continued sludge processing. 

In addition, processing the salt waste will itself require tank space. That waste has been 

concentrated and precipitated over the years through the use of robust evaporator systems. In 

order to make the salt mobile for transferring and to adjust the salt concentrations (molarity) for 

processing and disposal, significant quantities of chemically treated water must be added. 

Although 33.8 Mgal of salt waste currently resides in the Tank Farms, approximately 84 Mgal of 

salt solution [3] will be processed. New-style tank space must be available for receiving and 

adjusting these solutions. 

The SWPF is scheduled to begin operation in approximately 2009. With enough tank space 

available, SWPF should be able to process a minimum of 5 Mgal of salt solution during the first 

year of operation, and 6 to 9 Mgal of salt solution per year thereafter, with a nominal annual 

 

19 “New-style tanks” refers to the 27 Type III waste tanks that were constructed and placed into service between 

1967 and 1986. These tanks have full secondary containment, i.e., they are a “tank within a tank.” These tanks were 

also stress-relieved at the time of construction to address nitrite-induced stress corrosion cracking that led to the 

formation of leak sites in the old-style tanks. There have been no leak sites in the new-style tanks. 20 There are currently three operating waste evaporator systems in the SRS Tank Farms. The volume reduction 

capabilities of the evaporator systems are the key to continued processing of sludge and receipts of influents. 

Specifically, the evaporators must have adequate space in their associated concentrate receipt tanks to receive the 

concentrated waste from the evaporator. If the concentrate receipt tank for an evaporator system becomes filled with 

saltcake and a viable, alternative receipt tank is not available, the system cannot operate. Or, if the evaporator 

system concentrates that waste to the maximum extent possible and no other new-style tank space exists into which 

to transfer the concentrated solution, then the system must shut down. 
February 28, 2005 

Page 36 of 94 

processing rate of approximately 7 Mgal [4]. This will allow DOE to complete waste removal 

from the remaining 49 active waste tanks at SRS by 2019, as contemplated by the SRS 

Environmental Management Program Performance Management Plan [19]. 

This aggressive processing rate, however, is predicated on two key assumptions: 1) that the salt 

waste has been staged and is ready to feed through SWPF at the time of startup; and 2) that 

sufficient new-style tank space will be available to continue to process and stage batches of 

SWPF feed. Specifically, it assumes that multiple new-style tanks are available to be used to 

blend salt solutions so that they meet the WAC for SWPF feed, store the blended feed while it is 

being characterized, and stage the waste to feed SWPF. The processing rate dictates the amount 

of necessary space; larger processing rates require more dedicated processing space. This is 

because the time it takes to blend and characterize the unprocessed solution is relatively slow 

compared to the feed rate. If adequate processing space is not available to prepare salt feed 

solutions, SWPF production rates will be reduced. A reduction in SWPF production rates will 

slow the rate at which waste is removed from the old-style tanks and thus delay reducing the risk 

to the occupational worker, the public, and the environment. 

Thus, failure to proceed expeditiously with salt processing will have many significant 

consequences on DOE’s efforts to accelerate reduction of the risks posed by the continued longterm storage of this waste within the aging infrastructure of the Tank Farms. The problem is 

worse than a day-for-day slippage, because either continuing sludge removal operations without 

salt processing capacity or stopping them early in order to preserve tank space will result in more 

than a day-for-day loss of time. If DOE continues sludge removal operations without removing 

some salt waste, the remaining available waste tank space will decrease and the ability to 

efficiently treat tank waste for disposal also diminishes. Instead of being able to process large 

batches of material, small batches must be used. Once the available space decreases to a certain 

volume, significant programmatic delays will occur. For instance, it is estimated that a one-year 

delay in salt processing will result in a three- to four-year delay in the overall program execution 

with an associated significant increase in cost (estimated to be in excess of $1 billion). Likewise, 

if DOE were to cease sludge removal operations now in order to preserve tank space, similar 

program delays and cost impacts would result due to the requirement to recover tank space for 

full-capacity SWPF operations and to go through a complex restart process for DWPF. 

Either continuing sludge removal without initiating salt processing or halting sludge removal in 

order to preserve tank space and restarting these operations later also significantly increases 

operational risks. If sludge removal were simply continued without some removal of salt waste, 

as recognized by the Defense Nuclear Facilities Safety Board (DNFSB) [20], the reduction in 

usable storage space will require increasingly frequent and complex waste transfers and reduce 

the amount of contingency space reserved for emergency inter-tank transfers in the event a tank 

leaked. This translates to additional risk to occupational workers, the public, and the 

environment. If sludge removal and processing were stopped now, significant facility 

maintenance and modifications would be necessary in both the Tank Farms and in DWPF for 
February 28, 2005 

Page 37 of 94 

resumption of operations. These maintenance and modification activities would result in 

significant increased worker exposure to radiation and radiological contamination hazards21. 

5.2.2.b Interim Salt Processing Strategy 

Given these exigencies and recognizing that SWPF cannot be constructed, permitted, and 

operated until approximately 2009, DOE developed a two-part interim processing approach that 

is tailored to the physical attributes of the different waste forms and the existing available 

technologies22. This approach will allow ongoing risk reduction activities through sludge 

removal to continue at its current accelerated pace until SWPF is operational without resulting in 

back-end delays, thereby reducing risks to public health and safety, and the environment, as well 

as occupational risks. 

The Interim Salt Processing Strategy will be implemented so as to minimize the curies in the 

waste that will be disposed of at SDF following treatment using the interim technologies. This is 

accomplished by: 1) processing only that salt waste volume necessary to provide the minimum 

tank space needed to support efficient DWPF and SWPF operations and 2) carefully selecting the 

salt waste that will be processed from the tanks containing lower activity waste (with the majority 

of the total curies disposed of in SDF being associated with the shorter-lived Cs-137)23 [5]. 

The Interim Salt Processing Strategy provides for treatment of these lower activity streams to 

remove the Cs-137 and other actinides with a lower contribution to total activity utilizing DDA 

and ARP/MCU technologies. This tank selection process limits the total estimated radioactivity 

inventory permanently disposed of in SDF vaults to an estimated 3 MCi [4]. Due to some 

uncertainty associated with the characterization of the saltcake waste, the actual curie content of 

this volume of material may be as high as 5 MCi [4]. Nearly all of these curies are associated 

with Cs-137, which has a half-life of approximately 30 years. 

 

21 DOE would not have the current tank space problem if salt processing had been initiated as originally planned. 

That did not occurred for two reasons. First, the initial technology DOE selected for salt processing, In-Tank 

Precipitation (ITP), was not successful. In January 1998, DOE determined that the ITP process could not be safely 

operated at the throughputs required. DOE then performed studies, including a Supplemental EIS completed in June, 

2001, to select viable alternative technologies for salt waste disposition activities. These activities were suspended in 

2003 when the Idaho District Court issued its decision in Natural Resources Defense Council v. Abraham, 271 F. 

Supp. 2d 1260 (D. Idaho. 2003) invalidating the provisions in DOE’s Order 435.1 that set out DOE’s criteria for 

classifying waste from reprocessing as “waste incidental to reprocessing” rather than “high-level radioactive waste.” 

The criteria the District Court struck down were a central underpinning of DOE’s entire salt strategy. DOE has only 

recently been in a position to proceed with this strategy as a result of the enactment of Section 3116. 22 The interim processing approach was developed after input from the South Carolina Governor’s Nuclear Advisory 

Council and SCDHEC and is referred to as the Interim Salt Processing Strategy. 23 Extensive tank space modeling projections were performed to determine optimal selection of salt waste to be 

processed as part of Interim Salt Processing given the desire to minimize curies disposed in SDF. 
February 28, 2005 

Page 38 of 94 

5.2.2.c DDA, ARP and MCU 

DOE identified three technologies for removing radionuclides from the salt waste that it could 

deploy during the interim period between now and approximately 2009 when SWPF is slated to 

come online. These are DDA, ARP and MCU. 

These processes were selected for their ability to reduce Cs-137, Sr-90, and actinide 

concentrations. DOE assessed projected tank space availability versus needs during the time 

between now and the start-up of SWPF and determined that insufficient space existed for 

sustained sludge disposition activities or for feed preparation for SWPF. Options for nearer term 

deployment of treatment technologies were evaluated and, based upon the technology selection 

process used for SWPF, DOE decided to pursue early small-scale implementation of the same 

actinide and Cs removal technologies (i.e., ARP and MCU). Capacity limitations imposed by the 

re-use of existing facilities necessitated initiation of salt removal prior to availability of the smallscale facilities. DDA was selected as the best alternative to augment ARP/MCU capacities until 

SWPF was operational. 

DOE anticipates using all three technologies in combination in approximately 2007, when ARP 

and MCU are expected to come online. In the meantime, upon issuance of a final 3116 

Determination, DOE anticipates using DDA alone to process some of the lowest activity salt 

waste. 

DDA

The deliquification and dissolution processes involved in DDA will remove substantial amounts 

of Cs-137, as well as some insoluble Sr and actinides, from this already relatively low-curie 

material24. Through deliquification, DDA will remove approximately 50% [21] of the Cs-137 and 

its daughter product Ba-137m from the saltcake targeted for dissolution. The liquid removed 

from the saltcake will be transferred to other tanks for future processing by SWPF. The saltcake 

will then be dissolved, after which it will be managed in a manner that will facilitate the settling 

of the insoluble sludge solids (e.g., Sr-90 and actinides) entrained in the salt solution and 

minimize the transfer of these materials to SDF. This is done by transferring the dissolved salt 

solution to a staging tank where the solids are allowed to settle before transfer to the Saltstone 

Facility feed tank. After being processed through DDA and after being solidified in a grout 

matrix, the solidified low-activity salt waste will have concentrations below Class C limits for all 

radionuclides. DOE can begin to use DDA immediately. 

ARP and MCU

ARP and MCU are expected to come online in approximately 2007. ARP and MCU will remove 

an additional approximately 92% [22] of the Cs-137/Ba-137m following DDA while also 

removing insoluble solids which contain the majority of the Sr and actinides. The ARP facilities 

will also have the capability to remove soluble Sr and actinides through MST strikes. If the 

 

24 This process will not be used for Tank 48 waste. 
February 28, 2005 

Page 39 of 94 

soluble actinides in the original salt solution are sufficiently low (i.e., below Class C 

concentration limits), to achieve the necessary tank space recovery prior to SWPF start-up, the 

stream will only be filtered prior to being sent to MCU. The filtering step will remove the 

majority of insoluble solids and is a necessary precursor to processing the salt solution through 

MCU25. Additionally, while neither ARP nor MCU preferentially targets the removal of Se-79, 

Tc-99, Sn-126, and I-129, ARP’s filtration process will remove a majority of the insoluble 

species of these radionuclides. 

Use of ARP, MCU, and DDA

Once ARP and MCU are online, in approximately 2007, DOE anticipates using them on waste 

that has already undergone DDA to process approximately 2.1 Mgal of salt waste that are 

expected to result in an estimated 0.3 MCi disposed of in SDF vaults from these facilities [4]. By 

processing this salt solution through ARP and MCU, rather than just DDA, the total number of 

curies ultimately disposed in SDF vaults will be decreased by an estimated 3.1 MCi. 

DOE anticipates using DDA alone to process approximately 6.0 Mgal of some of the lowest curie 

salt solution, resulting in an estimated 1.7 MCi disposed in SDF vaults from these streams [4]26. 

This is because if salt processing is delayed until the ARP/MCU facilities become operational in 

approximately 2007, this still will result in significant delay to DOE’s sludge removal efforts. 

This, in turn, is because DOE has an immediate need for usable working space in specific newstyle tanks within the Tank Farms to continue these activities. This particular space is needed 

because of the proximity of these tanks to the salt processing facilities and the existing transfer 

line infrastructure, as well as the need to have viable concentrate receipt tanks for the evaporator 

systems associated with sludge batch preparation and concentration of the recycle waste stream 

returning from DWPF. Without use of DDA to remove salt waste until ARP and MCU are 

online, there will still be insufficient space created in the new-style tanks to ensure continued 

risk-reduction at DWPF at its current rate of processing and sufficient feed batches cannot be 

prepared to ensure that SWPF can start-up operations at the rate of 5 Mgal during the first twelve 

months of operation. In addition, the increased waste inventory will result in increased safety 

risks associated with the operational activities of the Tank Farms as identified by the DNFSB 

[20]. 

After ARP/MCU are on-line, DOE will continue to use DDA alone in the interim period for some 

of the salt waste. This is because the capacity of the combined ARP/MCU facilities is only 1.5 

Mgal salt solution per year and these facilities, therefore, cannot be used to process enough salt 

waste to create salt the tank space needed. 

 

25 The current Interim Salt Processing Strategy does not generally contemplate MST strikes of the salt solutions that 

will be batched through ARP/MCU but an 8-hour MST strike will be performed if necessary to meet Class C limits 

for disposal of DSS in SDF or if throughputs can be maintained at 1.5 Mgal per year even if strikes are not necessary 

to meet Class C concentration limits. 

26 These figures do not include the materials in Tank 48, which are unique and are discussed further later in this 

document. 
February 28, 2005 

Page 40 of 94 

Tank 48 Waste

In addition to the low-activity salt wastes that DOE proposes to process during the interim period 

using combinations of DDA, ARP, and MCU, DOE also proposes to create additional tank space 

during this interim period by removing a unique low-activity salt waste in Tank 48, aggregating it 

with DWPF recycle, and disposing of the aggregated salt waste stream in SDF as a solidified 

low-activity salt waste. 

Tank 48, a new-style waste tank that is strategically located in close proximity to the salt 

processing facilities, is currently being used to store approximately 0.24 Mgal [2] of relatively 

low-activity salt solution containing approximately 19,000 Kg [9] of potassium and cesium TPB 

salts generated during the 1983 ITP27 process demonstration and subsequent operation of the ITP 

facility from 1995 through 1996. ITP was DOE’s earlier unsuccessful salt processing 

technology. It produced a salt waste that contains organic compounds not found in the other 

tanks at SRS. These organic compounds prevent DOE from using any of the facilities discussed 

in prior portions of this document to remove radionuclides. 

The organic nature of TPB salts makes the Tank 48 waste incompatible with the existing Tank 

Farm facilities or future salt waste treatment facilities. The decomposition of TPB produces 

flammable vapors, predominantly benzene. Flammable vapors are a concern because they could 

initiate a vapor deflagration. The high vapor pressure of benzene allows evaporation within the 

current operating temperature ranges for waste tanks. Tank 48 has unique, restrictive controls to 

limit the potential of TPB decomposition. These controls include low temperature and a 

restriction on addition of any material into the tank. The rate of evolution of benzene increases 

with increased temperature and in the presence of catalysts that are present in other Tank Farm 

waste. Therefore, the presence of catalysts precludes combining this waste with other Tank Farm 

wastes and the need to maintain relatively low temperatures precludes mixing with materials that 

may be processed through one of the three existing evaporator systems. 

For the reasons noted above, this material cannot be concentrated into saltcake using existing 

facilities. Accordingly, DDA cannot be used to remove cesium. Likewise, the need to control 

temperature precludes treatment of this waste through ARP/MCU or SWPF because the effluents 

from these facilities receive subsequent high temperature processing through the DWPF melter. 

In addition, current TPB concentration in Tank 48 is likely to adversely affect glass chemistry 

and impact canister throughput and waste loading in the glass. 

DOE has performed extensive evaluation of alternative technologies for processing the Tank 48 

salt waste. In conjunction with the Savannah River National Laboratory, treatment options 

within existing site facilities were evaluated and, in a parallel effort, vendor proposals for the 

design and installation of a separate waste treatment unit were solicited. Options were explored 

 

27 The In-Tank Precipitation process was a technology that was designed to separate the high-activity and lowactivity waste fractions of the salt waste by mixing sodium tetraphenylborate and monosodium titanate. The sodium 

tetraphenylborate targeted the separation of cesium and the monosodium titanate targeted the separation of strontium 

and actinides. During full scale operations, the In-Tank Precipitation process could not achieve production goals and 

safety requirements for processing tank waste. 
February 28, 2005 

Page 41 of 94 

for organic destruction within Tank 48 (in-situ) including hydrolysis, both chemical (e.g., 

Fenton’s Reagent) and thermal, and catalysis (e.g., palladium). All of these options require 

taking the waste to an acidic state which is incompatible with the carbon steel materials of 

construction associated with Tank 48 (as well as all other SRS waste tanks). External options 

were explored that utilized existing site facilities including both precipitate and acid hydrolysis at 

DWPF. The DWPF options were eliminated due to direct conflict with the need for these same 

facility spaces for receipt of the high-level waste effluents from ARP/MCU and SWPF. DOE 

also explored construction of a separate waste treatment unit to remove the organics and 

concluded it would be extremely costly, would take years to build and permit and hence for all 

these reasons would be impractical. Thus, there are currently no proven, practical options for 

removing additional radionuclides from the Tank 48 waste that would allow the disposal of the 

removed radionuclides using DWPF, and likewise there are no practical approaches on the 

horizon that would allow this to be accomplished. 

Tank 48 currently contains a relatively small quantity of salt waste, approximately 0.24 Mgal, 

and a relatively small number of curies, approximately 0.8 MCi, when compared to other waste 

tanks. By processing the 0.24 Mgal of existing Tank 48 waste, a 1.3 Mgal waste tank would be 

made available for salt processing activities. No other available new-style tank that could 

substitute in place of Tank 48 contains salt waste that can be disposed of and result in less than 

0.8 MCi being sent to SDF. In addition, Tank 48 is uniquely well situated to assist in the overall 

salt processing strategy. It is located in close proximity to the salt processing facilities and 

already has dedicated transfer lines to the cells that will house the ARP reaction tanks. For this 

reason, Tank 48 is physically ready to serve as the ARP/MCU feed tank with only minimal 

modifications. Tank 48 also has an existing pump and dedicated transfer line that connects Tank 

48 directly to Tank 49. Tank 49 is currently slated to be the feed tank for SWPF. Considerable 

infrastructure work would need to be completed to set up any other tank as the feed tank for the 

ARP and to provide the transfer flexibility associated with Tank 48. 

Accordingly, given the unavailability of current, proven, or contemplated practical technologies 

to remove radionuclides from the salt waste in Tank 48, given the relatively low radioactive 

content of this material, given that its disposal after mixing it with other low-activity salt waste 

would free up considerable key tank space by adding fewer curies to SDF than other salt waste 

DOE has been able to identify that would free up the same amount of tank space, and given that 

when mixed with the low-activity waste to make adjustments for Saltstone Facility processability 

requirements (and stabilized at an approximate one-to-one ratio with grout , the solidified waste 

will be below Class C concentrations, it is appropriate, under the “maximum extent practical” 

consideration, to process Tank 48 waste without further removal of radionuclides as part of 

DOE’s overall interim salt waste strategy. 

In sum, DOE’s salt waste processing strategy will result in removal of highly radioactive 

radionuclides to the maximum extent practical. As explained above, in all, the plan will result in 

removal of approximately 218 to 220 MCi of the approximately 223 MCi [2] present in the SRS 

salt waste. DOE’s best estimate is that approximately 3 MCi, mostly the short-lived radionuclide 

Cs-137 and its daughter Ba-137m, will be solidified and disposed at SDF (although due to the 

uncertainty associated with the current characterization of the saltcake waste, the actual curie 
February 28, 2005 

Page 42 of 94 

content of this material may be as high as 5 MCi [4]). Collectively, DDA, ARP, MCU, and 

SWPF will remove approximately 98.7% of the curies from salt waste for vitrification in DWPF 

and ultimately permanent disposal in a deep geologic repository. 

To process the vast majority and the more radioactive salt waste, DOE will use the SWPF, which 

will use state-of-the-art removal technologies, as soon as this facility is constructed and 

operational. In the meantime, DOE will use combinations of the interim technologies to process 

the minimal amounts of particularly low-activity salt waste that must be removed to create the 

tank space necessary to allow DOE to continue to vitrify sludge and to be in a position to operate 

SWPF at close to capacity as soon as it is operational. These technologies are the best available 

during this timeframe to remove radionuclides from enough of this waste to create the needed 

space, and all will remove significant amounts of radionuclides from this waste. Accordingly, 

DOE’s salt waste processing strategy will remove highly radioactive radionuclides to the 

maximum extent practical. 
February 28, 2005 

Page 43 of 94 

6.0 THE WASTE DOES NOT EXCEED CONCENTRATION LIMITS IN 10 CFR 61.55 

Section 3116 of the NDAA provides in relevant part: 

 [T]he term “high-level radioactive waste” does not include waste from 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...[.] 

The NRC regulations at 10 CFR 61.55 categorize low-level radioactive waste intended for near 

surface disposal as Class A, Class B, or Class C. To make the waste classification determination, 

the waste concentrations for specified radionuclides are compared to concentration limits defined 

in 10 CFR 61.55. The radionuclides and their associated limits are specified in two separate 

tables within 10 CFR 61.55 which are reproduced below. 

Table 6.1: 10 CFR 61.55 Table 1 

Radionuclides (Long-lived) Concentration 

(Ci/m3

) 

C-14 8 

C-14 in activated metal 80 

Ni-59 in activated metal 220 

Nb-94 in activated metal 0.2 

Tc-99 3 

I-129 0.08 

Alpha Emitting Transuranic (TRU) nuclides with half-life 

greater than 5 years 100(1)

Pu-241 3,500(1)

Cm-242 20,000(1)

1 Units are in nanocuries per gram. 
February 28, 2005 

Page 44 of 94 

Table 6.2: 10 CFR 61.55 Table 2 

Concentration, curies per cubic 

meter Radionuclides (Short-lived) Column 1 

[Class A] 

Column 2 

[Class B] 

Column 3 

[Class C] 

Total of all nuclides with less than 5 year half-life 700 (1) (1) 

H-3 40 (1) (1) 

Co-60 700 (1) (1) 

Ni-63 3.5 70 700 

Ni-63 in activated metal 35 700 7000 

Sr-90 0.04 150 7000 

Cs-137 1 44 4600 

1

 There are no limits established for these radionuclides in Class B or C wastes. Practical considerations such as 

the effects of external radiation and internal heat generation on transportation, handling, and disposal will limit 

the concentrations for these wastes. These wastes shall be Class B unless the concentrations of other nuclides in 

the table determine the waste to be Class C independent of these nuclides. 

Estimated average concentrations for the grouted salt waste associated with the DDA, 

ARP/MCU, and SWPF processes were calculated [23] for radionuclides identified in 10 CFR 

61.55, Tables 1 and 2. These concentrations are compared to 10 CFR 61.55 Class C 

concentration limits in Tables 6.3 through 6.8 below. Tables 6.3 and 6.4 show the comparison 

done for the grouted salt waste associated with the DDA process, Tables 6.5 and 6.6 show the 

comparison done for the grouted salt waste associated with the ARP/MCU process, and Tables 

6.7 and 6.8 show the comparison done for the grouted salt waste associated with the SWPF 

process. Concentrations in these tables are shown in curies per cubic meter (Ci/m3

) in the 

grouted salt waste unless otherwise noted. 

Table 6.3: DDA Waste Stream Comparison to 10 CFR 61.55 Table 1 

Radionuclides (Long-lived) 

10 CFR 61.55 

Class C Limit 

(Ci/m3

) 

Estimated 

Average 

Concentration 

(Ci/m3

) 

% of 10 

CFR 61.55 

Class C 

Limit 

C-14 8 0.00009 0.001%

C-14 in activated metal 80 (1) (1)

Ni-59 in activated metal 220 (1) (1)

Nb-94 in activated metal 0.2 (1) (1)

Tc-99 3 0.002 0.06%

I-129 0.08 0.00001 0.02%

Alpha Emitting Transuranic (TRU) nuclides with halflife greater than 5 years 100(2) 64(2) 64%

Pu-241 3,500(2) 19(2) 0.5%

Cm-242 20,000(2) 0.005(2) 0.00003% 1

 Not present in the grouted salt waste in the Saltstone Disposal Facility. 

2

 Units are in nanocuries per gram. 
February 28, 2005 

Page 45 of 94 

Table 6.4: DDA Waste Stream Comparison to 10 CFR 61.55 Table 2 

Radionuclides (Short-lived) 

10 CFR 61.55 

Class C Limit 

(Ci/m3) 

Estimated 

Average 

Concentration 

(Ci/m3

) 

% of 10 CFR 

61.55 Class C 

Limit 

Total of all nuclides with less than 5 year half-life (1) (1) (1)

H-3 (1) (1) (1)

Co-60 (1) (1) (1)

Ni-63 700 0.00008 0.00001%

Ni-63 in activated metal 7000 (2) (2)

Sr-90 7000 3 0.04%

Cs-137 4600 27 0.6% 1

 There are no limits established for these radionuclides in Class B or C wastes. Practical considerations such as the 

effects of external radiation and internal heat generation on transportation, handling, and disposal will limit the 

concentrations for these wastes. These wastes shall be Class B unless the concentrations of other nuclides in the table 

determine the waste to be Class C independent of these nuclides. Because the Saltstone Disposal Facility is a low level 

waste disposal facility, the grouted salt waste must meet or be lower than the concentration limits for Class C. 2

 Not present in the grouted salt waste in the Saltstone Disposal Facility. 

Table 6.5: ARP/MCU Waste Stream Comparison to 10 CFR 61.55 Table 1 

Radionuclides (Long-lived) 

10 CFR 61.55 

Class C Limit 

(Ci/m3) 

Estimated 

Average 

Concentration 

(Ci/m3

) 

% of 10 

CFR 61.55 

Class C 

Limit 

C-14 8 0.00003 0.0003% 

C-14 in activated metal 80 (1) (1)

Ni-59 in activated metal 220 (1) (1)

Nb-94 in activated metal 0.2 (1) (1)

Tc-99 3 0.00004 0.001%

I-129 0.08 0.000006 0.008%

Alpha Emitting Transuranic (TRU) nuclides with halflife greater than 5 years 100(2) 8(2) 8%

Pu-241 3,500(2) 10(2) 0.3%

Cm-242 20,000(2) 0.005(2) 0.00003% 1

 Not present in the grouted salt waste in the Saltstone Disposal Facility. 

2

 Units are in nanocuries per gram. 
February 28, 2005 

Page 46 of 94 

Table 6.6: ARP/MCU Waste Stream Comparison to 10 CFR 61.55 Table 2 

Radionuclides (Short-lived) 

10 CFR 61.55 

Class C Limit 

(Ci/m3) 

Estimated 

Average 

Concentration 

(Ci/m3

) 

% of 10 CFR 

61.55 Class C 

Limit 

Total of all nuclides with less than 5 year half-life (1) (1) (1)

H-3 (1) (1) (1)

Co-60 (1) (1) (1)

Ni-63 700 0.00003 0.000004%

Ni-63 in activated metal 7000 (2) (2)

Sr-90 7000 0.0002 0.000003%

Cs-137 4600 9 0.2% 1

 There are no limits established for these radionuclides in Class B or C wastes. Practical considerations such as the 

effects of external radiation and internal heat generation on transportation, handling, and disposal will limit the 

concentrations for these wastes. These wastes shall be Class B unless the concentrations of other nuclides in the table 

determine the waste to be Class C independent of these nuclides. Because the Saltstone Disposal Facility is a low level 

waste disposal facility, the grouted salt waste must meet or be lower than the concentration limits for Class C. 2

 Not present in the grouted salt waste in the Saltstone Disposal Facility. 

Table 6.7: SWPF Waste Stream Comparison to 10 CFR 61.55 Table 1 

Radionuclides (Long-lived) 

10 CFR 61.55 

Class C Limit 

(Ci/m3) 

Estimated 

Average 

Concentration 

(Ci/m3

) 

% of 10 

CFR 61.55 

Class C 

Limit 

C-14 8 0.00003 0.0003%

C-14 in activated metal 80 (1) (1)

Ni-59 in activated metal 220 (1) (1)

Nb-94 in activated metal 0.2 (1) (1)

Tc-99 3 0.00004 0.002%

I-129 0.08 0.00003 0.04%

Alpha Emitting Transuranic (TRU) nuclides with halflife greater than 5 years 100(2) < 10(2) < 10%

Pu-241 3,500(2) 8(2) 0.2%

Cm-242 20,000(2) 0.005(2) 0.00003% 1

 Not present in the grouted salt waste in the Saltstone Disposal Facility. 

2

 Units are in nanocuries per gram. 
February 28, 2005 

Page 47 of 94 

Table 6.8: SWPF Waste Stream Comparison to 10 CFR 61.55 Table 2 

Radionuclides (Short-lived) 

10 CFR 61.55 

Class C Limit 

(Ci/m3) 

Estimated 

Average 

Concentration 

(Ci/m3

) 

% of 10 CFR 

61.55 Class C 

Limit 

Total of all nuclides with less than 5 year half-life (1) (1) (1)

H-3 (1) (1) (1)

Co-60 (1) (1) (1)

Ni-63 700 0.00003 0.000004%

Ni-63 in activated metal 7000 (2) (2)

Sr-90 7000 0.0002 0.000003%

Cs-137 4600 0.01 0.0002% 1

 There are no limits established for these radionuclides in Class B or C wastes. Practical considerations such as the 

effects of external radiation and internal heat generation on transportation, handling, and disposal will limit the 

concentrations for these wastes. These wastes shall be Class B unless the concentrations of other nuclides in the table 

determine the waste to be Class C independent of these nuclides. Because the Saltstone Disposal Facility is a low level 

waste disposal facility, the grouted salt waste must meet or be lower than the concentration limits for Class C. 2

 Not present in the grouted salt waste in the Saltstone Disposal Facility. 

In accordance with 10 CFR 61.55, the determination of the class of the waste is made by one of 

four different methods utilizing Tables 6.1 and 6.2. The relevant NRC regulations are reproduced 

below: 

1) If the waste contains only the long-lived radionuclides listed in Table 6.1 above (10 

CFR 61.55 Table 1), then classification is determined by 10 CFR 61.55 (a)(3): 

10 CFR 61.55 (a)(3) – “Classification determined by long-lived radionuclides. If 

radioactive waste contains only radionuclides listed in Table 1, classification shall be 

determined as follows: (i) If the concentration does not exceed 0.1 times the value in 

Table 1, the waste is Class A. (ii) If the concentration exceeds 0.1 times the value in 

Table 1 but does not exceed the value in Table 1, the waste is Class C. (iii) If the 

concentration exceeds the value in Table 1, the waste is not generally acceptable for 

near-surface disposal. (iv) For wastes containing mixtures of radionuclides listed in 

Table 1, the total concentration shall be determined by the sum of fractions rule 

described in paragraph (a)(7) of this section”. 

2) If the waste contains only the short-lived radionuclides contained in Table 6.2 above 

(10 CFR 61.55 Table 2), then classification is determined by 10 CFR 61.55 (a)(4): 

10 CFR 61.55 (a)(4) – “Classification determined by short-lived radionuclides. If 

radioactive waste does not contain any of the radionuclides listed in Table 1, 

classification shall be determined based on the concentrations shown in Table 2. 

However, as specified in paragraph (a)(6) of this section, if radioactive waste does not 

contain any nuclides listed in either Table 1 or 2, it is Class A. (i) If the concentration 

does not exceed the value in Column 1, the waste is Class A. (ii) If the concentration 

exceeds the value in Column 1, but does not exceed the value in Column 2, the waste 

is Class B. (iii) If the concentration exceeds the value in Column 2, but does not 
February 28, 2005 

Page 48 of 94 

exceed the value in Column 3, the waste is Class C. (iv) If the concentration exceeds 

the value in Column 3, the waste is not generally acceptable for near-surface disposal. 

(v) For wastes containing mixtures of the nuclides listed in Table 2, the total 

concentration shall be determined by the sum of fractions rule described in paragraph 

(a)(7) of this section”. 

3) If the waste contains a mixture of short-lived and long-lived radionuclides, some of 

which are listed in Table 6.1 above (10 CFR 61.55 Table 1), and some of which are 

listed in Table 6.2 above (10 CFR 61.55 Table 2), classification is determined by 10 

CFR 61.55 (a)(5): 

10 CFR 61.55 (a)(5) – “Classification determined by both long- and short-lived 

radionuclides. If radioactive waste contains a mixture of radionuclides, some of 

which are listed in Table 1, and some of which are listed in Table 2, classification 

shall be determined as follows: (i) If the concentration of a nuclide listed in Table 1 

does not exceed 0.1 times the value listed in Table 1, the class shall be that determined 

by the concentration of nuclides listed in Table 2. (ii) If the concentration of a nuclide 

listed in Table 1 exceeds 0.1 times the value listed in Table 1 but does not exceed the 

value in Table 1, the waste shall be Class C, provided the concentration of nuclides 

listed in Table 2 does not exceed the value shown in Column 3 of Table 2”. 

4) If the waste does not contain any of the radionuclides listed in either of the tables then 

classification is determined by 10 CFR 61.55 (a)(6): 

10 CFR 61.55 (a)(6) – “Classification of wastes with radionuclides other than those 

listed in Tables 1 and 2. If radioactive waste does not contain any nuclides listed in 

either Table 1 or 2, it is Class A”. 

The low-activity salt waste streams to be disposed at SDF will contain a mixture of short-lived 

and long-lived radionuclides from Tables 6.1 and 6.2. Therefore, the waste concentration limits 

will be determined in accordance with 10 CFR 61.55 (a)(5). As provided in 10 CFR 61.55 (a)(5), 

the radionuclide limits shown in Table 6.1 above (10 CFR 61.55 Table 1) and radionuclide limits 

shown in Column 3 of Table 6.2 above (10 CFR 61.55 Table 2) are applicable. 

All waste treated through interim processing or SWPF and disposed at SDF will meet the 

concentration limits established in 10 CFR 61.55 for Class C waste as shown in Tables 6.3 

through 6.8 above. Mixing dry chemicals with the low-activity salt waste stream to form grout is 

designed to stabilize the waste for disposal. All waste treated through interim processing or 

SWPF and disposed at SDF will be mixed with dry chemicals in the approximate ratio of 1:1 and 

solidified in a manner to achieve reasonable homogeneity. Thus, consistent with the guidance of 

the NRC’s Branch Technical Position [24], the radionuclide concentrations in the low activity 

salt waste stream will be reasonably homogeneous for the grouted salt waste disposed in SDF 

[25,26]. 
February 28, 2005 

Page 49 of 94 

The Saltstone Facility WAC will include limits established to ensure that concentration limits for 

Class C low-level waste in Section 10 CFR 61.55 are protected. 

The Saltstone Facility WAC establishes the physical, chemical, and radionuclide limits for all 

waste to ensure that, at a minimum, the following are not exceeded: 

• Limits established by the Performance Assessment (PA)/Special Analysis (SA)28 

• SDF operating permit limits (State issued) 

• Documented Safety Analysis (DSA) limits 

• Processability limits29

Limits established in the WAC are set to ensure that the most restrictive limits from the criteria 

listed above are not exceeded. Therefore, the WAC and SRS sampling, analysis, and quality 

assurance procedures provide added assurance that the NRC Class C waste disposal limits will be 

met. 

For example, to ensure compliance with the Saltstone Facility WAC, all generating facilities are 

required to develop a Waste Compliance Plan (WCP) which describes the controls or procedures 

imposed to meet the requirements set forth in the WAC. In addition to the generating facility 

requirements, the Saltstone Facility is required to provide testing and maintain records necessary 

to support certification of the resulting grout as suitable for disposal at SDF as defined in permits 

issued by the State of South Carolina and in DOE orders. The requirements and responsibilities 

associated with the Saltstone Facility WAC program are outlined in the Saltstone Facility Waste 

Acceptance Criteria Program Requirements [27]. 

The salt solution arising from the treatment of salt waste per the Salt Processing Strategy must 

meet the Saltstone Facility WAC to be accepted for processing and disposition at the Saltstone 

Facility. WAC limits will be established to ensure that NRC Class C waste disposal limits set out 

in 10 CFR 61.55 are not exceeded. Therefore, the Saltstone Facility WAC program adds 

assurance that criterion (a)(3)(A) of NDAA Section 3116 is met. 

 

28 See Section 7.2.1. 

29 Processability limits in this context refer to chemical, physical, and/or radiological limits placed on the incoming 

low-activity waste stream to ensure the ability of SPF and SDF to safely handle the waste stream and to produce an 

acceptable grout product at the necessary throughputs. 
February 28, 2005 

Page 50 of 94 

7.0 THE WASTE WILL BE DISPOSED OF IN ACCORDANCE WITH THE 

PERFORMANCE OBJECTIVES SET OUT IN 10 CFR PART 61, SUBPART C 

Section 3116 of the NDAA provides in relevant part: 

[T]he term “high-level radioactive waste” does not include waste from reprocessing of spent 

nuclear fuel that the Secretary of Energy...in consultation with the Nuclear Regulatory 

Commission...determines – 

*** 

(3)(A) [W]ill 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[.] 

7.1 10 CFR PART 61 PERFORMANCE OBJECTIVES

Subpart C of 10 CFR Part 61 lists one general requirement and four performance objectives, 

which are reproduced below. 

7.1.1 Section 61.40, General Requirement 

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 Sections 61.41 through 61.44. 

7.1.2 Section 61.41, Protection of the General Population from Releases of Radioactivity 

Concentrations of radioactive material which may be released to the general environment in 

ground water, surface water, air, soil, plants, or animals must not result in an annual dose 

exceeding an equivalent of 25 millirems to the whole body, 75 millirems to the thyroid, and 25 

millirems to any other organ of any member of the public. Reasonable effort should be made to 

maintain releases of radioactivity in effluents to the general environment as low as is reasonably 

achievable. 
February 28, 2005 

Page 51 of 94 

7.1.3 Section 61.42, Protection of Individuals from Inadvertent Intrusion 

Design, operation, and closure of the land disposal facility must ensure protection of any 

individual inadvertently intruding into the disposal site and occupying the site or contacting the 

waste at any time after active institutional controls over the disposal site are removed. 

7.1.4 Section 61.43, Protection of Individuals During Operations 

Operations at the land disposal facility must be conducted in compliance with the standards for 

radiation protection set out in part 20 of this chapter, except for releases of radioactivity in 

effluents from the land disposal facility, which shall be governed by Section 61.41 of this part. 

Every reasonable effort shall be made to maintain radiation exposures as low as is reasonably 

achievable. 

7.1.5 Section 61.44, Stability of the Disposal Site After Closure 

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. 

7.2 GENERAL REQUIREMENT (10 CFR 61.40) 

The general provision at 10 CFR 61.40 calls for “reasonable assurance” that exposures are within 

the limits of the four subsequent performance objectives - 10 CFR 61.41, 10 CFR 61.42, 10 CFR 

61.43, and 10 CFR 61.44 – as discussed below. Within each ensuing discussion of these four 

performance objectives, the general provision of 10 CFR 61.40 is also addressed. 

7.2.1 Protection Of The General Population (10 CFR 61.41) 

10 CFR 61.41 states: 

Concentrations of radioactive material which may be released to the general environment in 

ground water, surface water, air, soil, plants, or animals must not result in an annual dose 

exceeding an equivalent of 25 millirems to the whole body 75 millirems to the thyroid, and 

25 millirems to any other organ of any member of the public. Reasonable effort should be 

made to maintain releases of radioactivity in effluents to the general environment as low as 

is reasonably achievable30. 

 

30 The NRC has stated: “The dose methodology used in 10 CFR 61 Subpart C is different from that used in the newer 

10 CFR 20 Subpart E. However, the resulting allowable doses are comparable and NRC expects DOE to use the 

newer methodology in 10 CFR 20 Subpart E. Part 61 is based on International Commission on Radiation Protection 

Publication 2 (ICRP 2) and Part 20 is based on ICRP 26” [26]. The methodology defined in ICRP 26 calculates dose 

in total effective dose equivalent (TEDE) versus the organ doses of the earlier methodology. Radiological doses are 

also calculated in TEDE in the SA in accordance with the newer methodology. 
February 28, 2005 

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To determine whether the salt waste will be disposed of in conformity with the 10 CFR Part 61 

Subpart C performance objective relating to protection of the general population, DOE has used 

two performance assessment documents prepared in 1992 and 2002. These documents and the 

manner of their use are described further below. DOE is currently supplementing these 

documents with an updated Special Analysis (SA). DOE will take account of this updated SA in 

making its final 3116 Determination with respect to the salt waste. 

In 1992, as part of the process for siting and building the Saltstone Facility, DOE conducted a 

Performance Assessment (Saltstone PA) [28] of that facility. The PA evaluated whether doses 

the general public would be likely to receive as a result of disposal of the waste anticipated to be 

disposed of in this facility would meet performance standards specified by DOE Order 5820.2A, 

the Order then governing disposal of low-level radioactive waste. To do so, the PA simulated 

migration of radionuclides from the disposal of the solidified waste form through the disposal 

vaults into the surrounding environment to a hypothetical groundwater well located 100 meters 

from the disposal facility. 

In 1992, the waste to be disposed of in the Saltstone Facility was expected to consist of a 

saltstone grouted waste produced by mixing salt waste from the tanks that had been processed 

through a different salt processing technology, the ITP, together with a small amount of salt 

waste from the Effluent Treatment Facility, with specially-formulated grout. This waste 

inventory was expected to be considerably lower for fission products and actinides than the 

currently-planned inventory. 

The SRS has compiled an extensive site-specific hydrogeological database which includes the 

area around the Saltstone Facility using test borings and groundwater wells to understand how 

water flows in the site. It also has information on the chemical and physical composition of the 

soils through which the groundwater and radionuclides would flow. Based on this and other 

materials, DOE developed groundwater flow and transport models for the unsaturated and 

saturated zones using site-specific adsorption and permeability information as well as information 

from the scientific literature. The models also incorporated laboratory and other data to model 

the physical and chemical properties of the grouted waste form [25]. 

The PA results indicated that the vast majority of the radionuclides never reach the 100 meter 

point of assessment because their movement is sufficiently retarded by a combination of the soils, 

the saltstone grout, and the waste chemistry such that they decay first. The movement of many 

longer-lived radionuclides, such as U, Pu, and Tc-99 is also substantially retarded because they 

are only slightly soluble in the waste form. The PA analyzed two cases for groundwater: the 

degraded case which assumed that the grout in the vaults developed cracks upon closure of the 

SDF; and the intact case which assumed that the vault in the grout structure remained intact. 

Based on the results of the performance assessment, the calculated doses from estimated 

radionuclide concentrations in groundwater were at least two orders of magnitude below the 4 

mrem per year limit used in the PA analysis for all radionuclides, even if the facility degraded. 

The peak dose from the intact case was 0.001 mrem per year at greater than two million years 

after closure. The peak dose from the degraded case was 0.03 mrem per year at 15,000 years. 
February 28, 2005 

Page 53 of 94 

The PA also included qualitative rationale that showed that the air pathway would be 

insignificant and would not contribute to the doses to the member of the public. 

In 2002, DOE prepared an SA 31 updating the analysis in the PA to take into account that DOE 

was considering using the facility to dispose of salt waste from the tanks processed through the 

DDA and ARP processes. While DOE expected the salt waste processed through DDA and ARP 

to meet Class C concentration limits, these limits were higher than those of the waste DOE had 

originally analyzed. Therefore, DOE concluded that it needed to verify that if this DDA and 

ARP-processed waste were disposed of in the Saltstone Facility, doses to the general public 

would still meet the DOE performance objectives of 25 mrem per year to a member of the 

general public from all pathways, 10 mrem per year to a member of the general public for the air 

pathway alone, and the Environmental Protection Agency (EPA) Maximum Contaminant Levels 

(MCLs) for the groundwater pathway, consistent with the Federal Facilities Agreement for SRS. 

In the SA, calculations were performed to determine disposal limits on the maximum quantity 

and concentration of radionuclides that could be disposed of in SDF using DOE revised salt 

processing plans, without exceeding the DOE performance objectives set forth above. The SA 

calculated disposal limits for the radionuclides that were identified by a screening process based 

on the National Council on Radiation Protection and Measurements (NCRP) screening 

methodology [29]. The identified radionuclides were the same as those identified in the PA with 

the addition of Np-237 for the groundwater pathway. 

For each of those radionuclides other than Np-237, the SA calculated the DOE disposal limit for 

the SDF by using the peak doses in the PA to derive an estimated dose per curie attributable to 

each radionuclide. The SA then divided the DOE groundwater performance measure by the dose 

per curie to obtain the DOE disposal limit for the SDF for each radionuclide for the groundwater 

pathway. For Np-237, the SA calculated a dose per curie at 10,000 years using a simplified 

groundwater flow and transport model, and calculated the DOE disposal limit by dividing the 

DOE performance objective by the dose per curie at 10,000 years. For the air pathway, the SA 

calculated doses for H-3 and C-14, using the MAXIGASP code which is a SRS site-specific 

implementation of the CAP-88 code. These two radionuclides were selected for analysis because, 

based on expert judgment, they were the only two expected to contribute to dose from the air 

pathway. Using the SA doses for H-3 and C-14, the SA also calculated disposal limits for the 

SDF. 

The SA then verified that the proposed salt waste inventory contemplated in 2002 would meet the 

DOE disposal limit derived in the SA and hence would be DOE’s dose-based performance 

objective for the general public. 

For this draft 3116 Determination, DOE has estimated the doses from the groundwater pathway 

and the air pathway for each of the three salt processing streams and then compared the sum of 

 

31 The changes that precipitated the development of the SA as an upgrade to the PA included: 1) increased 

inventories of important radionuclide; 2) decrease in the design thickness of the vault configuration; 3) analytical 

corrections to more realistically predict doses in the intruder scenario; and 4) the addition of a one-meter thick barrier 

over the vault roof. 
February 28, 2005 

Page 54 of 94 

these doses to the NRC performance objectives. To do so, this draft 3116 Determination has 

used the DOE SDF disposal limits from the SA [25] and the expected composition of the salt 

waste to be disposed of in the SDF, as explained further below, to estimate the dose to a member 

of the general population. This draft 3116 Determination estimates the doses using the ICRP 26 

methodology to compare with the NRC performance objective of 25 mrem per year to the whole 

body. 

For this draft 3116 Determination, DOE estimated the doses to the general public from 

groundwater by essentially reverse-engineering the dose for each radionuclide based on the limits 

it had derived in the SA and then summing the doses appropriately. Specifically, DOE knew 

because of the way it had derived the SA disposal limits that if it disposed of a quantity of a 

radionuclide equal to the limit for that radionuclide as set out in the SA, disposal of that amount 

of that radionuclide would produce a dose equal to the DOE performance objective. 

Accordingly, to derive the dose that would result from each waste produced by each part of the 

salt waste processing strategy, DOE first divided the quantity of each radionuclide it projected 

would be present in that waste by the disposal limit for that radionuclide calculated in the SA, to 

obtain a fraction. DOE then summed the fractions obtained for all beta/gamma radionuclides and 

then multiplied that sum of the fractions by the DOE performance objective of 4 mrem per year 

to approximate a dose to the receptor32 from beta/gamma-emitters. DOE used a similar approach 

for alpha-emitting radionuclides using the DOE performance objective of 25 mrem per year. The 

resulting doses from the beta/gamma-emitting and the alpha-emitting radionuclides were then 

summed to obtain the total dose. The resulting maximum doses for the beta-gamma and alphaemitting radionuclides are negligible over 1,000 years from all three salt waste processes (DDA, 

ARP/MCU, and SWPF). Over 10,000 years, the maximum effective whole body beta/gamma 

dose is approximately 0.2 mrem per year. Over 10,000 years, the maximum effective whole body 

alpha-emitter dose is 0.0015 mrem per year33 [30]. 

The dose from the air pathway calculated for this draft 3116 Determination is insignificant. H-3 

and C-14 are the only radionuclides that contribute to the air pathway and they exist in relatively 

low concentrations. DOE calculated the estimated doses by using the disposal limits in the SA 

and a similar methodology to that described above for groundwater. The DOE performance 

objective used to derive the air pathway limit is an annual dose of 10 mrem to a hypothetical 

receptor located 100 meters from the disposed waste. Therefore, the air pathway dose can be 

estimated by multiplying the sum-of-fractions of the limits by 10. This results in an estimated 

dose of approximately 0.003 mrem per year at both 1,000 years and 10,000 years after closure of 

SDF. 

 

32 The receptor is a hypothetical adult member of the public assumed to be located 100 meters from the disposed 

waste in the direction of groundwater flow and is assumed to drink two liters of water per day for a year from that 

groundwater. The groundwater analysis considered one exposure pathway, that of drinking the water directly. Other 

potential exposure pathways, such as use of the groundwater for irrigating a garden and eating the produce from the 

garden, were not analyzed. 

33 This approach is conservative because it assumes that the maximum groundwater concentration of each 

radionuclide occurs at the same time, which is not the case (e.g., the maximum concentration for Tc-99 occurs at 

2,400 years and that for C-14 occurs at 7,300 years). The doses were derived from the peak concentrations of the 

degraded case analyzed in the PA. [25]. 
February 28, 2005 

Page 55 of 94 

To estimate the dose from all pathways for this draft 3116 Determination, the doses from the 

groundwater pathway and the air pathway are summed. For comparison with the 10 CFR 61.41 

performance objective of 25 mrem per year from all pathways combined, the sum of the 

groundwater pathway dose for both beta-gamma and alpha-emitting radionuclides and the air 

pathway dose at 1,000 years after closure is approximately 0.003 mrem per year, which is less 

than 0.01 percent of the NRC performance objective. At 10,000 years, the sum of the 

groundwater pathway dose for both beta-gamma and alpha-emitting radionuclides and the air 

pathway dose, at 10,000 years after closure is approximately 0.2 mrem per year, which is less 

than one percent of the NRC performance objective. 

Because 10 CFR 61.41 also has limits for doses to the thyroid and other key organs, these doses 

are also calculated for this draft 3116 Determination. Using conversion factors specified by the 

U. S. Environmental Protection Agency, Federal Guidance Report No. 11, EPA 520/1-88-020, 

Limiting Values of Radionuclides Intake and Air Concentrations and Dose Conversion Factors 

for Inhalation, Submersion, and Ingestion, September 1988, annual doses to the thyroid and other 

critical organs were calculated for each relevant radionuclide on an organ-by-organ basis. The 

peak annual dose to the thyroid is approximately 6.3 mrem per year compared to the 10 CFR 

61.41 limit of 75 mrem per year. The peak annual dose to any other organ is approximately 0.04 

mrem per year compared to the 10 CFR 61.41 limit of 25 mrem per year. 

Therefore, doses to the whole body, the thyroid and any other organ of a member of the public 

will not exceed the performance objective set forth in 10 CFR 61.41. 

The NRC general provision at 10 CFR 61.40 calls for “reasonable assurance ... that exposures to 

humans are within the limits established in the performance objectives” [16]. The analysis 

provides reasonable assurance that the performance objective for protection of the general 

population will be met because the 100-meter point of assessment used in the analyses for both 

the groundwater and air pathways is conservative. The SRS Land Use Plan [31] requires Federal 

ownership and control of the site well beyond 100 years after closure of SDF. DOE 5400.5 

precludes release of the area unless the radiological hazard meets the requirements of DOE 

5400.5 Chapter 4, which requires virtual perpetual DOE control34. No unrestricted use of the 

land or groundwater will be permitted for the central portion of the site, which includes SDF. 

Thus, a member of the public could not contact the groundwater in the vicinity of SDF. Due to 

 

34 Although for the purposes of this analysis, it is assumed that there is possible loss of institutional control for 

periods beyond 100 years which results in unrestricted access of the site. DOE requirements do not permit 

termination of necessary controls or unrestricted access to areas requiring radiological control. The Atomic Energy 

Act (AEA), and the Department of Energy Organization Act (DOA), assigns to the Department the responsibility to 

protect the public, the environment, and property from the hazards associated with its research, development, 

production, or other activities. This responsibility, which lacks any limit in time, includes protecting the public and 

environment from radiation or radioactive material. DOE policy (DOE P 454.1, Use of Institutional Controls) and 

requirements (DOE 5400.5, Radiation Protection of the Public and Environment) mandate continued control of 

property until the radiological hazard associated with this property is reduced to levels at which regulation under the 

AEA is no longer needed to ensure protection of the public and environment. By these requirements, DOE 

establishes “durable institutional controls” that are effectively similar to those required in 10 CFR 20.1403(e)(2)(ii) 

for sites requiring long-term control. 
February 28, 2005 

Page 56 of 94 

the restrictions in the SRS Land Use Plan, a member of the public could only contact potentially 

contaminated surface water off-site, approximately six miles from the facility at the mouth of 

Upper Three Runs creek. Furthermore, concentrations of SDF radionuclides in that surface water 

will be less than that assessed in the SA at 100 meters from SDF due to decay and other natural 

processes. 

The NRC performance objective also provides that reasonable effort should be made to maintain 

releases of radioactivity in effluents to the general environment as low as reasonably achievable 

(ALARA). The solidified saltstone waste form and the SDF vault design work together to reduce 

releases from SDF to the environment to very low levels, as demonstrated by the dose 

calculations above. The engineered grout and the SDF vaults are designed to retard migration of 

the radionuclides from the grouted waste form [32]. 

Therefore, disposal of the solidified low-activity salt waste streams at SDF will meet the limits in 

the NRC performance objective in 10 CFR 61.41. 

7.2.2 Protection of Individuals from Inadvertent Intrusion (10 CFR 61.42) 

10 CFR 61.42 states: 

Design, operation, and closure of the land disposal facility must ensure protection of any 

individual inadvertently intruding into the disposal site and occupying the site or contacting 

the waste at any time after active institutional controls over the disposal site are removed. 

To determine whether disposal of the salt waste in the SDF would meet this performance 

objective, DOE used the 2002 SA in a manner similar to the way it used that document to 

evaluate compliance with the 10 CFR 61.41 performance objective related to the protection of the 

general public. As discussed previously, the SA updated the 1992 PA. 

As a matter of background, the 1992 PA analyzed two cases for inadvertent intrusion: the 

degraded case which assumed that the grout in the vaults began to lose their physical integrity 

immediately upon closure of the SDF; and the intact case which assumed that the vault in the 

grout structure remained intact. Four chronic exposure scenarios were analyzed in detail for both 

in the PA: two agricultural scenarios, a residential scenario, and a well-drilling scenario. In all 

exposure scenarios considered for inadvertent intruder, the maximum peak dose was 0.6 mrem 

per year at greater than 100 years for the intact case. The maximum predicted dose for the 

degraded case was 50 to 110 mrem per year at greater than 1,000 years. 

The 2002 SA did not analyze the intact and degraded cases as was done in the PA. Rather the SA 

analyzed a more realistic case which included degradation of the engineered barriers over time. 

The SA analyzed only the residential intruder scenario for chronic exposure35 based on the thencurrent design of the SDF. Under that design, the vaults would be covered with a one-meter thick 

 

35 The SA [25] evaluates chronic exposure scenarios because, as explained in the SA and the references cited therein, 

chronic scenarios would bound acute scenarios 
February 28, 2005 

Page 57 of 94 

concrete barrier and a clay/gravel drainage system comprised of approximately 1.6 feet of clay with 

an overlying approximately 0.5 foot layer of gravel. Above the clay/gravel drainage system, a 

minimum of one foot of backfill would be placed. Above this layer of backfill, a moisture barrier 

of approximately 2.5 feet of clay and an overlying layer of approximately one foot of gravel 

would be installed. A second backfill layer, approximately 2.5 feet thick, would be placed over 

the moisture barrier. Finally, an approximately 0.5 foot layer of topsoil would be placed on the 

top layer of backfill to complete closure of SDF. This sequence of layers would provide a 

minimum of 9.6 feet of cover for each vault [25]. Therefore, this cover system when combined 

with the structure of the vault would make excavation into the waste during 10,000 years 

extremely unlikely and, thus, the agricultural and drilling scenarios were not analyzed. 

The SA evaluated the residential scenario at 10036, 1,000 and 10,000 years after SDF closure, and 

developed disposal limits for radionuclides at each of these points of time. In the residential 

scenario, the intruder is assumed to excavate no more than three meters in building a home over a 

disposal vault. The resident scenario assumes that excavation into the saltstone grout will not 

occur because the intruder encounters the one-meter concrete barrier over the vault roof that 

cannot be penetrated at 100 years after closure with excavation equipment typically used in the 

vicinity of SRS, and because the sub-surface roof of the SDF vaults will be below the typical 

maximum depth of a residential excavation at 1,000 and 10,000 years after SDF closure. 

Exposure in the resident scenario is assumed to occur after the home is constructed. The only 

relevant dose pathway would be external radiation exposure from radionuclides in the waste 

while residing in the home because the presence of uncontaminated cover material above the 

waste would preclude inhalation or ingestion exposure. 

Evaluation of the scenario at 100 years, when the engineered barriers (i.e., the grout above the 

saltstone, the vault roof, and the grout above the roof) are assumed to be intact, results in the 

intruder’s home being constructed on top of the concrete barrier. This scenario was used to 

determine disposal limits for shorter-lived radionuclides in the waste. Evaluation of the 

residential scenario at 1,000 and 10,000 years, when the engineered barriers are assumed to have 

partly failed (i.e., have partially lost their physical integrity) and provide less deterrence to 

intrusion, resulting in a lesser thickness of shielding above the waste, was used to determine 

disposal limits for longer-lived radionuclides. The thickness of uncontaminated material above 

the waste was assumed to be the same at these two later times because the upper 0.9 meter of the 

cover would have eroded and the depth of the intruder’s excavation would be limited to three 

meters. The residential scenario at 1,000 years is important for radionuclides having longer-lived 

decay products. The residential scenario at 10,000 years is important only for those radionuclides 

with decay products whose activities increase with time beyond 1,000 years. 

 

36 Human intrusion was assumed to be precluded in the first 100 years following closure based on the assumption 

that SRS boundaries remain unchanged and the land remains under Federal government ownership and control 

utilizing the existing site security system (e.g., fences, controlled site access, prohibited residential use of the land or 

groundwater). This assumption is consistent with the SRS Future Land Use Plan [31], which requires continual 

Federal ownership and control of the site well beyond 100 years. No unrestricted use of the land or groundwater will 

be permitted for the central portion of the site, which includes SDF. 
February 28, 2005 

Page 58 of 94 

The residential scenario was evaluated in the SA using the DOE 100 mrem per year dose limit to 

calculate DOE disposal limits for SDF, in a manner similar to the approach used in the SA for 

calculating disposal limits based on projected dose to the general public (from the groundwater 

and air pathways). DOE established these limits by calculating the radionuclide concentration in 

the low-activity salt waste that, when processed into saltstone grout and disposed of in SDF, 

would result in a dose to the intruder of 100 mrem in a year for each radionuclide. DOE then 

examined the radionuclide inventory anticipated to be disposed of against those limits and 

concluded that the inventory was below the limits and, therefore, that there was reasonable 

assurance that the SDF would protect the inadvertent intruder. 

For the purposes of this draft 3116 Determination, DOE has derived estimated doses to the 

inadvertent intruder over a 1,000 and 10,000 year period using the radionuclide limits developed 

in the SA and the currently planned radionuclide inventory in the waste streams to be disposed of. 

DOE used the same methodology for developing these doses that was used to develop the 

projected dose to the general public. The resulting maximum estimated dose over the period 

ending 1,000 years after SDF closure is 0.09 mrem per year and over the period ending 10,000 

years after closure is 0.13 mrem per year, thereby demonstrating that there is reasonable 

assurance of protection of the inadvertent intruder. 

In addition, a qualitative assessment of the engineered barriers provides reasonable assurance that 

the SDF will protect any individual inadvertently intruding into the SDF and occupying the site 

or contacting the waste after active institutional controls are removed. Although detailed closure 

plans will be further developed in the future, current plans include the following actions. Backfill 

of native soil will be placed around the vaults. The vaults will be covered with a one-meter thick 

concrete barrier and a clay/gravel drainage system comprised of approximately 1.6 feet of clay with 

an overlying approximately 0.5 foot layer of gravel. The clay/gravel drainage system is intended to 

prevent the buildup of perched water above the vaults. Above the clay/gravel drainage system, a 

geotextile fabric to maintain layer separation from overlying backfill and a minimum of one-foot 

of backfill will be placed. Above this layer of backfill, a moisture barrier will be installed. This 

upper moisture barrier will consist of approximately 2.5 feet of clay and an overlying layer of 

approximately one-foot of gravel. A geotextile fabric will also be placed on this upper gravel 

layer and a second backfill layer, approximately 2.5 feet thick, will be placed over the moisture 

barrier. Finally, an approximately 0.5 foot layer of topsoil will be placed on the top layer of 

backfill to complete closure of the SDF. This sequence of layers will provide a minimum of 9.6 

feet of cover for each vault [25]. These actions in conjunction with the structural integrity of the 

SDF will persist well into the future. As shown above, these barriers will minimize potential 

doses for any intruder to provide reasonable assurance that any inadvertent intruder would be 

protected from any significant radiation exposure at both the 1,000 and 10,000 year assessment 

time. Moreover, the SRS Land Use Plan [31] requires Federal ownership and control of the site 

well beyond 100 years after closure of the SDF. DOE 5400.5 precludes release of the area unless 

the radiological hazard meets the requirements of DOE 5400.5 Chapter 4, which requires virtual 

perpetual DOE control. No unrestricted use of the land or groundwater will be permitted for the 

central portion of the site, which includes SDF. Moreover, as shown above, in the unlikely event 

that there were inadvertent human intrusions, the resulting doses would be extremely low. 
February 28, 2005 

Page 59 of 94 

Thus, the disposal of the solidified low-activity salt waste in SDF meets the NRC inadvertent 

intruder performance objective in 10 CFR 61.42. 

7.2.3 Protection of Individuals During Operations (10 CFR 61.43) 

10 CFR 61.43 states: 

Operations at the land disposal facility must be conducted in compliance with the standards 

for radiation protection set out in part 20 of this chapter, except for releases of radioactivity 

in effluents from the land disposal facility, which shall be governed by Section 61.41 of this 

part. Every reasonable effort shall be made to maintain radiation exposures as low as is 

reasonably achievable. 

The performance objective in 10 CFR 61.43 cross-references “the standards for radiation 

protection in Part 20”. Consistent with Section 3116 (a) of the NDAA, the cross-referenced 

“standards for radiation protection” in 10 CFR Part 20 that are considered in detail in this draft 

3116 Determination are the dose limits for the public and the workers during disposal operations 

set forth in 10 CFR 20.1101(d), 10 CFR 20.1201(a)(1)(i), 10 CFR 20.1201(a)(1)(ii), 10 CFR 

20.1201(a)(2)(i), 10 CFR 20.1201(a)(2)(ii), 10 CFR 20.1201(e), 10 CFR 20.1208(a), 10 CFR 

20.1301(a)(1), 10 CFR 20.1301(a)(2), and 10 CFR 20.1301(b)37. As will be discussed in the 

following sections, these dose limits correspond to the dose limits in 10 CFR Part 835 and 

relevant DOE Orders which establish DOE regulatory and contractual requirements for DOE 

facilities and activities. The following subsections show that disposal operations at SDF for the 

 

37 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 subsection (a) 1 through (3) of Section 3116 are those which pertain to classification and disposal of waste, and 

the radiation protection standards for disposal. The Joint Explanatory Statement of the Committee of Conference in 

Conference Report 108-767, accompanying H.R. 4200 (the NDAA), also confirms that 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 Part 20 (as cross-referenced in the performance objective at 10 CFR 

61.43) which are relevant in the context of Section 3116 of the NDAA are the dose limits for radiation protection of 

the public and the workers during disposal operations, and not those which address general licensing, administrative, 

programmatic, or enforcement matters administered by NRC for NRC licensees. Accordingly, this draft 3116 

Determination addresses in detail the radiation dose limits for the public and the workers during disposal operations 

that are contained in the provisions of 10 CFR Part 20 referenced above. Although 10 CFR 20.1206(e) contains 

limits for planned special exposures for adult workers, there will not be any such planned special exposures for 

disposal operations at SDF for the solidified salt waste. Therefore, this limit is not discussed further in this draft 

3116 Determination. Likewise, 10 CFR 20.1207 specifies occupational dose limits for minors. However, there will 

not be minors working at SDF who will receive an occupational dose. Therefore, this limit is not discussed further in 

this draft 3116 Determination. 
February 28, 2005 

Page 60 of 94 

solidified salt waste streams meet these dose limits and that doses will be maintained as low as 

reasonably achievable38 [26]. 

7.2.3.1 Air Emissions Limit for Individual Member of the Public (10 CFR 20.1101(d)) 

The NRC regulation at 10 CFR 20.1101(d) provides in relevant part: 

[A] constraint on air emissions of radioactive material to the environment, excluding 

Radon-222 and its daughters, shall be established ... such that the individual member of 

the public likely to receive the highest dose will not be expected to receive a total 

effective dose equivalent in excess of 10 mrem (0.1 mSv) per year from these emissions. 

DOE similarly limits doses from air emissions to the public to 10 mrem per year in DOE Order 

5400.5. DOE is also subject to and complies with the Environmental Protection Agency 

requirement in 40 CFR 61.92, which has the same limit39. The estimated dose per year from 

airborne emissions to the maximally exposed individual member of the public located at or 

beyond the SRS site boundary from all operations at SRS ranged from 0.04 mrem to 0.07 mrem 

from 1999 through 2003 [33, 34, 35, 36, 37]40. These values for all of the SRS operations (not 

just disposal operations at SDF) are well below the dose limit specified in 10 CFR 20.1101(d). 

DOE anticipates that the dose from air emissions from SDF following the receipt of the lowactivity salt waste streams associated with DDA, ARP/MCU, and SWPF operations will be a 

small fraction of the dose from all operations at SRS. 

7.2.3.2 Total Effective Dose Equivalent Limit for Adult Workers (10 CFR 20.1201(a)(1)(i)) 

The NRC regulation at 10 CFR 20.1201(a), concerning occupational dose limits for adults, 

provides in relevant part: 

(a) [C]ontrol the occupational dose to individual adults, except for planned special 

exposures...to the following dose limits. 

(1) An annual limit, which is the more limiting of – 

(i) The total effective dose equivalent being equal to 5 rems (0.05 Sv); 

 

38 In addition, 10 CFR Part 835, like Part 20 for NRC licensees, includes requirements that do not set dose limits, 

such as requirements for radiation protection programs, monitoring, entrance controls for radiation areas, posting, 

records, reporting, and training. 

39 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 per year. 

40 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. The historical doses shown above do not include the incremental 

increase from disposal of the low-activity salt waste from disposal of DDA, ARP/ MCU, and SWPF in SDF. 
February 28, 2005 

Page 61 of 94 

*** 

DOE’s regulation at 10 CFR 835.202 (a)(1) has the same annual dose limit for the annual 

occupational dose to general employees41. For the occupational dose to adults during disposal 

operations at SDF, the TEDE per year will be controlled as low as reasonably achievable below 5 

rem [38]. In this regard, occupational doses to workers have been well within the annual limits 

specified in 10 CFR 1201 (2) for all work activities at SRS [39]. The highest dose received by an 

SRS worker in 2003 was 1808 mrem TEDE. There was close to zero total exposure received by 

the SRS workforce for the SPF and SDF activities. Since 1998, the highest dose received by an 

SRS worker has been at or below 1808 mrem per year. The total dose received by workers at 

SPF and SDF since 1998 is 35 mrem, which is well below the limit specified in 10 CFR 

20.1201(a). Furthermore, operations will continue to consist predominantly of mechanical 

mixing of low-activity salt solutions with cementitious material and then mechanical transfer to 

SDF; therefore, little hands-on work at SDF is anticipated. Thus, the total effective dose 

equivalent to workers from disposal of the solidified low-activity salt waste at SDF is expected to 

remain well below the NRC limit. 

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

20.1201(a)(1)(ii)) 

The NRC regulation at 10 CFR 20.1201(a), concerning occupational dose limits for adults, 

provides in relevant part: 

(a)[C]ontrol the occupational dose to individual adults, except for planned special 

exposures...to the following dose limits. 

(1) An annual limit, which is the more limiting of – 

*** 

(ii) The sum of the deep-dose equivalent and the committed dose equivalent to 

any individual organ or tissue other than the lens of the eye being equal to 

50 rems (0.5 Sv). 

The dose limit specified in 10 CFR 20.1201(a)(1)(ii) is the same as that specified in 10 CFR 

835.202 (a)(2). For the occupational dose to adults during disposal operations at SDF, the sum of 

 

41 DOE’s regulation requires that the occupational dose per year for general employees shall not exceed both a total 

effective dose equivalent of 5 rems 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 rems. NRC’s regulation 

specifies that either of these two limits shall be met by NRC licensees, whichever is more limiting. This draft 3116 

Determination will show that DOE will meet the more stringent of the dose limits in 10 CFR Part 835 and the 

relevant dose limits in 10 CFR Part 20 (see subsection 7.2.3 of this draft 3116 Determination for listing of relevant 

dose limits in 10 CFR Part 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 10 CFR 835.202(a)(2)), are discussed in 

separate subsections in this draft 3116 Determination. 
February 28, 2005 

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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 as low as reasonably achievable below a 

maximum of 50 rem per year [38]. SRS Engineering Standard 01064, Radiological Design 

Requirements [40], provides that the design basis annual occupational exposure limits for any 

organ or tissue other than the eye cannot exceed 10 rem per year, which is well below the NRC 

limit of 50 rem per year. Furthermore, operations will predominantly consist of mechanical 

mixing of low-activity salt solutions with cementitious material and then mechanical transfer to 

SDF, so that little hands-on work, and little, if any, associated exposure, is anticipated for 

disposal operations in SDF. 

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

20.1201(a)(2)(i)) 

The NRC regulation at 10 CFR 20.1201(a), concerning occupational dose limits for adults, 

provides in relevant part: 

(a) [C]ontrol the occupational dose to individual adults, except for planned special 

exposures...to the following dose limits. 

*** 

(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 at 10 CFR 835.202 (a)(3). For the occupational dose to adults during disposal 

operations at SDF, the annual dose limit to the lens of the eye will be controlled to as low as 

reasonably achievable below maximum of 15 mrem per year [38]. SRS Engineering Standard 

01064, Radiological Design Requirements [40], provides that the design basis annual 

occupational exposure limits for the lens of the eye cannot exceed 3 rem per year, which is well 

below the NRC limit of 15 rem per year. Furthermore, as noted previously, operations 

predominantly consist of mechanical mixing of low-activity salt solutions with cementitious 

material and then mechanical transfer to SDF; therefore, little hands-on work or exposure is 

anticipated during disposal operations at SDF. 

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

Extremities for Adult Workers (10 CFR 20.1201(a)(2)(ii)) 

The NRC regulation at 10 CFR 20.1201(a), concerning occupational dose limits for adults, 

provides in relevant part: 

(a) [C]ontrol the occupational dose to individual adults, except for planned special 

exposures...to the following dose limits. 
February 28, 2005 

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*** 

(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 disposal 

operations at SDF, 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 as low as reasonably achievable 

below a shallow-dose equivalent of 50 rem per year [38]. 

7.2.3.6 Limit on Soluble Uranium Intake (10 CFR 20.1201(e)) 

The NRC regulation at 10 CFR 20.1201(e), concerning occupational dose limits for adults, 

provides in relevant part: 

(e) In addition to the annual dose limits, ... limit the soluble uranium intake by an 

individual to 10 milligrams in a week in consideration of chemical toxicity [.] 

In addition to the annual dose limits to adults during disposal operations at SDF, the soluble 

uranium intake by an individual is controlled to less than 10 milligrams (mg) per week. DOE 

Order 440.1A requirements 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 per cubic meter which is the same as noted in 10 CFR 20 Appendix B 

footnote 3) or the Occupational Safety and Health Administration (OSHA) Permissible Exposure 

Limit (PEL) (0.05 mg per cubic meter) [41]. The OSHA PEL limit for soluble uranium, which 

equates to a soluble uranium intake of 2.4 mg per week, is the more restrictive of the two and 

therefore is the limit imposed for disposal operations at SDF. Accordingly, the soluble uranium 

intake, if any, during disposal operations at the SDF will be controlled to 2.4 mg per week, which 

is below the NRC limit in 10 CFR 20.1201(e). Moreover, the low-activity waste disposed of in 

the SDF will be in a solid form, and thus there will not be soluble uranium in the solidified waste 

susceptible to worker intake. In addition, as noted previously, disposal operations will generally 

be performed mechanically, without worker exposure during normal mechanical operations. 

7.2.3.7 Dose Equivalent to an Embryo/Fetus (10 CFR 20.1208(a)) 

The NRC regulation at 10 CFR 20.1208(a), concerning the dose equivalent to an embryo/fetus, 

provides in relevant part: 
February 28, 2005 

Page 64 of 94 

(a) [E]nsure that the dose equivalent to the embryo/fetus during the entire pregnancy, due to 

the occupational exposure of a declared pregnant woman, does not exceed 0.5 rem (5 

mSv). 

DOE’s regulation at 10 CFR 835.206 (a) has the same dose limit. For the occupational dose to an 

embryo/fetus during disposal operations at SDF, 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 [38]. 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 carefully 

track exposure. 

7.2.3.8 Dose Limits for Individual Members of the Public (10 CFR 20.1301(a)(1)) 

The NRC regulation at 10 CFR 20.1301(a), concerning dose limits for individual members of the 

public, provides in relevant part: 

(a) [C]onduct operations so that - 

(1) The total effective dose equivalent to individual members of the public ...does not 

exceed 0.1 rem (1 mSv) in a year, exclusive of the dose contributions from 

background radiation, from any medical administration the individual has 

received, from exposure to individuals administered radioactive material and 

released...., from voluntary participation in medical research programs, and from 

the ...disposal of radioactive material into sanitary sewerage [.] 

DOE Order 5400.5 II.1.a similarly limits public doses to less than 100 mrem per year. However, 

DOE’s application of the limit is more restrictive in that it requires DOE 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 disposal 

operations at SDF, the TEDE limit to an individual member of the public will be controlled to 

less than 0.1 rem per year. The air pathway is the predominate pathway for doses to the public 

from SRS operations, including disposal operations at SDF, and, as discussed in subsection 

7.2.3.1 of this draft 3116 determination, doses from the air pathway to members of the public 

have been, and are expected to continue to be, well below the 0.1 rem annual limit specified in 10 

CFR 20.1301(a) [38]. 

7.2.3.9 Dose Limits for Individual Members of the Public (10 CFR 20.1301(a)(2)) 

The NRC regulation at 10 CFR 20.1301(a), concerning dose limits for individual members of the 

public, provides in relevant part: 

(a) [C]onduct operations so that – 

 * * * 
February 28, 2005 

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(2) The dose in any unrestricted area from external sources, exclusive of the dose 

contributions from patients administered radioactive material and released ..., 

does not exceed 0.002 rem (0.02 millisievert) in any one hour. 

DOE’s regulation at 10 CFR 835.602 establishes the expectation that the TEDE in Controlled 

Areas will be less than 0.1 rem in a year. For individual members of the public during disposal 

operations at SDF, operations will be conducted such that the dose in any unrestricted area from 

external sources, exclusive of the dose contributions from patients administered radioactive 

material, will be less than 0.00005 rem per hour above background. Westinghouse Savannah 

River Company (WSRC) Manual 5Q, Chapter 2, Article 232, also restricts the TEDE in 

Controlled Areas to less than 0.1 rem in a year. 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 in a year or greater. In addition, SRS has established 

Radiological Buffer Areas around posted Radiological Areas. Standard SRS practice is to 

assume a 2,000 hour per year continuous occupancy at the outer boundary of these areas and; 

therefore, the dose rate at a Radiological Buffer Area boundary is 0.00005 mrem/hr (0.1 

rem/2,000 hr = 0.00005 rem/hr). Since the Controlled Area encompasses a Radiological Buffer 

Area, it is ensured that the dose in the Controlled Area (but outside of Radioactive Material Areas 

and Radiological Buffer Areas) will be less than 0.1 rem in a year [42]. Therefore, SRS 

implementation of the provisions at 10 CFR 835.602 and 10 CFR 835.603 limit provides a more 

restrictive limit than the dose limit specified in 10 CFR 20.1301(a)(2). 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 per year. 

7.2.3.10 Dose Limits for Individual Members of the Public (10 CFR 20.1301(b)) 

The NRC regulation at 10 CFR 20.1301(b), concerning dose limits for individual members of the 

public, provides in relevant part: 

(b) If ... members of the public [are permitted] to have access to controlled areas, the limits 

for members of the public continue to apply to those individuals. 

DOE’s regulation at 10 CFR 835.208 has the same dose limit. The TEDE limit to an individual 

member of the public granted access to controlled areas42 during disposal operations at SDF will 

be controlled to 0.1 rem per year [38]. Furthermore, training is required for individual members 

of the public for entry into controlled areas. In addition, to ensure no member of the public 

 

42 10 CFR 20.1003 defines restricted areas as an area, access to which is limited ... for the purpose of protecting 

individuals against undue risks from exposure to radiation and radioactive materials. This is the same as the 

definition in 10 CFR 835.2 for a controlled area. 
February 28, 2005 

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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 per year43.

7.2.3.11 As Low As Reasonably Achievable (10 CFR 20.1003) 

The NRC regulation at 10 CFR 20.1003 defines ALARA (As Low as Reasonably Achievable) 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...[.] 

DOE has a similar requirement, and DOE’s regulation at 10 CFR 835.2 defines ALARA as “... 

the approach to radiation protection to manage and control exposures (both individual and 

collective) to the work force and to the general public to as low as is reasonable...”. For 

radiological work activities during disposal operations at SDF, 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 at 10 CFR 

835.101(c) requires the contents of each Radiation Protection Program (RPP) to include formal 

plans and measure for applying the as-low-as-reasonably-achievable process to occupational 

exposure. 

7.2.3.12 Reasonable Assurance 

Measures that provide reasonable assurance that disposal operations at SDF will comply with the 

applicable dose limits and with the ALARA provisions include: (1) the documented RPP; (2) the 

DSA; (3) design; (4) regulatory and contractual enforcement mechanisms; and (5) access 

controls, training, and dosimetry. These measures are discussed in the following paragraphs. In 

addition, the following discusses the exposure history at SRS and at SDF. 

7.2.3.13 SRS Radiation Protection Program 

The Department of Energy regulates occupational radiation exposure at its facilities through 10 

CFR Part 835, Occupational Radiation Protection. Part 835 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 Part 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. 

 

43 10 CFR 20.1301(d) allows licensees to request NRC authorization to allow an individual member of the public to 

operate up to an annual dose limit of 0.5 rem (5 mSv). 10 CFR 835 is more restrictive for the dose to an individual 

member of the public with a limit of 0.1 rem maximum annual dose as discussed in Subsection 7.2.3.8. Therefore, 

this limit is not discussed further in this waste determination. 
February 28, 2005 

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Pursuant to Part 835, activities at SRS, including disposal operations at SDF, must be conducted 

in compliance with the documented RPP for SRS as approved by DOE. 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 Part 835. The RPP also includes 

formal plans and measures for applying the ALARA process. 

The requirements of Part 835 as contained in the approved RPP are incorporated in the WSRC’s 

Standards and Requirements Implementing Document system. The Standards and Requirements 

Implementing Document system links the requirements of Part 835 to the site-level and lowerlevel implementing policies and procedures that control radiological work activities conducted 

across the site. These requirements are primarily contained in the WSRC 5Q Manual, 

Radiological Control, and its lower tier manuals, e.g., WSRC 5Q1.1, Radiation and 

Contamination Control Procedures Manual, and WSRC 5Q1.2, Radiation Monitoring 

Procedures Manual. These procedures control the planning of radiological work, the use of 

radiation monitoring devices by employees, the bioassay program, the air monitoring program, 

the contamination control program, the ALARA program, the training of general employees, 

radiological workers, Radiological Control Inspectors, and health physics professionals and 

technicians, and the other aspects of an occupational radiation protection program as required by 

Part 835. 

7.2.3.14 Documented Safety Analysis 

A DSA [43] has been approved by DOE for operation of SPF and SDF in accordance with 10 

CFR Part 830. As the first step in the development of the DSA, a formal Consolidated Hazards 

Analysis (CHA) [44] was performed at the Saltstone Facility to evaluate the potential risk of 

operations to the workers and the public. The CHA was performed by a group of approximately 

20 subject matter experts, with expertise in the fields of operations, engineering, industrial 

hygiene, radiological protection, environmental compliance, and maintenance. 

The CHA 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 Saltstone Facility are evaluated against the criterion listed in DOE-STD-1027-92, Hazard 

Categorization and Accident Analysis Techniques for Compliance with DOE Order 5480.23, 

Nuclear Safety Analysis Reports, to determine the overall hazard classification of the facility. It 

was determined by the CHA team that the hazard classification of the Saltstone Facility was 

Hazard Category 3, which is the lowest hazard classification and denotes a potential for only 
February 28, 2005 

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localized consequences to workers at the facility and no potential for significant consequences to 

other workers at the site or to members of the public. 

During the third and final phase of the CHA, 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 was qualitatively determined, 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 those events which 

presented negligible risk to the facility workers and public. As a result of the hazard evaluation 

for the Saltstone Facility, all normal operational events were determined to present negligible risk 

to the workers and public (i.e., exposure < 5 rem to facility workers), and were thus removed 

from further evaluation. For purposes of this CHA, the waste inventory and curie concentrations 

were assumed to be greater than currently planned for the DDA, ARP/MCU, and SWPF streams. 

The DSA analyzed the hazards that were identified in the CHA that could impact facility workers 

during normal operations and accident conditions, and specifically included radiation exposure 

hazards. The DSA identified the basis for derivation of the Saltstone Facility Technical Safety 

Requirements (TSR) [45] and also discussed summary descriptions of the key features of safety 

management programs at SRS as they pertain to SDF. 

The Saltstone Facility TSR document identified the administrative controls that are necessary to 

achieve safe operations at SDF. In part, these TSR administrative controls require: (1) that a 

facility manager be assigned who is accountable for safe operation and in command of activities 

necessary to maintain safe operation; (2) that personnel who carry out radiological controls 

functions for SDF have sufficient organizational freedom to ensure independence from operating 

pressures; (3) that SDF personnel receive initial and continuing training including radiological 

control training; and (4) that an RPP shall be prepared consistent with 10 CFR Part 835. The 

DSA determined that the administrative controls identified in the TSR are sufficient to ensure 

worker protection in accordance with 10 CFR Part 835. 

In addition, the design requirements for SPF and SDF implemented 10 CFR Part 835 [46] and, in 

particular, implemented ALARA principles. The design is currently being upgraded to reflect the 

radionuclide concentrations in the low-activity waste streams to be received at SPF and SDF 

from planned Interim Salt Processing facilities and SWPF. While the upgraded design is not yet 

complete, based on the current SPF and SDF design, it is estimated that occupational exposures 

for SPF and SDF workers will be at least an order of magnitude lower than the 10 CFR Part 835 

dose limit of 5 rem per year during both Interim Salt Processing and SWPF operation. 

7.2.3.15 Radiological Design for Protection of Occupational Workers and the Public 

New SRS radiological facilities and facility modifications including the ongoing Saltstone 

Facility modifications are designed to meet the requirements of 10 CFR 835 Subpart K, Design 

and Control [47]. SRS Engineering Standard 01064, Radiological Design Requirements [40], 
February 28, 2005 

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provides the requirements necessary to ensure compliance with 10 CFR Part 835. The standard 

refers to 10 CFR Part 835, DOE orders, DOE standards, DOE handbooks, national consensus 

standards, SRS manuals, SRS engineering standards, SRS engineering guides, and site operating 

experience in order to meet the 10 CFR Part 835 specific requirements and additional 

requirements to ensure the design provides for protection of the worker and the environment. 

The standard covers the full spectrum of radiological design requirements and not just radiation 

exposure limits. The following are the specific areas addressed in the standard: radiation 

exposure limits; facility and equipment layout; area radiation levels; radiation shielding; internal 

radiation exposure; radiological monitoring; confinement; and ventilation. 

The design requirements for several of the important sections of the standard are highlighted in 

order to understand the design limits and philosophy for SRS designs. The first area of interest is 

the radiation exposure limits. The following is an excerpt of the standard which presents the 

exposure limits and philosophy for both external and internal radiation exposure. 

“During the design of new facilities or modification of existing facilities, the 

design objective for controlling personnel exposure from external sources of 

radiation in areas of continuous occupancy (2000 hours per year) shall be to 

maintain exposure levels below an average of 0.5 mrem per hour and as far below 

this average as is reasonably achievable. The design objectives for exposure rates 

for potential exposure to a radiological worker where occupancy differs from the 

above shall be As Low As Reasonably Achievable (ALARA) and shall not exceed 

the external limits in Table 5-1. Regarding the control of airborne radioactive 

material, the design objective shall be, under normal conditions, to avoid releases 

to the workplace atmosphere and in any situation, to control the inhalation of such 

material by workers to levels that are ALARA; confinement and ventilation shall 

normally be used [6.3, 6.12]. Table 5-1 summarizes the design basis external 

radiation exposure limits. 

 Design Basis Annual Occupational Radiation Exposure Limits 

 Type of Exposure Limit (rem) 

 Whole Body TEDE 1.0 

 Internal CEDE 0.5 

 Lens of Eye 3 

 Extremity 10 

 Any Organ (other than eye) or 

Tissue 

10 

 To meet the Site’s no deliberate intake policy, engineered controls will be 

evaluated and implemented to ensure that, under normal operating conditions, no 

worker will receive a deliberate intake of radionuclides (i.e., CEDE=0 rem). As a 

result, the TEDE limit will be independent of the CEDE limit. The 0.5 rem 

CEDE limit in Table 5.1 is to be applied to potential intakes from anticipated 
February 28, 2005 

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potential releases or anticipated off-normal maintenance. Under these anticipated 

potential conditions, engineered controls will be evaluated and implemented to 

minimize the potential for workers to receive intakes that will exceed the 0.5 rem 

CEDE. This evaluation will not take credit for the use of respiratory protection. 

The dose to any member of the public or a minor exposed to radiation at a DOE 

facility shall not exceed 0.1 rem TEDE in a year.” 

The facility design also incorporates radiation zoning criteria in order to ensure the exposure 

limits presented above are met by providing adequate radiation shielding. Areas in which nonradiological workers are present are assumed to have continuous occupancy (2,000 hours per 

year) and are designed to a dose rate less than 0.05 mrem per hour to ensure that the annual dose 

is less than 100 mrem. Other zoning criteria are established to ensure radiological worker doses 

are ALARA and less than 1,000 mrem per year to meet the 10 CFR 835.1002 design 

requirements. 

The design is also required to provide necessary radiological monitoring or sampling for airborne 

and surface contamination to ensure the engineered controls are performing their function and, in 

the event of a failure or upset condition, workers are warned and exposures avoided. 

Radiological protection personnel ensure the requirements of the standard are addressed and 

presented in design summary documentation. The incorporation of all the radiological design 

criteria in the engineering standard ensures the requirements of 10 CFR Part 835 are met and the 

design provides for the radiological safety of the workers and environment. 

7.2.3.16 Regulatory and Contractual Enforcement 

Any violation of the requirements in 10 CFR Part 835 is subject to civil penalties pursuant to 

section 234A of the Atomic Energy Act of 1954, as amended, 42 USC 2011 et seq., as 

implemented by DOE regulations in 10 CFR Part 820. In addition, the requirements in 10 CFR 

Part 835 and all applicable DOE Orders are incorporated into all contracts with DOE contractors, 

including WSRC, the DOE contractor for disposal operations at SDF as well as other operations 

at SRS. DOE enforces these contractual requirements through contract enforcement measures, 

including the reduction of contract fees. 

7.2.3.17 Access Controls, Training, Dosimetry, and Monitoring 

Training or an escort is required for individual members of the public for entry into controlled 

areas. In addition, use of dosimetry is required if a member of the public is expected to enter a 

controlled area and exceed 0.05 rem per year to ensure no member of the public exceeds radiation 

exposure limits [48]. 

In addition, worker radiation exposure monitoring is performed for all workers expected to 

receive 100 mrem per year from internal and external sources of radiation to provide assurance 
February 28, 2005 

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that no worker exceeds radiation exposure limits and that all radiation dose are maintained as far 

below the limits as is reasonably achievable [48].

7.2.3.18 Occupational Radiation Exposure History for Savannah River Site 

The effectiveness of the radiation protection programs, including the effectiveness of oversight 

programs to ensure they are implemented properly is demonstrated by the occupational radiation 

exposure results as documented for 2003 [49]. The highest dose received by an SRS worker in 

2003 was 1808 mrem TEDE compared to the DOE Administrative Control Limit of 2000 mrem 

per year and the 10 CFR Part 835 limit of 5000 mrem per year. There was close to zero total 

exposure received by the SRS workforce for the SPF and SDF activities. Since 1998, the highest 

dose received by an SRS worker has been at or below 1808 mrem per year. The total dose 

received by workers at SPF and SDF since 1998 is 35 mrem. 

In addition, for all work activities, the average TEDE exposure for workers receiving a TEDE 

dose at SRS has been 75 mrem per year or less since 2001 [50]. It is expected that exposures for 

workers at SDF will be at or below this average based on design requirements and past 

experience with SDF operation. 

7.2.4 Long-Term Stability of the Disposal Site (10 CFR 61.44) 

10 CFR 61.44 states: 

The disposal facility must be sited, designed, used, operated, and closed to achieve long-term 

stability of the disposal site and to eliminate to the extent practicable the need for ongoing 

active maintenance of the disposal site following closure so that only surveillance, 

monitoring, or minor custodial care are required. 

The SDF is currently an operational low-level waste disposal facility to which additional disposal 

vaults will be constructed on a just-in-time basis. The facility is not scheduled for closure until 

completion of salt waste disposition currently estimated for 2019. A final closure plan will be 

developed in the future, as appropriate, to support closure of SDF. As demonstrated below, SDF 

will meet the performance objective at 10 CFR 61.44 for long-term stability of the disposal site. 

SITING

SDF is located within an area designated as Z-Area in the north-central portion of SRS. Z-Area 

was chosen as a waste disposal site based on considerations of depth to the water table, distance 

to surface water and the public, available surface area, surface topography, and its proximity to 

the salt processing facilities. The Saltstone Facility is located approximately 1.0 miles from the 

salt processing facilities and approximately 6.2 miles from the nearest SRS site boundary. ZArea lies on a local topographic high, at approximately 290 feet above mean sea level [10]. Land 

use in the area is classified as heavy industrial. Land within a 5-mile radius of SDF is entirely 

within the SRS boundaries and is used for either industrial purposes or as forested land. Z-Area 
February 28, 2005 

Page 72 of 94 

is bounded by two streams: Upper Three Runs and McQueen Branch, a tributary to Upper Three 

Runs. McQueen Branch is located approximately 1.0 miles to the East and 0.75 miles to the 

Northeast of Z-Area, and Upper Three Runs is located approximately 1.0 miles to the Northwest 

[10]. Under Z-Area, the minimum depth to the water table from the ground surface (290 feet 

elevation) in any given year is estimated to be at 43 feet (i.e., at 247 feet above mean sea level) 

on the basis of water table fluctuations [11]. Per SCDHEC Regulation R.61-107.16, the bottom 

of all current and future disposal vaults must be located at least two-feet above this historic high 

water table. 

There are no lakes or flow control structures on Upper Three Runs or its tributaries. The 

Probable Maximum Flood value for Upper Three Runs is 175 feet above mean sea level, which is 

substantially below the permitted maximum depth of the Saltstone vaults [10]. The 500-year and 

100-year flood plains are located at 153.1 feet and 151.7 feet above mean sea level, respectively 

[12]. 

Two major earthquakes have occurred within 100 miles of SRS. The largest known earthquake 

to affect SRS was the Charleston earthquake of 1886, with an epicenter approximately 90 miles 

from SRS and a magnitude of 6.6 on the Richter Scale. It is estimated that an earthquake of this 

magnitude would result in a peak ground acceleration of 0.10g at SRS [13]. A seismic evaluation 

of Z-Area shows that the soils beneath Z-Area are not susceptible to significant liquefaction for 

earthquakes having a peak ground acceleration less than or equal to 0.17g [14]. The second 

earthquake occurred approximately 90 – 100 miles from SRS, with an estimated magnitude of 

4.5. 

Most of the soils in the vicinity of SDF are sandy over a loamy or clayey subsoil. The dominant 

vegetation near SDF is forest with types ranging from scrub oak to cypress with pine being the 

primary forest in the area [6]. The siting of SDF, including the additional vaults where the 

solidified low-activity salt streams will be disposed, is such that it provides long-term stability 

consistent with this performance objective.

DESIGN

SDF currently contains two large concrete vaults divided into cells. Each of the cells will be 

filled with solidified waste. The grout used to solidify the waste provides primary containment of 

the waste and the walls, floor, and roof of the vaults provide secondary containment. 

Approximately 15 feet of overburden have been removed to prepare and level the site for vault 

construction. All vaults will be built at or slightly below the grade level that exists after the 

overburden and leveling operations are complete. The bottom of the saltstone grout monoliths 

will be at least five feet above the historic high water table beneath SDF thus, avoiding disposal 

of waste in a zone of water table fluctuation. 

Run-on and runoff controls are installed to minimize site erosion during the operational period. 

Approximately 160 acres are bounded by the Saltstone Facility perimeter fence. The impervious 

portion of this area is approximately 5% of the total Saltstone Facility property. Stormwater 

runoff from the impervious area of the Saltstone Facility is channeled through a pipe to the Z-01 
February 28, 2005 

Page 73 of 94 

sedimentation basin. McQueen Branch receives any discharges from the Z-01 basin. The basin 

is inspected at least once every three months. Because this is a relatively small drainage area, no 

special measures for runoff control are required [51]. 

The current vault (Vault #4) has dimensions of approximately 200 feet wide, by 600 feet in 

length, by 26 feet in height. The vault is divided into 12 cells, with each cell measuring 

approximately 100 feet by 100 feet. The vault is covered with a sloped, permanent roof that has a 

minimum thickness of four inches, and a minimum slope of 0.24 inches/foot. The vault walls are 

approximately 1.5 feet thick, with the base mat having a thickness of two feet. Operationally, the 

cells of the vault will be filled to a height of approximately 25 feet with solidified waste, and then 

a layer of uncontaminated grout, with an average thickness of two feet, will be poured to fill in 

the space between the solidified waste and the sloped roof to provide structural support. 

The six cells within Vault #4 that will be used during Interim Salt Processing have a leachate 

collection system installed within the vault walls. This prevents hydraulic pressure build-up 

against the vault walls. A sheet drain system is installed on the cell walls with a 12” pipe at the 

bottom of the walls to collect the leachate. Each cell has a drain line that can be accessed from 

the exterior of the cells. Prior to Interim Salt Processing, modifications are being made to these 

cells to install a pump and piping system to transfer the collected leachate from the cells to the 

grout transfer line upstream of the cells. The leachate will be pumped at a low rate into the grout 

line for re-introduction to the cells during grout-fill operations. 

The other current vault (Vault #1) has the dimensions of approximately 100 feet wide, by 600 

feet in length, by 25 feet in height. The vault is divided into six cells, with each cell measuring 

approximately 100 feet by 100 feet. 

DOE is currently evaluating design alternatives for future SDF vaults. These new vaults will be 

designed to meet all of the same standards as the existing vaults. The new design will provide the 

same level or a greater level of protection against infiltration of water, migration of radioactive 

contaminants, structural integrity, and radiation shielding. New designs are being considered 

primarily to maximize the processing capacity of the facility, to simplify the operation of the 

vaults, and to minimize the cost of construction. Therefore, the existing vaults are, and the future 

vaults will be, designed to achieve the required long-term stability of the disposal site. 

The Saltstone Facility was originally designed for processing materials with lower radionuclide 

concentrations than those currently planned during Interim Salt Processing. In preparation for 

Interim Salt Processing, several modifications in the design and operation of the facility are being 

undertaken. For example, several inches of grout which does not contain the low-activity waste 

stream will be poured on top of the low-level grout in the inactive vault cells to reduce the 

amount of exposure to facility workers (which will be primarily in the SPF rather than SDF 

portion of the facility) due to sky shine. 
February 28, 2005 

Page 74 of 94 

USE/OPERATION

SDF is permitted as an Industrial Solid Waste Landfill (ISWLF) site, as defined by SCDHEC 

Regulations R61-66 and R.61-107.16. Active disposal operations are planned to occur at SDF 

until around 2019. Following the filling of each vault cell, monitoring of the vaults occurs. 

Except for erosion control purposes, backfilling around the vaults will likely not be done prior to 

the completion of disposal activities. This will allow the vaults to be visually monitored for 

several years prior to closure. Any liquid that may accumulate as a result of rain water 

infiltration will be drained and returned to the process to avoid creating bulges in the vault walls. 

Routine surveillances of the filled vaults for structural integrity and soundness are conducted 

[39]. 

CLOSURE

The stability of the SDF closure design is an important element for meeting SDF performance 

objectives. The SDF facility design is focused on minimizing the contact of water with the 

stabilized waste form. The saltstone grout material in the stabilized waste form will help retard 

the migration of rainwater to the waste. The existing storm water control systems will be 

maintained (including monitoring, surveillance, and minor custodial care activities). To further 

ensure long-term stability of SDF, the land in Z-Area which includes SDF, will remain under the 

ownership of the Federal Government. The three counties making up SRS have zoning 

restrictions that prevent the purchase of property or the approval of building permits at SRS. SRS 

is zoned “Department of Energy ownership”. Residential use of this land will be prohibited via 

continued land use leasing restrictions [31]. No unrestricted use of the land or groundwater will 

be permitted for SDF. 

Closure operations will begin near the end of the active SDF disposal period, in approximately 

2019. Although final closure plans for SDF will be developed in the future, the following actions 

are contemplated today. Backfill of native soil will be placed around the vaults. The vaults will be 

covered with a one-meter thick concrete barrier and a clay/gravel drainage system comprised of 

approximately 1.6 feet of clay with an overlying approximately 0.5 feet layer of gravel. The 

clay/gravel drainage system is intended to prevent the buildup of perched water above the vaults. 

Above the clay/gravel drainage system, a geotextile fabric to maintain layer separation from 

overlying backfill and a minimum of one-foot of backfill will be placed. Above this layer of 

backfill, a laterally extensive moisture barrier will be installed. This upper moisture barrier will 

consist of approximately 2.5 feet of clay and an overlying layer of approximately one-foot of 

gravel. A geotextile fabric will also be placed on this upper gravel layer and a second backfill 

layer, approximately 2.5 feet thick, will be placed over the moisture barrier. Finally, an 

approximately 0.5 feet layer of topsoil will be placed on the top layer of backfill to complete 

closure of SDF. This sequence of layers will provide a minimum of 9.6 feet of cover for each 

vault [25]. 

Final closure of SDF will be accomplished by constructing a drainage system and revegetating 

the site. The drainage system will consist of a system of rip-rap lined ditches that intercept the 

gravel layer of the moisture barrier. These ditches will divert surface runoff and water 
February 28, 2005 

Page 75 of 94 

intercepted by the moisture barrier away from the disposal site. The drainage ditches will be 

constructed between rows of vaults and around the perimeter of SDF. The topsoil will be 

revegetated with bamboo which is a shallow-rooted climax species that evapotranspirates yearround in the SRS climate, thus removing moisture from the soil and decreasing the infiltration 

into the underlying disposal system [25]. Therefore, SDF will be 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 so that only surveillance, monitoring, or minor custodial care are 

required. 

Thus, the Saltstone Disposal Facility complies with the NRC long-term stability performance 

objective in 10 CFR 61.44. 
February 28, 2005 

Page 76 of 94 

8.0 STATE-ISSUED PERMIT

Section 3116 of the NDAA provides in relevant part: 

[T]he term “high level radioactive waste” does not include waste from reprocessing of spent 

nuclear fuel that the Secretary of Energy...in consultation with the Nuclear Regulatory 

Commission...determines – 

(3)(A) ... [W]ill be disposed of- 

*** 

(i) 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 ...[.] 

SDF is operated as a Class 3 Industrial Solid Waste Landfill (Permit #025500-1603). The State 

of South Carolina permitted SDF under its authority pursuant to Subpart D of the Resource 

Conservation and Recovery Act and Chapter 61, Article 107 (Solid Waste Management) of the 

South Carolina Health and Environmental Control Regulations, Subpart 16, Industrial Solid 

Waste Landfills. DOE is applying for modification of this permit to incorporate disposal of the 

salt waste discussed in this draft 3116 Determination. DOE will not dispose of the solidified salt 

waste covered by this draft 3116 Determination until SCDHEC issues the appropriate permit 

modification. 
February 28, 2005 

Page 77 of 94 

9.0 CONCLUSION

As shown in the preceding sections of this draft 3116 Determination, the solidified low-activity 

salt waste is not high-level waste based on the considerations set forth in §3116 (a) of the NDAA 

and may be disposed of as low-level waste at SRS in accordance with Section 3116 of the 

NDAA. This draft 3116 Determination will be finalized after DOE has completed consultation 

with the NRC, and, although not so required by Section 3116, after public review and comment. 
February 28, 2005 

Page 78 of 94 

10.0 REFERENCES

 

1 “Ronald W. Reagan National Defense Authorization Act for FY 2005,” Section 3116, 2004. 

2 Ledbetter, L. S., CBU-PIT-2004-00024, “12/01/04 – December Monthly WCS Curie and 

Volume Inventory Report,” Revision 0, December 9, 2004. 

3 Rios-Armstrong, M. A., CBU-PIT-2005-00031, “Decontaminated Salt Solution Volume to be 

Transferred to the Saltstone Disposal Facility from Salt Treatment and Disposition Activities,”

Revision 0, February 13, 2005. 

4 Mahoney, M. J. and D. P. Chew, CBU-PIT-2004-00004, “Savannah River Site Tank Waste 

Processing Summary,” Revision 0, December 9, 2004. 

5 Mahoney, M. J. and P. D. d’Entremont, CBU-PED-2004-00027, “Interim Salt Processing 

Strategy Planning Baseline,” Revision 0, August 27, 2004. 

6 “Salt Processing Alternatives Final Supplemental Environmental Impact Statement,”

DOE/EIS-0082-S2, June 2001. 

7 Letter, Roberson, J. H. to G. P. Rudy, “Savannah River Salt Processing Strategy,” October 9, 

2001. 

8 Romanowski, L. B., CBU-SPT-2003-00224, “Tank 41 Low Curie Salt Process History and 

Data Analysis”, Revision 0, December 18, 2003. 

9 Fowler, R. C., CBU-PED-2004-00013, “Estimate of Total Tetraphenylborate in Tank 48,”

April 6, 2004. 

10 Westinghouse Savannah River Company, WSRC-IM-2004-00008, “DSA Support Document – 

Site Characteristics and Program Descriptions,” Revision 0, October 2004. 

11 Cook, J. R., DPST-83-607, “Estimation of High Water Table Levels at the Saltstone Disposal 

Site (Z-Area),” Revision 0, June 27, 1983. 

12 Chen, K. F., WSRC-TR-99-00369, “Flood Hazard Recurrence Frequencies for C-, F-, E-, S-, 

H-, Y-, and Z-Areas,” September 30, 1999. 

13 URS/Blume, DPE-3699, “Update of Seismic Design Criteria for the Savannah River Plant, I 

Geotechnical,” 1982. 

14 McHood, M., K-CLC-Z-00001, “Liquefaction Potential for Saltstone Disposal Facility Vault 

No. 4,” Revision 0, August 22, 2002. 
February 28, 2005 

Page 79 of 94 

 

15 National Academy of Sciences, “Alternatives for High-Level Waste Salt Processing at the 

Savannah River Site,” 2000. 

16 “Code of Federal Regulation 10 CFR 61, Licensing Requirements for Land Disposal of 

Radioactive Waste,” 1-1-04 Edition. 

17 P-SPC-J-00001, “Feed Strategy and Product and Secondary Waste Specification,” Revision 

0, December 22, 2004. 

18 Petersen, C. A., WHC-SD-WM-TI-699, “Technical Basis for Classification of Low-Activity 

Waste Fraction from Hanford Site Tanks,” Revision 2, September 1996 [Sections 4.2.4, 4.2.5, 

4.2.7, and 4.2.9]. 

19 WSRC-RP-2002-00245, “Environmental Management Program – Performance Management 

Plan,” Revision 6, August 7, 2002. 

20 Conway, J. T. to K. E. McSlarrow, Defense Nuclear Facility Safety Board memorandum dated 

June 18, 2004, “Safety Impacts of Suspending Salt Disposition at the Savannah River Site.”

21 Hopkins, M. D. and S. C. Shah, CBU-SPT-2004-00038, “Curie Calculation for Salt 

Strategies,” Revision 0, February 22, 2004. 

22 Campbell, S. G., “Preliminary Material Balance for the Modular CSSX Unit,” CBU-SPT2004-00059, Revision 1, June 22, 2004. 

23 d’Entremont, P. D., CBU-PIT-2005-00013, “Radionuclide Concentrations in Saltstone,”

Revision 2, February 24, 2005. 

24 “NRC Branch Technical Position (BTP) on Concentration Averaging and Encapsulation,” 

January 17, 1995. 

25 Cook, J. R., Kocher, D. C., McDowell-Boyer, L., and E. L. Wilhite, WSRC-TR-2002-00456, 

“Special Analysis: Reevaluation of the Inadvertent Intruder, Groundwater, Air, and Radon 

Analyses for the Saltstone Disposal Facility,” Revision 0, October 23, 2002. 

26 U.S. NRC, “Decommissioning Criteria for the West Valley Demonstration Project (M-32) at 

the West Valley Site; Final Policy Statement,” February 2002. 

27 SSF-ENG-2001, “Saltstone Facility (SSF) Waste Acceptance Program Requirements (U),”

Revision 0, September 15, 2003. 

28 Westinghouse Savannah River Company, WSRC-RP-92-1360, “Radiological Performance 

Assessment for the Z-Area Saltstone Disposal Facility (U),” Revision 0, December 18, 1992. 
February 28, 2005 

Page 80 of 94 

 

29 National Council on Radiation Protection and Measurements, NCRP Report No. 123, 

“Screening Models for Releases of Radionuclides to Atmosphere, Surface Water, and Ground, 

Volumes 1 and 2,” 1996. 

 

30 Cook, J. R., WSRC-RP-2005-01402, “Estimated All Pathways and Inadvertent Intruder Doses 

from Saltstone Disposal Using Updated Salt Waste Compositions,” Revision 0, February 24, 

2005. 

31 U. S. Department of Energy, SRS Long Range Comprehensive Plan,” December 2000. 

32 Kaplan, D. I., WSRC-RP-2003-00362, “Estimated Duration of the Subsurface Reducing 

Environment Produced by the Z-Area Saltstone Disposal Facility,” Revision 2, April 25, 2003. 

33 WSRC-TR-99-00299, “Savannah River Site Environmental Report for 1999.”

34 WSRC-TR-2000-00328, “Savannah River Site Environmental Report for 2000.” 

35 WSRC-TR-2001-00474, “Savannah River Site Environmental Report for 2001.” 

36 WSRC-TR-2003-00026, “Savannah River Site Environmental Report for 2002.” 

37 WSRC-TR-2004-00015, “Savannah River Site Environmental Report for 2003.” 

38 WSRC Radiological Control Manual 5Q, Chapter 2, Part 1. 

39 200-Z-0001, “Saltstone Outside Facilities Roundsheets,” Revision 3, September 12, 2003. 

40 WSRC-IM-95-1, SRS Engineering Standards Manual, Standard 01064, Radiological Design 

Requirements, Revision 2, January 25, 2005. 

41 DOE 440.1A, “Worker Protection Management for DOE Federal and Contractor 

Employees.”

42 WSRC Radiological Control Manual 5Q, Chapter 2, Article 232. 

43 Westinghouse Savannah River Company, WSRC-SA-2003-00001, “Saltstone Facility 

Documented Safety Analysis,” Revision 2, November 2004. 

44 Westinghouse Savannah River Company, WSRC-TR-2001-00574, “Saltstone Facility 

Consolidated Hazards Analysis (U),” Revision 4, September 2004. 

45 Westinghouse Savannah River Company, WSRC-S-TSR-Z-00002, “Saltstone Facility 

Technical Safety Requirements,” Revision 1, August 2003. 
February 28, 2005 

Page 81 of 94 

 

46 White, T. L., WSRC-TR-2003-00348, “Saltstone Low Curie Salt (LCS) Project Radiological 

Design Summary Report,” Revision 0, September 2003. 

47 U.S. Department of Energy, Code of Federal Regulations 10 CFR 835, Occupational 

Radiation Protection. 

48 WSRC Radiological Control Manual 5Q, Chapter 5, Article 511. 

49 ESH-RPS-2004-00020, “Westinghouse Savannah River Company Radiological Performance 

Indicators (U),” February 2004. 

50 DOE/EH-0688, “DOE Occupational Radiation Exposure,” 2003.

51 Sutherland, J. M., CBU-PIT-2005-00047, “Closure Business Unit Pollution Prevention Plan – 

Saltstone (SS) Z-01 Stormwater Outfall,” October 14, 2004. 
February 28, 2005 

Page 82 of 94 

APPENDICES
February 28, 2005 

Page 83 of 94 

APPENDIX 1: SYSTEM DESCRIPTION44

HISTORY

Most of the tank waste inventory currently stored at SRS is a complex mixture of 

chemical and radioactive waste generated during the acid-side separation of special 

nuclear materials and enriched uranium from irradiated targets and spent fuel using the 

Plutonium Recovery and Extraction (PUREX) process in F-Canyon and the modified 

PUREX process in H-Canyon (HM process). Waste generated from the recovery of Pu238 in H-Canyon for the production of heat sources for space missions is also included. 

The variability in both nuclide and chemical content is due to the fact that waste streams 

from the 1st cycle (high heat) and 2nd cycle (low heat) extractions from each canyon were 

stored in separate tanks to better manage waste heat generation. When these streams 

were neutralized with caustic, the resulting precipitate settled into four characteristic 

sludges presently found in the tanks where they were originally deposited. The soluble 

portions of the 1st and 2nd cycle waste were similarly partitioned but have and continue to 

undergo blending in the course of waste transfer and staging of salt waste for evaporative 

concentration to supernate and saltcake. Historically, fresh waste receipts have been 

segregated into four general categories in the SRS Tank Farms: PUREX high activity 

waste (HAW), PUREX low activity waste (LAW), HM HAW, and HM LAW. Because 

of this segregation, settled sludge solids contained in tanks that received fresh waste are 

readily identified as one of these four categories. Fission product concentrations are 

about three orders of magnitude higher in both PUREX and HM HAW sludges than the 

corresponding LAW sludges. Because of differences in the PUREX and HM processes, 

the chemical compositions of principal sludge components (iron, aluminum, uranium, 

manganese, nickel, and mercury) also vary over a broad range between these sludges. 

Combining and blending salt solutions has tended to reduce soluble waste into blended 

PUREX salt and concentrate and HM salt and concentrate, rather than maintaining four 

distinct salt compositions. Continued blending and evaporation of the salt solution 

deposits crystallized salts with overlying and interstitial concentrated salt solution in salt 

tanks located in both Tank Farms. More recently, with transfers of sludge slurries to 

sludge washing tanks, removal of saltcakes for tank closure, receipts of DWPF recycle, 

and space limitations restricting full evaporator operations, salt solutions have been 

transferred between the two Tank Farms. Intermingling of PUREX and HM salt waste 

will continue until processing in the SWPF can begin.

 

44 As explained in footnote 15, this draft 3116 Determination concerns the disposal as low-level waste of 

the solidified low-activity salt waste streams in the SDF. This appendix provides additional background 

information to enhance the reader’s understanding and includes discussions of other wastes and processes 

which are not the subject of this draft 3116 Determination and are not covered by §3116 of the NDAA. 

Nothing in this appendix should be construed to expand the scope of this draft 3116 Determination or the 

scope of §3116 of the NDAA. 
February 28, 2005 

Page 84 of 94 

BACKGROUND

Approximately 36.4 Mgal [1] of radioactive waste are currently stored in underground 

waste storage tanks as waste by-products generated from activities conducted at SRS 

since the site became operational in 1951. Continued long-term storage of these 

radioactive wastes poses an environmental risk. Therefore, since 1996, DOE and its 

contractor, WSRC, have been removing waste from tanks, pre-treating it, vitrifying it, 

and pouring the vitrified waste into canisters for long-term disposal in a Federal 

repository. As of December 2004, 1,774 canisters of waste have been vitrified for 

shipment to a Federal repository when the repository is licensed and operational. All 

canisters vitrified to date contain sludge-only waste. 

The Liquid Waste and Waste Solidification System is the integrated series of facilities at 

SRS that safely manage the existing waste inventory and safely disposition waste stored 

in the tanks into a final glass or grout form. This system includes facilities for storage, 

evaporation, waste removal, pre-treatment, vitrification, and disposal. These facilities are 

briefly described in the text that follows. 

TANK STORAGE

Since it became operational in 1951, SRS, a 300-square-mile DOE complex located in 

the State of South Carolina, has produced nuclear material for national defense, research, 

medical, and space programs. The separation of fissionable nuclear material from 

irradiated targets and fuels resulted in the generation of large quantities of radioactive 

waste which are currently stored onsite 

in large underground waste storage 

tanks. The waste was converted to an 

alkaline solution, metal oxides settled 

as sludge, and supernate evaporated to 

form saltcake. 

SRS has a total of 51 underground 

waste storage tanks, all of which were 

placed into operation between 1954 

and 1986. There are four types of 

waste tanks – Types I through IV. 

Type III tanks are the newest tanks and 

were placed into operation between 

1969 and 1986. There are a total of 27 

Type III tanks. These tanks meet 

current Environmental Protection 

Agency (EPA) requirements for full 

secondary containment and leak 

detection. The remaining 24 tanks do 

not have full secondary containment and do not meet EPA requirements for secondary 

Tanks under construction. Note tank size relative 

to construction workers. Later, dirt is backfilled 

around the tanks to provide shielding. 
February 28, 2005 

Page 85 of 94 

containment. Type I tanks are the oldest tanks and were constructed between 1952 and 

1953. Type II waste tanks were constructed between 1955 and 1956. There are eight 

Type IV tanks that were constructed between 1958 and 1962. Two of these Type IV 

tanks, Tanks 17 and 20 in F Tank Farm, have been isolated, operationally closed, and 

grouted. Twelve tanks without secondary containment have a history of leakage [2]. 

Sufficient waste has been removed from these tanks such that there are currently no 

active leak sites. The first tank without secondary containment began receiving waste in 

1954. This tank is still in service. 

Approximately 36.4 Mgal of radioactive waste, containing 426 MCi [1] of radioactivity, 

are currently stored in 49 active waste storage tanks located in two separate locations, H 

Tank Farm (29 tanks) and F Tank Farm (20 tanks). This waste is a complex mixture of 

insoluble metal hydroxide solids, commonly 

referred to as sludge, and soluble salt 

supernate. The supernate volume is reduced 

by evaporation, which also concentrates the 

soluble salts to their solubility limit. The 

resultant solution crystallizes as salts. The 

resulting crystalline solids are commonly 

referred to as saltcake. The saltcake and 

supernate combined are referred to as salt 

waste (combined 33.8 Mgal). 

The sludge component of the radioactive 

waste represents approximately 2.6 Mgal (7% 

of total) of waste but contains approximately 

203 MCi (48% of total). The salt waste 

makes up the remaining 33.8 Mgal (93% of 

total) of waste and contains approximately 

223 MCi (52% of total). Of that salt waste, 

the supernate accounts for 17.4 Mgal and 211 

MCi of the 223 MCi total salt related curies. 

The saltcake accounts for the remaining 16.4 

Mgal and 12 MCi [1]. The sludge contains 

the majority of the long-lived (half-life > 30 

years) radionuclides (i.e., actinides) and 

strontium. The sludge is currently being 

stabilized in DWPF through a vitrification 

process that immobilizes the waste in a borosilicate glass matrix. 

Salt waste is dissolved in the liquid portion of the 

waste. It can be in normal solution as Supernate (top 

picture) or, after evaporation, as salt cake (bottom 

picture) or concentrated supernate. The pipes in all 

the pictures are cooling coils. 
February 28, 2005 

Page 86 of 94 

Radioactive waste volumes and radioactivity inventories reported herein are based on the 

WCS [3] database (see Appendix 2), which includes the chemical and radionuclide 

inventories on a tank-by-tank basis. WCS is a dynamic database frequently updated with 

new data from ongoing operations such as decanting and concentrating of free supernate 

via evaporators, preparation of sludge batches for DWPF feed, waste transfers between 

tanks, waste sample analyses, and influent receipts such as F and H-Canyon waste and 

DWPF recycle. Volumes and curies referenced in this evaluation are current as of 

December 1, 2004 [1]. 

Approximately 95% [1] of the salt waste 

radioactivity is short-lived (half-life 30-years or 

less) Cs-137 and its daughter product, Ba-137m, 

along with lower levels of actinide 

contamination. Depending on the particular 

waste stream (e.g., canyon waste, DWPF 

recycle waste), the cesium concentration may 

vary. The precipitation of salts following 

evaporation can also change the cesium 

concentration. The concentration of cesium is 

significantly lower than non-radioactive salts in 

the waste, such as sodium nitrate and nitrite; 

therefore, the cesium does not reach its 

solubility limit and only a small fraction 

precipitates [4]. As a result, the cesium concentration in the saltcake is much lower than 

that in the liquid supernate and interstitial liquid fraction of the salt waste. 

Waste Tank Composite Inventory (As of 12/01/04) [1] 

Saltcake

Sludge

Volume

36.4 Million

Gallons (Mgal)

Curies

223 MCi

(52%)

203 MCi

(48%)

426 Million

Curies (MCi)

211 MCi

(49%)

12 MCi

(3%)

Salt Supernate

33.8 Mgal

(93%)

2.6 Mgal

(7%)

17.4 Mgal

(48%)

16.4 Mgal

(45%)

Sludge consists of insoluble solids that settle to the 

bottom of a tank. Note the offgas bubbles, 

including hydrogen generated from radiolysis. 
February 28, 2005 

Page 87 of 94 

WASTE TANK SPACE MANAGEMENT

To make better use of available tank storage capacity, incoming liquid waste is 

evaporated to reduce its volume. This is critical because most of the SRS Type III waste 

storage tanks are already at or near full capacity. Since 1951, the Tank Farms have 

received over 140 Mgal of liquid waste, of which over 100 Mgal have been evaporated, 

leaving approximately 36.4 Mgal in the storage tanks. Projected available tank space is 

carefully tracked to ensure that the Tank Farms do not become “water logged”, a term 

meaning that so much of the usable Type III compliant tank space has been filled that 

normal operations and waste removal and processing operations cannot continue. A 

portion of tank space must be reserved as contingency space should a new tank leak be 

realized. Waste receipts and transfers are normal Tank Farm activities as the Tank Farms 

receive new or “fresh” waste from the H Separations Canyon, stabilization programs 

from both the F- and H-Canyon facilities, liquid waste from DWPF processing (typically 

referred to as “DWPF Recycle”), and wash water from sludge washing. The Tank Farms 

also make routine transfers to and from waste tanks and evaporators. Currently, there is 

very little “fresh” waste that has not had the water evaporated from it to its maximum 

extent. The working capacity of the Tank Farms has steadily decreased and this trend 

will continue until salt processing becomes operational or the system becomes water 

logged. Three evaporator systems are currently operating at SRS - the 2H, 3H, and 2F 

systems. 

WASTE REMOVAL FROM TANKS

During waste removal, water that has been chemically treated to prevent corrosion of the 

carbon steel waste tanks, inhibited water (IW), is added to the waste tanks and agitated by 

slurry pumps. If 

the tank contains 

salt, this IW, and 

agitation if 

required, dilutes 

the concentrated 

salt or re-dissolves 

the saltcake. If the 

tank contains 

sludge, this water, 

and agitation, 

suspends the 

insoluble sludge 

particles. In either 

case, the resulting 

liquid slurry, which 

now contains the dissolved salt or suspended sludge, can be pumped out of the tanks and 

transferred to waste treatment tanks. 

Typical Waste Removal equipment includes three 

to four 45-foot long slurry pumps and one transfer 

pump or jet. Note the substantial structural steel 

required to support the loads in the picture above. 

At right is the typical installation of a transfer 

pump (Tank 8) requiring difficult, high-risk entries 

into High Level Waste Tanks. 
February 28, 2005 

Page 88 of 94 

Waste removal is a multi-year process. First, each waste tank must be retrofitted with 

slurry and transfer pumps, infrastructure to support the pumps, and various service 

upgrades (power, water, air, and/or steam). These retrofits can take between two and four 

years to complete. Then, the pumps are operated to slurry the waste. Initially, the pumps 

operate near the top of the liquid and are lowered sequentially to the proper depths as 

waste is slurried and transferred out of the tanks. Waste removal activities remove the 

bulk of the waste to prepare the tank for closure. 

SAFE DISPOSAL OF THE WASTE

The goal is to convert all of the waste into one of two final waste forms: Glass, which 

will contain 99% of the radioactivity; and saltstone grout, which will contain most of the 

volume. Each of the waste types at SRS needs to be treated to accomplish disposal in 

these two waste forms. The sludge must be washed to remove non-radioactive salts that 

would interfere with glass production. The washed sludge can then be sent to DWPF for 

vitrification. The salt must be treated to separate the bulk of the radionuclides from the 

non-radioactive salts in the waste. Starting in approximately 2009, this separation will be 

accomplished in SWPF. However, until the startup of SWPF, Interim Salt Processing 

will be used to accomplish this activity. 

SALT PROCESSING

A final DOE technology selection for salt solution processing was completed and a Salt 

Processing Environmental Impact Statement (EIS) ROD was issued in October 2001. 

The ROD designated CSSX as the preferred alternative to be used to separate cesium 

from the salt waste. In parallel, SRS evaluated the implementation of other salt 

processing alternatives for specific waste portions that would not need to be processed in 

a CSSX facility (SWPF) and could be processed before SWPF was operational. 

The Salt Processing Planning Baseline calls for using four different processes to treat 

salt: 

• Deliquification, Dissolution, and Adjustment - for salt in selected tanks 

that are relatively low in radioactive content (e.g., Tank 41), the treatment 

of deliquification (i.e., extracting the interstitial liquid) is sufficient to 

produce a salt that meets the Saltstone Facility WAC. Deliquification is an 

effective decontamination process because the primary radionuclide in salt 

is Cs-137, which is highly soluble. By exploiting the solubility of Cs-137 

and isolating the insoluble fraction, a low-level waste salt stream can be 

produced. 

• Actinide Removal Process - for salt in selected tanks (e.g., Tank 28), even 

though extraction of the interstitial liquid reduces Cs-137 and soluble 

actinide concentrations, the Cs-137 or actinide concentrations of the 

resulting salt are too high to meet the Saltstone Facility WAC. Salt from 
February 28, 2005 

Page 89 of 94 

these tanks first will be sent to ARP. In ARP, MST is added to the waste 

as a finely divided solid. Actinides are sorbed on the MST and then 

filtered out of the liquid to produce a low-level waste stream that is sent to 

MCU. If the soluble actinides in the original salt solution are sufficiently 

low, then the stream will not require the MST strike and will only be 

filtered prior to being sent to the MCU. After SWPF startup, ARP will 

send clarified salt solution to SWPF for cesium removal. 

• Modular CSSX Unit - for tanks with salt that is too high in activity for 

deliquification to sufficiently reduce Cs-137 concentrations, the salt in 

these tanks must be further treated to reduce the concentration of Cs-137 

using the CSSX process. After approximately 2009, this will be done in a 

new facility, SWPF. However, so that some of these wastes can be treated 

before 2009, DOE will build a small-scale modular CSSX unit. Salt to be 

processed will first be processed through ARP and then through the 

modular unit. This unit will allow processing of salt waste with higher Cs137 concentrations at a relatively low rate. 

• Salt Waste Processing Facility - this is the full-scale CSSX process. The 

facility incorporates both the ARP and CSSX process in a full-scale 

shielded facility capable of handling salt with high levels of radioactivity. 

After startup of SWPF in approximately 2009, all remaining salt waste will 

be processed through this facility. 

SLUDGE PROCESSING

Sludge is “washed” to reduce the amount of non-radioactive soluble salts remaining in 

the sludge slurry. The processed sludge is called “washed sludge.” During sludge 

processing, large volumes of wash water are generated and must be volume-reduced by 

evaporation. Over the life of the waste removal program, the sludge currently stored in 

tanks at SRS will be blended into separate sludge “batches” to be processed and fed to 

DWPF for vitrification. 
February 28, 2005 

Page 90 of 94 

DWPF VITRIFICATION

Final processing for the washed sludge and salt 

waste occurs at DWPF. This waste includes 

MST/sludge from ARP or 

SWPF, the cesium strip 

effluent from MCU or SWPF, 

and the washed sludge slurry 

from Extended Sludge 

Processing. In a complex 

sequence of carefully 

controlled chemical reactions, 

this waste is blended with glass 

frit and melted to vitrify it into a borosilicate glass 

form. The resulting molten glass is poured into 

stainless steel canisters. As the filled canisters cool, 

the molten glass solidifies, immobilizing the radioactive waste within the glass structure. 

After the canisters have cooled, they are permanently sealed, and the external surfaces are 

decontaminated to meet United States Department of Transportation requirements. The 

canisters are then ready to be stored on an interim basis on-site in the Glass Waste 

Storage Building (GWSB), pending shipment to a Federal Repository for permanent 

disposal. A low-level recycle waste stream from DWPF is returned to the Tank Farms. 

DWPF has been fully operational since 1996. 

SALTSTONE: ON-SITE DISPOSAL OF LOW-LEVEL WASTE

The Saltstone Facility, located in Z-Area, consists of two facility segments: SPF and 

SDF. SPF is permitted as a 

wastewater treatment facility 

per SCDHEC Regulations 

R.61-67. SPF receives and 

treats the salt solution to 

produce grout by mixing the 

LLW liquid stream with 

cementitious materials 

(cement, flyash, and slag). A 

slurry of the components is 

pumped into the disposal 

vaults, located in SDF, where the saltstone grout solidifies into a monolithic, nonhazardous, solid LLW form. SDF is permitted as an ISWLF site, as defined by SCDHEC 

Regulations R61-66 and R.61-107.16. 

View of the Saltstone Facility

Canisters being received (prior to being 

filled with radioactive glass) 

Sample of Vitrified 

Radioactive Glass 
February 28, 2005 

Page 91 of 94 

The facility will contain many large concrete vaults divided into cells. Each of the cells 

will be filled with solid saltstone grout. The grout itself provides primary containment of 

the waste, and the walls, floor, and roof of the vaults provide secondary containment. 

Approximately 15 feet of overburden were removed 

to prepare and level the site for vault construction. 

All vaults will be built at or slightly below the grade 

level that exists after the overburden and leveling 

operations are complete. The bottom of the saltstone 

grout monoliths will be at least 5 feet above the 

historic high water table beneath the Z-Area site, thus, 

avoiding disposal of waste in a zone of water table 

fluctuation. Run-on and runoff controls are installed 

to minimize site erosion during the operational 

period. 

The current vault (Vault #4) has the dimensions of approximately 200 feet wide, by 600 

feet in length, by 26 feet in height. The vault is divided into 12 cells, with each cell 

measuring approximately 100 feet by 100 feet. The vault is covered with a sloped, 

permanent roof that has a minimum thickness of 4 inches, and a minimum slope of 0.24 

inches/foot. The vault walls are approximately 1.5 feet thick, with the base mat having a 

thickness of 2 feet. Operationally, the cells of the vault will be filled to a height of 

approximately 25 feet with saltstone, and then a layer of uncontaminated grout, with an 

average thickness of 2 feet, will be poured to fill in the space between the saltstone grout 

and the sloped roof. The other current vault (Vault #1) has the dimensions of 

approximately 100 feet wide, by 600 feet in length, by 25 feet in height. The vault is 

divided into 6 cells, with each cell measuring approximately 100 feet by 100 feet. 

Closure operations will begin near the end of the active disposal period in the SDF, i.e., 

after most or all of the vaults have been constructed and filled. Backfill of native soil will 

be placed around the vaults. The present closure concept includes two moisture barriers 

consisting of clay/gravel drainage systems along with backfill layers and a shallowrooted bamboo vegetative cover. 

Construction of the Saltstone Facility and the first two vaults was completed between 

February 1986 and July 1998. The Saltstone Facility started radioactive operations June 

12, 1990. Future vaults will be constructed on a “just-in-time” basis in coordination with 

salt processing production rates. 

View of a Saltstone Vault
February 28, 2005 

Page 92 of 94 

 

REFERENCES

1 Ledbetter, L. S., CBU-PIT-2004-00024, “12/01/04 – December Monthly WCS Curie 

and Volume Inventory Report,” Revision 0, December 9, 2004. 

2 Waltz, R. S., CBU-LTS-2004-00104, “SRS High Level Waste Tank Leaksite 

Information (8/04),” August 2004. 

3 Hester, J. R., WSRC-TR-96-0264, “High Level Waste Characterization System 

(WCS),” Revision 0, December 1996. 

4 Brooke, J. N., Peters, J. F., and K. Stahell, WSRC-TR-99-00358, “Hydrological 

Methods Can Separate Cesium from Nuclear Waste Saltcake,” 1999. 
February 25, 2005 0800 hrs 

Page 93 of 94 

APPENDIX 2: WASTE CHARACTERIZATION SYSTEM

WCS is an electronic information system that tracks waste tank data, including projected 

radionuclide and chemical inventories, based on sample analyses, process histories, 

composition studies, and theoretical relationships. The system, initially developed in 

1995, tracks the concentrations of 96 chemicals and 41 radioactive nuclides in each of the 

SRS waste tanks [1]. The 41 nuclides were primarily selected based on their impact to 

waste tank source term, inhalation dose potential, or on the E-Area vault Waste 

Acceptance Criteria. WCS consists of two functional areas: a sample data repository and 

waste characterization. The waste characterization contains inputs and calculations used 

to represent real-time conditions in the waste tanks. These real-time conditions are used 

to determine compliance (as input to Safety Basis calculations) with several nuclear 

safety administrative control programs such as the waste tank Flammability Control 

Program and the waste tank Corrosion Control Program. 

WCS is routinely updated to reflect changes in waste inventory resulting from the receipt 

of new waste from the SRS Canyon facilities, the Effluent Treatment Facility, or recycle 

waste from DWPF; the transfer of waste for processing at DWPF or SDF, or from the 

receipt of new characterization information in the form of sample analysis results, field 

measurements, waste tank video inspections, or technical reports. The waste tank 

contents in WCS are also adjusted to reflect the impacts of evaporator operation and to 

account for waste tank to waste tank transfers. 

Prior to new data being entered into WCS, a Data Integrity Review Team evaluates the 

new data to determine its technical validity. The core team members are engineers or 

scientists with expertise in relevant fields such as flammability, corrosion, organic 

chemistry, or criticality [2]. The Data Integrity Review Team routinely meets to perform 

evaluations of new characterization information. As part of the review the new data is 

compared to previous measurements of the same parameter in the same waste tank, to 

previous measurements of the same parameter from other waste tanks, to the expected 

value for that parameter based on the waste tank’s processing history, and to the relative 

concentrations of other parameters in the same data set to determine if the new data can 

be accepted as a valid representation of actual conditions. Once the Data Integrity 

Review Team has deemed the data to be valid, it is approved for entry into the waste 

characterization portion of WCS and into the data repository. 

WCS is a controlled database that complies with strict Quality Assurance requirements. 

These requirements apply to not only data entry but also to configuration control of the 

database software itself. All data entries into WCS must be independently verified. A 

Software Quality Assurance Plan governs the methods that must be utilized to maintain 

the database baseline configuration and to ensure the proper handling of the database [3]. 

To address backup and recovery, including disaster recovery, WCS is hosted on a server 

that provides daily backup of all WCS files. 
February 25, 2005 0800 hrs 

Page 94 of 94 

 

REFERENCES

1 Ledbetter, L. A., CBU-PIT-2004-00022, “Waste Characterization System Radionuclide 

and Chemical Inventory Selection Technical Basis,” December 2004. 

2 Ledbetter, L. A., WSRC-TR-2003-00048, “Waste Characterization System Program 

Description Document,” Revision 2, June 2004. 

3 Manz, R. W., B-SQP-H-00041, “Closure Business Unit LWDP Waste Characterization 

System Software Quality Assurance Plan,” September 2004.