December 28, 2005

Mr. Charles E. Anderson

Principal Deputy Assistant Secretary

Office of Environmental Management

U.S. Department of Energy

1000 Independence Avenue, S.W.

Washington, DC 20585

SUBJECT: TECHNICAL EVALUATION REPORT FOR DRAFT WASTE DETERMINATION

FOR SALT WASTE DISPOSAL

Dear Mr. Anderson:

Under Section 3116 of the Ronald W. Reagan National Defense Authorization Act for Fiscal

Year 2005 (NDAA), the Secretary of Energy, in consultation with the U.S. Nuclear Regulatory

Commission (NRC), may determine that certain radioactive waste resulting from reprocessing

of spent nuclear fuel is not high-level waste. By letter dated March 31, 2005, the U.S.

Department of Energy (DOE) submitted the “Draft Section 3116 Determination, Salt Waste

Disposal, Savannah River Site” to the NRC for review. The NRC staff has performed a

technical review to determine whether there is reasonable assurance that applicable criteria of

the NDAA can be met by the waste management approach proposed in DOE’s draft Section

3116 waste determination. The NRC’s assumptions, analysis, and conclusions are presented

in the attached Technical Evaluation Report (TER). 

Based on the information provided by DOE to the NRC in letters dated March 31, June 30, July

15, September 15, and September 30, 2005, the NRC staff has concluded that there is

reasonable assurance that the applicable criteria of the NDAA can be met provided certain 

assumptions made in DOE's analyses are verified via monitoring. The assumptions described

in Section 4.3.1 of the TER are important to demonstrating that the performance objectives in

10 CFR 61, Subpart C, can be met and fall into the following general categories: wasteform and

vault degradation, the effectiveness of infiltration and erosion controls, and estimation of the

radiological inventory. The NDAA requires NRC, in coordination with the State of South

Carolina, to monitor disposal actions taken by DOE for the purpose of assessing compliance

with the performance objectives of 10 CFR 61, Subpart C. Consequently, NRC requests that

DOE develop proposed approaches that DOE will use to address the areas identified in Section

4.3.1. NRC will then, in coordination with the State, develop a program by which NRC will

monitor DOE’s implementation of the approaches.
Mr. Anderson - 2 -

It is important to note that the NRC’s conclusions presented in this TER are based on the 

information provided by DOE. If, in the future, DOE determines it is necessary to revise its

assumptions, analysis, design, or waste management approach and those changes are

important to meeting the criteria of the NDAA, DOE should consult once again with NRC

regarding the enclosed TER. The NRC looks forward to continuing to work cooperatively with

DOE in implementing the Section 3116 requirements. If you have any questions or need

additional information regarding this TER, please call me at 301-415-7437, or call Anna

Bradford, senior project manager on my staff, at 301-415-5228.

Sincerely,

/RA/

Larry W. Camper, Director

Division of Waste Management

 and Environmental Protection

Office of Nuclear Material Safety 

 and Safeguards
Mr. Anderson - 2 -

It is important to note that the NRC’s conclusions presented in this TER are based on the

information provided by DOE. If, in the future, DOE determines it is necessary to revise its

assumptions, analysis, design, or waste management approach and those changes are

important to meeting the criteria of the NDAA, DOE should consult once again with NRC

regarding the enclosed TER. The NRC looks forward to continuing to work cooperatively with

DOE in implementing the Section 3116 requirements. If you have any questions or need

additional information regarding this TER, please call me at 301-415-7437, or call Anna

Bradford, senior project manager on my staff, at 301-415-5228.

Sincerely,

/RA/

Larry W. Camper, Director

Division of Waste Management

 and Environmental Protection

Office of Nuclear Material Safety 

 and Safeguards

DISTRIBUTION: DWMEP r/f EPAD r/f

ML053010225

OFFICE DWMEP DWMEP DWMEP DWMEP

NAME A. Bradford D. Esh A. Ridge M. Thaggard

DATE 11/1/05 11/1/05 11/1/05 12/20/05

OFFICE DWMEP:SC OGC DWMEP DWMEP:DD

NAME R. Whited S. Treby S. Flanders L.W. Camper

DATE 12/20/05 11/4/05 12/22/05 12/28/05
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U. S. NUCLEAR REGULATORY COMMISSION 

TECHNICAL EVALUATION REPORT FOR 

THE U.S. DEPARTMENT OF ENERGY SAVANNAH RIVER SITE

DRAFT SECTION 3116 WASTE DETERMINATION FOR

SALT WASTE DISPOSAL

December 2005
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i

TABLE OF CONTENTS

EXECUTIVE SUMMARY . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . -v1. INTRODUCTION . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . - 1 -

1.1 Background . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . - 1 -

1.2 DOE-SRS Salt Waste Processing Strategy . . . . . . . . . . . . . . . . . . . . . . . . . - 1 -

1.3 Waste Determination Criteria . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . - 5 -

1.4 NRC Review Approach . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . - 8 -

1.5 Previous Reviews Performed for SRS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . - 9 -

2. CRITERION ONE . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . - 10 -

2.1 Salt Waste Disposal . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . - 10 -

2.2 NRC Review and Conclusions (Criterion One) . . . . . . . . . . . . . . . . . . . . . - 10 -

3. CRITERION TWO . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . - 10 -

3.1 Waste Inventory . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . - 10 -

3.2 NRC Evaluation – Waste Inventory . . . . . . . . . . . . . . . . . . . . . . . . . . . . . - 13 -

3.3 Identification of Highly Radioactive Radionuclides . . . . . . . . . . . . . . . . . . - 13 -

3.4 NRC Evaluation – Identification of Highly Radioactive Radionuclides . . . . - 14 -

3.5 Radionuclide Removal Efficiencies . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . - 15 -

3.6 NRC Evaluation – Radionuclide Removal Efficiencies . . . . . . . . . . . . . . . - 18 -

3.7 Selection of Treatment Processes . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . - 20 -

3.7.1 Alternative Treatment Technologies . . . . . . . . . . . . . . . . . . . . . . . . - 21 -

3.7.2 Alternatives to the Proposed Two-Phase Approach . . . . . . . . . . . . - 22 -

3.8 NRC Evaluation – Selection of Treatment Processes . . . . . . . . . . . . . . . . - 25 -

3.9 NRC Review and Conclusions (Criterion Two) . . . . . . . . . . . . . . . . . . . . . - 29 -

4. CRITERIA THREE (A) AND THREE (B) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . - 31 -

4.1 Assessment of Waste Classification . . . . . . . . . . . . . . . . . . . . . . . . . . . . . - 32 -

4.1.1 Waste Classification . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . - 32 -

4.1.2 NRC Evaluation - Waste Classification . . . . . . . . . . . . . . . . . . . . . - 32 -

4.2 Performance Assessment to Demonstrate Conformance with Performance

Objectives . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . - 34 -

4.2.1 Performance Assessment Overview . . . . . . . . . . . . . . . . . . . . . . . - 35 -

4.2.2 Source Term . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . - 37 -

4.2.3 NRC Evaluation – Source Term . . . . . . . . . . . . . . . . . . . . . . . . . . . - 38 -

4.2.4 Infiltration and Erosion Control . . . . . . . . . . . . . . . . . . . . . . . . . . . . - 40 -

4.2.4.1 Infiltration . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . - 40 -

4.2.4.2 Erosion Control . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . - 41 -

4.2.5 NRC Evaluation – Infiltration and Erosion Control . . . . . . . . . . . . . - 45 -

4.2.5.1 NRC Evaluation - Infiltration . . . . . . . . . . . . . . . . . . . . . . . - 45 -

4.2.5.2 NRC Evaluation - Erosion Control . . . . . . . . . . . . . . . . . . . - 47 -

4.2.6 Wasteform and Concrete Vault Degradation . . . . . . . . . . . . . . . . . - 48 -

4.2.7 NRC Evaluation – Wasteform and Concrete Vault Degradation . . - 51 -

4.2.8 Release and Near-Field Transport . . . . . . . . . . . . . . . . . . . . . . . . . - 53 -
-ii4.2.9 NRC Evaluation – Release and Near-Field Transport . . . . . . . . . . - 53 -

4.2.10 Hydrology and Far-Field Transport . . . . . . . . . . . . . . . . . . . . . . . . . - 58 -

4.2.11 NRC Evaluation – Hydrology and Far-Field Transport . . . . . . . . . . - 60 -

4.2.12 Dose Methodology . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . - 62 -

4.2.13 NRC Evaluation – Dose Methodology . . . . . . . . . . . . . . . . . . . . . . - 63 -

4.2.14 Overview of Performance Objectives . . . . . . . . . . . . . . . . . . . . . . . - 63 -

4.2.15 Protection of the Public . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . - 64 -

4.2.16 NRC Evaluation – Protection of the Public . . . . . . . . . . . . . . . . . . - 72 -

4.2.17 Protection of Intruders . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . - 80 -

4.2.18 NRC Evaluation – Protection of Intruders . . . . . . . . . . . . . . . . . . . - 82 -

4.2.19 Protection of Individuals During Operations . . . . . . . . . . . . . . . . . . - 84 -

4.2.20 NRC Evaluation – Protection of Individuals During Operations . . . - 86 -

4.2.21 Site Stability . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . - 86 -

4.2.22 NRC Evaluation – Site Stability . . . . . . . . . . . . . . . . . . . . . . . . . . . - 87 -

4.3 NRC Review and Conclusions [Criterion Three (A)] . . . . . . . . . . . . . . . . . . . . - 87 -

4.3.1 Factors Important to Assessing Compliance 

with 10 CFR 61, Subpart C . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . - 90 -

5. CONCLUSIONS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . - 92 -

APPENDIX A. NRC STAFF RECOMMENDATIONS . . . . . . . . . . . . . . . . . . . . . . . . . . . . - 93 -

CONTRIBUTORS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . - 95 -

REFERENCES . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . - 97 -

LIST OF ABBREVIATIONS AND ACRONYMS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . - 104 -
-iiiLIST OF TABLES

Table 1 Average Composition of Primary Physical Phases in DOE-SRS Tank Waste . . - 12 -

Table 2 Overview of Salt Waste Processing Efficiencies . . . . . . . . . . . . . . . . . . . . . . . . - 16 -

Table 3 Projected Removal Efficiencies of Radionuclides Considered in Detail (in %) . . - 17 -

Table 4 Summary of Risk and Cost Information for Three Waste 

 Treatment Approaches . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . - 24 -

Table 5 Comparison of DDA, ARP/MCU, and SWPF Waste with 

 Class C Concentration Limits [d’Entremont and Drumm, 2005](1) . . . . . . . . . . - 33 -

Table 6 Estimated Source Term for the Saltstone Disposal Facility . . . . . . . . . . . . . . . . - 38 -

Table 7 Summary of DOE Results Compared to Performance Objectives . . . . . . . . . . . - 65 -

Table 8 Summary of DOE Sensitivity Analyses for Vault 4 . . . . . . . . . . . . . . . . . . . . . . . - 68 -

Table 9 Description of DOE Sensitivity Analyses . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . - 70 -

Table 10 Comparison of Projected Vault 4 Inventory, Total Estimated SDF Inventory, 

 and Average Vault Inventory . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . - 74 -

Table 11 Scaled Results for the Sensitivity Scenarios . . . . . . . . . . . . . . . . . . . . . . . . . . . - 75 -

LIST OF FIGURES

Figure 1 Location of the Savannah River Site . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . - 2 -

Figure 2 DOE Salt Waste Processing Strategy . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . - 4 -

Figure 3 Location of the Z-Area . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . - 6 -

Figure 4 Photograph of Existing SDF Vault 4 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . - 7 -

Figure 5 Engineered Closure Cap Side View . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . - 41 -

Figure 6 Engineered Closure Cap Layers . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . - 42 -

Figure 7 Comparison of PORFLOW and Slag Lysimeter Data . . . . . . . . . . . . . . . . . . . . - 79 -

Figure 8 Resident Intruder Scenario . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . - 81 -
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-vEXECUTIVE SUMMARY

In March 2005, the U.S. Department of Energy (DOE) submitted the “Draft Section 3116

Determination, Salt Waste Disposal, Savannah River Site” (SRS) for review by the U.S. Nuclear

Regulatory Commission (NRC), as required by the Ronald W. Reagan National Defense

Authorization Act for Fiscal Year 2005 (NDAA). The draft waste determination addresses salt

waste that DOE proposes to remove from the high-level waste (HLW) tank farms, treat through

various processes, and dispose of on site in the Saltstone Disposal Facility (SDF). This

Technical Evaluation Report (TER) presents information on DOE’s salt waste processing

strategy, the applicable review criteria, and the NRC’s review approach, as well as the NRC’s

analysis and conclusions with respect to whether there is reasonable assurance that DOE’s

proposed approach can meet certain requirements of the NDAA for determining that waste is

not HLW. The NRC is not providing regulatory approval of DOE’s waste determination activities

and DOE is responsible for determining whether the waste streams addressed in the draft

waste determination are HLW.

Based on the information provided by DOE to the NRC in letters dated March 31, June 30, July

15, September 15, and September 30, 2005, the NRC staff has concluded that there is

reasonable assurance that the applicable criteria of the NDAA can be met provided certain 

assumptions made in DOE's analyses are verified via monitoring. These assumptions, which

are described in Section 4.3.1 of the TER, are important to demonstrating that the performance

objectives in 10 CFR 61, Subpart C, can be met and fall into the following general categories:

wasteform and vault degradation, the effectiveness of infiltration and erosion controls, and

estimation of the radiological inventory. The NDAA requires NRC, in coordination with the State

of South Carolina, to monitor disposal actions taken by DOE for the purpose of assessing

compliance with the performance objectives of 10 CFR 61, Subpart C. Consequently, NRC

requests that DOE develop proposed approaches that DOE will use to address the areas

identified in Section 4.3.1. NRC will then, in coordination with the State, develop a program by

which NRC will monitor DOE’s implementation of the approaches.

There are 51 HLW tanks at the SRS site, two of which are operationally closed. The remaining

49 tanks currently hold approximately 1.34 x 108

 L (36.4 Mgal) of waste and 15.8 x 1018 Bq (426

MCi) of radioactivity. The tank waste consists of a mixture of insoluble metal hydroxide solids,

referred to as sludge, and soluble salt supernate and crystallized salt, the combination of which

is referred to as salt waste. The salt waste comprises approximately 1.25 x 108

 L (33.8 Mgal) 

(93% of the total in the tank farm) and 8.25 x 1018 Bq (223 MCi) (52% of the total in the tank

farm). DOE estimates that approximately 1.1 x 1017 to 1.9 X 1017 Bq (3 to 5 MCi) of salt waste

would be disposed of in the SDF. The vast majority of the activity of the tank farm waste would

be sent to the onsite Defense Waste Processing Facility (DWPF) for processing and eventual

disposal in a Federal repository.

DOE intends to process the salt waste to segregate the low-activity fraction by using a twophase, three-part approach. The first phase (called Interim Salt Processing) will involve two

steps to treat the lower activity salt waste: (1) processing of a minimal amount of the lowest

activity salt waste though a process involving Deliquification, Dissolution, and Adjustment (DDA)

of the waste, and (2) processing of a minimal amount of additional waste with slightly higher

activity levels using an Actinide Removal Process (ARP) and a Modular Caustic Side Solvent

Extraction (CSSX) Unit (MCU), along with DDA. The second phase would involve the
-viseparation and processing of the majority of the salt waste using a much larger facility, the Salt

Waste Processing Facility (SWPF).

The NDAA contains three criteria for determining that waste is not HLW. The first is that the

waste does not require permanent isolation in a deep geologic repository for spent fuel or HLW. 

This criterion allows for the consideration that waste may require disposal in a deep geologic

repository even though the other criteria of the NDAA can be met. Consideration could be

given to those circumstances under which geologic disposal is warranted to protect public

health and safety and the environment; for example, unique radiological properties of the waste. 

Given that there is reasonable assurance that DOE’s proposed approach can meet the other

criteria in the NDAA, including the performance objectives of 10 CFR 61, Subpart C, and that

there appears to be no other properties of the waste that would require deep geologic disposal,

the NRC finds that there is reasonable assurance that Criterion One of the NDAA can be met.

The second criterion of the NDAA is that the waste has had highly radioactive radionuclides

removed to the maximum extent practical. To assess conformance with Criterion Two, the

NRC assessed the estimated waste inventory, the identification of highly radioactive

radionuclides, radionuclide removal efficiencies, and the selection of treatment processes and

approaches. NRC’s conclusions regarding Criterion Two are based on the following

assumptions: (1) the DDA and ARP/MCU interim treatment processes will not be used to treat

significantly larger volumes or higher activity waste than proposed in the draft waste

determination, (2) the actual radionuclide removal efficiencies of the ARP/MCU and SWPF

treatment processes will meet or exceed the reported lower bounds of the projected removal

efficiencies; (3) during the DDA process, dissolution of saltcake and solids settling will be

performed such that the amount of sludge entrained in salt waste after DDA processing does

not exceed DOE’s estimate of 200 mg/L salt waste, which was identified by DOE as being a

conservative estimate; (4) during DDA processing, the lower bound of 30% removal of

supernate is met or exceeded for deliquification; (5) financial data used to support DOE’s cost

estimates are reasonable; (6) the proposed dates of operation of the ARP/MCU and SWPF

facilities will be met; and (7) the Tank 48 waste can be processed safely in the Saltstone

Processing Facility (SPF), and the wasteform will meet the Class C limits and perform as

expected. Based on the assumptions listed above, the NRC has concluded that there is

reasonable assurance that DOE can meet Criterion Two with its proposed approach. 

The third criterion of the NDAA is that the waste is disposed of in compliance with the

performance objectives contained in NRC regulations at 10 CFR 61, Subpart C. The NDAA

provides for additional consultation with the NRC if the waste is found to exceed Class C

concentration limits, which is not the case for the salt waste evaluated in the draft waste

determination, and also provides for disposal pursuant to State-approved closure plans or

State-issued permits. These types of State closure plans and permits are outside of NRC’s

authority and are not addressed in this TER. Subpart C of 10 CFR 61 sets requirements for

protection of the public, the inadvertent intruder, and individuals during operation, and also

provides for site stability. To assess conformance with Criterion Three, the NRC evaluated: (1)

the waste classification, (2) DOE’s performance assessment, (3) the source term, (4) infiltration

and erosion control, (5) wasteform and concrete vault degradation, (6) release and near-field

transport, (7) hydrology and far-field transport, (8) dose methodology, (9) the safety

assessment and radiation protection program for individuals during operations, and (10) the

strategy to ensure stability of the disposal site after closure. 
-viiNRC’s conclusions regarding Criterion Three are based on the following assumptions: (1) more

realistic modeling of waste oxidation and release of technetium from an oxidized layer of waste

will result in predicted doses significantly lower than those projected in the DOE sensitivity

analysis for 100% oxidation of the waste; (2) the hydraulic conductivity of degraded saltstone

and vault concrete will not be larger than 1 x 10-7 cm/s (1 x 10-1 ft/yr); (3) field-scale physical

properties (e.g., hydraulic conductivity, effective diffusivity) of as-emplaced saltstone are not

significantly different from the results of laboratory tests of smaller-scale samples performed to

date; (4) cracking from any mechanism will not be significantly more extensive than currently

predicted by DOE; (5) the numerical modeling results for moisture flow through fractures in the

concrete and saltstone located in the vadose zone will be confirmed by future model support

(i.e., fracture flow will not occur); (6) the previous Documented Safety Analysis bounds the

impacts to individuals during operations that may be estimated from the proposed waste

streams; (7) active institutional controls will be maintained for 100 years; (8) current projections

of the radiological concentrations of the waste are greater than or equal to actual

concentrations for highly radioactive radionuclides; (9) future tests will confirm that the physical

properties of samples that contain organic materials similar to Tank 48 waste are consistent

with non-organic containing samples; (10) the erosion control design that DOE eventually

implements will not deviate significantly from the information submitted to the NRC in response

to the NRC’s request for additional information; (11) development and use of accurate

information for the moisture characteristic curves of concrete and saltstone will not significantly

increase currently estimated release rates; and (12) gas phase transport of oxygen will not

significantly increase oxidation of technetium in the saltstone, including transport in fractures.

Based on the assumptions listed above, the NRC has concluded that there is reasonable

assurance that DOE can meet Criterion Three, which by reference incorporates the

performance objectives of 10 CFR 61, Subpart C. DOE should assess the factors important to

assessing compliance with 10 CFR 61, Subpart C, that the NRC staff has identified in this TER

and carefully review the assumptions that the NRC conclusions are based upon. The DOE

analysis demonstrates that more realistic modeling of infiltration, water contact with the waste,

waste oxidation, and radionuclide release in an unsaturated and potentially fractured system is

needed, because conservative modeling yields results that are higher than the dose limits

provided in 10 CFR 61, Subpart C for protection of the public. Likewise, adequate model

support is essential to ensuring that public health and safety can be protected with reasonable

assurance. As discussed in detail in this TER, the deterministic DOE base case result does not

have adequate model support and is not clearly conservative. Considering the uncertainty in

many key parameters, DOE’s current deterministic base case should not be used as the basis

for developing inventory limits. A revised base case should be based on the projected average

vault inventory and orientation of multiple disposal vaults, include the expected magnitude and

timing of climate change from the natural cycling of climates, include the expected magnitude

and rate of oxidation of waste, consider liquid and gas flow in fractures that may develop, and 

account for the questionable moisture characteristic curve information for concrete and

saltstone that was used in the previous analysis. 

For a broader and more detailed discussion of DOE’s approach and NRC’s analyses and

conclusions, please see the appropriate sections of the TER, as the full discussion is not

replicated here in the Executive Summary. All of the conclusions reached by the NRC are

based on the draft Section 3116 waste determination dated March 2005, DOE’s responses to

NRC’s request for additional information, supporting references, and information provided

during meetings between DOE and NRC. If, in the future, DOE determines it is necessary to
-viiirevise its assumptions, analysis, design, or waste management approach and those changes

are important to meeting the criteria of the NDAA, DOE should consult once again with NRC

regarding the findings in this TER. It should also be noted that NRC is providing consultation to

DOE as required by the NDAA and the NRC is not providing regulatory approval in this action. 

DOE is responsible for determining whether the waste is HLW. This NRC assessment is a sitespecific evaluation and is not a precedent for any future decisions regarding non-HLW or

incidental waste determinations at SRS or at other sites.

 
- 1 -

1. INTRODUCTION

1.1 Background

The Savannah River Site (SRS) is a 780-square kilometer (300-square mile) site located in

western South Carolina and owned by the U.S. Department of Energy (DOE) (see Figure 1). 

The site began operation in 1951 and has produced nuclear material for national defense,

research, medical, and space programs. The waste from reprocessing of spent nuclear fuel for

defense purposes has been commingled with non-reprocessing waste resulting from the

production of targets for nuclear weapons and production of material for space missions. 

Significant quantities of radioactive waste are currently stored on site in large underground

waste storage tanks.

SRS has a total of 51 underground waste storage tanks, all of which were placed into operation

between 1954 and 1986. Twenty seven of the tanks meet current U.S. 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. 

Twelve tanks without secondary containment have a history of leakage, but sufficient waste has

been removed from these tanks such that there are currently no active leak sites. Two of the

tanks, which did not have secondary containment, have been closed and grouted. Of the 49

remaining operational tanks, 29 are in the H Tank Farm and 20 are in the F Tank Farm.

Approximately 1.34 x 108

 L (36.4 Mgal) of waste, containing 1.58 x 1019 Bq (426 MCi) of

radioactivity, is stored in the tanks at SRS. The waste is a mixture of insoluble metal hydroxide

solids, referred to as sludge, and soluble salt supernate. The supernate volume has been

reduced by evaporation, which also concentrates the soluble salts to their solubility limits. The

resultant solution crystallizes as salts, and the resulting solid is referred to as saltcake. The

saltcake and supernate combined are referred to as salt waste, and comprise 1.25 x 108

 L (33.8

Mgal) (93% of the total) and contains 8.25 x 1018 Bq (223 MCi) (52% of the total). The sludge

comprises 9.6 x 106

 L (2.6 Mgal) of waste (7% of the total) and contains 7.51 x 1018 Bq (203

MCi) (48% of the total). 

DOE intends to remove, stabilize, and dispose of the waste, and close all 49 of the operational

tanks. The sludge is currently being stabilized in the Defense Waste Processing Facility

(DWPF) through a vitrification process that immobilizes the waste in a borosilicate glass matrix

for eventual disposal in a Federal repository. In order to continue to have adequate tank farm

space to support DWPF operations and start up of the Salt Waste Processing Facility (SWPF), 

DOE has indicated it needs to remove a portion of the salt waste in the near term. This salt

waste will be treated and disposed of on site. DOE’s approach for treating and disposing of the

salt waste is evaluated by the U.S. Nuclear Regulatory Commission (NRC) staff in this

Technical Evaluation Report (TER).

1.2 DOE-SRS Salt Waste Processing Strategy

The 49 operational tanks at SRS are constructed of carbon steel and consist of four different

types of designs. Type III, the newest design, has a full-height secondary tank and forced

water cooling. There are 27 Type III tanks. Type I and II tanks have five-foot-high secondary 
- 2 -

Figure 1 Location of the Savannah River Site
- 3 -

pans and forced cooling. Type IV tanks have a single steel wall and do not have forced cooling. 

Only Type III tanks meet EPA requirements for full secondary containment and leak detection. 

The waste in the 49 tanks consists of sludge, supernate, and saltcake. DOE intends to remove

the salt waste and dispose of the low activity fraction on site in the Saltstone Disposal Facility

(SDF).

DOE intends to separate the salt waste to segregate the low-activity fraction by using a twophase, three-part approach (see Figure 2). The first phase (called Interim Salt Processing) will

involve two parts to treat the lower activity salt waste: (1) processing of a minimal amount of the

lowest activity salt waste through a process involving Deliquification, Dissolution, and

Adjustment (DDA) of the waste, and (2) processing of a minimal amount of additional waste

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

Caustic Side Solvent Extraction (CSSX) Unit (MCU), along with DDA. The second phase would

involve the separation and processing of the majority of the salt waste using the SWPF. 

The DDA process will be the first interim process used and involves the following steps: (1)

removing the supernate from above the saltcake; (2) extracting interstitial supernate from within

the saltcake; (3) dissolving the saltcake and transferring the resulting salt solution to a settling

tank; and (4) transferring the salt solution to the Saltstone Production Facility (SPF) for mixture

with cement, slag, and fly ash to form a grout which is then disposed of in the SDF. DOE

intends to use DDA on those tanks that contain the lowest activity material. In addition, during

the same time period that DOE plans to use DDA, DOE plans to process a unique salt waste

stream currently stored in Tank 48. This waste consists of approximately 8.8 x 105

 L (0.24

Mgal) of salt solution containing 19,000 kg (8600 lbs) of potassium and cesium

tetraphenylborate salts generated during an earlier, unsuccessful attempt to use a different

technology for in-tank waste processing. DOE does not believe the waste in Tank 48 can be

processed to remove radionuclides because of the explosion hazard posed by organic

chemicals in the waste, and instead plans to combine the waste with another salt waste stream,

currently planned to be the low-activity liquid recycle waste stream from DWPF, and transfer it

to the SPF for processing and eventual disposal in the SDF. DOE indicated that the DDA

process would be used from 2005 until approximately 2009, when the SWPF is expected to

become operational.

The ARP/MCU process will be used to remove radionuclides in slightly higher activity salt waste

than that proposed for the DDA process only. ARP/MCU will be used to treat some salt waste

after it has undergone the DDA process. The ARP will use monosodium titanate (MST) strikes

and cross-flow filtration to remove and concentrate the insoluble actinides. The high activity

waste stream would be sent to the DWPF for vitrification and the low activity waste stream

would be sent to the SPF for mixture with cement, slag, and fly ash and then disposal in the

SDF. The MCU will use the CSSX process to remove cesium from the salt waste by using

organic solvents to complex with cesium atoms in the waste stream. The solvent and salt

solution will be fed to centrifugal contractors to ensure mixing, and the cesium will then be

stripped and sent to DWPF for vitrification. The lower activity salt solution will be sent to the

SPF for eventual disposal in the SDF. The ARP/MCU process would be used from 2007 until at

least 2009, and may also be used to augment actinide removal capacity during operation of the

SWPF.
- 4 -

Figure 2 DOE Salt Waste Processing Strategy

The SWPF will use the same technologies as ARP and MCU to remove strontium, actinides,

and cesium from the salt waste. However, due to its large scale, the SWPF will provide a much

higher decontamination factor for cesium and have a higher throughput. The SWPF will consist

of three operations: (1) Alpha Strike Process (ASP); (2) CSSX; and (3) Alpha Finishing Process 

(AFP). Like the ARP, the ASP will remove soluble strontium and actinides by sorption onto

MST. The resultant slurry will be filtered to concentrate the MST and insoluble solids, with the

filtrate sent forward through the SWPF and the concentrated MST/sludge sent to DWPF. The 

second SWPF operation will use the CSSX process to remove cesium, with the strip effluent

sent to DWPF and the remaining waste stream sent to the SPF or to AFP if additional

strontium and actinide removal is required to meet the SDF Waste Acceptance Criteria (WAC). 

AFP will be the final step in which an additional MST strike may be performed, if needed. The

SWPF will be capable of processing 1.8 x 107

 L (5 Mgal) of salt solution during the first year of

operation and 2.2 x 107

 to 3.3 x 107 L (6 to 9 Mgal) of salt solution per year each following year. 

The SWPF is planned to be operational by 2009 and continue to operate until the completion of

salt processing in approximately 2019.

The SPF will mix the low-activity salt solution with dry chemicals (cement, slag, and fly ash) to

form a grout mixture. The grout mixture will be pumped into engineered disposal vaults in the

SDF, where the grout will solidify into a monolithic solid low-level waste form called “saltstone.” 

The SDF will consist of several large concrete vaults divided into cells. The saltstone grout

itself provides primary containment of the waste, and the vaults provide secondary containment. 
- 5 -

Several inches of uncontaminated grout will be poured on top of the saltstone grout to reduce

skyshine. Vaults will be constructed as needed as salt waste processing continues. 

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

and July 1998, and radioactive operations began in June 1990. Since that time, it has operated

on an intermittent basis to treat and dispose of low-level waste from other on-site facilities. 

However, the SPF was originally designed to process wastes with lower activities than those

currently proposed. Therefore, DOE is undertaking modifications to the design and operation of

the facility, including additional shielding and equipment redesign to allow remote repair.

The SDF is located in the Z-area of SRS and is approximately 10 km (6.2 mi) from the nearest

SRS site boundary on a well-drained local topographic high (see Figure 3). The Z-area is

bounded by two streams: Upper Three Runs is approximately 1.6 km (1.0 mi) to the Northwest

and McQueen Branch, a tributary to Upper Three Runs, is approximately 1.6 km (1.0 mi) to the

East and 1.2 km (0.75 mi) to the Northeast. The facility has been sited and designed to ensure

that even under extreme wet conditions, the water table will not rise to the base of the vaults. 

The Probable Maximum Flood on Upper Three Runs or its tributaries is 53 m (175 ft) above

mean sea level which is below the planned maximum depth of the SDF vaults.

The two existing vaults (Vaults 1 and 4) are constructed of reinforced concrete containing slag. 

Vault 4 has dimensions of 60 m (197 ft) wide, by 180 m (590 ft) long, by 8 m (26 ft) high. The

vault is divided into 12 cells of approximately 30 m (98 ft) by 30 m (98 ft). The vault is covered

by a permanent roof with a minimum thickness of 10 cm (3.9 in) and a minimum slope of 0.19

cm/m (0.15 in/ft) from the midplane of the vault roof parallel to the short axis. The vault walls

are approximately 0.46 m (1.5 ft) thick and the base mat is 0.61 m (2.0 ft) thick. Figure 4

provides a photograph of the existing SDF Vault 4. The SDF closure concept is described in

Section 4.2.4. Up to 14 additional vaults may be required to dispose of the total volume of

saltstone produced and the vaults will be covered by an engineered cap at final closure.

1.3 Waste Determination Criteria 

Since 1969, the NRC has recognized the concept of incidental waste or waste incidental to

reprocessing (WIR). The concept underlying WIR is that certain wastes can be managed

based on their risk to human health and the environment, rather than the origin of the wastes. 

For wastes that originate from reprocessing of spent nuclear fuel, some are highly radioactive

and need to be treated and disposed of as high-level waste (HLW). Other reprocessing waste

does not pose the same risk to human health and the environment, and therefore does not

need to be disposed of as HLW. This type of waste is termed “incidental waste” or “WIR.” 

DOE uses “waste determinations” to evaluate whether reprocessing waste is HLW or incidental

waste. 

The original incidental waste criteria were approved by the Commission in the Staff

Requirements Memorandum (SRM) dated February 16, 1993, in response to SECY-92-391,

ADenial of PRM 60-4 B Petition for Rulemaking from the States of Washington and Oregon

Regarding Classification of Radioactive Waste at Hanford@. These criteria are described in the

March 2, 1993, letter from R. Bernero, NRC, to J. Lytle, DOE as follows (NRC, 1993): (1) The

waste has been processed (or will be further processed) to remove key radionuclides to the

maximum extent that is technically and economically practical; (2) The waste will be 
- 6 -

Figure 3 Location of the Z-Area
- 7 -

Figure 4 Photograph of Existing SDF Vault 4

incorporated in a solid physical form at a concentration that does not exceed the applicable

concentration limits for Class C LLW as set out in 10 CFR Part 61; and (3) The waste is to be

managed, pursuant to the Atomic Energy Act, so that safety requirements comparable to the

performance objectives set out in 10 CFR Part 61, are satisfied.

In the May 30, 2000, SRM on SECY-99-0284, “Classification of Savannah River Residual Tank

Waste as Incidental,” the Commission indicated that a more generic, performance-based 

approach should be taken in regard to reviewing WIR determinations (NRC, 2000a). In effect,

cleanup to the maximum extent that is technically and economically practical and demonstration

that performance objectives could be met (consistent with those which the Commission

demands for the disposal of LLW) should serve to provide adequate protection of the public

health and safety and the environment. In the “Final Policy Statement for the Decommissioning

Criteria for the West Valley Demonstration Project at the West Valley Site,” dated February 1,

2002, the Commission noted the criteria that should be applied to the incidental waste

determinations at West Valley (NRC, 2002): 

(1) The waste should be processed (or should be further processed) to remove key

radionuclides to the maximum extent that is technically and economically

practical; and 

(2) The waste should be managed so that safety requirements comparable to the

performance objectives in 10 CFR Part 61, Subpart C, are satisfied. 
- 8 -

In October 2004, the Ronald W. Reagan National Defense Authorization Act for Fiscal Year

2005 (NDAA) was signed into law. Section 3116 of the NDAA allows DOE to continue to use a

process to determine that waste is not HLW and, among other things, requires that DOE

consult with NRC on its non-HLW determinations. However, the NDAA is applicable only to

South Carolina and Idaho and does not apply to waste transported out of those States. The

NDAA establishes the following criteria for determining that waste is not HLW:

(1) The waste does not require permanent isolation in a deep geologic repository for

spent fuel or HLW;

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

extent practical; and

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

section 61.55 of title 10, Code of Federal Regulations, and will be disposed of--

(i) in compliance with the performance objectives set out in subpart

C of part 61 of title 10, Code of Federal Regulations; and

(ii) pursuant to a State-approved closure plan or State-issued permit,

authority for the approval or issuance of which is conferred on the

State outside of this section; or

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

61.55 of title 10, Code of Federal Regulations, but will be disposed of–

(i) in compliance with the performance objectives set out in subpart

C of part 61 of title 10, Code of Federal Regulations;

(ii) pursuant to a State-approved closure plan or State-issued permit,

authority for the approval or issuance of which is conferred on the

State outside of this section; and

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

the Commission

These are the criteria used by DOE in its draft waste determination for salt waste disposal.

1.4 NRC Review Approach

In February 2005, DOE-HQ and NRC established an Interagency Agreement (IA) that provided

a mechanism for reimbursement for NRC activities in FY05, as required by the NDAA (NRC,

2005b). The IA also provides a statement of work that includes NRC’s technical review of

DOE’s draft waste determination for salt waste processing at SRS. 

NRC’s review was based on DOE’s “Draft Section 3116 Determination, Salt Waste Disposal,

Savannah River Site.” A non-publicly-available version of the draft waste determination was

submitted by DOE on February 28, 2005, because DOE had not completed its required internal

security review. On March 31, 2005, DOE submitted a publicly-available version of the draft

waste determination, along with approximately 60 references, and it is that version on which the

NRC staff based its review (DOE, 2005). The publicly-available version of the draft waste

determination contained only minor, non-technical changes from the non-publicly-available

version. The NRC staff performed a technical review of the information and sent a request for

additional information (RAI) to DOE on May 26, 2005 (NRC, 2005a). The RAI contained

questions concerning areas such as waste characterization, removal of highly radioactive
- 9 -

radionuclides, engineered barriers, and saltstone degradation. NRC and DOE held open

meetings on June 8 and June 20, 2005, to discuss DOE’s preliminary proposed approaches to

responding to the RAI. In letters dated June 30 and July 15, 2005, DOE submitted its RAI

responses, as well as a Performance Objectives Demonstration Document (PODD), a revised

Special Analysis to supplement the site’s performance assessment (PA), and approximately

150 additional references. On July 27 and August 17-18, 2005, NRC and DOE met to discuss

DOE’s RAI responses. During those meetings, NRC requested that DOE provide additional

information supporting the RAI responses. DOE responded by submitting “Response to Action

Items from Public Meetings Between NRC and DOE to Discuss RAI for the Savannah River

Site” in two separate letters dated September 15 and 30, 2005, along with approximately 50

supporting references. All of the documents submitted by DOE are publicly available and are

referenced under the docket number PROJ0734 in the NRC’s Agencywide Document Access

and Management System (ADAMS).

NRC reviewed the draft waste determination and supporting documentation to assess whether

it had sound technical assumptions, analysis, and conclusions with regard to meeting the

criteria in the NDAA and thus, that DOE’s proposed management of salt waste is protective of

public health and safety and the environment. This approach is consistent with the one

proposed by the NRC staff in SECY-05-0073, dated April 28, 2005, and approved by the

Commission in the Staff Requirements Memorandum dated June 30, 2005. This TER

addresses each of the applicable criteria in the NDAA and presents the NRC staff’s approach,

assumptions, and conclusions, as well as factors that the NRC staff believes may be important

to meeting the performance objectives of 10 CFR 61, Subpart C.

NRC’s conclusions are dependent on the assumptions discussed in the TER and if DOE revises

its assumptions, analysis, design, or proposed waste management approach, DOE should

consult once again with NRC regarding the findings in this TER. It should be noted that NRC

staff is providing consultation to DOE as required by the NDAA and the NRC is not providing

regulatory approval in this action. DOE is responsible for determining whether the waste is

HLW. This NRC assessment is a site-specific evaluation and is not a precedent for any future

decisions regarding non-HLW or incidental waste determinations at SRS or at other sites.

1.5 Previous Reviews Performed for SRS

In December 1996, DOE requested that NRC review the DOE-SRS methodology for

determining that residual tank waste was WIR. A Memorandum of Understanding between

NRC and DOE was established in July 1997. Following the establishment of the MOU, NRC

began its review of the “Regulatory Basis for Incidental Waste Classification at the Savannah

River Site High-Level Waste Tank Farms.” The NRC reviewed the information provided by

DOE and issued a RAI in June 1998. In September 1998, DOE responded to the RAI and a

public meeting was held in April 1999 to resolve some of the remaining outstanding issues. 

Following the public meeting, DOE submitted supplementary RAI responses in April 1999. 

NRC issued its final letter to DOE in June 2000. This previous WIR review used Criterion One 

and Criterion Three of the criteria established in 1993 (i.e., not the criteria in the NDAA) and

was for waste remaining in tanks, not for waste being removed from the tanks and disposed of

elsewhere. Therefore, this prior review is not directly applicable to the salt waste disposal draft

waste determination.
- 10 -

2. CRITERION ONE

The waste does not require permanent isolation in a deep geologic repository for spent fuel or

HLW

2.1 Salt Waste Disposal 

Criterion One allows for the consideration that waste may require disposal in a geologic

repository even though the two other criteria of the NDAA may be met. Consideration could be

given to those circumstances under which geologic disposal is warranted in order to protect

public health and safety and the environment; for example, unique radiological characteristics of

waste or non-proliferation concerns for particular types of material. 

2.2 NRC Review and Conclusions (Criterion One)

Given the analysis in the following sections of this TER, which indicates that there is reasonable

assurance that DOE can meet the applicable criteria in the NDAA based on certain

assumptions, and the fact that there is no indication that other considerations would warrant

disposal of the salt waste in a geologic repository, the NRC staff finds that there is reasonable

assurance that Criterion One can be met by DOE.

3. CRITERION TWO

The waste has had highly radioactive radionuclides removed to the maximum extent practical

To evaluate compatibility with this criterion, NRC staff evaluated DOE's tank inventory

estimates, estimated salt waste composition, process for identifying highly radioactive

radionuclides, predicted radionuclide removal efficiencies, waste treatment technology

selection, and alternatives to the proposed two-phase, three-part process. 

3.1 Waste Inventory 

The tank waste inventory (as of 12/1/04) was 1.34 x 108

 L (36.4 Mgal) containing 1.58 x 1019 Bq 

(426 MCi) of radioactivity. The radiological composition of the tank waste stored in the 49

operational underground carbon steel waste storage tanks can vary substantially from tank to

tank and among the three primary physical phases of waste (salt supernate, saltcake, sludge). 

DOE’s practice at SRS has been to use evaporator systems to reduce the volume of supernate

and concentrate the waste. During the evaporation process, salt waste forms two distinct

phases: concentrated supernate solution and solid saltcake (collectively referred to as salt

waste). Supernate is a concentrated salt solution (primarily nitrate and nitrite salts) that rests

on top of the saltcake and fills the pore space of the saltcake. Saltcake is the crystallized

precipitate formed when the salts reach their solubility limit. Because the concentrations of

cesium and most other radioactive elements in the supernate are significantly lower than the

concentrations of non-radioactive isotopes of sodium, they do not precipitate as salts and their

concentrations in the saltcake are significantly lower than their concentrations in the supernate. 

Other than C-14, Na-22, and Al-26, radionuclides are not expected to be incorporated into salt

crystals to a significant degree. Other radionuclides that are physically located in the saltcake
- 11 -

layer are believed to be contained within entrained sludge. Sludge is a complex mixture of

insoluble metal hydroxide solids that have precipitated in the tank. Although most of the sludge

settles to the bottoms of the tanks, some insoluble metal hydroxides are suspended in the

supernate and also incorporated in the saltcake. 

To keep track of tank inventories, DOE has developed a Waste Characterization System

(WCS), which is an electronic information system that is used to track waste tank data based

on sample analyses, process histories, composition studies, and theoretical relationships. The

system is used to track 96 chemicals and 41 radionuclides in each of the waste tanks. The

system is updated routinely to reflect changes in tank inventory resulting from the receipt of

new waste, tank to tank transfers, and evaporator operation. The WCS also is updated with the

results of analyses of samples of supernate, salt, and sludge (Tran, 2005). In 2003 and 2004,

a special campaign was completed to improve estimates of the radiological composition of the

supernate and salt. The technical objectives of DOE’s sampling approach are to understand

the physical, chemical, and radiological characteristics of material stored in the waste tanks to

support selection of appropriate options for waste treatment and disposal. In some cases

sampling is limited because of access restrictions and costs associated with sampling waste

from HLW tanks. 

Sampling of the waste in the HLW tanks directly supports the projected composition of waste

generated from the DDA, ARP/MCU, and SWPF for disposition in the saltstone facility. DOE

estimates that the salt waste treatment and processing strategy will result in 6.27x 108

 L (170

Mgal) of grout containing 1.1 x 1017 Bq (3 MCi) of radioactivity to be disposed of in the SDF. 

The SDF has established WAC to ensure that the concentration of waste accepted by the

facility is suitable for disposal. The saltstone facility WAC establishes the radionuclide limits for

the waste, in addition to physical and chemical property limits. DOE plans to demonstrate

compliance with the saltstone WAC prior to transferring waste from Tank 50 (the saltstone feed

tank) to the SPF by sampling the waste in the feed tank or with material balance calculations

and process knowledge (Ketusky, 2005). Compliance with the SDF WAC is discussed in

further detail in Sections 4.2.2 and 4.2.3.

Table 1 provides the composition of supernate, sludge entrained in saltcake, and the sludge

layer averaged over the two tank farms for the radionuclides considered in detail by DOE (see

Section 3.3). However, waste composition varies significantly from tank to tank. For example,

sludge formed from high-heat waste from the first canyon cycle typically has fission product

concentrations that are approximately three orders of magnitude greater than the fission

product concentrations in sludge formed from low-heat waste from the second canyon cycle

(WSRC, 2005b).

In addition to the low-activity salt waste that will be processed during the interim period, DOE

has proposed to create additional tank space by removing material from Tank 48, combining it

with DWPF recycle, and directly sending the material to the SPF to be solidified into a

low-activity saltstone wasteform with no additional processing. Tank 48 is a new style tank with

a strategic location with respect to the salt processing facilities. It currently contains

approximately 8.8x 105

 L (0.24 Mgal) of salt solution that contains approximately 19,000 kg

(8600 lbs) of potassium and cesium tetraphenylborate (TPB) salts from the unsuccessful InTank Precipitation (ITP) process (Ledbetter, 2004; Fowler, 2004). The Tank 48 waste is

estimated to contain 3 x 1016 Bq (0.8 MCi) of activity.
- 12 -

Table 1 Average Composition of Primary Physical Phases in DOE-SRS Tank Waste

(Based on information in Tran [2005])

Radionuclide Supernate 

(Ci/L)

Sludge Entrained in

Saltcake (Ci/L)

Sludge Layer 

(Ci/L)

Cs-137* 2.43 8.19 x 10-3 9.62 x 10-1

Sr-90* 6.12 x 10-4 1.88 x 10-1 1.65 x 101

Se-79 9.84 x 10-7 not reported 1.47 x 10-4

Tc-99 3.65 x 10-4 not reported 2.52 x 10-3

I-129 1.97 x 10-7 not reported 1.01 x 10-8

Sn-126 4.97 x 10-6 not reported 1.92 x 10-4

U-232 3.17 x 10-10 not reported 5.32 x 10-8

U-233 2.93 x 10-8 not reported 9.31 x 10-6

U-234 4.64 x 10-8 not reported 3.54 x 10-6

U-235 9.24 x 10-10 3.70 x 10-9 1.51 x 10-7

U-236 3.95 x 10-9 not reported 5.82 x 10-7

U-238 7.51 x 10-8 8.30 x 10-8 6.12 x 10-6

Np-237 4.69 x 10-8 not reported 9.36 x 10-6

Pu-238 6.35 x 10-4 1.20 x 10-3 1.90 x 10-1

Pu-239 3.73 x 10-5 4.00 x 10-5 4.53 x 10-3

Pu-240 1.00 x 10-5 not reported 1.96 x 10-3

Pu-241 4.15 x 10-4 not reported 1.18 x 10-1

Pu-242 1.04 x 10-8 not reported 3.10 x 10-6

Am-241 4.01 x 10-6 not reported 3.43 x 10-2

Am-242m 2.31 x 10-9 not reported 1.97 x 10-5

Cm-244 1.60 x 10-6 not reported 1.37 x 10-2

Cm-245 1.58 x 10-10 not reported 1.35 x 10-6

* Includes daughter
- 13 -

3.2 NRC Evaluation – Waste Inventory 

To estimate the inventory of various radionuclides in the tank farm system, DOE uses the WCS

to account for additions of waste to the tank farm and transfers between tanks. The WCS

system appears to be an effective tool to track the quantities and composition of the waste in

the tank farm. However, a number of inputs to the WCS are based on assumptions, process

knowledge, or other forms of indirect information. Despite campaigns to improve the estimates

of radiological composition in the salt and supernate in 2003 and 2004, sampling remains

limited. For example, although Cs-137 is relatively well characterized compared to other

radionuclides, DOE estimates that, because of uncertainties in salt waste characterization, the

total amount of activity sent to saltstone could be 70% higher than the expected value of

1.1 x 1017 Bq (3 MCi) (DOE, 2005). As discussed in Section 3.6, the NRC staff concludes that

the greatest amount of uncertainty in the composition of the waste sent to the SDF results from

uncertainties in the tank inventories and uncertainties in the results of the DDA process. NRC’s

evaluation of DOE’s plan to control the uncertainty in the SDF inventory is discussed in Section

4.2.3. 

3.3 Identification of Highly Radioactive Radionuclides

In its March 2005 submittal, DOE identified “radionuclides considered in detail” as Cs-137 (and

its daughter Ba-137m), Sr-90 (and its daughter Y-90), Se-79, Tc-99, I-129, Sn-126, and

actinides (isotopes of U, Pu, Am, Np, and Cm). DOE stated in its draft waste determination that

the selection of radionuclides considered in detail was based on scientific expertise, knowledge,

and health physics principles as applied to the SRS salt waste, and that the list of radionuclides

considered in detail included those radionuclides in Tables 1 and 2 of 10 CFR 61.55 that are in

the SRS salt waste in quantities such that they may be important to meeting the performance

objectives in 10 CFR 61, Subpart C. Highly radioactive radionuclides were not explicitly defined

in the March 2005 submittal (DOE, 2005). In its RAI, NRC asked DOE to provide a list of the

radionuclides it classified as highly radioactive radionuclides in the context of the SRS saltstone

waste determination (NRC, 2005a). In its RAI, NRC also indicated that the NRC staff believes

that highly radioactive radionuclides are those radionuclides that contribute most significantly to

risk to the public, workers, and the environment.

In response to NRC’s RAI, DOE indicated that it believes highly radioactive radionuclides in the

context of Section 3116 to be those radionuclides that, using a risk-informed approach,

contribute most significantly to radiological risk to the workers, the public, and the environment

(WSRC, 2005b). In its RAI response and supporting document, DOE indicated that it based its

identification of highly radioactive radionuclides on the characteristics of saltstone formed from

untreated salt waste (i.e., salt waste that had been solidified without application of DDA,

ARP/MCU, or SWPF treatment). The approach developed by DOE involved two types of

comparisons. First, DOE compared the concentrations of radionuclides in solidified untreated

salt waste to NRC Class A concentration limits. Then, DOE calculated public, intruder, and site

worker doses that would result if salt waste were solidified and placed in the SDF without

treatment and compared each of these doses to 10% of the appropriate low-level waste

performance objective (10 CFR 61.41, 61.42, or 61.43). Based on the first comparison, DOE

concluded that Cs-137, Sr-90, and alpha-emitting transuranic nuclides were the only

radionuclides in untreated solidified salt waste that would exceed NRC Class A concentration

limits (10 CFR 61.55). Second, DOE used the pathway dose conversion factors it calculated for
- 14 -

its deterministic base-case PA for Vault 4 to predict the dose to a member of the public from

radionuclides in untreated solidified waste (Cook et al., 2005). DOE concluded that, in its basecase analysis, no radionuclides in untreated solidified salt waste would result in a dose to a

member of the public that exceeded 10% of the 0.25 mSv/yr (25 mrem/yr) public dose limit (10

CFR 61.41). Third, DOE concluded that Cs-137 was the only radionuclide in untreated

solidified waste that would exceed a dose of 0.50 mSv (50 mrem), or 10% of the 5 mSv

(500 mrem) used in the Draft Environmental Impact Statement for 10 CFR 61 (see Section

4.2.18), to an inadvertent intruder. Fourth, DOE concluded that Cs-137 (including its daughter,

Ba-137m) was the only radionuclide in untreated salt waste that would result in a worker

gamma dose in excess of 500 mrem/yr, or 10% of the 50 mSv (5 rem/yr) limit (10 CFR 61.43). 

Finally, DOE noted that Sr-90, Cs-137, and alpha-emitting TRU nuclides were the only

radionuclides driving worker inhalation dose. On the basis of these comparisons, in its RAI

response (WSRC, 2005b), DOE identified Sr-90, Cs-137, and alpha-emitting transuranic

nuclides (alpha-emitting isotopes of Pu, Am, Np, and Cm) as highly radioactive radionuclides in

the context of the SRS saltstone Section 3116 draft waste determination.

In the RAI response (WSRC, 2005b) and supporting document (WSRC, 2005c), DOE indicated

that, although Se-79, I-129, Tc-99, Sn-126, and uranium isotopes were radionuclides

considered in detail in the original waste determination, they were not highly radioactive

radionuclides in the context of the draft waste determination for SRS salt waste because DOE

believed that concentrations of these nuclides in untreated salt waste were sufficiently low that

they did not present a significant radiological risk to members of the public, site workers, or the

environment. In addition, DOE noted that the concentrations of I-129 and Tc-99 in untreated

salt waste are expected to be below NRC Class A limits (10 CFR 61.55). However, after the

submittal of the RAI response, DOE added Se-79, I-129, and Tc-99 to its list of highly

radioactive radionuclides for the draft waste determination because DOE determined that

Se-79, I-129, and Tc-99 dominated the predicted doses to members of the public in several

sensitivity analysis cases that DOE performed at the request of NRC (see Sections 4.2.15 and

4.2.16) (WSRC, 2005a). In addition, in its written response to action items from the public

meetings between NRC and DOE to discuss the RAI (WSRC, 2005a), DOE indicated that only

four alpha-emitting transuranic nuclides, Pu-238, Am-241, Cm-244, and Pu-239, were being

retained on its list of highly radioactive radionuclides in the context of the Section 3116 draft

waste determination for SRS salt waste. In a supporting document (WSRC, 2005c), DOE

indicated that these four nuclides account for approximately 99% of the transuranic alpha

activity in untreated salt waste. Thus, DOE’s final list of highly radioactive radionuclides in the

context of the Section 3116 draft waste determination for SRS salt waste includes Cs-137 (and

its daughter, Ba-137m), Sr-90 (and its daughter, Y-90), Pu-238, Am-241, Cm-244, Pu-239, Se79, I-129, and Tc-99.

3.4 NRC Evaluation – Identification of Highly Radioactive Radionuclides 

The definitions of “highly radioactive radionuclides” used by NRC staff and by DOE appear to

be consistent. Specifically, NRC staff and DOE agree that highly radioactive radionuclides are

those radionuclides that contribute most significantly to radiological risk to the public, workers,

and the environment. To identify highly radioactive radionuclides, DOE compared the

inventories of radionuclides in untreated salt waste (i.e., salt waste solidified without application

of DDA, ARP/MCU, or SWPF treatment) to NRC Class A concentration limits (WSRC, 2005c). 

DOE also compared predicted doses resulting from untreated solidified salt waste to workers,
- 15 -

members of the public, and inadvertent intruders to NRC limits (WSRC, 2005c). Based on this

approach, DOE identified highly radioactive radionuclides as Cs-137 (and its daughter, Ba137m), Sr-90 (and its daughter, Y-90), and alpha-emitting isotopes of Pu, Am, Np, and Cm. 

Based on the results of sensitivity analyses performed after the RAI response submittal, DOE

added Se-79, Tc-99, and I-129 to its list of highly radioactive radionuclides (WSRC, 2005a). 

NRC staff evaluated DOE's final selection of highly radioactive radionuclides and has concluded

that the combination of approaches used by DOE to identify highly radioactive radionuclides in

the context of the draft waste determination for SRS saltstone waste is appropriate.

Although DOE identified Se-79, Tc-99 and I-129 as highly radioactive radionuclides, DOE also

concluded that “the associated risks of Se-79, Tc-99 and I-129 are so low that it would not be

sensible or reasonable to target these radionuclides for further removal” (WSRC, 2005a). 

While the NRC staff agrees that DOE’s base-case PA results are consistent with this

conclusion (see Sections 4.2.15 and 4.2.16), the NRC staff believe that the consideration given

to sensitivity and uncertainty analysis results must be related to the amount of model support

provided for the base-case results. As discussed in Section 4.2.16, the NRC staff has

concluded that there are realistically conservative cases in which doses due to Se-79 and I-129

may be significantly greater than DOE’s predicted base-case result. In addition, as discussed

in Section 4.2.16, the NRC staff has concluded that there may be more uncertainty in the

potential dose from Tc-99 than appears to be recognized by DOE. The NRC staff concludes

that DOE’s use of sensitivity analysis to identify highly radioactive radionuclides is appropriate

and that it is inappropriate to discount potential doses from highly radioactive radionuclides 

solely on the basis of base-case results, especially in cases in which the base-case has

insufficient support (see Section 4). 

3.5 Radionuclide Removal Efficiencies

Section 1.2 provides an overview of DOE’s strategy for salt waste processing. As previously

described, DOE intends to process the salt waste to segregate the low-activity fraction by using

a two-phase, three-part approach (see Figure 2). The first phase (called Interim Salt

Processing) will involve two parts to treat the lower activity salt waste: (1) processing of a

minimal amount of the lowest activity salt waste through the DDA process, and (2) processing

of a minimal amount of additional waste with slightly higher activity levels using the ARP and

MCU, along with DDA. The second phase would involve processing of the majority of the salt

waste using the SWPF. 

Each part of the salt waste processing strategy (DDA, ARP/MCU, SWPF) involves unit

operations and processes capable of removing highly radioactive radionuclides. Table 2

provides an overview of the activities of waste DOE plans to treat with the proposed salt waste

treatment processes. The activity totals are dominated by the fission products Cs-137 and

Sr-90 and their daughters. Because treatment efficiency in each of the primary processes is

sensitive to the physical phase of the radionuclides in the waste and because some of the unit

processes are radionuclide-specific, the removal efficiencies for individual radionuclides

considered in detail may be different than the total activity removal efficiencies shown in

Table 2. Table 3 provides the expected efficiencies for removal of radionuclides DOE

considered in detail in the draft waste determination by each of the primary treatment

processes. In general, more soluble radionuclides (e.g., I-129, Tc-99, Cs-137) can only be

removed by chemical processes whereas relatively insoluble radionuclides (e.g., Sr-90, 
- 16 -

Table 2 Overview of Salt Waste Processing Efficiencies

Process Total Input (MCi) Total Output to SDF (MCi) Removal Efficiency (%)

DDA 8.8 2.5 79 (3)

ARP/MCU 3.4 0.3 88

SWPF 217.1 0.2 99.9

Total 223 (1) 3.0 (2) 98.7

(1) The total is not the sum of the column values because some output from DDA will be sent to the

SWPF

(2) Because of uncertainty in the inventory, this value may be as high as 5.0 MCi

(3) As discussed in the text, the removal efficiency reported by DOE has been adjusted so that it does not

reflect the contribution of Tank 48 waste.

actinides) also can be removed by physical processes such as settling and filtration. Some

radionuclides considered in detail (e.g., Se-79, Sn-126) have significant fractions in both the

liquid and solid phases.

Of the 9.3 x 1016 Bq (2.5 MCi) that DOE predicted will be sent to SPF from the DDA process,

3 x 1016 Bq (0.8 MCi) originates in Tank 48 waste (DOE, 2005). The waste in Tank 48 is

incompatible with the existing treatment processes because of organic components of the TPB

salts. Flammable vapors (benzene) are produced from the decomposition of TPB, and the rate

of decomposition increases with increasing temperature and the presence of catalysts. The

presence of catalysts in other tank farm wastes and the need to maintain relatively low

temperatures precludes mixing the Tank 48 waste with other waste or processing it through the

proposed treatment systems. Because of the potentially problematic interactions of Tank 48

waste with other tank farm waste or with the proposed treatment processes, Tank 48 waste will

be sent to SPF without treatment; that is, there will be no removal of highly radioactive

radionuclides from the Tank 48 waste before it is solidified. In Table 2, the removal efficiency

reported by DOE for the DDA process, which included Tank 48 waste as part of the DDA total,

has been adjusted to reflect activity removal only in the waste that will be treated with the DDA

process. 

Table 3 shows predicted efficiencies of removal of radionuclides considered in detail by DOE. 

DOE's predicted removal efficiencies are shown for the radionuclides for which removal

efficiencies were provided. For radionuclides for which DOE did not supply predicted removal

efficiencies, NRC staff estimated removal efficiencies. The methods and assumptions that

NRC staff used to calculate removal efficiencies are discussed in Section 3.6. DOE predicts

that two-thirds of the entrained sludge in salt waste treated with the DDA process will be

removed by settling after saltcake dissolution. In addition, DOE predicts that the deliquification

process will remove half of the interstitial supernate in saltcake (Pike, 2005). In calculating

DDA removal efficiencies for Cs, Sr, and actinides, DOE estimated that 66% of the insoluble

radionuclides and 50% of the dissolved radionuclides would be removed by the DDA process

(WSRC, 2005c). DOE predicted that approximately 100% of the insoluble radionuclides would

be removed by filtration during the ARP/MCU and SWPF processes. In addition, DOE

predicted that 91% of the dissolved Cs, 99.2% of the dissolved Sr, 92.3% of the dissolved Pu,

41% of the dissolved Am, and 41% of the dissolved Cm would be removed during ARP/MCU 
- 17 -

Table 3 Projected Removal Efficiencies of Radionuclides Considered in Detail (in %)

Radionuclide DDA (1,2) ARP/MCU (2,4,5) SWPF (5) Overall (7)

Cs-137 50 (3) 91 (3) 99.998 (3) 99

Sr-90 66 (3) 99.997 (3) 99.98 (3) > 99

Se-79 40 59 59 60 (8)

Tc-99 4.2 6.3 6.3 5.7

I-129 0.03 0.05 0.05 < 0.05 (9)

Sn-126 18 27 27 27

U-232 46 (4) 72 72

U-233 53 (4) 82 80

U-234 36 (4) 58 47

U-235 46 (4) 71 70

U-236 45 (4) 70 66

U-238 37 (4) 59 58

Np-237 54 (4) 86 84

Pu-238 63 (3) 98.1 (3) 95.5 (3) 94

Pu-239 59 (3) 96.4 (3) 91.6 (3) 91

Pu-240 58 97 94 93

Pu-241 60 98 95 95

Pu-242 60 98 96 95

Am-241 66 (3) 99.3 (3) 99.7 (3) > 99

Am-242m 66 99 99.8 > 99

Cm-242 66 99 99 > 99

Cm-243 66 99 99 > 99

Cm-244 66 (3) 99.3 (3) 99.8 (3) 99

Cm-245 66 98 99 99

alpha-emitting TRU 63 (3) 98.1(3) 96 (3) 95

(1)

Unless otherwise noted, removal efficiencies were calculated by assuming removal of 66% of the insoluble fraction

of radionuclides and SWPF removal efficiencies of 50% of the soluble radionuclides, as discussed in Section 3.6. 

(2) Unless otherwise noted, removal efficiencies were calculated based on the phase partitioning of average salt

waste (WSRC, 2005c). As discussed in Section 3.6, actual removal efficiencies of the DDA and ARP/MCU

processes may differ significantly from the estimated removal efficiencies for radionuclides for which phase

partitioning in the average untreated salt waste is not representative of phase partitioning in the waste sent to DDA

or ARP/MCU. 

(3) Removal efficiency reported by DOE (WSRC, 2005c). 

(4) Insufficient information was provided for NRC staff to estimate a removal efficiency, as discussed in Section 3.6. 

(5) Unless otherwise noted, removal efficiencies were calculated based on removal of 100% of the insoluble fraction

and DOE's predicted removal efficiencies for soluble radionuclides for the appropriate process (WSRC, 20005c). 

(6) As described in Section 3.6, the removal efficiencies for dissolved radionuclides were those for the use of ARP

with MST (WSRC, 2005c). 

(7) Unless otherwise specified, overall removal efficiencies were calculated based on the inventory of radionuclides in

salt waste (WSRC, 2005c) divided by the projected inventory in the SDF (d’Entremont and Drumm, 2005).

(8) The overall removal efficiency appears to be greater than removal efficiency of individual processes because of

differences in rounding.

(9) Removal efficiency was calculated based on removal of insoluble I-129 because the removal efficiency calculated

based on the predicted saltstone inventory was affected by a rounding error. 
- 18 -

treatment (WSRC, 2005c). DOE predicted that 99.998% of the dissolved Cs, 95% of the

dissolved Sr, 82% of the dissolved Pu, 78% of the dissolved Am, 58% of the dissolved Np, and

26% of the dissolved U would be removed during SWPF treatment (DOE, 2005). DOE

indicated that dissolved Se, I, Tc, and Sn were not specifically targeted for removal by salt

waste treatment processes (DOE, 2005). 

The ARP/MCU process can be used with or without MST strikes to remove soluble Sr and

actinides. The removal efficiencies used to calculate the removal of dissolved radionuclides

were those predicted for the use of ARP with MST (WSRC, 2005c) because DOE has stated

that it intends to use MST strikes for all batches of waste processed with ARP as long as the

necessary processing throughput can be maintained (WSRC, 2005b). However, if the target

throughput cannot be maintained and if the solidified waste would still meet Class C

concentration limits, ARP may be run in a “filtration only” mode without MST strikes. The

primary effect of not using MST strikes during ARP treatment would be to lower the expected

efficiencies of removal of Pu-238 (from 98.1% to 75%), Pu-239 (from 96.4% to 54%), and total

alpha-emitting TRU (from 98.1% to 78%). The removal efficiencies of Sr-90, Am-241, and Cm244 would decrease slightly but would remain above 98%.

3.6 NRC Evaluation – Radionuclide Removal Efficiencies

Table 3 shows predicted efficiencies of removal of radionuclides considered in detail by DOE.

As discussed in Section 3.5, DOE estimated that 66% of the insoluble radionuclides and 50% of

the dissolved radionuclides would be removed by the DDA process (WSRC, 2005c). This

approach to the calculation of the removal of dissolved radionuclides by the DDA process

appears to underestimate DDA removal of Cs-137 and overestimate DDA removal of other

dissolved radionuclides. The approach appears to underestimate the removal of Cs-137

because the factor of 0.5 is applied to the entire dissolved fraction of Cs-137 even though the

free supernate is expected to be removed entirely rather than only being approximately 50%

removed as the interstitial supernate is (Pike, 2005). For example, the May 2005 WCS report

indicates that approximately 30% of the total supernate in the tanks DOE plans to treat with

DDA (Tanks 25, 28, and 41) is free, rather than interstitial (Tran, 2005). If it is assumed that

30% of the supernate is free and that 100% of the free supernate is removed and later treated

at the SWPF, the nominal efficiency of removal of Cs-137 by DDA would be 65% rather than

DOE's reported value of 50%. DOE's approach is also based on the assumption that all of the

dissolved radionuclides in the interstitial supernate removed from the saltcake during

deliquification are subsequently removed from the waste stream and are not sent to the SDF. 

This assumption is appropriate for Cs because the supernate removed from the tanks by

deliquification is held for treatment in the SWPF, which, DOE predicts, will remove 99.998% of

Cs-137. However, this approximation would overestimate the removal of other soluble

radionuclides that are not removed by downstream processes. For example, I-129 is present

almost entirely in supernate, but is not removed during SWPF treatment; therefore, although

approximately half of the I-129 may be removed from the saltcake during deliquification, it is not

removed from the extracted supernate during subsequent processing and ultimately is sent to

the SDF.

In calculating removal efficiencies, DOE also made the approximation that the phase

partitioning of radionuclides in waste being sent to each primary process was well represented

by the phase partitioning of the radionuclides in the average untreated salt waste (WSRC,

2005c). While this is a reasonable approximation for the calculation of SWPF removal
- 19 -

efficiencies because most of the waste will be processed through SWPF, it is a more tenuous

assumption for both DDA and ARP/MCU processes because a relatively small fraction of the

waste will be processed through these processes. If the partitioning of a radionuclide between

the solid and liquid phases in batches of waste treated with the DDA and ARP/MCU processes

is significantly different than the partitioning of the radionuclide in the average untreated salt

waste, the actual removal efficiencies may be significantly different than the projected removal

efficiencies shown in Table 3. This potential discrepancy is unlikely to affect the observed

removal efficiencies of radionuclides that are either strongly soluble (e.g., Cs-137, I-129, Tc-99)

or strongly insoluble (e.g., Sr-90, actinides), but could significantly affect the observed removal

efficiencies of radionuclides whose partitioning can vary significantly from tank to tank (e.g.,

Se-79).

For radionuclides for which DOE did not provide DDA removal efficiencies, NRC staff calculated

removal efficiencies based on the following assumptions: (1) 66% of the insoluble fraction of

each radionuclide is removed through settling, (2) half of the total volume of free and interstitial

supernate is removed by deliquification, as in DOE's calculation, and (3) the predicted

efficiencies of the removal of dissolved radionuclides from supernate separated from saltcake

during the DDA process are given by the removal efficiencies DOE reported for the removal of

dissolved radionuclides by the SWPF. Neither of the assumptions about the removal of

dissolved radionuclides has a significant effect on the predicted removal of Sr-90 and the

actinides by the DDA process because they are relatively insoluble and are removed primarily

through settling. Furthermore, assumptions about the amount of supernate that is removed

during the DDA process do not affect the predicted removal of highly soluble radionuclides that

are not removed by SWPF (e.g., I-129 and Tc-99). Because of these insensitivities to the

assumptions discussed, for radionuclides other than Cs-137, calculated removal efficiencies

were within 4% of DOE estimates irrespective of whether it was assumed that all of the

supernate was interstitial or that 30% was free.

NRC staff used a similar method to calculate removal efficiencies for the ARP/MCU and SWPF

processes. For isotopes for which removal efficiencies were not provided by DOE, NRC staff

calculated removal efficiencies based on the removal of 100% of the insoluble fraction of the

isotope and the ARP/MCU or SWPF removal efficiency that DOE predicted for the soluble

phase of the appropriate element (DOE, 2005; WSRC, 2005c). Because DOE did not provide

efficiencies of removal of dissolved U or Np by the ARP/MCU process, NRC staff could not

calculate predicted removal efficiencies of isotopes of these elements by the ARP/MCU

process. Because DOE indicated that dissolved Se-79, I-129, Tc-99, and Sn-126 were not

targeted for removal by salt waste treatment processes, and that removal of these

radionuclides would occur primarily because of removal of insoluble radionuclides (DOE, 2005),

NRC staff assumed that neither the ARP/MCU nor the SWPF processes would remove any of

the dissolved fraction of these radionuclides.

For comparison, NRC staff calculated an "overall" removal efficiency for each radionuclide

(Table 3) based on the predicted inventory of a radionuclide in salt waste (WSRC, 2005c)

divided by the projected inventory of the radionuclide in the SDF (d'Entremont and Drumm,

2005). Although DDA treatment is expected to remove a lower fraction of most radionuclides

than the ARP/MCU and SWPF processes, the overall treatment efficiency for many

radionuclides is similar to the SWPF removal efficiency because a relatively small fraction of the

total inventory of each radionuclide is expected to be processed with DDA alone (Table 2).
- 20 -

In addition to uncertainties in tank inventories, uncertainties in SDF inventory will reflect

uncertainties in the performance of each waste treatment process. Uncertainties in the extent

of deliquification that will be achieved, the amount of sludge that will become entrained in salt

waste, and the effectiveness of the DDA settling step to remove entrained sludge from DDA

waste will contribute to the uncertainty in concentrations of radionuclides in DDA waste sent to

the SDF. DOE estimated that deliquification will remove from 30% to 70% of the interstitial

supernate and that settling will remove from 50% to 85% of the entrained sludge from the salt

waste and concluded that the overall removal efficiency of the DDA process varied by

approximately ±20% of the nominal value for both soluble and insoluble radionuclides (WSRC,

2005c). DOE also noted that the uncertainty in the amount of sludge that would be entrained in

salt waste is more difficult to estimate, and that the concentration of entrained sludge in DDA

waste prior to settling could vary from approximately 0 milligrams of sludge per liter salt waste

to several grams of entrained sludge per liter salt waste (WSRC, 2005b). The uncertainty

associated with the extent of radionuclide removal by the ARP/MCU is on the order of only a

few percent for Cs-137, Sr-90, Am-241, and Cm-241, but is greater for Pu-238 and Pu-239

because Pu is more evenly partitioned between the sludge and supernate; the uncertainty in the

predicted efficiency of removal of both Pu-238 and Pu-239 by the ARP/MCU process is

approximately 25% if MST sorption is not used and approximately 10% if MST sorption is used. 

Reported uncertainties in the efficiency of removal of radionucides by the SWPF are less than

1% for Sr-90 and Cs-137, and approximately 5% or less for Pu-238, Pu-239, Am-241, and

Cm-244.

Based on the information about uncertainties in tank inventories and removal processes

provided by DOE, the NRC staff concludes that the greatest amount of uncertainty in the

composition of the waste sent to the SDF results from uncertainties in the tank inventories and

uncertainties in the results of the DDA process. NRC staff note that uncertainties in the DDA

process are expected to have the largest effect on the SDF inventories of radionuclides that the

ARP/MCU or SWPF treatment processes remove with a high removal efficiency. For example,

because DOE predicts that approximately 90% of the Cs-137 in the SDF will originate from the

DDA process (d'Entremont and Drumm, 2005), and that Cs-137 is primarily partitioned into the

supernate (WSRC, 2005c), the amount of deliquification achieved during the DDA process will

have a significant effect on the SDF inventory of Cs-137. Similarly, DOE predicts that

approximately 77% of the Sr-90 in the SDF will originate from the DDA process (d'Entremont

and Drumm, 2005). Because DOE predicts that approximately 100% of the Sr-90 in salt waste

will be insoluble (WSRC, 2005c), the SDF inventory of Sr-90 will be sensitive to the amount of

sludge entrained during the DDA process and the effectiveness of settling during the DDA

process. Thus NRC staff concludes that efforts to maximize the amount of deliquification

achieved, to minimize the amount of sludge entrained during DDA, and to maximize the amount

of sludge that settles during DDA processing will have a significant effect on the inventory of

highly radioactive radionuclides in the SDF and the demonstration that highly radioactive

radionuclides have been removed to the maximum extent practical. Uncertainties in the SDF

source term and process for determining whether waste will meet the SPF WAC are discussed

with respect to the source term used in the PA calculations (Sections 4.2.2 and 4.2.3).

3.7 Selection of Treatment Processes 

In developing its strategy for salt waste processing and disposal, DOE evaluated technologies

available to remove radionuclides from salt waste. In addition, DOE compared alternatives to
- 21 -

the proposed two-phase, three-part approach by evaluating cases in which waste processing

would be delayed until either the ARP/MCU process or the SWPF becomes operational.

3.7.1 Alternative Treatment Technologies

The CSSX technology used in both the SWPF and MCU process for removal of Cs-137 was

selected from approximately 140 alternatives that fell into 11 general technological categories

(WSRC, 1998a). Based on the results of the original evaluation, conducted in 1998, DOE

initially concluded that the CSSX alternative was technologically immature (DOE, 2001a). 

However, DOE resumed consideration of CSSX based on subsequent research and

development results and a recommendation by the National Academy of Sciences (NAS) (NAS,

2000). In the original 1998 evaluation, technology selection was based on expert judgment. 

Each technology was judged based on the unknowns associated with design and

implementation (i.e., level of development of the technology), risk of not meeting the waste

treatment goals, potential risks to meeting the necessary budget and schedules, and safety

risks, including nuclear safety risks, process hazards, and potential accidents (WSRC, 1998a). 

In 2000 and 2001, the NAS reviewed DOE’s technology evaluations and selections (NAS, 2000;

NAS, 2001). NAS concluded that a reasonable set of Cs removal technologies had been

evaluated and provided input for continued technology development.

Both the ARP/MCU and SWPF removal process will remove Sr-90 and actinides by sorption

onto MST and subsequent filtration. The use of MST sorption for Sr-90 and actinide removal

was part of the ITP process that DOE originally planned to use to treat salt waste. Each of the

four alternative Cs removal technologies discussed in the 2001 Savannah River Site Salt

Processing Alternatives Final Supplemental Environmental Impact Statement (DOE, 2001b)

was paired with MST sorption for Sr and actinide removal.

As discussed in Section 2.1, the DDA process relies on deliquification to remove Cs-137, which

is primarily dissolved in supernate, from the saltcake for subsequent treatment by the SWPF. 

DOE considered using other technologies for deliquification, including using positive pressure

and flushing. However, no alternate technologies were determined to be viable because of

safety concerns, tank space limits, or chemistry constraints. In addition, the DDA process relies

on settling to remove entrained sludge from the dissolved saltcake. In 2003, DOE evaluated

both settling and various filtration options for the removal of insoluble radionuclides present in

entrained sludge. Although DOE expects filtering to achieve approximately 100% removal of

insoluble radionuclides and expects settling to remove only approximately 66% of insoluble

radionuclides, DOE chose to use settling during DDA because DOE determined that filtration

facilities could not be constructed and become operational by the time DDA processing was

required (WSRC, 2005b).

In its response to the action items, DOE indicated that it did not target Se-79, Tc-99, or I-129 for

removal but that the solid portion (approximately 60% of Se-79, 6% of Tc-99, and 0.05% of I129) would be removed by a combination of settling and cross-flow filtration (see Section 3.5). 

DOE stated that removal of the dissolved portion of these radionuclides is impractical because

of the low technical maturity of applicable technologies. DOE also stated that they have not

conducted significant research and development activities regarding removal of these

radionuclides from tank waste (WSRC, 2005a).
- 22 -

In addition to attempting to optimize technology selection, DOE plans to limit the radiological

contribution from both the ARP/MCU and DDA interim processes by using these processes to

treat only the lowest activity tank waste. For example, DOE has indicated that, within

operational constraints, tanks with the lowest activity supernate will be selected for DDA

processing. In addition, DOE indicated that tanks that have received high-heat waste will not

be selected for DDA processing because sludge formed from high-heat waste has fission

product concentrations that are three orders of magnitude greater than fission product

concentrations in sludge formed from low-heat waste (WSRC, 2005b).

DOE evaluated several alternative technologies for processing Tank 48 waste. Options for insitu organic destruction were considered, including hydrolysis, both chemical and thermal, and

catalysis (e.g., palladium). However, the Tank 48 waste is contained in a carbon steel tank and

therefore cannot be acidified, as each of these options would require, because of the potential

for tank corrosion. DOE also considered treating the waste outside of the tanks using existing

facilities. The waste cannot be processed in the DWPF melter because the breakdown of TPB

in sufficient quantities could pose safety concerns. Construction of a separate waste treatment

unit to remove the organics was considered but was found to be extremely costly and would

require years to build and permit. DOE concluded that there are currently no proven, practical

options for removing additional radionuclides from the Tank 48 waste or treating the organics to

allow processing through the DDA, ARP/MCU, or SWPF. In the review of information

submitted in response to the RAI, NRC staff identified a report that indicated that in-tank

peroxide oxidation of Tank 48 waste may be technically and economically practical (Lambert et

al., 2005). However, a separate DOE report stated that initial corrosion testing performed to

support the technical evaluation of the process indicated that below a pH of 11, pitting corrosion

of the tank may occur. Therefore the option was not technically practical (Zapp, 2004).

3.7.2 Alternatives to the Proposed Two-Phase Approach

DOE proposes to implement the selected technologies in a two-phase, three-part approach. In

general, this approach is consistent with recommendations of the Defense Nuclear Facilities

Safety Board (DNFSB) (DNFSB, 2001) and NAS (NAS, 2000), which both suggested that DOE

should evaluate direct disposal of low activity wastes as an alternative to treating all of the

waste with the same processes. Initially, DOE’s primary consideration for removal of highly

radioactive radionuclides was that substantial amounts of tank space are required to remove

and prepare higher activity tank waste for disposal. Creation of sufficient tank space to allow

DOE to maintain sustained risk-reduction activities, including the vitrification of high-activity

sludge waste through the DWPF for eventual disposal in a federal repository for HLW, is the

primary basis for the need to apply interim processing (DDA, ARP/MCU) prior to the SWPF

coming on line in 2009. Preparation and removal of waste for disposal is expected to achieve

risk-reduction through two primary means: reduction of high-activity waste stored in old-style

tanks and reduction of the number of transfers and operations at the tank farm facility that could

result in spills or accidental exposures to workers. Reducing the amount of waste stored in oldstyle tanks is expected to reduce risk because twelve tanks without secondary containment

have a history of leakage; sufficient waste has been removed from these tanks such that there

are currently no active leak sites. In addition, reducing tank inventory is expected to reduce

worker risk because, when the tank farm is close to capacity, the number of operations and the

complexity of the sequence of operations needed to manage the waste increases substantially

(DNFSB, 2001).
- 23 -

In the RAI, NRC staff asked for additional information to support DOE’s conclusion that the use

of less efficient interim treatment measures before completion of the more efficient SWPF is

consistent with the NDAA criterion of removal of highly radioactive radionuclides to the

maximum extent practical. In response, DOE provided a comparison of risk-related metrics

between three cases: a baseline case (DDA, ARP/MCU, SWPF), a case employing ARP/MCU

and SWPF (no DDA), and a case employing SWPF only. DOE predicted the following five riskrelated metrics for each case: dose to the public, dose to the inadvertent intruder, dose to the

worker, risk in terms of old-style tank-years, and risk in terms of curie-years. Cost comparisons 

were provided, including the costs associated with slowing down DWPF waste processing. 

DOE also estimated the cost of shutting down the DWPF, but concluded that, in each case, it

was preferable to slow down DWPF waste processing to avoid DWPF shutdown. Table 4 is a

summary of the risk and cost information provided for the three cases considered. 

Although the construction of the ARP/MCU facilities and the implementation of the DDA

process will require resources, DOE expects that implementing interim treatment processes will

reduce operational costs by reducing the number of years of operation of the HLW system (i.e.,

the tank farms, DWPF, SWPF, SPF, and SDF). DOE expects that deferring treatment until the

ARP/MCU or the SWPF is operational will prolong the operation of the HLW system for two

reasons. The first is simply that waste treatment will begin at a later time. The second is that

the time required to treat waste at the SWPF will increase if interim processing is not performed

because less tank space will be available to prepare waste for SWPF treatment and it will,

therefore, take more time for the SWPF to process waste at its full capacity. The costs

associated with performing limited interim processing (ARP/MCU and SWPF only) or no interim

processing (SWPF only) are an additional $1.0 billion or $1.5 billion, respectively (2004 dollars). 

All-pathway public doses are not changed significantly because the all-pathway public doses

are dominated by the radionuclides I-129 and Se-79 and the treatment processes were not

designed to selectively remove these radionuclides. Intruder doses are lower in the cases

without interim processing or with only limited interim processing than they are in the baseline

case because, in both alternate cases, the SDF would not contain DDA waste. Intruder doses

are expected to be higher if the SDF contains DDA waste because DOE predicts that Cs-137

dominates the dose to an inadvertent intruder (Cook, et al., 2005) and DDA waste has a higher

Cs-137 concentration than either ARP/MCU or SWPF waste (d’Entremont and Drumm, 2005). 

In the limited interim treatment case and no interim treatment case, worker doses increase by

24% and 35% as compared to the baseline case, respectively. Because the tanks are aging,

there is a risk associated with the potential loss of containment. In particular, some of the oldstyle tanks have developed leaks due to stress corrosion cracking at welds. DOE has a tank

inspection program to evaluate the integrity of the tanks. In addition, where possible, the

relative humidity within the tank vaults is limited to reduce the rate of general corrosion of the

carbon steel. Because the probability of the development of leaks and the effects of DOE’s

tank inspection and maintenance programs are difficult to quantify, the risk of continued use of

the old style tanks has been quantified with a metric of old-style tank years rather than

estimates of the effects of potential leaks. This metric represents the cumulative amount of

time that tanks without full secondary containment would contain waste under each treatment

option. The limited interim processing case represents a 25% increase (to 300 tank-years)

whereas the no interim processing case represents a 42% increase over the baseline case. 

While old-style tank years is not a direct measure of risk because practices are in place to

minimize the future loss of containment of a tank or tanks, this metric does provide some

indication of the potential for risk or the need for expensive alternatives to remediate the effects

of a possible leak. The effects of each case on the life cycle of the entire tank farm are 
- 24 -

Table 4 Summary of Risk and Cost Information for Three Waste Treatment Approaches

(based on d’Entremont et al., 2005)

Metric

Baseline

(DDA, ARP/MCU,

SWPF)

Limited Interim

Processing

(ARP/MCU, SWPF)

No Interim

Processing

(SWPF)

Additional cost

(millions, in 2004

dollars)

Not Applicable 1,000 1,500

Public dose - During

operations (mrem/yr)

0.19 0.19 0.19

Intruder dose (mrem) 9 0 0

Public dose - After

operations (mrem/yr)

2.3 2.3 2.3

Worker dose - total,

all workers (rem)

890 1100 1200

Storage risk (oldstyle tank-years)240 300 340

Storage risk (curieyears)3.7x 109 4.7 x 109 5.3 x 109

quantified in terms of curie-years, which is the sum of the number of years each tank is in

service multiplied by the activity in the tank. Because waste treatment is deferred in the cases

in which no interim processing or limited interim processing is performed, the number of tank

farm curie-years increases to 25% greater than the baseline value in the limited interim

processing case and 42% greater than the base line value in the case in which no interim

treatment is performed. 

In addition to evaluating cases in which limited or no interim processing was performed, DOE

evaluated the possibility of creating the tank space necessary to support SWPF operation at full

capacity by building additional tanks. Based on estimates developed for Hanford, DOE

estimated that it would take approximately seven years and cost $300 million (2001 dollars) to

construct four new tanks at SRS (WSRC, 2005b). DOE noted the $300 million cost of creating

four new tanks would be approximately twice the remaining lifecycle cost of the ARP/MCU

facility (i.e., lifecycle costs of the ARP/MCU facility less associated costs that have already been

incurred) (d’Entremont et al., 2005). Furthermore, the $300 million cost of building new tanks

without performing interim processing does not reflect the total cost associated with deferring

processing until additional tanks could be built because it does not include costs associated

with slowing down waste processing at the DWPF when tank space becomes limiting. In

addition, DOE predicts that by the time the new tanks could be constructed and placed into

operation, the SWPF will become operational, thereby eliminating the need for additional tank

space. For these reasons, DOE concluded that building additional tanks would not be a

practical way to alleviate the tank space shortage.
- 25 -

3.8 NRC Evaluation – Selection of Treatment Processes

To evaluate whether DOE planned to remove the identified highly radioactive radionuclides to

the maximum extent practical, NRC staff evaluated information regarding technologies and

processes available for waste treatment. As described in Section 3.7.1, WSRC, under contract

to DOE in 1998, evaluated over 140 technologies to remove Cs-137 from SRS salt waste, and

both DOE-SRS and DOE headquarters reviewed the results of this investigation. In addition,

DOE headquarters requested and received input from NAS about the technologies selected

and the selection process itself. Technologies were selected on the basis of expert judgment

about the level of development of the technologies and the associated uncertainties in the

technologies' ability to meet the waste treatment goals while minimizing potential impacts to

worker safety and project schedules. NRC staff evaluated the original WSRC review of existing

Cs removal technologies, the record of decision for selection of CSSX removal, and the NAS

reviews of the technology selection and selection process, and concludes that the selection of

CSSX removal technology appears to be appropriate.

In addition to Cs-137, DOE identified Sr-90 and four alpha-emitting isotopes of Pu, Am, Np, and

Cm as highly radioactive radionuclides in the context of the Section 3116 draft waste

determination for SRS salt waste. In both the ARP/MCU and SWPF processes, as described in

Section 1.2, DOE intends to use cross-flow filtration to remove the insoluble fraction of Sr-90

and actinides and relies on sorption onto MST and subsequent filtration to remove the dissolved

fraction of Sr-90 and actinides. DOE presented little information about the selection of the MST

process. NRC staff recognizes that DOE may have based the selection of the MST sorption

process on a thorough evaluation of Sr-90 and actinide removal processes that was not

presented to NRC staff. The NAS 2000 and 2001 reviews (NAS, 2000; NAS, 2001) noted that

the MST sorption technology appeared to be promising but had potential technical deficiencies

that SRS staff were, at the time, working to resolve. In addition, the 2001 NAS report indicated

that SRS staff were considering two alternatives to MST sorption for Sr-90 and actinide

removal. Based on this information and the assumption that the high expected efficiencies of

removal of Sr-90 and actinides by MST sorption with filtration that were reported by DOE

(WSRC, 2005c) will be achieved, NRC staff concludes that the use of MST sorption with

subsequent filtration to remove Sr-90 and actinides appears to be appropriate.

As noted in Section 3.5, the ARP facility can be used in a "filtration only" mode in which MST

strikes are not used. DOE originally proposed to use MST as part of the ARP/MCU process

only if additional actinide removal was necessary for the grouted waste to meet Class C limits

(DOE, 2005). In response to a RAI from NRC asking for information to support the conclusion

that using MST sorption only if required to meet Class C limits was consistent with removal of

highly radioactive radionuclides to the maximum extent practical, DOE indicated MST sorption

would be used for all batches of waste processed by the ARP/MCU procedure as long as the

necessary processing throughput could be maintained (WSRC, 2005b). In addition, DOE

indicated that criteria for the application of MST strikes would be developed as operational

experience with the ARP/MCU process was gained and effects on operational throughput were

better known. NRC staff believes it is appropriate to specify criteria for the application of MST

sorption strikes as operational experience is gained because operational information is

necessary to be able to compare the risks and costs of delaying treatment to the benefits of the

additional amount of actinide removal that is accomplished with MST strikes. NRC staff
- 26 -

concludes that operation of ARP/MCU without MST strikes will be acceptable if required by

processing throughput requirements because (1) some actinide removal will be accomplished

by filtration without MST strikes, (2) a relatively small amount of waste is expected to be

processed by ARP/MCU as compared to the amount of waste to be processed by the SWPF,

and (3) the use of MST is not necessary for consistency with DOE's evaluation of the class of

the waste or the SDF's ability to meet the performance objectives because these outcomes

were calculated based on the assumption that MST strikes would not be used during ARP/MCU

treatment.

As discussed in Section 3.7.1, DOE considered technologies for deliquification and solids

separation to be used in the DDA process. In addition, DOE presented information about the

selection of tanks to be treated with DDA before ARP/MCU and the SWPF are operational. 

NRC evaluated information regarding the technologies that will be used in the DDA process as

well as the criteria used to select tank waste that will be treated with DDA. Because the DDA

process is the only proposed process that does not use filtration and is, therefore, not expected

to achieve approximately 100% removal of insoluble radionuclides, it is the only process for

which the removal efficiencies will be affected significantly by the amount of sludge entrained in

dissolved salt waste. Because the predicted radionuclide removal efficiencies of the ARP/MCU

and SWPF demonstrate that removal of Cs-137, Sr-90, and actinides can be accomplished to a

greater degree with different treatment technologies, the use of technologies used in the DDA

process must be considered in terms of DOE's two-phase, three-part process. NRC staff

concludes that the selection of waste with low supernate activity and relatively small volumes of

low-heat waste sludge, in addition to the treatment of a relatively small amount of waste with

the DDA process, are important to the acceptability of the DDA interim process. NRC staff

notes that, because a relatively small amount of waste will be treated by DDA, the overall

removal of most radionuclides from the salt waste is driven by removal in the SWPF (Table 3). 

NRC staff also notes that, while the removal of soluble radionuclides that are significantly

removed by downstream processes (e.g., Cs-137) is affected by the extent of deliquification

achieved during DDA processing, the removal of other soluble radionuclides is not affected by

the degree of deliquification because, whether or not they are separated from the saltcake and

processed at the SWPF, they ultimately are sent to the SDF (e.g., Tc-99, I-129). 

In response to the action items (WSRC, 2005a), DOE indicated that it did not target Se-79, Tc99, or I-129 for removal because DOE predicted those radionuclides would cause a small dose

to the public. However, as discussed in Section 3.4, the NRC staff has concluded that DOE

has provided insufficient support for its base case to be able to discount the potential dose

contributions of Se-79, Tc-99, and I-129 (also see Section 4). NRC concludes that DOE’s

decision to include Se-79, Tc-99, and I-129 as highly radioactive radionuclides is appropriate

and that Criterion Two requires removal of these radionuclides to the maximum extent practical. 

DOE concluded that removal of the dissolved phase of Se-79, Tc-99, or I-129 would be

impractical because treatment technologies were immature. To support the conclusion that no

economically practical alternatives were available to remove Se-79, Tc-99, or I-129 from salt

waste, DOE cited an evaluation of the feasibility of removal of various radionuclides from

Hanford site tank waste that was completed in 1996 (WHC, 1996). The cited document

identified technologies to remove both dissolved Se-79 and Tc-99 that had been tested at the

bench scale, but omitted at least one technology that had successfully removed Tc-99 from

simulated and actual decontaminated SRS salt waste in bench-scale tests (WSRC, 1994). The

cited document indicated that, at the time it was completed, there were no technically practical

technologies available to remove Se-79, Tc-99, or I-129 from Hanford tank waste. Based on
- 27 -

the information provided by DOE, and DOE’s statement that it has not performed significant

research and development activities regarding removal of these radionuclides, NRC concludes

that implementation of additional treatment technologies would not be practical within the

constraints imposed by DOE’s proposed schedule. Because DOE indicated that it did not

target Se-79, Tc-99, or I-129 for removal, NRC's conclusion that additional removal would not

be practical is not based on DOE's assessment of existing technologies, but rather on

consideration of the impacts of delays to the proposed schedule that would result from changes

to the treatment process design to implement applicable technologies. As discussed in Section

3.7.2, DOE has indicated that delays to the proposed schedule would cause significant

increases in project cost and worker risk. NRC concludes that the proposed removal of Se-79,

Tc-99, and I-129 is consistent with removal to the extent practical because, even if DOE were

aware of technologies that could remove dissolved Se-79, Tc-99, or I-129 from salt waste,

delays associated with implementing these technologies at this point in the design of the

treatment processes would cause unjustified increases in project cost and worker risk. 

Because the conclusion is based on the impacts of delays caused by implementation of

additional treatment technologies, any changes to the proposed schedule that would allow

applicable technologies to be developed or implemented without causing additional delays

would be expected to impact this conclusion. 

The waste that will receive the least amount of treatment before being sent to SDF is the waste

in Tank 48. Although the Tank 48 waste represents only 0.3% of the volume of the salt waste, it

represents 27% of the activity projected to be disposed of in the SDF. DOE was unable to

develop a practical option for treatment of the Tank 48 waste, other than directly processing it in

the SPF and disposing of the resultant material in the SDF. Because Tank 48 contains a

relatively small amount of activity (3 x 1016 Bq [0.8 MCi]) compared to other tanks that could be

emptied to create as much space, disposing of Tank 48 waste in the SDF creates a relatively

large amount of new-style tank space per Ci sent to SDF. DOE evaluated alternatives for

treating other tank wastes with DDA processing and found that creation of a comparable

amount of tank space by treating waste in an alternate tank with DDA processing would result

in a greater activity of waste being sent to the SDF. In addition, because of its location, Tank

48 is the planned feed tank for the ARP/MCU process and is an integral part of the plan for

staging feed for the SWPF. Based on the following factors, NRC staff concludes that there is

reasonable assurance that Tank 48 waste has been processed to the maximum extent

practical: (1) the Tank 48 waste represents a small fraction (0.4% by activity) of the total waste

in or projected for the tank farm facilities, (2) other treatment options were considered and were

determined to not be technically practical, (3) Tank 48 is a new style tank containing a relatively

small amount of waste (3 x 1016 Bq [0.8 MCi]) and has a strategic location with respect to tank

farm operations and, (4) DOE estimated that the saltstone resulting from processing Tank 48

waste would result in a wasteform that meets Class C concentration limits (see Section 4.1.1). 

In addition to evaluating the selection of processes to remove radionuclides from salt waste,

NRC staff also evaluated information regarding alternatives to DOE's proposed two-phase,

three-part approach. Because SWPF will be able to remove a greater fraction of most highly

radioactive radionuclides than either DDA or ARP/MCU, NRC requested information to support 

DOE's conclusion that using DDA and ARP/MCU rather than treating all of the waste with the

SWPF is compatible with the criterion for removal of highly radioactive radionuclides to the

maximum extent practical. 
- 28 -

In its draft waste determination, DOE stated that removal of radionuclides "to the maximum

extent practical" will vary depending not only on technologies but also the overall costs and

benefits, and allows consideration of factors such as environmental, health, timing or other

exigencies, considerations that may ensue from delay of waste removal, and reasonable

availability of proven technologies. NRC staff agrees that multiple factors can and should be

considered when weighing what is the maximum extent practical for any given situation. In the

1993 petition denial establishing its WIR criteria, the NRC stated that, with regard to

decommissioning of reprocessing facilities, both economic factors and technical factors could

be considered, as long as there is adequate protection of public health and safety. In previous

incidental waste reviews performed for DOE, NRC reviewed information on monetary costs as

well as worker doses (NRC, 2003a). In its RAI for this draft waste determination, staff

requested additional quantitative information to support DOE's assertion that the proposed

approach would remove highly radioactive radionuclides to the maximum extent practical.

In response to the RAI, DOE performed an analysis of the economic and mission impacts of

alternative approaches in which waste processing would be postponed until either the

ARP/MCU processes or the SWPF becomes operational. DOE analyzed a "limited interim

processing" case in which the DDA process would not be used and a "no interim processing"

case in which neither DDA nor ARP/MCU interim processes would be used.

Because the radionuclide removal efficiencies achieved by the interim processes are lower than

the predicted radionuclide removal efficiencies that will be achieved by the SWPF, use of

interim processes will result in higher radionuclide concentrations and total activity in the SDF

than cases in which waste treatment is delayed until the SWPF becomes operational. The

increased future risk to the public due to the disposal of greater quantities of highly radioactive

radionuclides in the SDF must be evaluated in the context of the risks, costs, and mission

impacts associated with cases in which interim processing is not performed. There are two

types of risks associated with deferring waste processing until the SWPF becomes operational. 

The first is the risk to the public associated with the possibility that a leak will develop in an

old-style tank. It is difficult to estimate the risk to the public associated with a leak in a HLW

tank because it is difficult to predict the probability that a leak will develop, and because, if a

leak were to be detected, it is likely that mitigative measures would be used to limit the risk to

the public. The second type of risk is risk to facility workers. Cumulative radiological risk to

facility workers is expected to increase if waste treatment is deferred because the tank farm

would operate for a longer period of time (Table 4) (d'Entremont et al., 2005). In addition, if

less space is available in the tank farm to prepare waste for DWPF processing, radiological risk

to workers is expected to increase because the number of tank-to-tank transfers necessary to

maintain operation of the DWPF would increase (DNFSB, 2001). 

Thus, the primary considerations as to whether employing interim processing technologies that

are less efficient while waiting for the more efficient technologies to become available in

achieving removal of highly radioactive radionuclides to the maximum extent practical are cost

and risk transferral. Employing the interim processing is expected to reduce worker risk, while

increasing future risks to an inadvertent intruder (Table 4). The risk to a member of the public

that does not intrude on the site is not expected to change if waste treatment is deferred until

the SWPF is operational (Table 4) because the predicted dose to a member of the public in

DOE's base-case PA is dominated by Se-79 and I-129 (see Sections 4.2.15 and 4.2.16), which

are expected to be removed to a similar extent by the interim processes and by the SWPF. It is
- 29 -

difficult to compare worker risk to public risk because worker risk is accepted by the individual

while public risk is imposed. 

The additional costs associated with deferring waste treatment until the SWPF becomes

operational are easier to quantify. As discussed in Section 3.7.2, DOE expects that deferring

treatment until the ARP/MCU or SWPF become operational will increase the total costs of

treating the waste primarily because of the added operational costs of operating the HLW

system for additional time. DOE predicts that the cost associated with the baseline option is

significant ($4.9 billion in 2004 dollars), but that the cost associated with performing only limited

interim processing (i.e., use of ARP/MCU but no use of DDA) ($5.8 billion in 2004 dollars) and

the costs associated with postponing waste treatment entirely until the SWPF becomes

operational ($6.3 billion in 2004 dollars) would be even greater (Table 4). Therefore, if DDA

processing is not used, dose to an inadvertent intruder is expected to decrease by 0.09 mSv/yr

(9 mrem/yr) at a cost of approximately $1.0 billion (2004 dollars) while worker risk is increased

by 210 person-rem. If no interim processes are used, dose to an inadvertent intruder is

expected to decrease by 0.09 mSv/yr (9 mrem/yr) at a cost of approximately $1.5 billion dollars

(2004 dollars) while worker risk is increased by 310 person-rem. Because the NRC staff

believes there is reasonable assurance that worker and public performance objectives will be

satisfied if DDA and ARP/MCU are used to process the volume of low-activity waste described

in the draft waste determination and supporting documents (d'Entremont and Drumm, 2005),

NRC staff concludes that expenditure of an additional $1.0 billion (2004 dollars) to $1.5 billion

(2004 dollars) to decrease public exposures further below the performance objectives would not

be warranted (see Sections 4.2.16 and 4.2.18).

Use of the interim processes to treat more waste or higher activity waste than proposed would

be expected to increase the inventory of highly radioactive radionuclides in the SDF, which

would be expected to affect the demonstration that the performance objectives can be met. 

Because the demonstration that the use of the interim processes is consistent with the criterion

of removal of radionuclides to the maximum extent practical depends in part on the use of the

interim processes to treat a relatively small volume of the lowest activity waste, use of the

interim processes to treat more or higher activity waste than proposed would be expected to

impact whether Criterion Two can be met.

Because DOE predicts the primary worker risk reduction and cost savings achieved by

implementing interim waste treatment processes result from reducing the number of years of

operation of the HLW system (d'Entremont et al., 2005), changes to the proposed waste

treatment schedule would be expected to change the cost/benefit analysis used to support the

conclusion that highly radioactive radionuclides will be removed from the salt waste to the

maximum extent practical. In addition, because schedule constraints were cited as the primary

reason for some of DOE’s technology selections, it appears that changes to the proposed

schedule could affect DOE’s choice of salt waste treatment technologies.

3.9 NRC Review and Conclusions (Criterion Two)

The NRC staff concludes that Criterion Two will be met for SRS salt waste. This conclusion is

based primarily on the NRC staff’s four intermediate conclusions, as follows: (1) DOE used an

appropriate process to identify highly radioactive radionuclides in the context of the Section

3116 draft waste determination for SRS salt waste (Section 3.4), (2) relatively high removal
- 30 -

efficiencies are predicted for most of the highly radioactive radionuclides in SRS salt waste

(Section 3.6), (3) appropriate processes were used to evaluate technologies for the treatment of

Tank 48 waste and additional removal of radionuclides from the waste is not technologically or

economically practical (Sections 3.7.1 and 3.8), and (4) the additional costs and radiological

risks to workers associated with delaying salt waste treatment until the ARP/MCU or SWPF is

operational do not appear to be justified by the small reduction in the risk to an inadvertent

intruder and the public that could be achieved by delaying treatment (Section 3.8). As

discussed in Section 3.8, the NRC staff agrees with DOE’s position that it is appropriate to

consider multiple factors when evaluating the practicality of radionuclide removal in the context

of Criterion Two. For example, although DOE predicts that very low efficiencies of removal of I129 and Tc-99 will be achieved, the NRC staff concludes that the removal of these

radionuclides is not inconsistent with Criterion Two because implementation of additional

treatment technologies was determined not to be practical within the constraints of DOE’s

proposed schedule. Similarly, NRC staff considered DOE’s need to create space in new-style

tanks when evaluating DOE’s selection of waste treatment technologies and its proposed twophase, three-part approach. The primary benefit of using interim processes, as identified by

DOE, is that these processes will create enough compliant tank space to support operation of

the SWPF at close to its full capacity when it becomes operational. Significant changes in the

proposed schedule would be expected to impact the evaluation of which technologies could be

implemented in time to create the tank space necessary to support SWPF start up, as well as

the relative costs and benefits of implementing various treatment technologies. In addition, the

NRC staff notes that the conclusion that the proposed two-phase, three-part approach is

consistent with Criterion Two is based in part on the use of the interim processes to treat a

relatively small volume of the lowest activity waste. Use of the interim processes to treat more

waste or higher activity waste than proposed would be expected to affect both the

demonstration that the performance objectives can be met and the demonstration that Criterion

Two can be satisfied. 

NRC staff used the following assumptions in assessing conformance with Criterion Two:

1. The DDA and ARP/MCU interim treatment processes will not be used to treat significantly

more waste or higher activity waste than proposed (see Section 3.8).

2. The actual radionuclide removal efficiencies of the ARP/MCU and SWPF treatment

processes will meet or exceed the reported lower bounds of the projected removal

efficiencies (see Sections 3.6 and 3.8).

3. During the DDA process, dissolution of saltcake and solids settling will be performed such

that the amount of sludge entrained in salt waste after DDA processing does not exceed

DOE’s estimate of 200 mg/L salt waste, which was identified by DOE as being a

conservative estimate (WSRC, 2005c) (see Section 3.6).

4. During DDA processing, deliquification will be performed such that the lower bound of

predicted deliquification (30%) (WSRC, 2005c) is met or exceeded (see Section 3.6).

5. Financial data used to support DOE’s cost estimates are reasonable (see Section 3.8).

6. The proposed dates of operation of the ARP/MCU and SWPF facilities will be met (see

Section 3.8).
- 31 -

7. The Tank 48 waste can be processed safely in the SPF, and the wasteform will meet the

Class C limits and perform as expected (see Section 3.8).

NRC staff concludes the following with respect to Criterion Two:

1. DOE’s conclusion that highly radioactive radionuclides will be removed to the extent

practical by the proposed two-phase, three-part process, including the proposed process for

the management of Tank 48 waste, is reasonable. 

4. CRITERIA THREE (A) AND THREE (B)

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

61.55 of title 10, Code of Federal Regulations, and will be disposed of--

(i) in compliance with the performance objectives set out in subpart C of part

61 of title 10, Code of Federal Regulations; and

(ii) pursuant to a State-approved closure plan or State-issued permit,

authority for the approval or issuance of which is conferred on the State

outside of this section; or

(B) exceeds concentration limits for Class C low-level waste as set out in section 61.55 of

title 10, Code of Federal Regulations, but will be disposed of--

(i) in compliance with the performance objectives set out in subpart C of part

61 of title 10, Code of Federal Regulations;

(ii) pursuant to a State-approved closure plan or State-issued permit,

authority for the approval or issuance of which is conferred on the State

outside of this section; and

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

Commission

Whether Criterion 3(A) or Criterion 3(B) is applicable depends on the class of the waste being

assessed in the waste determination. In the case of the draft waste determination for salt

waste disposal at SRS, Criterion 3(A) is applicable because the final waste form disposed of in

the SDF is not expected to exceed Class C concentrations (see Section 4.1). Therefore,

Criterion 3(B) is not evaluated in this TER. NRC staff also did not assess 3(A)(ii) because the

State of South Carolina has authority over State-approved closure plans and State-approved

permits.

Criterion 3(A) refers to the performance objectives of 10 CFR Part 61, Subpart C, which require

assessment of protection of the general population from releases of radioactivity, protection of

individuals from inadvertent intrusion into the waste, protection of individuals during operations,

and evaluation of the stability of the disposal site after closure. Protection of the general

population (including intruders) is typically evaluated through a PA calculation that takes into

account the relevant physical processes and the temporal evolution of the system. The NRC

staff’s assessment of DOE’s PA is presented in Section 4.2.
- 32 -

4.1 Assessment of Waste Classification

(3)(A) (The waste) 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, 

4.1.1 Waste Classification

10 CFR 61.55 classifies low-level radioactive waste intended for near surface disposal as Class

A, B, or C. Table 5 provides a comparison of the radiological composition of the average waste

resulting from the DDA, ARP/MCU, and SWPF processes with the Class C concentration limits

found in Tables 1 and 2 in 10 CFR 61.55. The Class C concentrations in Tables 1 and 2 in 10

CFR 61.55 were not provided for every radionuclide, only the dominant radionuclides expected

in a commercial low-level waste stream. The saltstone wasteform will contain a mixture of longand short-lived radionuclides, and therefore 10 CFR 61.55 (a)(5) was applied to determine

waste classification. 10 CFR 61.55 (a)(5) states:

“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 determined by the concentration of the radionuclides

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.”

The waste treated through interim processing and the SWPF is estimated to meet the

concentration limits established in 10 CFR 61.55 for Class C waste as shown in Table 5. The

wasteform produced from waste processing is a formulation of cement, fly ash, and blast

furnace slag mixed with waste in an approximate ratio of 1:1 to produce a solidified wasteform

to achieve stability of the waste for disposal (see Section 4.2.6). The waste and solidifying

materials will be mixed during processing to achieve reasonable homogeneity. The NRC

Branch Technical Position on concentration averaging is being applied by DOE to calculate the

average concentration of the final wasteform (NRC, 1995). 

4.1.2 NRC Evaluation - Waste Classification

In developing the waste classification system in 10 CFR 61, consideration was given to longlived radionuclides whose potential hazard will persist long after institutional controls, improved

waste form, and deeper disposal have ceased to be effective and short-lived radionuclides for

which the requirements on institutional controls, waste form, and disposal methods are

effective. Tables 1 and 2 of 10 CFR 61.55 provide the concentration limits for classifying lowlevel radioactive waste intended for near surface disposal as Class A, B, or C. The waste

classification system was originally developed primarily to facilitate waste generators’

assessment as to whether their material was suitable for packaging and transport to a waste

disposal facility, and secondarily to identify technical requirements for a waste disposal facility

developer. The 10 CFR 61 classification system is based on the concentration of radionuclides 
- 33 -

Table 5 Comparison of DDA, ARP/MCU, and SWPF Waste with 

 Class C Concentration Limits [d’Entremont and Drumm, 2005](1) 

Radionuclide

Class C

Limit (Ci/m3

)

DDA Waste

(Ci/m3)

ARP/MCU Waste

(Ci/m3)

SWPF Waste

(Ci/m3)

Long-Lived

C-14 8 0.001 0.001 0.0008

Tc-99 3 0.006 0.03 0.06

I-129 0.08 0.000004 0.00001 0.00003

Alpha Emitting

Transuranic (TRU)

nuclides with half-life

greater than 5 years (2)

100 41 2 <10

Pu-241 (2) 3500 2 0.1 7

Cm-242 (2) 20000 0.00005 0.00008 0.0001

Short-Lived

Ni-63 700 0.002 0.0009 0.0002

Sr-90 7000 0.1 0.02 0.002

Cs-137 4600 22 8 0.005

(1) Values in table are different from those in the draft waste determination (March 2005) due to updated values

provided in d’Entremont and Drumm (June 2005).

(2) Units are nanocuries per gram

in waste and is intended to limit exposures to the inadvertent intruder, whereas protection of the

general population is addressed via limits on the total radionuclide activity disposed of at a 

facility. It is expected that a waste disposal facility would provide a demonstration of how its

disposal facility design can achieve the Subpart C performance objectives for the expected

waste stream. It would not be unexpected that there may be radionuclides, not listed in Tables

1 or 2, that are present in a waste stream that cause a significant risk. Likewise, because

different disposal techniques or systems are possible, waste that does not meet Class C

concentration limits can be determined to be suitable for near surface disposal by applying the

alternate waste classification criteria found at 10 CFR 61.58. Therefore satisfaction of the

performance objectives is a necessary element to achieving safety whether Class C limits are

satisfied or not (see Section 4.2.10). 

Table 5 shows the expected average radionuclide concentrations for the DDA, ARP/MCU, and

SWPF waste streams. The waste is expected to contain both short- and long-lived

radionuclides, therefore application of 10 CFR 61.55 (a)(5) is appropriate for determining waste

classification. The sum of fractions for the long-lived radionuclides is 0.41, 0.03, and 0.10 for

the DDA, ARP/MCU, and SWPF waste streams, respectively. The sum of fractions for the

short-lived radionuclides is 0.005, 0.002, and 2 x 10-6 for the DDA, ARP/MCU, and SWPF,

respectively. The long-lived radionuclides are more limiting in defining the waste classification,

although all waste streams are expected to be Class C or lower. Although not expected to 
- 34 -

occur, uncertainty in radionuclide concentrations could result in some batches of DDA waste

not meeting Class C concentration limits (see Section 3.2).

DOE has appropriately applied the NRC Branch Technical Position on concentration averaging

for the saltstone wasteform (NRC, 1995). The wasteform produced from waste processing is a

formulation of cement, fly ash, and blast furnace slag mixed with waste in an approximate ratio

of 1:1 to produce a solidified wasteform to achieve stability of the waste for disposal. Based on

the information provided by DOE, the NRC staff agrees that the waste and solidifying materials

will be mixed during processing to achieve reasonable homogeneity. Although there is

uncertainty in the characterization data, the NRC staff believes the waste treated through

interim processing and the SWPF will likely meet the concentration limits established in 10 CFR

61.55 for Class C waste as shown in Table 5. 

4.2 Performance Assessment to Demonstrate Conformance with Performance Objectives 

Typically a PA is developed to demonstrate whether the performance objectives have been

met. A PA is a quantitative evaluation of potential releases into the environment and the

resultant radiological doses. Depending on the computational tools used by the analyst, the PA

may be a single integrated model, or it may represent an analysis approach for integrating and

evaluating a collection of other models. Performance assessments involve the integration of

process models to identify and propagate impacts and uncertainties between models. Process

models are used to evaluate physical and chemical phenomenon such as release of

radionuclides from wasteforms, degradation of engineered components, and transport of

radionuclides through environmental media, among others. The PA documentation will

commonly provide the justification for the data used, a description of the models used,

verification of and support for the models, and an evaluation of the impact of data and model

uncertainty. To evaluate uncertainty, a variety of techniques typically are used, including

deterministic analysis with sensitivity analysis, and probabilistic analysis with uncertainty and

sensitivity analyses. The sensitivity analysis may then be used to conduct a risk-informed

evaluation through the in-depth review of those parameters and processes most important to

system performance in relation to meeting the performance objectives. The following TER

sections provide a discussion of the main components of DOE’s PA to demonstrate compliance

with the performance objectives for the saltstone waste disposal facility.

In general, all approaches to PA calculations (e.g., deterministic, probabilistic) have their

advantages and disadvantages. A deterministic approach can be very valuable when the

analysis is clearly conservative because it makes the demonstration of meeting the

performance objectives more straightforward and it can be significantly easier to interpret

results and explain them to stakeholders. While deterministic analysis can be a suitable

methodology for PA, it can also present a challenge for a system that responds in a highly

nonlinear fashion with changes in the independent variables. In addition, when there are

numerous inputs (e.g., data or models) that are uncertain, the evaluation of the impacts of the

uncertainties on the decision can be difficult. Typical one-off type of sensitivity analysis where a

single parameter is increased or decreased will only identify local sensitivity within the

parameter space, such that it may not clearly identify the risk implications. A probabilistic

approach can have distinct advantages when there are a number of uncertainties that may

significantly influence the results of a PA.
- 35 -

The term “conservatism,” as used with respect to PA, is a relative term that needs to be placed

within the proper context. Conservatism is typically defined mainly with respect to what is

known or sometimes with standard practices that have been demonstrated to yield acceptable

performance. In this regard, model support (i.e., information that supports the results of a

model) of process model results plays a key role in developing confidence in the output of PA

calculations. Because of the long time periods involved with the analysis, PA models cannot be

validated in the traditional sense. However, multiple methods for developing confidence in the

model projections can be used, including: laboratory experiments, alternative modeling

approaches, field measurements, natural analogs, and expert elicitation, among others. The

amount of model support provided should be commensurate with the risk reduction being

provided by the natural and engineered system. Multiple lines of evidence are strongly

encouraged when the risk reduction of the systems being evaluated is large. 

4.2.1 Performance Assessment Overview

The Radiological Performance Assessment (Cook and Fowler, 1992) for the SDF was originally

prepared to fulfill the requirements of DOE Order 5820.2A in 1992. This DOE order has since

been superseded by DOE Order 435.1, Radioactive Waste Management. The 1992 PA was

followed by an addendum in 1998 (Fowler, 1998) to address and incorporate comments from

the DOE Performance Assessment Peer Review Panel and DOE headquarters. In 2002, a

special analysis (Cook, et al., 2002) was performed to update the 1992 PA based on new

information from the construction phase of the SDF and to provide the compliance

determination for DOE Order 435.1 (DOE, 1999a, 1999b, 1999c). Using what it considered the

latest information on the SDF feed solutions, updated modeling methods, and updated closure

cap design and evaluations, DOE developed the 2005 special analysis (Cook, et al., 2005) as

the latest iteration in the PA process. As stated by DOE in Rosenberger, et al. (2005), the 2005

special analysis supplements the analyses in the 1992 PA and supersedes the analyses of the

2002 special analysis.

To evaluate the long-term performance of the SDF in the 1992 PA, DOE developed site-specific

conceptual models to consider exposure pathways and scenarios of potential importance,

potential releases from the facility to the environment, and potential doses to a receptor as a

result of exposure to radioactive contaminants. In the 1992 PA, DOE assumed that the

inventory (source term) was composed of the decontaminated salt solution stream from the InTank Precipitation Facility and the concentrate stream from the Effluent Treatment Process

(Cook and Fowler, 1992). The 1992 PA evaluated waste material in SDF Vaults 1 and 4 and

projected the results for up to 14 future vaults. The groundwater pathway was the primary

quantitative analysis for post-closure performance, and a qualitative evaluation concluded that

the air pathway was insignificant in comparison. The resident farmer intruder scenario was

evaluated qualitatively in the 1992 PA. The PORFLOW code developed by Analytical and

Computational Research Inc. was the primary computer code used to evaluate the groundwater

pathway (Runchal and Sagar, 1991). PORFLOW was used to simulate groundwater flow and

contaminant transport and to calculate maximum concentrations of radionuclides in

groundwater at locations beyond the 100-m (328-ft) buffer zone used in the 1992 PA. 

Estimates of dose from direct ingestion of contaminated groundwater were then made by

multiplying the radionuclide concentrations by a radionuclide-specific dose conversion factor of

annual dose per unit concentration. The 1992 PA concluded that the groundwater dose

calculations were well within the performance limits of DOE and therefore an all-pathways
- 36 -

evaluation was not completed because DOE had reasonable assurance that the all-pathways

dose would be within the performance limits (Rosenberger, et al., 2005).

The 2005 special analysis (Cook, et al., 2005) was intended to evaluate the SDF against the

specific performance objectives of DOE Order 435.1 and 10 CFR Part 61, Subpart C in an allpathways analysis. The 2005 special analysis also was intended to update the 1992 PA by

using the current estimates of total inventory expected to be disposed of in Vault 4 and to

recalculate the groundwater and air transport exposure pathways (Rosenberger, et al., 2005). 

The all-pathways analysis included residential and agricultural pathways (i.e., resident farmer

scenario) and sensitivity evaluations. The 2005 special analysis did not alter the 1992 PA

conceptual model of the migration of radionuclides from the saltstone wasteform in Vault 4 to

the environment via diffusion and advection (Cook, et al., 2005).

The 2005 special analysis included 45 radionuclides (and their progeny) selected with a

screening procedure recommended by the National Council on Radiation Protection and

Measurements (NCRP) (NCRP, 1996), but only addressed the inventory projected for Vault 4. 

The PA methodology is summarized as follows. The groundwater pathways analysis used a

sequence of three computer codes. The Hydrologic Evaluation of Landfill Performance (HELP)

code [U.S. Environmental Protection Agency (EPA), 1994a,b] was used to calculate water

infiltration over time through the upper geosynthetic clay layer. The PORFLOW code

(Analytical and Computational Research, Inc., 2002) was then used to calculate groundwater

transport over time below the upper geosynthetic clay layer and in the saturated zone, as well

as to calculate dissolved radionuclide fluxes to and within the saturated zone to the 100-m

(328-ft) well location. The third step in the process uses the LADTAP XL© spreadsheet code

(Simpkins, 2004) to calculate doses to the resident farmer from contaminated groundwater use

at 100 m (330 ft) down-gradient through a number of residential and agricultural exposure

pathways.

The air pathways exposure analysis was performed independently of the groundwater pathway

analysis and used a sequence of two computer codes. The PORFLOW code (Analytical and

Computational Research, Inc., 2002) was used to calculate the gaseous radionuclide diffusion

from the vault to the ground surface. The EPA Clean Air Act Assessment Package-1988

(CAP88) code (EPA, 2002) was then used to calculate airborne transport and the resultant

dose from possible air pathways (e.g., direct plume shine, inhalation, and ingestion of

vegetables, meat, and milk exposed to the airborne radioactivity). Eight radionuclides were

considered for the air pathway analysis (Rosenberger, et al., 2005). Doses from the air and

groundwater pathways were then summed for each radionuclide to provide an all-pathways

dose for comparison to the 10 CFR Part 61, Subpart C, performance objective.

Inadvertent intruder analysis was conducted quantitatively in the PA presented in the 2005

special analysis. A resident intruder scenario was evaluated using a software tool developed by

DOE specifically for the inadvertent intruder evaluation at the Savannah River Site (Koffman,

2004). The software tool calculated the radionuclide-specific concentrations that would be

encountered by an intruder by disregarding leaching and only considering decay for the amount

of contaminant remaining at the site at the time of intrusion (Cook, et al., 2005).
- 37 -

4.2.2 Source Term

The source term, as defined here, is the radiological composition and quantity of material to be

disposed of in the SDF. DOE estimates that the salt waste treatment and processing strategy

will result in 6.27 x 108

 L (170 Mgal) of grout containing 1.1 x 1017 Bq (3 MCi) of radioactivity to

be disposed of in the SDF. The SDF has established WAC to ensure that the concentrations of

waste accepted by the facility is suitable for disposal. The saltstone facility WAC establishes

the radionuclide limits for the waste. A key element of DOE’s approach to managing

uncertainties associated with development of the source term for the PA calculations is to

sample the material sent to the feed tank prior to transfer to the SPF.

In response to the RAI, DOE provided the sampling plan for the saltstone feed tank (Ketusky,

2005). The plan was developed prior to and during the development of the WAC. DOE intends

to develop a Waste Compliance Plan (WCP), in which a detailed sampling plan will be finalized. 

Tank 50 is envisioned to be the receipt tank for waste to be transferred to the SPF. Possible

waste stream influents include but are not limited to: Effluent Treatment Process (ETP)

effluents, DWPF recycle, DDA solution, and post treatment dissolved salt solution from the ARP

and MCU. Three main types of samples are to be obtained from Tank 50, namely: DDA pull

sample, Grout pull sample, and operations (OPS) pull sample. Samples are to be taken from

Tank 50 for a variety of uses, including: vault classification, grout qualification and formulation,

verification that the grout will be nonhazardous, verification that the grout will meet processing

and quality requirements, and compliance with liquid/solid chemistry and radiological contents

requirements. The term pull sample is used to describe a bulk sample of up to 4 liters of

material that then may be used for a variety of analyses. Sample frequency is from 1 per 5

years for a DDA pull sample to quarterly for Grout and OPS pull samples.

Material balance will be used to maintain the waste in Tank 50 compliant with the WAC. The

material to be transferred to the SPF will be sampled prior to transfer to Tank 50 (as

applicable). The characterization of the material transferred to the SPF will be maintained by

the material balance. DOE indicated that process knowledge can be used in numerous aspects

of the sampling strategy. In particular, based on the proposed permit with the State of South

Carolina, WAC samples would not need to be taken from Tank 50 prior to sending material to

the SPF. 

The source term used in DOE’s current PA calculations is based on estimates of the projected

radiological inventory. The source term estimates have undergone considerable modification

since the original PA was developed in 1992 as a result of new characterization information,

new techniques to calculate concentrations of some radionuclides, and a new processing

strategy to treat the waste. Table 6 provides the estimated total inventory (in curies) to be

disposed of in the SDF for each radionuclide (Rosenberger et al., 2005, pg. 17). The total

volume of material to be disposed of is 6.4 x 105

 m3 (2.26 x 107 ft3). Section 3.1 provides a

description of the approach to estimating the waste composition prior to disposal in the SDF.

DOE’s approach to ensuring that the waste can achieve the performance objectives is to use

the PA to develop inventory limits. The expected waste inventory (product of volume and

radiological concentration of the waste) is compared to the inventory limits and a sum-offractions is calculated. If the sum-of-fractions is less than one, then the waste can satisfy the

performance objective. As indicated above, sampling of the waste feed material prior to
- 38 -

Table 6 Estimated Source Term for the Saltstone Disposal Facility

Radionuclide Inventory (Ci) Radionuclide Inventory (Ci)

H-3 9.43 x 103

 Ra-228 1.04 x 10-1

C-14 5.20 x 102

 Ac-227 1.91 x 10-5

Na-22 5.05 x 103

 Th-229 7.53 x 10-3

Al-26 2.35 x 101

 Th-230 3.53 x 10-2

Ni-59 2.85 Pa-231 5.32 x 10-5

Co-60 1.10 x 102

 Th-232 1.04 x 10-1

Ni-63 2.51 x 102

 U-232 3.09 x 10-2

Se-79 8.94 x 101

 U-233 2.22 

Sr-90 7.43 x 103

 U-234 7.72 

Y-90 7.43 x 103

 U-235 1.35 x 10-1

Nb-94 4.22 x 10-3 U-236 3.03 x 10-1

Tc-99 3.31 x 104

 U-238 5.19 

Ru-106 2.28 x 103

 Np-237 2.12 

Rh-106 2.28 x 103

 Pu-238 1.36 x 104

Sb-125 9.24 x 103

 Pu-239 6.55 x 102

Te-125m 2.26 x 103

 Pu-240 1.75 x 102

Sn-126 4.51 x 102

 Pu-241 7.03 x 103

Sb-126 6.30 x 101

 Pu-242 1.81 x 10-1

Sb-126m 4.50 x 102

 Am-241 9.50 x 101

I-129 1.80 x 101

 Am-242m 5.27 x 10-2

Cs-134 2.71 x 103

 Pu-244 7.96 x 10-4

Cs-135 4.67 Am-243 2.18 x 10-2

Cs-137 1.35 x 106

 Cm-242 1.05 x 10-1

Ba-137m 1.28 x 106

 Cm-243 2.67 x 10-2

Ce-144 6.27 Cm-244 8.72 x 101

Pr-144 6.27 Cm-245 8.58 x 10-3

Pm-147 4.14 x 103

 Cm-247 5.15 x 10-12

Sm-151 4.55 x 103

 Cm-248 5.36 x 10-12

Eu-152 2.20 x 101

 Bk-249 6.31 x 10-19

Eu-154 9.74 x 102

 Cf-249 4.79 x 10-11

Eu-155 2.57 x 102

 Cf-251 2.47 x 10-1

Ra-226 1.30 x 101

 Cf-252 5.32 x 10-14

Total 2.74 x 106

transfer to the SPF or calculation of the radiological waste composition will be used to perform

the sum-of-fractions calculations for the relevant performance objectives.

4.2.3 NRC Evaluation – Source Term

NRC staff evaluated the information supporting DOE’s estimated inventory. Cs-137 and its

short-lived daughter Ba-137m are arguably the most well-characterized radionuclides in the

tank farm waste. DOE estimated that the total activity of Cs-137 to be disposed of at the SDF

would be approximately 1.1 x 1017 Bq (3 MCi), but could possibly range as high as 1.9 x 1017 Bq
- 39 -

(5 MCi) (DOE, 2005). A direct estimate of the uncertainty in the other radionuclides was not

provided, but it would be expected, on a relative basis, to be larger as a result of more limited

characterization and in some cases more complicated partitioning between the physical

components of the waste (e.g., supernate, saltcake, sludge). 

As indicated in Section 3.1, the estimation of tank inventory (which directly influences the

eventual SDF inventory) and the radiological composition of each physical phase, is

complicated by a number of factors. A number of inputs used to develop the tank farm

inventory are based on assumptions, process knowledge, or other forms of indirect information. 

Even with the 2003 and 2004 campaigns to improve the estimates of the radiological

composition of the supernatant and salt phases, the sampling of a number of highly radioactive

radionuclides is fairly limited. Therefore there is uncertainty in the radiological composition of

the waste feed that will be sent to the SPF. To mitigate this uncertainty, DOE has developed a

sampling plan to: (1) provide characterization of the waste streams input to the saltstone feed

tank, (2) ensure that the grout recipe can be modified as needed to compensate for changes in

the salt, organic, or solids content of the salt solutions, (3) verify that the waste is

nonhazardous, and (4) verify compliance with the WAC (Ketusky, 2005). Although the

frequency of sampling is provided in the plan, it is not clear that the sampling frequency for

various types of samples is consistent with the schedule for batch waste preparation

operations. DOE indicates that process knowledge will be used for chemical and radiological

characterization of the transfers to Tank 50, and that material balance may then be used to

estimate the chemical and radiological composition of the material transferred to the SPF. 

Tank 50 contents would be re-baselined with data from samples analyzed quarterly for

chemical characteristics and semi-annually for radiological characteristics (Ketusky, 2005).

Process knowledge can be used in the development of inventory estimates; however, past

experience with tank farm characterization has demonstrated that process knowledge can

result in significant deviations from actual waste stream composition as determined by direct

sampling. As indicated by DOE, the uncertainty in Cs-137 activity sent to the SPF may be as

large as approximately 70% of the current best estimate, and Cs-137 is one of the most wellcharacterized radionuclides in the tank farm system. The sampling plan needs to consider the

uncertainty in the methods and data used to estimate the radiological composition of the waste

sent to the SPF. If a factor of safety was applied to the inventory limits generated with the PA,

then the likelihood that material that exceeds the WAC would be sent to the SPF (because of

indirect methods of estimating composition) could be reduced. However, DOE has not

indicated that it intends to apply any additional measures to account for the uncertainty in the

waste characterization. The least uncertain method is to perform adequate sampling of each

batch of waste in the feed tank prior to transferring the waste to the SPF to ensure that the

WAC can be met. Sampling of each batch of waste in the feed tank prior to transfer to the SPF

is needed unless it can be demonstrated through a predictive exercise that indirect methods

can adequately predict the concentrations of radionuclides in the saltstone feed tank.

Use of an adequate sampling plan combined with a sum-of-fractions approach to compare the

disposal of the actual saltstone inventory to the performance objectives is an acceptable

approach to managing uncertainties in the estimation of the radiological source term for the PA.
- 40 -

4.2.4 Infiltration and Erosion Control

The SDF closure concept is illustrated in Figures 5 and 6. After each vault is filled with

saltstone, interim closure will involve pouring a clean grout layer between the saltstone and

overlying concrete roof. Final closure of the disposal facility will be accomplished by

constructing a thick, multi-layer engineered cap, a drainage system, and performing revegetation of the site. The drainage system will consist of rip-rap lined ditches that intercept the

gravel layer of the moisture barrier. These ditches will divert surface runoff and water

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 the SDF. The topsoil will be

re-vegetated with bamboo that is expected to quickly establish a dense ground cover and limit

the growth of pine trees, the most deeply rooted naturally occurring plant type at the Savannah

River Site. Through evapotranspiration, the bamboo is expected to remove a large amount of

moisture from the soil and decrease infiltration into the underlying disposal system.

The engineered cap serves two primary purposes: to limit infiltration to the waste and to limit

erosion. Both are discussed in detail below. By limiting water contact with the waste, the

release of radioactivity from advection can be limited. The erosion control barrier will be located

approximately one meter below the top of the final closure cap. Approximately 3.2 m (10.5 ft) of

material will be between the erosion control barrier and the top of the waste. Most agricultural

or resident intruder scenarios consider a nominal excavation depth of 3 m (9.8 ft), therefore 

proper design, implementation, and function of the erosion control barrier would essentially

eliminate direct contact with the waste for intruder receptors, reducing the exposure pathways

to primarily that of direct radiation exposure. The remaining material (upon erosion down to the

erosion control barrier) will also provide shielding from gamma radiation for intruder receptors. 

The top layer of the cap will be topsoil and backfill on which bamboo will be grown. Bamboo

provides an effective method to reduce infiltration through evapotranspiration. Below the

erosion barrier will be an upper drainage layer and a geosynthetic clay layer (GCL), which are

designed to limit infiltration.

4.2.4.1 Infiltration

DOE recognizes that limiting the amount of water flowing through the saltstone vaults and

contacting the saltstone is important for limiting releases of radionuclides from the vault into the

accessible environment. DOE proposes to reduce water fluxes that contact the waste using a

multiple barrier system, including: (1) vegetative reduction of infiltration using a dense ground

cover for a species with large evapotranspiration demand during institutional control, (2) a

closure cap designed to shed water from the vaults using redundant drainage layers, and

(3) favorable hydraulic characteristics of the vault and wasteform that limit both advective

transport and molecular diffusion within water. The closure cap consists of three functional 

components separated by backfill, including: (1) an erosion barrier, (2) an upper drainage unit

intended to shed water far from the vault, and (3) a lower drainage unit intended to route water

around the top and sides of the vault into a buried drainage system.
- 41 -

Figure 5 Engineered Closure Cap Side View

Each of these barriers is assumed to degrade over time. Institutional control of the initial

ground cover (bamboo) is assumed to be maintained for 100 years. Once institutional control

ceases, ground cover is assumed to revert to native pine forest after 200 years. Reversion to

pine forest would decrease evapotranspiration, causing an increase in net infiltration. Closurecap degradation is defined as altering the thickness or hydraulic properties of each of the cap

layers over time. Three degradation processes were considered: (1) erosion, which removes

topsoil and upper backfill; (2) pine forest root penetration, which punctures the upper GCL, and

(3) colloidal clay migration, which moves colloidal clay from backfill into the pores of the

underlying drainage layers. Saturated hydraulic conductivity was measured for the porous

medium used for the drainage layers, and the manufacturer provided a value for the GCL. 

Changes in hydraulic properties were inferred from the degradation processes.

4.2.4.2 Erosion Control

The upper layers of the engineered cap will primarily serve as erosion barriers and will consist

of a 1 m (3 ft) thick soil cover and a 0.3 m (12 in) rock layer on the top slopes of the disposal

cell. The rock erosion barrier will be placed on top of the middle backfill and geotextile fabric to

form a barrier to erosion if a gully were to form in the upper soil layer. However, the upper soil

layer is designed to be stable during the occurrence of very rare flood events, such that a

considerable margin of safety exists for the prevention of erosion. For the side slopes, erosion

protection will be provided by a 0.6 m (24 in) rock layer placed directly on top of the slopes. 
- 42 -

Figure 6 Engineered Closure Cap Layers

The computation of peak flood discharges for various site design features was performed by

DOE in several steps. These steps included: (1) selection of a design rainfall event;

(2) determination of infiltration losses; (3) determination of times of concentration;

(4) determination of appropriate rainfall distributions, corresponding to the computed times of

concentration; and (5) calculation of flood discharge. Input parameters were derived from each

of these steps and were then used to determine soil cover slopes and rock sizes for erosion

protection.

One of the phenomena most likely to affect long-term stability is surface water erosion. DOE

used a Probable Maximum Precipitation (PMP) computed by deterministic methods, rather than

statistical methods, and the PMP was based on site-specific hydrometeorological

characteristics. The PMP has been defined as the most severe reasonably possible rainfall

event that could occur as a result of a combination of the most severe meteorological
- 43 -

conditions occurring over a watershed. PMP values were estimated by DOE using

Hydrometeorological Report No. 51 (HMR-51) (COE, 1978) and HMR-52 (COE, 1982). These

reports also provide information on distributing the rainfall that falls over a particular drainage

area. A 1-hour PMP of 48.8 cm (19.2 in) and a 5-minute PMP of 15.7 cm (6.2 in) were used by

DOE as a basis for estimating a Probable Maximum Flood (PMF) for the smaller areas at the

site such as the top and side slopes. 

The determination of the peak runoff rate is dependent on the amount of precipitation that

infiltrates into the ground during its occurrence. If the ground is saturated from previous rains,

very little of the rainfall will infiltrate and most of it will become surface runoff. The loss rate is

highly variable, depending on the vegetation and soil characteristics of the watershed. 

Typically, all runoff models incorporate a variable runoff coefficient or variable runoff rates. 

Commonly-used models such as the U.S. Bureau of Reclamation (USBR) Rational Formula

(USBR, 1977) incorporate a runoff coefficient (C); a C value of 1 represents 100% runoff and

no infiltration. Other models such as the U.S. Army Corps of Engineers Flood Hydrograph

Package HEC-1 (COE, 1988) separately compute infiltration losses within a certain period of

time to arrive at a runoff amount during that time period. In computing the peak flow rate for

the small drainage areas at the site, DOE used the Rational Formula (USBR, 1977). In this

formula, the runoff coefficient was assumed to be 0.8; that is, DOE assumed that very little

infiltration would occur. 

The time of concentration is the amount of time required for runoff to reach the outlet of a

drainage basin from the most remote point in that basin. The peak runoff for a given drainage

basin is inversely proportional to the time of concentration. If the time of concentration is

assumed to be smaller, the peak discharge will be larger. Times of concentration and/or lag

times are typically computed using empirical relationships. Times of concentration for the riprap

design were estimated by DOE using the Kirpich Method (USBR, 1977). This method is

generally accepted in engineering practice.

After the PMP is determined, it is necessary to determine the rainfall intensities corresponding

to shorter rainfall durations and times of concentration. A typical PMP value is derived for

periods of about one hour. If the time of concentration is less than one hour, it is necessary to

extrapolate the data presented in the various hydrometeorological reports to shorter time

periods. To accomplish this, DOE used a procedure recommended in HMR-52. This

procedure involves the determination of rainfall amounts as a percentage of the one-hour PMP,

and computes rainfall amounts and intensities for very short periods of time. To calculate peak

flood flows, the 5-minute PMP rainfall intensity was determined to be about one-third of the 1-Hr

PMP, or about 15.7 cm (6.2 in). 

To estimate PMF peak discharges for the top and side slopes, DOE used the Rational Method

(Chow, 1959). This method is a simple procedure for estimating flood discharges and is

recommended in NUREG-1623 (Johnson, 2002). In using the Rational Method, DOE assumed

a runoff coefficient equal to 0.8. DOE also assumed a flow concentration factor of about 5 and

a factor of safety of about 1.35 that would be used to increase the flows for design purposes. 

For a maximum top slope length of 140 m (450 ft) (with a slope of 0.015) and a side slope

length of 18 m (60 ft) (with a slope of 0.33), DOE estimated the peak flow rates to be about

0.29 m3

/s-m width (3.1 cubic feet per second per foot of width [cfs/ft]) for the top slope and 0.33

m3

/s-m width (3.5 cfs/ft) for the side slope. PMF flow rates for overland flow for the
- 44 -

downstream apron were estimated by DOE and are slightly less than the flow rates for the side

slopes. 

The ability of a riprap layer to resist the velocities and shear forces associated with surface

flows over the layer is related to the size and weight of the stones which make up the layer. 

Typically, riprap layers consist of a mass of well-graded rocks which vary in size. Because of

the variation in rock sizes, design criteria are generally expressed in terms of the median stone

size, D50, where the numerical subscript denotes the percentage of the graded material that

contains stones of less weight. For example, a rock layer with a minimum D50 of 10 cm (4 in)

could contain rocks ranging in size from 1.9 cm (0.75 in) to 15 cm (6 in); however, at least 50%

of the weight of the layer will be provided by rocks that are 10 cm (4 in) or larger.

The top portion of the cell will be protected by a 1 m (3 ft) thick soil layer. DOE proposes that

the soil cover will be stable and will provide the necessary erosion protection for a 10,000-year

period. The top slope was evaluated for erosional stability by DOE using the permissible

velocity procedure discussed in NUREG-1623. Using a peak discharge of about 0.29 m3

/s-m

(3.1 cfs/ft), the maximum velocity was computed to be about 0.79 m/s (2.6 ft/s), less than the

maximum permissible velocity of 0.91 m/s (3.0 ft/s). However, to increase the margin of safety,

in the unlikely event that a gully would occur in the soil cover for any reason, DOE proposes to

provide a riprap layer under the 1 m (3 ft) soil cover to prevent further vertical erosion into the

geotextile and drainage layers. This size of this riprap was computed assuming that a flow

concentration factor of 5 would occur, and a factor of safety of 1.35 would be applied. Using

the design procedure recommended in NUREG-1623, DOE computed the required rock size to

resist further gullying to be about 6.6 cm (2.6 in). DOE proposes to place a 0.3 m (12 in) layer

of riprap with a D50 of about 10 cm (4 in) under the soil cover. For the side slopes of the cell,

DOE proposes to use a 0.6 m (24 in) layer of rock with an average D50 of 0.25 m (10 in). 

Methods discussed in NUREG-1623 were used to determine a required rock size of 20 cm

(7.7 in). 

The design of the apron for the cell must be adequate to withstand forces from several different

phenomena and is based on the following general concepts: (1) provide riprap of adequate size

to be stable against overland (downslope) flows produced by the design storm (PMP), with

allowances for turbulence along the downstream portion of the toe; (2) provide uniform and/or

gentle grades along the apron and the adjacent ground surface such that runoff is distributed

uniformly onto natural ground at a relatively low velocity, minimizing the potential for flow

concentration and erosion; (3) provide an adequate apron length and quantity of rock to allow

the rock apron to collapse into a stable configuration if gullying occurs and erodes toward the

site, and (4) provide an apron with adequate rock size to resist flows that will occur laterally

along the apron. Using design criteria suggested in NUREG-1623, DOE determined that riprap

with a D50 of 0.25 m (10 in), a layer thickness of 0.91 m (36 in), and a width of 6 m (20 ft) would

be placed in the apron area. 

Depending on the rock source, variations occur in the sizes of rock available for production and

placement, and it is therefore necessary to ensure that these variations in rock sizes are not

extreme. Examples of acceptable gradations are provided in NUREG-1623. The estimated

average D50 values to be used as the basis for the design of well-graded mixture of rock to

resist the shear forces of the PMF peak discharge were determined to be 10 cm (4 in) for the

top slopes and 25 cm (10 in) for the side slopes and apron. Riprap gradations and layer

thicknesses were developed by DOE using criteria suggested in NUREG-1623. 
- 45 -

Rock durability is defined as the ability of a material to withstand the forces of weathering. 

Factors that affect rock durability are: (1) chemical reactions with water, (2) saturation time, (3)

temperature of the water, (4) scour by sediments, (5) windblown scour, (6) wetting and drying,

and (7) freezing and thawing. To assure that the rock used for erosion protection remains

effective for long periods of time, potential rock sources must be tested and evaluated. A

procedure for determining the acceptability of a rock source is presented in NUREG-1623. In

general, rock durability testing is performed using standard test procedures, such as those

developed by the American Society for Testing and Materials (ASTM). The ASTM publishes

and updates an Annual Book of ASTM Standards (ASTM, 1995), and rock durability testing is

usually performed using these standardized test methods.

In accordance with the criteria outlined in NUREG-1623, DOE provided information regarding

testing, inspection, and quality control procedures to be used for the erosion protection

materials. The information included programs for durability testing, gradation testing, and rock

placement. DOE provided information regarding the testing program that will be used to

document the durability of the rock source selected. Using the scoring procedure discussed in

NUREG-1623, DOE intends to test the rock and to use only rock that achieves a minimum

score of 80. DOE's proposed rock durability testing program includes the following tests: (1)

bulk specific gravity, (2) absorption, (3) sodium sulfate soundness, (4) L.A. abrasion at 100

cycles, and (5) Schmidt hardness.

DOE proposes that rock gradation testing will be performed in accordance with standard ASTM

test procedures, as suggested in NUREG-1623. DOE provided a placement program where

riprap will be placed in accordance with ASTM standard test procedures and procedures

suggested in NUREG-1623. Further, because the rock layer will be covered with soil, it is

important to assure that the rock voids are uniformly filled and that differential settlements are

minimized. To accomplish this, DOE intends to fill the rock voids with a grout mixture (flowable

fill). Placement of this type of material will greatly enhance the stability of the rock layer and

adds a considerable margin of safety to the overall design. 

4.2.5 NRC Evaluation – Infiltration and Erosion Control

It is important to note that design for long-term infiltration control and design for long-term

erosion control have different objectives and different degradation mechanisms. Acceptability

of a design for one does not necessarily mean that an acceptable design has been achieved for

the other.

4.2.5.1 NRC Evaluation - Infiltration

The DOE approach to assessing the long-term performance of the engineered cap to limiting

infiltration is based on a reasonable conceptual model. DOE considered three degradation

processes: (1) erosion, which removes topsoil and upper backfill; (2) pine forest root

penetration, which punctures the upper GCL; and (3) colloidal clay migration, which moves

colloidal clay from the backfill into the pores of the underlying drainage layers. The approach of

using the HELP model with assumptions about pine tree root penetrations over time is a

reasonable conceptual model for estimating deterioration of the upper GCL over time. The

amount of infiltration and the timing of infiltration through the upper GCL results in long-term

near surface infiltration rates that are not significantly lower than natural recharge rates may be

expected to be for the site. 
- 46 -

NRC staff requested that DOE provide its estimate of the amount of infiltration available to the

wasteform that may be diverted due to the low permeability vault and saltstone. This type of

information would provide an indication of how much credit was being taken by DOE for longterm cap performance. DOE provided a figure of fractional flow through the saltstone as a

function of time for the base case PA (WSRC, 2005a). The fractional flow reaches a maximum

of approximately 2 x 10-4 at 10,000 years. It is not clear how effective the lower drainage layer

of the engineered cap would be at limiting infiltration to the waste at higher infiltration rates

through the cap or for larger amounts of degradation to the saltstone and vaults 

Independent analysis by the NRC staff suggests that several factors mediate the closure cap

performance in limiting fracture flows. The upper GCL is the first key engineered barrier, and

will fail as roots penetrate the layer. The lower drainage layer is the second key barrier, and

could fail as it is plugged with colloidal clay. A lower GCL is placed on the roof of the vaults;

however, DOE has taken no performance credit for the lower GCL in the analysis. The lower

GCL mediates flow rates and could control whether or not a fracture is filled with a porous

medium, both largely depending on whether the GCL ruptures during settlement. The DOE

analysis suggests that it is credible to consider the scenario where the GCL allows the fracture

to be filled with the overlying porous medium, while exposure to the environment degrades the

saltstone near the fracture. This scenario would be credible for every fracture that stretches the

GCL beyond its limits, either by opening at the top or by vertical displacement. Since the

fracture would be filled with the same material as the drainage layer, there would be no capillary

effect precluding flow into the fracture, although flow rates might be small. There is no known

relevant experience for the long-term performance of GCLs.

An independent NRC analysis suggests that the model used by DOE to estimate the effects of

colloidal clay plugging the lower drainage layer may not be conservative. If the hydraulic

properties of the lower drainage unit are modified uniformly throughout the layer to reflect

partial plugging by colloidal clay, open-fracture flow may be allowed in less than a third of the

time predicted by the DOE model, in which complete plugging occurs in a discrete sublayer

within the lower drainage unit. The NRC staff expects that colloidal clay transport will result in

the highest degree of clay deposition occurring at the top of the lower drainage layer, with clay

deposition that may be smeared from the sharp front implied by the DOE model. The NRC staff

recognizes that the rates of colloidal clay transport used by DOE are based on literature values

(Phifer and Nelson, 2003), while the soils at the SRS have low pH values that may modify

colloidal transport. Because the integrity of the lower drainage layer behavior affects every

fracture, uncertainty in drainage layer behavior is expected to be important to performance. 

Uncertainty in the estimated rate of degradation of the lower drainage layer could be reduced

through further studies or could be mitigated through additional conservatism in cap design. 

The DOE model for lateral diversion in the lower drainage layer, as implemented, may

exaggerate lateral diversion in the PA exercises due to the calculational approach. The

approach represents the degraded lower drainage layer as an equivalent anisotropic medium in

the PORFLOW simulations. The horizontal and vertical hydraulic conductivities of the lower

drainage layer are calculated assuming flow is in series and in parallel, respectively, which NRC

staff agrees is a reasonable approach. However, it is only appropriate to perform this

calculation using saturated hydraulic conductivities when the medium is saturated. For

example, the ratio of horizontal-to-vertical conductivity using the DOE approach increases from

1.1 to 233 over the period of simulation. When using representative suction heads of 100,

1,000, and 5,000 cm (3.3, 33, and 164 ft) in both media simultaneously, the maximum ratio of
- 47 -

horizontal- to-vertical unsaturated conductivity is 280, 3.2, and 1.2, respectively. This

discrepancy will presumably reduce predicted doses, since the hydraulic conductivities of the

vault are small, and it may not have a large effect unless fracture flows are initiated.

Aside from the shortcomings noted above, NRC staff believes the DOE approach to assessing

long-term cap performance is reasonable. However, DOE provided limited model support to

justify the numerical modeling results for long-term infiltration through the engineered cap. 

Recent experience with field-scale observations of engineered caps at a number of different

types of sites suggest that, in many cases, performance may be significantly less than

predicted by numerical modeling (Albright, et al., 2004). Model support is needed to ensure

that the idealized and simplified numerical calculations are appropriately representing the realworld system. The rate at which the lower drainage layer plugs is determined by the rate at

which colloidal clay migrates into the layer. With the current modeling approach, the likelihood

of flow in fractures (if present) is determined by the saturation of the lower drainage layer, which

in turn is primarily driven by the hydrologic properties of the lower drainage layer. As shown in

Section 4.2.16, both the amount of infiltration to the waste and the likelihood of fracture flow can

significantly influence whether the disposal system can meet the performance objective of

protection of the public. The processes being modeled are highly uncertain, and adequate

justification is needed that the modeling has appropriately accounted for uncertainties or is

sufficiently conservative based on the model support.

4.2.5.2 NRC Evaluation - Erosion Control

The primary purpose of the erosion control system is to ensure that a thick cover of soil is

maintained over the waste to eliminate exposure pathways to hypothetical inadvertent intruders.

To mitigate the potential effects of surface water erosion, the NRC staff considers that it is very

important to select an appropriately conservative rainfall event on which to base the flood

protection designs. Further, the staff considers that the selection of a design flood event should

not be based on the extrapolation of limited historical flood data, due to the unknown level of

accuracy associated with such an extrapolation. No recurrence interval is normally assigned to

the PMP; however, the staff has concluded that the probability of such an event being equaled

or exceeded is very low. Accordingly, the PMP is considered by the NRC staff to provide an

reasonable design basis. DOE used HMR-51 and HMR-52 to estimate PMP amounts and staff

review of these procedures indicate that the PMP amounts are reasonable for the small

drainage areas at the site. 

Based on a review of the calculations, including the time of concentration, rainfall intensity, and

runoff, the staff concludes that the estimates for peak flood discharges are conservative and

are, therefore, reasonable. In particular, for the design calculations DOE used a runoff

coefficient of 0.8. Based on the conservatism associated with the use of a 5-minute PMP of

15.7 cm (6.2 in) and a resulting rainfall intensity of 190 cm/hr (74 in/hr), the staff concludes that

this is a reasonable assumption. Times of concentration for the riprap design were estimated

by DOE using the Kirpich Method (USBR, 1977). This method is generally accepted in

engineering practice and is considered by the staff to be appropriate for estimating times of

concentration at this site. Based a review of the calculations provided, the staff concludes that

the time of concentration values used by DOE were appropriately derived. To calculate peak

flood flows, the 5-minute PMP rainfall intensity was determined to be about one-third of the 1-hr

PMP, or about 15.7 cm (6.2 in). Based on a review of this aspect of the flooding determination,

the staff concludes that the computed peak rainfall intensities are reasonable.
- 48 -

Staff reviewed the DOE calculations for stability of the top slope, side slopes, and apron. 

Based on review of the calculations, the conservatism associated with the design discharge,

and the ability of the cover to provide adequate protection against the computed velocities and

shear stresses, the staff concludes that the top slope should be stable under extreme flooding

conditions. The proposed rock size for the side slope is adequate, and the apron design is

acceptable (based on the conservatism associated with the computation of flow rates). Based

on the acceptability of using procedures suggested in NUREG-1623, the riprap design should

achieve the objectives. In addition, the proposed information and procedures are sufficient to 

ensure acceptable placement of the riprap and fill into the riprap voids. 

DOE’s proposed gradations are adequate to assure acceptable protection. Based on review of

the durability test procedures and ASTM gradation test procedures, the staff concludes that the

programs are appropriate. The gradation testing program will ensure that rock layers with

acceptable gradations are provided. 

Based on review of the information submitted by DOE and on independent calculations, the

NRC staff concludes that DOE has provided sufficient information to conclude that the erosion

protection design is adequate to provide reasonable assurance of long-term stability of the

closure cap for erosion control purposes. It should be emphasized that DOE has provided a

conceptual design, and this design could change in the future, depending on the number and

location of vaults. Staff’s conclusions on this conceptual design is based largely on the

conservatism associated with the design, as proposed. There may be several individual design

parameters which may not be acceptable if some of the other conservative design parameters

are changed or removed. When making design changes, DOE should ensure appropriate

conservatism is maintained.

4.2.6 Wasteform and Concrete Vault Degradation

DOE evaluated several alternatives for the final disposal of decontaminated salt waste and the

preferred method it selected was the conversion of the waste to a cementitious form suitable for

shallow land disposal on the SRS. A cementitious wasteform was selected because it is: (1)

resistant to leaching; (2) tolerant to variations in waste composition, waste loading, and

radiation levels; (3) produced from low cost and readily available inorganic materials; and

(4) suited to high volume production and emplacement using known concrete production

technology (Cook and Fowler, 1992). The bases for this decision were discussed in the

Environmental Impact Statement for Defense Waste Processing Facilities at the Savannah

River Site (DOE, 1982). Subsequent to this decision, DOE undertook extensive studies to

optimize the production process and the solid waste product and to select an appropriate landfill

design and location. The final wasteform evolved from a mixture of salt solution and cement to

its present formulation containing a blend of slag, fly ash, and cement or lime mixed with the

salt solution. The disposal technology has evolved from a series of simple trenches that would

be filled with waste and covered with native soil to the current design of above-grade concrete

vaults that provide an engineered secondary containment barrier for the waste. The following

description is taken mostly from Cook and Fowler (1992) and Cook, et al. (2005).

The current wasteform, referred to as saltstone, is a solid that is the product of chemical

reactions between a salt solution and a blend of cementitious materials (blast furnace slag, fly

ash, and a lime source). The hydration of slag in the saltstone releases sulfide species,

predominantly S2!

, into the pore fluid, which imposes a strongly reducing redox potential on the
- 49 -

system and chemically binds several contaminants as insoluble species. Thus, the propensity

of these contaminants to leach from the solid wasteform is reduced. A broad range of

wasteform precursor compositions has been shown via leach testing to yield saltstone that

qualifies as nonhazardous waste, as defined by EPA. The recommended composition for

saltstone, using a nominal salt solution blend, is 47 wt% salt solution (29 wt% salt), 25 wt % of

Grade 120 slag, 25 wt% of Class F fly ash, and 3 wt% of Type II Portland cement or lime.

Chemical reactions between the components in the salt solution and the dry blend changes the

composition of the saltstone compared to the starting materials. Cr (VI) species, technetium

(VII) species, and salt solution contaminants that form sparingly soluble sulfides (Hg, Co, Ni,

Zn, Tc, Ru, Rh, Sb, Sn) react with components of the slag to form insoluble phases that are not

readily leached from the saltstone wasteform. Water and soluble Al (III), Ca (II), Ba (II), and Sr

(II) species are incorporated into the cement matrix as the dry materials hydrate and the

saltstone sets. Tc (VII) is believed to react with components of the slag to form Tc2S7, whereas

Cr (VI) in the salt solution is believed to be reduced to a lower oxidation state and precipitated

as a Cr (III) hydroxide solid. Strontium and barium are incorporated as aluminosilicates within

the aluminosilicate structure of the cement paste. These less soluble forms effectively fix the

contaminants, thus reducing their potential release through groundwater pathways.

When first prepared, the saltstone grout has the consistency and flow characteristics of latex

paint and is readily pumped from the SPF to a cell in a disposal vault. The disposal vaults

serve as a form to support the saltstone grout until it sets into a solid, monolithic wasteform. 

The saltstone itself provides primary containment of the waste, but the walls, floors, and roof of

the vaults provide a secondary containment barrier for the contaminants in the waste and

isolate the waste from direct contact with the environment. 

In the proposed disposal site layout, up to 14 additional vaults made of reinforced concrete will

be constructed. These vaults will each have dimensions of approximately 60 m (197 ft) wide by

180 m (591 ft) long by 7.6 m (25 ft) high, divided into twelve cells approximately 30 m (98 ft)

wide by 30 m (98 ft) long by 7.6 m (25 ft) high. One vault that has already been constructed

and filled with material, designated as Vault 1 on the site layout plan, is approximately

30 m (98 ft) wide by 180 m (591 ft) long by 7.6 m (25 ft) high, divided into six cells with the

same cell dimensions as those in the larger vaults. Operationally, the cell of the vault will be

filled to a height of approximately 7.5 m (25 ft) with saltstone, and then a layer of

uncontaminated grout will be poured to fill in the space between the saltstone and the sloped

roof. Vaults 1 and 4 have already been built at the SDF. These two vaults are made of

reinforced concrete containing slag. Because the current saltstone formulation also includes

slag, the use of slag in the construction of future vaults to mitigate the release of certain

contaminants is redundant and may be reduced or eliminated to reduce costs and to improve

the resistance to cracking due to shrinkage during curing. 

As mentioned previously, the saltstone wasteform provides primary containment of the

contaminants, and the concrete vault serves as a secondary containment barrier mitigating the

release of contaminants to the environment. The performance of these barriers is important in

order to meet the NRC performance objectives in 10 CFR Part 61, Subpart C. Thus, the

methods used to analyze the long-term performance of the SDF must account for potential

mechanisms of contaminant release from the facility, and the potential mechanisms for loss of

integrity of the SDF engineered barriers. At a conceptual level, mechanisms that lead to the

release of contaminants from the saltstone consist of dissolution, diffusion, dispersion, and
- 50 -

advection. The rates of contaminant release by these mechanisms are controlled principally by

(1) the moisture flux to the saltstone through the vault from the overlying soil, (2) the solubility

and the diffusive and sorption properties of the contaminants, and (3) the chemical properties

(e.g., redox condition) and physical characteristics (e.g., hydraulic conductivity, presence, and

density of fractures) of the wasteform and vault. For example, the DOE analysis relies on the

presence of blast furnace slag in the saltstone to produce chemical conditions that result in low

solubilities and increased retardation for certain contaminants. The hydraulic conductivity of the

vault and saltstone will affect the rate at which contaminants can migrate to the surrounding

soil. Degradation of the vaults and monoliths through chemical (e.g., sulfate attack,

carbonation, leaching, and rebar corrosion) or physical (e.g., cracking due to settlement of

foundation soil, seismic activity) processes is likely to increase contaminant release due to

increased advection of contaminants.

In conducting its PA of the facility, DOE considered the various mechanisms of release to

estimate the source term and release of contaminants. Both diffusive and advective transport

processes were addressed. To model contaminant transport in the near field, there was a need

to estimate the contaminant concentrations in the pore fluid based on the concentrations in the

saltstone. However, relating inventory in the saltstone to pore fluid concentration is complicated

by various processes, such as precipitation/dissolution reactions, aqueous complex formation,

and sorption. DOE acknowledged that these processes are poorly understood and difficult to

quantify for the SDF. Thus, concentrations in the saltstone pore fluid were derived using

models that considered only reversible linear sorption (i.e., distribution coefficients [Kd values]). 

In its 1992 PA (Cook and Fowler, 1992), DOE used geochemical analyses to derive the initial

pore solution concentrations of Tc-99, I-129, and nitrate, but applied Kd values for saltstone

constituents considered by DOE to be relatively unimportant with respect to acceptable

performance of the facility. In its 2005 special analysis (Cook, et al., 2005), DOE used Kd

values taken from a report by Bradbury and Sarott (1995), which was considered to provide the

best conservative Kd values for cement-water systems available in literature. The Kd values

selected from the report of Bradbury and Sarott (1995) are values that apply to reducing

environments. In selecting those values, DOE assumed that the rate of oxidation of the

saltstone was sufficiently slow such that the saltstone remained reducing throughout the 10,000

year simulated performance due to the presence of slag.

Degradation that alters the integrity and permeability of the saltstone monoliths and vaults also

was considered. Thus, scenarios with an intact vault and a degraded vault were evaluated. 

However, DOE recognized that “the timing and extent of degradation are not readily predictable

due to enormous uncertainties in conditions for thousands of years” (Cook and Fowler, 1992). 

Thus, in the 1992 PA (Cook and Fowler, 1992), cracking of the vaults and monolith was chosen

to represent the increased permeability of the waste and the vaults. For simplicity, cracking

was represented by vertical cracks that fully penetrate the vault and saltstone, with a spacing of

3 m (10 ft) and an average aperture of 0.005 m (0.016 ft). The 1992 PA also included

sensitivity and uncertainty analyses on various processes and related parameters to determine

their significance to facility performance. The parameters to which the results were most

sensitive were the hydraulic conductivity and diffusivity of concrete and saltstone. For degraded

vaults, the results were most sensitive to the depth of perched water, crack spacing, and the Kd

values. 
- 51 -

In the 2005 special analysis (Cook, et al., 2005), cracks in the saltstone were not included in the

analysis on the basis that cracks will open either at the top or at the bottom and will be pinched

closed at the opposite end, preventing the development of through-wall cracks (Peregoy, 2003). 

Instead, degradation of the closure system, the saltstone wasteform, and the vault, regardless

of the mechanism, was represented by using material properties that varied for each time

interval. For example, it was assumed that the hydraulic conductivities of saltstone and

concrete would increase as time proceeds, which would gradually increase water percolation

through the vault. The hydraulic conductivities of saltstone and concrete in the time interval of

0 to 100 years were set to 1 x 10!11 and 1 x 10!12 cm/s (1 x 10!5

 and 1 x 10!6 ft/yr), respectively,

and were increased to 1 x 10!9

 cm/s (1 x 10!3 ft/yr) over a 10,000-year period through eight

steady-state stages. The degradation rate for concrete was assumed to be faster than for

saltstone because the concrete will be exposed to the environment and would be more

vulnerable to attack by sulfate, chloride, and other chemical reactions. DOE acknowledged that

the assumptions regarding the changes in hydraulic properties over time are based on

professional judgment because actual data over the time periods of interest do not exist.

4.2.7 NRC Evaluation – Wasteform and Concrete Vault Degradation

The hydraulic conductivities of nondegraded saltstone and concrete vault used in the PA were

based on limited data. The hydraulic conductivity of the nondegraded concrete vault (1.0 ×

10!12 cm/s [1.0 × 10!6

 ft/yr]) used in the analyses was based on measurements by Core

Laboratories, Inc. of samples from the Savannah River Site E-area low-level waste disposal

vault (WSRC, 1998). The hydraulic conductivity of nondegraded saltstone (1.0 × 10!11 cm/s

[1.0 × 10!5

 ft/yr]) was based on measurements on saltstone grout specimens by Core

Laboratories, Inc. (Yu, et al., 1993). For comparison purposes, values of hydraulic

conductivities of concrete and cement pastes available in the literature were tabulated by DOE

(WSRC, 2005a). The tabulated values indicate that 1.0 × 10!12 cm/s (1.0 × 10!6

 ft/yr) for

nondegraded vault is on the low end of the range of literature values and is similar in magnitude

to that of high performance concrete that uses a super plasticizer. Higher values (1.1 × 10!10 to

2.3 × 10!9

 cm/s [1.1 × 10!4 to 2.4 × 10!3 ft/yr]) of hydraulic conductivity were measured for

samples taken from the Savannah River Site Z-area saltstone vault. DOE explained that the

higher hydraulic conductivity for the Z-area vault samples was due to the lesser quality of

workmanship in the construction of the Z-area vault compared to the E-area vault. Data from

(Malek et al., 1985) for earlier formulations of saltstone provided hydraulic conductivity values

that ranged from 2.5 × 10-6 cm/s [2.5 ft/yr] to <1.0 × 10-11 cm/s [1.0 × 10-5 ft/yr] for nondegraded

saltstone. Measurements were presented at different curing times. It is not clear if the

variability in the measurements is a function of curing time or if the variability results from the

measurement technique. The data suggest that longer curing times may result in higher

permeabilities, although it cannot be determined whether the result is an artifact of the

uncertain measurement technique. Because the hydraulic conductivity data being relied upon

by DOE represents limited samples over a small range of curing times, DOE should consider

performing additional tests for the current saltstone formulation.

DOE indicated a range of saltstone compositions over which acceptable saltstone can be

produced based on results of Toxicity Characteristic Leaching Procedure (TCLP) tests (WSRC,

2005b). In the RAI, NRC requested justification that the physical properties of saltstone would

not vary significantly over the broad range of composition for which the TCLP tests were

performed. DOE indicated that the nominal composition of saltstone has not varied significantly

on a daily or batch basis (WSRC, 2005b). The larger range of compositions were intended to
- 52 -

accommodate the State-approved operating permit for the facility. Because there was a fairly

significant amount of variability in the TCLP test results, significant deviation from the nominal

saltstone composition may result in a saltstone wasteform composition for which tests have not

been performed to develop hydraulic conductivity and effective diffusivity values. 

Because of the potentially problematic interactions of Tank 48 waste with other tank farm waste

or with the proposed treatment processes, DOE has indicated Tank 48 waste will be sent to

SPF without treatment; that is, there will be no removal of highly radioactive radionuclides from

the Tank 48 waste before it is solidified. DOE conducted TCLP tests on samples containing

Tank 48 waste and the results were found to be acceptable. However, the technical basis (e.g.,

hydraulic conductivity and effective diffusivity) for the performance of a saltstone wasteform

containing Tank 48 waste has not been developed. 

In the 2005 special analysis (Cook, et al., 2005), the degradation of the concrete vault and

saltstone was modeled by increasing the hydraulic conductivity of the vault and saltstone to

1.0 × 10!9

 cm/s (1.0 × 10!3 ft/yr) at 10,000 years. This value was selected based on

professional opinion of DOE staff. The professional opinion is based on: (1) an assessment of

potential degradation mechanisms identified in published literature; (2) a simplified assessment

of the consequences of these mechanisms; (3) a determination that freeze/thaw and erosion

will not occur because the vault and wasteform will remain buried over the period of

performance; and (4) the assumption that the effect of environmental factors such as acid rain

and leaching by infiltrating water would be significantly less than the effect of cracking (WSRC,

2005a). Recent sensitivity analyses presented by WSRC (2005a) demonstrated that an

increase in degradation rate results in an increase in calculated dose. For example, increasing

the final hydraulic conductivity to a value typical of a clayey-sandy soil (1.0 × 10–6 cm/s [1.0

ft/yr]), instead of the value used in the 2005 special analysis, increased the calculated dose by

a factor of 320 (WSRC, 2005a). The calculated dose (when properly scaled, see Section

4.2.16) would be approximately 1.1 mSv/yr (110 mrem/yr) when assuming a hydraulic

conductivity of 1.0 × 10–6 cm/s (1.0 ft/yr), which would exceed the performance objective of 0.25

mSv/yr (25 mrem/yr). Furthermore, an increase in hydraulic conductivity would also adversely

affect the reducing capacity of the saltstone due to diffusion or advection of oxygen into the

monolith. A combination of increased fluid flow and oxidizing conditions would lead to a 

calculated dose higher than 110 mrem/yr.

In general, the NRC staff agree with the qualitative assessment of the degradation mechanisms

for saltstone. However, given that: (1) the calculated releases from the SDF are sensitive to the

values of hydraulic conductivity of the vault and saltstone; and (2) “the timing and extent of

degradation are not readily predictable due to enormous uncertainties in conditions for

thousands of years” (Cook and Fowler, 1992, Section 3.1.3.5), it would be useful to reduce the

uncertainties associated with the hydraulic conductivity and long-term integrity of the vault and

saltstone. Additional laboratory measurements of initial hydraulic conductivity, as well as longterm tests or monitoring studies designed to evaluate the long-term durability of the saltstone

and concrete vault, would help reduce these uncertainties. The vaults into which the saltstone

will be poured contain rebar and fill pipes, in the current design, that would be exposed to the

environment. Unless mitigated by future design modifications, consideration should be given to

the effects of rebar and fill pipe corrosion on the integrity and projected lifetime of the vaults in

future degradation calculations. In the response to action items, DOE indicated that if

penetrations in the vault provide an unacceptable moisture flow pathway, the closure plan for

the facility will be revised to provide modifications to the design (WSRC, 2005a).
- 53 -

4.2.8 Release and Near-Field Transport

DOE assumed that the vault and saltstone would have several fractures due to differential

settlement, with fractures widening gradually over time. Fracture characteristics were based on

a structural analysis (Peregoy, 2003) that suggests that the cracks will be preferentially created

at joints between concrete pours, and fully penetrate the vault and saltstone, but will only be

open on one end (either top or bottom). Vertical offsets between adjacent blocks were not

estimated. The cementitious materials of the vault and wasteform were assumed to gradually

increase their hydraulic conductivity over time to account for degradation. Hydraulic

conductivity measurements were obtained on concrete and saltstone core samples; these

measured values are used for early times. Changes over time were based on engineering

judgment. An analysis suggesting that the suction heads were so large that fractures would not

flow under steady conditions (Cook, et al., 2005, Appendix A.4) was used to justify neglecting

fractures in most analyses. In other analyses, fractures were imposed as a series of discrete

features with properties similar to gravel.

DOE’s groundwater pathways analysis of radionuclide releases used a sequence of three

computer codes. The HELP code (EPA, 1994a,b) was used to calculate net infiltration and

water balances from the ground surface through the upper geosynthetic clay liner. The

PORFLOW code (Analytical and Computational Research, Inc., 2002) used the water fluxes

through the upper GCL calculated by HELP as an upper boundary condition and calculated

water fluxes in the vadose zone below the upper GCL, within the vault, and in the vadose zone

above the saturated zone to calculate a series of steady-state flow fields. In a subsequent step,

PORFLOW used the flow fields to calculate the time history of radionuclide releases to the

saturated zone. A separate PORFLOW model was used to calculate a steady-state regional

water balance. The time history of radionuclides released to the vadose zone was combined

with the steady-state saturated zone water fluxes to calculate radionuclide transport in the

saturated zone. 

Water balances in the vadose zone were calculated as a series of steady-state periods,

representing discrete intervals over the analysis. For each period, hydraulic properties were

imposed to account for degradation of the closure cap and vault, the HELP model calculated

steady-state water fluxes past the upper GCL, and the PORFLOW model calculated steadystate water fluxes below the upper GCL. These water fluxes are held fixed for the duration of a

steady-state period while radionuclide release and transport was calculated using PORFLOW. 

At the end of each period, a new steady-state flow field was created, concentrations were

adjusted to account for change in cell saturation, and the transport simulation continued. This

procedure can result in non-physical transport responses at the start of a new period that would

be smoothed out if the flow fields changed gradually over time.

A number of sensitivity analyses were performed to evaluate the impact of uncertainty in

parameters or models. Sections 4.2.15 and 4.2.16 provide a discussion of the DOE analyses

and the NRC staff interpretation of those analyses.

4.2.9 NRC Evaluation – Release and Near-Field Transport

DOE is concerned with three primary groundwater-pathway release scenarios: (1) diffusive

release via concentration gradients in the pore fluid of the saltstone to the environment outside

the vaults; (2) bulk movement of water through the saltstone matrix; and (3) focused release via
- 54 -

fracture flow in a few discrete gaps caused by settlement. DOE suggests that fracture flow

does not occur until the lower drainage layer fails through plugging resulting in increased

saturation that can overcome the capillary barrier effect imposed by the presence of the

differing properties of the geologic and engineered materials (Rosenberger, et al., 2005). The

DOE conclusion is that the vault system will likely meet performance objectives if the closure

cap and vault have properties similar to those used by DOE for its base case. However, there

are certain areas in which the DOE modeling assumptions are not strongly supported, and

plausible alternative assumptions in these areas may lead to situations that will not meet the

performance objectives.

The NRC staff’s understanding of the physical processes relevant to water flow and

radionuclide releases are summarized in order to put comments on the DOE modeling into

context. The following description does not consider releases from disruptive processes (e.g.,

human intrusion) or gaseous releases through the air pathway. The staff believes that

concentrations within the pore water are spatially uniform within most of the saltstone matrix as

long as the emplacement stream is unchanged during emplacement. Since the movement of

water is slow within the matrix, dissolved radionuclide concentrations should have sufficient

time to reach equilibrium with the radionuclides within the saltstone matrix. However,

radionuclide concentrations may be significantly altered near the edges of the saltstone matrix

due to interactions with the external environment. These interactions include movement of

water, oxygen, and radionuclides across the saltstone boundary.

Movement of radionuclides out of the saltstone matrix will occur as diffusion from the edge of

the saltstone through the concrete vault into the accessible environment. Advective transport

will also occur as water moves across the boundary of the saltstone, carrying dissolved

radionuclides with the water. Advective transport tends to counteract diffusive transport where

water enters the saltstone and augments diffusive transport where water exits.

Similar effects will regulate the movement of oxygen into the saltstone, potentially creating an

oxidized layer that allows certain radionuclides (e.g., Tc-99) to be more mobile where the

saltstone meets the accessible environment. Oxygen will diffuse in the gas phase more readily

than the liquid phase and may move into the vault more quickly than dissolved radionuclides

are released, unless the vault materials are completely liquid-saturated or a gas-phase diffusion

barrier is incorporated into the vault design. DOE did not simulate transport of the gas phase;

instead, DOE assumed that, outside of the saltstone block, infiltrating water was in equilibrium

with atmospheric concentrations of oxygen and then simulated transport of dissolved oxygen in

saltstone based on the assumption that the saltstone would remain completely saturated with

water. If the saltstone is partially saturated, and oxygen can move through the saltstone in the

gas, rather than the liquid phase, it is reasonable to use a coefficient of diffusion for oxygen that

is larger than for water-saturated conditions. If the rate-limiting portion of slag capacity

consumption is due to transport, the implication is that the oxidized rind may be thicker than

estimated by DOE.

In an alternative conceptual model, conditions at the edge of the saltstone matrix may play the

dominant role in determining radionuclide releases. When water fluxes are small (as in the

DOE base case), diffusion is the primary release mechanism and releases may occur around

the entire vault boundary. When water fluxes are large, advection is the primary release

mechanism and calculated releases are dominated by advection out of the bottom of the vault. 

Since essentially all oxidation occurs where the radionuclides contact the external environment,
- 55 -

a highly oxidized rind over a few volume percent of the saltstone may govern release rates,

which is quite different from the situation where the same total oxidation is uniformly spread

throughout the saltstone matrix. Diffusive transport is the primary transport mechanism in the

DOE base case. Because diffusive transport is directly proportional to the concentration

gradient in the fluid phase and fluid phase concentrations of the highly radioactive radionuclide

Tc-99 are many orders of magnitude higher under oxidizing conditions compared to reducing

conditions, it would be expected that releases from the saltstone would be sensitive to the rate

of formation of the oxidized layer.

The other primary potential release mechanism occurs because of flowing water within a

fracture. A certain amount of water will imbibe from the fracture into the saltstone matrix when

the matrix has a capillary suction less than necessary to maintain flowing water in the fracture;

hence, releases will be due to radionuclides diffusing against the direction of water movement. 

However, the fracture also provides a potential pathway for oxygen to move into the saltstone

matrix so that oxidizing conditions may be found at the fracture surface. 

In the PORFLOW simulations of release from the saltstone and vaults, DOE used moisture

characteristic curve data in the description of moisture flow under unsaturated conditions

(WSRC, 2005b). In the RAI, NRC staff questioned the use of the data because of anomalously

high increases in gas flow during the laboratory test (NRC, 2005a). Upon re-examination of the

data, DOE acknowledged that the testing cannot be used as the basis for moisture

characteristic curves for the numerical modeling. DOE assessed the impact of the use of the

questionable data by assuming the relative permeability of the saltstone and concrete was 1.0

regardless of the saturation, and found that dose increased by a factor of 4 over the base case

results (WSRC, 2005a). The NRC believes the sensitivity analysis performed to address this

concern was bounding, but that either the impacts should be considered as part of the base

case or additional tests should be performed to replace the questionable data.

DOE’s demonstration that the disposal system could meet the performance objectives was

largely determined by several factors, including: (1) water passing through the saltstone matrix

is limited; (2) radionuclide diffusion rates are limited; (3) water passing through fractures is

limited; and (4) oxidation near saltstone boundaries is limited.

The dose significance of water flowing through the saltstone matrix is clearly demonstrated in

several sensitivity scenarios modeled by DOE (sensitivity scenarios 12, 23, 24, 25, 27, and 28

in WSRC, 2005a), in which increased mean annual precipitation and increased vault saturated

hydraulic conductivities have a large influence on dose even when oxidation does not occur. 

Water flowing through the matrix determines advective releases from the vault together with

radionuclide concentrations at the edge of the vault where the water exits. Based on

scenarios 4 through 11 (see Table 8), dose uncertainty is more sensitive to uncertainty in the

saltstone hydraulic properties than in the concrete properties, as might be expected from the

much larger path length within saltstone. However, the vault and saltstone wasteform act 

independently as barriers to moisture movement in the DOE modeling, such that the largest

increases in dose are observed for scenarios when both barriers are degraded. Increasing the

mean annual precipitation by 25%, which is not unreasonable for representing unexpectedly

poor closure-cap performance, increases doses by a factor of 23. The NRC staff interpretation

of the sensitivity results is that the water fluxes removing radionuclides by advection are largely

determined by (1) hydraulic properties of the inner core of the saltstone and to a lesser extent

the concrete vault and (2) closure-cap performance. In other words, if the inner core remains
- 56 -

substantially undegraded, advective releases would be relatively insensitive to degradation of

the concrete and outer fringe of the saltstone, since most of the lower exit boundary would be

under a low-permeability umbrella. Degradation of the inner core is difficult to predict because

of a lack of data about the long term performance of cementitious materials. However, the

NRC staff believes that degradation is likely to proceed to a greater extent near the accessible

environment.

The dose significance of radionuclide diffusion is examined in scenarios 14 through 18 (WSRC,

2005a), in which an order-of-magnitude increase in the diffusion coefficient in both concrete

and saltstone increases doses by a factor of 14, while increasing either coefficient

independently has a much smaller response. The diffusion coefficient value used for soil, while

not expected to influence transport significantly, is at the low end of values reported for

common cations and anions in water from the cited reference (Domenico and Schwartz, 1990),

reasonably accounting for tortuosity effects. It appears that a unit conversion error may have

reduced the soil value by an order of magnitude in DOE’s simulations. Table 4-6 in

(Rosenberger et al., 2005) provides a molecular diffusion coefficient used in the modeling of

158 cm2

/yr when the correct value based on 5 x 10-5 cm2/s should be 1580 cm2/yr. It is not

believed that the molecular diffusion coefficient of soil is a key factor in estimating release rates

to the saturated zone. The molecular diffusion coefficients used for concrete and saltstone

were 4 orders of magnitude smaller than the range reported in the cited reference, but DOE

leaching experiments and external literature suggest that the diffusion coefficients are

reasonable for intact concrete. Additional diffusive behavior occurs due to mechanical

dispersion of moving fluids, which would also potentially increase diffusive releases. DOE does

not justify the neglect of mechanical dispersion, but the NRC staff notes that mechanical

dispersion is more significant as length scales and water velocities increase, and therefore

should not be significant over release-significant scales on the order of a meter with the slow

velocities expected within the saltstone and concrete. In scenarios where the saltstone and

vault were assumed to deteriorate hydrologically, DOE did not consider an increase in the 

diffusion coefficient over time as the concrete vault and saltstone degrade or develop cracks. 

Cracking would be expected to significantly shorten the diffusive path length and could result in

increases in the effective diffusion coefficient depending on the frequency and severity of the

cracking. Since both the effective diffusion coefficient and hydraulic conductivity are intimately

related to pore structure in cementitious materials, it is reasonable to expect that both

coefficients would be subjected to some increase over time.

The significance of water flowing through fractures is examined in scenarios 31 and 32, in

which three discrete features are included in the simulation as vertical strips of saturated gravel,

with both fractures and matrix assumed to have artificially saturated hydraulic conductivity in

order to maximize flows. When completely reducing conditions were assumed, doses

increased by a factor of 73 (essentially at the regulatory limit when results are appropriately

scaled. See Section 4.2.16), while assuming completely oxidized conditions increased doses a

further factor of 7 because of an overall factor of 810 increase in Tc-99 releases. An additional

scenario completed by DOE but not provided in their table of sensitivity scenarios simulates the

effect of filling the fracture with 2.5 cm (1.0 in) of native soil, showed a factor of 23 increase in

dose (without oxidation). These results should be comparable to filling the fracture with the

gravel drainage layer material, since the two materials would have similar hydraulic conductivity

above 1,000 cm (33 ft) of suction head. DOE dismisses fracture flows as not credible because

of an analysis that indicates fracture flows would be negligible if the lower drainage layer does

not develop sufficient saturation such that there is a head difference to overcome the air entry
- 57 -

pressure of the fractures. The NRC staff agrees with the approach of the analysis, but

disagrees with the interpretation because of the limited consideration of a number of sources of

uncertainty and limited model support for the numerical predictions. 

Model support is needed to justify the model results that flow in fractures in the vaults and

saltstone will not occur in the unsaturated system. A number of real world observations

suggest that moisture will flow in joints or fractures under conditions that may not be predicted

by numerical models. For example, observations at the tank farm facility at a DOE facility in

Idaho suggest that water flows from unsaturated soils through joints in the vaults surrounding

the high-level waste tanks (Lockie, 2002). It is postulated that flow occurred primarily as a

result of a rapid influx of snow melt or precipitation events. While the system at SRS may not

be completely analogous, there are many similarities and the observations suggest that the

temporal and spatial averaging used in the calculations must be consistent with the expected

real-world system. Modeling to justify the lack of flow through fractures would typically include

the following features: (1) heterogeneity in material properties; (2) temporal variations in

saturation, especially resulting from rapid transport of infiltration through root holes to the lower

drainage layer; (3) infilling of fractures and joints with porous media; (4) offset of the essentially

impermeable vault/saltstone on either sides of the fractures; (5) variability in the aperture of the

joints and fractures; (6) sensitivity to grid size in the simulations, especially at the interface of

the fracture and porous media; and (7) variability in moisture characteristic curve parameters. 

Ideally, this type of analysis would be performed probabilistically because of the many sources

of uncertainty. The modeling results must be supported by appropriate information, such as a

blind prediction of observations at analogous systems, lab- and field-scale experiments, or

other forms of model support that take into account site-specific behavior. 

The dose significance of the oxidation state near matrix boundaries is examined in

scenarios 20–22, 25, 29, 30, and 33 (WSRC 2005a). The only radionuclide examined by DOE

that is significantly affected by the oxidation state is Tc-99, with negligible doses under reducing

conditions. Under the assumption of complete oxidation but otherwise nominal conditions,

there was a factor of 67 increase in dose relative to the reduced state and the calculated dose

was very close to the regulatory limit. Under wetter conditions (scenarios 25 and 30), the

assumption of complete oxidation resulted in a factor of 180 increase to dose relative to the

reduced state, with failure to meet the regulatory limit (when results are appropriately scaled. 

See Section 4.2.16) regardless of the oxidation state. Under highly degraded conditions

(scenario 33), complete oxidation results in a factor of 110 increase to dose relative to the

reduced state. The NRC staff agrees with the following conclusions made by DOE: (1) it is

unlikely that more than a fraction of the saltstone mass will oxidize during the performance

period, (2) the oxidized zone will likely be where the saltstone is near the accessible

environment and along fractures, and (3) calculations with more representative oxidation

conditions will likely fall between the two extremes. The DOE interpretation is that the bulk

saltstone matrix, essentially unaffected by oxidation, will govern Tc-99 releases, while the NRC

staff interpretation is that a thin rind of oxidized saltstone may greatly increase Tc-99 releases

relative to the DOE conceptual model. 

The sensitivity simulations suggest that Vault 4 performance is largely controlled by: (1)

closure-cap performance eliminating fracture flow and reducing matrix flow; (2) degradation

rates in the inner core of the saltstone matrix increasing hydraulic conductivity; (3) degradation

rates in the outer fringe of the saltstone matrix and concrete vault increasing diffusion and
- 58 -

advection; and (4) oxidation of the saltstone near boundaries with the accessible environment. 

Degradation and oxidation of the concrete and saltstone are discussed further in Section 4.2.7.

The strategy followed by DOE is to use a series of steady-state flow fields to calculate releases

and subsequent transport, apparently without adjusting the fluid densities to account for the

dissolved salts within the fluid. Using steady-state flow fields is readily justified if the entire flow

system equilibrates quickly to changes in boundary conditions relative to the period of analysis. 

The saltstone matrix has not been demonstrated to respond quickly, and it would be expected

that response times could easily be measured in thousands of years with the nominal properties

used by DOE. If the suction head in the saltstone matrix is initially much greater than the

steady state condition (the matrix is drier than calculated), it may be conservative to use a

series of equivalent steady-state flows because capillary effects will tend to draw water into the

saltstone, thereby reducing releases relative to the steady-state calculation. Alternatively,

releases could be enhanced if the matrix is initially wetter than the steady-state flow estimates

since water in the matrix must leave to achieve the long-term conditions. There are no

measurements of the initial conditions in the saltstone. The neglect of fluid density (e.g., brine)

in the flow calculations may also affect release rates and transport in the vadose zone. PA

calculations could account for brine releases simply by performing transient simulations with

coupled density-dependent flow and salt transport, where the modeled salt consists of the suite

of dissolved constituents. This could automatically account for the slow response time of the

saltstone and concrete matrix.

4.2.10 Hydrology and Far-Field Transport

For the purposes of this review, the far field is defined as that portion of the contaminant

transport pathway that starts at the water table. Near-field and vadose zone processes are

discussed in previous sections of this report, but discussion of Kd’s used for vadose zone

transport will be included in this section.

In the recent special analysis, DOE modeled contaminant groundwater transport from the water

table below Vault 4 to the compliance boundary 100 m down-gradient using the flow and

transport code PORFLOW (Cook, et al., 2005). The localized, steady-state groundwater model

used for Vault 4 was based on the regional model described by Flach (2004). Site descriptive

information, including hydrogeology, is detailed by Cook and Fowler (1992, Section 2). The

aquifers affected by contaminants from Vault 4 in the model are largely unconsolidated, Tertiary

Coastal Plain sands with variable clay contents and some clay layers. In the immediate vicinity

of Vault 4, the water table is approximately 23 m (75 ft) below the planned ground surface

above the vault and approximately 12 m (40 ft) below the bottom of the vault. For PORFLOW,

a “reduced” local model, based on the larger regional model of Flach (2004), was constructed

for a 640- by 580-m (2,100- by 1,900-ft) area around Vault 4. A horizontal grid with 30 m (100-

ft) elements was constructed and vertical grid elements of variable height were selected to

reflect hydrostratigraphy (Cook, et al., 2005, Section A.3). Compliance nodes that intersect the

groundwater zones most affected by Vault 4-derived contaminants were assigned for

monitoring model contaminant concentrations. For radionuclides, these nodes were around

100 m (328 ft) to the east-northeast (i.e., down-gradient) of the Vault 4 footprint. Flow modeling

showed that contaminants from Vault 4 will be restricted to the lower unit of the Upper Three

Rivers Aquifer in the flow path to the 100-m (328-ft) compliance boundary; therefore,
- 59 -

compliance nodes were restricted to that unit. The underlying Gordon Confining Unit will

prevent transport to the Gordon Aquifer. 

The inputs to the saturated zone PORFLOW model consist of time histories of contaminant

fractional release rates from the vadose zone model. The release rate, which is normalized to

the source concentration, is recalculated as a concentration term based on the water volume of

the model source nodes. For radionuclides, the model output is fractional concentration in units

of pCi/L/Ci of source inventory in groundwater at each of the compliance nodes. For

demonstrating compliance with the 10 CFR 61.41 performance objectives, DOE uses the

PORFLOW outputs to calculate groundwater doses from multiple pathways (Rosenberger,

et al., 2005). Model doses (per curie of source radionuclide) are based on peak groundwater

concentrations, conservatively combining all peak radionuclide concentrations for the

dose calculation regardless of when they occur.

The key parameters affecting model radionuclide transport are saturated hydraulic conductivity

(hydrostratigraphic unit specific) and sorption coefficient, or Kd (radionuclide-specific). 

Saturated zone hydraulic conductivity values for the special analysis were based on

characterization data and calibrated to well data (Flach and Harris, 1999; Flach, 2004,

Figure 2-6). Model sorption coefficients with references were reported in Rosenberger, et al.

(2005, Table 4-12) and further discussed was provided in Westinghouse Savannah River

Company (2005b, pp. 277–292). For vadose zone (below the disposal vault) and saturated

zone transport modeling, tabulated soil values were used for all units except the Gordon

Confining Unit, for which clay values were used (Pickett, 2005). For many radionuclides, the

literature compilation for sand soils by Sheppard and Thibault (1990) was cited. For all other

radionuclides (except tritium and californium, which are risk-insignificant), site-specific

information was used in developing sorption coefficients. For plutonium DOE adopted an

oxidation state model that distributed plutonium isotopes into two species representing less

mobile Pu (III+IV) and more mobile Pu (V+VI). The ratio between the species was based on

field measurements of oxidation and reduction rates.

The PORFLOW model calculated short groundwater travel times from the water table below

Vault 4 to the compliance nodes of less than 10 years (Cook, et al., 2005, Figures A–40 and

A–42). For the special analysis base case, fractional concentration time histories at the

100-m (328-ft) compliance boundary for radionuclides are presented in Figures A–47 through

A–76 of Cook, et al. (2005). All concentration values are very low and, except for radionuclides

completely depleted by decay, rise at the end of the 10,000-year simulation period. The latter

observation is true even for radionuclides such as Tc-99 that have very low soil Kd’s and should

therefore travel very quickly through the saturated zone. This relationship demonstrates that

the saturated zone does not provide an effective barrier for relatively poorly sorbing

radionuclides.

DOE (Cook, et al., 2005; Rosenberger, et al., 2005; WSRC, 2005a,b) provided limited

sensitivity studies directed at saturated zone parameters. Recharge and hydraulic

conductivities of the Gordon Confining Unit were varied in WSRC (2005b, pp. 270–276) to

represent uncertainty in saturated zone flow rates; modest sensitivity to dose was found (a dose

increase by a factor of 1.7 for the worst case). No analyses were sensitive to Kd in isolated

subvault vadose or saturated zones. Sorption coefficients in vadose and saturated zone natural

media were decreased in two of the recently reported sensitivity studies (WSRC, 2005a,

scenarios 19 and 26), but the values also were decreased in saltstone and concrete in the
- 60 -

same analyses. However, there is indirect evidence that model results are much more sensitive

to saltstone and vault characteristics than to geosphere sorption coefficients (WSRC, 2005a). 

For example in scenario 26, reduction of Kd values in all media throughout the model system by

a factor of ten increased dose by a factor of approximately 40. However, cases in which the

Tc-99 Kd in the saltstone and vault alone was decreased due to oxidizing conditions

(scenarios 21, 22, and 29) yielded far greater dose increases, as did cases involving saltstone

hydraulic properties and oxidizing conditions (scenarios 30, 32, 33). Because (1) Tc-99 is

responsible for these calculated dose increases, (2) the natural system Tc-99 Kd is already low

(0.1 mL/g), and (3) saturated zone hydraulic and sorption characteristics are better constrained

(i.e., uncertainty is reduced) than the equivalent long-term parameters for the disposal system,

dose is not apparently highly sensitive to a reasonable range of saturated zone transport

characteristics in this model system.

Another indication of saturated zone performance can be seen in the sensitivity analysis for

water usage inside the 100-m (328-ft) boundary (Rosenbereger, et al., 2005, Section 8.3.3). In

this analysis, peak vadose zone annual radionuclide releases to the water table were diluted

into the top mesh layer of the saturated zone and not subjected to transport and retardation. 

This water volume formed the basis for an all-pathways groundwater dose calculation; the dose

was mainly due to Se-79. For comparison to the base case groundwater pathways analysis

(Rosenberger et al., 2005, Section 4.1), which used a different inventory, it is convenient to look

at dose per activity values for Se-79. The sensitivity case neglecting saturated zone transport

yielded a value of 3.5 x 10-14 mSv/Bq-yr (0.13 mrem/Ci-yr) for Se-79, whereas the base case

value was 6.8 x 10-14 mSv/Bq-yr (0.025 mrem/Ci-yr). In other words, unrealistically neglecting

any retardation and attenuation effects of the saturated zone increased dose by only a factor of

five. In light of sensitivity studies of vault and saltstone parameters (WSRC, 2005a), this effect

is minimal. Neglecting retardation and attenuation in the saturated zone had a small effect

because those radionuclides with low natural system Kd values (e.g., Tc-99, Se-79, I-129)-and

which are therefore less affected by saturated zone transport-also tend to dominate dose in the

base case (or in disposal system sensitivity cases for Tc-99) because of their low Kd values in

the disposal system.

4.2.11 NRC Evaluation – Hydrology and Far-Field Transport

The PORFLOW implementation of groundwater flow and transport modeling for Vault 4 is well

supported, appropriately integrated with regional and local hydrogeological conditions, and well

suited for the purpose. NRC staff note, however, that the model may be overpredicting the

ability of the saturated zone to dilute the vadose zone-derived contaminant plume. According to

WSRC (2005a, Figure 18-2), the concentration of relatively unretarded I-129 decreases by a

factor of greater than 1 x 105

 during transport from the vadose zone to the 100-m (328-ft)

compliance location. Figures A–43 and A–44 of Cook, et al. (2005) show that horizontal fluxes

in the saturated zone are much greater than the vertical influxes from the vault, which staff

agrees will result in dilution of the concentrations. However, the model grid layers in the lower

aquifer above the Gordon Confining Unit are about 6-m (20-ft) thick vertically, while particle

traces from adjacent corners show a vertical extent of about 1.2 m (4 ft) (Cook, et al., 2005,

Figure A–43). The actual plume would be expected to be similar to the envelope of particle

traces, while in numerical simulations radionuclide mass is smeared over one or more elements

in the vertical dimension. Because PORFLOW calculates average concentration within a grid

cell, a simple mass balance suggests that peak concentrations are diluted by at least a factor of

five due to the coarse vertical grid resolution. A smaller vertical grid dimension would result in
- 61 -

significantly higher concentrations being calculated in certain cells, rather than diluting the

contaminants throughout a larger volume. Similar considerations apply in the horizontal

dimension. A finer model grid may give more realistic contaminant concentrations. A reduction

in the grid spacing in the portion of the model immediately surrounding Vault 4 and the

compliance nodes to verify that peak concentrations are properly calculated should be

considered. In practice, grid refinement might be accomplished by using a separate nested grid

aligned with the groundwater flow direction for transport calculations while using boundary

conditions from the regional grid. Alternately, DOE can ensure that assumptions regarding

dilution from withdrawals in a well would result in dilution effects as large or larger than that

introduced from the grid refinement.

As described on page A-54 of Cook, et al. (2005), the model vadose zone contaminant flux is

distributed evenly among the 12 source nodes at the water table. New sensitivity analyses

have considered the possibility of release and transport along large-scale fractures in the

saltstone and vault (WSRC, 2005a, scenarios 31 and 32). Focused release and transport of

contaminants from the disposal system could result in more localized pulses than would be

represented by the vault footprint. Averaging these pulses over the footprint is another

potential source of dilution. If further analyses were to conclude that large-scale fracture flow

through the disposal system were credible, the horizontal resolution of the model grid should

be reevaluated.

DOE performed a sensitivity analysis on plume interaction between Vaults 1 and 4 that

suggested only a 25% increase in dose from inclusion of the Vault 1 plume (Cook, et al., 2005,

Section 7.5.1). That report also mentioned further consideration of plume interactions in future

PAs. These considerations should include all planned disposal systems and should consider

effects beyond the compliance location for any one vault. For example, the dose from two

intersecting plumes farther than 100 m (328 ft) from the two sources could potentially exceed

the dose at any single 100-m (328-ft) compliance location.

Saturated hydraulic conductivities used in the model are based on extensive characterization

and well calibration. Calculated dose is not strongly sensitive to these parameters. The

PORFLOW model appears to have appropriately bounded any uncertainty in hydraulic

conductivities in the saturated zone.

As discussed in the previous section, calculated dose is less sensitive to order-of-magnitude

changes in Kd than to changes in Tc-99 Kd alone due to oxidizing conditions. In addition, the

Tc-99 Kd for the vadose and saturated natural systems is so low that lower values will not

strongly affect dose. Therefore, a risk-informed evaluation of sorption coefficients will address

uncertainties or variability on an order-of-magnitude scale. It appears that, in most cases, DOE

has appropriately chosen Kd values for natural system transport simulations. Where

site-specific data were lacking, literature values that tended toward the low end of ranges were

selected. For example, in Rosenberger, et al. (2005), model soil values were chosen from the

sand data in Sheppard and Thibault (1990), instead of from data more representative of more

sorptive loam, clay, and organic soils. In addition, in nearly every case, a selected site-specific

value was lower than the corresponding literature value. A notable exception is uranium, for

which a Kd of 800 mL/g was selected from pH-dependent, field-derived concentration ratios

between soils and coexisting waters (Serkiz and Johnson, 1994). This value is much higher

than the Sheppard and Thibault (1990) sand value of 35 mL/g and the value of 40 mL/g used in

some previous SRS models (Serkiz and Johnson, 1994). A number of uncertainties are
- 62 -

attendant on application of concentration ratios as such Kd values, including the uranium

sorption mechanism (e.g., precipitation versus surface complexation); redistribution during

sampling, storage, and handling; and applicability of F and H seepage basin conditions to

groundwater flow in the Z-area subsurface. These uncertainties, in light of the lack of more

site-specific data obtained under better-controlled conditions, suggest that 800 mL/g may not

be suitably conservative. Under current model conditions, the Kd for uranium does not appear

to be risk-significant; however, if the PA is significantly changed, the uranium Kd should be

reassessed.

Aquifer geochemical conditions can directly affect radionuclide sorption. In the current model,

release of contaminants to the saturated zone appears unlikely to significantly change saturated

zone water chemistry. This is because of the apparently high degree of dilution afforded by the

saturated zone (WSRC, 2005a), such that, nitrate concentrations and pH would not change

appreciably due to the chemical plume emanating from the disposal system. However, DOE

has not provided an analysis of this potential effect. If subsequent calculations suggest that

dilution at the water table may be less than previously predicted, potential chemical effects on

saturated zone sorption coefficients should be considered. A straightforward means for

assessing this issue would be a sensitivity analysis that lowers only chemically sensitive

saturated zone Kd values to appropriate levels. In addition, available information does not

address the potential for vadose zone sorption to be less effective than expected due to

contaminant plume chemistry. Again, this question could be addressed by a sensitivity analysis

focused only on vadose zone natural material Kd values.

4.2.12 Dose Methodology

The dose methodology used by DOE in the SDF PA process was the application of dose

conversion factors. This methodology is widely used in PAs and consists of multiplying the

radionuclide concentration in air, water, or soil that a receptor might be exposed to through any

of the various pathways by the dose conversion factor specific to that ingestion or inhalation

process and radionuclide.

The groundwater pathway dose calculation was conducted by DOE using the LADTAP XL©

spreadsheet code (Simpkins, 2004) discussed previously. The dose factors used in LADTAP

XL© are those specified by DOE in PAs conducted throughout the DOE complex in two

separate documents on external and internal dose conversion factors for calculation of dose to

the public (DOE, 1988a,b). The calculation process and the dose factors for the groundwater

pathway PA used are described in Cook, et al. (2005); Rosenberger, et al. (2005); and Jannik

(2005).

The air pathway dose calculation is conducted within the widely used EPA computer code

CAP88 (EPA, 2002). CAP88 was used to calculate dose factors in units of mrem/yr per Ci/yr

released at the ground surface, Savannah River Site boundary, and 100 m (328 ft) from Vault 4

(Cook, et al., 2005). This dose conversion factor methodology has been used in other PAs

conducted by DOE, EPA, and NRC.

In the intruder analysis performed by DOE for the SDF PA, radionuclide dose conversion

factors from the Federal Guidance Reports developed by EPA (EPA, 1993, 1988) were used

(Cook, et al., 2005; Rosenberger, et al., 2005). Ingestion and inhalation dose conversion

factors were taken from Federal Guidance Report No. 11 (EPA, 1988), and external dose
- 63 -

conversion factors were taken from Federal Guidance Report No. 12 (EPA, 1993). Both of

these reports provide 50-year committed effective dose equivalents per unit of activity based on

the exposure pathway (inhalation, ingestion, or external) and the specific radionuclide. The

intruder analysis dose conversion factors were included in the intruder analysis code developed

by DOE (Koffman, 2004).

4.2.13 NRC Evaluation – Dose Methodology

The dose methodology implementation of the SDF PA is well supported and suited for the

purpose. Numerous NRC guidance documents provide recommendations on the approach and

use of the specific dose conversion factors used in the SDF PA process. These include

NUREG–1573 (NRC, 2000b), which provides guidance on the use of pathway dose conversion

factors for calculating doses via the potential exposure pathways, and NUREG–1757 (NRC,

2003b, Volume 2, Appendix I), which provides guidance on the use of specific dose conversion

factors such as those developed by the EPA and published in Federal Guidance Report Nos. 11

and 12 (EPA, 1988, 1993).

4.2.14 Overview of Performance Objectives

The NDAA establishes the applicable criteria for determining that waste is not HLW, including

that the waste will be disposed of in compliance with the performance objectives set out in

subpart C of part 61 of title 10, Code of Federal Regulations. The performance objectives

provide criteria to ensure that the public, workers, and environment will be protected from

releases of radioactivity.

10 CFR 61.41, “Protection of the general population from releases of radioactivity,” states:

“Concentrations of radioactive material which may be released to

the general environment in ground water, surface water, air, soil,

plants, or animals must not result in an annual dose exceeding an

equivalent of 25 millirems to the whole body, 75 millirems to the

thyroid, and 25 millirems to any other organ of any member of the

public. Reasonable effort should be made to maintain releases of

radioactivity in effluents to the general environment as low as is

reasonably achievable.” (NRC, 2001a)

The 0.25-mSv/yr (25-mrem/yr) limit applies for the post-closure period of a disposal facility. 

The other radiological control limits of 10 CFR Part 20, “Standards for Protection Against

Radiation,” apply during facility operation (NRC, 2001b). 

10 CFR 61.42, “Protection of individuals from inadvertent intrusion,” 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.”
- 64 -

Although a particular dose limit is not specified in this performance objective, compliance with

the technical requirements of Part 61 and, in particular, with the classification system of 10 CFR

61.55, is considered to provide adequate protection to intruders at a near-surface land disposal

facility. In the Draft Environmental Impact Statement for Part 61 (NRC, 1981), NRC used a 5-

mSv (500-mrem) dose limit to an acute inadvertent intruder to establish the concentration limits

and other aspects of the waste classification system. In addition, Part 61 does not specify a

time for institutional controls in the performance objectives, but does require, in 10 CFR

61.59(b), that “... controls may not be relied upon for more than 100 years.”

10 CFR 61.43, “Protection of individuals during operations,” states:

“Operations at the land disposal facility must be conducted in

compliance with the standards for radiation protection set out in

part 20 of this chapter, except for releases of radioactivity in

effluents from the land disposal facility, which shall be governed by

61.41 of this part. Every reasonable effort shall be made to

maintain radiation exposures as low as is reasonably achievable.”

This performance objective applies to both the public and to disposal facility workers.

10 CFR 61.44, “Stability of the disposal site after closure,” 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 stability performance objective is consistent with a premise of Part 61 that the facility must

be sited, designed, used, operated, and closed with the intention of providing permanent

disposal. A disposal facility should not require long-term maintenance and care. Stability is

particularly important considering the requirements in 10 CFR 61.59(b) that “... institutional

controls must not be relied upon for more than 100 years following transfer of control of the

disposal site to the owner.”

Table 7 provides an overview of the DOE results compared to the performance objectives of 10

CFR 61, Subpart C. The results are explained in more detail in the sections that follow.

4.2.15 Protection of the Public

The public is represented by an adult member of a farming community (e.g., resident farmer)

that lives in a residence downgradient of the SDF. During the operational and institutional

control periods, it is assumed that the individual resides at the SRS site boundary. After active

institutional controls cease at 100 years, the member of the public resides 100 m downgradient

from Vault 4. As previously indicated, up to 14 additional vaults would be needed to dispose of

saltstone grout. Groundwater modeling to assess the impact from all of the vaults considering

the potential for overlap of plumes from individual vaults at any potential receptor location was 
- 65 -

Table 7 Summary of DOE Results Compared to Performance Objectives

Part 61 Performance Objective Performance

Limit

DOE PA

Result

(mrem/yr)

61.41 All-pathways dose to public 25 mrem/yr 2.3(2)

61.42 Intruder - Resident scenario 500 mrem/yr 21

61.42 Intruder - Agricultural scenario(1) 500 mrem/yr 149

61.42 Intruder - Post drilling scenario(1) 500 mrem/yr 31

61.43 Protection of individuals during operations 5 rem/yr(3) NA(4)

(1) Because of the long-term performance of the engineered cap to prevent erosion and the long-term

durability of the saltstone wasteform, DOE believes these scenarios are not credible intruder scenarios. 

Results were provided for these scenarios as part of sensitivity analysis.

(2) This result is for all of the current and projected radiological inventory being placed in one vault.

(3) There are numerous numerical limits for protection of individuals during operations, the value shown is

the limit for protection of workers.

(4) NA = not applicable. Not calculated in the DOE PA.

not completed at the time the draft waste determination was submitted. Therefore, to

demonstrate conformance with the performance objectives, DOE assumed that all of the

radiological inventory expected to be distributed over the future vaults was located within the

existing Vault 4. A member of the public is assumed to use water from a well for domestic

purposes following the institutional control period. The well is assumed to be located where the

maximum concentration of radionuclides in the groundwater is predicted to occur. Modeled

contaminant concentrations are not diluted as a result of clean water entering the well if the

extraction of contaminated water from the well exceeds the amount of contaminated water

available. The groundwater pathway calculations were performed in two steps. First the flux to

the water table for parent radionuclides and progeny were estimated with vadose zone flow and

transport simulations. Changes to properties of the engineered cap were estimated through

additional analysis and input to the vadose zone flow and transport calculations. Using the

estimated fluxes from the vadose zone modeling, saturated zone flow and transport modeling

was used to estimate the groundwater concentrations at a hypothetical well. It is assumed that

contaminated water from the well is used for the following purposes: (1) drinking water, (2)

pond water (in which fish are raised and recreational activities occur), and (3) irrigation for

growing vegetables and raising livestock to provide meat and milk. The gaseous flux of

radionuclides to the land surface over time was estimated by DOE and used to calculate the

doses from direct plume shine, inhalation, and ingestion of vegetables, meat, and milk exposed

to airborne radioactivity. For most radionuclides the dominant pathways were water

consumption, vegetable consumption, and meat and milk consumption; however, important

pathways may be radionuclide- and scenario-specific. The dose assessment was more

complex than summarized here (Cook and Fowler, 1992).

The dose from all radionuclides that met the screening criteria and all pathways was computed

by assuming that the maximum concentration in contaminated groundwater and air occurred at

the same time for each radionuclide. The whole body dose to a member of the public was
1

 The dose methodology used in 10 CFR Part 61, Subpart C [based on International

Commission on Radiological Protection Publication 2 (ICRP 2)], is different than that used in the

newer ICRP 26. However, the resulting allowable doses are comparable, and DOE-SRS used

the newer methodology in ICRP 26.

- 66 -

 estimated to be 0.023 mSv/yr (2.3 mrem/yr) over the 10,000 year analysis period, which does

not exceed the 10 CFR Part 61 limit of 0.25 mSv/yr (25 mrem/yr) to the whole body.1

 Over 99

percent of the dose was from Se-79 and I-129, principally from the ingestion pathway. EPA

(1988) values for ingestion dose conversion factors were used to determine the doses to other

organs by determining the ratio of the organ dose conversion factors to the whole body factor

and multiplying by the known whole body dose (Rosenberger et al., 2005). The results

indicated that the dose to the thyroid would be 0.046 mSv/yr (4.6 mrem/yr) and the dose to any

other organ would be 0.053 Sv/yr (5.3 mrem/yr), which are below the performance objectives

of 0.75 mSv/yr (75 mrem/yr) and 0.25 mSv/yr (25 mrem/yr) for the thyroid and any other organ,

respectively.

As indicated in Section 4.2.1, DOE’s PA analysis was deterministic. Initially, limited sensitivity

analysis was performed to evaluate the impact of uncertainty and to identify key parameters or

conceptual models. In response to the RAI and subsequent meetings with NRC staff, DOE

completed an expanded sensitivity analysis to address a number of parameters, including

combinations of parameters that may represent alternative scenarios differing from DOE’s base

case scenario (WSRC, 2005a). NRC staff had suggested that the sensitivity cases be defined

such that they would provide an understanding of how reliant the dose results were on various

parameters, models, or assumptions, with special emphasis on the engineered cap and

vault/wasteform. Thirty five sensitivity cases were run by DOE and the main areas evaluated

were: (1) infiltration through the engineered cap, (2) the hydraulic conductivities of the vault and

saltstone and the rate at which they degrade, (3) the effective diffusion coefficients of the

saltstone and vault, (4) the impact of fracture flow through cracks in the vault, (5) the saturation

state of the vault and saltstone, and (6) the oxidation state of the vault and saltstone.

The SDF is a controlled release facility, such that proper closure to meet the objective of

limiting moisture through the waste and to limit deterioration of the wasteform is an integral part

of the long-term acceptability of the salt waste management strategy. DOE’s original base case

was based on the assumption that essentially no deterioration of the engineered cap or

wasteform over the 10,000 year analysis would occur; therefore, water available to the waste

was limited to 0.175 cm/yr (0.0689 in/yr). This small amount of infiltration represented less than

half a percent of projected precipitation, whereas for many vegetated humid sites natural

recharge values are commonly in the range of 15 - 40% of precipitation. Because the

engineered components of the system did not degrade, releases were limited to diffusive

releases and the doses to the public receptor through the groundwater pathway were limited to

1 x 10-5 mSv/yr (0.001 mrem/yr)(Cook and Fowler, 1992). It should be noted, that the result

from 1992 (i.e., 1 x 10-5 mSv/yr [0.001 mrem/yr]) is much lower than the current result (0.023

mSv/yr [2.3 mrem/yr]) because of three primary factors: (1) more degradation is assumed in the

current analysis, (2) the inventory in the current analysis is much larger, and (3) the current

analysis is an all-pathways dose assessment compared to the previous analysis which only

calculated the dose from drinking water ingestion. DOE’s revised base case did consider

deterioration of the engineered system (e.g., engineered cap and vault/wasteform) over time

(WSRC, 2005a).
- 67 -

The current closure concept has a geosynthetic cover system. After the institutional control

period, infiltration is predicted to gradually increase over time as the closure system degrades,

primarily from the intrusion of deep-rooted plants (e.g., pine trees) and siltation of drainage

layers (e.g., from colloidal clay migration). The engineered cap has an upper GCL and a lower

GCL. The upper GCL was assumed to degrade over time by pine tree penetration (Phifer and

Nelson, 2003). Root penetration of the GCL by pine trees was simulated to result in holes

through the GCL that would act as direct water pathways for infiltration from the upper drainage

layer to the lower backfill layer. After precipitation events, saturated conditions were predicted

to occur around the holes such that a radius of influence would form that would be much

greater than the radius of the hole itself. As a result, a small area of holes in the GCL can

greatly reduce the lateral flow of water in the drainage layer and increase the vertical flow of

water into the lower backfill. Holes covering only approximately 0.3% of the GCL resulted in

infiltration near that of typical background infiltration. DOE estimated that infiltration through the

upper GCL would be 1.6 cm/yr (0.63 in/yr) shortly after closure, increase to approximately 35

cm/yr (14 in/yr) by year 2000 after closure, and would remain relatively constant until year

10,000. Alternate land use scenarios resulted in a range of infiltration rates after degradation of

the upper GCL (i.e., 16 to 54 cm/yr [6.3 to 21 in/yr]).

As discussed in Sections 4.2.6 and 4.2.8, the vault and saltstone are designed to have a very

low hydraulic conductivity to greatly limit releases into the natural system and eventually into the

groundwater aquifer. Based on tests performed at Core Laboratories, the hydraulic conductivity

of the vault and saltstone are assigned values of 1 x 10-12 cm/s (1 x 10-6 ft/yr) and 1 x 10-11 cm/s

(1 x 10-5 ft/yr), respectively (Yu et al., 1993). Fly ash is added to the saltstone to lower the

redox potential of the pore water, thereby effectively eliminating the risk from Tc-99 if reducing

conditions are maintained. In common cementitious systems, Tc-99 has a Kd of 1000 ml/g

under reducing conditions but a Kd of approximately 1 ml/g under oxidizing conditions (Bradbury

and Sarott, 1995). Instead of attempting to predict the long-term evolution of the hydrologic

properties of the vault and saltstone, DOE used professional judgment to assign degraded

values for hydraulic conductivity of the vault and saltstone of 1 x 10-9 cm/s (1 x 10-3 ft/yr) at

10,000 years. The timing for the increase in hydraulic conductivity was implemented as multiple

step functions in the PA. Sensitivity analysis was used to identify the importance of the

assigned value of the degraded hydraulic conductivity to meeting the performance objective.

Table 8 provides a summary of sensitivity runs performed by DOE, and Table 9 is a companion

table that provides a more detailed description of each analysis. The results in Table 8 are for

the current and projected Vault 4 inventory based on DOE’s proposed approach to waste

treatment (WSRC, 2005b). As previously indicated, 35 sensitivity scenarios were run. The

main areas evaluated were: infiltration through the engineered cap, the hydraulic conductivities

of the vault and saltstone and the rate at which they degrade, the effective diffusion coefficients

of the saltstone and vault, the impact of fracture flow through cracks in the vault, the saturation

state of the vault and saltstone, and the oxidation state of the vault and saltstone. Because the

analysis approach was deterministic, combinations of uncertainties were evaluated in order to

better assess their impact. For the first nineteen sensitivity runs performed, four radionuclides

were evaluated: H-3, C-14, Se-79, and I-129. The additional sixteen runs had additional

radionuclides evaluated depending on the goal of the analysis. The additional radionuclides

considered were: Tc-99, U-238, Np-237, Pu-238, Pu-239, Cs-137 and Sr-90, although not all of

these radionuclides were considered in each simulation. The radionuclides considered by DOE

were limited because of the additional effort required to verify the accuracy of the results.
- 68 -

Table 8 Summary of DOE Sensitivity Analyses for Vault 4

Scenario Description Result

(mrem/yr)

RadionuclidesLargest

Contributors

1 Base case 0.048 Se-79, I-129

2 Optimistic cover degradation 0.02 Se-79

3 Pessimistic cover degradation 0.27 Se-79, I-129

4 Optimistic vault degradation 0.031 Se-79

5 Pessimistic vault degradation 0.051 Se-79, I-129

6 Optimistic initial vault conductivity 0.032 Se-79

7 Pessimistic initial vault conductivity 0.051 Se-79, I-129

8 Optimistic saltstone grout degradation 0.037 Se-79

9 Pessimistic saltstone grout degradation 0.23 Se-79, I-129

10 Optimistic initial saltstone grout conductivity 0.038 Se-79

11 Pessimistic initial saltstone grout conductivity 0.24 Se-79, I-129

12 Pessimistic cover, vault, grout degradation 4.0 I-129, Se-79

13 Vault and grout saturated 0.18 Se-79, I-129

14 Optimistic vault diffusion 0.035 Se-79, I-129

15 Pessimistic vault diffusion 0.17 Se-79, I-129

16 Optimistic saltstone grout diffusion 0.039 Se-79, I-129

17 Pessimistic saltstone grout diffusion 0.064 Se-79

18 Pessimistic vault/saltstone grout diffusion 0.67 Se-79, I-129

19 Pessimistic Kd values 36 C-14, Se-79, I-129

20 Reducing conditions in vault at all times. Only

Tc-99 was considered in the analysis

1.6 x 10-13 Tc-99

21 Oxidizing conditions (Kd Tc-99 = 1 mL/g). Only

Tc-99 was considered in the analysis 

3.2 Tc-99

22 Oxidizing conditions (Kd Tc-99 = 0 mL/g). Only

Tc-99 was considered in the analysis 

90 Tc-99

23 Final Kh vault/saltstone to 1 x 10-6 cm/s (1ft/yr) 16 I-129, Se-79, Tc-99

24 Increased precipitation (climate change) 1.1 Se-79, I-129

25 Increased precipitation, pessimistic vault and

saltstone degradation

6.5 I-129, Se-79

26 Decreased Kd values by a factor of 10 1.8 I-129, Se-79

27 Pessimistic initial saltstone grout conductivity and

pessimistic saltstone grout degradation

0.43 I-129, Se-79

28 Pessimistic initial vault hydraulic conductivity 0.11 Se-79, I-129

29 Ox 100% loss of reducing conditions. Only Tc-99

and U-238 were included in the analysis

(hydraulic parameters equivalent to base case

values)

3.2 Tc-99
Scenario Description Result

(mrem/yr)

RadionuclidesLargest

Contributors

- 69 -

29 Red 0% loss of reducing conditions. Only Tc-99 and

U-238 were included in the analysis (hydraulic

parameters equivalent to base case values)

1.6 x 10-13 Tc-99

30 Oxidizing saltstone and grout combined with

increased infiltration and pessimistic vault and

saltstone grout degradation. Only Tc-99 and U238 were included in the analysis

1200 Tc-99

31 Vault/grout 100% saturated, reducing conditions 3.5 Se-79

32 Vault/grout 100% saturated, oxidizing conditions 26 Tc-99, Se-79

33 Ox Increased infiltration, hydraulic conductivity,

diffusivity. 100% oxidizing conditions

34,000 Tc-99, I-129, Se-79

33 Red Increased infiltration, hydraulic conductivity,

diffusivity. 100% reducing conditions

310 I-129, Se-79, Tc-99
- 70 -

Table 9 Description of DOE Sensitivity Analyses

Scenario Description

1 Base case

2 Optimistic cover degradation - Peak infiltration from 36 cm/yr (14 in/yr) to 18 cm/yr (7

in/yr)

3 Pessimistic cover degradation - Peak infiltration from 36 cm/yr (14 in/yr) to 53 cm/yr (21

in/yr)

4 Optimistic vault degradation - Saturated hydraulic conductivity of the vault changes from

1 x 10-12 to 1 x 10-10 cm/s (1 x 10-6 to 1 x 10-4 ft/yr) over 10,000 years

5 Pessimistic vault degradation - Saturated hydraulic conductivity of the vault changes

from 1 x 10-12 to 1 x 10-8 cm/s (1 x 10-6 to 1 x 10-2 ft/yr) over 10,000 years

6 Optimistic initial vault conductivity - Initial saturated hydraulic conductivity of the vault

set to 1 x 10-13 cm/s (1 x 10-7 ft/yr)

7 Pessimistic initial vault conductivity - Initial saturated hydraulic conductivity of the vault

set to 1 x 10-11 cm/s (1 x 10-5 ft/yr)

8 Optimistic saltstone grout degradation - Saturated hydraulic conductivity of the grout

changes from 1 x 10-11 to 1 x 10-10 cm/s (1 x 10-5 to 1 x 10-4 ft/yr) over 10,000 years

9 Pessimistic saltstone grout degradation - Saturated hydraulic conductivity of the grout

changes from 1 x 10-11 to 1 x 10-8 cm/s (1 x 10-5 to 1 x 10-2 ft/yr) over 10,000 years

10 Optimistic initial saltstone grout conductivity - Initial saturated hydraulic conductivity of

the saltstone grout set to 1 x 10-12 cm/s (1 x 10-6 ft/yr)

11 Pessimistic initial saltstone grout conductivity - Initial saturated hydraulic conductivity of

the saltstone grout set to 1 x 10-10 cm/s (1 x 10-4 ft/yr)

12 Pessimistic cover, vault, grout degradation - Peak Infiltration from 14 in/year to 21

in/year. Saturated hydraulic conductivity of the vault goes from 1 x 10-12 to 1 x 10-8

cm/sec (1 x 10-6 to 1 x 10-2 ft/yr) over 10,000 years. Saturated hydraulic conductivity of

saltstone grout changes from 1 x 10-11 to 1 x 10-8 cm/sec over 10,000 years (1 x 10-5 to

1 x 10-2 ft/sec)

13 Vault and grout saturated - Relative hydraulic conductivity set to 1 for both the vault and

saltstone grout

14 Optimistic vault diffusion - Vault diffusion coefficient set to 1 x 10-9 cm2

/s (3 x 10-5 ft2/yr)

15 Pessimistic vault diffusion - Vault diffusion coefficient set to 1 x 10-7 cm2

/s

16 Optimistic saltstone grout diffusion - Saltstone diffusion coefficient set to 5 x 10-10 cm2

/s

(2 x 10-5 ft2

/yr)

17 Pessimistic saltstone grout diffusion - Saltstone diffusion coefficient set to 5 x 10-8 cm2

/s

(2 x 10-3 ft2

/yr) 

18 Pessimistic vault/saltstone grout diffusion - Vault diffusion coefficient set to 1 x 10-7

cm2/s, Saltstone grout diffusion coefficient set to 5 x 10-8 cm2/s (2 x 10-3 ft2/yr)

19 Pessimistic Kd values - Partition coefficients for C-14, Se-79, and I-129 set to 0 ml/g

20 Reducing conditions in vault at all times - Tc Kd set to 1000 ml/g. Only Tc-99 was

included in the analysis

21 Oxidizing conditions (Kd Tc-99 = 1 ml/g). Only Tc-99 was included in the analysis
Scenario Description

- 71 -

22 Oxidizing conditions (Kd Tc-99 = 0 ml/g). Only Tc-99 was included in the analysis

23 Final Kh vault/saltstone to 1 x 10-6 cm/s (1 ft/yr) - Concrete Kh increases from 1 x 10-12

to1 x 10-6 cm/s (1 x 10-6 to 1 ft/yr) with a degradation rate constant, α = 3. Saltstone

grout Kh increases from 1 x 10-11 to 1 x 10-6 cm/s (1 x 10-5 to 1 ft/yr) with a degradation

rate constant, α = 2.5.

24 Increased precipitation (climate change) - Increase the average precipitation used in the

base case (i.e., 124 cm/yr [48.8 in/yr]) by 25%.

25 Increased precipitation, pessimistic vault and saltstone degradation - combined

scenarios 5, 9, and 24

26 Decrease Kd values in all media by a factor of 10

27 Pessimistic initial saltstone grout conductivity and pessimistic saltstone grout

degradation - combined scenarios 9 and 11

28 Pessimistic initial vault hydraulic conductivity - Concrete Kh decreases from 1 x 10-8

cm/s to 1 x 10-6 cm/s (1 x 10-2 to 1 ft/yr) over 10,000 years

29 Ox 100% loss of reducing conditions - Oxidizing Kd values for Tc and U (Bradbury and

Sarott, 1995). Only Tc-99 and U-238 were included in the analysis

29 Red 0% loss of reducing conditions (base case redox assumption) - Only Tc-99 and U-238

were included in the analysis. Reducing Kd values for Tc and U (Bradbury and Sarott,

1995)

30 Oxidizing saltstone and grout combined with increased infiltration and pessimistic vault

and saltstone grout degradation - combined scenarios 25 and 29. Only Tc-99 and U238 were included in the analysis

31 Vault/grout 100% saturated, reducing conditions. Assume the vault and saltstone grout

exhibit large-scale cracking at a 9.1 m (30 ft) nominal spacing. Assume vault, saltstone

grout, and fractures are fully saturated. Implemented by redefining the water retention

and relative permeability curves, such that both are 1.0 regardless of suction head.

32 Vault/grout 100% saturated, oxidizing conditions - Same as scenario 31 except for

oxidizing conditions

33 Ox Increased infiltration, hydraulic conductivity, diffusivity. Oxidizing conditions. Infiltration

to the vault is set to 25 cm/yr (9.8 in/yr) throughout the simulation and the closure cap

drains silted to allow infiltration to go to the saltstone grout. Hydraulic conductivity of

the vault and saltstone grout are set to 5 x 10-7 cm/s (5 x 10-1 ft/yr) throughout the

simulation. Effective diffusivity for the vault and saltstone grout are increased by a

factor of 10 over the base case. 100% oxidizing conditions at the start of the

simulation.

33 Red Increased infiltration, hydraulic conductivity, diffusivity. Reducing conditions. Same as

scenario 33 Ox but for reducing conditions throughout the simulation.
- 72 -

DOE’s interpretation of the results in Table 8 are that the Vault 4 special analysis and the

results of the sensitivity studies clearly demonstrate that the solidified salt waste from the

projected waste stream for Vault 4 can be disposed in Vault 4 and that all applicable

performance measures, including the 0.25 mSv/yr (25 mrem/yr) all-pathways public dose limit,

will be met. DOE concluded that the results demonstrate that the saltstone disposal system is

very robust in that the parameters which are hardest to measure or predict have relatively little

impact on the calculated dose. The results of the sensitivity analyses show that the projected

dose from the disposal of saltstone is most dependent on the amount of precipitation and

resultant infiltration of water through the saltstone grout and the oxidation-reduction condition of

the vault and saltstone grout. These parameters are the ones that DOE has a high degree of

confidence will perform as described in the special analysis (Cook et al, 2005). DOE has

indicated that it is important that the models provide a realistic representation of the physical

and chemical processes that will occur within the saltstone disposal vaults and surrounding

environment for the next 10,000 years (WSRC, 2005a).

4.2.16 NRC Evaluation – Protection of the Public

DOE has used an all pathways dose assessment to show conformance with the performance

objectives established for the public. As indicated in Section 4.2.15, DOE generated results

assuming all of the inventory (both existing and projected) would be placed within Vault 4 (0.023

mSv/yr [2.3 mrem/yr]) as well as for the expected Vault 4 inventory (0.00048 mSv/yr [0.048

mrem/yr]). The former case represents an unrealistic scenario that was generated to

demonstrate that the performance objectives could be achieved for the disposal system

regardless of how the radiological inventory may be distributed in the disposal facility. The

latter scenario represents the risk from Vault 4 if DOE were to proceed with the waste treatment

and disposal process as currently envisioned. In both cases, the physical and chemical

retention capabilities of the system are identical, the only difference is the projected inventory. 

The base case peak Total Effective Dose Equivalent (TEDE) to a member of the public of 0.023

mSv/yr (2.3 mrem/yr) is well within the performance objective of 0.25 mSv/yr (25 mrem/yr) in 10

CFR Part 61.41 (Protection of General Population from Releases of Radioactivity). There is a

substantial difference between the total inventory projected for the SDF and that projected for

Vault 4, as discussed below. 

Overall, DOE’s calculational methodology for performing the PA to evaluate protection of the

public is generally consistent with NRC guidance. There are a few areas related to the

approach to PA where there is some divergence between NRC and DOE; those areas are

discussed in detail in the paragraphs that follow. The selection of the receptor as a future

resident that may perform agricultural activities at the site after institutional controls are

assumed to cease (100 years) is reasonable considering past regional land use practices prior

to development of the SRS. Assuming a member of the public will use water from a well for

domestic purposes and that the well is assumed to be located where the maximum

concentration of radionuclides in the groundwater are predicted to occur is consistent with

NRC’s definition of an average member of the critical group. The groundwater pathway

calculations were performed in two steps. The stepwise approach to performing the

groundwater calculations is reasonable assuming quality controls are implemented to ensure

that errors are not introduced. NRC staff did not identify any errors of this type during its 

review.
- 73 -

The analysis performed by DOE was deterministic. The approach by DOE to evaluate

combinations of uncertain parameters with a deterministic analysis is reasonable; however, use

of a probabilistic approach can provide many insights into this type of analysis (NRC, 2000b). 

DOE’s model support is limited in a number of areas that the sensitivity analyses identified as

impacting system performance, as discussed in detail below. 

DOE=s initial PA provided limited sensitivity analyses. NRC requested an expansion of the

sensitivity analyses in its RAI (NRC, 2005a). Because the PA analyses were deterministic,

DOE provided a series of analyses to evaluate the impact of key uncertainties. For complex

modeling and long-term projections of system performance, it is essential that uncertainties are

evaluated and their impacts factored into decision making (NRC, 2000b). In general, if

deterministic modeling is used, it should be reasonably conservative such that a subject matter

expert, with minimal interaction with those who performed the assessment, could come to a

conclusion that the analysis was conservative. The key uncertainties evaluated by DOE were

infiltration through the engineered cap, the hydraulic conductivities of the vault and saltstone

and the rate at which they degrade, the effective diffusion coefficients of the saltstone and

vault, the impact of fracture flow through cracks in the vault, the saturation state of the vault and

saltstone, and the oxidation state of the vault and saltstone. Based on current information,

NRC agrees that DOE has evaluated the key uncertainties.

Before discussing NRC’s interpretation of DOE’s sensitivity analysis results, it is important to

address the inventory used in the sensitivity analysis and the potential for multiple sources. 

Table 10 provides a comparison of the inventory values for the three radionuclides that appear

to be most significant to achieving protection of the public through the groundwater pathway. 

As shown in Table 10, the inventory used by DOE in the sensitivity analysis calculations is

projected to be approximately a factor of 3 less than would be expected for the average vault

containing salt waste from the tanks. The purpose of the sensitivity analysis was to address

uncertainties and provide justification that DOE’s salt waste disposal activities could meet the

performance objectives; therefore, the use of estimated Vault 4 inventory was inappropriate. 

As previously indicated, as many as 14 additional vaults would be needed to dispose of

saltstone material. DOE’s groundwater modeling to assess the impact from all of the vaults,

considering the potential for overlap from plumes of individual vaults at any given potential

receptor location, was not completed at the time of the draft waste determination. DOE

indicated during the August 17-18, 2005, meeting that their professional opinion was that the

impact on groundwater concentrations for the public receptor for the entire SDF would be

approximately a factor of 2 larger than for an individual vault (Cook, 2005). 

Considering the differences between the projected Vault 4 inventory and the average vault, and

the potential for multiple sources (plume overlap from multiple vaults), the sensitivity results

presented in Table 8 need to be scaled, at a minimum, to increase by a factor of 6.6 before the

significance of the results is interpreted. The 6.6 value is derived by applying a factor of 3.3 for

the difference in inventory between the values for projected Vault 4 and the average vault and a

factor of 2 to account for overlap of the plumes from multiple disposal vaults. Ideally, these

scaled results would have been generated by model calculations. However, in the absence of

model results a simple linear scaling should be sufficient for discussion purposes, given the 
- 74 -

Table 10 Comparison of Projected Vault 4 Inventory, Total Estimated SDF Inventory, 

 and Average Vault Inventory 

Radionuclide Projected Vault 4

(Ci)

Total SDF

(Ci)

Average Vault(1)

(Ci) Ratio(2)

Se-79 1.96 89.4 6.4 3.3

I-129 0.44 18.0 1.3 3.0

Tc-99 716 33,100 2360 3.3

(1) The average vault value is generated by dividing the total SDF inventory by the projected number of

vaults (14).

(2) Ratio = Average Vault / Projected Vault 4

current modeling approach to release from the source term (i.e., doses should scale somewhat

linearly with inventory when using a Kd-based release model). Table 11 is a list of the sensitivity

scenarios provided and the scaled dose result based on the above considerations.

The demonstration that the SDF can meet the performance objective for protection of the public

is dependent on the long-term (10,000 year) performance of the engineered closure cap to

greatly limit infiltration and the concrete vaults and cementitious saltstone wasteform to limit

radionuclide release. The initial analysis performed by DOE had assumed essentially no

degradation of the main engineered components of the disposal system (engineered cap,

vaults, cementitious wasteform) over 10,000 years in the base case analysis (Cook and Fowler,

1992). In response to the NRC RAI and subsequent meetings with NRC staff, DOE performed

analyses to evaluate various degrees of degradation of the engineered components of the

disposal system (WSRC, 2005a, b). In providing an interpretation of the sensitivity results,

DOE indicated that it considers many of the analyses to represent highly unrealistic conditions. 

In some instances, NRC staff agrees with the DOE discussion of the likelihood of the sensitivity

scenario (i.e., some are clearly unrealistic based on current information and understanding). 

The major difficulty in interpreting the DOE sensitivity analysis results is the mixture of

potentially optimistic and pessimistic conditions or assumptions, and the amount of support

provided for the modeling results. Roughly a third of scaled results of the scenarios evaluated

in the sensitivity analyses approach or exceed the 0.25 mSv (25 mrem/yr) dose limit, some by a

large margin. Of these, 70% represent analyses where 100% oxidizing conditions were

assumed at the time of facility closure. The following bullets provide a synthesis of NRC’s

assessment of the realism of DOE’s sensitivity scenarios.

• NRC agrees with DOE that the assumption of 100% oxidation of the waste at year zero

is an unrealistic assumption. Scaling of the sensitivity results to the amount of oxidation

expected by DOE over 10,000 years (3 to 8%) is confounded by the fact that this

amount of oxidation is projected for the next 10,000 years (i.e. it will not occur

instantaneously) and will likely occur at exposed surfaces rather than throughout the

bulk of the material. Assuming 0% oxidation (DOE base case) is clearly nonconservative while assuming 100% oxidation is unrealistic.
- 75 -

Table 11 Scaled Results for the Sensitivity Scenarios

Scenario Description

Result

(mrem/yr)

1 Base case 0.32

2 Optimistic cover degradation 0.13

3 Pessimistic cover degradation 1.8

4 Optimistic vault degradation 0.20

5 Pessimistic vault degradation 0.34

6 Optimistic initial vault conductivity 0.21

7 Pessimistic initial vault conductivity 0.34

8 Optimistic saltstone grout degradation 0.24

9 Pessimistic saltstone grout degradation 1.5

10 Optimistic initial saltstone grout conductivity 0.25

11 Pessimistic initial saltstone grout conductivity 1.6

12 Pessimistic cover, vault, grout degradation 26

13 Vault and grout saturated 1.2

14 Optimistic vault diffusion 0.23

15 Pessimistic vault diffusion 1.1

16 Optimistic saltstone grout diffusion 0.26

17 Pessimistic saltstone grout diffusion 0.42

18 Pessimistic vault/saltstone grout diffusion 4.4

19 Pessimistic Kd values 240

20 Reducing conditions in vault at all times. Only Tc-99 was included

in the analysis 

1.0 x 10-12

21 Oxidizing conditions (Kd Tc-99 = 1 mL/g). Only Tc-99 was included

in the analysis 

21

22 Oxidizing conditions (Kd Tc-99 = 0 mL/g). Only Tc-99 was included

in the analysis 

590

23 Final Kh vault/saltstone to 1 x 10-6 cm/s (1 ft/yr) 110

24 Increased precipitation (climate change) 7.3

25 Increased precipitation, pessimistic vault and saltstone

degradation

43

26 Decreased Kd values by a factor of 10 12

27 Pessimistic initial saltstone grout conductivity and pessimistic

saltstone grout degradation

2.8

28 Pessimistic initial vault hydraulic conductivity 0.73

29 Ox 100% loss of reducing conditions. Only Tc-99 and U-238 were

included in the analysis

21
Scenario Description

Result

(mrem/yr)

- 76 -

29 Red 0% loss of reducing conditions (base case redox assumption). 

Only Tc-99 and U-238 were included in the analysis

1.0 x 10-12

30 Oxidizing saltstone and grout combined with increased infiltration

and pessimistic vault and saltstone grout degradation. Only Tc-99

and U-238 were included in the analysis

7900

31 Vault/grout 100% saturated, reducing conditions 23

32 Vault/grout 100% saturated, oxidizing conditions 170

33 Ox Increased infiltration, hydraulic conductivity, diffusivity. 100%

oxidizing conditions.

220,000

33 Red Increased infiltration, hydraulic conductivity, diffusivity. 100%

reducing conditions.

2000

• The amount and type of degradation experienced by the saltstone wasteform over 10,000

years was evaluated in a number of analyses. The amount of degradation assigned by

DOE was based on professional judgment, informed by qualitative consideration of

degradation mechanisms and other factors. In DOE’s base case, the hydraulic conductivity

of the saltstone was assumed to change from 1 x 10-11 cm/s to 1x 10-9 cm/s (1 x10-5 to 1

x10-3 ft/yr), the effective diffusivity remained unchanged throughout the analysis at 5 x 10-9

cm2/sec (2 x 10-4 ft2/yr), and the wasteform experienced no oxidation. In scenario 25, the

hydraulic conductivity of the saltstone and vault were assumed to degrade to 1 x 10-8 cm/s

(1 x 10-2 ft/yr) and the precipitation was increased by 25%. Even with no oxidation of the

wasteform, the scaled result is 0.43 mSv/yr (43 mrem/yr), which would exceed the 0.25

mSv/yr (25 mrem/yr) performance objective. This result demonstrates that proper design of

the engineered barriers and careful implementation of the design is essential to meeting the

performance objective for protection of the public. In addition, the modeling results or

assumptions for the degradation of the saltstone, vaults, and engineered cap must be

supported by adequate information.

• The initial values for the hydraulic conductivity of the saltstone and vaults are based on very

limited measurements of small-scale samples. The thermal and mechanical conditions

experienced by the field-scale saltstone wasteform during curing may be significantly

different than the conditions experienced by the samples that have been tested. While the

sensitivity analyses demonstrate that the model results are not very sensitive to the initial

values of hydraulic conductivities, testing performed on earlier saltstone formulations exhibit

much larger hydraulic conductivities and possibly indicate temporal variability associated

with curing time (Malek et al., 1985).

• Fracture flow was assessed in scenario 31 for no oxidation of the waste and scenario 32 for

complete oxidation of the waste. Fractures extending through the vaults were introduced

with a 9 m (30 ft) spacing parallel to the short axis of the vault based on analysis performed

of settlement and seismic effects (Peregoy, 2003). The fractures, saltstone, and vaults

were assumed to be saturated. The scaled result for scenario 31 and 32 were 0.23 mSv/yr
- 77 -

(23 mrem/yr) and 1.7 mSv/yr (170 mrem/yr), respectively. Assuming 100% saturation of the

system and complete oxidation of the waste is not realistic.

• Some degree of natural (non-human induced) climate change can be expected to occur

over the next 10,000 years at the SRS. Scenario 24 demonstrates that the current model

results are sensitive to estimates of infiltration through the engineered cap. Assuming a

25% increase in precipitation at year zero is unrealistic. Assuming no change in the climate

(considering the model sensitivity) over the next 10,000 years (DOE base case) is nonconservative.

• Relative permeability values developed for the saltstone and vaults were determined to

have been obtained from experimental results that are likely in error. Sensitivity scenario 13

assumed that the vault and grout were saturated throughout the analysis. This assumption

is unrealistic. The impacts of using the questionable data was bound by a factor of 4.

Scenario 33 was completed at the request of NRC. NRC staff wanted to understand what the

conservative upper bound for the risk to the public from the disposal of saltstone may be. The

scenario combined an infiltration rate of 25 cm/yr through the upper GCL, an assumption of

complete siltation of the lower drainage layer of the engineered cap, an assumed value for

hydraulic conductivity of the vault and saltstone of 5 x 10-7 cm/s (5 x 10-1 ft/yr), an increase of

effective diffusivity of the vault and saltstone by a factor of 10, and either completely reducing or

completely oxidizing conditions. DOE describes this scenario as “extremely conservative in that

it represents a disposal system that has no closure cap and no vault and in which the saltstone

grout had properties similar to SRS sandy clay soil” (WSRC, 2005a). The NRC staff agrees

that it is overly conservative to assume that complete oxidation of the saltstone will occur, but

does not believe that this scenario is extremely conservative if it is assumed there is no

oxidation. NRC staff interprets this analysis to represent a disposal system that has an

engineered cap that has not performed as designed, resulting in an infiltration rate consistent

with observations of natural recharge rates at the site. The hydraulic conductivity and effective

diffusivity values used for the vault and saltstone do not represent a system with no vault and

grout having properties of SRS sandy clay soil, but instead represent degraded cementitious

materials with properties that are roughly 100 times better than a SRS sandy clay soil and

closer to a controlled compacted kaolin or a natural clay deposit (Phifer and Collard, 2003). 

The analysis demonstrates that DOE must appropriately design the engineered systems, and

develop the model support to confirm that it will perform as designed, or, based on current

modeling, it is questionable that the performance objective for protection of the public can be

met. The analysis also demonstrates that more realistic modeling of infiltration, water contact

with the waste, waste oxidation, and radionuclide release in an unsaturated and potentially

fractured system is needed, because conservative modeling does not yield acceptable results.

In (WSRC, 2005a), DOE provided normalized sensitivity as a metric to interpret the sensitivity

analysis results. Using the normalized sensitivity measure eliminated the effects of units and

the absolute magnitude of each parameter. However, the challenge with using normalized

sensitivity is that the results can be influenced by the range of the parameter that is evaluated. 

For example, the dose response to a parameter may not be linear. Therefore if an arbitrary

large range is selected, the sensitivity measure can be in effect diluted. Sensitivity should be

considered with respect to the level of knowledge of the range of the parameter being

investigated to avoid this impact if possible.
- 78 -

The deterministic DOE base case result does not have adequate model support and is not 

conservative. Considering the uncertainty in many key parameters, it should not be used as the

basis for demonstrating that there is reasonable assurance that the performance objective for

protection of the public can be met or for developing inventory limits. A revised base case

should be based on the projected average vault inventory for salt waste from the tanks and the

orientation of multiple disposal vaults, include the expected magnitude and timing of climate

change from the natural cycling of climates, include the expected magnitude and rate of

oxidation of waste, consider liquid and gas flow in fractures (that may develop), and account for

the questionable moisture characteristic curve information for concrete and saltstone that was

used in the previous analysis. 

Model support is essential for deterministic modeling that is not clearly conservative because

uncertainties are not represented in a deterministic analysis. Model support is information that

provides confidence that the numerical model results are adequately representing the behavior

of the actual system. Model support is commonly called model validation in the technical

literature, but NRC recognizes that some aspects of PA are not amenable to scientific model

validation (e.g., the long time frames) (NRC, 2000b). Model or software verification is a

process to ensure that the numerical calculations are correct, but it does not ensure that the

model results are appropriate for the given application. It does appear that DOE has performed

adequate software verification for the main products it used to complete the PA (e.g.,

PORFLOW, HELP, internally developed intruder analysis software). NRC requested additional

model support for key areas of DOE’s PA in the RAI (NRC, 2005a). DOE stated that the

information provided in (WSRC, 2005b) encompasses all currently available field data providing

model support for the simulation results. A path forward that outlined how DOE intended to

develop adequate model support was discussed in the August 17, 2005, meeting; however, a

path forward was not provided in the written response to the action items (WSRC, 2005a). 

The following example is intended to illustrate the types of information that could be developed

to provide adequate model support. One of the key elements of DOE’s PA is the reduction of

Tc-99 in the wasteform by the addition of slag. As previously discussed, the sensitivity

analyses demonstrate quite clearly that the rate and extent of oxidation of the wasteform is a

key factor in meeting the protection of the public performance objective. DOE has performed

basic research to evaluate whether the slag would result in Tc-99 being contained in a reduced

form, and installed field-scale saltstone lysimeter tests with and without slag (Cook and Fowler,

1992). The slag-based lysimeters showed a much lower release rate of Tc-99 compared to the

lysimeters without slag. At the request of NRC, DOE provided a comparison of a PORFLOW

simulation with analytical data, that showed an underprediction by the model of roughly a factor

of 10 (WSRC, 2005a). Figure 7 provides the comparison of the PORFLOW and slag lysimeter

data. DOE explained the underprediction was the result of the analytical measurements being

performed for gross non-volatile beta-gamma emitters, and therefore the results show

essentially background. However, the data shows an increasing trend starting around day 250

that is consistent with the PORFLOW simulation, therefore it is unclear that the analytical data

represents background. In addition to providing information on field-scale release rates, the

slag lysimeters may contain valuable information about the rate of oxidation of the saltstone. 

Currently, DOE’s estimates for the amount of oxidation of the saltstone over 10,000 years are

based primarily on numerical modeling results. It may be possible to exhume and characterize

a saltstone lysimeter. The depth of the penetration of the oxidation front should be able to be

estimated and it would provide excellent model support for a key element of DOE’s PA.
- 79 -

Figure 7 Comparison of PORFLOW and Slag Lysimeter Data

In order to better risk-inform the staff=s review, the NRC staff developed a PA model applicable

to the SDF using the software platform GoldSim (Kossik, 2005). The model was used to

evaluate sensitivity of the PA results and to corroborate, in a general sense, the DOE-SRS

calculational results. Utilization of the model by the staff allowed a more focused review on

those technical aspects of the problem more likely to influence the risks. With similar input

values for parameters and assumptions, the output from the simulations in NRC’s independent

analysis was in general agreement with DOE’s PA.

NRC staff concludes that there is reasonable assurance that the performance objectives in 10

CFR 61.41 can be met, including the provision of as low as reasonably achievable (ALARA)

releases of radioactivity to the general environment, if the following conditions are satisfied: (1)

DOE will perform more realistic modeling of waste oxidation and moisture movement with

results that demonstrate release rates are acceptable, (2) DOE will perform confirmatory

activities that demonstrate the hydraulic conductivity of the saltstone and vault will not degrade

to values larger than 1 x 10-7 cm/s (1 x 10-1 ft/yr), and (3) DOE will enhance model support for

the areas discussed above. The ALARA provision is not part of the PA calculation, as the PA is

the means to generate results to compare to performance objectives. Through demonstration

of Criterion Two (the waste has had highly radioactive radionuclides removed to the maximum
- 80 -

extent practical), DOE can satisfy the intent of the ALARA provision to maintain releases of

radioactivity to the environment as low as reasonably achievable.

4.2.17 Protection of Intruders

In estimating doses to inadvertent intruders to the site after the 100 year period of institutional

control, it is assumed that individuals could establish a residence and that the intruders have no

previous knowledge of waste disposal activities at the site. For assessing whether the

performance objective of 10 CFR 61.42 could be met, DOE’s analysis used the projected

inventory for Vault 4 (Rosenberger, et al., 2005). DOE indicated that the Vault 4 inventory was

expected to bound the inventory for other vaults because the dominant radionuclide for the

intruder scenarios was Cs-137 and the DDA waste stream, which has the highest Cs-137

concentration, will be located in Vault 4. Direct intrusion into the SDF facility was assumed to

occur under one of three scenarios: agriculture, resident, and post-drilling. Acute and chronic

exposures were considered. However, only chronic exposures were evaluated for the SDF

because of the longer exposure times associated with chronic scenarios. 

Figure 8 is a visualization of the resident intruder pathway. The resident intruder scenario is

assumed to be a credible occurrence at any time after the active institutional controls are

relinquished. It is expected that institutional controls of various form (active, passive) will be

maintained at the site longer than the 100 years assumed in the analysis. Surveillance and

maintenance will be used during the period of institutional controls to repair any degradation to

the engineered cap. In the resident intruder scenario, the intruder is assumed to excavate a

foundation for a home to a maximum depth of 3 m (10 ft). Over time erosion will lower the

ground surface until the erosion barrier layer of the engineered cap becomes exposed. Thus

the amount of shielding will decrease over time and radioactive decay of long-lived

radionuclides will produce increasing quantities of daughter products. The intruder analysis

was performed in ten year steps from year 100 to year 10,000 after closure. The inadvertent

intruder analysis was performed by an automated code developed by Savannah River National

Laboratory (SRNL) (Koffman, 2004). 

The erosion barrier layer has been designed to maintain a 3 m (10 ft) thickness of soil over the

top of the vaults for 10,000 years (Cook, et al., 2005). The erosion barrier is to be constructed

of durable material sized to remain in place during the maximum rainfall event expected over

the 10,000 year design (Phifer and Nelson, 2003). Sections 4.2.4 and 4.2.5 provide a detailed

discussion of the erosion control design. The agricultural scenario is not possible if the erosion

barrier performs as designed, because the soil layer prevents waste from being exhumed.

The reinforced concrete vault roof and saltstone waste form are assumed to present a barrier

that would discourage drilling activities, and therefore drilling is not considered a plausible

scenario. The thick cover system provides protection from physical weathering. Initially and for

many years after construction, the vaults and wasteform will present a dramatically more

difficult media through which to drill. The concrete and saltstone grout will undergo chemical

degradation over time which will slowly alter the properties of the materials. However it is

expected that even altered concrete and saltstone grout will present a sharp contrast to the

native sands and clays. The well-driller intruder is not deemed credible for the SDF based on

the long-term physical properties of the materials and the regional practices for well

development. 
- 81 -

Figure 8 Resident Intruder Scenario

The projected Vault 4 inventory and the inventory limit for each radionuclide was used to

estimate the dose contribution regardless of the time of occurrence. The dose contributions

were then summed to estimate the total dose, yielding 0.21 mSv/yr (21.7 mrem/yr), which is

approximately 4% of the 5 mSv/yr (500 mrem/yr) performance objective. The four largest

contributing radionuclides were Cs-137 (92%), Sn-126 (3.7%), Al-26 (2.9%), and U-234 (0.7%). 

A geometrical reduction factor was applied to account for the probability of placing a house on

the area between the vaults, thereby reducing the exposure to the “average” intruder (Cook et

al., 2005). Without taking into account the geometrical reduction factor, the resident intruder

dose would be 0.36 mSv/yr (36 mrem/yr). 

The agricultural and drilling scenarios were assessed by DOE as part of sensitivity studies in

response to the NRC RAI, in part to address NRC concerns regarding the long-term

performance of the erosion control barrier (WSRC, 2005a). For the drilling scenario, it was

assumed at 1,000 years after closure that a well is drilled through the disposal vault. The

subsurface material exhumed during the drilling included some of the saltstone waste. The

exhumed material was assumed to be mixed with soil in a garden and the intruder is exposed to

the waste through a variety of pathways (e.g., direct radiation from contaminated soil, ingesting

contaminated food from the garden, direct ingestion of contaminated soil, and inhalation of
- 82 -

radionuclides attached to soil particles). A dilution factor of 0.02 was applied to account for

mixture of the exhumed waste in the vegetable garden (Rosenberger et al., 2005). The sum of

fractions derived for this scenario was 0.31 resulting in an estimated post-drilling intruder dose

of 0.31 mSv/yr (31 mrem/yr). For the agricultural scenario, it was assumed that the erosion

barrier did not function as designed, such that it eroded at the same rate as the native soil. 

This allowed the agricultural scenario to occur after a sufficient depth of the closure cover had

eroded away. The intruder was assumed to build a home directly on top of the disposal units,

and the foundation was assumed to extend into the disposal units. Waste exhumed in this

scenario is assumed to be indistinguishable from native soil. Exposure pathways are similar to

the post drilling scenario. A dilution factor of 0.2 was applied to account for mixture of the

exhumed waste with native materials (Rosenberger et al., 2005). Based on an earlier estimate

of projected Vault 4 inventory, the sum of fractions for this scenario was 1.49 (based on 1

mSv/yr [100 mrem/yr]) which is equivalent to 1.49 mSv/yr (149 mrem/yr) and less than the 5

mSv/yr (500 mrem/yr) NRC performance objective for inadvertent intruders (Cook et. al, 2005). 

4.2.18 NRC Evaluation – Protection of Intruders

It is difficult to predict future actions of humans over hundreds to thousands of years. The

intruder scenarios are a regulatory construct designed to ensure protection of the public from

unlikely events while eliminating speculation about future human activities. The intruder

analyses assume that humans will partake in normal land use activities consistent with regional

practices when the active institutional controls are no longer enforced (i.e., disrupt the waste at

100 years with no consideration of the likelihood of occurrence). DOE anticipates they will

maintain active control of the site longer than 100 years. A numerical performance objective is

not provided in 10 CFR Part 61.42, however a dose limit of 5 mSv (500 mrem) per year was

described in the Draft Environmental Impact Statement for 10 CFR Part 61 for development of

waste classification requirements and is applied here for intruder scenarios (NRC, 1981). 

DOE considered reasonable intruder scenarios to evaluate protection of inadvertent intruders

and demonstrate that the performance objectives in '61.42 (Protection of Individuals from

Inadvertent Intrusion) could be achieved. Intruder protection is based on chronic exposures

associated with the resident intruder scenario. In the resident intruder scenario, a home is

excavated above the disposal units and the receptors are exposed to direct radiation that has

been attenuated by the intact saltstone vault roof, soil, and the foundation of the house. 

Ingestion pathways are eliminated from consideration by the presence of at least 3 m of soil

over the disposal vaults by the engineered closure cap (see Section 4.2.5). Intruder doses for

the resident scenario are calculated to be less than the 5 mSv/yr (500 mrem/yr) performance

objective. 

DOE indicated that “the use of the average concentrations of radionuclides in a disposal vault,

rather than the maximum concentrations at any location in a vault, is appropriate when an

inadvertent intruder would access a vault at random locations” (Cook et al., 2002). From a risk

perspective, the statement is correct. However, the information provided in (DOE, 2005) shows

that each waste stream may in fact be different classes of waste (Class A, B, or C). Thus the

risk from each type of vault should be provided, unless the waste streams are going to be

mixed prior to emplacement in the vaults. The reduced likelihood of the scenario occurring is

already accounted for in the application of a 5 mSv/yr (500 mrem/yr) limit to the intruder

scenarios as compared to the application of a 0.25 mSv/yr (25 mrem/yr) limit to the member of
- 83 -

the public, which is the nominal scenario. Use of the average concentration is not appropriately

protective if the volume of more highly-concentrated waste would fill an area that is consistent

with the exposure scenario. If the volume of waste is considerably smaller than the area used

in the exposure scenario, then averaging would be appropriate. DOE applied a reduction factor

of 0.6 to account for the likelihood that the residence may be located in the area between vaults

on the disposal facility. As indicated with respect to waste concentrations, the likelihood of the

scenario occurring is accounted for in application of the higher limit. Because the resident

intruder doses are dominated by Cs-137 and DOE used the estimated inventory for Vault 4, use

of the average concentration for Vault 4 is protective for the resident intruder. The Vault 4

inventory is projected to contain 1.2 x 106

 Ci of the 3 x 106 Ci total Cs-137 to be disposed of in

the SDF. Elimination of the reduction factor would increase the estimated resident intruder

dose from 0.217 mSv/yr (21.7 mrem/yr) to 0.36 mSv/yr (36 mrem/yr), which is well within the 5

mSv/yr (500 mrem/yr) performance objective. This type of reduction factor should not be

applied in future waste determinations. 

The magnitude of the intruder doses are strongly influenced by the amount of shielding

provided by the vaults and closure cap. For example, in an earlier analysis DOE estimated that

a reduction in the amount of shielding from 0.75 m to 0.5 m at year 100 would increase doses

by about a factor of 13 (Cook et al., 2002). It was also estimated that as a result of much larger

Cs-137 inventory than projected in 1992, the intruder doses would exceed the performance

objective by a large margin at 100 years with inadequate shielding (Cook et al, 2002). DOE has

designed a thick, persistent closure cap to ensure that adequate shielding will be provided. 

Because Cs-137 has a relatively short half life (30 years) compared to the analysis time frame

(10,000 years), the risk to the resident intruder decreases rapidly and would meet acceptable

values on the order of a few hundred years after disposal facility closure even with reduced 

shielding (e.g., failure of the engineered cap). There is a higher degree of confidence that an

erosion control barrier can be designed and will perform acceptably for hundreds of years

compared to tens of thousands of years.

DOE assessed the agricultural and drilling scenarios as part of sensitivity studies in response to

the NRC RAI, in part to address NRC concerns regarding the long-term performance of the

erosion control barrier (WSRC, 2005a). The sum of fractions derived for the drilling scenario

was 0.31 resulting in an estimated post-drilling intruder dose of 0.31 mSv/yr (31 mrem/yr). The

dose for the agricultural scenario was 1.49 mSv/yr (149 mrem/yr), which is less than the 5

mSv//yr (500 mrem/yr) NRC performance objective for inadvertent intruders. The results for the

post-drilling and agricultural scenario are based on estimates for projected Vault 4 inventory

that can be considerably less than than the average inventory expected for each of the vaults. 

Cook et al., 2002 used an inventory of 2.65 Ci for Sn-126 and 98.2 Ci for Tc-99. However, the

average inventory of the vaults (assuming 14 vaults) would be approximately 32.1 Ci for Sn-126

and 2357 Ci for Tc-99 (d’Entremont and Drumm, 2005). Therefore, scaling the agricultural

intruder doses to account for estimated average inventory of the vaults would result in

approximately 6.31 mSv/yr (630 mrem/yr) for Sn-126 and 2.21 mSv/yr (220 mrem/yr) for Tc-99. 

Based on current inventory estimates and if the erosion control barrier was to not perform as

designed, these results indicate that the long-term intruder doses would likely approach or

exceed the performance objective for protection of inadvertent intruders. Given the robust

erosion control design, the likelihood that the agricultural intruder scenario is applicable is

significantly reduced, though not completely eliminated given the uncertainty in projecting the

performance of engineered systems over very long periods of time. In the event that the

erosion control design does not function properly, the dose is not excessive. The agricultural
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intruder dose would exceed the value Sn-126 due to the contribution from other radionuclides. 

However, the results for Tc-99 and Sn-126 can not be directly summed because of differences

in the timing of when the peaks would be expected to occur.

DOE’s approach to estimating intruder doses is to sum the peak contribution to dose by

radionuclide regardless of the time of the peak dose. This approach should yield a

conservative estimate of dose because the time of the peak dose from different radionuclides

would be expected to be different. DOE could consider summing the contribution from each

radionuclide at each point in time and reporting the peak of the sums. The approach being

used by DOE may be overly conservative under certain circumstances.

It is expected that the maximum acute dose a hypothetical resident intruder would receive is

0.36 mSv/yr (36 mrem/yr), which demonstrates reasonable assurance that the performance

objective for protection of intruders can be met. Considering the physical properties of the

vaults and saltstone, the NRC agrees that the post-drilling scenario would be unlikely,

especially for the next 1,000 years. If a drilling event were to occur after 1,000 years, the

performance objective could still be met. The erosion control design should eliminate the

occurrence of the agricultural scenario in the near-term, and greatly reduce the likelihood of

occurrence in the long-term.

4.2.19 Protection of Individuals During Operations

The performance objective in 10 CFR 61.43 cross-references “the standards for radiation

protection in Part 20”. DOE’s approach to demonstrating protection of individuals during

operations (10 CFR 61.43) is to cross-walk the relevant DOE regulation or limit with that

provided in 10 CFR Part 20 and demonstrate that the DOE regulation provides an equivalent

level of protection. The cross-referenced “standards for radiation protection” in 10 CFR Part 20

(USNRC, 2005) that are considered in detail 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). 

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.

A number of measures will ensure that exposure of individuals during operations are

maintained ALARA. These include: (1) a documented Radiation Protection Program (RPP), (2)

a Documented Safety Analysis (DSA), (3) design of the SDF and SPF, (4) regulatory and

contractual enforcement mechanisms, and (5) access controls, training, and dosimetry

(Rosenberger et al., 2005). The discussion that follows was provided in Rosenberger et al.,

2005.

A DSA has been approved by DOE for operation of the SPF and SDF in accordance with 10

CFR Part 830 (WSRC, 2004a). As the first step in the development of the DSA, a formal

Consolidated Hazards Analysis (CHA) (WSRC, 2004b) was performed 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 basic phases: hazard identification; hazard classification; and hazard

evaluation. During the hazard identification phase, all possible radiological and chemical
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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 that the hazard classification of the Saltstone Facility was Hazard Category 3,

which is the lowest hazard classification and denotes a potential for only 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 the 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 then analyzed the hazards that were identified

in the CHA that could impact facility workers during normal operations and accident conditions. 

The design requirements for SPF and SDF implemented 10 CFR Part 835 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. Dose rate calculations are being

performed to determine the shielding requirements for the facility. While the upgraded design

is not yet complete, based on the current SPF and SDF design, DOE estimates 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 50 mSv per year (5 rem per year) during both Interim

Salt Processing and SWPF operation (WSRC, 2005b).

The effectiveness of the radiation protection programs at the Savannah River Site has been

demonstrated by past occupational exposure results. Since 1998, the highest dose received by

an SRS worker has been 18.08 mSv (1808 mrem) TEDE compared to a DOE Administrative

Control Limit of 20.0 mSv per year (2000 mrem per year) and the 10 CFR Part 835 limit of 50.0

mSv per year (5000 mrem per year). There has been close to zero total exposure for the SPF

and SDF workforce. The total dose received by workers at the SPF and SDF since 1998 is

0.35 mSv (35 mrem) (Rosenberger et al., 2005).

The air pathway is the predominate pathway for doses to the public from SDF operations. 

Doses from the air pathway to members of the public have been, and are expected to continue

to be, well below the 1.0 mSv (0.1 rem) annual limit. 
- 86 -

4.2.20 NRC Evaluation – Protection of Individuals During Operations

DOE has provided adequate information that individuals will be protected during operations. 

DOE provided a detailed cross-walk of the relevant DOE regulations to those provided in 10

CFR Part 20, which is referenced in the 10 CFR 61.43 performance objective. NRC agrees that

an equivalent level of protection is provided by the relevant DOE regulations or limits to the

requirements found 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). In addition, a number of

measures are applied to ensure that exposure of individuals are maintained as low as

reasonably achievable including: (1) a documented Radiation Protection Program (RPP), (2) a

Documented Safety Analysis (DSA), (3) design of the SDF and SPF, (4) regulatory and

contractual enforcement mechanisms, and (5) access controls, training, and dosimetry. 

In general, the activities at the SPF and SDF involve inert materials and common, lowtemperature, low-energy industrial operations. The public will be located a significant distance

(several kilometers) from the facilities during operations and active security is maintained to

prevent inadvertent access to the site. The NRC agrees with DOE that the risk to the public

during operations should be minimal, and the relevant regulatory limits can be achieved.

The existing DSA for the SPF does not address the organic material in the waste streams

resulting from MCU and SWPF, or the processing of the organic salt waste stream from Tank

48. To prepare for future activities, a hazard analysis and DSA revision process will be

completed to provide the safety basis for the activities. WSRC is in the process of obtaining

information to understand the potential hazards associated with the processing of organic

materials in the SPF and SDF. The information will be considered when the CHA process is

applied to the future waste streams. 

4.2.21 Site Stability

The saltstone facility is located approximately 10 kilometers (6.2 mi) from the nearest site

boundary. The site was selected for low-level waste disposal because of its location on a welldrained topographic high. The probable maximum flood for the nearby surface water body

(Upper Three Runs) is 53 m (175 ft) above mean sea level, which is substantially below the

planned maximum depth of the SDF vaults (WSRC, 2004a). The SDF has been sited in an

area of relatively low seismic activity. The largest known earthquake to affect SRS was the

Charleston earthquake of 1886. The epicenter was approximately 144 km (90 mi) from SRS

and had a magnitude of 6.6 on the Richter Scale. It has been estimated that an earthquake of

this magnitude would result in a peak ground acceleration of 0.10g at SRS, whereas a seismic

evaluation showed that the soils beneath Z-Area are not susceptible to significant liquification

for earthquakes having a peak ground acceleration less than or equal to 0.17g (URS/Blume,

1982), (McHood, 2002).

Two features provide the primary stability for the SDF: (1) the wasteform and engineered vaults

and (2) the engineered cap and erosion control design. The wasteform and vaults are

cementitious materials designed to provide a solid monolith with little void space. There is

expected to be a small amount of shrinkage of the saltstone resulting in a small gap (< 1 cm)

between the saltstone and the walls of the disposal vaults. The use of cement and concrete in

the wasteform and disposal vault will eliminate differential settlement commonly observed in
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low-level waste disposal. Although cracking of the vaults and wasteform may be expected over

long periods of time, significant structural collapse is not anticipated. It is currently envisioned

that a thick, multi-layer engineered cap will be placed over the disposal vaults at closure. An

erosion design with an erosion control barrier will be used to ensure that at least 3 m (10 ft) of

soil will be maintained over the disposal vaults. There are no known biotic disruptive processes 

that would significantly impact the disposal system at the SDF site.

The SDF is an operational low-level waste disposal facility for which additional disposal vaults

will be constructed for disposal of the tank salt waste. Because final closure will not occur until

2019, a final closure plan has not yet been developed. Actions that are contemplated for final

closure have been given in Phifer and Nelson, 2003 and in Cook et al., 2005. The 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. To further ensure long-term stability of

the disposal site, the land in Z-Area will remain under the ownership of the Federal

Government. The three counties making up the SRS have zoning restrictions that prevent the

purchase of property or the approval of building permits at SRS. Residential use of this land

will be prohibited via continued land use leasing restrictions.

4.2.22 NRC Evaluation – Site Stability

DOE’s plans to place the saltstone grout wasteform in concrete vaults eventually covered by a

thick, engineered cap specifically designed for erosion control appear sufficient to indicate that

there is reasonable assurance that the performance objective of '61.44 (Stability of the

Disposal Site After Closure) can be met. The wasteform and vaults will likely contain minimal

void space, therefore differential settlement and the associated negative effects on waste

isolation would be eliminated. The SDF is not located in the flood plain of nearby surface water

bodies and seismic impacts are expected to be limited to cracking of the disposal units along

construction joints.

Erosion is the main disruptive process that may influence the stability of the disposal site. As

indicated in Section 4.2.5.1, DOE has provided sufficient information to conclude that the

erosion protection design is adequate to provide reasonable assurance of long-term stability of

the closure cap for erosion control purposes.

4.3 NRC Review and Conclusions [Criterion Three (A)]

The NRC staff concludes that there is reasonable assurance that Criterion Three (A) will be met

for disposal of salt waste in the saltstone disposal facility. As a result of reviewing the DOE

analysis and supporting technical information, staff concludes that the saltstone disposal facility,

saltstone wasteform, and the engineered closure cap must be appropriately designed and the

designs appropriately implemented in order to satisfy the performance objectives of 10 CFR 61

Subpart C, in particular 10 CFR 61.41. As with any disposal action involving a persistent

hazard, there is significant uncertainty in the projected performance of engineered systems

over extended periods of time. DOE has taken or will be taking actions to mitigate those

uncertainties. 

The NRC staff’s conclusion is based on a number of assumptions about technical aspects of

the disposal facility and wasteform. Many of these assumptions relate to the mechanisms and
- 88 -

rates of degradation of the disposal facility’s engineered features. NRC staff conclusions are

conditional on verification of the assumptions. For the SDF, demonstration that 10 CFR 61.41,

“Protection of the General Population from Releases of Radioactivity” can be met appears to

pose the biggest challenge. 

As discussed in previous sections of this TER, maintaining Tc-99 in the wasteform in a reduced

form to limit its release to the groundwater pathway is the key factor in achieving the

performance objective. The DOE wasteform formulation contains blast furnace slag in an

attempt to immobilize technetium. DOE field lysimeter experiments have shown that the

release of technetium from wasteform samples that contain slag is significantly slower than the

release of technetium from samples that do not contain slag. NRC staff assumed that more

realistic modeling of waste oxidation and release of technetium from an oxidized layer of waste

will result in predicted doses that are significantly lower than those projected in the DOE

sensitivity analysis for 100% oxidation of the waste (see Section 4.2.9). DOE currently

estimates that 3 to 8% of the wasteform will oxidize during the 10,000 year performance period;

therefore the fractional rate of wasteform oxidation is likely to be small enough to allow 10 CFR

61.41 to be satisfied.

Limiting the amount of water flow through the wasteform is important to achieving 10 CFR

61.41. NRC staff have assumed that the wasteform will not have a bulk degraded hydraulic

conductivity greater than 1 x 10-7 cm/s. Bulk degradation of the wasteform can occur from a

variety of mechanisms, as discussed in Section 4.2.9. While the vault contains rebar and other

embedded carbon steel components, the saltstone wasteform is free of these materials and

therefore would not be expected to experience significant changes in bulk hydraulic properties

from reinforcement corrosion. The thick engineered cover will protect the wasteform from

freeze thaw deterioration. Most other degradation mechanisms are initiated by the flux of

species to the vaults, such that the degradation would occur at exposed surfaces. Because of

the low hydraulic and effective diffusivity properties of the initial materials, it is unlikely that

mechanisms that result in a shrinking core type of attack would result in an increase of the bulk

hydraulic conductivity of the wasteform to greater than 1 x 10-7 cm/s in 10,000 years. However,

this conclusion is based on short-term observations and tests of materials that may not be

sufficiently analogous to the saltstone wasteform. Additional studies by DOE to justify the

assumed long-term degradation rates of the wasteform may significantly reduce uncertainty.

In this TER, the NRC staff has identified factors important to assessing compliance with 10

CFR 61, Subpart C, including improvements in future modeling and the associated support to

justify the predictions of the modeling. 

NRC staff used the following assumptions in assessing conformance with Criterion Three (A):

1. More realistic modeling of waste oxidation and release of technetium from an oxidized

layer of waste will result in predicted doses significantly lower than those projected in the

DOE sensitivity analysis for 100% oxidation of the waste (see Section 4.2.9). 

2. The hydraulic conductivity of degraded saltstone and vault concrete will not be larger

than 1 x10-7 cm/s (1 x 10-1 ft/yr) (see Section 4.2.7).
- 89 -

3. Field-scale physical properties (e.g., hydraulic conductivity, effective diffusivity) of asemplaced saltstone are not significantly different from the results of laboratory tests of

smaller-scale samples performed to date (see Section 4.2.7).

4. Cracking from any mechanism will not be significantly more extensive than currently

predicted by DOE (see Sections 4.2.6 and 4.2.9).

5. The overall numerical modeling results for moisture flow through fractures in the

concrete and saltstone located in the vadose zone will be confirmed by future model

support (i.e., fracture flow will not occur) (see Section 4.2.9).

6. The previous DSA bounds the impacts to individuals during operations that may be

estimated from the proposed waste streams (see Section 4.2.20).

7. Active institutional controls will be maintained for 100 years (see Section 4.2.18).

8. Current projections of the radiological concentrations of the waste are greater than or

equal to actual concentrations for highly radioactive radionuclides (see Section 4.2.3).

9. Future tests will confirm that the physical properties of samples that contain organic

materials similar to Tank 48 waste are consistent with non-organic containing samples

(see Section 4.2.7).

10. The erosion control design that DOE eventually implements will not deviate significantly

from the information submitted to the NRC in (WSRC, 2005a) and the associated

references (see Section 4.2.5.1).

11. Development and use of accurate information for the moisture characteristic curves of

concrete and saltstone will not significantly increase currently estimated release rates

(see Section 4.2.9).

12. Gas phase transport of oxygen will not significantly increase oxidation of technetium in

the saltstone (including transport in fractures) (see Section 4.2.9).

NRC staff concludes the following with respect to Criterion Three (A):

1. There is reasonable assurance that the performance objective of 10 CFR 61.41 can be

met, including the provision of ALARA releases of radioactivity to the general

environment assuming that the aforementioned assumptions relevant to the

performance objective are verified.

2. There is reasonable assurance that the performance objective of 10 CFR 61.42 for

protection of individuals from inadvertent intrusion can be satisfied, given that a design

for long-term erosion control is implemented which greatly reduces the likelihood of an

agricultural intruder scenario occurring.

3. There is reasonable assurance that the performance objective of 10 CFR 61.43 for

protection of individuals during operations can be met. During operations, individuals

are protected by DOE regulations which were demonstrated to provide protection
- 90 -

comparable to 10 CFR Part 20. In addition, a number of measures are applied to

ensure that exposure of individuals are maintained as low as reasonably achievable

including: (1) a documented radiation protection program, (2) a Documented Safety

Analysis, (3) design of the SPF and SDF, (4) regulatory and contractual enforcement

mechanisms, and (5) access controls, training, and dosimetry.

4. DOE’s plans to place the saltstone grout wasteform in concrete vaults eventually

covered by a thick, engineered cap specifically designed for erosion control are

sufficient to indicate that there is reasonable assurance that the performance objective

of 10 CFR 61.44 (Stability of the Disposal Site After Closure) can be met. 

4.3.1 Factors Important to Assessing Compliance with 10 CFR 61, Subpart C

In general, verification of the assumptions made in assessing whether the performance

objectives of 10 CFR 61, Subpart C, can be met should be performed by DOE. However, some

of the assumptions made in the analysis, if incorrect, could lead to noncompliance with the

performance objectives. It is these types of assumptions that the NRC plans to monitor as part

of its responsibilities under the NDAA. These assumptions fall into the following general

groups: wasteform and vault degradation, the effectiveness of infiltration and erosion controls,

and estimation of the radiological inventory. The NRC staff concludes that for this draft waste

determination, the following factors are important to assessing whether DOE’s disposal actions

will be compliant with the performance objectives of 10 CFR 61, Subpart C: 

1. The rate of waste oxidation and release of technetium from an oxidized layer of

saltstone will be a key determinant of the future performance of the saltstone disposal

facility and therefore whether 10 CFR 61.41 can be met. More realistic modeling will be

important to achieving the performance objectives, and adequate model support is

essential to providing the technical basis for the model results. It will be important to

ensure that gas phase transport of oxygen through fractures will not significantly

increase oxidation of technetium in the saltstone.

2. The extent of degradation that may influence the hydraulic isolation capabilities of the

saltstone and vaults will be a key factor in assessing whether the SDF can meet 10 CFR

61.41. Degradation mechanisms that may result in the hydraulic conductivity of

degraded saltstone and vault concrete being larger than 1 x 10-7 cm/s (1 x 10-1 ft/yr)

need to be evaluated with multiple sources of information (e.g., modeling, analogs,

experiments [especially field scale and long-term], expert elicitation) to ensure that they

are unlikely to occur. It will be important to ensure that field-scale physical properties

(e.g., hydraulic conductivity, effective diffusivity) of as-emplaced saltstone are not

significantly different from the results of laboratory tests of smaller-scale samples

performed to date. It will be important to perform additional laboratory measurements of

hydraulic conductivity because the data being relied upon represent limited samples that

had a small range of curing times. In addition, because there was a fairly significant

amount of variability in the TCLP test results, if DOE deviates significantly from the

nominal saltstone composition, DOE should perform additional tests for hydraulic

conductivity and effective diffusivity that justify the parameter values used over the

range of compositions.
- 91 -

3. Adequate model support is essential to assessing whether the saltstone disposal facility

can meet 10 CFR 61.41. The model support for: (1) moisture flow through fractures in

the concrete and saltstone located in the vadose zone, (2) realistic modeling of waste

oxidation and release of technetium, (3) the extent and frequency of fractures in

saltstone and vaults that will form over time, (4) the plugging rate of the lower drainage

layer of the engineered cap, and (5) the long-term performance of the engineering cap

as an infiltration barrier is key to confirming performance assessment results.

4. The erosion control design is important to ensuring that 10 CFR 61.42 can be met

because it eliminates pathways and scenarios for intruder dose assessments. 

Implementation of an adequate design that does not deviate significantly from the

information submitted to the NRC in (WSRC, 2005a) and the associated references is

important, or if it does deviate significantly that it is reviewed by NRC staff to ensure that

the revisions are consistent with long-term erosion control design principles.

5. The infiltration control design is important to ensuring that 10 CFR 61.41 can be met

because the release of contaminants to the groundwater is predicted to be sensitive to

the large reduction in infiltration provided by the infiltration control. It is important to

ensure that the design can be implemented and will perform as designed.

6. Implementation of an adequate sampling plan is important to ensuring that 10 CFR

61.41 and 10 CFR 61.42 can be met. It is important to assess results of future sampling

and confirm that current projections of the concentrations of highly radioactive

radionuclides in treated salt waste (or grout) are greater than or equal to actual

concentrations of highly radioactive radionuclides in treated salt waste (or grout).

7. To ensure that Tank 48 waste can be safely managed, future tests of the physical

properties of samples that contain organic materials similar to Tank 48 waste will need

to confirm that the properties of the wasteform made from this waste will provide for

suitable wasteform performance such that the disposal system will be able to meet the

performance objectives. The technical basis should, at a minimum, include tests for

hydraulic conductivity and effective diffusivity.

8. Predicted removal efficiencies of highly radioactive radionuclides by each of the planned

salt waste treatment processes are a key factor in determining the radiological inventory

disposed of in saltstone. The inventory, in turn, is a important factor in the

determination that 10 CFR 61.41 and 10 CFR 61.42 can be met.

This is not an all-inclusive list of factors that may need to be monitored by the NRC to assess

compliance with the performance objectives, but rather is based on DOE’s current planned

approach and NRC’s current analysis of DOE’s approach. Therefore, the factors that need to

be monitored may change as DOE implements its disposal plans. A monitoring plan of DOE’s

disposal actions for the waste assessed in this draft waste determination will be developed, in

coordination with the State, that will present the details of NRC’s planned future monitoring

activities. 
- 92 -

5. CONCLUSIONS

As discussed in detail in previous sections, the NRC staff has conducted a technical analysis of

DOE’s draft waste determination for salt waste disposal at the SRS. NRC staff concludes that

there is reasonable assurance that DOE can meet the criteria provided in the NDAA in Section

3116 (a)(1), (a)(2), and (a)(3)(A)(i). This conclusion is based on information presented in

DOE’s draft Section 3116 waste determination dated March 2005, DOE’s responses to NRC’s

RAI, supporting references, and information provided during meetings between NRC and DOE. 

If, in the future, DOE determines it is necessary to revise its assumptions, analysis, design, or

waste management approach and those changes are important to meeting the criteria of the

NDAA, DOE should consult once again with the NRC regarding the conclusions of this TER. It

should be noted that NRC staff is providing consultation to DOE as required by the NDAA and

the NRC is not providing regulatory approval in this action. DOE is responsible for determining

whether the waste is HLW. This NRC assessment is a site-specific evaluation and is not a

precedent for any future decisions regarding non-HLW or incidental waste determinations at

SRS or at other sites.

DOE should assess the factors that the NRC staff have identified as important to assessing

compliance with 10 CFR 61, Subpart C, and carefully review the assumptions that the NRC

conclusions are based upon. The DOE analysis demonstrates that more realistic modeling of

infiltration, water contact with the waste, waste oxidation, and radionuclide release in an

unsaturated and potentially fractured system is needed, because conservative modeling does

not yield acceptable results. Likewise, adequate model support is essential to confirm the

model results and to ensure that public health and safety can be protected with reasonable

assurance. The deterministic DOE base case result does not have adequate model support

and is not conservative. Considering the uncertainty in many key parameters, it should not be

used as the basis for developing inventory limits. A revised base case should be based on the

projected average vault inventory and orientation of multiple disposal vaults, include the

expected magnitude and timing of climate change from the natural cycling of climates, include

the expected magnitude and rate of oxidation of waste, consider liquid and gas flow in fractures

that may develop, and account for the questionable moisture characteristic curve information

for concrete and saltstone that was used in the previous analysis. 
- 93 -

APPENDIX A. NRC STAFF RECOMMENDATIONS

During its consultative review of the “Draft Section 3116 Determination, Salt Waste Disposal,

Savannah River Site” dated March 2005, the NRC staff made some observations regarding

DOE’s waste management approach. As a result of these observations, the NRC staff has

developed some recommendations for DOE’s consideration. The purpose of the

recommendations is to communicate actions that DOE might consider to further enhance its

approach for management of the salt waste at SRS, as well as the approach for future waste

determinations. As stated in this TER, the NRC staff has concluded that there is reasonable

assurance that DOE will meet the applicable criteria of the NDAA if certain assumptions made

in DOE’s analyses are verified. Thus, it is the NRC staff’s view that implementation of these

recommendations is not necessary to meet the criteria in the NDAA. The recommendations are

based on the information provided in the TER and a more detailed discussion of the underlying

bases for the recommendations can be found in the referenced sections.

1. DOE should continue to base identification of highly radioactive radionuclides in part on

the results of sensitivity or uncertainty analyses (see Section 3.4).

2. If warranted by operating experience, DOE should consider operational practices or

process modifications to minimize sludge carryover and entrainment in salt and

supernate during the DDA process (see Section 3.6). 

3. Because DOE predicts salt treatment will continue at SRS until at least 2019 and

because DOE estimates that only approximately 1% of the Se-79, Tc-99, and I-129 that

will be disposed of in the SDF will result from DDA waste (d’Entremont and Drumm,

2005), DOE may have the opportunity to implement processes to improve removal of

these radionuclides. Processes that remove these radionuclides, which are expected to

dominate doses to members of the public from the SDF, could reduce long-term risk to

the public even if they are implemented several years after the proposed approach has

begun (see Sections 4.2.15 and 4.2.16). Therefore, DOE should consider further

evaluation of the practicality of additional removal of Se-79, Tc-99, and I-129 from SRS

salt waste (e.g., performing additional basic research to develop technologies for the

removal of Se-79, Tc-99, and I-129) to limit the inventory of long-lived radionuclides and

the uncertainties in dose resulting from uncertainty in the long-term performance of the

disposal system (see Section 3.8).

4. DOE should continue to assess the balance between removal of radionuclides that

cause worker risk and are short-lived (e.g., Cs-137) and removal of radionuclides that

cause potential future public risks and are long-lived (e.g., Se-79, Tc-99, I-129) in future

radionuclide removal strategies (see Sections 3.5, 3.8, and 4.2.16). 

5. DOE should consider adopting a probabilistic approach for future performance

assessments (see Sections 4.2 and 4.2.16).

6. In developing its sampling plan, DOE may want to consider how adjusting the inventory

limits could assist in establishing the appropriate number of samples needed to confirm

the actual waste sent to the SPF does not exceed the WAC (see Section 4.2.3).
- 94 -

7. DOE should consider addressing the degradation of the lower drainage layer through

further studies to reduce uncertainty or through additional conservatism in cap design

(see Section 4.2.5).

8. To help reduce uncertainties, DOE should consider performing additional laboratory

measurements of initial hydraulic conductivity, as well as long-term tests or monitoring

studies designed to evaluate the long-term durability of the saltstone and concrete vault

(see Section 4.2.7).

9. Unless mitigated by future design modifications, DOE should explicitly consider rebar

and fill pipe corrosion on the integrity and projected lifetime of the vaults in future

degradation calculations. In the response to action items, DOE indicated that if

penetrations in the vault provide an unacceptable moisture flow pathway, the closure

plan for the facility will be revised to provide modifications to the design (see Section

4.2.7). 

10. In developing model support to justify the assumption of lack of flow through fractures,

DOE should consider: (1) heterogeneity in material properties, (2) temporal variations in

saturation, especially resulting from rapid transport of infiltration through root holes to

the lower drainage layer, (3) infilling of fractures and joints with porous media, (4) offset

of the essentially impermeable vault/saltstone on either sides of the fractures, (5)

variability in the aperture of the joints and fractures, (6) sensitivity to grid size in the

simulations, especially at the interface of the fracture and porous media, and (7)

variability in moisture characteristic curve parameters (see Sections 4.2.9 and 4.3.1). 

11. In future performance assessments, DOE should consider the impact of the use of

steady-state flow fields and the impact of fluid densities on estimated release rates (see

Section 4.2.9).

12. In saturated zone modeling, DOE should consider reducing the grid spacing in the

portion of the model immediately surrounding Vault 4 and the compliance nodes to

verify that peak concentrations are not being arbitrarily diluted by the grid discretization

(see Section 4.2.11).

13. DOE should consider that averaging contaminant releases over the footprint of the vault

may not be realistic for scenarios with discrete releases (e.g., flow in widely-spaced

fractures) (see Section 4.2.11).

14. If the performance assessment is significantly changed in the future, DOE should

develop additional technical basis for the uranium Kd to support the selected value (see

Section 4.2.11). 

15. If subsequent calculations suggest that dilution at the water table may be less than

previously predicted, DOE should consider potential chemical effects on saturated zone

sorption coefficients (see Section 4.2.11). 
- 95 -

CONTRIBUTORS

NRC Staff

Project Manager for Salt Waste Determination Review

Anna H. Bradford

Senior Project Manager

M. Eng., Environmental Engineering, Johns Hopkins University

B.S., Mechanical Engineering, Virginia Tech

Lead Technical Reviewer

David W. Esh

Senior Systems Performance Analyst

Ph.D., Environmental Engineering, The Pennsylvania State University

M.S., Nuclear Engineering, The Pennsylvania State University

B.S., Nuclear Engineering, The Pennsylvania State University

B.S., Physics, Lebanon Valley College

Technical Reviewer

A. Christianne Ridge

Systems Performance Analyst

Ph.D., Environmental Engineering, University of California, Berkeley

M.S., Environmental Engineering, Cornell University

B.A., Physics, Drew University

Center for Nuclear Waste Regulatory Analyses

Lane D. Howard

Senior Research Engineer

M.S., Nuclear Engineering, Texas A&M University

B.S., Civil Engineering, Texas A&M University

Roberto T. Pabalan

Institute Scientist

Ph.D., Geochemistry and Mineralogy, The Pennsylvania State University

B.S., Geology, University of the Philippines

David A. Pickett

Principal Scientist

Ph.D., Geology, California Institute of Technology

M.S., Geology, California Institute of Technology

B.A., Geology, Rice University

Stuart A. Stothoff

Senior Research Scientist

Ph.D., Civil Engineering, Princeton University

S.B., Civil Engineering, Massachusetts Institute of Technology
- 96 -

Consultants

Terry L. Johnson

NRC Consultant

B.S., Civil Engineering, West Virginia University
- 97 -

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- 104 -

LIST OF ABBREVIATIONS AND ACRONYMS

AFP Alpha Finishing Process 

ALARA As Low As Is Reasonably Achievable

Am americium

ARP Actinide Removal Process

ASP Alpha Strike Process

Bq becquerel

Bq/mL becquerel/milliliter

C carbon

CFR Code of Federal Regulations

Ci curie

cm centimeter

Cs cesium

CSSX Caustic Side Solvent Extraction

DDA Deliquification, Dissolution, and Adjustment

DNFSB Defense Nuclear Facilities Safety Board

DOE U.S. Department of Energy

DOE-SRS U.S. Department of Energy, Savannah River Site

DUST-MS Disposal Unit Source Term – Multiple Species

DWPF Defense Waste Processing Facility

EPA U.S. Environmental Protection Agency

HLW high-level radioactive waste

I iodine

IA Interagency Agreement

ICRP International Commission on Radiological Protection

in. inches

ITP In-Tank Precipitation

GCL geosynthetic clay layer

Kd distribution coefficient

Kh hydraulic conductivity

L liters

LLW low-level radioactive waste
- 105 -

NDAA National Defense Authorization Act of Fiscal Year 2005

m meters

m3 cubic meters

MCi million curies

MCU Modular CSSX Unit

mL/g milliliters/gram

MOU Memorandum of Understanding

mrem millirem

MST Monosodium Titanate

mSv millisievert

NAS National Academy of Sciences

Np neptunium

NRC U.S. Nuclear Regulatory Commission

PA performance assessment

pCi/mL picocuries/milliliter

PMF Probable Maximum Flood

PMP Probable Maximum Precipitation

Pu plutonium

RAI request for additional information

SDF Saltstone Disposal Facility

SPF Saltstone Production Facility

Sr strontium

SRM Staff Requirements Memorandum

SRS Savannah River Site

Sv sievert

SWPF Salt Waste Processing Facility

TBq terabecquerel

TPB tetraphenylborate

Tc technetium

TCLP Toxicity Characteristic Leaching Procedure

TEDE Total Effective Dose Equivalent

TER Technical Evaluation Report

TRU transuranic
- 106 -

WAC Waste Acceptance Criteria

WCS Waste Characterization System

WIR waste-incidental-to-reprocessing

yr year