Technical Evaluation Report 

For the Revised Performance Assessment 

for the Saltstone Disposal Facility at the 

Savannah River Site, South Carolina 

Final Report 

U.S. Nuclear Regulatory Commission 

Office Of Federal And State Materials 

 And Environmental Management Programs 

Washington, DC 20555-0001 

APRIL 2012 

 
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i 

Table of Contents 

LIST OF FIGURES .................................................................................................................... III

LIST OF TABLES ....................................................................................................................... V

ABBREVIATIONS/ACRONYMS ............................................................................................... VII

EXECUTIVE SUMMARY ............................................................................................................ X

1. INTRODUCTION ................................................................................................. 1

1.1 Monitoring Background ........................................................................................ 1

1.2 Current Review .................................................................................................... 8

1.3 Site Overview .................................................................................................... 10

2. PROTECTION OF THE GENERAL POPULATION ........................................... 14

2.1. Performance Assessment Overview .................................................................. 14

2.1.1 Description of Deterministic PORFLOWTM Cases ................................................ 14

2.1.2 Description of GoldSim® Probabilistic Analysis .................................................... 18

2.1.3 NRC Evaluation – PA Approach .......................................................................... 18

2.2 Source Term and Inventory ............................................................................... 21

2.2.1 Source Term and Inventory ................................................................................. 21

2.2.2 NRC Evaluation – Source Term and Inventory .................................................... 27

2.3 Scenario Selection and Receptor Group ............................................................ 28

2.3.1 Period of Performance and Institutional Controls ................................................. 28

2.3.2 Scenario Identification ......................................................................................... 28

2.3.3 Identification of Relevant Features and Processes .............................................. 29

2.3.4 Receptor Characteristics ..................................................................................... 30

2.3.5 NRC Evaluation – Scenario Selection and Receptor Groups ............................... 31

2.4 Infiltration and Erosion Control .......................................................................... 31

2.4.1 Local Meteorology and Climatology ..................................................................... 32

2.4.2 Infiltration and Erosion Control ............................................................................ 32

2.4.3 NRC-Evaluation – Infiltration and Erosion Control ............................................... 35

2.5 Disposal Unit Design and Performance ............................................................. 39

2.5.1 Design and Construction of the Disposal Units .................................................... 39

2.5.2 Modeling of Disposal Unit Performance ............................................................... 43

2.5.3 NRC Evaluation – Disposal Unit Design and Performance .................................. 56

2.6 Waste Form ....................................................................................................... 66

2.6.1 Description of Waste Form .................................................................................. 66

2.6.2 Quality Assurance for Waste Form ...................................................................... 69

2.6.3 Modeling of Waste Form Degradation ................................................................. 69

2.6.4 NRC Evaluation – Waste Form ............................................................................ 71

2.7 Source Term Release and Near-Field Transport ............................................... 86

2.7.1 Source Term Release Models ............................................................................. 86

2.7.2 Sorption Coefficients ........................................................................................... 89

2.7.3 Modeling of Flow in the Waste Form and Near Field ........................................... 91

2.7.4 NRC Evaluation – Release and Near-Field Transport.......................................... 95

2.8 Hydrology and Transport ................................................................................. 117

2.8.1 Description of Site Hydrology ............................................................................ 117

2.8.2 Modeling of Hydrology and Far-Field Transport ................................................. 120

2.8.3 NRC Evaluation – Hydrology and Far-Field Transport ....................................... 131

2.9 Air Transport of Radionuclides ........................................................................ 138
ii 

2.9.1 Modeling of Transport of Radionuclides in the Air Pathway ............................... 138

2.9.2 NRC Evaluation – Air Transport ........................................................................ 140

2.10 Biosphere Characteristics and Dose Assessment ........................................... 141

2.10.1 DOE Approach to Modeling Exposure Pathways and Public Dose .................... 141

2.10.2 DOE Biosphere Model Input Parameter Values ................................................. 142

2.10.3 NRC Evaluation – Biosphere Pathway and Dose Calculations .......................... 143

2.11 Computer Models and Computer Codes .......................................................... 147

2.11.1 Computer Models and Computer Codes Used by DOE ..................................... 147

2.11.2 Probabilistic Uncertainty and Sensitivity Analysis .............................................. 152

2.11.3 DOE’s QA Program for Computer Codes .......................................................... 154

2.11.4 NRC Evaluation – Computer Models and Computer Codes .............................. 154

2.12 ALARA Analysis .............................................................................................. 164

2.12.1 ALARA Analysis ................................................................................................ 164

2.12.2 NRC Evaluation – ALARA Analysis ................................................................... 165

2.13 Protection of the General Population from Releases of Radioactivity .............. 167

2.13.1 DOE Dose Calculations ..................................................................................... 167

2.13.2 DOE Uncertainty and Sensitivity Analysis for Protection of the General 

Population ......................................................................................................... 170

2.13.3 NRC Evaluation – Evaluation of Performance Assessment Results for 

Protection of the General Public ........................................................................ 174

3. PROTECTION OF INADVERTENT INTRUDERS............................................ 200

3.1 Assessment of Inadvertent Intrusion ................................................................ 200

3.2 NRC Evaluation – Protection of Intruders ........................................................ 204

4. PROTECTION OF INDIVIDUALS DURING OPERATIONS ............................. 209

4.1 Protection of Individuals during Operations ..................................................... 209

4.2 NRC Evaluation – Protection of Individuals during Operations ........................ 210

5. SITE STABILITY ............................................................................................. 212

5.1 Site Stability..................................................................................................... 212

5.2 NRC Evaluation – Site Stability ....................................................................... 215

6. CONCLUSIONS .............................................................................................. 218

6.1 Compliance with the Performance Objectives of 10 CFR 61, Subpart C .......... 219

6.1.1 Protection of the General Population from Releases of Radioactivity ................ 219

6.1.2 Protection of Individuals from Inadvertent Intrusion ........................................... 220

6.1.3 Protection of Individuals during Operations ....................................................... 221

6.1.4 Stability of the Disposal Site after Closure ......................................................... 222

6.2 Key Monitoring Factors .................................................................................... 223

6.3 Path Forward ................................................................................................... 225

REFERENCES ....................................................................................................................... 227

CONTRIBUTORS ................................................................................................................... 245

APPENDIX A MONITORING FACTORS ............................................................................... 247

APPENDIX B RESULTS OF NRC STAFF SENSITIVITY ANALYSES PERFORMED 

WITH A SIMPLE ANALYTICAL DUAL-KD MODEL .......................................... 264 
iii 

List of Figures 

Figure 1: NRC Interpretation of DOE Base Case Conceptual Model..........................................xv

Figure 1.3-1: Location of the Z-Area ......................................................................................... 11

Figure 1.3-2: Vault 4 at the SDF ............................................................................................... 12

Figure 1.3-3: Vault 1 at the SDF ............................................................................................... 12

Figure 1.3-4: Disposal Cells 2A and 2B .................................................................................... 13

Figure 1.3-5: Areal Picture of SDF ............................................................................................ 13

Figure 2.3-1: Locations of the Compliance Points and the Sectors Analyzed ............................ 29

Figure 2.4-1: Conceptual illustration of the SDF Closure Cap ................................................... 33

Figure 2.4-2: Conceptual illustrations of the components of the SDF closure caps and 

additional engineered layers above the disposal units .............................................................. 33

Figure 2.5-1: Fraction of Disposal Unit Concrete that Remains Reduced ................................. 48

Figure 2.5-2: DOE Case A modeled Hydraulic Conductivity and Diffusion Coefficient for 

HDPE Lining of FDCs After Closure ......................................................................................... 50

Figure 2.5-3: DOE Case A modeled Hydraulic Conductivity and Diffusion Coefficient for 

HDPE/GCL Composite Layer on the Roof and Below the Floors of the FDCs After 

Closure ..................................................................................................................................... 50

Figure 2.5-4: Respective Material Zones for PORFLOWTM modeling ........................................ 54

Figure 2.5-5: FDC wall-floor interface including proposed design enhancements (as of 

November, 2010) ...................................................................................................................... 55

Figure 2.5-6: FDC lower corner detail for PORFLOWTM modeling ............................................ 55

Figure 2.6-1: Eh–pH Diagram Showing Tc Speciation ............................................................... 83

Figure 2.7-1: Vertical Darcy velocity through saltstone in Vault 4 predicted with the DOE 

PORFLOWTM model. .............................................................................................................. 112

Figure 2.7-2: Vertical Darcy velocity through saltstone in an FDC predicted with the DOE 

PORFLOWTM model ............................................................................................................... 113

Figure 2.7-3: Moisture Characteristic Curves Adapted from the PA and 

WSRC-STI-2006-00198 .......................................................................................................... 114

Figure 2.7-4: Characteristic curves for the intact and fractured saltstone and concrete as 

adapted from the 2009 PA and SRNL-STI-2009-00115 .......................................................... 115

Figure 2.8-1: Location of Z-Area within the GSA and Surface Water Locations ...................... 119

Figure 2.8-2: Conceptual Model for Flow on the GSA near SDF ............................................. 120

Figure 2.8-3: Perspective View of Regional GSA/PORFLOWTM Flow and Transport 

Model Domain ........................................................................................................................ 122
iv 

Figure 2.8-4: SDF/PORFLOWTM Saturated Zone Model with 100-m (330-ft) Compliance 

Evaluation Sectors .................................................................................................................. 123

Figure 2.8-5: GSA/PORFLOWTM Model Boundary Conditions ................................................ 124

Figure 2.8-6: Calibrated Horizontal Hydraulic Conductivity Assignments to the Regional 

GSA/PORFLOWTM Model ....................................................................................................... 128

Figure 2.8-7: Calibrated Vertical Hydraulic Conductivity Assignments to the Regional 

GSA/PORFLOWTM Model ....................................................................................................... 128

Figure 2.8-8: Locations of Sources at the Saltstone Disposal Facility ..................................... 137

Figure 2.11-1: Model Linkages in DOE’s Performance Assessment for the SDF .................... 149

Figure 2.13-1: Doses from DOE Case K with variations in assumed Kds for Tc in oxidized 

and reducing saltstone............................................................................................................ 169

Figure 2.13-2: Vault 4 vertical Darcy velocity through the saltstone matrix (excludes 

fractures) predicted with the DOE PORFLOWTM model for Case A, the Synergistic Case, 

and the Increased Hydraulic Conductivity Case ...................................................................... 179

Figure 2.13-3: FDC vertical Darcy velocity through the saltstone matrix (excludes 

fractures) predicted with the DOE PORFLOWTM model for Case A, the Synergistic Case, 

and the Increased Hydraulic Conductivity Case ...................................................................... 180

Figure 2.13-4: Water shedding around Vault 4 in DOE Case A at 0 and 8,000 years after 

site closure (left top and bottom) as compared to an NRC interpretation of know 

conditions (top right) or potential future conditions considering uncertainty (bottom right) ....... 182

Figure 2.13-5: Modeled volumetric flow per unit thickness through Vault 4 from 5,500 to 

6,000 years after SFD closure predicted by the DOE PORFLOWTM model for Case A, 

Cases K/K1/K2, and select DOE sensitivity analyses ............................................................. 183

Figure 2.13-6: Modeled volumetric flow per unit thickness through Vault 4 from 5,500 to 

6,000 years after SFD closure predicted by the DOE PORFLOWTM model for the 

synergistic case. ..................................................................................................................... 184

Figure 2.13-7: Intermediate results from DOE PORFLOWTM model for Case K, showing 

Tc release from Saltstone and re-concentration in the FDC floor and walls ............................ 185

Figure 2.13-8: Comparison of annual fractional release from saltstone predicted by 

NRC’s simple analytical average-Kd and dual-Kd models, using DOE Case K1 fracturing, 

near-field flow, and Kd values ................................................................................................. 193

Figure 2.13-9: Comparison of annual fractional release from saltstone and from the nearfield domain (i.e., saltstone, disposal units, and unsaturated soil) from DOE Case K1 

model output files with the NRC predicted saltstone oxidation for Case K1 fracturing. ............ 193
v 

List of Tables 

Table 1: Key Review Results from the NRC's Technical Evaluation of the 2009 PA .................. xii

Table 1.1-1: Conditions of NRC's 2005 Findings Regarding the NDAA Section 3116(a) 

Criteria ........................................................................................................................................ 3

Table 1.1-2: Key Factors Resulting from NRC Review of 2005 Saltstone PA ............................. 4

Table 1.1-3: History of Onsite Observation Activities .................................................................. 7

Table 1.1-4: Saltstone Disposal Facility Open Issues ................................................................. 8

Table 1.2-1: NRC/DOE Meetings during PA Review ................................................................... 9

Table 2.1-1: Description of Deterministic Cases Modeled by DOE ........................................... 15

Table 2.2-1: Projected Inventory at Time of Closure ................................................................. 22

Table 2.2-2: Revised Inventory of Ra-226 and its ancestors in Cases K, K1, and K2 ............... 25

Table 2.2-3: Estimates of Ra-226 and Th-230 inventory in Vault 4 ........................................... 25

Table 2.5-1: Vault 1 Concrete Formulations .............................................................................. 39

Table 2.5-2: Vault 4 Concrete Formulations .............................................................................. 40

Table 2.5-3: Potential FDC Concrete Formulation .................................................................... 42

Table 2.5-4: Initial Disposal Unit Concrete Properties Assumed by DOE PA Cases and 

Cases K, K1, K2 ....................................................................................................................... 44

Table 2.5-5: Chemical Transition Times for Cementitious Materials in Disposal Units .............. 47

Table 2.5-6: Initial Fracture Spacing in Cases K, K1, and K2 .................................................... 47

Table 2.5-7: Cumulative Area of Holes in HDPE Encasing Disposal Cells ................................ 51

Table 2.6-1: Formula of Saltstone and Clean Grout .................................................................. 67

Table 2.6-2: Saltstone Hydraulic and Physical Properties Used in Case A ............................... 67

Table 2.6-3: Saturated Hydraulic Conductivity and Effective Diffusivity Used in Selected 

Cases ....................................................................................................................................... 70

Table 2.6-4: Modeled and Measured Porosity (η), Bulk Density (ρb), and Particle Density 

(ρp) Values for Saltstone ........................................................................................................... 78

Table 2.6-5: Modeled and Measured Initial Hydraulic Conductivity Values for Saltstone........... 80

Table 2.7-1: Initial (or for Some Materials Time Invariant) Material Properties Used in 

PORFLOWTM Vadose Zone Modeling ...................................................................................... 94

Table 2.7-2: Saltstone Kd Value for Tc under reducing conditions ........................................... 104

Table 2.7-3: Saltstone Kd Value for Tc under oxidizing conditions .......................................... 105

Table 2.7-4: Kd values (mL/g) assumed by DOE for I for cementitious materials .................... 106

Table 2.7-5: Kd values (mL/g) assumed by DOE for Ra for cementitious materials ................. 107
vi 

Table 2.7-6: Kd values (mL/g) assumed by DOE for Se for cementitious materials ................. 108

Table 2.7-7: Kd values (mL/g) assumed by DOE for Sr for cementitious materials .................. 109

Table 2.8-1: Saturated Zone Material Properties DOE used in PORFLOWTM Modeling .......... 126

Table 2.8-2: Comparison of Estimated and GSA/PORFLOW-Modeled Baseflows .................. 129

Table 2.8-3: Dispersivities Used in DOE’s SDF/PORFLOWTM models .................................... 133

Table 2.8-4: Ratio of All Sources to Individual Source Peak Doses ........................................ 136

Table 2.11-1: Aquifer Exposure Probabilities and Relative Concentrations ............................. 160

Table 2.13-1: Sector B Doses from DOE’s Deterministic Analyses ......................................... 171

Table 2.13-2: Sector I Dose Results from DOE’s Deterministic Analyses ............................... 172

Table 2.13-3: Estimated Peak Vault 4 Dose from Ra-226 that Ingrows from Th-230 .............. 188

Table 2.13-4: Variations of DOE Case K1 PORFLOWTM model. ............................................. 190

Table 3.1-1: Description of Intruder Cases Evaluated by DOE................................................ 202

Table 3.1-2: Results of Intruder Assessment Performed by DOE ........................................... 203

Table A-1: Key Monitoring Factors ......................................................................................... 248

Table A-2: PA Maintenance Items .......................................................................................... 257 

Table B-1: Peak annual fractional release rates estimated with a simple analytical dual-Kd 

model for cases in which flow through saltstone is limited by the saltstone hydraulic 

conductivity...........................................................................................................267 

Table B-2: Peak annual fractional release rates estimated with a simple analytical dual-Kd 

model for cases in which flow is based on DOE PORFLOWTM Case K1 flow through 

saltstone...............................................................................................................268 

Table B-3: Peak annual fractional release rates predicted with DOE’s Case K, K1, and K2 

PORFLOWTM models and an NRC-modified version of Case 

K1.......................................................................................................................268 

 
vii 

ABBREVIATIONS/ACRONYMS 

ACI American Concrete Institute 

ALARA As Low As Is Reasonably Achievable 

Am americium 

ARP Actinide Removal Process 

ASP Alpha Strike Process 

ASTM American Society for Testing and Materials 

Ba barium 

BGS Below Ground Surface 

C carbon 

CFR Code of Federal Regulations 

Cl chlorine 

CNWRA Center for Nuclear Waste Regulatory Analyses 

Cr chromium 

Cs cesium 

CSH Calcium silicate hydrate 

CSSX Caustic Side Solvent Extraction 

DCF Dose Conversion Factor 

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 

DRF Dose release factor 

DUST-MS Disposal Unit Source Term B Multiple Species 

DWPF Defense Waste Processing Facility 

Eh Reduction Potential 

EPA U.S. Environmental Protection Agency 

EXAFS Extended X-ray absorption fine structure 

FACT Flow and Contaminant Transport 

FDC Future Disposal Cell 

FTF F-Tank Farm 

GCL Geosynthetic Clay Layer 

GCU Gordon Confining Unit 

GSA General Separations Area 

HDPE High Density Polyethylene 

HELP Hydrologic Evaluation of Landfill Performance 

HLW high-level radioactive waste 

HTF H-Tank Farm 

I iodine 

IA Interagency Agreement 

IAEA International Atomic Energy Agency 

ICRP International Commission on Radiological Protection 

ITP In-Tank Precipitation 

Kd distribution coefficient 
viii 

Kh hydraulic conductivity 

LFRG Low-Level Waste Disposal Facility Federal Review Group 

LLW low-level radioactive waste 

MCU Modular Caustic Side Solvent Extraction Unit (Modular CSSX Unit) 

MEI Maximally exposed individual 

MOP Member of the Public 

MOU Memorandum of Understanding 

MST Monosodium Titanate 

NAS National Academy of Sciences 

NCRP National Council on Radiation Protection and Measurements 

NCRP National Council on Radiological Protection 

NDAA Sect. 3116, National Defense Authorization Act for Fiscal Year 2005 

Np neptunium 

NRC U.S. Nuclear Regulatory Commission 

PA Performance Assessment 

PGA Peak Ground Acceleration 

PMF Probable Maximum Flood 

PMP Probable Maximum Precipitation 

Pu plutonium 

QA Quality Assurance 

QAMP Quality Assurance Management Plan 

QC Quality Control 

Ra radium 

RAI Request for Additional Information 

Rh rhodium 

Sb antimony 

SCDHEC South Carolina Department of Health and Environmental Control 

SCHWMR South Carolina Hazardous Waste Management Regulations 

SDF Saltstone Disposal Facility 

Se selenium 

SPF Saltstone Production Facility 

SQAP Software quality assurance plan 

Sr strontium 

SRM Staff Requirements Memorandum 

SRS Savannah River Site 

SWPF Salt Waste Processing Facility 

Tc technetium 

TCCZ Tan Clay Confining Zone 

TCLP Toxicity Characteristic Leaching Procedure 

TEDE Total Effective Dose Equivalent 

TER Technical Evaluation Report 

TPB tetraphenylborate 

TRU transuranic 

U uranium 
ix 

UTR Upper Three Runs 

UTR-LZ Upper Three Runs Lower Zone 

UTR-UZ Upper Three Runs Upper Zone 

UV Ultraviolet 

WAC Waste Acceptance Criteria 

WCS Waste Characterization System 

WD Waste Determination 

WIR waste-incidental-to-reprocessing 

XANES X-Ray absorption near edge structure 

Yr Year 
x 

EXECUTIVE SUMMARY 

On November 23, 2009, the U.S. Department of Energy (DOE) submitted the “Performance 

Assessment for the Saltstone Disposal Facility at the Savannah River Site” (2009 PA) to the 

U.S. Nuclear Regulatory Commission (NRC) for review, as required by Section 3116(b) of the 

Ronald W. Reagan National Defense Authorization Act for Fiscal Year 2005 (NDAA or the Act). 

Section 3116(b) of the NDAA requires NRC to monitor DOE’s disposal actions concerning 

certain wastes associated with spent fuel reprocessing that DOE, in consultation with NRC, has 

determined to be non-High Level Waste (HLW). Although radioactive material resulting from the 

reprocessing of spent nuclear fuel typically is defined as HLW, DOE may determine certain 

reprocessing waste is non-HLW, or Waste Incidental Reprocessing (WIR), if it does not need to 

be disposed of as HLW to manage the risks it poses. The 2009 PA is an update to DOE’s 

February 28, 2005 Performance Assessment (PA) performed in support of the “Section 3116 

Determination, Salt Waste Disposal, Savannah River Site” (DOE-WD-2005-001). The updated 

PA includes new information about issues addressed in the 2005 PA, and information about 

changes to the disposal cell design as well as an improved format for readability and technical 

clarity. Per the definition of NRC monitoring in the NDAA, this review focuses on DOE’s 

compliance with the third criterion of the Act, which is that the disposal actions must comply with 

the performance objectives of NRC’s LLW disposal facility regulations as presented in the Title 

10 of the Code of Federal Regulations (CFR) Part 61, Subpart C. The purpose of DOE’s PA is 

to demonstrate that its waste disposal strategy for the SDF remains in compliance with these 

performance objectives. 

Approximately 136 million liters (36 million gallons) of liquid waste resulting from reprocessing 

spent nuclear fuel is stored in 49 underground carbon steel tanks at SRS. This waste is 

separated into two streams based on activity. The high activity fraction is HLW and is made into 

a glass waste form. The low activity fraction, called salt waste, is treated to reduce the 

concentrations of certain key radionuclides and then mixed with dry materials (i.e., cement, blast 

furnace slag, and fly ash) to form a grout waste form called saltstone. Saltstone is disposed of 

in underground disposal units in the Saltstone Disposal Facility (SDF). The performance of the 

SDF is the subject of DOE’s revised PA and of this review. At the time of this review 24 million 

liters (6.3 million gallons) of salt waste has been disposed in Vault 4, equaling 42 million liters 

(11 million gallons) of saltstone (SRR-ESH-2011-00014). 

NRC has conducted a review and confirmatory analysis of the 2009 PA and has documented 

this review and analysis in this Technical Evaluation Report (TER). NRC’s review results are 

being provided to DOE in accordance with its monitoring role under the NDAA and are not 

intended to represent any regulatory authority related to DOE’s disposal activities. 

The NRC staff concludes it has reasonable assurance that waste disposal at the SDF meets the 

10 CFR 61 performance objectives for protection of individuals against intrusion (§61.42), 

protection of individuals during operations (§61.43), and site stability (§61.44). However, based 

on its evaluation of DOE’s results and independent sensitivity analyses conducted with DOE’s 

models, the NRC staff no longer has reasonable assurance that DOE’s disposal activities at the 

SDF meet the performance objective for protection of the general population from releases of 
xi 

radioactivity (§61.41). Although the NRC staff cannot conclude that the performance objective 

in §61.41 is met, based on DOE’s results and NRC’s own independent analyses, the potential 

dose to an off-site member of the public from DOE’s disposal actions is still expected to be 

relatively low (i.e., approximately 1 mSv/yr [100 mrem/yr], the public dose limit in §20.1301)1

. 

DOE concludes that any dose greater than 0.25 mSv/yr (25 mrem/yr) would occur more than 

10,000 years after site closure; however, the staff disagrees with many of the assumptions in 

DOE’s model. The staff expects that any exceedance of the §61.41 limit would occur many 

years after site closure but finds that DOE has not provided a sufficient basis for DOE’s 

conclusion that any exceedances would occur beyond 10,000 years. In accordance with 

NUREG-1854, NRC Staff Guidance for Activities Related to U.S. Department of Energy Waste 

Determinations, the time for which the 0.25 mSv/yr (25 mrem/yr) dose limit in §61.41 must be 

met is generally2

 10,000 years. Therefore, it seems likely that DOE could provide additional 

information or take mitigative actions in the short term that would provide reasonable assurance 

that salt waste disposal at the SDF meets the 10 CFR 61 Subpart C performance objectives. 

NRC has identified these items as monitoring factors in this TER. 

In December 2005, the NRC staff documented a similar review of DOE’s 2005 PA for the SDF. 

At that time, the NRC staff concluded that it had reasonable assurance that salt waste disposal 

at the SDF would meet the performance objectives of 10 CFR 61 provided certain assumptions 

in DOE’s analyses were verified during monitoring. During the current review, the NRC staff 

carefully evaluated information related to these assumptions (i.e., information regarding 

saltstone oxidation, saltstone and disposal unit hydraulic conductivity, field-scale properties of 

as-emplaced saltstone, saltstone fracturing, numerical modeling of flow through fractures, 

radionuclide concentrations, moisture characteristic curves, and erosion control), as well as new 

factors of importance to the modified disposal plans and revised conceptual model. The 

outcome of the NRC staff’s 2005 review was a list of 8 Key Monitoring Factors with which NRC 

tracks DOE’s SDF disposal actions and development of supporting information. In the current 

review, staff has taken a similar approach to identify Key Monitoring Factors, in that staff has 

focused on risk-significant issues. However, based on staff’s monitoring experience since the 

previous review, staff determined that making individual Key Monitoring Factors more specific 

(i.e., smaller in scope), though leading to a slightly larger number of individual Key Monitoring 

Factors, would facilitate monitoring. Following completion of this TER, the staff will revise its 

 

1

 §61.41 states “Concentrations of radioactive material which may be released to the general environment in ground 

water, surface water, air, soil, plants, or animals must not result in an annual dose exceeding an equivalent of 25 

millirems to the whole body, 75 millirems to the thyroid, and 25 millirems to any other organ of any member of the 

public.” NRC has evaluated compliance using a dose limit of 0.25 mSv/yr (25 mrem/yr) Total Effective Dose 

Equivalent (TEDE) consistent with the approach discussed in final rule (66 FR 55752) “Because each of the organs 

had the same limit under the older system even though each had a different level of radiosensitivity, it is very difficult 

to directly compare the old standards with the new standards. As noted in the proposed rule, the Commission 

considers 0.25 mSv/yr (25 mrem/yr) TEDE as the appropriate dose limit to compare with the range of potential doses 

represented by the older limits that had whole body dose limits of 0.25 mSv/yr (25 mrem/yr).” The DOE performance 

assessment and NRC’s review have used the most updated dose factors consistent with Commission direction in 

SRM-SECY-01-0148 to calculate the potential dose. 

2

 NUREG-1854 also indicates that assessments beyond 10,000 years may be necessary in some cases to 

demonstrate that the disposal of certain types of waste does not result in high impacts to future generations. 
xii 

monitoring plan for the SDF. Appendix A of this TER lists the new Key Monitoring Factors and 

provides details about each. 

Table 1 below presents the NRC staff’s key review results related to the 10 CFR 61 

performance objectives. 

Table 1: Key Review Results from the NRC's Technical Evaluation of the 2009 PA 

Performance 

Objective Key Results

§61.41 

Protection of 

the General 

Population 

from Releases 

of 

Radioactivity

• Based on the results of the NRC’s review, the NRC staff does not have 

reasonable assurance that DOE’s disposal actions at the SDF meet this 

performance objective. 

• The NRC staff does not find Case A to be an appropriate compliance case 

because it does not accurately reflect current site conditions, does not account for 

the full ranges of measured values of key parameters or expected differences 

between laboratory and as-emplaced values, and does not appropriately account 

for potential changes in parameter values with time. 

• In response to NRC concerns about DOE’s Case A, DOE supplied Cases K, K1, 

and K2. Contrary to DOE’s characterization of these new cases as overlypessimistic sensitivity analyses, the NRC staff believes these cases are based on 

a combination of both overly-optimistic and apparently conservative assumptions. 

NRC staff has relied heavily on Case K1 in its review and conclusions because it 

resolves many of the concerns NRC staff has about using Case A as a 

compliance case. 

• The NRC staff finds that the timing of DOE’s predicted peak dose to an off-site 

general member of the public in Case K1 (approximately 0.9 mSv/yr [90 mrem/yr] 

at 12,900 years after site closure) is sensitive to, and delayed by assumptions 

about, saltstone fracture growth, use of an average Kd value to simulate release of 

Tc-99, and seemingly overly-optimistic assumptions about Tc retention in disposal 

unit concrete. 

• Because of the large uncertainty in the predicted timing of the 0.9 mSv/yr 

(90 mrem/yr) peak dose from Tc-99 and the expectation that the predicted peak in 

DOE’s Case K1 model was delayed by unsupported assumptions, the NRC staff 

could not conclude it had reasonable assurance that the dose would meet the 

0.25 mSv/yr (25 mrem/yr) limit for 10,000 years after site closure based on DOE’s 

Case K1 results. 

• The NRC staff performed additional analyses with DOE’s PA model of sourceterm release to determine whether it could reduce the uncertainty in the timing or 

magnitude of the Case K1 peak dose. These additional analyses, however, still 

led to predicted Tc-99 peak doses that exceeded the performance objective dose 

limit of 0.25 mSv/yr (25 mrem/yr). 

• An additional attempt by NRC staff to develop a simplified release model to avoid 

an artifact of DOE’s Case K1 Tc release model led to peak fractional release rates 

into the environment similar to the DOE’s Case K1 fractional release rate. 
xiii 

Performance 

Objective Key Results

§61.42 

Protection of 

Individuals 

from 

Inadvertent 

Intrusion 

• Based on the results of the NRC’s review, the NRC staff has reasonable 

assurance that DOE’s disposal actions at the SDF meet this performance 

objective. 

• The NRC staff finds that the scenarios and pathways analyzed by DOE for the 

assessment of this performance objective are appropriate based on the regional 

practices near SRS. 

• The NRC staff finds DOE’s approach in its sensitivity analysis to evaluate the 

dose to an individual who drills into a disposal unit and spreads drill cuttings on 

the land surface to be reasonable. 

• The NRC staff concludes that the Case K1 chronic intruder dose predicted by 

DOE is likely to be conservative (i.e., overestimate the potential intruder dose). 

The NRC staff therefore concludes that the dose to an inadvertent intruder 

through the groundwater pathways is likely to be less than 5 mSv/yr 

(500 mrem/yr), provided that key assumptions in Case K1 analysis are true. 

• The results of the reviewed sensitivity cases indicate that the additional dose an 

intruder would receive from being exposed to drill cuttings containing saltstone 

would not be significant compared to the dose to an intruder from using 

groundwater on site, which the NRC staff expects to meet the 5 mSv/yr 

(500 mrem/yr) dose limit used in promulgation of §61.42. 

• The NRC staff expects the dose to an intruder who drills directly into saltstone 

from the drill cuttings to meet the performance objective for protection of an 

inadvertent intruder. 

§61.43 

Protection of 

Individuals 

during 

Operations 

• Based on the results of the NRC’s review in its 2005 TER, onsite 

observations conducted in October 2007 and March 2008, and NRC’s review 

of the annual SRS Environmental Reports, the NRC staff has reasonable 

assurance that DOE’s disposal actions at the SDF meet this performance 

objective. 
xiv 

Performance 

Objective Key Results

§61.44 

Stability of the 

Disposal Site 

after Closure 

• Based on the uncertainty in certain risks associated with this performance 

objective, aspects of site stability will be included in NRC's revised 

monitoring plan. However, based on the results of the NRC’s review, the 

NRC staff has reasonable assurance that DOE’s disposal actions at the SDF 

meet this performance objective. 

• The NRC staff concludes the saltstone waste form will provide a monolithic 

structure, minimize void space, and prevent collapse and differential settlement 

that could occur due to consolidation of the waste. 

• NRC staff evaluated the dynamic settlement that would result from an earthquake 

with a 10,000-year return period and found it was unlikely to cause significant 

disruption to the SDF. 

• NRC staff determined that floods are unlikely to disrupt the SDF because the 

10,000-year flood level for the Upper Three Runs basin near the SDF is 

significantly below the lowest planned elevation of a disposal unit at the SDF. 

• Much of SRS, including the SDF, is underlain by calcareous sediment in the 

Santee formation resulting in the presence of “soft zones.” Historically, some of 

these zones have consolidated, resulting in depressions on the land surface. 

DOE concluded that consolidation of a soft zone would have minimal effects on 

the stability of a disposal unit at the SDF. 

• Sinks identified elsewhere at SRS are comparable in size to a four-pack of Future 

Disposal Cells (FDCs). If a sink developed under a four-pack, local infiltration 

could increase and disposal units could fracture. However, because the history of 

sink development at the SRS is unclear and the potential development of soft 

zones under the SDF also is uncertain, the probability a sink would develop at the 

SDF within 10,000 years of closure is uncertain. 

• Recent studies predict greater static settlement in the SDF than addressed by 

DOE in the PA. 

• In general, the NRC staff found more uncertainty in the potential effects of static 

settlement due to loading of the subsurface layers and settlement due to 

calcareous zones present or potentially developing under the SDF than that 

resulting from the potential effects of earthquakes, floods, and erosion at the SDF. 

The following section provides a high-level overview of the staff’s findings as described in the 

TER. Although the section offers helpful details of the staff’s conclusions, for a broader and 

more detailed discussion of the 2009 PA and NRC staff’s evaluation, please see the appropriate 

sections of the TER. 

High-level Overview of NRC Conclusions on the 2009 PA

In 2005, the NRC reviewed the analyses the DOE performed to support its Waste Determination 

for the SDF. The main NRC findings for long-term performance in 2005 were that to contain the 

major risk driver, Tc, over the long-term, (1) the as-placed waste form would need to chemically 

reduce the Tc sufficiently to make it immobile, and (2) the waste form and other engineered 

features would need to limit water flow into the waste form. Water flow into the facility could 

bring oxygen into the waste form and oxidize the Tc, allowing release. In addition, the water 

would provide transport out of the waste form for any mobilized radionuclides, including Tc. 

After the waste determination was issued in 2006, DOE gathered additional information 
xv 

regarding their assumptions for long-term performance while producing saltstone operationally. 

After short-term performance issues with the existing vault design, DOE redesigned the vaults 

and referred to the new disposal units as “future disposal cells.” DOE developed revised 

analyses for the existing vaults and future disposal cells to address the new designs and the 

results of new experiments, as well as performance issues (such as cracking in the walls of 

Vault 4). DOE provided this revised PA to NRC in November 2009. 

Figure 1: NRC Interpretation of DOE Base Case Conceptual Model 

DOE’s revised PA evaluates a base case (alternately, called Case A), which DOE indicates 

reflects its expectation of long-term performance. In this case, the upper cover is assumed to 

degrade with time, but there is no degradation of the saltstone and limited degradation of the 

disposal units assumed over a 20,000 year evaluation period (Figure 1). The walls in Vaults 1 

and 4 are assumed to be initially cracked, while the new future disposal cells walls, like the floor 

and roof experience only minimal degradation. Due to a combination of the material properties 

assumed and the lack of roof degradation, Case A continues to shed over 99% of the water 

around the disposal units from 400 to 10,000 years after site closure. This significant limitation 

on modeled water contact with the waste limits the modeled transport of radionuclides out of the 

waste. In addition, because oxidation from gas-phase transport of oxygen is not included in 

DOE’s base case, the flow restriction significantly limits the modeled oxidation of saltstone, 

further limiting Tc mobilization and release. 

DOE also supplied a number of alternate assessments, which evaluated the impacts of 

increased degradation of one or more barriers (as compared to the base case). For example, in 

Cases B and C, DOE evaluates the impact of fast-flow paths, such as gaps between the 

saltstone and disposal unit walls. However, in almost all of the alternate assessments provided 

with the PA, the roof remained largely intact so the shedding of over 99% of the water occurred, 

which made the cases less useful for evaluating importance of the barriers. In one case, the 

Synergistic Case, DOE evaluated multiple degradation mechanisms, including roof degradation. 
xvi 

Although more water flows to the disposal unit in the synergistic case (as compared to the base 

case), because the saltstone matrix is assumed to have a very low hydraulic conductivity, most 

of the flow is diverted through the disposal unit walls (for the FDCs) or saltstone fractures (for 

Vault 4). Although the significant hydraulic degradation of the disposal unit walls appears to be 

pessimistic, intermediate model results show that the mathematical model does not reflect 

DOE’s stated purpose for the Synergistic Case because the wall degradation lowers the 

predicted dose. Because almost all of the infiltrating water bypasses most of the saltstone 

inventory in the cases DOE supplied with the PA, these cases do not appear to realistically 

predict potential doses from saltstone. In addition to these deterministic cases, DOE also 

performed probabilistic analyses. However, because of concerns about the design and 

implementation of the probabilistic model, the NRC staff did not rely on the probabilistic model in 

its compliance evaluation. 

NRC’s evaluation focused largely3

 on the technical factors related to the oxidization and 

mobilization of Tc. The staff questioned support for several of DOE’s assumptions in its base 

case analysis, including (1) the lack of saltstone fractures, given that cracking of saltstone 

already has been observed; (2) the performance provided by the roof and lower drainage layer 

in shedding over 99% of the water around the disposal units throughout a 10,000 year 

performance period; and (3) the basis for a number of parameters (e.g., hydraulic conductivity, 

Tc sorption coefficients), because recent research does not support DOE assumptions. The 

NRC sent two rounds of questions, called request for additional information. In the second 

round, NRC requested a revised base case to address assumptions in DOE’s base case that 

the NRC staff had concluded were unrealistically optimistic or inadequately supported. For 

example, the NRC staff requested that the revised base case represent degradation of the 

engineered disposal units and saltstone over time (e.g., cracks developing which allows 

additional water to oxidize saltstone and release the mobilized Tc). 

DOE responded to this request with Case K4

, which addresses a number of NRC concerns. 

The roof and disposal unit walls degrade over time, developing cracks and increased water flow. 

The saltstone develops a number of through-going cracks, which results in the saltstone 

becoming oxidized sooner than in earlier DOE cases. The fractures are not represented 

explicitly in the flow model, but water transport through the system is greater than DOE 

assumes in the base case because of greater assumed hydraulic conductivity in the saltstone 

and disposal unit concrete. Although more water flows through the disposal units and saltstone 

in Case K, DOE assumes that the disposal unit concrete acts as a significant barrier to Tc 

release. Because DOE changed the way Tc sorption was modeled from a discrete-fracture 

model to a model based on an average sorption coefficient, Tc does not leave the system 

 

3

 A series of questions focused on Ra because it was a risk driver in Case A and several of the alternate analyses 

DOE submitted with the PA. Since that time, DOE has revised its projected inventory of Ra-226 and its ancestors, 

Th-230 and U-234, such that DOE no longer predicts radium to be a risk-driver. 

4

 DOE provided the NRC three cases related to Case K: Case K, Case K1, and Case K2. The only differences 

between these three cases are the Kd values used to represent Tc sorption in oxidizing and reducing cementitious 

materials (saltstone and disposal unit concrete). When this distinction is not important (e.g., when discussing 

hydraulic properties, which are the same in all three cases), the NRC staff uses the term “Case K” to refer to all three 

cases. The NRC staff differentiates between Cases K, K1, and K2 when discussing values specific to Tc-99 

(e.g., Tc-99 Kd values, release rates, or doses). 
xvii 

through oxidized pathways. Instead, in the Case K model, intermediate model outputs show Tc 

is retained in the disposal unit floor for thousands of years until it nears complete oxidation. 

Thus, in the Case K model, the disposal unit floor captures nearly all of the Tc, like a filter. For 

example, in the FDCs, this filtering effect leads to the disposal unit floor having a modeled Tc 

concentration approximately 13 times greater than the original concentration in the saltstone 

waste itself. The disposal unit floors also are modeled as releasing the Tc much more gradually 

than the saltstone does, which lowers the predicted peak dose. Like the Synergistic Case, the 

Case K mathematical model does not seem to represent DOE’s stated purpose for the case. 

Specifically, although the description of the case indicates increased degradation of the disposal 

unit floor (compared to DOE’s base case), the change in the mathematical model for Tc 

retention in the floor appears to have caused an unintended significant improvement in floor 

performance. There is no support provided for the modeled behavior of the disposal unit floor. 

In all likelihood, the disposal unit floor will develop flow pathways through cracks and joints. The 

NRC staff expects these pathways will oxidize quickly and will not significantly retain Tc. 

Because of DOE assumptions about (1) the way cracks develop over time (i.e., most of the 

cracks develop after 8,000 years) and (2) Tc retention in the disposal unit floor, predicted peak 

doses are delayed by several thousand years to approximately 12,000 - 14,000 years. DOE 

calculated the peak doses to be between 0.5 mSv/yr (50 mrem/yr) and 0.9 mSv/yr (90 mrem/yr). 

The NRC staff concludes that the information supporting the delay of the peak by these two 

assumptions is weak and, therefore, there is not reasonable assurance that these peak doses 

will occur after 10,000 years. That is, with more supportable assumptions, it is likely the peak 

dose will be predicted to occur before 10,000 years. However, the NRC staff noted that certain 

modeling assumptions in Case K (e.g., the suddenness of fracturing and the suddenness of 

release in the Tc release model DOE used) appeared to be pessimistic (i.e., to over estimate 

dose). The NRC staff performed independent analyses to evaluate whether these assumptions, 

when combined with DOE’s more optimistic assumptions (i.e., the superior chemical 

performance of the disposal unit floors), still led to an over estimate the predicted dose. The 

NRC staff concluded that, based on current information, the predicted dose is between 

approximately 0.25 mSv/yr (25 mrem/yr) and approximately 1 mSv/yr (100 mrem/yr). 

DOE has indicated that it continues to believe that the limited degradation in Case A, their 

original base case, appropriately models the future behavior of the system. DOE also indicated 

it believes Case K is extremely pessimistic. NRC disagrees with this characterization and 

concludes that, with certain exceptions (e.g., the chemical performance of the disposal unit 

floors), Case K provides a more realistic estimate of the future behavior of the system than 

DOE’s selected base case. 

Conclusions in this TER are based on the NRC staff’s review of the 2009 PA dated 

October 29, 2009, DOE responses to the NRC staff’s requests for additional information, 

supporting references, and information provided during publicly-documented meetings and 

teleconferences between DOE and NRC, as well as NRC staff analyses as documented in this 

TER. If, in the future, DOE determines it is necessary to revise the assumptions, analysis, 

design, or waste management approach related to any aspects of the disposal strategy at the 

SDF in a way that may affect DOE’s compliance with the 10 CFR 61 performance objectives, 
xviii 

DOE should provide details of these changes to the NRC once approved by the sites regulatory 

body (South Carolina Department of Health and Environmental Control). This NRC staff 

assessment is not a precedent for any future decisions regarding non-HLW or incidental waste 

determinations at SRS or other DOE sites. 
1 

1. Introduction 

1.1 Monitoring Background 

In October 2004, the National Defense Authorization Act for Fiscal Year 2005 (NDAA) was 

signed into law. Section 3116 of the NDAA requires that (a) U.S. Department of Energy (DOE) 

consult with U.S. Nuclear Regulatory Commission (NRC) on its non-High-Level Waste (HLW) 

determinations5

 and, once a determination is complete, that (b) NRC monitor DOE’s disposal 

actions to assess compliance with 10 CFR Part 61, Subpart C, low-level waste disposal facility 

performance objectives6

. As stated in Section 3116(a), for waste to be determined to be 

incidental to reprocessing rather than HLW, DOE’s disposal actions must meet the following 

criteria: 

(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 

§61.55, and will be disposed of— 

(i) in compliance with the performance objectives set out in subpart C of 10 CFR 

Part 61; 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 §61.55, but will 

be disposed of– 

(i) in compliance with the performance objectives set out in subpart C of 10 CFR 

Part 61; 

(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 

Following DOE’s completion of the basis documents for disposal, in consultation with the NRC, 

DOE begins disposal activities and NRC moves into the role of monitoring these disposal 

actions, per Section 3116(b). As monitor, NRC, in coordination with the covered State, is 

required to assess DOE’s compliance with subparagraphs (A) and (B) of subsection (a)(3) listed 

above. 

 

5

 A non-HLW determination (or waste determination [WD]) is a technical analysis that DOE uses to evaluate whether 

waste is incidental to reprocessing, or alternatively, is HLW, based on the risks it poses. A WD documents whether 

DOE’s proposed disposal action will meet the applicable incidental waste criteria. A WD often is supplemented by a 

performance assessment (PA). 

6

 Section 3116 of the NDAA is applicable only to South Carolina and Idaho and does not apply to waste transported 

out of those States. 
2 

As stated above, as monitor, NRC assesses compliance with 10 CFR 61, Subpart C, low-level 

waste disposal facility performance objectives. The objectives are as follows: 

§61.40: General requirement - Land disposal facilities must be sited, designed, 

operated, closed, and controlled after closure so that reasonable assurance 

exists that exposures to humans are within the limits established in the 

performance objectives in §61.41 through §61.44. 

§61.41: Protection of the general population from releases of radioactivity - 

Concentrations of radioactive material which may be released to the general 

environment in ground water, surface water, air, soil, plants, or animals must not 

result in an annual dose exceeding an equivalent of 25 millirems to the whole 

body, 75 millirems to the thyroid, and 25 millirems to any other organ of any 

member of the public. Reasonable effort should be made to maintain releases of 

radioactivity in effluents to the general environment as low as is reasonably 

achievable. 

§61.42: Protection of individuals from inadvertent intrusion - Design, 

operation, and closure of the land disposal facility must ensure protection of any 

individual inadvertently intruding into the disposal site and occupying the site or 

contacting the waste at any time after active institutional controls over the 

disposal site are removed. 

§61.43: Protection of individuals during operations - Operations at the land 

disposal facility must be conducted in compliance with the standards for radiation 

protection set out in part 20 of this chapter, except for releases of radioactivity in 

effluents from the land disposal facility, which shall be governed by §61.41 of this 

part. Every reasonable effort shall be made to maintain radiation exposures as 

low as is reasonably achievable. 

§61.44: Stability of the disposal site after closure - The disposal facility must 

be sited, designed, used, operated, and closed to achieve long-term stability of 

the disposal site and to eliminate to the extent practicable the need for ongoing 

active maintenance of the disposal site following closure so that only 

surveillance, monitoring, or minor custodial care are required. 

In March 2005, DOE submitted its draft basis document for the Saltstone facility, initiating 

NRC’s review under Section 3116(a) of the NDAA. In December 2005, following multiple 

interactions between the agencies, the NRC published its review of the draft basis document for 

the Saltstone facility in a technical evaluation report (NRC, 2005). 

The NRC’s 2005 review concluded with a finding of reasonable assurance for each of the 

Section 3116(a) criteria, with conditions (Table 1.1-1). 
3 

Table 1.1-1: Conditions of NRC's 2005 Findings Regarding the NDAA Section 3116(a) 

Criteria 

Criterion 2005 Conclusions 

One Given the conclusion in the 2005 TER that the NRC staff has reasonable 

assurance that salt waste disposal at the Saltstone Disposal Facility (SDF) meets 

the performance objectives of 10 CFR 61, Subpart C, based on certain 

assumptions, and the lack of other considerations that necessitate disposal of 

the salt waste in a geologic repository, the NRC staff concluded it had 

reasonable assurance that salt waste meets Criterion One. 

Two DOE’s conclusion that highly radioactive radionuclides would 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, was found to be 

reasonable. 

Three The NRC concluded it had reasonable assurance that salt waste disposal at the 

SDF would meet the performance objectives of 10 CFR Part 61, Subpart C, with 

the following provisions: 

§61.41: The assumptions [as stated in Section 4.3 of the NRC’s 2005 TER 

(NRC, 2005)] relevant to the performance objective are verified 

during monitoring. 

§61.42: A design for long-term erosion control is implemented, which greatly 

reduces the likelihood of an agricultural intruder scenario occurring. 

§61.43: During operations, individuals are protected by DOE regulations, 

which were demonstrated to provide protection 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 Saltstone Production Facility (SPF) and SDF, (4) regulatory 

and contractual enforcement mechanisms, and (5) access controls, 

training, and dosimetry. 

§61.44: None. 

One of the main outcomes of the 2005 review (NRC, 2005) was the development of monitoring 

factors, for use in NRCs monitoring role under NDAA Section 3116(b). In general, verification of 

DOE’s assumptions made in assessing whether the 10 CFR 61 performance objectives can be 

met should be performed by DOE. However, because some of the assumptions made in the 

analysis, if incorrect, could lead to noncompliance with the performance objectives, NRC has 

monitored these assumptions as part of its responsibilities under the NDAA. These 

assumptions fall into the following general groups: waste form and vault degradation, the 
4 

effectiveness of infiltration and erosion controls, and estimation of the radiological inventory. 

The NRC staff concluded that certain factors are important to assessing whether DOE’s 

disposal actions will be compliant with the performance objectives. Based on the review 

described in the 2005 TER (NRC, 2005), the NRC staff crafted the following eight “key” factors 

critical to NRC monitoring of the Saltstone facility. 

Table 1.1-2: Key Factors Resulting from NRC Review of 2005 Saltstone PA 

Key Factor Details 

1. Oxidation 

of Saltstone 

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 SDF and 

therefore whether §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. Hydraulic 

Isolation of 

Saltstone 

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 §61.41. Degradation mechanisms that may result in the hydraulic 

conductivity of degraded saltstone and vault concrete being larger than 

1x10-7 cm/s (1x10-1 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 Toxicity Characteristic Leaching 

Procedure (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. 

3. Model 

Support 

Adequate model support is essential to assessing whether the SDF can meet 

§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 

PA results. 
5 

Key Factor Details 

4. Erosion 

Control 

Design 

The erosion control design is important to ensuring that §61.42 can be met 

because it eliminates pathways and scenarios for intruder dose assessments. 

Implementation of an adequate design that (1) does not deviate significantly 

from the information submitted to the NRC in CBU-PIT-2005-00203 Rev 1 and 

the associated references, or, (2) if it does deviate significantly, is reviewed by 

NRC staff to ensure that the revisions are consistent with long-term erosion 

control design principles is important. 

5. Infiltration 

Barrier 

Performance 

The infiltration control design is important to ensuring that §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. Feed Tank 

Sampling 

Implementation of an adequate sampling plan is important to ensuring that 

§61.41 and §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. Tank 48 

Waste form 

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 waste form made from this 

waste will provide for suitable waste form 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. Removal 

Efficiencies 

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 an 

important factor in the determination that §61.41 and §61.42 can be met. 

Following the January 2006 completion of the DOE’s final waste determination for salt waste 

disposal (DOE-WD-2005-001), the NRC developed a monitoring plan (NRC, 2007a). The plan 

describes activities designed to monitor DOE’s disposal activities as they relate to the eight key 

factors identified as part of NRC’s PA review (Table 1.1-2) as well as other relevant activities 

that were identified early in the monitoring process (e.g., review of environmental monitoring 

data and worker dose records). 

The NRC’s monitoring plan, finalized in May 2007 (NRC, 2007a), describes two primary types of 

monitoring activities the NRC staff performs: (1) technical reviews of DOE data and analyses 

and (2) onsite visits to observe DOE's disposal actions. With regard to the role of technical 
6 

reviews under monitoring, DOE’s disposal activities at the site are often complex processes, 

which rely on technical documents as basis for disposal. Should any changes or revisions be 

made to key technical documents (e.g., to describe use of new technologies or methods, or to 

reflect changes requested by SC DHEC in its regulatory role or by NRC in its monitoring role), 

assumptions are reviewed again to confirm continued compliance with the performance 

objectives. Since the Saltstone PA is a critical element of DOE demonstration of compliance 

with the performance objectives, any updates to the Saltstone PA must also be assessed under 

NRC’s monitoring role. In coordination with the covered State, NRC’s onsite observation 

program involves trips to the site to observe and review certain operations that might affect 

compliance with the performance objectives as they are being performed. The NRC staff 

observes operational details, evaluates the implementation and purpose, and reviews 

documentation helpful to assessing the impact the process has on compliance. 

Section 3.1.9 of the NRC monitoring plan for Saltstone, Performance Assessment Process 

Review, provides an explanation of how the NRC staff evaluates revisions made to the PA 

(NRC, 2007a). In brief, the staff uses an approach similar to that used in the original technical 

evaluation of the waste disposal actions (NRC, 2005). Where practical, the NRC staff performs 

its own independent assessments of the more risk-significant aspects of the performance 

assessment. Sensitivity and uncertainty analyses can be especially useful in determining the 

effects and importance of changes to the PA models. When a deterministic analysis is used for 

a PA, the NRC staff evaluates whether the analysis provides a demonstrably conservative 

estimate of potential doses, or DOE has provided appropriate support and bases for key 

assumptions or parameters in the PA. Determining the important features of a deterministic 

analysis may be difficult, which is one reason why it may be advantageous for the staff to 

undertake its own independent assessment. If a probabilistic analysis is used, the staff must 

ensure that appropriate parameter distributions are used and that the analyses appropriately 

account for parameter uncertainty and possible parameter correlation. 

As of this report, NRC has conducted 11 onsite observations since inception of monitoring in 

2006. Each observation covers a range of topics applicable to one or more performance 

objectives (Table 1.1-3). In addition, NRC has conducted numerous technical reviews topics 

covered in the monitoring plan. 

On occasion, an issue of relatively high risk significance arises during monitoring activities that 

could affect compliance. These Open Issues require additional follow-up by the NRC staff or 

additional information from DOE to address questions that the NRC staff has raised regarding 

DOE disposal actions. Issues such as these retain the title of Open until adequately addressed 

by DOE. 

Since the first onsite observation, a total of four Open Issues have been documented. Of these, 

one Open Issue has been closed (Open Issue 2007-3), and three remain open (Table 1.1-4). 

NRC monitoring of the SDF has continued during the development of this document. NRC staff 

has conducted five onsite observations and has participated in multiple meetings regarding 

DOE disposal activities since receiving the 2009 PA. 
7 

Table 1.1-3: History of Onsite Observation Activities 

Date Discussion Topics Open Issue Activity 

October 2007 Grout Formulation and Placement 

Vault Construction 

Waste Sampling 

Radiation Protection Program 

2007-1 Opened 

2007-2 Opened 

2007-3 Opened 

March 2008 Saltstone Characterization 

Vault Operation and Characterization 

Waste Sampling and Inventory 

Radiation Protection Program 

2007-3 Closed 

July 2008 Saltstone Characterization 

Waste Sampling 

NRC Staff Technical Reviews 

March 2009 Disposal Cell Construction 

Radionuclide Inventory 

NRC Staff Technical Reviews 

2009-1 Opened 

June 2009 Disposal Cell Construction 

August 2009 Performance Assessment Process Review 

February 

2010 

Disposal Cell Construction 

SPF Operation 

PA Process Review 

Radionuclide Inventory 

April 2010 Disposal Cell Construction 

July 2010 Saltstone Quality Assurance Plan 

Hydro-test results on Cell 2A and 2B 

Saltstone Core Samples 

May 19, 2010 Saltstone Inadvertent Transfer 

January 2011 Vault 4 Integrity 

SPF Operations 

April 2011 Radionuclide Inventory 

Waste Oxidation and Tc Release Research 

Disposal Cell Construction 

Topics from Previous Observations 
8 

Table 1.1-4: Saltstone Disposal Facility Open Issues 

Number Brief Description of Issue Identified

2007-1 

(Open)1

DOE should determine the hydraulic and chemical properties of asemplaced saltstone grout. 10/2007 

2007-2 

(Open)1

DOE should demonstrate that intrabatch variability, flush water additions 

to freshly poured saltstone grout at the end of each production run, and 

additives used to ensure processability are not adversely affecting the 

hydraulic and chemical properties of the final saltstone grout. 

10/2007 

2007-3 

(Closed) 

DOE should reassess the risk significance of the as-built conditions of 

Vault 4 in light of the presence of contaminated seeps on the exterior 

wall of Vault 4. 

10/2007 

2009-1 

(Open)1

DOE should demonstrate that (1) Tc-99 in salt waste is strongly retained 

in saltstone grout and (2) the sorption of dissolved Tc-99 onto saltstone 

grout and vault concrete is consistent with the Kd values for Tc-99 

assumed in the PA. 

03/2009 

1

 Issues remaining Open prior to review of the 2009 PA will be incorporated into the NRC staff’s revision to 

its monitoring plan for the SDF. These actions will be closed upon issuance of this TER. 

1.2 Current Review 

In 2008, DOE decided to revise the PA for the SDF to reflect a new vault design, new vault and 

grout testing information, emergent facility conditions, and areas of uncertainty identified by 

NRC staff during its review of the 2005 Saltstone PA (NRC, 2008b). In May 2008, DOE hosted 

an interagency meeting between DOE, SC DHEC, and NRC in which DOE provided an 

overview of the revised design and received comments from both NRC and SC DHEC 

(NRC, 2008b). Following this meeting, once approved by SC DHEC, DOE began construction 

of the new disposal vaults (referred to as Future Disposal Cells [FDCs]). 

In November 2009, DOE submitted a new PA7

 for the SDF (SRR-CWDA-2009-00017) to the 

NRC for review in the NRC’s capacity as a monitor of DOE’s disposal actions at the SDF in 

accordance with Section 3116(b) of the NDAA. In accordance with these responsibilities, the 

NRC staff began review of the 2009 Saltstone PA in November 2009. The purpose of the 

review is to assess whether DOE’s disposal actions, as described in the PA, meet the 

performance objectives of 10 CFR 61, Subpart C. The review included two Requests for 

Additional Information (RAI). The RAI process under the NDAA typically involves the NRC 

preparing an RAI within 90 days of receiving the PA and DOE responding to that RAI within 90 

days. The NRC sent the first RAI (RAI-2009-01) on March 31, 2010 (NRC, 2010b), to which 

DOE responded on July 22, 2010 (SRR-CWDA-2010-00033). After review of DOE’s responses, 

the NRC staff concluded that some of these responses did not fully address the NRC’s 

 

7

 Throughout this TER, the staff references the new PA by either its DOE document number 

(SRR-CWDA-2009-00017), “the PA,” or “the 2009 PA.” 
9 

questions and concerns with assumptions in DOE’s expected degradation case (or base case). 

The NRC sent a second RAI (RAI-2009-02) on December 15, 2010 (NRC, 2010i), to which DOE 

responded in draft form (SRR-CWDA-2011-00044 Rev. 0) and discussed at a public meeting on 

April 26, 2011, and then submitted a final version on August 26, 2011, 

(SRR-CWDA-2011-00044 Rev. 1). To respond to the NRC’s concerns with DOE’s base case, 

DOE developed a new case called Case K. 

The NRC and DOE staffs met frequently throughout the review process, and held additional 

meetings to discuss the NRC’s concerns and the details of the new Case K. The meetings in 

which the NRC staff conducted and/or took part are listed below in Table 1.2-1. All of the 

summaries of these meetings are available to the public and can be found via the associated 

Agencywide Documents Access and Management System (ADAMS) Accession numbers in the 

third column. With the exception of the technical exchange meetings, all of the meetings listed 

in Table 1.2-1 were open to members of the public. 

In a letter dated May 20, 2011, following DOE’s decision to respond to the NRC’s concerns by 

developing the new Case K, DOE requested NRC acceptance of the new approach. On 

July 20, 2011, in response to this letter, the NRC stated that final conclusions on the new 

approach could only be reached after NRC staff completes its detailed review of DOE’s written 

responses to RAI-2009-02 (NRC, 2011i). 

Although Case K resolves many of the NRC staff’s concerns with DOE’s selected base case, 

the NRC staff identified both overly-optimistic and potentially overly-pessimistic assumptions in 

this new case. The impact of these assumptions was not immediately clear because the Case 

K analysis supplied by DOE did not include an uncertainty or sensitivity analysis that addressed 

these assumptions (NRC, 2011g). In addition, the staff noted unexpected intermediate results 

in the Case K computational model files. To better understand the uncertainty in the Case K 

results, the effects of conflicting optimistic and pessimistic assumptions, and the causes of the 

unexpected intermediate results, the staff performed its own confirmatory analyses related to 

Case K. 

Table 1.2-1: NRC/DOE Meetings during PA Review 

Meeting Date Meeting Purpose Meeting Summary 

(ADAMS Acc.) 

December 8, 2009 Commencement Meeting for NRC Review of 

Revised PA 

ML100050554 

April 19, 2010 Clarification Discussion Following DOE Receipt of 

NRC’s RAI 

ML101440307 

September 2, 2010 Clarification Discussion Following NRC Receipt of 

DOE’s RAI Response 

ML102980289 

January 27, 2011 Detailed Discussion Following DOE Receipt of 

NRC’s Second RAI 

ML110341424 
10 

Meeting Date Meeting Purpose Meeting Summary 

(ADAMS Acc.) 

April 27, 2011 Discussion Regarding DOE Potential Responses 

to Second RAI 

ML111950042 

May 3, 2011 Technical Exchange #1 – NRC/DOE Staff-Level 

Discussion Concerning Base Case 1

ML111440310 

May 5, 2011 Technical Exchange #2 – NRC/DOE Staff-Level 

Discussion Concerning Base Case 1

ML111440270 

May 12, 2011 Technical Exchange #3 – NRC/DOE Staff-Level 

Discussion Concerning Base Case 1

ML111440857 

May 17, 2011 Technical Exchange #4 – NRC/DOE Staff-Level 

Discussion Concerning Base Case 1

ML111440864 

May 19, 2011 Technical Exchange #5 – NRC/DOE Staff-Level 

Discussion Concerning Base Case 1

ML111440873 

June 2, 2011 Culmination of Technical Exchanges and 

Discussion of Path Forward 

ML111780433 

1 A technical exchange is a staff-level, non-public discussion involving no decision-making. Summaries 

from technical exchanges are available to the public. 

1.3 Site Overview 

The 780 square kilometer (300 square mile) DOE SRS, located in south-central South Carolina, 

began operation in 1951 producing nuclear material for national defense, research, medical, 

and space programs. Waste produced at the site from spent nuclear fuel reprocessing 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, which were placed into operation between 1954 and 1986. Of the original 

51 storage tanks, 49 are still operational. The waste stored in the tanks at SRS 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. 

DOE removes the salt waste, treats it to remove highly radioactive radionuclides to the 

maximum extent practical, and disposes the low activity fraction on site in the SDF. 

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

SRS site boundary on a well-drained local topographic high (Figure 1.3-1). 
11 

Figure 1.3-1: Location of the Z-Area (based on CBU-PIT-2005-00146) 

The Saltstone Facility consists of two facility segments, the SPF, which receives and treats salt 

solution to produce solidified saltstone, and the SDF, which consists of existing vaults and 

projected FDCs used for the final disposal of the solidified saltstone. The SPF is permitted by 

the State of South Carolina as a wastewater treatment facility. The SDF is permitted by the 

State as a Class 3 Landfill. 

Vaults 1 and 4, which were constructed in the 1980s, are constructed of reinforced concrete 

containing blast furnace slag. Vault 4 (Figure 1.3-2) is 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). 

The roof is sloped, with a minimum slope of 0.19 cm/m (0.15 in/ft) from the middle of the 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. Vault 1 (Figure 1.3-3) is 30.5 m (100 ft) wide, by 183 m (600 ft) long, 

by 8.2 m (27 ft) high. It is divided into six cells of approximately 30 m (98 ft) by 30 m (98 ft). 

Vault 1 originally had a temporary rolling roof. This temporary roof was later removed, and 

three of the cells in Vault 1 have been filled with grout and covered with a permanent roof. 

Nearly six of the FDCs (Figure 1.3-4) were fully constructed prior to publication of this TER. The 

diameter of the cells is 45.7 m (150 ft), with an interior height of 6.7 m (22 ft) (this height will 

increase to 7.16 m (23.5 ft) at the center of the cell). The cylindrical, reinforced concrete cells 

are constructed below grade. Native soil is used to backfill around the completed cells. The 

remaining capacity of Vault 4 and up to a total of 64 FDCs may be required to dispose of the 
12 

total volume of saltstone produced. An engineered cap will be added to the site at closure to 

cover the disposal units. For perspective, Figure 1.3-5 shows the entire SDF. 

Figure 1.3-2: Vault 4 at the SDF (SRR-CWDA-2010-00013) 

Figure 1.3-3: Vault 1 at the SDF (2009 PA) 
13 

Figure 1.3-4: Disposal Cells 2A and 2B (and Approximate Design of FDC’s) at the SDF 

(SRR-CWDA-2011-00082) 

Figure 1.3-5: Areal Picture of SDF 
14 

2. Protection of the General Population 

2.1. Performance Assessment Overview 

DOE conducted the 2009 PA to evaluate the potential long-term dose to a member of the public 

(i.e., to assess compliance with the performance objective in §61.41). In this PA, the dose to an 

offsite receptor located 100 m from the disposal site was estimated using projected releases 

into the air and groundwater from waste disposed in Vaults 1, 4, and the FDCs. DOE also 

performed an intruder assessment (which DOE evaluated as part of the PA) to evaluate the 

potential dose to an inadvertent intruder who inhabits the site after the end of institutional 

controls (i.e., to assess compliance with §61.42). This evaluation is described in Chapter 3. 

The 2009 PA consisted of a deterministic analysis for a case that DOE considers to represent 

expected future conditions, deterministic analyses of alternate cases, and a probabilistic 

sensitivity and uncertainty analysis. The infiltration rates through the upper layers of the closure 

cap (i.e., from the vegetative cover layer down to the lower backfill layer) were determined using 

the Hydrologic Evaluation of Landfill Performance (HELP) code. The flow through the lower 

layers of the closure cap (i.e., layers below the foundation layer in Figure 2.4-2) is modeled 

using the PORFLOWTM code with the infiltration rates from the HELP code acting as a boundary 

condition. In the deterministic cases, PORFLOWTM was also used to model the fluid flow and 

contaminant transport in the disposal units and the vadose and saturated zones to determine 

the concentrations of radionuclides in the groundwater. A GoldSim® model that was run in 

deterministic mode was then used to convert these groundwater concentrations to doses. The 

air pathway analysis was performed in a separate PORFLOWTM calculation and is described in 

Section 2.9. Sensitivity and uncertainty analyses were performed using a probabilistic 

GoldSim® model that was developed by DOE. As discussed in more detail in Section 2.11, the 

probabilistic model used near-field flow rate information that was calculated by PORFLOWTM in 

the deterministic cases, and factors affecting the near field flow rates were not evaluated 

probabilistically. 

2.1.1 Description of Deterministic PORFLOWTM Cases 

The deterministic cases evaluated by DOE using the PORFLOWTM code are described in 

Table 2.1-1. These cases consisted of the case that DOE considers to be the “base case” or 

expected case (i.e., Case A) as well as other cases that are considered by DOE to be sensitivity 

cases. 
15 

Table 2.1-1: Description of Deterministic Cases Modeled by DOE 

Case Description of Key Model Assumptions Reference 

Case A DOE considers this the “base case,” or expected case.

Key assumptions include: 

− Saltstone intact for the duration of the assessment 

− Vault 1 and 4 walls degraded at t=0 

− FDC walls intact for the duration of the assessment

SRR-CWDA-2009-00017 

(p 249) 

Case B Same properties as Case A except: 

− Fast flow path present in the sheet drain system 

(i.e., along the walls from the roof through the floor) 

for Vault 4 and the FDCs 

SRR-CWDA-2009-00017 

(p 249) 

Case C Same properties as Case A except: 

− Fast flow path present in the sheet drain system 

(i.e., along the walls from the roof through the floor) 

for Vault 4 and the FDCs 

− Fast flow through cracks in saltstone 

SRR-CWDA-2009-00017 

(p 249) 

Case C 

PA-4 Case 

Same assumptions as Case C, but additional 

radionuclides were included 

SRR-CWDA-2011-00044 

(p 22) 

Case D Same properties as Case A except: 

− Capillary break present at sheet drains in Vault 4 

and the FDCs 

SRR-CWDA-2009-00017 

(p 249) 

Case E Same properties as Case A except: 

− Saltstone severely degraded with a hydraulic 

conductivity of 1.7x10-3 cm/s 

SRR-CWDA-2009-00017 

(p 249) 

No Closure 

Cap (with 

credit for 

the lower 

lateral 

drainage 

layer) 

Same properties as Case A except: 

− Cap layers (composite hydraulic barrier, lateral 

drainage layers, and erosion control layer) modeled 

with the properties of the surrounding soil 

SRR-CWDA-2009-00017 

(p 541) 

10x Sulfate 

Attack 

Same properties as Case A except: 

− Diffusion coefficient used in prediction of sulfate 

attack on the concrete in the disposal unit walls and 

floors increased by factor of 10 

SRR-CWDA-2009-00017 

(p 542) 

No Sulfate 

Attack 

Same properties as Case A except: 

− There is assumed to be no damage from ettringite 

formation to the disposal unit walls and floors. 

Disposal units do not degrade over time 

SRR-CWDA-2009-00017 

(p 542) 
16 

Case Description of Key Model Assumptions Reference 

Oxidized 

Concrete 

Same properties as Case A except: 

− Vault 1 and 4 walls oxidized at closure 

SRR-CWDA-2009-00017 

(p 550) 

Increased 

Saltstone 

Hydraulic 

Conductivity 

Same properties as Case A except: 

− Hydraulic conductivity of saltstone increased to 

1x10-7 cm/s (from 2x10-9 cm/s) 

SRR-CWDA-2009-00017 

(p 551) 

Synergistic 

Case 

Evaluates the effect of an assumed increase in several

degradation mechanisms: 

− Closure cap degraded at a faster rate 

− Vault 1 and 4 walls initially middle aged oxidized 

concrete and transition to old aged concrete at 500 

years 

− Vault 1 and 4 walls degraded hydraulically to soil 

properties at 500 years 

− FDC concrete degraded chemically and 

hydraulically at 500 years 

− Saltstone assumed cracked at closure and oxygen 

diffuses into the monolith via cracks 

SRR-CWDA-2009-00017 

(p 548) 

Synergistic 

Case 

VP-1 

Same assumptions as the Synergistic Case, but 

assumed a relative permeability of 1 (i.e., moisture 

characteristic curves were not used) 

SRR-CWDA-2010-00033 

VP-1 

Synergistic 

Case 

PA-9 

Same assumptions as Synergistic Case, but additional 

radionuclides were included 

SRR-CWDA-2011-00044 

(p 106) 

Synergistic 

Case 

Updated 

PA-9 

Same assumptions as Synergistic Case - PA-9, but 

updated dose methodology was used 

SRR-CWDA-2011-00044 

(p 107) 
17 

Case Description of Key Model Assumptions Reference 

Case K Evaluates the effect of revising multiple aspects of the 

PA to address concerns raised by NRC in the RAIs. 

Key assumptions in this case include: 

− Saltstone saturated hydraulic conductivity 

assumed to have an initial value of 1x10-7 cm/s 

(base case value 2x10-9 cm/s) and to degrade to 

1x10-6 cm/s 

− Saltstone diffusivity assumed to degrade from 

1x10-7 cm2

/s (base case value) to 5x10-6 cm2/s 

− Moisture characteristic curves not used in 

determining flow through unsaturated cementitious 

material (i.e., relative permeability assumed to be 

1) 

− Number of pore volumes required for Eh and pH 

transitions decreased 

− An average Kd model (referred to by DOE as a 

“single porosity” model) abstracted from shrinking 

core results is used to model Tc release instead of 

an explicit shrinking core model to accommodate 

an increase in the postulated number of saltstone 

fractures 

− Increased degradation of disposal unit concrete 

− Inventory of Pu-238, U-234, Th-230, and Ra-226 

revised to lower values 

− Kd values for saltstone, disposal unit concrete, and 

soil updated 

− Tc Kd assumed to be 1000 mL/g for reduced 

saltstone and 10 mL/g for oxidized saltstone 

− Dose methodology updated (biotic transfer factors 

updated, chicken and egg pathway included, 25 

year buildup in soil assumed, leafy vegetables 

included in plant transfer factor) 

SRR-CWDA-2011-00044 

(p 75) 

Case K1 Same assumptions as Case K, but the Kd values for 

Tc were assumed to be 500 mL/g for reduced 

saltstone and 0.8 mL/g for oxidized saltstone 

SRR-CWDA-2011-00044 

(p 218) 

Case K2 Same assumptions as Case K, but the Kd value for Tc 

was assumed to be 500 mL/g for reduced saltstone. 

The oxidized saltstone Kd remained 10 mL/g for this 

case 

SRR-CWDA-2011-00044 

(p 219) 

 
18 

2.1.2 Description of GoldSim® Probabilistic Analysis 

In addition to the deterministic PORFLOWTM models, DOE also developed a probabilistic 

GoldSim® model to evaluate the uncertainty in model results and the sensitivity of predicted 

dose to various assumptions. The DOE GoldSim® model can either run Cases A - E individually 

or to run “all cases” (including Cases A - E). When the model is run in the “all cases” mode, a 

case is selected randomly for each realization based on DOE’s predicted probabilities for each 

of these cases occurring, which were based on DOE judgment (2009 PA; Table 5.6-3). The 

other deterministic cases listed in Table 2.1-1 were developed after the GoldSim® model, and 

are therefore not included in this model. The probabilistic model relies on output from the 

deterministic models as input for some parameters and processes. For example, the onedimensional near-field flow rates through the grouted systems are abstracted from twodimensional, deterministic PORFLOWTM models. 

As part of the development of the probabilistic model, DOE conducted a multi-step model 

adjustment process, which DOE referred to as benchmarking. During benchmarking, DOE 

modified the GoldSim® model to optimize the match between intermediate model results (i.e., 

flux rates and peak well doses) of corresponding simulations run in its deterministic and 

probabilistic models. GoldSim® parameters adjusted included (1) a pseudo-Kd for Tc-99, 

(2) flow multipliers for the upper aquifer zone, disposal unit floor, disposal unit wall, and grout, 

and (3) flow multipliers to adjust contributions from Vault 4, adjust flows to represent a flow 

divide at the site, and to adjust contributions from certain individual disposal units. 

2.1.3 NRC Evaluation – PA Approach 

NUREG-1854, the NRC Staff Guidance for Activities Related to U.S. Department of Energy 

Waste Determinations (NRC, 2007b) states that either a deterministic or probabilistic analysis 

can be used by DOE to demonstrate compliance with the performance objectives in 10 CFR 61, 

Subpart C, although probabilistic approaches are preferred for complex assessments. 

Probabilistic assessments are preferred because a deterministic model, without additional 

sensitivity analyses, gives no indication of the sensitivity of the results to certain parameters or 

of the importance of the uncertainty in the parameters. The use of a deterministic approach 

may result in the need for stronger justification of code input parameter values and may require 

further analysis of doses using upper or lower bounding conditions to gain insights into the 

range of dose estimates. Additionally, in a deterministic analysis, key parameters should be 

reasonably conservative or well justified unless adequate sensitivity analyses have been 

provided that demonstrate the overall risk significance of the parameters is small. These 

sensitivity analyses need to consider how uncertainties in different parameters are interrelated. 

If the sensitivity of interrelated parameters is only evaluated in “one-off” analyses in which one 

parameter is varied at a time, the potential effect of the parameters on the dose could be 

missed. For example, a degraded cap might also lead to an increased rate of degradation of 

the disposal units and waste form due to the increased infiltration. Additionally, the sensitivity 

analyses should consider the relative change in dose based on changes to the parameter, 

rather than absolute changes. 
19 

DOE’s overall approach of performing a deterministic base case analysis, deterministic 

sensitivity cases, and a probabilistic sensitivity/uncertainty analysis is consistent with the NRC’s 

guidance. However, NRC staff does not agree with some of the specific ways in which this 

approach was implemented, such as the selection of Case A as the base case. NRC staff also 

has significant concerns about key aspects of DOE’s probabilistic analysis. In particular, the 

NRC staff finds that DOE’s approach of using certain outputs of the deterministic model directly 

in the probabilistic model has obscured key factors that had the greatest effect on dose results. 

For example, the use of near-field flow rate information calculated by PORFLOWTM in the 

deterministic cases, limited the evaluation of factors affecting the near field flow rates in the 

probabilistic model. 

NRC staff disagrees with DOE’s determination that Case A represents the expected, or most 

probable, case (i.e., the base case). As described in the NRC RAIs (NRC, 2010b, i) the PA 

base case scenario is unrealistic and non-conservative for the following reasons: 

(i) The base case model is inconsistent with known conditions. Significant site 

characteristics that have not been adequately incorporated into the model include the 

following: 

• Fractured saltstone is not considered in the base case even though fracturing of 

saltstone has been observed. In addition, shrinkage has been observed and is 

not included in the model. (Section 2.6) 

• The PA models appear to be inconsistent with observed, advective contaminant 

releases from Vault 4. (Section 2.5) 

• Material interfaces have shown to be relevant to performance; however they are 

not considered in the PORFLOWTM model. (Section 2.5) 

(ii) The base case model does not adequately account for the observed range of 

experimentally measured values representing saltstone and disposal unit initial 

conditions, or the uncertainty in the predicted temporal evolution of conditions. NRC 

concerns with parameter and conceptual model uncertainty include the following: 

• The hydraulic conductivity and effective diffusion coefficient for saltstone are 

time-invariant as the base case model does not adequately account for temporal 

variation. (Section 2.6) 

• The modeled initial hydraulic conductivity of saltstone does not fully account for 

or attempt to bound differences between the properties of laboratory-prepared 

and full-scale, as-emplaced saltstone. (Section 2.6) 

• The PA does not account for the possible effects of potentially relevant disposal 

unit degradation mechanisms or provide a basis for discounting those 

mechanisms. (Section 2.5) 
20 

(iii) The base case does not have adequate technical bases. NRC concerns with the limited 

model support include the following: 

• Model support for geotextile filter fabrics and the lateral drainage layers is not 

commensurate with their expected long-term performance and risk significance. 

(Section 2.4) 

• The moisture characteristic curves implemented in the base case for intact and 

fractured cementitious materials, which significantly reduce flow, lack adequate 

support considering their risk significance. (Section 2.7) 

• The chemical stability of saltstone provides a significant barrier to transport; 

however, the basis for the Eh-pH evolution of cementitious materials is very 

limited. (Section 2.6) 

• The basis for the adopted Tc pseudo-Kd of 1,000 mL/g is inaccurate and 

insufficient. (Section 2.7) 

Many of the concerns described above for Case A apply to the other PORFLOWTM deterministic 

cases as well. However, in DOE’s response to comments PA-8 and SP-19 

(SRR-CWDA-2011-00044), DOE created new PORFLOWTM cases, Cases K, K1, and K2, to 

address these concerns. Of the cases listed in Table 2.1-1, the NRC staff considers Cases K, 

K1, and K2 to best represent the current and future expected conditions of the disposal units 

and the waste form. Of these three cases, NRC staff considers Case K1 to contain Kd values 

for Tc in saltstone that are the most realistic based on experimental data (Section 2.7). NRC 

staff is therefore relying strongly on Case K1 in its determination of compliance. Although NRC 

staff believes that this case is the best representation of the current and future expected 

conditions of the cases DOE modeled, NRC staff finds that there are some aspects of the 

modeling performed for this case that do not reflect the expected future behavior of the system. 

For example, Cases K, K1, and K2 use an average Kd (referred to by DOE as a “single porosity” 

model) approach to modeling the Tc Kd values for saltstone (Section 2.7). This approach results 

in an underprediction of the release rates when the fraction of saltstone oxidized is low and an 

overprediction of the release rates when the fraction oxidized is high. Also, as described in 

more detail in Section 2.5, intermediate outputs from the PORFLOWTM model indicate that the 

modeled disposal unit performance in Cases K, K1, and K2 is not consistent with the expected 

performance, and the modeled disposal unit performance significantly lowers the predicted peak 

doses. There is also significant uncertainty in the assumptions regarding the timing, rate, and 

final amount of fracturing (Section 2.6). The values assumed for these parameters significantly 

affect the predicted maximum dose as well as the timing of the dose in Cases K, K1, and K2. 

The NRC staff believes that the addition of a probabilistic model to the DOE PA is a valuable 

improvement to their PA and this type of modeling should be continued in the future. However, 

the NRC staff has a number of concerns with the specific probabilistic GoldSim® model used in 

this PA. As discussed in more detail in Section 2.11 and below, the concerns include difficulties 

caused by DOE’s approach of basing parts of the probabilistic model on outputs of the 
21 

deterministic model, adjustments made to the model based on the deterministic output, and 

apparent errors in the probabilistic model. As part of DOE’s approach, the deterministic flow 

rates from PORFLOWTM for Cases A - E were “hard wired” into the probabilistic model. As 

described in the list presented above, the NRC staff has concerns about some of the key 

assumptions in Cases A - E (e.g., hydraulic conductivity, moisture characteristic curves), and, 

as a result, NRC staff finds that the flow rates generated using PORFLOWTM for these cases 

are optimistic. Additionally, because deterministic flow results are “hard wired” into the 

probabilistic models for the near-field, the probabilistic models could not provide information 

about the relative importance of assumptions and parameters affecting flow and near field 

release, which the NRC staff expects to have a significant effect on model results. As 

discussed in Section 2.11, NRC staff has concerns with the benchmarking process used in the 

development of the GoldSim® model (i.e., the process used to modify the GoldSim® model to 

optimize the match between intermediate model outputs in the PORFLOWTM and GoldSim®

models). NRC staff also identified a number of potential errors in the GoldSim® model that 

could significantly affect the results (e.g., key features of Case C not being implemented when 

the model is run in “all cases” mode, incorrect linkages between the unsaturated zone cells, and 

an incorrect value for the Vault selector for Vault 4 [Section 2.11]). For these reasons, the NRC 

staff is not using the results of the DOE GoldSim® model as part of the assessment of 

compliance with the performance objectives. 

NRC staff concludes that Case K1 best represents the current and future expected conditions of 

the SDF. Therefore, the NRC staff is relying strongly on Case K1 in its determination of 

compliance. The NRC staff is not using the uncertainty and sensitivity analyses performed 

using the probabilistic model as part of the assessment of compliance. Instead, the NRC 

evaluation of uncertainty in dose and the sensitivity of dose to input assumptions is based on: 

(1) an evaluation of the results of DOE’s deterministic sensitivity cases, (2) an evaluation of 

intermediate model results from DOE’s deterministic sensitivity analyses, and (3) independent 

sensitivity analyses focused on Cases K, K1, and K2 (Section 2.13). 

2.2 Source Term and Inventory 

2.2.1 Source Term and Inventory 

The methodology used by DOE to estimate the closure inventory in the SDF is described in 

SRNS-J2100-2008-00004. The assumed inventory for Vault 1 is based on the current 

estimated inventory since there are no plans to add additional inventory to this vault. The 

assumed inventory for Vault 4 is based on the current inventory in Vault 4 plus the inventory 

intended for disposal in this vault in the future. The void spaces in the Vault 1 and Vault 4 walls 

are assumed to be filled with pore fluid containing the same concentration of radionuclides as in 

the saltstone. The purpose of this assumption is to evaluate the potential contamination in the 

vault wall weeping that occurred during saltstone placement in these vaults. In Vault 1, 

approximately 0.65% of the inventory is assumed to be in the wall, and in Vault 4, 0.5% of the 

inventory is assumed to be in the wall. The remainder of the inventory for these vaults is 

assumed to be located in the saltstone. For the FDCs, the inventory was determined based on 

the remaining tank farm inventory and assumptions about the removal efficiencies of the 
22 

treatment processes. The total predicted inventory for the FDCs is assumed to be divided 

evenly over all of the FDCs. 

The resulting inventory projected by DOE for the disposal units at the SDF is presented in 

Table 2.2-1. These projected inventories are based on the decay of the radionuclides to the 

date of October 1, 2030, the expected date of closure for the SDF. 

Table 2.2-1: Projected Inventory at Time of Closure 

Vault 1 

(Ci) 

Vault 4 

(Ci) 

Individual FDC

(64 FDCs total)

(Ci) 

Total SDF 

(Ci) 

Ac-227 1.60x10-5 1.70x10-7 2.70x10-5

Al-26 3.40x10-1 1.90x10-1 1.30x101

Am-241 4.70x10-4 1.30x102 1.40 2.20x102

Am-242m 6.70x10-2 5.90x10-4 1.00x10-1

Am-243 1.80 3.70x10-2 4.20 

Ba-137m 4.10 2.80x105 2.20x101 2.80x105

Bk-249 1.80x10-28 1.80x10-28 1.20x10-26

C-14 1.30 2.70x101 2.00 1.60x102

Ce-144 1.80x10-9 3.60x10-10 2.50x10-8

Cf-249 6.50x10-13 6.70x10-13 4.40x10-11

Cf-251 1.20 2.30x10-14 1.20 

Cf-252 1.80x10-18 1.80x10-18 1.20x10-16

Cl-36 7.60x10-4 3.00x10-3 4.20x10-4 3.10x10-2

Cm-242 6.70x10-2 6.30x10-19 6.70x10-2

Cm-243 2.10x10-1 2.10x10-4 2.20x10-1

Cm-244 1.30x102 9.50x10-1 1.90x102

Cm-245 9.20x10-1 2.40x10-4 9.40x10-1

Cm-247 3.90x10-6 7.10x10-14 3.90x10-6

Cm-248 1.20x10-13 7.40x10-14 4.90x10-12

Co-60 8.20x10-5 4.60x10-1 5.40x10-2 3.90 

Cs-134 5.20x10-1 1.50x10-5 5.20x10-1

Cs-135 5.40 1.30x10-4 5.40 

Cs-137 4.30 3.00x105 2.30x101 3.00x105
23 

Vault 1 

(Ci) 

Vault 4 

(Ci) 

Individual FDC

(64 FDCs total)

(Ci) 

Total SDF 

(Ci) 

Eu-152 1.80x10-3 9.70x10-2 9.80x10-2 6.40 

Eu-154 2.30x10-4 1.20x101 1.80 1.30x102

Eu-155 6.80x10-1 1.30x10-1 9.00 

H-3 6.10 2.60x102 3.00x101 2.20x103

I-129 1.10x10-1 2.80x10-1 3.80x10-1 2.50x101

K-40 7.60x10-4 3.00x10-3 4.20x10-4 3.10x10-2

Na-22 1.50x10-1 6.90x10-2 4.60 

Nb-93m 2.50x10-1 8.40 3.70x10-1 3.20x101

Nb-94 2.50x10-3 8.70x10-2 3.80x10-3 3.30x10-1

Ni-59 3.50x10-2 4.00x10-1 8.40x10-2 5.80 

Ni-63 7.80x10-1 2.20x101 2.40 1.80x102

Np-237 4.50x10-3 6.10x10-1 5.00x10-2 3.80 

Pa-231 9.30x10-5 9.80x10-7 1.60x10-4

Pd-107 1.90x10-3 5.00x10-2 5.60x10-3 4.10x10-1

Pm-147 4.10x10-1 7.70x10-2 5.30 

Pr-144 1.80x10-9 3.60x10-10 2.50x10-8

Pt-193 3.70x10-1 1.00x101 1.10 8.10x101

Pu-238 7.80x10-3 9.10x103 1.70x102 2.00x104

Pu-239 1.20x10-2 3.80x102 1.50x101 1.30x103

Pu-240 1.20x10-2 1.20x102 4.10 3.80x102

Pu-241 9.80x10-3 2.40x103 4.20x101 5.10x103

Pu-242 9.00x10-4 8.10x10-1 3.90x10-3 1.10 

Pu-244 1.60x10-2 1.60x10-5 1.70x10-2

Ra-226 6.40x10-7 4.10 7.80x10-7 4.10 

Ra-228 1.60x10-6 8.70x10-5 5.60x10-3

Rh-106 1.50x10-10 9.10x10-7 1.20x10-6 7.80x10-5

Ru-106 1.50x10-10 9.10x10-7 1.20x10-6 7.80x10-5

Sb-125 1.60x10-1 5.70 2.40x10-1 2.10x101
24 

Vault 1 

(Ci) 

Vault 4 

(Ci) 

Individual FDC

(64 FDCs total)

(Ci) 

Total SDF 

(Ci) 

Sb-126 1.40x10-1 9.00x10-1 1.20 7.80x101

Sb-126m 1.00 6.40 8.20 5.30x102

Se-79 3.00x10-1 4.60x101 1.40 1.40x102

Sm-151 4.20x101 5.90x101 3.80x103

Sn-126 1.00 6.40 8.20 5.30x102

Sr-90 6.90x10-3 2.40x105 3.70x101 2.40x105

Tc-99 1.10x102 5.80x102 5.40x102 3.50x104

Te-125m 3.80x10-2 1.40 5.80x10-2 5.20 

Th-229 3.00x10-1 2.50x101 3.90x10-2 2.80x101

Th-230 4.10x10-1 7.50 1.90x10-1 2.00x101

Th-232 3.20x10-4 1.40x10-3 9.00x10-2

U-232 4.40x10-2 3.10x10-4 6.40x10-2

U-233 2.80x10-1 2.40x101 3.70x10-2 2.70x101

U-234 2.80x10-1 2.60x101 1.30x10-1 3.50x101

U-235 3.20x10-3 4.70x10-1 3.00x10-3 6.70x10-1

U-236 3.20x10-3 7.70x10-1 1.60x10-2 1.80 

U-238 7.40x10-3 5.90x10-1 1.00x10-1 7.00 

Y-90 6.90x10-3 2.40x105 3.70x101 2.40x105

Zr-93 2.50x10-1 8.40 3.70x10-1 3.20x101

Total 1.30x102 1.10x106 1.00x103 1.10x106

1.0 MBq is 3.7x104 Ci

2009 PA; Tables 3.3-1, 3.3-3, 3.3-5, and 3.3-7

The inventory values listed in the above table were used in the PA calculations (2009 PA), with 

the exception of Cases K, K1, and K2, which used revised values for the inventory of Pu-238, 

U-234, Th-230, and Ra-226 (Table 2.2-2). Revised inventory values were used for Pu-238, 

Ra-226, Th-230, and U-234 for Vault 4 and for Ra-226 and Th-230 for the FDCs. The inventory 

of Pu-238 and U-234 in Cases K, K1, and K2 were revised for Vault 4 based on sample 

analyses for the waste already disposed in Vault 4 (SRR-CWDA-2011-00115). DOE stated that 

the inventories of Th-230 and Ra-226 were also revised in these cases to remove conservative 

assumptions made in the original estimation. 
25 

Table 2.2-2: Revised Inventory of Ra-226 and its ancestors in Cases K, K1, and K2 

Vault 4 Individual FDC 

Original 

Inventory 

(Ci) 

Case K 

Inventory 

(Ci) 

Original 

Inventory 

(Ci) 

Case K 

Inventory 

(Ci) 

Ra-226 4.10 1.0x10-3 7.80x10-7 1.3x10-5

Th-230 7.50 1.0x10-2 1.90x10-1 1.3x10-4

U-234 2.60x101 1.0x101 1.30x10-1 1.30x10-1

Pu-238 9.10x103 1.0x103 1.70x102 1.70x102

1.0 MBq is 3.7x104 Ci 

SRR-CWDA-2011-00044; Response to PA-8. 

In the estimation of the original inventory (SRNS-J2100-2008-00004), the inventory of Th-230 

and Ra-226 was determined based on an assumption of transient equilibrium with U-234. 

However, DOE noted in the response to comment IN-5 (SRR-CWDA-2011-00044) that this is a 

significant conservatism based on the relatively long half-lives and the relatively young age of 

the waste. In SRR-CWDA-2011-00115, the inventory of Ra-226 and Th-230 was reevaluated 

using two different approaches. Because the sample results for Ra-226 and Th-230 were below 

the analytical detection limit in all samples, a predicted inventory of these radionuclides was 

generated based on the assumption that the radionuclides were present at the level of the 

detection limit. Additionally, an alternate inventory of these radionuclides was generated based 

on the in-growth of these radionuclides from U-234. The results of these estimates are 

presented in Table 2.2-3. 

Table 2.2-3: Estimates of Ra-226 and Th-230 inventory in Vault 4 

Ra-226 (Ci) Th-230 (Ci) 

Estimated inventory based on analytical 

detection limits 

(SRR-CWDA-2011-00115; page 5) 

6.6 20 

Estimated inventory based on U-234 

inventory and in-growth rates 

(SRR-CWDA-2011-00115; page 5) 

7.6x10-4 (1) 7.6x10-3

Reported Inventory in Vault 4 as of 9/30/10 

based on an alternative determination 

(X-CLC-Z-00034) 

9.29 2.82x10-2

1.0 MBq is 3.7x104 Ci

(1) Note that Table 5 of SRR-CWDA-2011-00115 also reports an inventory of 0.7 

MBq (1.9x10-5 Ci) for Ra-226 based on a similar approach

Information on the inventory of Ra-226 and Th-230 disposed to date in Vault 4 has previously 

been reported by DOE in X-CLC-Z-00034 and X-CLC-Z-00027. The reported inventory of these 

radionuclides in these documents was generated using an alternative calculation based on the 
26 

original waste tank radionuclide concentrations, the volumes of interstitial liquid removed and 

salt dissolved, and the volumes of the batches transferred to SDF. The inventory of Ra-226 and 

Th-230 generated using this method is also given in Table 2-4. 

As can be seen from the information in Table 2-4, the inventory generated based on in-growth 

from U-234 is significantly smaller than the inventories based on the detection limit for both 

radionuclides. The reported inventory based on the special calculation was much lower than 

the one based on the detection limits for Th-230, but was higher for Ra-226. As discussed in 

the response to comment IN-1 (SRR-CWDA-2010-00033), DOE stated that the reported Ra-226 

inventory to date was calculated from original waste tank inventories that were based on 

analytical detection limits. 

In Section 5.2.2 of the 2009 PA, DOE identified key radionuclides, or the radionuclides which 

contribute the most significantly to dose, based on the peak all-pathways dose calculated at 

100 m using the Base Case (Case A) assumptions over a period of 20,000 years. DOE 

included any radionuclides with a calculated peak dose of more than 5x10-4 mSv/yr 

(0.05 mrem/yr) in the list of key radionuclides. The resulting radionuclides that DOE identified 

as key radionuclides were Tc-99, I-129, Ra-226, Np-237, and Pa-231. The radionuclides that 

contributed most to the calculated dose were I-129, Tc-99, and Ra-226 and its progeny for Case 

A as well as for the sensitivity cases other than Case K. In Case K, Ra-226 contributed less to 

the calculated dose due to the revised inventory for Ra-226 and its ancestors, particularly 

Th-230. 

The Case A and Cases K, K1, and K2 PORFLOWTM calculations included all of the 

radionuclides listed in Table 2-1. However, the other PORFLOWTM cases only included the key 

radionuclides (i.e., Tc-99, I-129, Ra-226, Np-237, and Pa-231) and the parents and progeny of 

these radionuclides (i.e., Ac-227, Pb-210, Pu-238, Th-229, Th-230, U-233, U-234, and U-235). 

In response to NRC RAIs, DOE reran the PORFLOWTM calculations for Case C and the 

Synergistic Case with 13 additional radionuclides to evaluate the potential dose from 

radionuclides not included in the initial calculations (SRR-CWDA-2011-00044). As described in 

the response to comment PA-4, the results for the Case C analysis with additional radionuclides 

were similar for Sector I for the 20,000 year time of analysis. The results for Sector B were 

similar until approximately 16,000 years after which the analysis that included additional 

radionuclides was higher. Similarly, the graphs presented in the response to comment PA-9 

(SRR-CWDA-2011-00044) indicated that the doses from the Synergistic Case and the PA-9 

case that included additional radionuclides were similar. However, the dose values presented 

for the Synergistic Case in this RAI response were higher than those presented elsewhere. A 

comparison of the PA-9 case to the Synergistic Case results presented in the original PA 

(DOE-WD-2005-001) shows a larger difference in dose between these two cases. The 

GoldSim® calculations included nearly all of the radionuclides listed in the above table, with the 

exception of radionuclides with short half-lives that did not require transport modeling 

(SRR-CWDA-2010-00033). 
27 

2.2.2 NRC Evaluation – Source Term and Inventory 

Because the dose predicted by the 2009 PA is directly related to the assumed inventory, the 

inventory disposed of at SDF was identified as a key monitoring factor in NRC’s previous review 

(NRC, 2005), and the NRC staff has been monitoring this inventory since that time. NRC staff 

will continue to monitor the inventory disposed of at the SDF by tracking the actual inventory 

disposed against the inventory in Table 2-2 (for Ra-226, Th-230, U-234, and Pu-238) and 

Table 2-1 (for all remaining radionuclides) as part of monitoring. The dose is a function of both 

the total dose, as well as the spatial distribution of the inventory among the different disposal 

units, so NRC staff will monitor both the total inventory disposed in the SDF as well as the 

inventory disposed in Vault 1, Vault 4, and in each individual FDC. If the total inventory or the 

inventory in an individual vault or FDC is higher than the assumed values, an analysis will need 

to be performed to understand the dose consequences of the increased inventory. Also, as part 

of monitoring, NRC staff will continue to review the methodology used to measure and 

determine the inventory to confirm that risk significant radionuclides are being assessed 

adequately. 

NRC staff notes that there is significant uncertainty in the inventory of Ra-226 and Th-230 due 

to the fact that these radionuclides are present in the salt waste at levels that are below the 

detection limit of the analytical method used by DOE. As described in more detail in 

Section 2.13, the dose due to these radionuclides is potentially significant if they are present at 

levels that are as high as the detection limit. Because of this, it is important for the inventory of 

these radionuclides to be well understood. NRC staff finds that the alternate method of 

evaluating the Th-230 and Ra-226 inventory used in the Case K evaluation (i.e., determining the 

Th-230 and Ra-226 inventory from the in-growth from U-234) is appropriate for waste from 

tanks that have not had thorium inputs. However, this method would result in an 

underprediction of the Th-230 and Ra-226 inventory for waste that had sources of thorium other 

than the in-growth from U-234. In such waste, the inventory of Th-230 and Ra-226 should be 

based on the detection limits if the measured concentrations are below the detection limits or on 

the measured concentrations if the concentrations are above the detection limits. NRC staff will 

continue to monitor the methodology used to determine the Th-230 and Ra-226 inventory as 

part of the monitoring process, though as noted in Section 2.13, the inventory of these 

radionuclides may be less risk significant if assumptions used in the Case K modeling for the 

sorption coefficients prove to be true. 

The NRC staff finds that the approach used by DOE to determine the key radionuclides using a 

case in which little degradation is assumed to occur, such as Case A, is potentially problematic 

because different radionuclides may be risk-significant in cases in which more degradation 

occurs. This approach is especially potentially problematic when the case used is not 

consistent with, and contains less degradation than, the known and future expected conditions. 

Additionally, the NRC staff was also concerned that the exclusion of the non-key radionuclides 

from many of the cases other than Case A could have led to an underestimation of the 

estimated dose. However, in response to NRC staff’s concerns, DOE included all of the 

radionuclides in the Case K analysis. DOE also performed an evaluation of the effect of adding 

additional radionuclides to the Case C and Synergistic Case analyses. These evaluations 
28 

indicated that the exclusion of the non-key radionuclides from many of the deterministic 

analyses likely did not have a large effect on the calculated dose. It therefore does not appear 

that DOE’s approach for determining the key radionuclides led to the exclusion of risk significant 

radionuclides in the 2009 PA. 

2.3 Scenario Selection and Receptor Group 

2.3.1 Period of Performance and Institutional Controls 

The period of performance assumed by DOE in the 2009 PA for the purpose of demonstrating 

compliance with the performance objectives was a period of 10,000 years following the closure 

of the site. For the purpose of the PA, DOE assumed that institutional controls that would last 

for a period of 100 years following closure of the site, which is assumed to occur in 2030. 

During the time of institutional controls, it is assumed that the site is owned and controlled by 

the federal government and access to the site is limited. Maintenance of the closure cap is also 

assumed to occur during the institutional control period. The assumption of a 100-year 

institutional control period was based on regulation (§61.59), not on DOE’s future plans for the 

site. In reality, the site may be owned and controlled by the federal government for a longer 

period of time than 100 years. 

DOE ran all transport models to at least 20,000 years for the purposes of determining peak 

concentrations that occur after the 10,000-year performance period. In addition, a deterministic 

PORFLOWTM analysis was performed for Case A for 40,000 years. This calculation only 

included the radionuclides identified as key dose contributing radionuclides 

(SRR-CWDA-2011-00044). A probabilistic analysis to 450,000 years was also run using the 

GoldSim® model. In the response to NRC RAIs, DOE noted that the long-term analyses are 

meant only to help identify the model’s sensitivity to certain parameters and their uncertainty, 

and are provided for trend spotting only (SRR-CWDA-2010-00033). 

2.3.2 Scenario Identification 

The scenario considered by DOE in evaluating the dose to the member of the public was the 

resident farmer scenario. In this scenario, it is assumed that a receptor drills a well into a 

groundwater aquifer at a location 100 m from the disposal unit and uses this water as a potable 

water source as well as for agriculture (i.e., irrigation and water for livestock) following the end 

of active institutional controls. The resident farmer is also assumed to use nearby streams for 

recreational activities, such as swimming and fishing. 

The point of compliance for the dose to the member of the public is the point of highest 

projected dose beyond a 100 m buffer zone surrounding the disposal facility. In the 2009 PA, 

the 100 m doses were determined in 12 sectors that are downgradient of the facility. A diagram 

of the SDF, the points of compliance for the member of the public and the intruder, and the 

sectors analyzed can be seen in Figure 2.3-1. 
29 

Figure 2.3-1: Locations of the Compliance Points and the Sectors Analyzed 

(2009 PA; Figure 4.4-70) 

2.3.3 Identification of Relevant Features and Processes 

The primary mechanism of transport evaluated in the assessment of the dose to the member of 

the public is leaching from the saltstone into the groundwater and subsequent human use. The 

gas phase diffusion of radionuclides from the waste to the surface was also evaluated. 

Contaminant disturbance caused by bio-intrusion or erosion was not considered in the PA 

because DOE did not believe these were credible mechanisms based on the depth waste (i.e., 

> 3.7 m (12 ft) below the erosion control barrier). 

The transport of the radionuclides from the SDF to a well at 100 m from the disposal facility and 

to outcrops at two streams, the McQueen Branch to the east and the Upper Three Runs to the 

north was modeled as flow through porous media. The groundwater pathway dose for each 

sector was based on peak groundwater concentration in that sector. Additionally, the 

recreational dose for the receptor was based on the seepline location with the maximum 

concentration for each radionuclide. PORFLOWTM modeling of the transport of radionuclides to 

the stream seeplines was only performed for the radionuclides identified by DOE as being key 

radionuclides (i.e., Tc-99, I-129, Ra-226, Np-237, Pa-231). For the other radionuclides, the 
30 

seepline concentrations were assumed to be equal to 35% of the 100 m concentration. This 

assumption was based on 35% being the bounding ratio of the seepline concentration to the 

100 m concentration for any radionuclide. 

The modeled release and transport of the radionuclides from the waste form is highly dependent 

on assumptions made regarding the release from the source term, the assumed degradation of 

the waste form and disposal units, the assumed degradation of the engineered cover, and the 

transport in the subsurface. More detail is provided about DOE’s identification of the relevant 

features and processes for these areas in other sections of this TER. A description of the 

features and processes included in the modeling of the release of radionuclides from the waste 

form is in Section 2.7, descriptions of the features and processes involved in the degradation of 

the waste form and disposal units are located in Section 2.6 and 2.5 respectively, a description 

of the modeled long term performance of the engineered cover is in Section 2.4, and a 

description of the transport in the subsurface is discussed in Section 2.8. 

2.3.4 Receptor Characteristics 

The dose assessment for the member of the public evaluated well water, surface water, and air 

pathways. The exposure pathways related to the use of potentially contaminated well water 

included: 

• direct ingestion of well water 

• ingestion of vegetables grown in soil that is irrigated with well water 

• ingestion of meat and milk from livestock that drink well water and eat fodder grown in 

soil irrigated with well water 

• ingestion and inhalation of well water while showering 

• dermal contact with well water while showering 

• inhalation of well water during irrigation 

• inhalation of dust from soil irrigated with well water 

• ingestion of soil irrigated with well water 

• direct radiation from soil irrigated with well water 

The exposure pathways considered for the use of surface water included: 

• direct radiation during recreational activities in the stream 

• incidental inhalation of stream water during recreational activities 

• incidental ingestion of stream water during recreational activities 

• dermal contact with stream water during recreational activities 

• ingestion of fish from the stream 

The accidental ingestion of water while swimming and showering was considered as part of the 

total well water ingested and was not considered separately. The dose due to the dermal 

absorption of radionuclides was not considered quantitatively because the radionuclides are 

generally absorbed poorly and DOE therefore considered this dose to be insignificant. 
31 

The ingestion of shellfish was not included in the dose assessment because the Upper Three 

Runs and McQueen branch are not significant sources of edible shellfish. The ingestion of wild 

game was also not included because livestock raised on the site would be more affected by the 

disposed radionuclides than game, which are more transient. Additionally, the ingestion of 

game by the receptor would offset the amount of ingestion of livestock, which could result in a 

lower total dose. The ingestion of poultry and eggs was not included in the initial PA by DOE 

(DOE-WD-2005-001). However, an evaluation of this dose was performed by DOE in response 

to comment B-2 (SRR-CWDA-2011-00044), and the dose from this pathway was not found to 

be significant based on the parameter values selected in this analysis (Section 2.10). 

The potential dose due to the diffusion of gaseous radionuclides through the waste form and 

overlaying materials was also evaluated (Section 2.9). 

2.3.5 NRC Evaluation – Scenario Selection and Receptor Groups 

In the PA, DOE assumed a period of performance of 10,000 years, though all analyses were run 

to 20,000 years, which was generally sufficient to capture the peak dose. NUREG-1854 

(NRC, 2007b) states “Generally, a period of 10,000 years after closure is sufficient to capture 

the peak dose from the more mobile, long-lived radionuclides and to demonstrate the influence 

of the natural and engineered systems in achieving the performance objectives (NRC, 2000). 

However, assessments beyond 10,000 years may be necessary to ensure (1) that the disposal 

of certain types of waste does not result in markedly high impacts to future generations or 

(2) evaluate waste disposal at arid sites with extremely long groundwater travel times.” DOE’s 

approach is consistent with this guidance. DOE’s assumption of a 100-year institutional control 

period is consistent with NRC guidance in NUREG-1854 and the regulations in §61.59(b). 

The NRC staff finds that DOE’s use of the resident farmer scenario and the pathways selected 

by DOE for this scenario are reasonable and appropriate based on regional practices near SRS. 

Additionally, NRC staff finds that the use of a 100 m point of compliance for the member of the 

public is consistent with the typical assumption of a 100 m buffer zone described in 

NUREG-1854 (NRC, 2007b). NRC’s evaluation of the features and processes included in the 

modeled release and transport (i.e., the release from the source term, the assumed degradation 

of the waste form and disposal units, the assumed degradation of the engineered cover, and the 

transport in the subsurface) are discussed in the portions of this TER that specifically address 

those areas. 

2.4 Infiltration and Erosion Control 

This section addresses the engineered soil covers (closure caps) DOE plans to use to limit 

infiltration and erosion at the SDF. In addition, the erosion control layer of the closure cap is 

expected to deter potential inadvertent intruders. The current closure cap design is preliminary 

and will be finalized closer to the time of SDF closure. 
32 

2.4.1 Local Meteorology and Climatology 

In the PA, DOE states that the climate at SRS is humid and subtropical. On average, 

approximately half of summer days have maximum temperatures exceeding 32°C (90°F). Less 

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

with temperatures below -7°C (20°F) are infrequent. Mean annual precipitation, consisting 

primarily of rainfall, is reported to be approximately 121 cm/yr [48 in/yr], with a range of 88 to 

183 cm/yr [35 to 72 in/yr] (DP-MS-87-126, as reported in WSRC-STI-2008-00244). 

2.4.2 Infiltration and Erosion Control 

Closure of the SDF will include the construction of two engineered closure caps over Vault 1, 

Vault 4, and the FDCs, as illustrated in Figure 2.4-1 (2009 PA; WSRC-STI-2008-00244). The 

closure caps are designed to (1) provide physical stabilization, (2) limit infiltration by promoting 

runoff, evapotranspiration, and the shedding of water around the vaults and disposal cells and, 

(3) act as an intruder deterrent. The closure cap design (WSRC-STI-2008-00244) consists, 

from the surface downward, of topsoil and backfill, an erosion barrier of coarse stone, a middle 

backfill layer, an upper drainage layer over a high-density polyethylene (HDPE)/Geosynthetic 

clay liner (GCL) composite layer, and additional backfill (Figure 2.4-2). In addition, several 

geotextile filter fabric layers in the cover, while not directly influencing flow, serve to separate 

some of the engineered layers (e.g., filter fabric above the drainage layers helps to slow 

drainage layer clogging with soil particles from layers above). In addition to the site closure 

caps, a lower lateral drainage layer will be placed directly over each disposal unit. For the 

FDCs, the lower lateral drainage layer will be underlain by a second HDPE/GCL composite 

layer (Figure 2.4-2). The maximum slope of the upper surface of the closure caps is 1.5%. The 

side slopes, with maximum slopes of 10.5%, will be covered with large stones for erosion 

control. 

After installation of the closure caps an initial 100 year institutional control period will begin, 

during which active maintenance will be conducted to prevent pine forest succession and to 

repair any significant erosion. In the PA, DOE does not assume any active maintenance will 

continue after the 100 year period post site closure. Although the closure caps are assumed to 

be initially vegetated with a persistent grass, such as Bahia (Paspalum notatum), the vegetation 

is assumed to evolve into a pine forest. The grass may later be replaced with bamboo, if DOE 

determines that bamboo will slow invasion of loblolly pine trees. DOE plans to slow the invasion 

of loblolly pines, if possible, because loblolly pines are expected to result in the degradation of 

the upper HDPE/GCL composite layer due to root depths up to 3.7 m (12 ft). DOE does not 

expect Loblolly pines to disrupt deeper layers (i.e., the lower hydraulic barrier layers) because 

the roots typically do not grow to that depth. 
33 

Figure 2.4-1: Conceptual illustration of the SDF Closure Cap 

(WSRC-STI-2008-00244; Figure 5) 

Figure 2.4-2: Conceptual illustrations of the components of the SDF closure caps and 

additional engineered layers above the disposal units (2009 PA; Figure 3.2-20) 
34 

2.4.1.1 Infiltration 

The potential for advective transport of radionuclides from saltstone is reduced by limiting 

infiltration. DOE used the HELP model (EPA-600-R-94-168a, EPA-600-R-94-168b) to estimate 

infiltration rates through the closure cap considering processes that degraded or altered the 

properties of the materials used to construct the closure cap, such as root penetration of the 

HDPE geomembrane and clogging of the drainage layer. Simulations of infiltration through the 

closure cap were based on a 100-year synthetic weather database for Augusta, Georgia that 

was modified with SRS-specific precipitation data. The annual precipitation in this data set 

ranged from 76 to 175 cm/yr [30 to 69 in/yr]. In both Case A and Case K, infiltration rates 

through the closure cap (i.e., from the surface through the foundation layer in Figure 2.4-2) 

reached steady-state conditions of 26.9 cm/yr [10.6 in/yr] at approximately 5,500 years after site 

closure. This steady-state value is approximately the same as the natural infiltration rate in the 

absence of a closure cap (WSRC-STI-2008-00244). Estimated average annual net infiltration 

through the closure cap was then used as an upper boundary condition to the SDF vadose zone 

model, which included the lower backfill layer, lower lateral drainage layer, HDPE and 

Geosynthetic Clay Liner (for the FDCs only), the disposal unit roofs, walls, and floors, and the 

saltstone waste form, as well as the underlying and surrounding backfill soil. 

The SDF “no closure cap” sensitivity analysis case indicates that the closure cap (i.e., layers 

from the surface to the Foundation Layer in Figure 2.4-2) is expected to delay the release of 

radionuclides from the SDF and reduce peak radionuclide doses within 10,000 years by 

approximately 25% as compared to Case A (2009 PA; Tables 2-13 and 2-14 and Figure 5.6-75). 

Although this case is called the “no closure cap” case, it does include the effects of the lower 

lateral drainage layer. 

2.4.1.2 Erosion Control 

The ability of the closure cap to reduce infiltration and deter intrusion for long time periods is 

dependent on erosion controls. The closure cap design consists of a minimum of 3 m [10 ft] of 

material above the backfill covering the vaults and disposal cells. Typically, agricultural and 

resident intruder scenarios include a nominal excavation depth of 3 m [10 ft]. Therefore, proper 

design, construction, and performance of the erosion barrier should limit surface water erosion 

and direct contact of the waste by potential inadvertent intruders. In addition, DOE assumes 

that the erosion barrier will prevent animal intrusion into the lower layers. DOE does not 

assume the erosion layer will prevent root penetration from pine trees. 

The erosion barrier is designed to limit erosion of the underlying cap layers, however the 

vegetative cover, topsoil, and upper backfill layer, which provide water storage and promote 

evapotranspiration, are susceptible to erosion. DOE performed scoping-level calculations to 

inform its design of the cap layers to prevent gully formation and ensure that soil loss would not 

impact closure cap performance. 

DOE evaluated the physical stability of the closure cap with respect to a probable maximum 

precipitation (PMP) event (WSRC-STI-2008-00244) as is consistent with NUREG-1623 
35 

(NRC, 2002b). The PMP is defined as the theoretically greatest depth of precipitation that is 

physically possible during a given period of time over a given area at a particular geographic 

location. Based on the PMP, the design criteria for the vegetative cover, erosion barrier, side 

slopes, and toe of the side slopes were evaluated. Although the methodology presented in 

NUREG-1623 addresses a 1,000 year timeframe, DOE stated in an RAI response for the FTF 

review that the SRS-specific PMP event provides assurance of closure cap stability against 

gullying for the 10,000 year compliance period (SRR-CWDA-2009-00054). In response to NRC 

staff's concern with the long-term performance of the side slopes (NRC, 2010b; IEC-3), DOE 

indicated that slumping of the side slope and down-slope creep of the riprap were not evaluated 

explicitly. DOE also indicated that the conservatisms in its approach to designing the side 

slopes are expected to account for any potentially adverse effects of additional degradation 

mechanisms (SRR-CWDA-2010-00033). In addition, DOE indicated that slope stability will be 

considered as part of the closure cap final design. 

Although the slopes and slope length for the topsoil and upper backfill layers of the closure caps 

have been designed to prevent gully formation due to a PMP event, these layers are still 

assumed to be subject to erosion (WSRC-STI-2008-00244). The projected long-term topsoil 

loss was determined according to the Universal Soil Loss Equation for both vegetative cover 

conditions (i.e., bahia grass and pine forest). DOE predicted approximately 3.3 cm [1.3 in] of 

soil loss for the topsoil and no reduction in the upper backfill layer within 10,000 years 

(WSRC-STI-2008-00244; Section 7.2). 

Riprap for the integrated drainage system ditches has not yet been sized due to the early phase 

of the project and resultant lack of a detailed closure cap drainage system layout. Riprap 

material for the erosion barrier, side slope, and toe of the side slope will be selected from local 

granite or mylonitic quartzite quarries. 

2.4.3 NRC-Evaluation – Infiltration and Erosion Control 

The designs for long-term infiltration and erosion control have different objectives and are 

subject to different degradation mechanisms. Acceptability of a design for one does not ensure 

that an acceptable design has been achieved for the other (e.g., designing the vegetative cover 

and topsoil to promote runoff may reduce infiltration in the near term, but may increase longterm erosion). 

2.4.3.1 NRC-Evaluation – Infiltration 

Dose typically is sensitive to infiltration because infiltration is directly related to the flux of 

contaminants into the groundwater. However, the assumed performance of a series of 

additional engineered barriers limits the sensitivity of the overall performance of the SDF to 

closure cap performance (Section 2.13.3). In particular, due to the hydraulic performance of 

(1) the lower lateral drainage layer and disposal unit roofs; (2) saltstone grout; and (3) the 

composite HDPE/GCL layer above the FDCs, approximately 99.9% of the infiltrating water is 

modeled as being shed around the disposal units from 1,000 to 10,000 years 

(SRR-CWDA-2011-00044). 
36 

DOE’s “no closure cap” sensitivity analysis is based on the assumption that the upper 

HDPE/GCL composite layer, the upper drainage layer, and the erosion control layer have the 

hydraulic properties of backfill soil at the time of site closure (2009 PA; Section 5.6.6.2). As a 

result, the infiltration rate is fixed at 41.8 cm/yr [16.5 in/yr] as a boundary condition into the 

PORFLOWTM vadose zone model (i.e., flow to layers below the foundation layer in 

Figure 2.4-2). However, the infiltrating water is still effectively shed around the disposal units 

(e.g., greater than 99.8% of the infiltrating water is predicted to be shed around the disposal 

units 8,000 years after closure). Consequently, the peak dose shifted only slightly earlier in time 

and the magnitude increased by less than 50%. If the assumptions regarding the hydraulic 

performance of the lower lateral drainage layer, disposal unit roofs, HDPE/GCL composite layer 

(for the FDCs only), and the saltstone grout are found to be optimistic during monitoring, the 

closure cap would become a more risk significant barrier. 

The upper and lower lateral drainage layers are designed to divert a significant portion of the 

infiltrating water away from the underlying disposal units. A geotextile filter fabric will be placed 

on top of the drainage layers to provide filtration between the underlying sand and the overlying 

backfill layers. DOE assumed that the degradation of the drainage layers (i.e., a reduction in 

hydraulic conductivity) will be controlled by colloidal infilling of the pore spaces within the 

drainage layers from the overlying backfill. As there is limited data regarding the service life of 

filter fabrics, NRC requested additional information about potential infilling of the drainage layers 

with larger particles the resulting potential decrease in drainage layer hydraulic conductivity 

(NRC, 2010i; IEC-8). The NRC staff was concerned that decreased hydraulic conductivity could 

limit the ability of the lateral drainage layer to shed water and lead to more water reaching 

saltstone than is assumed in the PA. In response, DOE provided a flow budget showing that 

the lower lateral drainage layer significantly limits infiltrating water (e.g., approximately 99,9% of 

the water is modeled as being shed around Vaults 1, 4, and the FDCs at 10,000 years 

(SRR-CWDA-2011-00044; IEC-8). DOE also conducted a sensitivity analysis to assess the risk 

significance of filter fabric by doubling the modeled infilling of the drainage layers 

(SRR-CWDA-2011-00044; PA-10). DOE's analysis demonstrated that the shedding of the 

water around the disposal units is much more sensitive to assumptions about the disposal unit 

roofs and the HDPE/GCL composite layer above the FDCs than it is to assumptions about 

drainage layer infilling. Specifically, for Vault 4 and for an FDC, doubling the infilling of the 

drainage layers increased the Darcy velocity through saltstone by approximately a factor of 3 or 

less at 10,000 years in Case A, and caused smaller increases in Case K. In comparison, the 

Darcy velocity through saltstone at 10,000 years after closure is between a factor of 100 and 

1,000 times greater in Case K than it is in Case A. The NRC staff agrees with DOE’s 

assessment that the degradation of disposal unit concrete and saltstone is more risk significant 

than the degradation of the drainage layers. However, a factor of two to three in difference in 

the Darcy velocity is more significant to a compliance assessment if predicted doses approach 

the relevant dose limit. 

As discussed in more detail in Chapter 5, the NRC staff determined the potential for differential 

settlement due to static settlement of the disposal units or the formation of sinks under the SDF 

may not have been fully evaluated. Differential settlement of disposal units could affect the 

performance of the closure cap and lower lateral drainage layer. Disruption of the upper 
37 

HDPE/GCL layer is not expected to affect system performance in ways not already addressed 

by increasing predicted infiltration to near natural infiltration rates. However, because the lower 

lateral drainage layer is modeled as performing well for the entire performance period, 

disruption of the disposal unit roofs or HDPE/GCL layers on the FDC roofs could cause the 

system behavior to deviate significantly from model predictions. 

The assumed saturated hydraulic conductivity for the foundation layer of 1.0x10-6 cm/s 

(3.3x10-8 ft/s) constrains the HELP model to a maximum infiltration rate of 31.6 cm/yr (12.4 in/yr) 

through the engineered cover (i.e., from the surface to the foundation layer). Consequently, the 

long-term steady-state infiltration rate, which ranges from 13.7 to 31.6 cm/yr (5.4 to 12.4 in/yr) is 

less than the background value of 37.7 cm/yr (14.9 in/yr). DOE indicated that the assumed 

saturated hydraulic conductivity for the foundation layer is appropriate, as the value is a design 

specification that is achievable for soil-bentonite blends (SRR-CWDA-2010-00033). An 

increase from DOE’s assumed long-term infiltration rate to the background value would result in 

an increase in radionuclide release and a decrease in the timing of the chemical transitions in 

the saltstone waste form. 

DOE modeling indicates that saturated conditions will occur above the upper composite layer in 

the closure cap. An initial hydraulic head of 9.55 cm (3.76 in) is predicted to develop on top of 

the HDPE geomembrane and increase until 5,400 years after closure, when it is predicted to 

range from 99.0 to 100 cm [39.0 to 40.0 in] until 10,000 years or more after closure 

(SRR-CWDA-2011-00044). DOE stated that conservative modeling assumptions (e.g., depth of 

evapotranspiration zone and degradation of the lateral drainage layer) resulted in estimates of 

head on the HDPE geomembrane that are bounding and conservative. Should the buildup of 

hydraulic head occur, DOE does not believe it would adversely impact the physical stability of 

the closure cap, vegetation, erosion, or the performance of the composite layer. Based on 

limited model support, it is difficult to assess: (1) the likelihood of hydraulic head buildup within 

the cover; or (2) its potential implications for closure cap performance. A more realistic 

representation of infiltration and saturation within the proposed closure cap is needed to assess 

the potential for buildup of hydraulic head. If an analysis containing a more realistic 

representation determines that the buildup of hydraulic head is realistic, an explicit evaluation of 

the physical stability of cover materials under this condition will be needed. 

In addition, when the cap is constructed, it is important for there to be a robust quality 

assurance (QA)/quality control (QC) program. In particular, the ability of the composite 

hydraulic barrier to limit infiltration early in the compliance period is dependent on construction 

quality. 

DOE indicated that the closure cap has a minimal impact on peak doses when other barriers to 

flow perform as designed (i.e., the lower lateral drainage layers above each disposal unit and 

the intended low permeability of disposal unit roofs and saltstone waste form). However, the 

NRC staff expects closure cap performance to be more risk significant if these other barriers do 

not perform as designed (Section 2.13.3). The long-term performance of the infiltration barrier 

is expected to be affected by several sources of uncertainty, including (1) degradation of the 

drainage layers and underlying disposal unit roofs or HDPE/GCL composite layers (for the 
38 

FDCs), (2) potential effects of differential settlement, (3) hydraulic degradation of the foundation 

layer, and (4) effects of hydraulic head buildup within the closure cap. The NRC staff concludes 

that more model support is needed for the long-term infiltration-limiting performance of the 

closure cap and infiltration-limiting engineered layers above the disposal units. Because of the 

importance of limiting the infiltration to system performance, the NRC staff will track these 

issues as monitoring factors. 

2.4.3.2 NRC-Evaluation – Erosion Control 

Erosion control is necessary to ensure that a thick cover of soil is maintained over the waste for 

protection of inadvertent intruders and to provide suitable conditions for the vegetative cover. 

To mitigate the potential effects of erosion by surface water, erosion protection designs must be 

based on an appropriately conservative rainfall event. DOE’s determination of the PMP event 

and the corresponding design criteria for the vegetative layer, erosion barrier, side slopes, and 

toe of the side slopes were consistent with NUREG-1623 (NRC, 2002b). This guidance 

document specifically addresses a 1,000-year timeframe rather than the 10,000-year 

compliance period. DOE stated that the SRS-specific PMP has a low probability of occurrence 

and is a bounding event, thereby providing assurance of physical stability of the closure cap 

design for the 10,000-year compliance period (SRR-CWDA-2011-00054). NRC staff 

determined that DOE’s estimates for the PMP are reasonable and that the probability of such an 

event being equaled or exceeded is very low. Accordingly, the PMP is considered by NRC staff 

to provide a reasonable design basis. 

Long-term maintenance of the topsoil and vegetative cover is important to closure cap 

performance as the average evapotranspiration rate (82.7 to 85.4 cm/yr [32.6 to 33.6 in/yr]) 

dominates the modeled water balance distribution for SRS precipitation (125 cm [49.1 in]). An 

evaluation of the cumulative effects from precipitation events over long time periods with respect 

to gully formation is needed to support predictions of long-term performance of the topsoil and 

vegetative layers. 

The NRC staff concludes that the closure cap, as designed, can provide adequate long-term 

erosion protection. However, because the cap will not be built until SDF closure, it remains for 

DOE to demonstrate that certain aspects of the designed performance can be achieved (e.g., 

evaluation of an acceptable rock source, the ability of an integrated drainage system to 

accommodate design features). Although the design will not be made final until closer to the 

time of site closure, verification that certain designed features can be implemented as designed 

is needed in advance of site closure to allow sufficient time to change the closure cap design or 

the assumptions regarding long-term erosion protection, if necessary. These modifications may 

be important in the overall evaluation of closure cap performance. 
39 

2.5 Disposal Unit Design and Performance 

2.5.1 Design and Construction of the Disposal Units

2.5.1.1 Vault 1 

Vault 1 is a rectangular, reinforced concrete disposal unit approximately 180 m (600 ft) long, 

30 m (100 ft) wide, and 8.2 m (27 ft) high. It contains two units, each 30 m by 90 m (100 ft by 

300 ft), which are separated by an 8 cm (3 in) gap. Each unit has three cells, each 

approximately 30 m by 30 m (100 ft by 100 ft). In total, Vault 1 comprises six cells total (Cells A 

to F), each 30 m by 30 m (100 ft by 100 ft). Vault 1 has a poured-in-place concrete roof with a 

minimum thickness of 15 cm (6 in). The walls are 46 cm (18 in) thick and the floor slab is 61 cm 

(24 in) thick. The walls and floor are constructed of reinforced concrete, using slightly different 

concrete formulations (Table 2.5-1). Underlying the floor slab is a 10 cm (4 in) thick concrete 

working slab. Vault 1 was built above grade and remains above grade, but it will be buried and 

covered with a closure cap at the time of site closure. 

Table 2.5-1: Vault 1 Concrete Formulations 

Ingredient 

Quantity (lb/yd3) 1

Working 

Slab 

Floor and 

Walls 

Roof 

ASTM C 150 Type II Cement 413 419 400 

Grade 120 ASTM C 989 Blast Furnace Slag 0 278 0 

Type F ASTM C 618 Fly Ash 73 0 70 

ASTM C 33 Sand 1,356 1,133 1,149 

No. 67 ASTM C 33 Aggregate 

(maximum 1.9 cm [0.75 in]) 1,698 1,798 1,900 

Water (maximum) 272 268 292 

Water to Cementitious Material Ratio 0.56 0.385 0.62 

Minimum Compressive Strength at 28 days 2,000 psi 4,000 psi 3,000 psi 

2009 PA; Table 3.2-1 

1

 1.0 lb/yd3 = 0.59 kg/m3

Vault 1 Cells A, B, and C have been filled and are covered by a permanent roof. Vault 1 

Cells D, E, and F are unfilled and have no roof. Vault 1 Cell A was filled with saltstone in 1994. 

While a temporary concrete roof was being installed, water was trapped between the saltstone 

and disposal unit walls (ESH-WPG-2006-00132). Pressure from the trapped water caused the 

wall to crack, at which point DOE observed liquid weeping from Vault 1 Cell A onto the ground. 

Vault 1 Cells B and C were closed in 1998. While a concrete roof was being installed over 

these cells, water also was trapped between the saltstone and disposal unit walls. 
40 

2.5.1.2 Vault 4 

Vault 4 is a rectangular, reinforced concrete disposal unit, approximately 180 m (600 ft) long, 

60 m (200 ft) wide, and 9 m (30 ft) high. It is divided into two units, each 60 m by 90 m (200 ft 

by 300 ft), which are separated by an 8 cm (3 in) gap. Each unit has 6 cells, each 

approximately 30 m by 30 m (100 ft by 100 ft). Thus, in total, Vault 4 comprises twelve cells 

(Cells A to L), each 30 m by 30 m (100 ft by 100 ft). Vault 4 Cells B to L are being filled with 

saltstone, whereas Cell A has drums of low-activity waste encapsulated by clean grout. Like 

Vault 1, Vault 4 was built and remains above grade, but will be buried and covered with a 

closure cap at site closure. 

Table 2.5-2: Vault 4 Concrete Formulations 

Ingredient 

Quantity (lb/yd3) 1

Working Slab Floor Slab and 

Walls to 7.6 m (25 ft) 

Walls above 7.6 m 

(25 ft) and Roof 

ASTM C 150 Type II 

Cement 413 419 466 

Grade 120 ASTM C 989 

Blast Furnace Slag 0 278 0 

Type F ASTM C 618 

Fly Ash 73 0 62 

ASTM C 33 Sand 1,356 1,133 1,190 

No. 67, ASTM C 33 

Aggregate (maximum 

1.8 cm [0.75 in]) 

1,698 1,798 1,800 

Water (maximum) 273 254 296 

Water to Cementitious 

Material Ratio 0.56 0.36 0.56 

Minimum Compressive 

Strength at 28 days 2,000 psi 4,000 psi 4,000 psi 

2009 PA Table 3.2-2 

1

 1.0 lb/yd3 = 0.59 kg/m3

Vault 4 was constructed in 1988 and initially included a 10 cm (4 in) thick concrete working slab, 

0.6 m (2 ft) thick reinforced concrete floor slab, and 46 cm (18 in) thick reinforced concrete 

walls. Construction joints are located on 9 m (30 ft) centers in the floor slab and walls 

(T-CLC-Z-00006). The concrete formulations used for the Vault 4 floors and walls are listed in 

Table 2.5-2. In 1996, prior to filling the disposal unit cells with grout, a permanent roof (pouredin place concrete over steel decking) was installed to reduce potential rainwater intrusion. Nine 

equally spaced roof support columns (25 cm [10 in] diameter concrete-filled carbon steel pipes) 

also were installed in each of Vault 4 Cells B to L. The concrete roof was painted with a heatdissipating coating, but no waterproof coat was applied. The roof joints allowed rainwater to 
41 

enter the disposal unit cells, which DOE believes is responsible for wall cracking similar to that 

in Vault 1. Specifically, DOE concluded the cracking in the Vault 1 and 4 walls is due to 

hydrostatic pressure buildup of liquids in the annuli between the saltstone grout and the walls 

(SRR-CWDA-2010-00033, comment VP-1; C-CLC-Z-00016). Subsequently, the joints were 

sealed and sheet drain systems were installed in eight of the disposal unit cells to remediate the 

liquid accumulation (ESH-WPG-2006-00132). 

2.5.1.3 Future Disposal Cells 

As of 2009, DOE planned to construct 64 FDCs at the SDF (2009 PA). Each disposal unit is a 

separate 46 m (150 ft) diameter cylindrical tank primarily made of sulfate-resistant concrete. 

The walls are made of pre-cast panels of sulfate-resistant concrete, a steel shell diaphragm, 

reinforcing bars, pre-stressing wires, and shotcrete fill. The reinforcing steel, pre-stressing wire, 

and steel shell diaphragm are made of carbon steel. The shotcrete is intended to protect the 

reinforcing bars and pre-stressing wires from corrosion. 

Construction on FDCs 2A and 2B began in 2009 and the cells began testing for water tightness 

beginning 2010 (Section 2.5.3.1). Both FDCs are 46 m (150 ft) in diameter and have an interior 

height of 6.7 m (22 ft) (7.2 m [23.5 ft] at the center). The current design for the remaining 62 

FDCs is similar to the design of 2A and 2B; a few notable differences are mentioned in this 

section. 

The concrete used in the FDC walls and floor serves as a chemical as well as a physical barrier 

to radionuclide release. The concrete includes blast furnace slag to create chemically reducing 

conditions to reduce the mobility of Tc-99 through the concrete. DOE stated that two potential 

FDC concrete mixes have been examined and produced acceptable results, but that further 

concrete mix testing is expected to be conducted during the operational phase of the SDF 

(2009 PA). One of these formulations, which may be used for the FDC floors and walls, is 

provided (Table 2.5-3). The concrete attributes DOE considers important to the FDC are (i) low 

hydraulic conductivity, (ii) high pH, (iii) low Eh, (iv) high degradation resistance, and (v) high 

sulfate attack resistance. 

FDC construction begins with pouring a 10 cm (4 in) thick concrete base, called the lower mud 

mat, on site. The lower mud mat is covered with a geosynthetic clay liner (GCL), composed of a 

layer of bentonite between two geotextile layers (Section 2.5.2.2). The GCL is then covered 

with a 2.5-mm (100-mil) HDPE liner, onto which an upper mud mat, made of Class III sulfateresistant concrete, is poured. The 10 cm (4 in) thick upper mud mat is designed to protect the 

HDPE liner during construction. Reinforcing steel bars are then put in place and a 20 cm (8 in) 

thick floor, also composed of Class III sulfate-resistant concrete, is poured. 

Cell walls are constructed from pre-cast sulfate resistant concrete panels with a minimum 

thickness of 20 cm (8 in). The concrete panels are poured onto the steel diaphragm panels 

after reinforcing steel bars are put in place. DOE expects that rust that formed on the inside and 

outside (NRC, 2009b; photos 22, 23, 25, and 44 - 47) of the steel diaphragm panels during 

construction will not accelerate corrosion cracking because the steel will be passivated by the 
42 

high pH environment created by the concrete wall, on the inside, and a layer of shotcrete on the 

outside of the steel shell. 

Table 2.5-3: Potential FDC Concrete Formulation 

Ingredient Quantity (lb/yd3) 1

ASTM C 150 Type V Cement 213 

Grade 100 ASTM C 989 Blast Furnace Slag 284 

ASTM C 1240 Silica Fume 47.3 

Type F ASTM C 618 Fly Ash 165.7 

ASTM C 33 Sand 911 

No. 67 Granite ASTM C 33 Aggregate 1,850 

Water (maximum; gal/yd3) 32.3 

Grace WRDA 35 (oz.cwt c+p) 5 

Grace Darex II (oz/cwt c+p) 0.4 to 0.5 

Grace Adva 380 (oz/cwt c+p) 3 to 4 

Maximum Water to Cementitious Material Ratio 0.38 

2009 PA Table 3.2-3 

1

 1.0 lb/yd3 = 0.59 kg/m3

The wall panels are joined with steel closure strips. A polysulfide material is injected into the 

area where the closure strip meets the steel diaphragm panels to form a watertight seal. Joints 

between the panels are filled with sulfate resistant concrete. After the wall panels are erected 

and joined, the structure is covered with a 2.5 cm (1 in) minimum layer of concrete. 

Pretensioning wires are then wrapped around the structure and tensioned and another layer of 

shotcrete is applied to protect the pretensioning wires. The walls are then entirely covered with 

2.5 mm (100 mil) HDPE sheets, which are sealed to each other and to the HDPE layer under 

the upper mud mat. 

Inside an FDC, the floor and walls are coated with a mat-reinforced epoxy-novolac 

thermosetting lining. The liner is not expected to provide a long-term hydraulic barrier, but is 

designed to protect the floor and wall concrete from sulfate in saltstone bleed water during the 

initial saltstone curing and for several decades afterward. DOE has included sheet drains 

between the liner and saltstone waste form to remove saltstone bleed water and any infiltrating 

water from the disposal units. The sheet drain is fastened to the floor with anchor bolts. 

Removal of bleed water and infiltrating water from the FDCs is intended to protect the FDCs 

from the type of hydraulic loading that caused fracturing of Vault 4 walls. The sheet drain 

system is designed to function during the operational period. Drain pipes leading from the sheet 

drain system will be filled with grout at the time of site closure (NRC, 2010i; IN-4). 

The 20 cm (8 in) thick FDC roof is supported by 48 36-cm (14-in) diameter columns made of 

carbon steel-reinforced sulfate-resistant concrete. The roof includes penetrations for ventilation, 
43 

sheet drain, thermocouple trees, closed circuit television, and grout poring. During the 

operational period, these penetrations will be sealed with neoprene gaskets. The external 

surface of the roof will be covered with an HDPE/GCL composite layer once the FDC is filled 

(Section 2.5.2.2). 

After testing cells 2A and 2B for water tightness, DOE made a number of design modifications 

to the FDCs that occurred after the PA was issued (WB00001K-058). In the second response 

comment VP-3 (SRR-CWDA-2011-00044), DOE discussed modifications to cells 2A and 2B 

that included cutting the anchor bolt penetrations flush with the floor, applying an interior 

coating, installing an additional 36 cm (14 in) of sulfate resistant concrete above the coating 

layer, and installing an exterior sulfate resistant curb around the perimeter of the disposal unit 

floor. In addition, DOE discussed that significant design modifications to the disposal units have 

been implemented based on the lessons learned from the hydrostatic testing (e.g., future 

disposal units do not include anchor bolt penetrations in the floor). Additional discussion of the 

hydrostatic testing and design modifications is provided in Section 2.5.3.1. 

2.5.2 Modeling of Disposal Unit Performance 

2.5.2.1 Cementitious Material Performance 

The initial physical properties assumed in the PA for the concrete in Vault 1, Vault 4, and the 

FDCs are presented in Table 2.5-4. The Vault 1 and 4 walls are modeled with an increased 

initial hydraulic conductivity to represent the cracks that have been observed in these vault 

walls. As discussed in more detail in Section 2.7, DOE used moisture characteristic curves 

along with the hydraulic conductivity values listed below in the modeling of flow through the 

system. 

DOE postulated that the physical degradation of disposal unit concrete will be dominated by 

sulfate attack from exposure to sulfate ions remaining in the saltstone porewater after curing. 

Sulfate reacts with cement paste and forms ettringite, an expansive mineral phase often 

associated with concrete spalling or cracking. DOE used the proprietary reactive transport code 

STADIUM® to simulate ettringite formation in disposal unit concretes exposed to solutions with 

three different sulfate concentrations (low, medium, and high sulfate concentration). The results 

of these simulations were used to derive an empirical equation for the position of the ettringite 

front as a function of sulfate concentration. Concrete degradation was estimated by assuming 

concrete ahead of the ettringite front retained its initial properties whereas concrete behind the 

front is cracked or otherwise degraded. Degradation due to sulfate attack was represented in 

the SDF performance assessment by increases in concrete hydraulic conductivity and diffusion 

coefficient. Completely degraded concrete was assumed to have properties similar to those of 

backfilled soil and, thus, not to be a barrier to contaminant release in comparison to the 

environment surrounding the disposal unit. 
44 

Table 2.5-4: Initial Disposal Unit Concrete Properties Assumed by DOE PA Cases and 

Cases K, K1, K2 

Material 

Porosity 

(%) 

Dry Bulk 

Density 

(g/cm3) 

Particle 

Density 

(g/cm3) 

Hydraulic 

Conductivity 

(cm/s) 

Effective 

Diffusion 

Coefficient 

(cm2/s) 

Medium quality 

concrete 

(Vault 1 roof) 

14.5 2.20 2.57 5.0x10-9 1.0x10-7

Medium quality 

concrete 

(Vault 4 roof) 

13.6 2.21 2.56 5.0x10-9 1.0x10-7

High quality concrete

(Vault 1 and 4 base) 12.0 2.24 2.55 3.1x10-10 5.0x10-8

Fractured walls 

(Vault 1 and 4) 12.0 2.24 2.55 1.7x10-1 (1) 5.0x10-8

Low quality concrete

(lower mud mats of 

FDCs) 

21.1 2.06 2.61 1.0x10-8 8.0x10-7

High quality concrete

(FDCs) 11.0 2.22 2.49 9.3x10-11 5.0x10-8

Based on 2009 PA; Table 4.2-16 

(1) In Case K, an initial hydraulic conductivity of 1x10-6

 cm/s was used for Vault 1 and 4 walls 

In Case A, sulfate degradation is assumed to cause the hydraulic conductivity of the Vault 1 roof 

and Vault 1 and 4 floors and walls to increase from their initial values by a factor of 

approximately 3 or less in 20,000 years. Sulfate degradation also is assumed to increase the 

effective diffusivity of the Vault 1 roof and Vault 1 and 4 floors and walls by a factor of 

approximately 4 or less in 20,000 years. However, sulfate attack traverses the full thickness (4 

inches [10 cm]) of the Vault 4 concrete roof in 10,000 years, after which the Vault 4 roof 

hydraulic conductivity and effective diffusivity become equal to that of backfill soil (i.e., a 

hydraulic conductivity of 4.1x10-5 cm/s and a diffusivity of 5.3x10-6 cm2

/s [PA; Table 4.2-14]). 

Also, in Case A, sulfate attack causes the hydraulic conductivity and effective diffusivity of the 

FDC concrete to increase from their initial values by a factor of approximately 6 or less in 

20,000 years (for the roof and floor) or 10,000 years (for the walls). The FDC wall hydraulic 

conductivity is modeled as equal to that of backfill soil after approximately 17,000 years. More 

detail on the modeled changes to the hydraulic conductivity and diffusivity with time is shown in 

Figures 4.2-36 to 4.2-40 of the PA. Additionally, Appendix E in SRNL-STI-2009-00115 provides 

more details on the assumed hydraulic conductivities at different time steps for Cases A - E. 

Some of the deterministic cases modeled by DOE considered alternate assumptions for the 

disposal unit concrete degradation. In the 10x Sulfate Attack Case, DOE assumed that the 

diffusion coefficient used to predict the amount of sulfate attack was increased by a factor of 10, 
45 

which led to a shorter modeled time to complete failure for the disposal unit concrete. In the No 

Sulfate Attack Case, the disposal unit properties were assumed to not degrade and to remain 

constant over time. In the Synergistic Case, the roof and floor of Vault 1 and 4 and the FDC 

concrete degrade hydraulically to soil properties 500 years after closure. As in Case A, the 

Vault 1 and 4 walls are assumed to be initially fractured (represented in the model as increased 

hydraulic conductivity and diffusivity). 

Additionally, in Case K, the Vault 1 and 4 walls were assumed to be initially degraded such that 

the hydraulic conductivity was set equal to 1x10–6 cm/sec, which is similar to the hydraulic 

conductivity of the upper vadose zone soil (PA; Table 4.2-14). The initial hydraulic conductivity 

of the other cementitious materials in the disposal units was assumed to be the same as 

described in Table 2.5-4. In Case K, the degradation of the disposal unit concrete was modeled 

non-mechanistically with increases in the hydraulic conductivity and diffusivity. Except for the 

Vault 4 roof and walls, the disposal unit concrete was modeled with hydraulic conductivity 

increasing with a log-linear relationship to a final value of 1x10-6 cm/s at 10,000 years and 

diffusivity increasing from the values provided in Table 2.5-4 to a value of 5x10-6 cm2

/s at 10,000 

years with a log-linear relationship. As previously noted, the Vault 4 walls were assumed to be 

degraded at closure. The Vault 4 roof was modeled similarly to the other disposal unit 

components but reached the degraded values (i.e., hydraulic conductivity of 1x10-6 cm/s and 

diffusivity of 5x10-6 cm2

/s) at 3,500 years after closure. 

In Section 4.4.2 of the PA, DOE acknowledged that credible fast flow paths include large-scale 

cracks due to seismic events or differential settlement, shrinkage cracks, vertical sheet drains, 

degraded steel columns and roof supports, roof penetrations, and bleed water drain piping. To 

assess the effects of these and other disposal unit concrete degradation mechanisms that could 

result in the formation of fast flow paths, several alternative scenarios were considered in the 

DOE PA. DOE stated each case or scenario was meant to capture the outcome rather than 

predict the cause of the outcome. Cases B, C, and D assumed that a gap forms between 

saltstone and the walls of Vault 4 and of the FDCs at time zero due to the presence of a sheet 

drain at the wall and connecting breaches through the roof and floor. These gaps in Cases B 

and C function as fast pathways for water flow. In Case D, the gap between the disposal unit 

wall and saltstone is not a fast pathway because it is modeled as a capillary break (i.e., an 

impediment to flow). Case C contains additional fast flow paths through the disposal unit from 

the roof through the floor. 

Chemical degradation of the vault and FDC concrete controls the chemical composition of pore 

fluids passing through the vault and FDC and ultimately influences the release of contaminants 

from the vault and FDC. DOE performed simulations using the Geochemist’s Workbench 

geochemical code to estimate the chemical degradation of the concrete 

(SRNL-TR-2008-00283). The simulations estimated changes in Eh and pH as a function of the 

number of pore volumes of infiltrate water that reacts with the cementitious material. The 

concrete mineralogy that was used as input to the thermodynamic simulations was derived 

using a normative calculation that assigned chemical constituents determined by bulk chemical 

analysis to specific mineral phases. The derived vault and FDC concrete mineralogy is 

comprised of calcium-silicate-hydrate (CSH), hydrotalcite, gibbsite, quartz, hematite, and 
46 

gypsum. To represent the reducing component of slag, pyrrhotite was included in the model, in 

an amount based on laboratory reducing capacity measurements (SRNS-STI-2008-00045). In 

the simulations, the concrete minerals were reacted with water having a composition based on 

an analysis of a sample from a water table monitoring well in the vicinity of the Saltstone 

Disposal Facility. The dissolved oxygen concentration in groundwater used in the simulations 

was 8 mg/L—higher than the 1.2 mg/L measured from the well water sample—based on 

assumed groundwater equilibrium with atmospheric oxygen. 

For all radionuclides except Tc, sorption coefficient values (Kd values) were based on specific Eh

and pH transitions in the disposal unit concrete. This process is similar to the process used to 

model radionuclides sorption in saltstone, which is described in more detail in Section 2.6 and 

2.7. The results of the geochemical simulations described above were used to estimate the 

number of times the liquid in the pore volume of the disposal unit concrete must be replaced 

with infiltrating water to cause these Eh and pH transitions. The estimated pore volumes 

required to cause the transitions in the disposal unit concrete were the same in all cases. 

However, the times at which these transitions were predicted to occur were modified in certain 

cases based on the predicted flow through the disposal unit concrete (Table 2.5-5). The 

geochemical modeling indicated that the Eh transitions from reducing to oxidizing conditions 

(from –0.46 to +0.57 V) at 2,994 and 3,230 pore volumes for the Vault 1 and 4 concrete and the 

FDC concrete, respectively. The predicted pH gradually decreases from an initial value of 11.0 

as the CSH phase in the cement dissolves. The modeled CSH phase in the vault concrete and 

FDC concrete is depleted at 3,660 and 4,206 pore volumes, respectively. 

In the modeling of Tc transport in the PA cases (i.e., all cases except Cases K, K1, and K2), the 

Eh of the disposal unit concrete was modeled using a more sophisticated approach that DOE 

refers to as a shrinking-core model. In this model, the reducing capacity of the concrete is 

explicitly tracked in each grid cell in the PORFLOWTM model. When oxygen entering the cell by 

liquid phase transport consumes all of the reducing capacity, the Tc Kd for the cell is changed 

from a high value to a low value, greatly increasing the mobility of the Tc. Additional detail 

regarding the shrinking-core model is provided in Section 2.7. 

In Cases K, K1, and K2, the Eh of the concrete was calculated externally to PORFLOWTM based 

on the diffusion of oxygen into the disposal unit concrete from fractures. The initial fracture 

spacing assumed in Cases K, K1, and K2 is presented in Table 2.5-6. In the RAI response 

describing Case K (SRR-CWDA-2011-00044, PA-8), DOE did not provide the final fracture 

spacing assumed for the disposal unit concrete or address whether the diffusion of oxygen from 

the exterior of the disposal units was included in the oxidation calculation. The fraction of the 

concrete oxidized with time was then used to determine an average Kd value for the concrete. A 

significant fraction of the disposal unit concrete is modeled as remaining reduced during a 

20,000 year analysis time (Figure 2.5-1). Additional detail regarding the average-Kd model is 

located in Section 2.7. 
47 

Table 2.5-5: Chemical Transition Times for Cementitious Materials in Disposal Units 

Cementitious 

Material 

Vault 1 Vault 4 FDCs 

Eh 

Transition 

(years) 

pH 

Transition 

(years) 

Eh 

Transition 

(years) 

pH 

Transition 

(years) 

Eh 

Transition 

(years) 

pH 

Transition 

(years)

Case A

Wall 20,781 21,043 15,519 16,018 16,334 16,753 

Floor >30,000 >30,000 >30,000 >30,000 22,498 23,274 

Oxidized Concrete Case 

Wall 0 21,043 0 16,018 16,334 16,753 

Floor 0 >30,000 0 >30,000 22,498 23,274 

Synergistic Case 

Wall 0 500 0 500 500 500 

Floor 0 500 0 500 500 500 

Cases K, K1, K2

Wall 8,297 9,545 9,219 10,671 7,756 8,232 

Floor 7,740 8,227 >40,000 >40,000 7,970 8,462 

2009 PA; Table 4.2-17 and Sections 5.6.6.5 and 5.6.6.6, and SRR-CWDA-2011-00044 Tables PA-8.8, 

PA-8.9, and PA-8.10 

Table 2.5-6: Initial Fracture Spacing in Cases K, K1, and K2 

Disposal Unit Vault 1 

(m) 

Vault 4 

(m) 

FDCs 

(m) 

Roof 10 10 41 

Wall 10 10 7 

Floor 10 10 41 

From SRR-CWDA-2011-00044 Table PA-8.3 
48 

Figure 2.5-1: Fraction of Disposal Unit Concrete that Remains Reduced (calculated from 

Kd values in Case K PORFLOWTM files) 

2.5.2.2 HDPE and HDPE/GCL Performance 

DOE plans to use HDPE/GCL composite layers on the roof and below the floor (i.e., between 

the upper and lower mud mat) of the FDCs. In addition, DOE plans to cover the walls in a layer 

of 2.5-mm (100-mil) HDPE. Thus the HDPE will form an outer lining material for the FDCs8

. 

The composite layer under the floor will extend 0.6 m (2 ft) up the side of the FDC walls and be 

welded to HDPE that is attached to the cylindrical sides of the FDCs. The HDPE on the sides 

will, in turn, be welded to the HDPE/GCL placed on the FDC roofs. The purpose of the HDPE 

surrounding the FDCs and the HDPE/GCL composite layer on the roof of the FDCs is to limit 

the flow of water and oxygen into the FDCs. The HDPE/GCL layer under the floors of the FDCs 

is designed to limit the flow of water out of the disposal units. In addition, DOE expects the 

HDPE to protect the FDC concrete from potentially corrosive soil components and carbonation. 

DOE also plans to include an HDPE/GCL composite layer in the closure cap. This section 

addresses the HDPE used to cover the FDCs and the HDPE/GCL composite layers on the roof 

and under the floor of the FDCs. The HDPE/GCL composite layer used in the closure cover is 

discussed in Section 2.4. 

 

8

 DOE does not plan to use HDPE as an outer liner for either Vault 1 or Vault 4. 
49 

In the PA, DOE indicated Case A uses an initial hydraulic conductivity9

 of 2x10-13 cm/s and an 

initial diffusivity of 4x10-11 cm2

/s to represent the HDPE lining the FDC walls (PA page 217). 

The HDPE/GCL composite layer in the FDC roof and walls is modeled with an initial hydraulic 

conductivity10 of 2.8x10-12 cm/s and an initial diffusivity of 1.2x10-10 cm2/s (PA page 186). These 

values are modeled as degrading significantly during the first 1,000 years after closure and 

more slowly until 20,000 years after closure (Figure 2.5-2 and Figure 2.5-3). The same 

hydraulic conductivity and diffusivity used in Case A is used to model the HDPE in Cases B - E, 

except that Cases B, C, and D include breaks in the disposal unit roofs and floors 

(Section 2.5.2.1) that include breaks in the HDPE. Case K uses the same values of HDPE and 

HDPE/GCL hydraulic conductivity and diffusivity, except for apparent pathways through the 

floor. As previously discussed, the fractures in the Case K conceptual model are represented 

as a decrease in the bulk hydraulic conductivity and are not modeled explicitly in PORFLOW. 

Although the description of Case K does not include a discussion of fast pathways 

(SRR-CWDA-2011-00044), the PORFLOWTM output files indicate fast pathways are included in 

Case K from the sheet drains and columns extending through the floor, HDPE/GCL composite 

layer, and mud mats (i.e., to the unsaturated soil). DOE did not specifically discuss 

assumptions about HDPE performance in the other modeled case (i.e., any of the other cases 

described in Table 2.1). 

To derive this projected HDPE performance (Figure 2.5-2), DOE considered various HDPE 

degradation mechanisms and selected those expected to cause the most significant 

degradation. DOE then estimated the number and size of defects that would form during the 

performance period (Table 2.5-7). Defects were classified as pinholes, holes, tears, small 

cracks, or cracks, depending on size. DOE then estimated hydraulic properties based on the 

total estimated area of defects per unit HDPE area (SRNL-STI-2009-00115, Table 23). In the 

PA, DOE indicates that the estimated degradation of the HDPE surrounding the FDCs is based 

on the same methods used to estimate HDPE degradation in the closure cap (Section 2.4). 

This procedure considered initial defects as well as degradation with time. 

 

9 The reported initial hydraulic conductivity of HDPE is significantly different from the value of 5.9x10-10 cm/s used in 

the first modeling times step, as presented in SRNL-STI-2009-00115, Appendix E (unnumbered table on page 234). 

These values differ by more than the expected difference between the initial values and the average values for the 

first time step. Both of these values differ from the initial value depicted in Figure 2.5-2. 

10 The reported initial hydraulic conductivity of HDPE is different from the value of 2.2x10-11 cm/s used for the 

HDPE/CGL composite on the roof and under the floor in the first modeling times step, as presented in 

SRNL-STI-2009-00115, Appendix E (unnumbered table on pages 246 and 248). These values may differ because 

the table in SRNL-STI-2009-00115 presents the average for the first time step rather than the true initial value.
50 

Figure 2.5-2: DOE Case A modeled Hydraulic Conductivity and Diffusion Coefficient for 

HDPE Lining of FDCs After Closure (2009 PA, Figure 4.2-42) 

Figure 2.5-3: DOE Case A modeled Hydraulic Conductivity and Diffusion Coefficient for 

HDPE/GCL Composite Layer on the Roof and Below the Floors of the FDCs After Closure 

(2009 PA, Figure 4.2-19) 

The estimate of initial defects is based on DOE’s QA procedures for installing the HDPE. DOE 

provided a description of the QA procedures that would be used for the HDPE in the HDPE/GCL 

layer above the individual FDC roofs (WSRC-STI-2008-00244). These procedures include 

100% visual inspection of the HDPE rolls prior to inspection and industry-recommended 

procedures for avoiding wrinkles. In addition, the QA procedures include 100% visual 

inspection of seams as well as 100% non-destructive vacuum or air pressure testing of seams. 

DOE also indicated that it would conduct periodic destructive testing of seams in accord with 

industry standards, and that the sites of destructive tests would be repaired and tested non-
51 

destructively after the repair. DOE did not indicate whether similar procedures will be used for 

the HDPE on the sides and in the floor of the FDCs. 

Table 2.5-7: Cumulative Area of Holes in HDPE Encasing Disposal Cells 

Year 

Interpolated 

Cumulative Area of 

Holes1

Total Cumulative 

Hole Size2 

(mm2/Hectare) 

Fraction of HDPE 

Membrane with 

Holes3

0 550 5.50x10-8 5.50x10-6

100 46500 4.65x10-6 4.65x10-4

180 122220 1.22x10-5 1.22x10-3

220 157140 1.57x10-5 1.57x10-3

300 226980 2.27x10-5 2.27x10-3

380 296820 2.97x10-5 2.97x10-3

460 366660 3.67x10-5 3.67x10-3

560 453960 4.54x10-5 4.54x10-3

1,000 838080 8.38x10-5 8.38x10-3

1,800 1536480 1.54x10-4 1.54x10-2

3,200 2758680 2.76x10-4 2.76x10-2

5,412 4689756 4.69x10-4 4.69x10-2

5,600 4853880 4.85x10-4 4.85x10-2

10,000 8695080 8.70x10-4 8.70x10-2

Adapted from WSRC-STI-2008-00244 Table 50 

1

 Using WSRC-TR-2005-00101, page D-6 

2

 Total Cumulative Hole Size = earlier year hole size + (year interpolated - earlier year) / 

(later year - earlier year) * (later year hole size - earlier year hole size) 

3

 Fraction of HDPE geomembrane with holes = Total cumulative hole size/10,000,000,000 

mm2

/Hectare 

DOE considered degradation of the HDPE layers by ultraviolet (UV) radiation, antioxidant 

depletion, thermal oxidation, high-energy irradiation, tensile stress cracking, attack from 

saltstone leachate, and biological degradation including microbial action, root penetration, and 

effects of burrowing animals. Of these degradation mechanisms, DOE concluded the 

mechanisms expected to cause the most degradation would be antioxidant depletion, thermal 

oxidation, and tensile stress cracking (WSRC-STI-2008-00244). As previously noted, in the PA 

DOE indicated that the degradation of the HDPE lining the exterior of the FDCs was based on 

the same evaluation of degradation made for the HDPE in the closure cap. For the HDPE in the 

closure cap, the estimated duration of antioxidant depletion, during which few defects are 

generated from oxidation, was based on recent studies (Mueller and Jakob, 2003). To estimate 

the creation of defects (i.e., pinholes, holes, tears, and cracks) in the closure cap HDPE from 
52 

the combination of antioxidant depletion, thermal oxidation, and tensile stress, DOE used the 

method of Needham et al. (2004) which was developed by the Environment Agency of England 

and Wales to support PAs for landfills (WSRC-STI-2008-00244). 

DOE expects the effects of other HDPE degradation mechanisms to be limited. For example, 

DOE expects degradation due to UV radiation during the time between FDC construction and 

backfilling to be limited by the incorporation of 2 to 3% carbon black and other ultraviolet 

chemical stabilizers into the HDPE. Attack from saltstone leachate is expected to be limited by 

use of the interior liner to limit bleed water penetration into the disposal units during curing 

(Section 2.5.1.3). DOE identified several methods to limit high-energy-irradiation induced 

damage of the HDPE layers, including shielding, lowering the level of oxygen exposed to HDPE, 

increasing antioxidant concentration in HDPE, use of thicker HDPE layer, and minimizing tensile 

stresses on HDPE layers. Of these, the current design calls for shielding (i.e., by the disposal 

unit walls), exposure to sub-surface (rather than atmospheric) oxygen concentrations, and use 

of thick (i.e., 2.5-mm [100-mil]) HDPE. DOE indicated that little information is available on longterm degradation of HDPE by fungi or bacteria, but that a common engineering textbook on 

geosynthetics (Koerner, 1998) describes HDPE as resistant to microbial degradation. DOE also 

expects the 2.5 mm (100 mil) HDPE used to cover the FDCs will be impervious to tree roots, 

except in areas with existing holes. This conclusion is based on Environmental Protection 

Agency (EPA) and industry experience with HDPE membranes used in landfill applications 

indicating that tree roots are effectively stopped even by thinner HDPE layers (e.g., 

approximately 8 mm [30 mil]). DOE also excludes the effects of burrowing animals from further 

consideration because DOE expects that burrowing animals will be deterred by the overlying 

erosion barrier (WSRC-STI-2008-00244). 

With respect to GCL degradation, DOE considered the effects of slope stability, freeze-thaw 

cycles, dissolution, divalent cations (e.g., Ca+2, Mg+2), desiccation (wet-dry cycles), and 

biological degradation (e.g., root penetration, burrowing animals) (WSRC-STI-2008-00244). 

Of these mechanisms, DOE concluded that degradation by divalent cations is likely to cause the 

most significant effects on the GCL performance. Divalent cations can cause GCL degradation 

because the divalent cations replace two Na+ ions in the primary component of the GCL 

bentonite (i.e., sodium-montmorillonite). This replacement results initially in clays with 

approximately half the swelling capacity and poorer hydraulic conductivity, and subsequently in 

additional minerals that also have poorer hydraulic conductivity than sodium-montmorillonite. 

To estimate the effects of divalent cations on the GCL, DOE performed geochemical modeling 

with Geochemist’s Workbench to evaluate the transformation of the sodium-montmorillonite in 

the GCL into other minerals as a function of the volume of water that has passed through the 

GCL. DOE used two different modeled water compositions: one representative of natural 

infiltrating water at the site and one representative of water equilibrated with portlandite 

(Ca(OH)2). As expected, DOE found that the water equilibrated with portlandite caused much 

faster modeled degradation than the water equilibrated with site soil. Specifically, approximately 

1,000 L/m2

 (100 L/ft2) were required to cause significant transformation of the sodiummontmorillonite when water equilibrated with soil was modeled (WSRC-STI-2008-00244, 

Figure 17). Significant transformation of sodium-montmorillonite was seen after less than 
53 

50 L/m2

 (5 L/ft2) of water equilibrated with portlandite was reacted with the same volume of the 

sodium-montmorillonite GCL (WSRC-STI-2008-00244, Figure 17). Because DOE judged there 

to be significant uncertainty in the geochemical modeling results, DOE did not directly apply the 

results and, instead, simply assumed the GCL layer to be degraded after the 100 year period of 

institutional controls. Specifically, DOE assumed the sodium montmorillonite GCL to be 

converted to calcium or magnesium montmorillonite with a saturated hydraulic conductivity 

approximately one order of magnitude higher (i.e., 5X10-8 cm/s). DOE used this hydraulic 

conductivity of the GCL alone in its determination of the hydraulic conductivity of the HDPE/GCL 

composite layer (Figure 2.5-3). 

DOE provided rationale, summarized here, for excluding the other identified degradation 

mechanisms from modeling (WSRC-STI-2008-00244). DOE judged that slope stability will not 

significantly affect the HDPE/GCL composite layer performance because of the relatively low 

slope of installation. DOE expects that freeze-thaw cycling will not cause significant 

degradation for two reasons. The first is that the GCL layers on the roof and under the floors of 

the disposal units will be installed significantly below the 25 cm (10 in) depth to which soil 

freezes in South Carolina. The second is that DOE indicates that the clay in GCLs deforms 

easily around ice lenses and heals well after thawing. Regarding dissolution, DOE reasons that 

the relatively low amount of water DOE expects to reach the GCL layers will not lead to 

significant dissolution of the sodium-montmorillonite in the GCL into other minerals 

(e.g., kaolinite and quartz). DOE does not anticipate that burrowing animals will cause 

significant degradation because DOE expects they will be stopped by the erosion control layer 

in the closure cap. Similarly, DOE expects tree roots to be effectively deterred by the HDPE 

layer of the HDPE/GCL composite, and to cause GCL degradation only in areas of existing 

HDPE holes. In addition, DOE also expects the HDPE layer of the HDPE/GCL composite to 

prevent significant deterioration due to desiccation. 

2.5.2.3 Integration and Interaction of Disposal Unit Materials 

The FDCs contain complicated joints with many material interfaces. The FDC walls, in 

particular, are composed of multiple separate pre-cast wall panels that are joined in the field. 

DOE abstracted and simplified various components of Vaults 1 and 4 and the FDCs into the 

near-field PORFLOWTM model, as illustrated in Figure 2.5-4. As an example of the type of 

simplifications DOE made to model the disposal units, the complexity of the FDC wall-floor 

interface, including proposed changes to Cells 2A and 2B (as of November, 2010) 

(SRR-ESH-2010-00101), is shown in Figure 2.5-5. The abstraction of the wall-floor interface for 

the FDCs for the PORFLOWTM modeling is shown in Figure 2.5-6 (SRR-ESH-2010-00101). In 

addition to the physical complexity of the joints, the joints include a variety of different types of 

materials (e.g., epoxy, polysulfide material). 

DOE developed Cases B and C (Section 2.5.2.1) to evaluate the dose impact of preferential 

pathways through the disposal units that may develop within a 10,000 year period of 

performance. As discussed in Section 4.4.2 of the PA, DOE designed these cases to capture 

the effect of preferential pathways through the disposal units, rather than explicitly represent all 

potential preferential pathways. 
54 

Figure 2.5-4: Respective Material Zones for PORFLOWTM modeling 

(adapted from 2009 PA) 
55 

Figure 2.5-5: FDC wall-floor interface including proposed design enhancements (as of 

November, 2010) (adapted from SRR-ESH-2010-00101) 

Figure 2.5-6: FDC lower corner detail for PORFLOWTM modeling (2009 PA) 
56 

2.5.3 NRC Evaluation – Disposal Unit Design and Performance 

2.5.3.1 Construction Quality and Testing 

The PA relies on the disposal units to provide secondary containment of the contaminants as 

well as limit waste form exposure to oxidizing chemical conditions. Therefore, inspection of the 

disposal units is important to identifying deviations from the disposal unit design and assumed 

performance within the PA, as discussed in the NRC’s monitoring plan (NRC, 2007a). 

In comment VP-1, NRC staff requested additional information to assess the applicability of the 

degradation mechanisms responsible for the observed fracturing of Vaults 1 and 4 to the FDCs 

and to other portions of Vaults 1 and 4. In its response (SRR-CWDA-2010-00033), DOE 

described design enhancements to the FDCs it believed would prevent the formation of cracks 

that were observed in Vaults 1 and 4 walls. These design enhancements include multiple 

engineered systems, structures, and components to address the potential buildup of liquids 

inside the FDCs, such as a system of sheet drains and associated drain water piping to drain 

liquids from the space between the saltstone monolith and FDC walls. Another is the cylindrical 

design that incorporates pre-stressing wires to counteract the stress of free-standing salt 

solution. Also, an interior epoxy-based coating is used to preclude liquids from penetrating the 

walls during filling operations and to protect the walls from sulfate attack during the 

approximately 30 year operational period when sulfate-bearing bleed water is the greatest 

concern. In addition, seismic restraints are included outside the concrete walls to address 

potential movement during a seismic event. Geotechnical investigations have been conducted 

(Chapter 5) as part of the overall FDC design and construction process to prevent potential 

differential settlement of the FDCs that might lead to cracking. 

During hydrotests of disposal units 2A and 2B, damp spots were observed in several locations 

at the wall-floor joint and at the interface of the floor and mud mat (SRR-ESH-2010-00101). The 

observation of a fluorescent dye tracer confirmed that the damp spots were due to disposal unit 

leakage (SRR-ESH-2010-00101). Based on interior and exterior inspections, leakage was 

determined to have occurred through cracks beneath the anchor bolts and around the interior 

curb (SRR-ESH-2010-00101). Recurring delamination of the epoxy coating was determined to 

be a likely contributor to the leakage (SRR-ESH-2010-00101). NRC Onsite Observation 

Reports (NRC, 2010d, h) provide additional information on the hydrotesting of disposal units 2A 

and 2B. In its second response to NRC comment VP-3 (SRR-CWDA-2011-00044), DOE 

described significant FDC design modifications that have been implemented after the FDC 

2A/2B hydrotesting. The modifications to disposal units 2A and 2B include cutting the anchor 

bolt penetrations flush with the floor, applying an interior coating, installing an additional 36 cm 

(14 in) of sulfate resistant concrete above the coating layer, and installing an exterior sulfate 

resistant curb around the perimeter of the disposal unit floor. DOE stated that these 

modifications were made to ensure the disposal units are constructed and operated in such a 

manner that they have properties consistent with the Case A assumptions. However, NRC staff 

has expressed concern that the installation of curbing to the exterior of the disposal units prior to 

the hydrotest would prevent the potentially weak interfaces that could have leaks from being 

visible during the hydrotest, thereby confounding any conclusions about the effectiveness of the 
57 

disposal unit repairs (NRC, 2010h). In addition, DOE discussed that significant design 

modifications to the disposal units, including units 7A - D11 have been implemented based on 

the lessons learned from the hydrostatic testing (e.g., future disposal units do not include anchor 

bolt penetrations in the floor). 

In addition to the design modifications, the hydrotest procedures were also modified 

(SRR-ESH-2010-00101) to accommodate changes to the allowable stresses to the disposal unit 

resulting from the installation of the curb that was added on the exterior of the disposal units. 

The maximum test height of water was revised from 6.7 m to 3.7 m (22 ft to 12 ft) and the test 

duration was extended to a minimum of 98 hrs to provide a sufficient time for the hydrotest to 

equate to a hydraulic head of 6.7 m (22 ft). After the disposal units are backfilled, DOE 

anticipates that a fill height of 6.3 m (20.6 ft) for a liquid with a specific gravity of 1.3 will not 

exceed allowable stresses. NRC staff agrees that backfilling will help reduce stresses on the 

disposal unit walls. However, similar to the installation of curbing, the presence of backfill 

during the filling of the disposal units with saltstone grout will prevent visual inspection and 

result in additional uncertainty in the disposal unit integrity. The incorporation of a liquid-sample 

collection system from beneath certain disposal units (State of SC, 2007) may increase 

confidence in the short-term performance of the disposal units. 

Based on the performance of the floors of FDC 2A and 2B during recent hydrotesting (due to 

cracking near anchor bolts), it is not clear that DOE’s Case A assumption that there are no 

fractures in the floors of Vaults 1 and 4 is realistic, because Vaults 1 and 4 also contain anchor 

bolts. The NRC staff suggested (NRC, 2010b; VP-3) that DOE include vault floor fractures in its 

base case or provide a basis for not including this feature in light of limited vault floor 

characterization (SRR-CWDA-2011-00044). In response, DOE described a visual inspection 

that was done on drainwater system anchor bolts in Vault 4 Cells B and H, which indicated no 

evidence of voids or cracking at the embedded anchor bolt locations. DOE stated that plans 

and procedures have been developed to perform regular visual disposal unit inspections to 

identify any structural deterioration that may require evaluation. NRC staff considers DOE’s 

response to cracking beneath the anchor bolts adequate to demonstrate that the floors do not 

currently have visible cracks and that DOE will monitor and address the potential development 

of visible cracks prior to filling the disposal units. However, significant uncertainty remains 

about the Case A assumption that the disposal unit floors will not fracture for thousands of years 

after closure. 

2.5.3.2 NRC Evaluation – Disposal Unit Cementitious Materials Performance 

The initial hydraulic conductivity and effective diffusion coefficient values assumed in the PA for 

the FDC are in the range of those observed in the literature for concretes for non-fractured 

concrete (CNWRA, 2008). The NRC staff finds that these values seem reasonable for the 

concrete in the FDCs. The Vaults 1 and 4 walls have already shown cracking and were 

modeled in the PA with an increased saturated hydraulic conductivity. A high hydraulic 

conductivity value (i.e., higher than the hydraulic conductivity of soil) was assumed for the 

 

11 DOE has changed the numbering of the FDCs since the PA was published. These FDCs are now referred to as 

2A/2B, 3A/3B, and 5A/5B. 
58 

Vault 1 and 4 walls in the PA cases (i.e., cases other than Case K, K1, and K2). However, as 

described in more detail in Section 2.7, the modeled flow was limited by the assumed moisture 

characteristic curves. The Vault 1 and 4 roofs and floors are assumed to have properties of 

medium and high quality concrete, respectively. As described above, the Vault 4 roof has 

previously leaked. Repairs were made by DOE to seal the joints in the roof. However, it is 

unclear how well these repairs will perform in the future. Additionally, as described in the 

previous section, although DOE inspected the Vault 4 floors, uncertainty remains in their future 

hydraulic performance because the floors may have fractures that are not readily visible but 

transmit water. 

In the PA cases, the degradation of the disposal unit concrete was based on sulfate attack 

modeling results performed using the STADIUM® code. The PA and supporting documents 

predict that all of the vault and FDC concrete components except the Vault 1 and 4 walls and 

Vault 4 roof would remain essentially intact with hydraulic conductivities increasing by less than 

an order of magnitude within 10,000 years. The estimated hydraulic conductivity of the Vault 4 

roof increased by approximately three orders of magnitude within 10,000 years, and the Vault 1 

and 4 walls were modeled as being degraded at closure. 

In comment VP-2 (NRC, 2010i), NRC noted that degradation mechanisms other than sulfate 

attack (e.g., alkali silica reaction and corrosion-induced degradation) were not included in the 

evaluation of the degradation of the Vaults 1 and 4 roof and floor, or the FDC roof, walls, and 

floors. The NRC staff specifically noted that the design of the FDCs includes significant 

amounts of carbon steel components (e.g., rebar, prestressing wires, diaphragms), which could 

corrode and lead to expansive reactions that could cause cracks to form in the concrete. 

In its second response to NRC comment VP-2 (SRR-CWDA-2011-00044), DOE provided 

additional information to support its assumption that degradation mechanisms other than sulfate 

attack will not significantly increase the degradation of the disposal unit concrete. DOE 

addressed (i) alkali–silica reaction, (ii) carbonation, (iii) chloride-induced corrosion, (iv) calcium 

leaching, (v) microbial degradation, (vi) freeze–thaw cycles, and (vii) cracking from seismic 

events, settlement, and external static loading. For example, DOE cited petrographic testing of 

concrete samples from the exterior and interior of the F- and H-Canyon structures that indicated 

alkali–silica reaction was not present in that concrete to date. DOE also indicated that, because 

the aggregate that will be used in the disposal unit concrete has a large component of relatively 

non-reactive granite, the potential for adverse alkali-silica reactions is limited. 

DOE used a concrete carbonation rate derived from carbonation depths measured in P-reactor 

concrete to estimate the time interval for the Vault 1 and 4 concrete to be carbonated and for 

rebar corrosion to initiate. The estimated time interval was in excess of 10,000 years, which is 

similar to the PA estimated concrete degradation time by sulfate attack. DOE also provided an 

estimation of the amount of chloride induced corrosion expected. DOE predicted that it would 

take over 6,000 years to reduce 5% of the rebar and over 10,000 years to reduce 10% of the 

rebar. DOE noted that this reduction could result in the formation of preferential or fast flow 

paths. 
59 

In the response to NRC comment VP-2, DOE stated that calcium leaching is not expected to be 

significant in the saltstone disposal units, based on an assessment provided in 

NUREG/CR-5542. This is consistent with NRC staff’s understanding of the potential 

degradation due to calcium leaching. However, the effects of calcium leaching may be 

important to the overall degradation of cementitious materials due to the potential coupling of 

multiple degradation mechanisms. DOE also stated that microbial organisms could potentially 

promote damage to cementitious materials and further research on this topic is planned. DOE 

stated that exposure of FDC concrete to freezing conditions is not expected after the operational 

period because, after the disposal units are filled, they will be surrounded with backfill and a 

closure cap will be installed. However, the NRC staff expects that some freeze-thaw damage 

could occur prior to the installation of the closure cap for Vaults 1, 4, and any portions of the 

FDCs that are covered by less than 25 cm (10 in) of soil, which is the depth to which soil freezes 

in South Carolina. 

DOE indicated cracking from seismic events, settlement, and external static load have been 

considered in the design and construction of the vaults and FDCs and macroscopic cracks in 

the exterior Vault 1 and 4 walls were modeled in the 2009 PA. However, the NRC staff notes 

that the formation of cracks in the Vault 1 and 4 walls were modeled assuming bulk degradation 

of the walls, not discrete fractures, and the Case A modeling included the use of moisture 

characteristic curves that NRC staff questioned the basis (Section 2.7). Additionally, the Case A 

model did not include fracturing of the Vault 1 and 4 floors. The NRC staff also disagreed with 

some aspects of the settlement and structural integrity analysis (Chapter 5). Specifically, recent 

geotechnical investigations for FDCs 2A/2B, and 7A, 7B, 7C, and 7D (Figure 2.8-8])12 predicted 

settlement values greater than the American Concrete Institute (ACI) standard (ACI-372) for the 

FDCs (K-ESR-Z-00001; K-ESR-Z-00002). Because neither the standard nor the DOE analysis 

addresses the consequences of exceeding these criteria, the risk-significance of this finding is 

unclear. Accordingly, the NRC staff will monitor the development of information related to 

settlement of the SDF due to static loading. 

DOE predicted the effects of settlement to result in cracking of the floor and the overlying 

saltstone grout at the construction joints (T-CLC-Z-00006). These predictions are inconsistent 

with the hydraulic conductivities that were assumed in the PA for Vault 4 (e.g., 3.84x10-10 cm/s 

at 10,000 yrs) (SRNL-STI-2009-00115, Appendix E). In addition, this structural integrity 

analysis assumed that the strength of the concrete and saltstone do not change with time 

(T-CLC-Z-00006). However, the exclusion of chemical degradation mechanisms which tend to 

decrease the strength of cementitious materials (e.g., sulfate attack, rebar corrosion, alkali silica 

reaction) from the structural integrity analysis could result in an underestimation of the amount 

of fracturing. Both physical and chemical degradation mechanisms can affect the long-term 

performance of disposal units. Therefore, coupling these mechanisms in an integrity analysis is 

necessary to ensure that predictions of disposal unit performance are not overly optimistic. 

 

12 DOE has changed the numbering of the FDCs since the PA was published. These FDCs are now referred to as 

2A/2B, 3A/3B, and 5A/5B. 
60 

Based on the information in this section, which describes a limited evaluation of degradation 

mechanisms for disposal units and the inconsistency between relevant studies and the PA, 

NRC staff finds the assumed long-term performance of disposal units in Case A to be 

unsupported. 

In addition to the physical and hydraulic properties of the disposal unit concrete, the chemical 

properties can also significantly affect the transport of radionuclides out of the disposal unit and 

into the environment. The modeled transport of many elements depends on the Eh and pH of 

the disposal unit concrete, which affect the sorption coefficient (Kd value). The three different 

approaches DOE used to model the effects of Eh and pH on radionuclide release from saltstone 

and transport through the disposal units are discussed in more detail in Sections 2.7.1 and 

2.7.4.1 and summarized here. 

In the pore-volume modeling approach DOE used to model the chemical properties of the 

cementitious materials for radionuclides other than Tc (in all cases), the chemical conditions of 

the whole component (i.e., saltstone monolith or disposal unit floor, walls, or roof) transition in a 

step change. That is, the Eh and pH of the whole component transition only when until enough 

water has traveled through the entire component to completely consume the reducing capacity 

(for Eh) or to consume the buffering capacity to a certain pH. However, in the real system, 

portions of the concrete are likely to undergo earlier Eh and pH transitions due to relative 

position and preferential flow through intact and degraded regions. Earlier transitions in certain 

regions would affect the modeled transport of elements that have significantly different chemical 

properties (e.g., solubility limit or Kd value) depending on the Eh and pH of the system13. The 

NRC staff notes that the dissolved oxygen concentration used in the calculation of the number 

of pore volumes required to cause Eh transitions is conservatively assumed to be in equilibrium 

with atmospheric oxygen. The concentration of oxygen in the subsurface may be lower (e.g., 

DOE reports a measured value of 1.2 mg/L from a groundwater monitoring well near the SDF). 

Thus the NRC staff finds the pore-volume approach appropriate for elements with limited 

sensitivity to system Eh and pH. 

The NRC staff finds that the explicit shrinking core approach that DOE used in Case A and the 

other PA cases to model the transport of Tc in the disposal units is more appropriate than the 

pore-volume approach appropriate for elements sensitive to system Eh or pH (e.g., Tc). In 

particular, the NRC staff finds that tracking releases from the oxidized and reduced fractions of 

saltstone separately appears to be the most appropriate implementation of DOE’s conceptual 

model of Tc-99 release. Regarding the oxidation modeling, the NRC staff finds DOE’s 

assumption that infiltrating water is equilibrated with atmospheric oxygen (i.e., 8 mg/L dissolved 

oxygen, equivalent to 1.06 meq/L) to be conservative because subsurface oxygen 

concentrations are likely to be lower (e.g., DOE reports a measured value of 1.2 mg/L from a 

groundwater monitoring well near the SDF). However, the NRC staff finds DOE’s basis for 

neglecting gas-phase transport of oxygen to saltstone surfaces, which relies on saltstone 

saturation, to be insufficient (Section 2.7.4.1). In addition, the NRC staff finds DOE’s omission 

 

13 The modeled release of the redox-sensitive radionuclide Tc-99 is not affected by this concern because, as 

previously described, Tc-99 release was modeled using a different approach. 
61 

of gas-phase oxygen transport in the PA cases is inconsistent with DOE’s Case K, K1, and K2 

results, which demonstrate that gas-phase oxygen can have a non-negligible effect on saltstone 

and disposal unit concrete oxidation. The relative importance of these two sources of oxygen 

(i.e., gas-phase oxygen and dissolved oxygen in infiltrating water) will depend on assumptions 

about water flow through the system as well as the concentrations of oxygen in soil, gas, and 

infiltrating water. 

In Cases K, K1, and K2, DOE separates modeling of oxidation from modeling of Tc release and 

transport. Specifically, in Cases K, K1, and K2 DOE models saltstone and disposal unit 

oxidation by calculating the fraction of the concrete that is oxidized based on the diffusion of 

oxygen into the disposal unit concrete through fractures. Unlike the explicit shrinking core 

model implemented in PORFLOW, the average-Kd model does include diffusion of oxygen from 

the gas phase into saltstone or the disposal unit concrete, however it neglects oxygen 

introduced into the saltstone or disposal unit in infiltrating water. In Cases K, K1, and K2, DOE 

assumes that the soil gas in contact with saltstone has atmospheric concentrations of oxygen. 

The NRC staff finds this to be a conservative assumption. However, it is unclear if this 

conservative assumption entirely compensates for omitting water in infiltrating water from the 

oxidation model. As previously noted, the relative importance of dissolved oxygen in infiltrating 

water depends on assumptions about the dissolved oxygen concentration and flow rate. 

In Cases K, K1, and K2, after the extent of oxidation is calculated, the sorption coefficient (Kd

value) used in the model is then based on a weighted average of the oxidized and reduced Kd

values and the fraction of the concrete that is oxidized (Section 2.7.1). As discussed in more 

detail in subsequent paragraphs and in Section 2.7.1.4, this “average-Kd” modeling approach 

results in significant unexpected modeled chemical performance. 

DOE considered a number of cases that included various assumptions for the physical, 

hydraulic and chemical degradation of the disposal unit concrete. Case A, which DOE 

considers to be the base case, contains limited physical and hydraulic degradation. As noted 

above, the NRC staff does not consider the assumptions and parameter values used in the 

Case A modeling for the disposal unit concrete to represent the current and future expected 

performance. In particular, the NRC staff did not agree with the minimal degradation assumed 

for the FDC concrete and the Vault 1 and 4 floors and roofs. The Case A modeling 

assumptions for the disposal unit concrete did not include all of the relevant degradation 

mechanisms, particularly those mechanisms that result in the formation of discrete fractures in 

the concrete. The NRC staff also noted that the modeled diffusive release through the Vault 4 

walls was inconsistent with the observed advective seeps that occurred during grout placement 

(NRC, 2007a). 

The disposal unit assumptions in Case C were similar to those in Case A, except that Case C 

accounts for a discrete fracture through the disposal unit. Similarly, the 10x Sulfate Attack Case 

considers a faster rate of degradation of the concrete due to an increased rate of sulfate attack. 

The Oxidized Concrete Case evaluates the dose if the Vault 1 and 4 walls are assumed to be 

oxidized at the time of closure. However, like Case A, these cases contain other modeling 

assumptions that the NRC staff does not agree with, such as the use of overly optimistic 
62 

moisture characteristic curves (Section 2.7), and overly optimistic assumptions for the chemical 

and hydraulic performance of the saltstone waste form (Section 2.6). 

The Synergistic Case (PA Section 5.6.6.5) non-mechanistically accounts for early hydraulic and 

chemical degradation of disposal unit roofs, floors, and walls. However, the degree of overall 

conservatism assumed in the sensitivity analysis is unclear because the Synergistic Case 

incorporates overly optimistic assumptions for the chemical and hydraulic performance of the 

saltstone waste form (Section 2.6). In its response to NRC comment VP-2 

(SRR-CWDA-2010-00033), DOE provided results of a revised Synergistic Case that 

demonstrated that use of the moisture characteristic curves did not significantly affect the 

synergistic case results (i.e., the assumption that the relative permeability is 1 did not 

significantly affect the results). However, this case still included the overly optimistic 

assumptions for the chemical and hydraulic performance of the saltstone waste form 

(Section 2.6). 

As noted in Section 2.1, the NRC staff considers the modeling performed in Case K, K1, and K2 

to most appropriately reflect current and future expected conditions at the SDF. These cases 

include significant bulk degradation of the disposal unit concrete, which is represented by an 

increase in the hydraulic conductivity of the cementitious materials to a hydraulic conductivity 

value that is typical for soil. However, it is unclear whether these cases include any discrete 

fractures in the disposal unit concrete. The response to NRC comment PA-8 

(SRR-CWDA-2011-00044) does not describe the presence of discrete fractures in this model. 

However, the Case K PORFLOWTM files provided to the NRC indicate that minor fracturing (i.e., 

a single discrete fracture) may have been included. Regardless, Cases K, K1, and K2 derive 

significant chemical performance from the disposal unit concrete that is inconsistent with the 

assumption that the disposal unit concrete is fractured (Section 2.13). As seen in Figure 2.5-1, 

even though significant hydraulic failure of the disposal unit concrete occurred in Case K, K1, 

and K2, a significant fraction of the disposal unit concrete remained reduced, even at long 

evaluation times. This resulted in a large amount of modeled sorption of Tc-99, which resulted 

in a lower rate of release and a delayed time of release from the disposal unit for this 

radionuclide. In reality, the flow through the disposal unit concrete will likely occur through 

fractures rather than through the bulk concrete. The NRC staff expects these fractures to 

become oxidized (i.e., less sorptive for Tc-99) more quickly than the bulk concrete and will not 

provide as much chemical performance. 

2.5.3.3 NRC Evaluation – HDPE and HDPE/GCL Composites 

In general, the NRC staff concludes that the hydraulic properties assigned to the HDPE/GCL 

composite layers on the FDC roof and under the FDC floors and the HDPE on the FDC walls 

appear to be reasonable. The NRC staff evaluated DOE’s consideration of potential sources of 

HDPE and GCL degradation and concluded most major potential degradation modes were 

considered. However, the NRC staff is concerned that the potential for differential settlement 

due to static settlement or consolidation of soft zones has not been evaluated fully and could 

impact the HDPE/GCL performance (Chapter 5). 
63 

DOE indicated that its estimated degradation of the HDPE lining the exterior of the FDCs was 

based on the same methods that were used to estimate degradation of the HDPE in the closure 

cap (PA page 217). The NRC staff finds the application of the methods used to estimate 

antioxidant depletion (Mueller and Jakob, 2003) were reasonable. Although considerable 

uncertainty in antioxidant depletion time is introduced by the uncertainty in the choice of an 

activation energy (e.g., WSRC-STI-2008-00244 Table 28), DOE’s selection appears to be 

reasonable. Specifically, in the context of the closure cap, the NRC staff does not expect DOE’s 

choice of the duration of antioxidant depletion (i.e., 275 years) (WSRC-STI-2008-00244 

Appendix I) to have a significant effect on performance during a 10,000 year compliance period. 

Although it is unclear whether DOE used the same choice with respect to the HDPE 

surrounding the FDCs, the evolution of the hydraulic conductivity and diffusivity of the FDC 

HDPE (Figure 2.5-2) indicates that DOE appears to have chosen an appropriate value of the 

duration of antioxidant depletion. 

With respect to estimation of the combination of the effects of antioxidant depletion, thermal 

oxidation and tensile stress cracking, in general, the NRC staff finds the application of the 

method of Needham et al. (2004) to be reasonable. The applicability of the method of 

estimating defects due to tensile stress to the vertical sections of HDPE on the FDC walls and 

the interfaces between the HDPE on the walls, roof, and under the floor is unclear. However, 

the NRC staff expects the method to apply to the HDPE in the HDPE/GCL layers on the roof 

and under the floor, which are expected to have the most significant effect on modeled HDPE 

performance (Section 2.7). 

With respect to HDPE degradation from exposure to saltstone and concrete leachates, as well 

as site groundwater, the NRC staff concludes that short-term tests do not indicate the potential 

for HDPE corrosion, and that long-term information is lacking. The leachate from the saltstone 

grout is expected to be an aqueous solution containing aluminum nitrate, sodium carbonate, 

sodium nitrate, sodium nitrite, sodium sulfate, and sodium phosphate and from concrete is 

expected to be an aqueous solution containing predominantly calcium hydroxide. The 

groundwater is expected to contain predominantly sodium and calcium cations, and chloride 

and carbonate anions. Of these, calcium hydroxide, sodium hydroxide, aluminum nitrate, 

sodium carbonate, and sodium chloride aqueous solutions are not expected to be particularly 

corrosive to HDPE (Schweitzer, 2004). No literature information could be found on the effects 

of sodium nitrite, sodium nitrate, sodium sulfate, and sodium phosphate aqueous solutions on 

HDPE. However, HDPE is not expected to be adversely affected by calcium sulfate, calcium 

nitrate, up to 50 % phosphoric acid, and up to 20 % nitric acid aqueous solutions 

(Schweitzer, 2004) below 60 °C. Although there is limited information about the effects of these 

chemicals on HDPE performance in the long term (i.e., thousands of years), the NRC staff 

concludes that the potential effects of these chemicals on the HDPE are accounted for, at least 

in part, by DOE’s modeled rapid degradation of HDPE hydraulic conductivity and diffusivity 

(Figure 2.5-2). 

These chemicals may have a more significant effect on the performance of the HDPE/GCL 

composite because of the importance of divalent cations to GCL degradation. Specifically, DOE 

found that a relatively small amount of water equilibrated with portlandite was expected to cause 
64 

significant transformation of the sodium-montmorillonite GCL into higher-hydraulic conductivity 

minerals (WSRC-STI-2008-00244, Figure 17). However, the effects of this transformation 

appear to be accounted for by DOE’s choice to increase the modeled hydraulic conductivity of 

the GCL layer to 5 x10-8 cm/s at 100 years after site closure. The NRC staff does not expect 

DOE’s selection of 100 years after site closure for this transformation to be important to 

modeled FDC performance. 

With respect to the other degradation GCL mechanisms DOE considered, the NRC staff finds 

the discussions of slope stability, desiccation, and biological factors to be appropriate. 

Regarding freeze-thaw cycling, the NRC staff agrees that, once installed, the GCL layers should 

be protected from freezing, but notes some of the GCLs will be installed several years prior to 

installation of the closure cap. Neglect of freeze-thaw degradation therefore does rely in part of 

DOE’s understanding that the GCLs will recover well from freezing. The discussion of the 

potential effects of sodium-montmorillonite dissolution did not appear to be complete. 

Specifically, DOE estimated the time required for complete dissolution of a unit area of the GCL 

under natural infiltration conditions (i.e., approximately 1,200 years). DOE then reasoned that 

since substantially less water was actually expected to reach the GCL layers over and under the 

FDCs, dissolution would not occur during the performance period. This reasoning appears to 

neglect the potential degradation of GCL performance from partial dissolution of the sodiummontmorillonite. However, the NRC staff concludes that dissolution is likely to cause less 

degradation than exposure to divalent cations and that the degradation is appropriately reflected 

in the evolution of hydraulic properties of the HDPE/GCL composite layer DOE used in its PA 

models (Figure 2.5-3). 

Use of a material with which there is limited long-term engineering experience and no natural 

analogues, such as HDPE, introduces conceptual model uncertainty. For example, if the HDPE 

performs better than expected, and forms few defects for thousands of years after placement, 

the saltstone in the FDCs could oxidize substantially from gas-phase transport of oxygen while 

being exposed to very little water. If the HDPE were then to begin to fail several thousand years 

after placement, when the closure cover and FDC roofs are already significantly degraded, the 

oxidized saltstone in the FDCs could be suddenly exposed to a much larger flow of water that 

could cause the release of a significant fraction of the Tc-99 inventory in a relatively short 

amount of time. The sudden failure of the HDPE on the roof and under the floor of the FDCs is 

expected to be mitigated to some extent by the GCL, which the NRC staff expects to fail more 

gradually. However, if both layers fail as the result of a disruptive event (e.g., an earthquake or 

formation of a sink), water flow through the FDC could increase significantly in a relatively short 

time. Thus, information regarding the potential for sudden failure of the HDPE/GCL layers is 

important to an evaluation of predicted site performance. 

2.5.3.4 NRC Evaluation – Integration and Interaction of Disposal Unit Materials 

DOE’s abstraction of the complex material interfaces present in the disposal units does not 

account for preferential pathways in Case A. The low hydraulic conductivity (e.g., 

3.60x10-10 cm/s [1.42x10-10 in/s] for the FDCs at 10,000 yrs) of the modeled material interfaces 

in Case A are representative of seamless and undegraded high-quality concrete (hydraulic 
65 

conductivity value from Appendix E of SRNL-STI-2009-00115). However, observations from 

disposal units 2A and 2B during the hydrotests and observations of seepage at interfaces in 

Vault 4 are indicative of the challenges associated with constructing water-tight joints. 

Furthermore, there is little support for the assumption that joints can remain essentially water 

tight over long periods of time. In particular, the inclusion of materials about which there is little 

information to support predictions of long-term performance (e.g., epoxy, polysulfide material) 

add uncertainty to the predictions of the long-term performance of the FDCs. 

Because Cases B, C, and the Synergistic Case capture a range of potential disposal unit 

degradation mechanisms, the NRC staff concludes that these cases are more defensible with 

respect to potential preferential pathways than Case A, although they have other limitations 

(Sections 2.1, 2.6 2.7, and 2.13). The complex design of the disposal units, in particular the 

FDCs, and the limited operating experience with novel materials warrants the inclusion of 

preferential pathways through the joints in a compliance case. As previously discussed, the 

PORFLOWTM files for Cases K, K1, and K2, appear to include minor fast pathways through the 

disposal unit floors (Section 2.5.3.2). However, the amount of chemical performance from the 

disposal unit concrete in these cases is inconsistent with significant flow through fast pathways 

though disposal unit joints. Because of the importance of potential fast pathways through the 

disposal units, the NRC staff will monitor the development of model support for the long-term 

performance of the disposal unit material interfaces. 

2.5.3.5 Disposal Unit Design and Performance Conclusion 

The performance of the disposal units can significantly affect the release of radionuclides from 

saltstone into the environment. In the 2005 TER (NRC, 2005), the NRC staff stated that 

degradation that may influence the hydraulic isolation capabilities of the saltstone and disposal 

units is a key factor in assessing whether the SDF can meet the performance objective in 

§61.41. The 2005 TER also stated that model support for the extent and frequency of fractures 

that will form in the disposal units was essential to assessing whether the saltstone disposal 

facility can meet the performance objective. 

With respect to the 2009 PA, the NRC staff concludes that the assumptions made in Case A, 

which DOE regards as its base case, for the performance of the disposal unit were not justified 

for the following reasons: 

• Case A included only minimal degradation of the FDC concrete and the Vault 1 and 4 

floors and roofs and did not account for all potentially relevant disposal unit degradation 

mechanisms; 

• Case A did not include fast pathways that could occur at the material interfaces (e.g., 

joints between the floor and walls, or between the wall panels) in the disposal unit or 

fractures in the disposal unit concrete; 

• Case A disposal unit performance assumptions were not consistent with the observed 

advective releases that occurred during saltstone placement in Vault 4. 
66 

The NRC staff finds that the modeling assumptions and parameters selected for the disposal 

unit concrete in Cases K, K1, and K2 are generally appropriate. However, as described above, 

while these cases are based on the assumption that the hydraulic conductivity of the disposal 

unit concrete degrades significantly, the chemical performance of the disposal units in these 

cases is inconsistent with any significant flow through fast pathways (e.g., flow through fractures 

in the concrete or joints). The NRC staff notes that DOE also performed a number of other 

sensitivity cases that included increased disposal unit degradation or the formation of fast paths 

out of the disposal unit. However, as described in other sections of this TER (Section 2.1, 2.6, 

2.7, 2.13), the NRC staff determined that other aspects of the modeling in these cases were not 

appropriate and limited the use of these cases in the NRC assessment of compliance with the 

performance objectives. 

DOE stated that PA maintenance activities include testing and research activities addressing 

(i) degradation mechanisms associated with cementitious hydraulic properties, (ii) moisture flow 

through fractures, and (iii) the extent and frequency of fractures in disposal units 

(SRR-CWDA-2011-00044). NRC staff expects that this research will be useful in providing 

model support for future revisions to the PA. 

NRC staff concludes that additional information is needed regarding the expected performance 

of the disposal units and has identified key monitoring factors related to the disposal unit 

performance. For example, because of the importance of the large difference in hydraulic 

conductivity between the disposal unit roofs and lower lateral drainage layer to disposal unit 

performance (Sections 2.7 and 2.13), the NRC will monitor model support for the long-term roof 

performance. Similarly, because of the uncertainty in the long-term behavior of HDPE and the 

importance of the HDPE/GCL layer overlying the FDC roofs (Sections 2.7 and 2.13), the NRC 

staff will monitor the development of information about HDPE and GCL performance. In 

addition, NRC staff thinks that additional model support related to the long-term performance of 

the non-cementitious materials used in the disposal units (e.g., epoxy and neoprene seals) may 

be a useful part of an assessment of the potential for fast paths out of the disposal unit. In 

particular, the NRC will monitor the development of model support for the long-term chemical 

performance of the disposal units (e.g., sorption of radionuclides in disposal unit fractures or 

joints). As DOE constructs additional disposal units, the NRC staff will monitor any changes to 

the FDC design. 

2.6 Waste Form 

2.6.1 Description of Waste Form 

Saltstone is a cementitious waste form made by mixing treated salt solutions from the SRS Fand H-Tank Farm liquid waste storage tanks with a dry mixture of blast furnace slag, fly ash, 

and cement. The treated salt solution is comprised mostly of sodium nitrate, sodium hydroxide, 

sodium nitrite, sodium aluminum hydroxide, sodium carbonate, and sodium sulfate. The solid 

components in the dry mixture are silicon dioxide, aluminum oxide, calcium oxide, magnesium 

oxide, and iron (III) oxide. Solidified saltstone forms a microporous matrix containing a salt 

solution, predominantly sodium, nitrate, and nitrite, within its pore structure. Chemically, 
67 

saltstone is alkaline and reducing. Clean grout, which is added to the disposal units to cap the 

saltstone, is made from process water and has a bulk composition similar to saltstone 

(Table 2.6-1). 

Table 2.6-1: Formula of Saltstone and Clean Grout

Ingredient 

Quantity (wt%) 

Saltstone Clean Grout 

Blast Furnace Slag (grade 100 or 120) 25 28 

Cement (ASTM C 150 Type II) or lime 3 6 

Fly Ash (Class F) 25 28 

Salt Solution (average 28% by weight salts) 47 N/A 

Water (maximum) N/A 38 

2009 PA; Table 4.2-7 

One of the key assumptions of Case A is that the hydraulic and physical properties of the waste 

form (Table 2.6-2) will essentially eliminate any advective flux of radionuclides out of the waste 

and limit releases to diffusion-dominated fluxes. In Case A, the dry porosity, bulk density, 

particle density, hydraulic conductivity, and effective diffusion coefficient in the waste form are 

assumed to remain constant for the entire 20,000 year analysis period. As discussed in 

Section 2.1, DOE used deterministic sensitivity analyses to test the effects of fracturing and 

increased bulk hydraulic conductivity and diffusivity. Case K represents the effects of fracturing 

on the flow of fluid through saltstone by modeling an equivalent increase in the bulk hydraulic 

conductivity14. 

Table 2.6-2: Saltstone Hydraulic and Physical Properties Used in Case A 

Property Value 

Hydraulic conductivity (cm/sec) 2.0×10−9

Effective diffusion coefficient (cm2/sec) 1.0×10−7

Porosity (%) 58.0 

Dry bulk density (g/cm3) 1.01 

Particle density (g/cm3) 2.40 

The properties were assumed to remain constant over time. 

Values taken from 2009 PA (Table 4.2-16)

DOE expects that, with a few notable exceptions, release of most contaminants from saltstone 

is controlled by sorption, which DOE represents with linear distribution coefficients (Kd values). 

 

14 Although Case K represents the effects of fracturing on saltstone hydraulic properties with changes in the bulk 

hydraulic conductivity, the effects on the fraction of saltstone oxidized is modeled by considering oxidation as 

proceeding from fracture faces. 
68 

DOE expects release of Tc to be limited by precipitation of Tc(IV) with sulfur supplied by the 

blast furnace slag in the dry premix. However, as discussed in greater detail in Section 2.7, 

DOE also represents Tc release with a sorption process. In addition, DOE indicated that Pu 

may be solubility limited in saltstone, but it represented Pu release in the PA by a sorption 

process. Kd values are element specific and, for some elements, change as the saltstone Eh

(oxidation potential or measure of reduction potential) and pH (measure of acidity or basicity) 

vary with time. 

The release of redox sensitive elements, most notably Tc, is strongly influenced by Eh

conditions. Tc is relatively insoluble and strongly sorbs under reducing conditions but is mobile 

under oxidized conditions. DOE included blast furnace slag in the saltstone dry mix to create 

reducing conditions that immobilize Tc-99. The use of blast furnace slag and the use of fly ash 

in the grout formulation also are known to have the added benefits of reducing heat evolution 

during hydration, decreasing hydraulic conductivity, and increasing strength and sulfate 

resistance. DOE measurements indicated slag has a specific reducing capacity of 820 μeq/g 

(SRNS-STI-2008-00045 and SRNL-STI-2009-00637) and solutions reacted with saltstone have 

redox potentials (Eh) as low as –585 mV (SRNL-STI-2010-00668). In addition, a field study 

conducted at SRS (Langton, 1988; WSRC–RP–92–1360) indicated that the addition of slag to 

saltstone essentially prevented Tc-99 leaching from a 210-liter (55-gallon) sample of saltstone 

for at least 2.5 years. The immobilization of Tc was ascribed to the formation of reducing 

conditions and the limited diffusion of oxygen within the core of the 210-liter (55-gallon) sample 

(Langton, 1988). The PA and supporting documents (e.g., SRNL-STI-2010-00668; also 

summarized in SRR-CWDA-2011-00044; SP-15 response) cite laboratory studies that suggest 

that Tc is in the reduced +4 oxidation state in saltstone and other cementitious materials 

containing blast furnace slag. However, the PA does not make an explicit assumption about the 

exact chemical forms of Tc in oxidizing or reducing saltstone. 

Chemical degradation of saltstone controls the chemical composition of pore fluids passing 

through the saltstone, and ultimately influences the release of contaminants from the waste 

form. In the PA, chemical degradation of saltstone is modeled by transitions from one defined 

age–redox state to another. The “age” of the saltstone is determined based on the pH of the 

pore fluid traversing through the material. A pH value between ~13.5 and 12.5 indicates 

Region I (young age), a pH of 11 or greater indicates Region II (middle age), and a lower pH 

level indicates Region III (old age) (Bradbury and Sarott, 1995). In the model, the saltstone is 

assigned an age–redox state based on the pH and Eh of the system, which transitions as the pH 

decreases (from Region II to Region III) and the Eh increases (from reducing to oxidizing 

conditions). In the PA, for modeling of all elements except Tc, changes in pH or oxidation state 

are treated as functions of the pore water volume that travels through the saltstone. As 

discussed in more detail in Section 2.7, the release model from Tc uses a shrinking-core model 

of saltstone oxidation. The pore water volume traveling through saltstone is dependent on the 

infiltration rate and the saltstone hydraulic conductivity. 
69 

2.6.2 Quality Assurance for Waste Form 

DOE does not directly discuss the QA program for the saltstone waste form in the PA. 

However, DOE has previously provided information on the QA for saltstone to the NRC during 

the monitoring process, particularly as part of addressing NRC Open Issues 2007-1 and 2007-2. 

A further discussion on DOE’s QA program for the saltstone waste form is located in 

Section 2.6.4.1. 

2.6.3 Modeling of Waste Form Degradation 

The Case A simulation did not consider physical degradation of saltstone, such as fracturing or 

increased hydraulic conductivity of the saltstone matrix, even though physical degradation may 

ultimately affect the rate of chemical degradation by influencing the rate of infiltration through 

the waste form. The Case A assumption that physical degradation of the saltstone will not 

occur during the performance period is based on a thermodynamic and mass balance analysis 

(WSRC-STI-2008-00236) that DOE characterized as preliminary research (NRC, 2008e). This 

work was based on a simulation, implemented with the Geochemist’s Workbench® geochemical 

code, of the formation of expansive mineral phases in saltstone due to exposure to rainwater or 

rainwater equilibrated with cement (WSRC-STI-2008-00236). The initial saltstone mineralogy 

assumed in the analysis comprised CSH, hydrotalcite, gibbsite, quartz, hematite, and gypsum, 

which is different from the mineralogy assumed in the analysis supporting the modeled Eh and 

pH transitions in saltstone (SRNL-TR-2008-00283). The author concluded that fracturing of 

saltstone is unlikely because expansive phase formation will cause a maximum 34% loss in 

porosity. However, this value is slightly higher than 30%, which is the median cement porosity 

for several cements that can be filled by expansive phases before fracturing can initiate (Tixier 

and Mobasher, 2003). Furthermore, the author acknowledged that the fraction of porosity that 

must be filled by expansive phases before fracturing can occur is uncertain 

(WSRC-STI-2008-00236). 

Case A also neglects other types of degradation, such as shrinkage cracking, steel corrosioninduced cracking, or dissolution of salts and low solubility matrix phases. However, alternative 

scenarios were considered in PA sensitivity analyses that evaluated the consequences of a 

hypothetical degraded saltstone condition. Case E is similar to Case A in that no fast flow paths 

are assumed to exist, but the saltstone is assumed to be severely degraded at time zero. In this 

scenario, the degraded saltstone was modeled by increasing the saltstone hydraulic 

conductivity to 1.7x10-3 cm/sec and modifying the moisture characteristic curves based on 

separate DOE analyses (SRNL-STI-2009-00115). In Case C and the Synergistic Case, 

fractures through the saltstone were included. 

Case K was constructed to provide additional understanding of the impacts of physical 

degradation of cementitious materials through time and other NRC concerns (Section 2.1). In 

Case K, complete physical degradation of saltstone was assumed to occur in 10,000 years, 

which was represented in the PA by increasing the saltstone saturated hydraulic conductivity 

and effective diffusivity with time. Case K applied a simplified, conservative assumption that 

saltstone saturation does not affect its hydraulic conductivity (i.e., the relative permeability is 
70 

always 1). With respect to oxidation of slag-bearing grout in Tc-99 transport modeling, Case K 

also assumed the saltstone grout will form fractures, with spacing that decreases to 0.1 m in 

10,000 years and a frequency that increases logarithmically with time. Values of saturated 

hydraulic conductivity and effective diffusivity assumed for Cases A, E, and K are listed in 

Table 2.6-3. 

Table 2.6-3: Saturated Hydraulic Conductivity and Effective Diffusivity Used in Selected 

Cases 

Case A Case E Case K 

Saturated 

Hydraulic 

Conductivity 

2.0x10-9 cm/sec; 

invariant with time 

1.7x10-3 cm/sec; 

invariant with time 

Initial value: 1.0x10-8 cm/sec; 

Value at 10,000 years: 1.0 

×10-6 cm/sec 

Effective 

Diffusivity 

1.0x10-7 cm2/sec; 

invariant with time 

1.0x10-7 cm2/sec; 

invariant with time 

Initial value: 1.0x10-7 cm2

/sec; 

Value at 10,000 years: 

5.0x10-6 cm2

/sec 

Values taken from 2009 PA (Table 4.2-16) and SRR-CWDA-2010-00044 

To determine the number of pore fluid volumes required to transition from one saltstone age–

redox state to another, DOE conducted thermodynamic simulations using the Geochemist’s 

Workbench® geochemical code (SRNL-TR-2008-00283). The simulations mimicked column 

experiments where at each simulation step a pore volume of infiltrate enters the column and 

equilibrates with the saltstone and an equal volume of reacted infiltrate is displaced and exits 

the column. The saltstone mineralogy that was used as input to the thermodynamic simulations 

was derived using a normative calculation that assigned chemical constituents determined by 

bulk chemical analysis to specific mineral phases. The derived saltstone mineralogy is 

comprised of calcium-silicate-hydrate (CSH), hydrotalcite, kaolinite, quartz, hematite, and 

gypsum. To represent the reducing component of slag, pyrrhotite was included in the model in 

an amount based on the measured saltstone reducing capacity15 (SRNS-STI-2008-00045). 

Three different infiltrating fluid compositions were modeled as reacting with saltstone. The first 

is groundwater unreacted with concrete, which represents the scenario with a fast flow path 

through the concrete in which unreacted groundwater contacts the saltstone. The second 

composition is groundwater equilibrated with an oxidized, Region III (old age) disposal unit 

concrete defined as “groundwater equilibrated with calcite.” The third infiltrating fluid 

composition was modeled as being equilibrated with an oxidized, Region II (middle age) future 

disposal cell concrete and defined as “groundwater equilibrated with CSH.” The groundwater 

composition used in the simulations is based on an analysis of a sample from a water table 

monitoring well in the vicinity of the SDF. The dissolved oxygen concentration in groundwater 

used in the simulations was 8 mg/L - higher than the 1.2 mg/L measured from the well water 

sample - based on assumed groundwater equilibrium with atmospheric oxygen. 

 

15The measured reducing capacity of simulated saltstone, which contained 23 wt% blast furnace slag, 

(SRNS-STI-2008-00045) was 822 μeq/g, which was approximately the same value as the reduction capacity Kaplan, 

et al. measured for 100% blast furnace slag. 
71 

The simulation results for Case A indicated that the Eh transitions from reducing to oxidizing 

conditions (from –0.45 to +0.6 V) at about the same number of pore volumes for the three 

infiltrate compositions (2,734, 2,775, and 2,806, respectively, for the first, second, and third 

infiltrating solution compositions). The pH transitions (from 11 to ~10) for the first and second 

fluid compositions occurred at 2,274 and 2,558 pore volumes, respectively. The simulation 

using the third infiltrating solution composition terminated before a pH transition occurred; a pH 

transition pore volume of 10,422 was derived based on extrapolation of the simulation results 

and an assumption that the pH transition occurs when the mass of CSH is completely 

consumed. The calculated fluid pore volumes for Eh transitions are similar for all three cases 

because the dissolved oxygen concentration in the three reacting fluids is the same, whereas 

the pH transitions are different because the pH values of the reacting fluids are different. The 

Case A results indicated that the saltstone remains in a reducing middle age state throughout 

the performance period. DOE considered an alternative scenario (Case K) in which the 

saltstone mineralogy has three-fourth less pyrrhotite and, therefore, three-fourth less reducing 

capacity than the saltstone considered in Case A16. The saltstone effective porosity used in 

determining the transition volumes for Case K was assumed to be 58%, which is based on 

measured values reported in SRNL-STI-2008-00421, compared to the value used in 

determining the transition volumes for Case A (42.3%) (SRNL-TR-2008-00283). Based on the 

lower reducing capacity and higher effective porosity of the saltstone compared to the Case A 

values, the pore volumes for the Eh and pH transitions used in the PA Case K scenario were 

decreased from Case A values to 505 and 7,608, respectively. The DOE analysis of Eh and pH 

transitions assumed that 100% of the minerals in the saltstone are available for reaction with 

infiltrating fluid. DOE acknowledged that if this assumption is not true, either because of 

fracturing or occlusion of portions of the cementitious materials from infiltrating fluid, the number 

of pore volumes required to reach Eh and pH transitions will be lower. In Case K, the fraction of 

Tc that is oxidized was determined based on the diffusion of oxygen into the monolith from 

fractures that form over time. 

2.6.4 NRC Evaluation – Waste Form 

2.6.4.1 NRC Evaluation – QA for Waste Form 

In October 2007, NRC conducted an onsite observation that addressed, in part, DOE’s quality 

assurance program related to the saltstone waste form (NRC, 2008a). As a result of that 

observation, the NRC staff concluded that certain aspects of DOE’s quality assurance program 

were satisfactory but that certain others should be tracked as open issues related to waste form 

variability. Specifically, in 2007, the NRC staff determined that DOE had a program for verifying 

that the dry bulk grout materials conform to applicable American Society for Testing and 

Materials (ASTM) standards and that the program is implemented effectively. The NRC staff did 

recommend that, because of the importance of reducing capacity of saltstone to achieving the 

 

16DOE reduced the saltstone reducing capacity in Case K because of an NRC staff concern regarding uncertainties in 

the saltstone reducing capacity. As stated in footnote-6, SRNS-STI-2008-00045 measured a reduction capacity of 

saltstone approximately equal to that of pure blast furnace slag, even though blast furnace slag comprised only 

approximately one-fourth of the saltstone mixture composition. 
72 

§61.41 performance objective, DOE should consider performing independent characterization of 

the slag upon receipt of the material rather than relying upon the vendor’s documentation. 

Although the quality assurance program for the composition of the dry bulk materials was 

determined to be implemented effectively, the NRC staff identified other aspects of the saltstone 

production and emplacement processes that may not be effectively controlled. Open issue 

2007-1 indicates that, as a result of variations in the composition of saltstone grout actually 

produced at the SRS SPF, DOE should determine the hydraulic and chemical properties of 

as-emplaced saltstone grout. Open issue 2007-2 specifies that DOE should demonstrate that 

intra-batch17 variability, flush water additions to freshly poured saltstone grout at the end of each 

production run, and additives used to ensure processability are not adversely affecting the 

hydraulic and chemical properties of the final saltstone grout. Open issue 2007-2 also indicates 

that DOE should show that the hydraulic and chemical properties of saltstone are consistent 

with the assumptions in the waste determination or show that any deviations are not significant 

with respect to demonstrating compliance with performance objectives. 

To address the first part of Open Issue 2007-2 (i.e., related to intra-batch variability), DOE 

studied the variability in the fractions of the main components in 3218 premix batches of the 

saltstone dry premix material (LWO-WSE-2009-00038). As part of its QA program, DOE 

controls the amount of fly ash, slag, and cement added to the saltstone dry premix material with 

“use every time” procedures. The saltstone dry premix material can be mixed with either 

automated or manual control. The 32 batches studied in 2009 (LWO-WSE-2009-00038) 

included 25 batches created with manual controls and 7 batches created with automated 

controls. There was no detectable difference between the batches created with automated and 

manual controls. In the PA, DOE reported that the 95% confidence interval for the fractional 

contribution of the main components extended ±3% from the nominal value for cement and ±2% 

of the nominal values for fly ash and blast furnace slag. Independent calculations that paired 

the ingredient masses with the total individual batch masses were essentially consistent with 

DOE’s results and yielded 95% confidence intervals that vary ±3% from the nominal fraction for 

cement but only ±1% from the nominal fractions for slag and fly ash. These values reflect small 

uncertainties in the mean fractional values. However, the 95% confidence interval reflects 

uncertainty in the mean rather than batch-to-batch variability. With a sufficient number of 

measurements, the uncertainty in the mean can be quite small even if the batch-to-batch 

variability is large. A more appropriate measurement of the batch-to-batch variability is the 

relative standard deviation of the fractional composition of the premix from batch to batch. 

Pairing the ingredient weights with the individual batch weights, which reduces measured 

variability in the fractional contribution of each of the ingredients, yields relative standard 

deviations of ±2% in blast furnace slag and fly ash and ±9% in cement. The study did not 

 

17 In this context, “intra-batch variability” refers to the variability within a lift of saltstone. A lift of saltstone is typically 

comprised of many of the individual premix batches described in the following paragraph. 

18 In the PA, DOE indicated that the saltstone batch variability study (LWO-WSE-2009-00038) included 31 batches. 

Although 31 batches appear to have been included in the summary statistics, data for 32 batches are provided in the 

data sheets included in the report. No reason was given for excluding the last batch, which did not appear to differ 

significantly from the other batches. Including the last batch did not affect the summary statistics significantly. 
73 

include statistics regarding any other aspects of Open Issue 2007-2, including the variability in 

the composition and mass of admixtures added to each batch or the salt waste to premix ratios. 

Like variations in the premix components, variations in the composition of salt waste may affect 

finished saltstone properties. To comply with South Carolina Hazardous Waste Management 

Regulations (SCHWMR), samples of saltstone made with samples of salt waste from the SPF 

feed tank (Tank 50) are checked quarterly with the EPA Toxic Characteristic Leaching 

Procedure (TCLP). Leachate taken with the TCLP is analyzed for the 8 listed toxic metals 

arsenic, barium, cadmium, chromium, mercury, lead, selenium, and silver. Because the 

unstabilized salt waste exceeds regulatory limits for several underlying hazardous constituents, 

the leachate also is tested for antimony, beryllium, nickel, and thallium. The TCLP results 

continue to show saltstone adequately retains the measured toxic metals and underlying 

hazardous constituents and is therefore a non-toxic, non-hazardous waste form in the context of 

the SCHWMR (e.g., SRNL-STI-2011-00262, SRNL-STI-2011-00561). 

Like the chemical properties, physical properties of finished saltstone also are affected by 

variations in salt waste composition. Caustic sludge washing prior to sludge vitrification is 

expected to increase the concentration of aluminate in salt waste treated in the SWPF as 

compared to salt waste treated with the ARP/MCU process (SRNL-STI-2009-00184). Near 

baseline levels (0.11 M for SWPF waste and 0.05 M for ARP/MCU waste), increased aluminate 

appears to decrease saltstone porosity, which would be expected to decrease the hydraulic 

conductivity, if the curing temperature is controlled at approximately 22oC (72oF) 

(SRNL-STI-2009-00184). However, at higher concentrations (0.45 M to 0.65 M), increased 

aluminate in simulated SWPF samples appears to decrease the Young’s modulus of cured 

saltstone, which is correlated with increased permeability (SRNL-STI-2009-00810). In addition, 

simulated ARP/MCU and SWPF samples with greater aluminate concentrations (i.e., greater 

than approximately 0.35 M in simulated SWPF samples and 0.22 M in simulated ARP/MCU 

simulated) appear to be more prone to cracking (SRNL-STI-2009-00184, 

SRNL-STI-2009-00546, SRNL-STI-2009-00810). Increasing the aluminate content of the salt 

waste also has been shown to increase the heat of hydration of saltstone 

(WSRC-STI-2007-00506 and SRNL-STI-2009-00546), which could increase curing 

temperatures if the curing temperature is not carefully controlled. Because curing temperature 

itself has a significant effect on saltstone properties, as discussed in greater detail later in this 

section, the net effect of increased aluminate concentrations is likely to depend in part on 

saltstone curing temperature control. This was confirmed by more recent research on the effect 

of aluminate on hydraulic conductivity (SRNL-STI-2011-00665) that found that a higher 

concentration of aluminate (0.28 M) reduced the hydraulic conductivity in samples that had the 

curing temperature controlled to 20oC or 40oC. 
74 

To begin to quantify the effects of these additional sources of variability, DOE studied the 

hydraulic and sorptive properties of laboratory-prepared saltstone samples prepared with 

different compositions and curing temperatures (SRNL-STI-2009-00419). Specifically, DOE 

examined the effects of admixtures (i.e., a set retarder and an anti-foam agent19), organic 

content, salt waste to premix ratio, aluminate concentration, and curing temperature on the 

porosity and saturated hydraulic conductivity of samples made with simulated ARP/MCU treated 

salt waste. Because of the small number of samples in each category (i.e., n = 3 or 6 for 

hydraulic conductivity and n=7 or 13 for porosity and density), the analysis of variance DOE 

performed had limited statistical power. In this study, DOE did not detect any effect of 

admixtures, organic content, or aluminate concentration. However, DOE did find statistically 

significant effects of salt waste to premix ratio and curing temperature. 

Specifically, DOE found that varying the salt waste to dry premix ratio from the nominal value of 

0.60 to 0.65 caused a six-fold increase in hydraulic conductivity (from 1.4x10-9 cm/s to 

8.4x10-9 cm/s) (SRNL-STI-2009-00419). This result was consistent with the results of previous 

studies that demonstrated an increase in saltstone porosity, which is expected to cause an 

increase in hydraulic conductivity, with increasing liquid waste to dry premix ratio 

(WSRC-STI-2007-00352). This effect may be important to long-term waste form performance 

because DOE allows water used to flush the saltstone emplacement lines to enter the disposal 

cells immediately after the saltstone is emplaced (i.e., before it is cured). As identified in NRC’s 

October 2007 monitoring report (NRC, 2008a), the extent to which this water mixes with the 

saltstone and affects the liquid to dry premix ratio is unknown. However, the results of DOE’s 

laboratory tests (SRNL-STI-2009-00419) indicate that even a small amount of mixing is likely to 

degrade the saltstone hydraulic conductivity. 

Curing temperature also appears to have a significant effect on the hydraulic properties of 

saltstone. Increased initial curing temperature from 22oC to 60oC (72oF to 140oF) appears to 

increase saltstone porosity (SRNL-STI-2009-00184) and decrease the Young’s modulus of 

simulated saltstone samples (SRNL-STI-2009-00810), both of which are correlated with 

increased hydraulic conductivity. When hydraulic conductivity was measured directly, DOE 

found that samples cured at 60oC (140oF) had an average hydraulic conductivity of 

8.0x10-7 cm/s, which is more than 500 times greater than similar samples prepared at 20oC 

(72oF) and 400 times greater than the base-case hydraulic conductivity assumed in the PA 

(SRNL-STI-2009-00419). DOE subsequently suggested that these results may have 

overrepresented the effects of curing temperature because the samples cured at 60oC (140oF) 

also were cured in a low-humidity environment. Grout cracking due to drying is well known 

(Pabalan et al, 2009). However, DOE previously found that saltstone samples cured at 90oC 

(190oF) in closed containers cracked during curing even though the samples did not dry 

(WSRC-TR-98-00337). Thus curing at 90oC (190oF) does appear to impact saltstone cracking 

 

19 DOE found that an anti-foam agent and set retarder were necessary to achieve the desired processability of 

saltstone made with H-canyon waste (WSRC-TR-2005-00149). A subsequent study confirmed the correlation 

between the fraction of H-canyon low-activity waste in the saltstone salt waste mixture and the need for an anti-foam 

agent and set retarder (SRNL-STI-2010-00522). The same study also found correlations between the fraction of 

Effluent Treatment Project waste and foam formation and General Purpose Evaporator waste and foam persistence 

(SRNL-STI-2010-00522). 
75 

through at least one mechanism in addition to drying. Although samples cured at 70oC (160oF) 

in a closed container were not visibly cracked, the effects on the hydraulic properties were not 

measured (WSRC-TR-98-00337), so it is unknown if the high temperature affected the hydraulic 

properties. Thus it is not certain that the elevated hydraulic conductivity DOE measured in 

samples cured at 60oC (140oF) in a low-humidity environment was entirely attributable to drying. 

Taken together, these results suggest the importance of good control of curing temperature to 

the hydraulic properties of as-emplaced saltstone. The results also suggest that laboratory 

samples will provide a good indication of the properties of field-scale saltstone only if the 

laboratory samples realistically reflect the curing temperature profile of field samples. 

Newer research considered the effect of variations in the curing temperature, water to cement 

ratio, aluminate concentration, and amount of fly ash on the hydraulic conductivity of simulated 

ARP/MCU samples (SRNL-STI-2011-00665). It was found that the cure temperature had the 

largest effect on the hydraulic conductivity. As described above, there was also some effect of 

the aluminate concentration on the hydraulic conductivity. The measured hydraulic conductivity 

values ranged from 2.9x10-9 to 4.9x10-8 cm/s for samples cured at 20oC, 1.0x10-9 to 

1.9x10-6 cm/s for samples cured at 40oC, and 6.2x10-8 to 1.3x10-6 cm/s for samples cured at 

60oC. The humidity was not controlled in these samples, so grout cracking due to drying could 

be contributing to the higher hydraulic conductivity in these samples. 

In October 2007, DOE indicated that thermocouples are used within the vault and saltstone to 

monitor curing temperatures (NRC, 2008a). At that time, the maximum observed temperature in 

saltstone during curing was approximately 50oC (120oF). Because of the potential significance 

of curing temperatures on the hydraulic properties of saltstone, the NRC staff inquired about the 

curing temperatures for saltstone grout during the April 26, 2011, monitoring observation 

(NRC, 2011k). Specifically, the NRC staff asked about the potential impact of high-aluminate 

salt waste on curing temperatures. DOE stated that cure temperature profiles for saltstone are 

being compiled. As indicated in the summary of the NRC staff’s April 2011 monitoring 

observation (NRC, 2011k), NRC staff will review the cure temperature profiles for saltstone 

when DOE compiles them following future testing. 

To begin to understand potential differences between the properties of laboratory-prepared and 

field-emplaced saltstone (Open Issue 2007-1), DOE collected nine core samples from Vault 4, 

Cell E in September 2008. Of the nine samples collected from Vault 4 in September 2008, one 

was used for bulk density and porosity measurements and three were used for chemical phase 

measurements (SRL-STI-2009-00804). DOE reported that the measured bulk density after 

saturation with simulated salt solution was 1.9 g/cm3

 and the porosity was 0.599 

(SRL-STI-2009-00804). X-ray diffraction measurements of three of the samples showed 

chemical phases typical of fly ash and calcium silicate hydrate (CSH). Based on these results, 

DOE concluded that the Vault 4 cored samples were consistent with laboratory-prepared 

samples (SRL-STI-2009-00804). However, permeability testing of five of the Vault 4 samples 

collected in September of 2008 showed that the saturated hydraulic conductivity of the Vault 4 

samples was significantly greater than the hydraulic conductivity of laboratory-prepared samples 

(SRNL-STI-2010-00657). Specifically, the hydraulic conductivity of the field samples had an 

arithmetic mean of 4.0x10-7 ± 1.9x10-7 cm/s (mean ± s.e.) and a median of 1.8x10-7 cm/s, which 
76 

is approximately two orders of magnitude greater than the saturated hydraulic conductivity DOE 

used in its base-case PA model (i.e. 2.0x10-9 cm/s) and 18 times greater than the initial 

hydraulic conductivity DOE used in Case K (i.e., 1.0x10-8 cm/s). 

In response to NRC’s second RAI, DOE hypothesized that the relatively high hydraulic 

conductivity of the core samples taken from Vault 4 Cell E is an artifact of the sample collection 

method (SRR-CWDA-2011-00044; page 177). However, samples taken from laboratoryprepared blocks of simulated saltstone with the same coring method that was used to collect 

samples from Vault 4 had hydraulic conductivities similar to the hydraulic conductivities of 

laboratory samples prepared in molds (i.e., average hydraulic conductivity of 4.1x10-9 cm/s for 

cored samples from laboratory-prepared blocks and an average of 5.4x10-9 cm/s for samples 

made in molds) (SRNL-STI-2010-00657). Thus the relatively high hydraulic conductivity of 

Vault 4 samples may not be attributable to the coring method, unless the effects of coring in the 

field and laboratory are significantly different despite the same coring method being used. 

In its PA, DOE identified Open Issues 2007-1 and 2007-2 related to the variability of asemplaced Saltstone as areas of ongoing and future work. Because DOE hypothesizes that 

coring samples of saltstone in the field artificially increases the hydraulic conductivity of the 

samples, DOE is developing a formed-core sampling technique. The formed-core technique 

entails placing sampling tubes in the saltstone vaults prior to saltstone grout pours, and 

removing the sampling tubes (with samples) after the saltstone has cured 

(SRNL-STI-2010-00167). Hydraulic conductivity tests of these samples may provide 

information relevant to Open Issue 2007-1, related to the hydraulic properties of field-emplaced 

saltstone. 

2.6.4.2 NRC Evaluation – Modeling of Physical and Hydraulic Properties of Waste 

Form 

The NRC staff reviewed the DOE analysis of the initial properties of saltstone and waste form 

degradation, as described in PA Section 4.2 and supporting documents, the information DOE 

provided in its responses to NRC RAIs, as well as literature information. The values of hydraulic 

conductivity, porosity, dry bulk density, and particle density assumed in Case A are based on 

laboratory measurements on saltstone simulants (SRNL-STI-2008-00421). The effective 

diffusion coefficient is based on a value for concrete recommended in WSRC-STI-2006-00198. 

The PA values assumed for the saltstone porosity, bulk density, and particle density are 

reasonable based on comparison with measured values (Table 2.6-4). 

The modeled hydraulic conductivity of saltstone in Case A (i.e., 2x10-9 cm/s) is generally in the 

range of the measured hydraulic conductivity of laboratory-prepared samples cured at low 

temperatures (Table 2.6-5). However, the hydraulic conductivity measured using groundwater 

equilibrated with vault concrete simulant as the permeating fluid instead of simulated saltstone 

pore fluid was almost an order of magnitude higher (1.5x10-8 cm/s) (WSRC-STI-2007-00649). 

The fluid flowing into and through the saltstone is expected to be more consistent with the 

groundwater equilibrated with vault concrete than simulated saltstone pore fluid, so this result 

may be more applicable than the hydraulic conductivity measured in other samples with 
77 

simulated saltstone pore fluid. The hydraulic conductivity values reported in 

SRNL-STI-2011-00665 for simulated ARP/MCU saltstone cured at 20oC (2.9x10-9 cm/s to 

4.9x10-8 cm/s) were also higher than the Case A value. As noted in Section 2.6.4.1, samples 

cured at elevated temperatures had measured hydraulic conductivity values as high as 

1.9x10−6

 cm/s (SRNL-STI-2011-00665), though this may be due to drying cracking in the 

laboratory. Additionally, the hydraulic conductivity of 2x10-9 cm/s assumed in Case A appears 

to be optimistic as compared to the measured hydraulic conductivity of cores taken from Vault 4 

(Table 2.6-5). As previously discussed, the hydraulic conductivity of the field samples had an 

arithmetic mean of 4.0x10-7 cm/s ± 1.9x10-7 cm/s (mean ± s.e.) and a median of 1.8x10-7 cm/s, 

which is approximately two orders of magnitude greater than the saturated hydraulic 

conductivity DOE used in its base-case PA model (i.e., 2.0x10-9 cm/s) and 18 times greater than 

the initial hydraulic conductivity DOE used in Case K (i.e., 1.0x10-8 cm/s). DOE has 

hypothesized that the measured values from Vault 4 samples are not representative of the 

actual hydraulic conductivity of saltstone in Vault 4 and instead represent an artifact of the 

coring method used. This hypothesis is not supported by the results of experiments that show 

samples cored from laboratory-prepared blocks using the same coring technique had measured 

hydraulic conductivity values similar to laboratory-prepared samples prepared in molds (i.e., not 

cored) (SRNL-STI-2010-00657). These results suggest the difference in the measured 

hydraulic conductivity of cored laboratory-prepared samples and cored field-emplaced samples 

may be attributable to the difference between laboratory and field saltstone preparation and 

curing rather than the coring method. Potential mechanisms for these differences include 

variations in curing temperature and the liquid to premix ratio in the field, as well as 

heterogeneity in physical properties at the field scale that is not captured in laboratory-scale 

samples. Alternately, the difference may be attributable to some difference between field and 

laboratory sampling even when the same sampling method is used (Section 2.6.4.1). If fieldprepared saltstone has a significantly greater saturated hydraulic conductivity than laboratoryprepared saltstone, and, consequently, a significantly greater saturated hydraulic conductivity 

than DOE used in Case A, Case A would likely significantly under-represent radionuclide 

release from the saltstone. Although the measured hydraulic conductivity in core samples also 

exceeds the initial value of hydraulic conductivity used in Case K, the increase in hydraulic 

conductivity with time used in Case K is expected to capture an appropriate range of hydraulic 

conductivities. 

The effective diffusivity value of 1x10-7 cm2

/s assumed by DOE in Case A was based on 

information in WSRC-STI-2006-00198. In this document, a literature review of effective 

diffusivity values reported values that ranged from 1.44x10-8 cm2

/s to 4x10-7 cm2/s. Based on 

this data, the authors concluded that 1x10-7 cm2

/s was a representative value for the effective 

diffusivity in ordinary quality concrete. In this same report, research on the effective diffusivity of 

simulated saltstone, which was prepared with a slightly different formulation than is currently 

used, was also summarized. These effective diffusivity values ranged from 8x10-10 cm2

/s to 

9x10-9 cm2

/s. A more recent study of the effective diffusion coefficient in simulated saltstone 

reported values less than 1x10-8 cm2

/s (SRNL-STI-2010-00515). Based on this information, 

NRC staff finds that the effective diffusion coefficient used by DOE is reasonable. However, 

NRC staff notes that research has not been performed on the effective diffusivity in as-
78 

emplaced saltstone, and, as noted above in the discussion on hydraulic conductivity, the 

physical properties of as-emplaced saltstone may differ significantly from lab-prepared 

simulated saltstone. 

Table 2.6-4: Modeled and Measured Porosity (η), Bulk Density (ρb), and Particle Density 

(ρp) Values for Saltstone 

Modeled Values Notes 

All Cases η: 58% 

ρb: 1.01 g/cm3

ρp: 2.40 g/cm3

Assumed invariant with time 

Measured Values Notes

SRNL-STI-2008-00421; 

Table 31 

η: 54 to 57% 

ρb: 1.04to 1.08 g/cm3

ρp: 2.32 to 2.48 g/cm3

Range of values measured for 

DDA saltstone cured for 28 or 90 

days1

η: 58 to 61% 

ρb: 0.95 to 1.01 g/cm3

ρp: 2.35 to 2.49 g/cm3

Range of values measured for 

ARP/MCU saltstone cured for 28 

or 90 days 

η: 56 to 60% 

ρb: 1.00 to 1.05 g/cm3

ρp: 2.35 to 2.53 g/cm3

Range of values measured for 

SWPF saltstone cured for 28 or 

90 days 

Pabalan, et al. (2011) η: 52 and 67% 

ρp: 2.47 g/cm3

Porosity of crushed and sieved 

DDA saltstone simulant used in 

two column leaching experiments

Crushed and sieved DDA 

saltstone simulant 

1

 DDA = Deliquification, Dissolution, and Adjustment salt simulant; ARP/MCU = Actinide Removal 

Process/Modular Caustic Side Solvent Extraction Unit salt simulant; SWPF = Salt Waste Processing 

Facility salt stimulant 

The NRC staff believes that the Case A assumption that saltstone will be hydraulically 

undegraded for 20,000 years is unrealistically optimistic, and that this assumption is inconsistent 

with observations of existing cracks in Vault 4 saltstone (SRNL-ESB-2008-00017, 

SRR-CWDA-2011-00105). In the response to NRC comment SP-2 (SRR-CWDA-2011-00044), 

DOE stated that they do not believe that there is evidence of these cracks extending below the 

surface of the monolith. However, in SRR-CWDA-2011-00105, it is noted that the depth and 

thickness of these cracks is not known. Therefore NRC staff believes that there is insufficient 

evidence that these cracks do not extend below the surface. Furthermore, in the PA, DOE 

considered only sulfate attack and did not provide an adequate basis for neglecting other types 

of degradation such as shrinkage cracking, steel corrosion-induced cracking, cracking due to 

settlement or earthquakes, and dissolution of salts and low solubility matrix phases. Of these 

mechanisms, DOE estimates of fracturing were found only for static and dynamic settlement of 

Vault 4 saltstone (T-CLC-Z-00006). The settlement analysis assumed static settlement caused 
79 

cracking in saltstone along Vault 4 construction joints (at 9.1 m [30 ft] intervals) and between the 

saltstone and vault walls. Dynamic settlement was assumed to cause cracking at 15.2 m (50 ft) 

intervals. The analysis calculated increases in crack length and width as a function of time, 

assuming 3 earthquakes occurred at random times between 100 and 10,000 years. 

The NRC staff also is concerned about the DOE conclusion that saltstone fracturing by 

expansive phases due to sulfate attack is unlikely. The conclusion is based on geochemical 

modeling results that are unsupported by comparisons with empirical data or observations. 

Also, the geochemical equilibrium model ignored reaction kinetics that could result in metastable 

product formation often associated with an increase in volume. The effects of organic additives 

or pozzolanic replacement on the dissolution and precipitation of cement-related compounds, 

which may have an effect on the generation of expansive phases, also were not considered. 

NRC issued comment SP-1 (NRC, 2010b, i) asking for additional justification for this DOE 

conclusion. In its response to NRC comment SP-1 (SRR-CWDA-2011-00044), DOE stated that 

research into saltstone material degradation is planned or ongoing. NRC staff still has concerns 

associated with saltstone fracturing by expansion and suggests future research could 

(i) consider the effect that sulfide from the blast furnace slag might have on the phases and 

reactions present in this system, (ii) include experiments that are designed to collect data on 

initial mineralogical conditions, fundamental thermodynamic data and reaction kinetics, and 

(iii) consider expansive phases produced by intermediate or metastable reaction products.

The NRC staff understands fracturing was addressed non-mechanistically in the PA sensitivity 

analyses by Case C and the Synergistic Case, which consider fractured saltstone, and by 

Cases B and D, which postulate a gap between the saltstone and disposal unit wall. However, 

NRC believes these cases contain other assumptions, which are not representative of the 

expected performance of the saltstone (e.g., use of overly-optimistic moisture characteristic 

curves and potentially overly-optimistic saltstone hydraulic conductivity). In NRC comment PA-8 

(NRC, 2010i), NRC staff expressed concern with DOE’s use of Case A as the base case. In its 

response to NRC concerns (SRR-CWDA-2011-00044; PA-8), DOE maintained that Case A is 

valid, however, DOE proposed an alternative scenario (Case K) in which saltstone grout will 

develop fractures with spacing that decreases and frequency that increases with time such that 

within 10,000 years after closure the saltstone has a hydraulic conductivity and diffusivity similar 

to soil. The Case K final fracture spacing at 10,000 years after closure is one thoroughgoing 

fracture every 10 cm (4 in). The semi-log fracture growth relationship used in Case K results in 

minimal cracking at early times followed by a rapid increase in the number of fractures between 

8,000 and 10,000 years after site closure. There is significant uncertainty in the rate and extent 

of fracturing because of the lack of experience with engineered sub-surface cementitious 

materials several thousand years old or applicable natural analogs. The NRC staff believes that 

a modeling approach in which the rate of fracturing increases with time is reasonable because, 

as the saltstone degrades, the increase in fractures would lead to increased flow through the 

monolith, which could then accelerate the formation of new fractures. However, a number of 

alternate fracture growth curves other that the semi-log model DOE used could be valid. These 

models could result in a different rate of fracture formation and, potentially, earlier significant 

fracturing. If earlier fracturing occurred, the dose peaks would move to earlier times. The NRC 
80 

staff evaluated the model sensitivity to the assumed rate of fracture growth, as discussed in 

Section 2.13. 

Table 2.6-5: Modeled and Measured Initial Hydraulic Conductivity Values for Saltstone 

Hydraulic 

Conductivity 

(cm/sec) 

Notes 

Modeled Values

Case A 2.0x10-9 Assumed invariant with time 

Case E 1.7x10-3 Assumed invariant with time 

Case K 1.0x10-8 Value at 10,000 years: 1.0x10-6 cm/sec 

Measured Values – Lab Prepared Samples

WSRC-STI-2007-00056 1.4x10-9 to 

3.4x10-9

MCU saltstone; measured using a beambending technique 

SRNL-STI-2008-00421; 

Table 30 

2.5x10-9 to 

9.3x10-9

Range of logarithmic averages of hydraulic 

conductivities measured for DDA, ARP/MCU, 

and SWPF saltstone; 28-day curing period1

9.6x10-11 to 

6.0x10-9

Range of logarithmic averages of hydraulic 

conductivities measured for DDA, ARP/MCU, 

and SWPF saltstone; 90-day curing period 

WSRC-STI-2007-00649 

1.5x10-8

MCU saltstone; groundwater equilibrated with 

vault concrete simulant was used as 

permeating fluid 

5.3x10-9 MCU saltstone; saltstone pore fluid simulant 

was used as permeating fluid 

SRNL-STI-2009-00419 

1.7x10-10 to

9.9x10-9

Range of hydraulic conductivities measured 

for various formulations of simulated saltstone

8.0x10-7 Sample cured at 60oC at low relative humidity 

SRNL-STI-2011-00665 

2.9x10-9 to 

4.9x10-8

ARP/MCU saltstone cured at 20oC without 

control of humidity; permeating fluid is low 

aluminate salt solution 2

1.0x10-9 to 

1.9x10-6

ARP/MCU saltstone cured at 40oC without 

control of humidity; permeating fluid is low 

aluminate salt solution 2

6.2x10-8 to 

1.3x10-6

ARP/MCU saltstone cured at 60oC without 

control of humidity; permeating fluid is low 

aluminate salt solution 2

Measured Values – Saltstone Core Samples 

SRNL-STI-2010-00657 3.9x10-7 Arithmetic average of 10 measurements of 5 

cores taken from Vault 4, Cell E 

1 DDA = Deliquification, Dissolution, and Adjustment salt simulant; ARP/MCU = Actinide Removal 

Process/Modular Caustic Side Solvent Extraction Unit salt simulant; SWPF = Salt Waste Processing 

Facility salt stimulant 

2

 Samples prepared using a range of water to cement ratios, aluminate concentration, and fly ash
81 

2.6.4.3 NRC Evaluation – Modeling of Chemical Properties of Waste Form 

The NRC staff is concerned that the DOE conclusion the saltstone will remain reducing and 

middle-aged through the performance period relies on the results of geochemical equilibrium 

modeling (SRNL-TR-2008-00283) that lacks adequate model support. NRC staff is concerned 

that the Eh transition to oxidizing conditions may occur sooner than DOE derived from its 

geochemical modeling. The model results have not been validated by comparison to 

experimental or other data independent of model calculations. In its response to NRC comment 

SP-12 (SRR-CWDA-2010-00033), DOE acknowledged that its geochemical modeling results 

lack experimental verification and that source term model support is one area of future DOE 

work, potentially including physical testing to address issues that were evaluated only through 

Geochemist’s Workbench® simulations. DOE also indicated in its response to NRC comment 

SP-15 (SRR-CWDA-2011-00044) that future work on saltstone degradation will include focus on 

the timing of saltstone transitions from reducing to oxidizing state. NRC staff is supportive of 

additional work to provide model support (e.g., pH and Eh measurements in accelerated 

physical testing using higher flow rates than anticipated in full-scale saltstone). 

Depending on which initial mineralogy is more appropriate, a different conclusion could be 

reached regarding the likelihood of expansive phase formation or the calculated pore volumes 

for Eh and pH transitions. In its second response to NRC comment SP-8 

(SRR-CWDA-2011-00044), DOE provided additional information on Eh and pH transitions times 

and variability in grout mineralogy. The Geochemist’s Workbench® code was shown to use very 

similar saltstone mineralogies in reaction path calculations even if the input mineralogies are 

different because the code seeks an initial equilibrium assemblage prior to proceeding with the 

reaction path calculation. In addition to differences in saltstone formulations, differences in bulk 

density and porosity also can lead to differences in the predicted initial mineralogy. Additional 

Geochemist’s Workbench® calculations, using two saltstone compositions with similar formulae 

but different bulk densities and porosities, showed that the pore volumes to reach the major Eh

transition was 39% different between the two mineralogies and the pore volumes to reach the 

major pH transition was 42% different 

The geochemical modeling of Eh transitions assumed all the reducing capacity in the saltstone is 

available for reaction with the infiltrate. Also, the reducing capacity of saltstone used in the 

model is based on laboratory measurements using slag samples that were finely ground to 

increase the reactive surface area (SRNS-STI-2008-00045). The NRC staff is concerned that 

these DOE measurements of saltstone reducing capacity (SRNS-STI-2008-00045) indicate the 

reducing capacity of saltstone is equivalent to the reducing capacity measured for pure blast 

furnace slag, even though the saltstone sample only contained 23 weight percent (wt%) blast 

furnace slag. The NRC staff also is concerned that, in actual field conditions, only a fraction of 

the slag will be accessible for reaction with the infiltrate and that the reactive surface area and 

reducing capacity of saltstone emplaced in the field are likely to be much smaller than that of 

finely ground laboratory samples. Thus, the longevity of reducing chemical conditions and the 

timing of release of redox-sensitive elements such as Tc, Pu, and Np, may be overestimated in 

the PA. In response to NRC comment PA-8 (SRR-CWDA-2011-00044) regarding the saltstone 

reducing capacity and Eh transition times assumed in the PA, DOE conducted additional 
82 

analysis (Case K) in which the saltstone reducing capacity was reduced by a factor of four to 

account for the saltstone composition (25% blast furnace slag) and the saltstone porosity was 

changed from 42.3% to 58% to correct the porosity used in the initial Eh and pH transition 

estimate (SRNL-TR-2008-00283) to the measured porosity used in the PA. Based on the lower 

reducing capacity and higher effective porosity of the saltstone compared to the Case A values, 

the Eh transition used in the PA Case K scenario was decreased from the Case A value of 2,806 

to 505 pore volumes. The NRC staff believes the decreased Eh transition time in Case K more 

reasonably reflects the expected evolution of the saltstone chemical environment. 

The DOE conclusion that Tc release will be mitigated by the addition of blast furnace slag to the 

saltstone mixture appears to be supported by data from DOE studies (Langton, 1988) and other 

published literature (Brodda, 1988; Gilliam, et al., 1990; Aloy, et al., 2007). Leaching 

experiments indicated that blast furnace slag cement spiked with Tc-99 retained 99.9 percent of 

the Tc after 500 days of leaching in a simulated medium-level waste solution (Brodda, 1988), 

while others showed Tc leachability decreased by several orders of magnitude by the addition of 

blast furnace slag to the grout (Gilliam, et al., 1990). The reduction in Tc leachability in the 

presence of blast furnace slag has been attributed to the formation of relatively insoluble TcO2

(Langton, 1988), Tc3S10 (Lukens, et al., 2005), or Tc2S7 (Liu, et al., 2007). The Tc sulfide solids 

are considered to have formed by reaction of Tc with sulfide species released into solution by 

slag hydration. 

However, as described in more detail below, other experiments have shown that a fraction of Tc 

may remain oxidized in saltstone formulations containing blast furnace slag. Thus, the NRC 

staff is concerned about the ability of saltstone to effectively reduce Tc, which is initially in the 

highly mobile Tc(VII) oxidation state, to its low solubility and highly sorptive Tc(IV) oxidation 

state. NRC staff has been monitoring information regarding Tc reduction in saltstone as part of 

Open Issue 2009-1 (NRC, 2009a). A DOE study (SRNS-STI-2008-00045) indicated that Tc 

apparently was not reduced in the experiments and the measured Tc Kd values (6.5 and 

13.0 mL/g) were significantly less than the value (1,000 mL/g) assumed for Tc(IV) species and 

used in the PA. Although DOE recently measured higher Kd values (711 mL/g after 22 days of 

equilibrium and 581 mL/g after 56 days) (SRNL-STI-2010-00668), H2(g), a strong reductant that 

could contribute to Tc reduction, was present in the experimental system. Because H2(g) is not 

expected to be representative of SRS field conditions, any reduction attributable to H2(g) would 

be an experimental artifact. In a different system, DOE measured a relatively low Tc Kd of 

139 mL/g under reducing conditions using an actual Vault 4 saltstone sample 

(SRNL-STI-2010-00667). DOE attributed the low Kd to the saltstone sample not being fully 

reduced because of the presence of low concentrations of atmospheric oxygen (30 ppm to 

60 ppm). In SRNL-STI-2010-00668, the authors also hypothesize that the sample color, which 

was olive green as opposed to black, indicated that the sample had become partially oxidized. 

The NRC staff believes that the color of the saltstone is insufficient evidence regarding the 

redox state of Tc because species other than Tc are likely responsible for the color change. 

Given that the saltstone sample was taken from Vault 4 nine months after emplacement, and 

the Tc was not fully reduced, the DOE result reinforces the NRC staff concern regarding the 

degree and the rate of Tc reduction by saltstone. 
83 

Figure 2.6-1: Eh–pH Diagram Showing Tc Speciation20

Tc is predicted to form a variety of different solids with sulfur depending on the system Eh and 

pH (Figure 2.6-1). In this diagram, the yellow region represents the stability field for Tc(IV) (i.e., 

the form of Tc that has low solubility and high sorption), and the blue region represents the 

stability field for Tc(VII). As noted in Section 2.6.1, DOE’s measurements of initial reducing 

conditions were also as low as -585 mV, which is in the region of stability for Tc(IV). 

Additionally, Center for Nuclear Waste Regulatory Analyses (CNWRA) column experiments 

using simulated saltstone determined initial Eh values as low as -460 mV and a pH of 

approximately 12, which is well within the stability field for Tc(IV) (Figure 2.6-1). However, after 

approximately 30 pore volumes flowed through the column, the Eh and pH values for the 

simulated saltstone transitioned to the stability region for the highly mobile Tc(VII) (TcO4

-

). The 

applicability of these column experiments to the expected performance of the as-emplaced 

saltstone is described in greater detail in a subsequent paragraph. 

Several studies of the redox state and speciation of Tc have shown the transformation of 

oxidized Tc(VII) to reduced Tc(IV) in blast furnace slag-containing grouts to be incomplete. 

Allen, et al. (1997) observed that, in blast furnace slag-containing cement mixtures with 

simulated radioactive waste solutions, Tc(VII)O4

−

 indicators persisted in XANES and EXAFS 

X-ray absorption spectra. Thirty to 90 percent of the Tc remained in the oxidized state even 

 

20 Calculated using Geochemist’s Workbench® ACT2. (TcO4

−

 Activity =10−8; SO4

2−

 Activity = 10−3). Diamond 

symbols are experimental Eh and pH values for simulated saltstone measured by CNWRA. Yellow areas indicate 

solid phase stability fields and blue areas indicate aqueous phase stability fields. 
84 

after one month (Allen, et al., 1997, Figure 2). The Tc(VII)O4

−

 was eliminated only when Na2S 

or FeS was added to the mixtures instead of blast furnace slag. The work of Lukens, et al. 

(2005) suggests that Tc reduction by blast furnace slag is chiefly due to reaction with sulfide, 

leading to precipitation of a Tc(IV) sulfide with the likely formula Tc3S10. Lukens, et al. (2005) 

prepared some of their Tc-doped grout samples with added Na2S to ensure complete Tc 

reduction; the Tc in these samples was initially entirely Tc(IV) sulfide, according to EXAFS data. 

This pretreatment is not consistent with expected conditions for emplaced blast furnace slagcontaining saltstone. Lukens, et al. (2005) prepared three additional samples using a Na2S 

solution that had become oxidized prior to the experiment. The specific sulfur species present 

in that solution is unknown. However, the authors estimated that the solution contained SO3

2- 

(0.1 M) and S2O3

2- (1.4 M). In SRNL-STI-2010-00668, it was stated that the reductant in these 

samples can not be the Na2S solution. However, the NRC staff believes that it is possible that 

the solution may have still contained some Na2S that could have reduced some of the Tc. 

Additionally, Lukens, et al. (2005) note that SO3

2- can react with Tc to form TcO2•2H2O. In these 

samples, approximately 40 percent of the Tc was initially in the form of TcO2•2H2O, indicating 

that the SO3

2- may have reacted with the Tc. Also, in these samples, 16 to 20 percent of the Tc 

was initially in the form of the oxidized pertechnetate ion (TcO4

-

) (Lukens, et al., 2005, Figure 4 

upper right; details of these results may be found in Shuh, et al., 2003). Even though there is 

some uncertainty in the effect of the presence of the oxidized Na2S solution on the reduction of 

Tc in these samples, the results show that blast furnace slag alone is not sufficient to ensure full 

Tc reduction. 

As pointed out by Kaplan, et al. (2011), in discussing the results found in Lukens, et al. (2005), 

the proportion of Tc that was in the oxidized form of TcO4

-

 subsequently decreased over time 

when the samples prepared using the oxidized Na2S solution were isolated from air. Lukens, et 

al. (2005) reported that TcO4

-

 fell below 10 percent of total Tc in 18 months in all three samples 

and down to 3 percent in 30 months in the one sample that was not exposed to air. Lukens, et 

al. attributed this reduction to the blast furnace slag in the grout mixture, though NRC staff 

believes the presence of the oxidized Na2S solution creates some uncertainty in this conclusion. 

The TcO4

- was not observed to be completely reduced in these samples. This is an important 

point because even a small amount of mobile Tc under nominally reducing conditions could lead 

to higher Tc releases than DOE has modeled. Additionally, although the length of time found by 

Lukens, et al (2005) for Tc to become 97% reduced (i.e., 30 months) is small compared to the 

analysis period in the PA (i.e., 20,000 years), the initial presence of oxidized Tc could result in 

larger releases at early times than predicted. It also should be noted that these samples were 

not exposed to air, while actual saltstone is not isolated from the environment. The extent of 

reduction may therefore be less in the as-emplaced saltstone. 

The NRC staff concludes that, while blast furnace slag certainly helps reduce Tc release by 

immobilizing it as Tc(IV) solids, it is not clear that all Tc will be initially in the reduced state in 

emplaced saltstone. There appears to be considerable uncertainty in the quantity of oxidized Tc 

potentially present in saltstone and other blast furnace slag-containing grouts with no additional 

reductants (see discussion in this section of Allen, et al., 1997, Lukens, et al., 2005). In 

addition, there remains uncertainty not only about the thermodynamic characteristics of the 
85 

Lukens, et al. (2005) proposed Tc(IV) sulfide, but also the relative quantities of different Tc(IV) 

solids in fresh saltstone. 

DOE also acknowledged that only a small concentration of O2(g) (e.g., 30 to 60 ppm) is needed 

to oxidize Tc(IV) to Tc(VII) (SRNL-STI-2010-00667). The oxidation rate will depend on the 

groundwater infiltration rate, the hydraulic properties of the saltstone, the ability of blast furnace 

slag to consume oxygen, and the kinetics of Tc oxidation. For example, bench-scale column 

experiments on Tc leaching from simulated saltstone performed at CNWRA indicated the 

system transitioned from reducing to oxidizing conditions after only tens of fluid pore volumes 

have passed through the column, which resulted in an enhanced release of Tc from the 

simulated saltstone. Although the bench-scale tests do not completely represent field 

conditions because of the relatively high flow rate through the system, the high reactive surface 

area of the crushed saltstone simulant, and the interaction of the particles with atmospheric 

oxygen during grinding, the results still indicate that the DOE model may be overestimating the 

length of time required for the Eh transition to oxidizing conditions and the model may be 

underestimating release at earlier times. Additional research that more closely represents 

actual conditions is needed to reduce the uncertainty in the release of Tc-99 from the saltstone 

waste form. 

2.6.4.4 Waste Form Conclusions 

The NRC staff concludes that the representation of the waste form in Case A is not realistic and 

does not represent current conditions or expected future degradation. Of the cases analyzed by 

DOE, NRC staff considers the modeled initial conditions and degradation of the waste form in 

Cases K, K1, and K2 to best reflect the current and future expected conditions of the waste form 

based on currently available data. However, NRC staff recognizes that there is significant 

uncertainty in the rate and extent of fracturing and there is considerable uncertainty in some of 

the key parameters related to the waste form performance. For example, little is known 

regarding the potential performance of the actual as-emplaced waste form. DOE performed 

research on some of these parameters since the 2006 Waste Determination 

(DOE-WD-2005-001) was issued. Additionally, in Section 8.2 of the PA, and in the response to 

NRC comment PA-7 (SRR-CWDA-2011-00044), DOE described their planned future research 

in these areas. NRC staff views this research as extremely useful and encourages continued 

research in these areas to ensure that the estimated dose does not under-predict the actual 

expected dose. 
86 

NRC staff concludes that additional information is needed on the performance of the asemplaced saltstone waste form. Specific areas in which continued research is needed include 

the: 

• hydraulic conductivity of as-emplaced saltstone 

• potential variability of as-emplaced saltstone properties with variations in the 

composition (e.g., water to cement ratio, presence of aluminate, changes to admixtures) 

• applicability of measured hydraulic properties (e.g., hydraulic conductivity, diffusivity) for 

laboratory-prepared samples to field-scale, as-emplaced saltstone (e.g., effects of scale, 

temperature, presence of admixtures, and reducing conditions) 

• effect of curing temperature profile for on the hydraulic properties of as-emplaced 

saltstone 

• expected fracturing in saltstone with time 

• leaching of radionuclides from as-emplaced saltstone 

• reduction of Tc in saltstone 

• expected current and future Eh and pH conditions in the saltstone 

NRC staff concludes that the items listed above are key monitoring factors (Appendix A). NRC 

staff had previously identified aspects of the performance of the waste form as key monitoring 

factors in the 2005 NRC TER and associated monitoring plan for saltstone (NRC, 2005, 2007a). 

Additionally, during monitoring NRC staff identified three Open Issues related to the 

performance of the waste form (Chapter 1). Due to the risk significance of the performance of 

the waste form, the NRC staff will continue monitoring this area. 

2.7 Source Term Release and Near-Field Transport 

Source term modeling estimates the partitioning in and release of radionuclides from the 

disposal unit. The near field is generally defined as the area surrounding the waste that may 

have moisture flow and chemical conditions (e.g., due to the presence of the waste or 

engineered barriers) significantly different from the natural system in which the waste disposal 

facility is located. In this document, the term “Near Field” is used to reference the closure cap, 

saltstone waste, disposal units, backfill, and natural soil above the water table. 

2.7.1 Source Term Release Models 

Radionuclide release from saltstone is modeled in all cases is diffusive and advective, with 

aqueous concentrations controlled by linear partitioning (represented with Kd values) from 

saltstone solids into pore water. Chemical properties of saltstone are expected to change over 

time as infiltrating groundwater depletes the reductive and buffering capacity of the saltstone 

matrix, causing the Eh to fall and the pH to rise. Because sorption of many radionuclides 

depends on these chemical properties, release of most radionuclides is represented with Kd

values that depend on step changes in the modeled redox state and pH of saltstone 

(Section 2.7.1.1). As discussed in Section 2.6, Tc-99 mobility is more sensitive to redox 

conditions than the mobility of other key radionuclides. Because of this sensitivity, Tc release 

was modeled differently from the release of other elements (Sections 2.7.1.2 and 2.7.1.3). 
87 

2.7.1.1. Pore-Volume Model 

For elements other than Tc, DOE models radionuclide release by applying a single Kd value to 

the entire saltstone monolith or disposal unit component (i.e., floor, roof, or walls). DOE models 

the evolution of these Kd values as step changes corresponding to certain changes in the 

saltstone or disposal unit Eh and pH. The redox state is broadly divided into “reduced” and 

“oxidized” conditions and the pH is divided into three categories associated with the age of the 

cementitious material. DOE indicates that the initial condition of saltstone is expected to be 

middle-age, so the “new” values are not actually used in the source term release model. The 

material is "middle" if the pH is above 11 and “old” if the pH drops below 11. Elements other 

than Tc are assigned a single Kd value in the saltstone or disposal unit component based on six 

redox-age pairs: reducing-new, reducing-middle, reducing-old, oxidizing-new, oxidizing-middle, 

and oxidizing-old. 

As discussed in Section 2.6.3, DOE performed geochemical modeling to determine the number 

of pore volumes required to transition from reduced to oxidized conditions and from high pH to 

lower pH conditions. Cases A and K use the same type of release model for elements other 

than Tc, but chemical transitions in saltstone are predicted to occur at fewer pore volume 

replacements in Case K because of higher porosity and lower modeled reducing capacity. 

Based on the lower reducing capacity and higher effective porosity of the saltstone compared to 

the Case A values, the pore volumes for the Eh and pH transitions used in saltstone in Case K 

decreased from Case A values to about 500 and 7,600 pore volumes, respectively. 

Because the relevant properties of the disposal unit concrete (e.g., porosity, reducing capacity) 

are the same in Cases A and K, the number of pore volume replacements required to cause the 

chemical transitions are the same. However, because more water flows through saltstone and 

disposal unit in Case K, the same number of pore volume replacements are achieved in less 

time (Table 2.5-5). Additional details regarding chemical transition times for disposal unit 

concrete and saltstone are found in Sections 2.5 and 2.6. 

2.7.1.2 Explicit Shrinking-Core Model for Tc Release 

As discussed in Section 2.6, Tc is much more mobile in its reduced Tc(IV) form than it is in its 

oxidized Tc(VII) state. Because release of Tc is more redox-sensitive than release of other key 

radionuclides, DOE modeled its release based on the gradual oxidation of saltstone with time 

rather than the whole-monolith step changes used for other elements modeled with the porevolume model. In Case A and the sensitivity cases included in the PA, DOE used what it calls a 

“shrinking core” model that it implemented explicitly in PORFLOW. In this model, the redox 

state of individual finite elements is tracked in PORFLOWTM as dissolved oxygen in inflowing 

water consumes the reducing capacity in the cementitious material. DOE assumed the 

infiltrating water has oxygen concentrations that would be in equilibrium with atmospheric 

oxygen (i.e., 8 mg/L or 1.06 meq e-

/L) (the NRC evaluation of this assumption is included in 

Section 2.7.3). Oxidation from gas-phase transport of oxygen is not included (i.e., the amount 

of reducing capacity consumed is limited by the water flowing into the waste form or disposal 

unit component). Tc is assigned a “pseudo-Kd” in each finite element based on the amount of 
88 

reducing capacity consumed. The “pseudo-Kd” ranges from 1,000 mL/g to 0.8 mL/g and 

essentially represents Tc as mobile in oxidized areas of the saltstone and immobile in reduced 

areas. 

2.7.1.3 Average-Kd Model for Tc Release 

In Cases K, K1, and K2, DOE assumed saltstone has an initial fracture spacing of 30 m in 

Vault 1, 61 m in Vault 4, and 41 m in the FDCs. The saltstone in each disposal cells is assumed 

to decrease to 0.1 m at 10,000 years after closure (SRR-CWDA-2011-00044, Tables PA-8.3). 

However, PORFLOWTM cannot easily be used to model either the high fracture frequency at 

10,000 years or a fracture frequency that increases with time. Therefore, the Case K 

PORFLOWTM model did not model Tc release from oxidized and reduced areas explicitly, as 

was done for Case A and the other cases presented in the original PA (Section 2.7.1.2). 

Instead, for Case K, saltstone oxidation was calculated separately from the PORFLOWTM model 

using the method of Smith and Walton (1990). In this method, oxidation is modeled as 

proceeding from saltstone edges and fracture faces using the oxygen concentration at these 

surfaces as a boundary condition. The calculation results in the fraction of saltstone that is 

oxidized as a function of time. The fraction of saltstone oxidized is then is used to calculate a 

weighted average of the Kd values for oxidizing and reducing saltstone. This graduallydecreasing average Kd is assigned to Tc throughout the entire saltstone monolith. DOE refers 

to this approach as a “single-porosity” approach. To avoid confusion with the use of the term 

“single-porosity” in fracture flow modeling, the NRC staff refers to the Case K Tc release model 

as an “average-Kd” approach. 

An analytical model (external to PORFLOW) is used to estimate fracture growth over time using 

a semi-log fracture growth relationship with fracturing assumed to terminate at 10,000 years at a 

final fracture spacing of 10 cm (4 in). In Cases K, K1, and K2, the reducing capacity in saltstone 

is assumed to be 25% of the Case A value (i.e., 0.206 meq e-

/g). The reducing capacity in the 

disposal units is unchanged from its Case A value (i.e., 0.240 meq e-

/g). The oxygen 

concentration at the saltstone or disposal unit surfaces is assumed to remain constant at 

atmospheric levels (i.e., 1.06 meq e-

/L). The reduction capacity and dissolved oxygen 

concentrations are used in combination with a diffusion coefficient of 1.0x10-7 cm2

/s to 

parameterize an oxidation model to estimate progression of the oxidation front along each 

fracture face for each fracture (or partial fracture) that grows in over time. 

For Cases K, K1, and K2, oxidation of disposal unit concrete was modeled as proceeding from 

floor and wall edges, similar to the modeling of saltstone oxidation. In addition, like the 

modeling of saltstone oxidation, oxidation is affected by an assumed increase in diffusivity with 

time (from the values provided in Table 2.5-2 for the various disposal unit components to 

5x10-6 cm2

/s at 10,000 years). As discussed in Section 2.5, DOE did not provide information 

about the fracturing assumptions used in the disposal unit concrete in these cases. 
89 

2.7.2 Sorption Coefficients 

2.7.2.1 Cementitious Materials 

DOE developed sets of sorption coefficients (Kd) that were used in the modeling radionuclide 

release and transport in saltstone and concrete in the PA for all radionuclides except for Tc-99 

(2009 PA; Table 4.2-18). In the PA cases (i.e., all cases except for Cases K, K1, and K2) the 

same sets of sorption coefficient data were used for both saltstone and the disposal unit 

concrete. 

Table 4.2-18 in the PA included six sets of Kd values that represent various ages and oxidation 

states of the cementitious materials (i.e., young age, middle age, and old age for both reduced 

and oxidized cementitious materials). However, DOE only used three of these sets in the 

PORFLOWTM calculations (i.e., reduced middle age, oxidized middle age, and oxidized old 

age). The cementitious materials that do not contain slag are initially assigned oxidized middle 

age Kd values, while those that contain slag are initially assigned the reduced middle age Kd

value. The derivation of the times required for the Eh and pH transitions to occur are discussed 

in Section 2.7.1.1. When the Eh transition time is reached, the Kd values corresponding to 

oxidized middle age are used, and when the pH transition time is reached, the oxidized old age 

Kd values are used (SRR-STI-2009-00115). 

DOE stated that the Kd values were based, where possible, on site-specific information, but 

literature reviews or engineering judgment were relied on in the absence of such data (2009 PA; 

page 213). The reference cited for the majority of the radionuclides was 

WSCR-STI-2007-00640, which contained a compendium of literature information on the Kd

values for cementitious materials as well as some site-specific Kd measurements for a reducing 

grout and an aged cement. Other references cited include SRNS-STI-2008-00045, which 

reports site-specific measurements of sorption onto simulated saltstone and FDC concrete, and 

WSRC-STI-2007-00640, which reports site-specific measurements for Pu Kd values for 

cementitious materials. 

For Cases K, K1, and K2, DOE developed revised Kd values for cementitious materials for many 

elements (SRR-CWDA-2011-00044, Tables PA-8.6 and PA-8.7). DOE stated that these values 

are based on recent site-specific experimental data and new analyses of existing data. For 

most elements the citation for the revised Kd values is the 2010 compilation of geochemical data 

applicable to SRS performance assessments (SRNL-STI-2009-00473; Table 18). The revised 

Kd values for Ba and Sr are based on site-specific and literature information summarized in 

SRNL-STI-2010-00667, and the revised Kd values for U are based on a site-specific Kd

measurements onto reducing grout and a weathered concrete sample (SRNL-STI-2010-00493). 

Additionally, for Cases K, K1, and K2, DOE developed saltstone specific Kd values for Ba and Sr 

based on Kd values measured in leaching from a core sample of saltstone from Vault 4 

(SRNL-STI-2010-00667). In the majority of cases, the new analyses resulted in the selection of 

higher Kd values compared to the 2009 PA base case, meaning that the newer calculations will 

result in delayed radionuclide release and slower transport. 
90 

The Tc Kd values for the PA cases (i.e., cases other than Cases K, K1, and K2), were used in a 

shrinking core model explicitly implemented in PORFLOWTM (Section 2.7.1.2). In the PA cases, 

the Tc Kd value for reduced cementitious materials was 1,000 mL/g. The PA notes that this Kd

value is intended to immobilize the Tc in the reduced cementitious materials. As the reducing 

capacity is consumed by dissolved oxygen in infiltrating water, the Tc is made mobile by using a 

Kd value of 0.8 mL/g. The transition of the Tc Kd value from 1,000 mL/g to 0.8 mL/g as the 

reducing capacity is consumed is shown in Figure 4.2-41 of the PA. 

As described in Section 2.7.1.3, in Cases K, K1, and K2, the Kd for Tc was implemented in the 

model using an “average-Kd approach.” In this approach, a weighted average of the reduced 

and oxidized Kd values for the cementitious materials was determined based on the fraction of 

the material that was oxidized during the given time step. In Case K, a Tc Kd value of 

1,000 mL/g is assumed for reduced cementitious materials, and a value of 10 mL/g is assumed 

for oxidized cementitious materials. DOE based the reduced value of 1,000 mL/g on sorption 

experiments performed for simulated saltstone in a reducing environment (2% H2(g)) 

(SRNL-STI-2010-00668). The oxidized value of 10 mL/g was based on desorption of Tc from a 

saltstone core sample under oxidizing conditions (SRNL-STI-2010-00667). In Case K1, a Tc Kd

value of 500 mL/g is assumed for reduced cementitious materials and a value of 0.8 mL/g is 

assumed for oxidized materials. In Case K2, a Tc Kd value of 500 mL/g is assumed for reduced 

cementitious materials and a value of 10 mL/g is assumed for oxidized materials. 

2.7.2.2 Soil Kd Values 

The Kd values DOE used to model radionuclide transport through (i) backfill, (ii) the vadose 

zone, and (iii) the saturated zone are presented in Table 4.2-15 in the PA. The values tabulated 

for backfill soil are applied also to “clayey” soils in the saturated zone, that is, soils with a 

hydraulic conductivity less than 1x10-7 cm/sec (2009 PA). The values tabulated for the vadose 

zone are applied also to “sandy” soils in the saturated zone, that is, soils with a hydraulic 

conductivity greater than 1x10-7 cm/sec (2009 PA). 

DOE stated that the Kd values were selected on the basis of site-specific sorption data, from 

compilations of literature data, or on expert judgment; preference was always given to SRS sitespecific data (2009 PA). In some cases, chemical homologues were used. The main reference 

for the Kd values was WSRC-TR-2006-00004 (Table 10), which contains a compendium of SRS 

site-specific and literature Kd values. Another DOE report, SRNL-TR-2009-00019, is cited for 

the Tc values in PA Table 4.2-15. This report summarizes recent data from an E-Area borehole 

and recommends new Tc Kd values for soils. 

For Cases K, K1, and K2, DOE adopted modified soil Kd values for some elements on the basis 

of new data or new analyses of the literature (SRR-CWDA-2011-00044; Table PA-8.4). The 

new references for the modified values are SRNL-STI-2009-00473, SRNL-STI-2010-00493, and 

SRNL-STI-2011-00011. SRNL-STI-2009-00473 is a new compilation of geochemical data, 

prepared as part of the PA maintenance program, relevant to a number of performance 

assessment efforts at SRS. The relevance of this new report to the data in 

SRR-CWDA-2011-00044, Table PA-8.4, is that the new compilation includes consideration of 
91 

more recent site-specific sorption experiments. SRNL-STI-2010-00493 provides new Cl and U 

Kd values based on recent sorption experiments using SRS soil samples. 

SRNL-STI-2011-00011 cites new site-specific Ra data and calculates Kd values for Ba as being 

midway between measured values for Sr and Ra. 

2.7.3 Modeling of Flow in the Waste Form and Near Field 

DOE used the PORFLOWTM computer code to model unsaturated flow and contaminant 

transport to determine the source-specific, time-dependent radionuclide contaminant flux 

entering the water table aquifer at SDF. Three separate near-field (or vadose zone) models 

were constructed representing Vault 1, Vault 4, and the FDCs. This section summarizes DOE’s 

approaches for near-field flow and transport model construction, material properties, and model 

validation. 

2.7.3.1 Near-Field and Vadose Zone Model Construction and Boundary Conditions 

The discrete features of SDF Vaults 1 and 4, and the sixty-four FDCs, are described in 

Section 2.5 of this TER. The discrete features of Vaults 1 and 4, and the FDCs were 

necessarily simplified for the purposes of PA modeling (Figure 2.5-6). A description of the 

simplified conceptual models represented in the PORFLOWTM model for each SDF source is 

discussed in Section 4.4.1 of the 2009 PA (page 243). Section 4.4.4.1.2, “Disposal Unit 

Modeling in PORFLOW” (2009 PA; page 277) also provides details on the numerical model 

construction and flow results for each vadose zone model. 

The long rectangular, Vaults 1 and 4, were represented in 2-D Cartesian coordinates as a 

transverse slice with inventory assigned to a nominal thickness of 1 length unit (i.e., 0.3 m [1 ft]). 

Only half of Vault 1 and 4 in the short dimension is modeled, taking advantage of symmetry 

about the centerline. The FDCs were modeled in 2-D cylindrical coordinates with an assumed 

“thickness” of one unit radian, implicitly assuming symmetry about the centerline axis. 

Cracks in Vault 1 and 4 walls are not modeled explicitly but rather are reflected in the initial and 

degraded hydraulic property assignments. To account for the potential for contaminants to be 

located in the vault walls, vault wall pore fluid is assumed to contain concentrations similar to 

those found in the vault cells with approximately 0.65 and 0.5 percent of the inventory estimated 

to be present in the Vault 1 and 4 walls, respectively. 

DOE places no-flow boundary conditions at the disposal unit centerlines and at the outer 

perimeter or radius of the Vault 1, 4, and FDC model domains, which extend at least 9.1 m 

(30 ft) beyond the perimeter of each disposal unit (2009 PA; page 283). Net infiltration rate and 

vadose zone thickness are naturally variable with time; however, short term variations (e.g., 

seasonal effects) in these parameters are not simulated. A time-dependent net infiltration flux 

that varies as a function of the degradation state of the engineered closure cap (Section 2.4 of 

this TER and 2009 PA; Table 3.2.7) is prescribed as the upper boundary condition to the nearfield and vadose zone PORFLOWTM model. The lower boundary of the model coincides with 

the water table where pressure head is set to zero. DOE specifies an outflow boundary 
92 

condition for radionuclide transport, whereby the flux leaving the vadose zone is by advection 

only. 

The vadose zone PORFLOWTM model domains use approximately 7,000 cells to represent up 

to 18 different material types each with their own hydraulic and transport properties. Different 

material types are also used to evaluate various scenarios such as flow through preferential 

pathways that may form in a portion of a single material, in addition to representing variation in 

hydraulic and transport properties of the various materials themselves. Grid resolution is stated 

to be a compromise between two competing objectives: (1) resolution of thin geometric features 

(e.g., sheet drains, HDPE-GCL liners) and sharp flow field transitions (e.g., ponded water 

flowing over roof edge), and (2) achieving reasonable computer storage and runtimes. 

The upper boundary of the near-field model also includes materials located physically above the 

top of the disposal units (e.g., lower lateral drainage layer and HDPE-GCL). The lower lateral 

drainage layer and HDPE-GCL are assumed to degrade over time (Sections 2.4 and 2.5). After 

approximately 19,000 years, the hydraulic conductivity and porosity of the lower lateral drainage 

layer are estimated to be similar to those for the overlying backfill layer (2009 PA; page 183). 

However, the maximum hydraulic conductivity of the HDPE-GCL at 20,000 years is still only 

assumed to be about a fraction of an inch per year (i.e., <1x10-8 cm/s [ <0.1 inches/yr]) after 

20,000 years (Figure 2.5-3). Additionally, an HDPE layer is applied to the walls of the FDCs 

with a hydraulic conductivity that peaks at around 8x10-8 cm/s (1 inch/yr) at 20,000 years 

(Figure 2.5-2). The vadose zone (below the top of the disposal units) comprises an upper and 

lower zone described in more detail below. 

Time step sizes were selected as a compromise between two competing objectives: 

(i) resolution of concentration peaks from relatively mobile species that migrate as a pulse, and 

(ii) achieving reasonable computer runtimes. A transport step size of 1 year was selected for 

vadose zone flux simulations with the exception of Tc-99 simulations that were run using a 

0.05 year transport step size to avoid numerical inaccuracies. A recording frequency of 1 year 

was used for the vadose zone transport simulations. A 2.5 year transport step size and 20 year 

recording frequency was selected for aquifer concentrations. 

2.7.3.2 Near-Field and Vadose Zone Transport Material Properties 

The cementitious materials simulated in the SDF PA modeling can be grouped into five 

categories as follows: 

1. Low quality concrete associated with the lower mud mats for the FDCs. 

2. Medium quality or ordinary concrete associated with the roof of Vaults 1 and 4. 

3. High quality concrete associated with the base of Vaults 1 and 4, and the concrete for 

the FDCs, including the upper mud mats. 

4. “Fractured concrete” associated with the existing walls of Vaults 1 and 4, which have 

experienced macroscopic cracking. 

5. Saltstone. 
93 

The initial hydraulic properties associated with each of the cementitious material categories are 

presented in Table 4.2-16 of the PA, and are summarized in Table 2.7-1 

Preferential pathways through cracks, fractures, or other discrete features were either implicitly 

represented as an increase in saturated hydraulic conductivity within a porous medium, or were 

explicitly represented in a porous medium formulation as discrete zones of high permeability 

depending on the feature and the case being analyzed. The materials palette used in 

PORFLOWTM SDF modeling is located in Table 4.4-10 of the 2009 PA (page 291). 

Non-zero diffusivities were assigned in the vadose zone transport modeling, while dispersivities 

were set to zero. A value of zero for the dispersivity would tend to minimize plume spreading 

leading to higher peak fluxes to the water table. 

DOE assumed that the disposal unit concrete will degrade with time due to exposure to sulfate 

in the salt waste (Section 2.5). In Case A, degradation of the hydraulic properties of 

cementitious materials such as the disposal unit roofs, walls, and floors is assumed to coincide 

with the creation of expansive phases (e.g., ettringite) that leads to cracking. Effective hydraulic 

properties are calculated based on a weighted average of degraded and intact material 

properties. In Cases K, K1, and K2, the degradation of the disposal unit concrete was modeled 

non-mechanistically with increases in the hydraulic conductivity and diffusivity. Increased 

degradation of the disposal unit concrete also is assumed in several of the deterministic 

sensitivity cases (Section 2.5). 

As discussed in detail in Section 2.6, the NRC staff concluded that DOE’s Case A assumption 

that saltstone does not degrade during a 20,000 year evaluation period is unrealistic. Cases K, 

K1, and K2 consider a scenario in which the saltstone monolith degrades hydraulically with an 

initial hydraulic conductivity of 1x10-8 cm/s that increases over time to 1x10-6 cm/s at 10,000 

years. Increased degradation of saltstone is also assumed in some of the deterministic 

sensitivity cases (Section 2.6). The Case A assumptions regarding engineered barrier and 

saltstone hydraulic performance result in greater than 99 percent of meteoric water that 

infiltrates through the cover being shed through the lower lateral drainage layer (Sections 2.5 

and 2.13). The result of the revised parameters in Cases K, K1, and K2 is a significant increase 

in flow through the saltstone matrix compared to the Case A scenario with average Darcy 

velocities around 10 cm/yr (4 in/yr) at 10,000 years and 20 cm/yr (8 in/yr) at 20,000 years. 

The near-field PORFLOWTM models utilize characteristic curves (relative permeability and 

suction head as functions of saturation) for cementitious materials and soils. The data used to 

develop these curves are taken from WSRC-STI-2006-00198, WSRC-STI-2007-00649, or 

SRNL-STI-2009-00115. Moisture characteristic curves used for cementitious materials in PA 

modeling are based on WSRC-STI-2006-00198, except for fractured concrete which is based 

on SRNL-STI-2009-00115. In the fractured cases, the fractures become unsaturated under 

certain conditions. Moisture characteristic curves are illustrated in Figures 4.2-23 through 

4.2-27 of the 2009 PA. The NRC staff’s evaluation of DOE’s approach to modeling unsaturated 

flow including development of moisture characteristic curves is discussed further in 

Section 2.7.4.4. In Cases K, K1, and K2, it was assumed that relative permeability and 
94 

saturation were equal to 1.0 for all suction levels and moisture characteristic curves were not 

used. 

Table 2.7-1: Initial (or for Some Materials Time Invariant) Material Properties Used in 

PORFLOWTM Vadose Zone Modeling 

Material

Saturated 

Effective 

Diffusion 

Coefficient 

(cm2/s)

Average 

Total 

Porosity 

(%)

Average 

Dry Bulk 

Density 

(g/cm3)

Average 

Particle 

Density 

(g/cm3)

Saturated 

Horizontal 

Hydraulic 

Conductivity 

(cm/s)

Saturated 

Vertical 

Hydraulic 

Conductivity 

(cm/s)

Backfill 5.3x10−6 35 1.71 2.63 7.6x10−5 4.1x10−5

Undisturbed 

Vadose 

Zone 

5.3x10−6

 39 1.62 2.66 3.3x10−4 9.1x10−5

Saltstone 

and Clean 

Grout Cap 

1.0x10−7

 58 1.01 2.4 2.0x10−9 2.0x10−9

High 

Quality 

Concrete 

5.0x10−8 121

112

2.241

2.222

2.551

2.492

3.1x10−10 (1)

9.3x10−11 (2)

3.1x10−10 (1)

9.3x10−11 (2)

Fractured 

Walls in 

Vaults 1 

and 4 

5.0x10−8

 12.0 2.24 2.55 1.7x10−1 1.7x10−1

Medium 

Quality 

Concrete 

1.0x10−7 14.53

13.64

2.203

2.214

2.573

2.564 5.0x10−9

 5.0x10−9

Low Quality 

Concrete5 8.0x10−7

 21.1 2.06 2.61 1.0x10−8 1.0x10−8

2009 PA (Tables 4.2-14 and 4.2-16) and WSRC-STI-2006-00198 (Table 5-18).

1

 Vaults 1 and 4 base. 

2

 FDCs. 

3

 Vault 1 roof. 

4

 Vault 4 roof. 

5

 Lower mud mats of the FDCs. 

The vadose zone thickness between the vaults and the underlying Upper Three Runs (Upper 

Zone) water table ranges from 10.9 m (35.6 ft) (for future disposal cells 10A to 10D 

[Figure 2.8-8]) to 14.6 m (48 ft) (for Vault 1). Although the vadose zone thickness is variable for 

FDCs as provided in Table 4.2-13 in the 2009 PA, page 191, the thickness is modeled as a 

constant based on the average thickness for all 64 FDCs of 12.8 m (42 ft) (2009 PA; page 188). 

The upper vadose zone is assumed to be composed of backfill materials with properties more 

similar to clay than sand. The undisturbed lower zone is assumed to have material properties 

similar to sand. In reality, the undisturbed materials below the vaults may consist of both native 
95 

upper and lower vadose zone materials. Because the properties of the upper vadose zone are 

more clayey and would tend to retard contaminant transport compared to backfill or lower 

vadose zone materials, DOE indicates the modeling assumptions that backfilled (rather than 

upper vadose zone) material surrounds the vaults and that all of the soil underneath the 

disposal unit is similar to lower vadose zone (rather than upper vadose zone) material is 

conservative. Vadose zone material properties are listed in Table 2.7-1. DOE assumes vadose 

zone material properties are constant, and do not change with time. Soil moisture characteristic 

curves are presented in the PA for backfill (Figure 3.2-17) and for lower vadose zone soil 

(Figure 4.2-22). A complete list of figures that illustrate moisture characteristic curves used in 

the SDF PORFLOWTM modeling is provided in Table I3-2 of the PA, along with the source of the 

data used to construct each curve. 

2.7.3.3 Near-field Model Validation 

DOE used characterization and monitoring data from uncapped E-Area (adjacent to F-Area) to 

validate aspects of the PORFLOWTM vadose zone model. These data included soil suction and 

water content, tracer test pore velocity, and tritium plume concentration. Soil suction data from 

E-Area indicate a range from 50 to 200 cm (20 to 79 in), and PORFLOWTM vadose zone 

modeling produces upper vadose zone soil suction values of 83 cm (33 in) and lower vadose 

zone values of 170 cm (67 in) (SRS-REG-2007-00002). Water content data suggest saturation 

ranges from 35 to 75 percent, and PORFLOWTM vadose zone modeling produces upper vadose 

zone saturation values of 91 percent and lower vadose zone values of 72 percent 

(SRS-REG-2007-00002). Field and laboratory tracer test experiments indicate a pore velocity 

of 114 cm/yr (45 in/yr), and PORFLOWTM vadose zone modeling produces upper vadose zone 

pore velocities of 86 cm/yr (34 in/yr) and lower vadose zone velocities of 109 cm/yr (43 in/yr) 

(SRS-REG-2007-00002). Comparison between the measured tritium plume concentration from 

a disposal trench at E-Area and PORFLOWTM vadose zone model results do not serve to 

validate the model, but PORFLOWTM results are generally consistent with measured data and 

the vadose zone model is not invalidated by the data (SRS-REG-2007-00002). 

2.7.4 NRC Evaluation – Release and Near-Field Transport 

2.7.4.1 NRC Evaluation – Release Models 

The three radionuclide release models DOE used in the PA and RAI responses (i.e., the pore 

volume model, explicit shrinking-core model, and average-Kd model) all are based on linear 

partitioning of radionuclides from saltstone solids into the pore water. Differences between 

these three approaches are discussed in Section 2.7.1 and evaluated later in this section. First, 

this section addresses the common assumption that radionuclide release and transport can be 

represented with linear partitioning (represented by Kd values). This section addresses two 

main implications of the linear partitioning assumption for release and transport modeling: 

(1) effects if laboratory tests are actually solubility limited, and (2) effects if radionuclides are 

solubility limited in the emplaced waste. 
96 

If the concentration of radionuclides dissolved in water is actually limited by their solubility 

instead of by sorption in laboratory tests intended to measure linear sorption coefficients (Kd

values), Kd values may be overestimated. Values may be overestimated because, at 

equilibrium, the dissolved radionuclide concentration cannot exceed its solubility limit (by 

definition). Therefore, if dissolved radionuclide concentrations in sorption experiments are 

actually limited by solubility, they will not increase as a radionuclide is added to the test system, 

even though the amount of the radionuclide in the solid phase increases. Because the sorption 

coefficient is the ratio of the solid concentration to the dissolved concentration, increases in the 

solid concentration without increases in the dissolved concentration cause the measured 

sorption coefficient to increase. This phenomenon is a particular concern for Pu-239 and 

Np-237, which were found to be solubility-controlled in experiments designed to measure 

sorption in cementitious material (SRNL-STI-2009-00636). In an RAI response 

(SRR-CWDA-2010-00033, SP-9), DOE indicated that sensitivity tests performed with the SDF 

GoldSim® model suggested the potential effect of overestimating these Kd values is small. 

Specifically, DOE found that setting Kd values for Pu-239 and Np-237 in cementitious material 

equal to zero increased their contribution to dose by less than a factor of three. However, 

because of NRC concerns about the GoldSim® model (Sections 2.11.4.2 and 2.11.4.3), the 

NRC staff concluded that a better basis is needed for Pu-239 and Np-237 sorption in 

cementitious materials. The NRC staff included improved support for Kd values in saltstone as a 

PA maintenance item. 

If a radionuclide is actually solubility limited in emplaced waste but its release is represented 

with a Kd value, the accuracy of the modeled release depends on the solid concentration in the 

actual waste. For the SDF, representing radionuclide release with a Kd instead of a solubility 

limit is likely to be of the most concern for Tc-99, because it is anticipated to be solubility-limited 

in reducing saltstone but is represented with a Kd value. DOE has not shown whether the 

aqueous concentrations calculated with the sorption coefficients used in the model are 

reasonable and reflect the physical processes that DOE believes actually control release (i.e., if 

the aqueous concentrations calculated with a Kd match the aqueous concentrations that would 

result from a solubility limitation). Therefore, the NRC staff performed a simple calculation to 

evaluate the reasonableness of the aqueous Tc concentration given a Kd of 1,000 mL/g (the 

appropriateness of the selected Kd values is evaluated in subsequent paragraphs). The 

parameters needed for this calculation are as follows: 

- Average FDC Tc inventory of 540 Ci (3.18x104

 g) (PA Table 3.3-5) 

- FDC saltstone diameter of 45.7 m (150 ft) and height of 6.10 m (20 ft) (PA Section 3.2.1.3.2) 

- Saltstone porosity of 0.58 and particle density of 2.42 g/cm3

 (PA Table 4.5-5) 

The NRC staff first considered a saturated system at equilibrium under reducing conditions. In 

this case, the Tc inventory is modeled as being distributed between the saltstone solid and the 

pore water according to a Kd of 1,000 mL/g. The resulting calculated aqueous Tc concentration 

would be 3x10-8 M. This value is within the range of measured Tc(IV) solubility limits in 

reducing, high-pH systems (e.g., Pilkington, 1990, Greenfield, et al., 1998, and Warwick, et 

al., 2007). Therefore, it appears that, the high Kd would yield aqueous concentrations roughly 

consistent with solubility control until the initial Tc Kd is significantly depleted. After the Tc 
97 

inventory is depleted, continued solubility control would predict a constant release rate while 

modeling with a Kd value will predict a decreasing release rate. This is a general problem 

associated with modeling solubility-limited release with a Kd model. However, because Tc 

release is likely to be more significantly affected by the rate of saltstone oxidation than the 

difference between a Kd model and solubility–limited release, the NRC staff concluded using a 

Kd model was appropriate in this version of the PA. 

Pore Volume Model

There are two types of concerns about the pore volume model: (1) concerns about the basis for 

the estimated number of pore volume flushes required to cause redox (Eh) and age (pH) 

transition times, and (2) concerns about using a whole-monolith step change to model changes 

in radionuclide mobility instead of modeling more gradual changes in the saltstone and disposal 

unit concrete. As discussed in more detail in Section 2.6, the NRC staff found the basis for 

predicted number of pore volumes required to cause certain pH and Eh changes to be 

insufficient in Case A. However, the NRC staff expects the transition times in Case K to more 

reasonably reflect the expected evolution of the saltstone chemical environment. 

The NRC staff found the pore-volume model DOE used for radionuclides other than Tc (in all 

cases) to be appropriate for radionuclides with little sensitivity to Eh or pH because they are 

generally insensitive to how Eh and pH evolution is modeled. The NRC staff questioned the 

applicability of the model to other radionuclides (i.e., redox or pH sensitive radionuclides) such 

as Np-237 and Pu-239 (NRC, 2010b, i; SP-13). In response, DOE indicated that only Cr, Rh, 

Sb, and Tc had large enough differences in their sorption coefficients under different pH and 

redox conditions to be sensitive to the use of a pore-volume model as compared to a model 

implementing a more gradual transition. DOE indicated that, except for Tc, none of these 

elements is likely to be dose significant. The NRC staff agrees that changing the source term 

release model for these radionuclides will not have a significant effect on peak dose unless 

other assumptions (e.g., regarding inventory or sorption coefficients) about these radionuclides 

change. 

Explicit Shrinking Core Model

In general, the NRC staff concludes the explicit shrinking-core model DOE used to model Tc-99 

release from saltstone in the PA Cases (i.e., all cases except Cases K, K1, and K2) is 

appropriate. In particular, the NRC staff finds that tracking releases from the oxidized and 

reduced fractions of saltstone separately appears to be the most appropriate implementation of 

DOE’s conceptual model of Tc-99 release. Regarding the oxidation modeling, the NRC staff 

finds DOE’s assumption that infiltrating water is equilibrated with atmospheric oxygen (i.e., 

8 mg/L dissolved oxygen, equivalent to 1.06 meq/L) to be conservative because subsurface 

oxygen concentrations are likely to be lower (e.g., DOE reports a measured value of 1.2 mg/L 

from a groundwater monitoring well near the SDF). 

The NRC staff finds DOE’s basis for neglecting gas-phase transport of oxygen to saltstone 

surfaces to be insufficient. Specifically, DOE indicated that gas-phase transport of oxygen to 
98 

saltstone faces is neglected in the PA Cases because saltstone is nearly completely saturated 

with water. The NRC staff does not believe this is an adequate basis for neglecting gas-phase 

transport of oxygen to saltstone faces because the significance of gas-phase transport is not 

necessarily that gas permeates saltstone, but that gas-phase oxygen serves as another source 

of oxygen beyond the finite quantity of oxygen in infiltrating water. The NRC staff expects that, 

in cases with sufficient water flow (e.g., the Synergistic Case) neglecting gas phase oxygen 

transport is mitigated by DOE’s assumption that the infiltrating water is saturated as if in 

equilibrium with atmospheric concentrations of oxygen. However, the NRC scoping calculations 

suggest that in Case A, water flow into saltstone is so limited that the amount of oxidation from 

inflowing water would be small relative to the potential oxidation from diffusion of gas-phase 

oxygen into saltstone. 

Average-Kd Model

DOE addressed some of the uncertainties of the shrinking core model and the Tc Kd values in 

Case K presented in SRR-CWDA-2011-00044 (Table PA-8.1). For example, the model allowed 

for more extensive cracking of the saltstone, providing a greater available surface area for 

oxidation. Unlike the explicit shrinking core model implemented in PORFLOW, the average-Kd

model does include diffusion of oxygen from the gas phase into saltstone or the disposal unit 

concrete, however it neglects oxygen introduced into the saltstone or disposal unit in infiltrating 

water. 

As explained in more detail in Section 2.7.1.3, the weighted average Kd is based on a 

calculation of saltstone oxidation performed outside of the PORFLOWTM near-field model. The 

average Kd value, which gradually decreases as saltstone is modeled as becoming more 

oxidized, is then applied to the whole monolith. In general, the NRC staff finds the average-Kd

approach yields results that are not consistent with DOE’s conceptual model of Tc release, in 

which Tc is relatively mobile in oxidized portions of saltstone and immobile in the reduced 

fraction. In particular, the NRC staff finds that with certain relevant parameter combinations, the 

average-Kd model predicts very little release of Tc as oxidation proceeds, and predicts a 

relatively sudden release of Tc when saltstone oxidation is nearly complete. This tends to both 

delay and overestimate the peak Tc release (Section 2.13.3). 

In theory, the average-Kd model rests on the assumption that the concentration of Tc in 

saltstone pore water will equilibrate between oxidized and reduced regions more rapidly than Tc 

is released from the oxidized region. In DOE’s description of the average-Kd model 

(SRR-CWDA-2011-00044) DOE indicates that “...the liquid concentration has been assumed to 

be in equilibrium between the oxidized and reduced regions,” that is, the Tc pore liquid 

concentration is the same in both the oxidized and reduced regions. This assumption is 

necessary to make the simplification from Equation 9 to Equation 10 in DOE’s mathematical 

development of the average-Kd model (SRR-CWDA-2011-00044, page 84). The NRC staff 

finds that this concentration equivalence to be unrealistic. First, the assumption that the Tc 

concentration in saltstone equilibrates between the oxidized and reduced regions before a 

significant amount of Tc is released from the oxidized region is inconsistent with the same 

mechanism (i.e., diffusion) governing both processes. That is, diffusion controls both the 
99 

release of Tc from the oxidized region and the diffusion of Tc from the oxidized to the reduced 

region. If Tc is removed advectively from the fractures, as the NRC staff expects, the aqueous 

concentration gradient from the oxidized region to the fracture would be similar (and slightly 

greater than) the aqueous concentration gradient from the oxidized region to the reduced 

region. Thus, Tc release would be expected to occur on a similar time scale as Tc equilibration 

between the oxidized and reduced regions. The NRC staff also finds that DOE has not provided 

a basis for assuming that the diffusion of Tc between the oxidized and reduced regions of 

saltstone is significantly faster than the diffusion of dissolved oxygen into saltstone. 

In response to NRC concerns about the average-Kd model (NRC, 2011g, h, l); DOE compared 

two simple GoldSim® models (SRR-CWDA-2011-00114). The models were designed to enable 

a comparison between the results of an average-Kd approach and an approach in which Tc 

release from the oxidized and reduced portions of saltstone are modeled with two separate Kd

values. DOE refers to this approach as a “dual-porosity” approach. To avoid potential 

confusion with the use of the terms “dual-continuum” and “double-porosity” used in flow 

modeling, the NRC staff refers to a model in which Tc release from reduced and oxidized 

fractions of saltstone is tracked with two Kd values as a “dual-Kd” approach. 

As a result of this comparison, DOE concluded that the average-Kd model provides a good 

approximation of the results of a more complex dual-Kd model. DOE also concluded that the 

average-Kd model underestimates release at early times and overestimates release at later 

times (SRR-CWDA-2011-00114). The NRC staff has three main concerns with DOE’s 

conclusions, each of which is described in more detail below: 

• The results of the two GoldSim® models presented by DOE do not agree as well as DOE 

suggests. 

• Although the results did not agree with each other well in many cases, the agreement would 

have been worse if the results were not artificially made more similar by a modeling artifact 

related to DOE’s method of representing decreasing fracture spacing. 

• Agreement between two GoldSim® models does not demonstrate that DOE’s PORFLOWTM 

model accurately represents Tc release. 

In addition, the NRC staff notes that a comparison between two GoldSim® models, or two 

PORFLOWTM models, is not a substitute for a comparison between the model output and 

experimental results. 

DOE presented the results of the two models on log scales, which minimizes the apparent 

difference between the results. However, in many cases, the results of the two models differed 

by approximately an order of magnitude. In all cases, the peak release from the average-Kd

model exceeded the peak release from the dual-Kd model. While apparently conservative, this 

modeling artifact can lead to inconclusive results if the peak dose exceeds the dose limit. 

DOE’s model representing the average-Kd approach was a simple one-dimensional model 

containing a column of 10 cells that releases from the last cell into the unsaturated zone. The 

flow rate through this column is based on DOE’s Case K Vault 4 PORFLOWTM model. 
100 

DOE’s dual-Kd model is a 20 x 20 grid of cells representing the intact saltstone, with an 

additional column of 20 cells representing a fracture. An oxidation front was assumed to 

proceed from the fracture face. As the oxidation front reaches the furthest edge of a column of 

cells, the Kd of Tc in the column of cells changes from its value in reduced material to its value in 

oxidized material; resulting in the mobilization of the Tc in the column. Tc was released from 

the oxidized column through diffusion to the fracture followed by advective transport out of the 

fracture and through diffusion back into the reduced region. As the system fractures, the 

modeled cell size decreases to represent smaller fracture spacing, but the inventory in the cells 

did not change. At later times and with larger amounts of degradation, the amount of back 

diffusion from the oxidized columns into the reduced ones was significant and rapid because of 

the very small diffusive lengths. This caused the dual-Kd model increasingly to mimic a stirred 

tank (i.e., equilibration of aqueous concentrations in different parts of the model), and, thereby, 

to increasingly mimic the average-Kd model. 

The NRC staff determined that modeling flow through the oxidized regions in the dual-Kd model, 

which the NRC staff considers reasonable, would limit the amount of diffusion from the oxidized 

region to the reduced region because the oxidized Tc is advectively transported out of the 

monolith. Limiting diffusion from the oxidized to the reduced region would cause less 

concentration of Tc in the shrinking reduced region, which would mitigate the pulse release 

characteristic of the average-Kd model (Figures 2.13-6 and 2.13-7). DOE performed a 

sensitivity analysis in which they did include flow through the oxidized region, but the results did 

not show much of a difference (i.e., compare SRR-CWDA-2011-00114; Figure 11 with 

Figure 1). This result was contrary to NRC results with a similar GoldSim® model, which 

showed a significant difference in the peak release rate from the dual-Kd model if advective flow 

was modeled through the oxidized region. The flow rate used in DOE’s sensitivity analysis was 

not provided in SRR-CWDA-2011-00114. The difference in the NRC and DOE results may be 

due to differences in the assumed flow rates. 

2.7.4.2 NRC Evaluation – Cementitious Material Sorption Coefficients (Kd values) 

In general, the approach of defining sets of Kd values that apply to both saltstone and other 

cementitious materials is reasonable. Like saltstone, the formulations for floors and walls of 

Vault 1, Vault 4, and the FDCs include the key reactive components blast furnace slag and 

cement (Tables 2.5-4 and 2.6-3). On the other hand, there are enough differences in the 

formulations that Kd differences must be assessed if using data on one type for modeling 

another. For example, unlike the concretes, saltstone contains no sand or aggregate, and was 

formed using a salt solution (PA; Table 4.2-7). DOE has been conducting laboratory studies to 

obtain saltstone-specific Kd values (e.g., SRNS-STI-2008-00045). The NRC staff believes that 

performing this type of research to determining medium-specific parameter values that 

recognize differences in formulation, texture, and processing will provide useful model support 

for the performance assessment. 

In addition, when conducting research on sorption coefficients, it is important to assess whether, 

given the laboratory results, a Kd approach will be appropriate for modeling release from 

saltstone throughout the simulated time period, particularly when solubility may be a factor in 
101 

controlling concentration. When performing Kd experiments it is also important to understand if 

the measured values reflect precipitation rather than sorption. The results of Kd or leaching 

experiments can be misleading and inappropriate to apply to transport modeling if the sorption 

is controlled by solubility in the experiment. 

In comment SP-11 (NRC, 2010i), the NRC staff raised the issue of Kd measurements for old age 

materials that were conducted using a calcite-saturated solution but did not account for the 

mineralogical characteristics of long-aged cementitious materials. As discussed in the staff’s 

review (NRC, 2010i) of the original DOE response (SRR-CWDA-2010-00033, SP-11), the NRC 

staff does not agree that the measured Kd values have been demonstrated to be appropriate or 

conservative for an extensively aged material. In response, DOE noted the challenge of 

identifying an appropriate surrogate for aged cementitious material, acknowledged the 

uncertainty in their approach, and implied that their long-range program plan for testing of 

cementitious material will consider additional experiments that could address sorption 

coefficients for aged materials. The NRC staff acknowledges the difficulty in finding appropriate 

surrogate materials and concludes that, because the radionuclides expected to be most 

significant to off-site dose from the SDF have relatively low sorption coefficients, the effect of 

using Kd values specific to old-age material may be limited. The NRC staff will review any 

information DOE develops on this issue as part of its PA maintenance program. 

A related and important issue is the question of the applicability for modeling oxidizing 

conditions of sorption data obtained from cementitious materials that still contain blast furnace 

slag. As discussed in the context of saltstone Tc release in Section 2.7.1, the DOE conceptual 

model for oxidizing conditions in any cementitious material assumes that the reducing agent 

(e.g., blast furnace slag) has been exhausted. Therefore, sorption experiments on material 

originally containing a reducing agent may not faithfully reflect modeled oxidizing conditions, 

even if conducted under air using an initially oxidizing aqueous solution. This observation has 

potential implications for measurements such as those in SRNS-STI-2008-00045, used as the 

basis for oxidizing conditions Kd values of redox-sensitive elements such as Tc and Pu. 

The NRC review of the Kd values focused on the radionuclides that had a greater potential to be 

risk significant. The NRC review of those radionuclides (i.e., Tc, I, Ra, Se, Sr) is described 

below. 

Technetium Kd Value for Cementitious Materials

The release of Tc from the saltstone monolith is sensitive to the oxidation state of the Tc 

(Section 2.6). The NRC staff has been monitoring information regarding Tc reduction in 

saltstone as part of Open Issue 2009-1. The modeling of the reduction and sorption of Tc (i.e., 

the Kd value for Tc) is therefore one of the most risk-significant portions of the PA. 

In the PA cases (i.e., the cases other than Cases K, K1, and K2), DOE used a “shrinking core” 

model for the release of Tc. In this model, a value of 1,000 mL/g was used for the Kd for Tc 

under reducing conditions to immobilize the Tc. Case K also uses a value of 1,000 mL/g Tc Kd
102 

value for cementitious materials under reducing conditions, while Cases K1 and K2 assume a 

value of 500 mL/g (Table 2.7-2). 

In comment SP-15 (NRC, 2010b, i), the NRC staff asked DOE to provide further support for the 

adopted Tc pseudo-Kd of 1,000 mL/g because this value was not consistent with Kd values 

measured for simulated saltstone under reducing conditions (e.g., values of 6.5 mL/g and 

91.3 mL/g reported in SRNS-STI-2008-00045). In the initial response to the comment 

(SRR-CWDA-2010-00033), DOE cited a “recommended” value of 5,000 mL/g as the basis for 

selecting a value of 1,000 mL/g. In the second RAI (NRC, 2010i), NRC staff noted that the 

5,000 mL/g value was measured for a formulation that included a strong reducing agent and is 

very different than the saltstone formulation. 

In the second response comment SP-15 (SRR-CWDA-2011-00044), DOE referred to a recent 

report summarizing literature and new laboratory data (SRNL-STI-2010-00668). The new 

sorption data reported in SRNL-STI-2010-00668 (originally reported by Lilley, 2010) are used by 

DOE to justify a Tc reducing Kd of 1,000 mL/g. In this report, Tc Kd values for simulated 

saltstone under reducing conditions were measured to be 711 mL/g after 22 days and 518 mL/g 

after 56 days. However, NRC staff notes that this measurement was performed in a 2% H2

atmosphere, so this result may not be applicable to the actual saltstone system. In fact, in 

SRNL-STI-2010-00668, the reported Kd values for Tc sorption onto concrete that does not 

contain slag (i.e., concrete that does not contain a reducing agent) were comparable to those 

seen for the simulated saltstone. DOE subsequently concluded that the 2% H2 atmosphere was 

responsible for the observed Tc reduction and sorption in these experiments 

(SRNL-STI-2011-00716). In a more recent report (SRNL-STI-2011-00716), a Tc Kd value of 

1,258 mL/g was reported. This measurement was also made in an atmosphere containing 0.1% 

H2 in the presence of a palladium catalyst to convert O2 to water through reaction with H2. 

Unlike the results of the study using 2% H2 (SRNL-STI-2010-00668), a concrete sample without 

slag in the 0.1% H2 atmosphere did not show Tc reduction or sorption (SRNL-STI-2011-00716). 

This result indicates that the observed reduction and sorption of Tc should be attributable to 

saltstone rather than an experimental artifact (SRNL-STI-2011-00716). However, the Tc Kd for 

saltstone under reducing conditions was also measured in a leaching experiment using a 

saltstone core sample (SRNL-STI-2010-00667). In this experiment, a Kd value of 139 mL/g was 

measured for Tc. This measurement was performed under an atmosphere of 30 to 60 ppm O2. 

Because DOE has not shown that the SDF will be completely free of even trace levels of O2, it is 

unclear if the results of the study performed in the 1% H2 atmosphere with a palladium catalyst 

(SRNL-STI-2011-00716) are applicable to the actual saltstone system. 

The Tc Kd information described above is summarized in Table 2.7-2. Based on this 

information, the NRC staff concludes that the assignment of a Kd of 1,000 mL/g to Tc in 

saltstone under reducing conditions is still not adequately supported. The Tc Kd measurements 

for simulated saltstone that have been performed to date have either not supported a Tc Kd

value of 1,000 mL/g for saltstone under reducing conditions, or the experiments have been 

performed in an atmosphere that is significantly different than for the actual as-emplaced 

saltstone. Furthermore, the use of a reducing Kd value requires that oxygen is essentially 

absent from the system, because studies have shown that even very small quantities of oxygen 
103 

(i.e., 30 to 60 ppm O2), in a system oxidize Tc in saltstone (e.g., SRNL-STI-2010-00668). DOE 

has not demonstrated that oxygen will be absent from the water-saltstone system when release 

is possible. Although the NRC staff finds that there is also insufficient basis for the Tc Kd value 

of 500 mL/g assumed in Cases K1 and K2, this Kd value is more defensible than the value of 

1,000 mL/g. 

In the PA cases (i.e., cases other than Cases K, K1, and K2), a Tc Kd value of 0.8 mL/g was 

assumed for the sorption of Tc to saltstone under oxidizing conditions. A value of 0.8 mL/g was 

also assumed in Case K1, while Cases K and K2 use a value of 10 mL/g (Table 2.7-3). 

Measurements of the Tc Kd value for simulated saltstone under oxidizing conditions show little 

sorption occurring. SRNS-STI-2008-00045 measured average Kd values of 0.25 and 

-0.02 mL/g, respectively, for partially oxidized saltstone in Ca(OH)2 and CaCO3 solutions, and 

the corresponding measurements for reducing saltstone under oxidizing conditions were also 

below 1 mL/g. In SRNL-STI-2009-00636, a measured Tc Kd value of 5 mL/g was measured for 

simulated saltstone under oxidizing conditions. This research is consistent with frequent 

observations of Tc mobility in non-reducing cementitious materials (e.g., 

WSRC-RP-2007-01122; Langton, 1988; Serne, et al., 1992). 

In SRR-CWDA-2011-00044 response SP-19, DOE stated that the new oxidizing saltstone Tc Kd

of 10 mL/g used for Cases K and K2 was based on a single measurement of 12 mL/g for 

adsorption from a saltstone core from Vault 4 (SRNL-STI-2010-00667). The saltstone core 

sample was only partially oxidized, meaning that blast furnace slag was likely still present in the 

solid. The staff notes that Tc release is affected not only by the redox character of the water 

used in the experiment; the effect of any remaining blast furnace slag in the saltstone could be 

to continue to sequester a fraction of the Tc even if the water itself is oxidizing. Because the 

conceptual model for saltstone Tc release assumes that oxidizing conditions are in effect when 

all the reducing agent (i.e., blast furnace slag) in the grout is exhausted, experiments that use 

saltstone still containing blast furnace slag do not faithfully reflect the oxidizing conditions 

simulated in the PA. This caveat applies also to the interpretation of other Tc release 

experiments such as column or monolith leaching (e.g., Langton, 1986, 1988; Harbour and 

Aloy, 2007; Gilliam, et al., 1990; Pabalan, et al., 2010). 

Based upon the above information (as summarized in Table 2.7-3), the NRC staff concludes 

that a Tc Kd value of 0.8 mL/g is more appropriate than 10 mL/g for saltstone under oxidizing 

conditions. The NRC staff therefore concludes that of Case K, K1, and K2, the case that has 

the most appropriate Tc Kd values is Case K1. 
104 

Table 2.7-2: Saltstone Kd Value for Tc under reducing conditions 

Kd Value 

(mL/g) Notes 

Modeled Values 

Case A 1000 

Case K 1000 

Case K1 500 

Case K2 500 

Measured Kd Values for Simulated Saltstone 

SRNS-STI-2008-00045 6.5 to 91.3 

Measurements made in solutions of Ca(OH)2 and 

CaCO3 that were purged with N2. DOE noted that 

the quality of the spectra in these measurements 

was compromised by a U-233 shift. 

SRNL-STI-2009-00636 

32 (2)

(9.1 to 56) 

Experiments performed in a glovebox with a 2% 

H2 environment. Measurement taken after 4 d and 

may not have reached steady state.1

SRNL-STI-2010-00668 

711 (2) Experiments performed in a glovebox with a 2% 

H2 environment. Measurement taken after 22 d.1

518 (2) Experiments performed in a glovebox with a 2% 

H2 environment. Measurement taken after 56 d.1

SRNL-STI-2011-00716 

1258 (2)

(757 to 1759)

Experiments performed in 0.1% H2 atmosphere in 

the presence of a palladium catalyst to convert O2

to water through reaction with H2). Measurement 

taken after 56 days.1

Measured Kd Values for Saltstone Core Sample

SRNL-STI-2010-00667 139 

Leached under N2 using a Ca(OH)2 solution. DOE 

noted that the nitrogen glovebag contained small 

amounts of O2. Measurement taken after 20 d. 

1

 Measurements performed using a solution of CaCO3 unless otherwise reported 2

 Measurements performed in an environment containing H2 (g) 

The Tc Kd value for saltstone under reducing conditions is one of the most important modeling 

parameters in the PA for the SDF. Because the NRC staff does not find that the value DOE 

used in the PA is adequately justified, the NRC staff will monitor the development of additional 

information regarding sorption-reduced saltstone (in addition to sorption in oxidized saltstone, as 

previously discussed). Also, the same Kd values were assumed for the disposal unit concrete 

as for the saltstone. Because the disposal unit concrete has a smaller fraction of slag than 

saltstone, it is expected that the sorption will be less for the disposal unit concrete. The NRC 
105 

staff will therefore also monitor the development of additional justification for sorption in reduced 

disposal unit concrete (in addition to the Kd in oxidized concrete). 

Table 2.7-3: Saltstone Kd Value for Tc under oxidizing conditions 

Kd Value 

(mL/g) Notes 

Modeled Values 

Case A 0.8 

Case K 10 

Case K1 0.8 

Case K2 10 

Measured Kd Values for Simulated Saltstone

SRNS-STI-2008-00045 

0.16 to 0.93 Measurements made in solutions of Ca(OH)2 and 

CaCO3. 

-0.02 to 0.25 

Simulated saltstone was partially oxidized prior to 

experiment. Measurements made in solutions of 

Ca(OH)2 and CaCO3. 

SRNL-STI-2009-00636 5.0 

(3.1 to 6.3) 

Measurement taken after 4 d and may not have 

reached steady state.1

Measured Kd Values for Saltstone Core Sample

SRNL-STI-2010-00667 12 Leached using a Ca(OH)2 solution. Purged with 

air continuously. Measurement taken after 20 d. 

1

 Measurements performed using a solution of CaCO3 unless otherwise reported. 

Iodine Kd Value for Cementitious Materials

The Kd values DOE assumed for I in the performance assessment are presented in Table 2.7-4. 

Cementitious material I Kd values were the subject of an NRC RAI (NRC, 2010b, i; SP-14). The 

main NRC concerns were that laboratory measurements suggesting no I sorption were 

neglected and that grout formulations used in experiments differed substantially from saltstone. 

The reducing middle age value (9 mL/g) initially appears reasonable in light of experimental 

data in SRNS-STI-2008-00045, and as discussed in the DOE response in comment SP-14 in 

SRR-CWDA-2011-00044, a later study using more appropriate saltstone and concrete samples 

that yielded values similar to 9 mL/g (SRNL-STI-2009-00636). The NRC staff notes, however, 

that the short-term I experiments presented in SRNL-STI-2009-00636 and cited in the comment 

SP-14 response (SRR-CWDA-2010-00033) appear to have been considered unreliable in the 

later compilation in the Lilley (2010) thesis. The long-term reducing experiments gave 

somewhat lower I Kd values of less than 5 mL/g (Lilley, 2010, Figure 4.26). Similarly, data on 

cementitious materials under oxidizing conditions included in SRNL-STI-2009-00636 were later 

not included in Lilley (2010). These data were not cited as the source for the oxidized values in 

the PA (Table 4.2-18); rather, the oxidized middle and old age values of 15 and 4 mL/g, 
106 

respectively, were ascribed to Kaplan and Coates (2007) (WSRC-RP-2007-01122). The Lilley 

(2010, Figure 4.23) long-term I experiments on more appropriate cement and saltstone samples 

yielded Kd clustering in the range 2 to 8 mL/g—lower than the adopted oxidized middle age 

value of 15 mL/g. 

Table 2.7-4: Kd values (mL/g) assumed by DOE for I for cementitious materials 

Young Aged Middle Aged Old Aged 

Reducing Conditions 

PA Cases (i.e., cases other 

than Cases K, K1, and K2) 5 9 0 

Cases K, K1, and K2 5 9 4 

Oxidizing Conditions 

PA Cases (i.e., cases other 

than Cases K, K1, and K2) 8 15 4 

Cases K, K1, and K2 8 15 4 

From Table 4.2-18 in the PA and Table PA-8.6 in SRR-CWDA-2011-00044

Bold values indicate Kd values that were used in the PORFLOWTM model 

The NRC staff notes that literature data on I retention are variable. Fuhrmann, et al. (2006) 

found very low I Kd values (0.7 and 0.8 mL/g) for fresh reducing grouts under air and showed 

that Portland cement was the main I sorptive material. Other sorption and diffusion studies 

have found that I can be effectively partially retained on cementitious materials (e.g., Atkins and 

Glasser, 1990; Bonhoure, et al., 2002; Mattigod, et al., 2001). 

The NRC staff concludes that the Kd values assumed for I for cementitious materials seem to be 

slightly higher than measured values. This could result in a minor underestimation of the 

predicted dose. The NRC staff notes that model support for the Kd values in the PA is important 

and NRC staff will be monitoring this under PA maintenance (i.e., NRC staff will monitor the Kd

values selected for I, and their related model support, in the next revision to the PA). 

Radium Kd Value for Cementitious Materials

The Kd values assumed by DOE for Ra for cementitious materials in the performance 

assessment are listed in Table 2.7-5. Whereas the Ra cementitious material Kd values for 

oxidizing conditions are based on the literature, the reducing values in the PA are set equal to 

Sr values. The contrasting sources of information are responsible for the unexpectedly large 

differences in the two sets of values for an element that is generally insensitive to oxidation 

state. In the DOE responses to the comment SP-14 (SRR-CWDA-2010-00033 and 

SRR-CWDA-2011-00044), DOE explained the source of the contrast and stated that they will 

consider future studies on the transport properties of this potentially risk-significant element. 
107 

The NRC staff agrees that the potential future studies would be useful, particularly because 

directly applicable literature data on Ra are lacking. 

Table 2.7-5: Kd values (mL/g) assumed by DOE for Ra for cementitious materials 

Young Aged Middle Aged Old Aged 

Reducing Conditions 

PA Cases (i.e., cases other 

than Cases K, K1, and K2) 0.5 3 20 

Cases K, K1, and K2 100 100 70 

Oxidizing Conditions 

PA Cases (i.e., cases other 

than Cases K, K1, and K2) 100 100 70 

Cases K, K1, and K2 100 100 70 

From Table 4.2-18 in the PA and Table PA-8.6 in SRR-CWDA-2011-00044

Bold values indicate Kd values that were used in the PORFLOWTM model

In SRR-CWDA-2011-00044 (Table PA-8.6) DOE presented Case K Kd values for Ra on 

reducing cementitious materials that are much higher than the previous values. The young Kd

was increased from 0.5 to 100 mL/g, the middle value from 3 to 100 mL/g, and the old value 

from 20 to 70 mL/g. These changes made the reducing Kd values identical to the oxidizing Kd 

values (PA; Table 4.2-18). As discussed in the SP-14 comment and response 

(SRR-CWDA-2011-00044), Ra is not expected to be redox sensitive and the reducing values 

were previously based only on Sr data. This approach is reasonable, but it must also be 

pointed out that there appears to still be no experimental Ra sorption data on saltstone or a 

similar formulation. Bayliss, et al., (1989) measured Kd values of 860 mL/g and higher on a 

mixture of ordinary Portland cement, blast furnace slag, and limestone aggregate. The inclusion 

of the aggregate and the absence of fly ash call into question the applicability of the data. 

In Case A, and many of the other cases, the dose from Ra-226 dominates the dose. The dose 

from Ra-226 in Case K was much smaller due to this case including a much smaller inventory of 

Ra-226 and its ancestors, Th-230 and U-234. 

Because of the potential risk significance of this radionuclide and the lack of applicable data, the 

NRC staff will monitor measurements made of the Kd of Ra for saltstone and for reducing 

disposal unit concrete. However, as discussed in Section 2.2, the inventory of Ra-226 and its 

ancestors is uncertain. If strong justification is provided for the reduced inventory, then this 

radionuclide might no longer be risk significant and this monitoring factor might no longer be 

necessary. 

Selenium Kd Value for Cementitious Materials

In the PA a Se Kd value of 300 mL/g was assumed for young age, middle age, and old age 

reduced cementitious materials (Table 2.7-6). However, a footnote to this table states that it 
108 

was discovered after the PORFLOWTM analysis were completed that the old aged Kd for 

reduced cementitious materials should have been 150 mL/g. In Cases K, K1, and K2, the Se Kd

for old aged reduced cementitious materials was revised from 300 to 150 mL/g. The previous 

value for young and middle aged material of 300 mL/g was retained. The basis for these values 

is provided in SRNL-STI-2009-00473 (Table 17), which refers to selenate sorption 

measurements (250 to 930 mL/g) by Johnson, et al. (2000). However, that study actually 

concluded that its measurements were for selenite, a more reduced form of aqueous Se. The 

Johnson, et al. (2000) data are, therefore, reasonable to consider for moderately reducing 

conditions. A number of studies (Sugiyama and Fujita, 1999; Ochs, et al., 2002; Baur and 

Johnson (2003); Pointeau, et al. 2004) support the strongly sorptive behavior of selenite onto 

cementitious materials, with typical Kd values of 100 mL/g and higher. Based on this 

information, the NRC staff concludes that the Se Kd values assumed for reduced cementitious 

materials appear to be reasonable. 

Table 2.7-6: Kd values (mL/g) assumed by DOE for Se for cementitious materials 

Young Aged Middle Aged Old Aged 

Reducing Conditions 

PA Cases (i.e., cases other 

than Cases K, K1, and K2) 300 300 300 

Cases K, K1, and K2 300 300 150 

Oxidizing Conditions 

PA Cases (i.e., cases other 

than Cases K, K1, and K2) 300 300 150 

Cases K, K1, and K2 300 300 30 

From Table 4.2-18 in the PA and Tables PA-8.6 and PA-8.7 in SRR-CWDA-2011-00044 

Bold values indicate Kd values that were used in the PORFLOWTM model

However, the NRC staff does not find that there is an adequate technical basis for the Se Kd

values of 300 and 150 mL/g for oxidized middle and old age conditions, respectively. The DOE 

response (SRR-CWDA-2011-00044, comment C-4) to a follow-up to an original RAI response 

(SRR-CWDA-2010-00033, comment C-4) did not provide sufficient information to show that 

selenite would be the important species under oxidizing conditions or that selenate, the 

expected oxidized species, would have Kd values in excess of the 30 to 79 mL/g range 

measured by DOE for simulated saltstone and FDC concrete (SRNS-STI-2008-00045). The 

NRC staff notes that, in Cases K, K1, and K2, DOE used an oxidized old age Se Kd of 30 mL/g 

(SRR-CWDA-2011-00044, Table PA-8.7) and Se-79 did not become an important dose 

contributor. However, in these cases, the saltstone did not transition to old age within 20,000 

years, and the disposal unit concrete did not transition to old age until 7,300 to 10,700 years. 

DOE did not test the sensitivity of Se dose to a lower Se Kd for oxidized middle age conditions. 

The NRC staff therefore concludes that better supported information is needed for the Kd for Se 

for saltstone and disposal unit concrete under oxidizing conditions. The NRC staff will monitor 

the development of this information. 
109 

Strontium Kd Value for Cementitious Materials

The Sr Kd values assumed in the original PA cases (Table 2.7-7) were based on site specific 

measurements as documented in WSRC-STI-2007-00640. The NRC staff finds that these 

values are appropriate. 

Table 2.7-7: Kd values (mL/g) assumed by DOE for Sr for cementitious materials 

Young Aged Middle Aged Old Aged 

Reducing Conditions 

PA Cases (i.e., cases other 

than Cases K, K1, and K2) 0.5 3 20 

Cases K, K1, and K2 (disposal 

unit concrete) 15 15 5 

Cases K, K1, and K2 

(saltstone) 1000 1000 NA 

Oxidizing Conditions 

PA Cases (i.e., cases other 

than Cases K, K1, and K2) 3 30 15 

Cases K, K1, and K2 (disposal 

unit concrete) 15 15 5 

Cases K, K1, and K2 

(saltstone) 1000 1000 NA 

From Table 4.2-18 in the PA and Tables PA -8.5, PA-8.6, and PA-8.7 in SRR-CWDA-2011-00044

Bold values indicate Kd values that were used in the PORFLOWTM model

In Cases K, K1, and K2, the Sr Kd values for reducing and oxidizing cementitious materials were 

revised (Table 2.7-7). This revision was based on recommended values in 

SRNL-STI-2010-00667. This report cited the same site-specific measurements that were used 

to determine the Kd values in the original PA calculations, so it is not clear to the NRC staff what 

the basis was for changing the recommended values. The NRC staff also notes that the revised 

value of 15 mL/g for middle aged reduced materials is higher (i.e., less conservative) than the 

site-specific measured value of 2.9 mL/g (WSRC-STI-2007-00640; Table 11). 

In Cases K, K1, and K2, the Kd values for Sr for saltstone were also changed upward on the 

order of two to three orders of magnitude to 1,000 mL/g. The new value was based on 

desorption experiments on an actual ground saltstone core sample from Vault 4, which had 

evidence for partial oxidation (SRNL-STI-2010-00667). The measured Kd values, were 5,728 

and 737 mL/g for oxidizing and reducing conditions, respectively. The measurements of 

desorption Kd values on an actual Vault 4 saltstone sample provide an important source of 

information on radionuclide release for performance assessment. However, it is not clear to the 

NRC staff that sufficient data have been obtained to support using the derived Sr values in 

performance assessments. For example, the investigators speculate that the dissolved Sr 

measurements may, in fact, be controlled by SrSO4 solubility rather than sorption. There is a 
110 

suggestion in the text that SrSO3 could also be a controlling solid, perhaps under more reducing 

conditions. In any case, if the amount of desorption observed in a leaching experiment is 

controlled by solubility, the experimental artifact will result in the calculated Kd value calculated 

being artificially high. Furthermore, in the measurement of the Kd for oxidizing conditions, the 

saltstone sample was only partially oxidized (SRNL-STI-2010-00667), meaning that blast 

furnace slag was likely still present in the solid. Because the conceptual model for saltstone 

radionuclide release assumes that oxidizing conditions are in effect when all the reducing agent 

(i.e., blast furnace slag) in the grout is exhausted, experiments that use saltstone still containing 

blast furnace slag do not necessarily reflect the oxidizing conditions simulated in the 

performance assessment. 

The revised Kd value for Sr for saltstone has the potential to significantly limit the modeled Sr 

release from the saltstone. Because of the importance of Sr releases to the potential dose to an 

inadvertent intruder (Chapter 3), the NRC staff will monitor the development of additional 

measurements for the sorption or leaching of Sr from saltstone. Additionally, the NRC staff will 

monitor the basis for the selection of the reduced middle age Kd value for the disposal unit 

concrete in Cases K, K1, and K2 as part of PA maintenance (Appendix A). 

2.7.4.3 NRC Evaluation - Soil Sorption Coefficients (Kd values) 

The NRC staff review of the soil Kd values used for the saltstone performance assessment (PA; 

Table 4.2-15 and SRR-CWDA-2011-00044; Table PA-8.4) focused on elements considered 

potentially risk significant. The NRC staff finds that the overall DOE approach to establishing Kd 

values for soil is appropriate (i.e., employing site-specific data when possible and relying on 

literature data when site-specific data are lacking or sparse). However, the NRC staff notes that 

it is preferable to use site-specific Kd values for radionuclides that are potentially risk significant 

due to the large variability in Kd values from site to site. 

The Tc Kd values in PA Table 4.2-15 were based on the analysis in SRNL-TR-2009-00019, 

which relied on the median of existing data and more recent Kd measurements on 24 samples 

from different depths in an E-Area borehole (SRNS-STI-2008-00286). The PA values of 1.8 and 

0.6 mL/g for clayey and sandy soils, respectively, appear reasonable in light of the new 

borehole dataset; in fact, these values are lower than nearly all the SRNS-STI-2008-00286 

measurements. 

In the PA calculations (i.e., in cases except for Cases K, K1, and K2), a Kd value of 0 mL/g (i.e., 

assumption of no sorption) was used for I in vadose/sandy soil. The NRC finds this assumption 

of no sorption to be appropriate for these potentially mobile elements in the absence of sitespecific data for vadose/sandy soil. The backfill/clayey soil Kd value of 0.6 mL/g for I was 

reasonably based on site-specific data. In Cases K, K1, and K2, for I, the soil Kd values were 

changed from 0 to 0.3 mL/g for sandy soil, and from 0.6 to 0.9 mL/g for clayey soil 

(SRR-CWDA-2011-00044, Table PA-8.4). DOE cited recent studies conducted with I and SRS 

soils (SRNL-STI-2009-00473, Table 16) as the basis for the revised values. This report 

describes research performed by Schwehr, et al. (2009) that showed that I sorbs to SRS soils 

more readily at lower aqueous I concentrations (i.e., those more reflective of environmental 
111 

conditions) and that more sorptive organo-I and iodate can be more abundant at low I 

concentrations. These observations support the somewhat higher Kd values for I assumed in 

Cases K, K1, and K2. 

The NRC staff finds that the Sr Kd soil values are appropriately based on site data as described 

in WSRC-TR-2006-00004. In this reference it is noted that other measurements of the Sr Kd on 

site found a very strong pH effect on the Sr Kd values, with lower Kd values being measured at 

lower pH values. Because the SDF will have large amounts of cementitious materials that are 

highly alkaline, NRC staff expects that the pH values downgradient of the SDF will not be low. 

In the original PA calculations, the Ra soil Kd values were set equal to the Sr Kd values using the 

chemical homologue argument. The Ra values are reasonable because subsequent sitespecific measurements yielded significantly higher values (SRNL-STI-2010-00527). The 

revised sorption dataset used in Cases K, K1, and K2 also incorporates increases in Ra Kd 

values, with the sandy value changed from 5 to 25 mL/g and the clayey value from 17 to 

185 mL/g (SRR-CWDA-2011-00044, Table PA-8.4). The changes were recommended in 

SRNL-STI-2011-00011 on the basis of measurements reported in SRNL-STI-2010-00527. The 

two Ra Kd values resulted from measurements at an ionic strength of 0.02 M (i.e., the 

approximate ionic strength of SRS groundwater) that included added Sr and considered a range 

of starting aqueous Ra concentrations. While there was some scatter in the sorption isotherms 

(SRNL-STI-2010-00527, Figures 3-1 and 3-2), the results are appropriately interpreted as 

central value numbers based on site-specific analyses. 

Se Kd values of 1,000 mL/g for both clayey and sandy soils were the subject of an NRC RAI 

(NRC, 2010b, i; FFT-3). In the RAI, NRC staff noted this Kd value was representative of low-pH 

soil that the Se sorption is less at neutral and high pH values. In the initial response to the RAI, 

DOE provided the results of a sensitivity analysis performed using the probabilistic GoldSim®

model in which the Se Kd values for the sandy and clayey soils set to 0 mL/g 

(SRR-CWDA-2010-00033). The results of this sensitivity analysis showed that changing the Se 

Kd values resulted in less than a 3% increase in the peak dose within 20,000 years. However, 

although 3% represents a small absolute increase in dose, it represents a large relative 

increase in the dose derived from Se-79. Also, as discussed in more detail in Section 2.11.4.1, 

the NRC staff has number of concerns about the GoldSim® model. In the response to the 

second RAI (SRR-CWDA-2011-00044), DOE provided additional information regarding the pH 

dependence of the Se Kd for soils and the pH in monitoring wells in Z area. DOE stated that Se 

Kd values will decrease sharply as the pH increases above pH 6; and decrease an order of 

magnitude as the pH value approaches 7. DOE further noted that individual pH readings for 72 

samples collected over a 5-year time period from 9 wells ranged from 4.0 to 7.8 and that out of 

the 72 measured pH values only six readings were above a pH of 7. However, a total of 30 out 

of the 72 samples have pH values that are greater than or equal to 6. Furthermore, the pH of 

water in the near field may be elevated by the presence of saltstone and the cementitious 

disposal units. It is therefore not clear to the NRC staff that there is a basis for assuming that 

the soils in Z area will have a low pH. Additionally, the NRC staff notes that the multiple 

measurements of a Se Kd of 1,041 mL/g in the original report cited by DOE for the Se Kd values 

(WSRC-STI-2006-00037; Table 6) appears unusual and is suggestive of an experimental 
112 

artifact that has not been explained. Based on above information, the NRC staff finds that the 

basis for choosing a Kd representative of low-pH soil as compared to a more neutral soil is 

unclear, especially in light of the potential for alkaline buffering of the vadose zone soils by the 

significant quantity of cementitious materials in the SDF. The NRC staff will therefore monitor 

the Kd value assumed for Se in sand and clay soils (Appendix A). 

2.7.4.4 NRC Evaluation of PORFLOWTM Modeling of Near-Field Flow 

The NRC staff reviewed DOE’s near-field model constructed with the PORFLOWTM modeling 

code and generally found DOE’s approach to near-field model to be reasonable. However, the 

NRC staff found that, in several cases, parameter values and modeling assumptions resulted in 

severely constrained flow through saltstone. Specifically, in Vault 4, the Darcy velocity through 

saltstone in Case A remains below 0.05 cm/yr for more than 10,000 years after closure 

(Figure 2.7-1). The flow through the saltstone matrix remains even lower in the Synergistic 

Case, in part because almost all of the water that reaches saltstone is predicted to flow through 

fractures (Figure 2.13-4). In the FDCs, the Darcy velocity through saltstone in the Increased 

Hydraulic Conductivity Case and Case E remain less than 0.1 cm/yr for 10,000 years after 

closure, and the flow through saltstone in Case A remains less than 0.01 cm/yr for 10,000 years 

after closure. As in Vault 4, the flow through the saltstone matrix in the Synergistic Case is 

similar to the flow through saltstone in Case A because almost all of the water moving through 

saltstone in the Synergistic Case flows through fractures. 

Figure 2.7-1: Vertical Darcy velocity through saltstone in Vault 4 predicted with the DOE 

PORFLOWTM model. Data for the Synergistic Case excludes fractures (other cases do 

not represent fractures explicitly). Data taken from DOE’s STAT.out files (NRC, 2010g). 

1E-4

1E-3

1E-2

1E-1

1E+0

1E+1

1E+2

0 5000 10000 15000 20000

Darcy Velocity (cm/yr)

Time (years)

Infiltration (except 

Synergistic Case)

Case K/K1/K2

Case E

Increased 

Hydraulic 

Conductivity Case

Synergistic Case 

(matrix only)

Case A
113 

Figure 2.7-2: Vertical Darcy velocity through saltstone in an FDC predicted with the DOE 

PORFLOWTM model. Data for the Synergistic Case excludes fractures (other cases do 

not represent fractures explicitly). Data taken from DOE’s STAT.out files (NRC, 2010g). 

As discussed in more detail in Sections 2.5 and 2.6, the NRC staff found that the assumed low 

hydraulic conductivities of saltstone and the disposal unit concrete used in Case A and many of 

the deterministic sensitivity cases were overly optimistic. Similarly, assumptions regarding the 

performance of the lateral drainage layer, the disposal unit roofs, and the HDPE/GCL layer (for 

the FDCs) resulted in a significant amount (i.e., greater than 99% of the infiltrating water being 

shed around the disposal units even 10,000 years after site closure in Case A and several of the 

deterministic sensitivity analyses (Section 2.5) (SRR-CWDA-2011-00044; IEC-8). As discussed 

in Sections 2.5 and 2.6, the NRC staff finds that Cases K, K1, and K2 more realistically 

represent the current and expected future hydraulic and physical properties of the saltstone and 

disposal units. In these cases, the modeled flow through the saltstone and disposal units does 

not appear to be inappropriately constrained (Figures 2.7-1, 2.7-2, and 2.13-1). 

The NRC staff also found that the moisture characteristic curves DOE used in the PA cases 

(i.e., cases other than Cases K, K1, and K2) to represent flow through unsaturated cementitious 

porous media appear to constrain flow unrealistically. In particular, in the PA cases, DOE 

assumes the relative permeabilities of intact saltstone and fractured cementitious materials 

decrease by several orders of magnitude with minor decreases in moisture content 

(NRC, 2010b; SP-3 and SP-4) (Figures 2.7-3 and 2.7-4). As a result, small decreases in the 

saturation of the material cause unrealistically large decreases in the modeled flow. The large 

decrease in the relative permeability with decreasing saturation appears to be unjustified 

because the moisture characteristic curves assumed in the PA for intact cementitious material 

are substantially different from those found in the literature (NRC, 2010b; SP-3). For this 

reason, the NRC staff is concerned that the large decreases in the modeled flow with small 

1E-4

1E-3

1E-2

1E-1

1E+0

1E+1

1E+2

0 5000 10000 15000 20000

Darcy Velocity (cm/yr)

Time (years)

Infiltration (except 

Synergistic Case)

Case K/K1/K2

Case E

Increased Hydraulic 

Conductivity Case

Synergistic Case 

(matrix flow only)

Case A
114 

changes in saturation could lead to an underestimate of radionuclide release from the waste 

form and an underestimate of dose. 

In response to NRC’s concern about the moisture characteristic curves in intact saltstone, DOE 

agreed that the moisture characteristic curve utilized in the PA for intact saltstone is inconsistent 

with the literature (SRR-CWDA-2010-00033; SP-3). To evaluate the impact of moisture 

characteristic curves, DOE performed a sensitivity case based on Case A with the relative 

permeability for intact saltstone fixed at 1.0 (i.e., the hydraulic conductivity was not reduced 

based on moisture characteristic curves) for the most risk significant radionuclides. The 

resulting contaminant release rate was approximately twice that of the original Case A result for 

an FDC within 10,000 years after site closure. For Vault 4, with the relative permeability equal 

to 1.0, the release rate of Tc-99 was almost doubled, while the I-129 and Ra-226 rates 

increased by less than 30% as compared to the original Case A values. DOE stated that these 

increases in release rates would not significantly impact the resulting dose to the member of the 

public. In addition, DOE performed a sensitivity case based on the Synergistic Case with a 

revised saltstone relative permeability of 1.0 and concluded that the sensitivity case showed no 

significant impact on the resulting dose (SRR-CWDA-2010-00033, VP-2). Based on these 

analyses, DOE concluded that the moisture characteristic curve used in the PA for intact 

saltstone does not appreciably impact the estimated dose to the member of the public. 

Figure 2.7-3: Moisture Characteristic Curves Adapted from the PA and 

WSRC-STI-2006-00198 (SRR-CWDA-2010-00033; Figure 1) 
115 

Figure 2.7-4: Characteristic curves for the intact and fractured saltstone and concrete as 

adapted from the 2009 PA and SRNL-STI-2009-00115 (SRR-CWDA-2010-00033; Figure 2) 

However, as NRC indicated in response to these DOE analyses (NRC, 2010i; SP-3), one-off 

sensitivity analyses may result in insignificant increases in the dose on an absolute basis; 

however, the effect on dose results may be significant when there are many outstanding 

uncertainties that are evaluated on a cumulative basis. In response to the second RAI 

(SRR-CWDA-2011-00044), the DOE provided a more comprehensive analysis, Case K, to 

address the NRC staff's concern with DOE's moisture characteristic curve for intact saltstone as 

well as other concerns. In this analysis, DOE fixed the relative permeability to one for the 

cementitious materials (i.e., moisture characteristic curves were not used). NRC staff finds that 

this approach is reasonable due to the (i) uncertainty in unsaturated flow through cementitious 

materials over long periods of time and (ii) limited amount of reduction in flow that would be 

expected for cementitious materials that are predicted by DOE to be near saturation. Case K 

also addressed NRC staff's concern with the moisture characteristic curves that were used to 

represent unsaturated flow through fractured saltstone and concrete. 

DOE used an analytical approach to develop curves for fractured cementitious materials in the 

PA based on the moisture characteristic curves that were to represent intact saltstone. Figure 2 

of the first RAI response (SRR-CWDA-2010-00033) shows the curves for fractured saltstone 

(used to represent the saltstone grout in Case E) and fracture concrete (used to represent the 

walls of Vaults 1 and 4 in all of the cases in the PA except the Synergistic Case) (Appendix E of 

SRNL-STI-2009-00115; SRR-CWDA-2010-00033). The figure illustrates that a decrease in the 

saturation of the walls of Vaults 1 and 4 from 100% to 99% results in an unrealistic decrease of 

nine orders of magnitude in relative permeability. The NRC staff's concern, as discussed in the 
116 

second comment SP-4 (NRC, 2010i), is that DOE's abstraction of unsaturated fracture flow 

significantly underestimates actual flow rates through a fractured system. As discussed in the 

previous paragraph, Case K addresses NRC staff's concern about the use of use moisture 

characteristic curves to represent fractured cementitious materials by setting the relative 

permeability equal to 1.0. 

2.7.4.5 Source Term Release and Near Field Transport Conclusions 

The NRC staff concludes that DOE’s Case A model appears to underestimate the release of 

radionuclides from saltstone because of overly optimistic assumptions about the physical and 

chemical properties of saltstone and disposal unit concrete. The NRC staff found the 

assumptions related to near-field flow in Cases K, K1, and K2 to be appropriate. However, the 

NRC staff concludes the average-Kd model DOE used to model Tc release in Cases K, K1, and 

K2 appears to delay and overestimate the peak release rate in a manner that makes it difficult to 

interpret the results of these cases. Most significantly, the NRC staff finds the Kd values DOE 

uses to represent Tc sorption in reduced cementitious materials, and the Kd DOE uses to 

represent Tc sorption in oxidized saltstone in Cases K and K2 to be unsupported. Of Cases K, 

K1, and K2, the NRC staff finds that Case K1 has Kd values for Tc for cementitious materials 

that are the most supported. Thus, because Cases K, K1, and K2 differ only in the selected Kd

values for Tc in saltstone and the disposal units, the NRC staff considers Case K1 to best 

represent the system. 

With respect to flow modeling, the NRC’s analysis of DOE’s near-field model indicates that 

Case A appears to severely constrain flow through saltstone (Figures 2.7-1 and 2.7-2). As 

discussed in detail in Section 2.6, the saltstone hydraulic conductivity that contributes to this low 

flow value is inconsistent with recent measurements and does not reflect any degradation of the 

hydraulic conductivity as the saltstone ages. Flow through the saltstone and disposal unit 

concrete in the PA Cases (i.e., cases other than Cases K, K1, and K2) were further limited by 

moisture characteristic curves that are inconsistent with curves found in the literature for similar 

materials. DOE demonstrated that the effect of the suspect moisture characteristic curves is 

limited to a factor of 2 in Case A (Section 2.7.3), however, the effect may be greater if other 

engineered barriers do not perform as modeled. Because of the large uncertainty in the 

potential degradation of saltstone, disposal unit concrete, and the HDPE/GCL layers on the 

FDCs, and because of the importance of these barriers to system performance (Section 2.13), 

the NRC staff will monitor the development of model support for the long-term performance of 

these materials. 

The NRC staff found the pore-volume model DOE used for elements other than Tc to be 

appropriate for less redox and pH sensitive elements. However, as described in Section 2.6, 

the NRC staff found the basis for the number of pore volumes required to cause certain Eh and 

pH transitions in saltstone and disposal unit concrete in the pore-volume model to be insufficient 

in the PA cases. The NRC staff found that the basis for the number of pore volumes assumed 

in Cases K, K1, and K2 to be reasonable. The NRC staff found the explicit shrinking core model 

DOE used to model Tc release from saltstone in PA cases (i.e., all cases but Case K, K1, and 

K2) to be appropriate for redox sensitive elements. However, as discussed below, the NRC 
117 

staff finds the Kd value used for Tc sorption in reducing conditions to be unsupported and 

concludes that the average-Kd model DOE used to represent Tc release in Cases K, K1, and K2 

is inconsistent with DOE’s conceptual model of Tc release. Furthermore, the NRC staff finds 

that the average-Kd model overestimates peak release rates to different extents with different 

parameter values, in a manner that makes it difficult to interpret the model results. That is, the 

extent to which the peak release rate is overestimated is not consistent between different cases 

or readily predictable. 

The NRC staff finds the Kd value used to represent Tc sorption in reduced saltstone and 

disposal unit concrete to be unsupported. The assumed Kd values are based on experiments 

that included a strong reducing agent that is not consistent with the current saltstone formulation 

or expected field conditions. Moreover, the value is inconsistent with DOE’s conclusion that 

only trace levels of oxygen (i.e., 30 to 60 ppm) are required to maintain Tc in its oxidized and 

mobile form. DOE has not provided a basis for assuming that these trace levels of oxygen will 

not be present in the unsaturated soils surrounding saltstone after closure. Furthermore, DOE 

has not shown that the potential oxidation of saltstone during operations (when the saltstone is 

exposed to atmospheric oxygen) is negligible. For these reasons, and because of the 

importance of the Kd value of Tc in reducing cementitious materials to SDF performance 

(Section 2.13), the chemical reduction and sorption of Tc in saltstone is a key monitoring factor. 

Specifically, the NRC staff will monitor the development of model support for the assumption 

that saltstone will chemically reduce Tc(VII) to Tc(IV) to the extent assumed in DOE’s PA model 

under the range of conditions to which saltstone is expected to be subjected during the 

compliance period. 

Similarly, as described in Section 2.7.4.2, the NRC staff finds that the Kd values for cementitious 

materials were not supported for Ra and Se. Additionally, the NRC staff finds that the Kd value 

assumed for Sr for saltstone in Cases K, K1, and K2 were not supported. The NRC staff also 

found that the Kd value for Se for sand and clay soils was not supported. Because of the 

potential risk significance of these radionuclides, NRC staff will monitor the development of 

additional model support for the saltstone Kd value for Ra, Se, and Sr, the disposal unit concrete 

Kd value for Ra and Se, and the sand and clay Kd values for Se. 

2.8 Hydrology and Transport 

2.8.1 Description of Site Hydrology 

The SDF hydrological system and far-field modeling approach and results are discussed in 

several PA sections (Section 3.1.5 “Hydrogeology”, Section 4.2.3.1.3 “Transport Model—

Saturated Zone”, Section 4.2.3.2.5 “Saturated Zone Hydraulic Properties”, Section 4.3.1.2, 

“PORFLOW”, Section 4.4.4.1 “PORFLOWTM Modeling Process”, and Section 5.2.1, 

“Groundwater Concentrations at 100 m”). Data used to define and parameterize the threedimensional groundwater flow and contaminant transport models for the General Separations 

Area (GSA) are found in the GSA database. This database includes elevations for tops of 

hydrostratigraphic units, sediment core descriptions, water levels in the upper and lower zones 

of the Upper Three Runs (UTR) aquifer (or UTR-UZ and UTR-LZ), water levels in the Gordon 

aquifer, and permeability data from laboratory tests, multiple and single well pump tests, and 
118 

slug tests. The location of the SDF on the GSA (Z-Area) and nearby surface water locations are 

depicted in Figure 2.8-1. 

In the area of the SDF, the hydrological system consists of three aquifers of interest: (1) the 

water table or UTR aquifer, which is split into upper and lower zones, (2) the Gordon aquifer, 

and (3) the Crouch Branch aquifer. The UTR and Gordon aquifers are expected to be impacted 

by radionuclides from the SDF facility. Contamination is not expected to affect the deeper 

Crouch Branch aquifer, however, because of an upward flow gradient between the Crouch 

Branch and Gordon aquifers near Upper Three Runs creek. Groundwater flow in the UTR 

aquifer is driven by recharge. In the vicinity of the SDF, groundwater from the UTR aquifer 

discharges or seeps to McQueen Branch to the east or Upper Three Runs creek to the north 

and west. The underlying Gordon aquifer is strongly influenced by its discharge to Upper Three 

Runs creek. The Gordon aquifer is recharged by downward leakage from the UTR aquifer 

above and upward leakage from the Crouch Branch aquifer below. Conceptually, recharge of 

the Gordon aquifer by the Crouch Branch aquifer is very small in comparison to recharge from 

the Upper Three Runs aquifer and is thus neglected in modeling. DOE’s conceptual model for 

flow in affected aquifers on the GSA is presented in Figure 2.8-2. The northwest half of the SDF 

straddles a groundwater divide between Upper Three Runs creek and McQueen Branch 

causing SDF contaminants to discharge to either creek depending on the SDF source location 

(2009 PA, pg 172). 
119 

Figure 2.8-1: Location of Z-Area within the GSA and Surface Water Locations 

(2009 PA, Figure 4.2-1) 
120 

Figure 2.8-2: Conceptual Model for Flow on the GSA near SDF. (2009 PA, Figure 3.1-24 

[Adapted from WSRC-TR-96-0399 Rev. 1, Vol. 2, Figure 13]) 

2.8.2 Modeling of Hydrology and Far-Field Transport 

2.8.2.1 Regional GSA/PORFLOWTM Model 

DOE previously constructed a regional GSA groundwater flow model to provide a common 

framework to perform various hydrogeological investigations for the GSA as documented in 

WSRC-TR-96-0399, Volumes 1 and 2 using the FACT computer code. The regional GSA flow 

model was migrated to PORFLOWTM version 5.95.0 in 2004 for use in various GSA PAs as 

documented in WSRC-TR-2004-00106. The PORFLOWTM version of the GSA regional flow 

model is hereinafter referred to as the GSA/PORFLOWTM model. Figure 2.8-3 shows a 

perspective view of the regional GSA/PORFLOWTM model. The GSA/PORFLOWTM saturated 

zone flow and transport models are used to simulate flow and contaminant transport of 

constituents released from the SDF to surface water. Seepline concentrations are extracted 

from the GSA/PORFLOWTM model for use in calculating surface water pathway doses in the 

PA. 
121 

The interior areal resolution of the GSA/PORFLOWTM model is 61x61 m2

 (200x200 ft2); 

peripheral grid cells at the margin of the model domain are larger. A maximum of 108-grid cells 

span the domain from east to west, and a maximum of 77 grid cells span the domain from north 

to south. Vertical resolution varies with hydrostratigraphic picks and topography. The UTR-UZ 

aquifer is represented by up to 10 cells in the vertical direction, and the vadose zone is 

considered to be part of this unit. The tan clay-confining zone (TCCZ) is represented by 2 cells 

in the vertical direction, and is assumed to be laterally continuous. The UTR-LZ aquifer is 

represented by 5 cells in the vertical direction. The underlying Gordon Confining Unit (GCU) 

and Gordon aquifer are each represented by 2 cells in the vertical direction. Thus, a maximum 

of 21 grid cells represent the GSA hydrogeology from ground surface to the base of the Gordon 

aquifer. The regional model domain comprises 102,294 cells (2009 PA). 

The hydrostratigraphy of the GSA/FACT and GSA/PORFLOWTM models are similar with some 

notable exceptions (WSRC-TR-2004-00106). Adjustments to the GSA/FACT model mesh were 

necessary to overcome limitations of the PORFLOWTM model in accurately representing the 

velocity field for highly distorted elements present in the FACT model or to accommodate 

specific project needs. Most notably, GSA/PORFLOWTM model layers are assumed to be 

truncated by the ground surface to prevent excessively thin layers that are present in the 

GSA/FACT model. Additionally, GSA/PORFLOWTM model layers above the TCCZ are nonuniformly distributed compared to their counterparts in the GSA/FACT model. Thinner layers 

just above the TCCZ were desirable for E-Area modeling where the water table is located just 

above the TCCZ. The layering below the TCCZ is essentially the same between the two 

models. 

2.8.2.2 Local SDF/PORFLOWTM Model 

The primary focus of the local SDF/PORFLOWTM saturated zone transport model is 

contaminant concentrations at locations 1 m (3 ft) and 100 m (330 ft) from the SDF site 

boundary. These concentrations are used for calculating groundwater pathway doses for the 

intruder (§61.42) and members of the public (§61.41), respectively. The areal resolution of the 

local SDF model is 15×15 m2

 (50×50 ft2). In the horizontal plane, each GSA/PORFLOWTM 

regional model grid cell is divided four ways in each coordinate direction into 16 local SDF 

transport model grid cells; vertical resolution, however, is preserved. This constitutes a 4×4×1 

mesh refinement. 
122 

Figure 2.8-3: Perspective View of Regional GSA/PORFLOWTM Flow and Transport Model 

Domain. (2009 PA, Figure 4.2-9 [taken from WSRC-TR-2004-00106, Figure 2-3]) 

The velocity field for the local SDF/PORFLOWTM model is generated with a mass-conserving 

linear interpolation scheme directly from the regional GSA/PORFLOWTM velocity model; thus, 

the local SDF/PORFLOWTM model does not require a separate flow model with its own 

boundary conditions and properties. Within the lateral confines of the local SDF/PORFLOWTM 

model, the velocity field includes the complete vertical extent of the regional GSA/PORFLOWTM 

model. Figure 2.8-4 illustrates the 100-m point-of-compliance used to evaluate the §61.41 

performance objectives, as well as particle tracks to provide information on groundwater flow 

directions and rates for conservative species released at various SDF sources. 
123 

Red Streamlines Represent Each Plume Centerline. A Series of Five-Year Time Markers 

is Noted with Blue Dots on the Red Streamlines. Alternating Blue and Green Dots Mark 

100-m Compliance Boundaries to Indicate Evaluation Sectors [Results are Binned by 

Sector]. 

Vault 1 Particle Tracks Vault 4 Particle Tracks 

Northern FDC Particle Tracks Southern FDC Particle Tracks

Figure 2.8-4: SDF/PORFLOWTM Saturated Zone Model with 100-m (330-ft) Compliance 

Evaluation Sectors (2009 PA, Figures 5.2-2 though 5.2-4) 

2.8.2.3 Boundary Conditions 

The GSA/PORFLOWTM saturated zone model domain encompasses the GSA and its surface 

water discharge points. Streams define three lateral domain boundaries: Upper Three Runs 

forms the northern boundary; McQueen Branch forms the eastern boundary; and Four Mile 

Branch forms the southern boundary (Figure 2.8-1). Four Mile Branch and McQueen Branch 

provide natural, no-flow boundary conditions for the Upper Three Runs aquifer (Figure 2.8-5). 

This aquifer unit is absent at the northern model boundary due to Upper Three Runs stream 

incision. The western boundary is arbitrarily chosen where hydraulic head values from a 

contour map of measured water elevations are prescribed. The Gordon aquifer only discharges 

to Upper Three Runs creek; Upper Three Runs creek provides a natural no-flow boundary for 

the Gordon aquifer on the north face of the model. Gordon aquifer hydraulic head values from a 

contour map are prescribed over the west, south and east faces of the model. The grid cells 
124 

that represent the base of the Gordon aquifer are prescribed a general head boundary 

condition; upwelling recharge from the Crouch Branch aquifer below is considered negligible. 

The upper surface of the PORFLOWTM saturated zone model is prescribed a spatially variable 

recharge/drain boundary condition (implemented using a prescribed flux boundary with the flux 

being a function of pressure head) with an average recharge rate of 37.3 cm/yr (14.7 in/yr) 

(WSRC-TR-2004-00106, pg 14 and Figure 3-7) representative of non-capped GSA sediments. 

Figure 2.8-5: GSA/PORFLOWTM Model Boundary Conditions 

Contaminant fluxes that exit the individual vadose zone transport models (i.e., one each for 

disposal cell) become an upper boundary condition for the local SDF saturated zone model with 

fluxes from the representative Vault 2 (or FDC) vadose zone model multiplied by 64 and loaded 

homogeneously under the footprints of the 64 individual FDCs. Each water table flux 

contribution from an individual SDF disposal unit is assigned to the aquifer transport grid by 

uniformly distributing the flux to those water table cells with centroids lying within the footprint of 

the disposal unit. Inspection of PORFLOWTM modeling files reveals that the fewest and largest 

number of elements representing the footprint of an individual FDC is four and nine, 

respectively, with the overwhelming majority of FDCs represented by seven elements. 
125 

2.8.2.4 Material Properties and Parameters 

The regional GSA/PORFLOWTM model used for the PA generally inherited its parameter set 

from that of the predecessor GSA/FACT model. The GSA/FACT parameter set was derived 

from the GSA database of hydrologic data, which includes hydraulic conductivity data from 

laboratory tests, pump tests (multiple and single well tests), and slug tests. The same approach 

used in the GSA/FACT modeling to assign initial hydraulic conductivity fields was also used 

during GSA/PORFLOWTM model construction. 

The initial (uncalibrated) hydraulic conductivity assignments to GSA/FACT model grid cells were 

based upon a correlation between hydraulic conductivity and total mud fraction (both calcareous 

and siliciclastic), where mud is the summation of silt and clay fractions (corrected Figures 

CC-FF-5.1 and 5.2 provided in the FTF RAI responses [SRR-SWDA-2011-00054]). Apparently, 

for both GSA/FACT and GSA/PORFLOW, known core lithologies (e.g., mud fraction) from 

discrete intervals were correlated with known field- or laboratory-measured hydraulic 

conductivities from the same intervals, although DOE does not clearly state this in the FTF RAI 

responses (SRR-SWDA-2011-00054). There were slight differences in the initial definition of 

GSA/FACT and GSA/PORFLOWTM model hydraulic conductivity fields due to differences in the 

model meshes and in the manner in which property assignments are made (i.e., at element 

nodes in FACT versus on element faces in PORFLOW), as described in WSRC-TR-2004-00106 

(pg 5). The initial GSA/PORFLOWTM hydraulic conductivity field was subsequently modified 

during model calibration. Modifications made to the hydraulic conductivity field during 

GSA/PORFLOWTM model calibration were different than those made during GSA/FACT model 

calibration, leading to differences between the two models in terms of the final calibrated 

hydraulic conductivities. 

Given that the vadose zone is included within the regional GSA/PORFLOWTM model mesh, 

water retention characteristic curves are also required. Pseudo-soil water retention 

characteristic curves, which exhibit greater linearity than real soil water curves, are used to 

simulate unsaturated flow from the land surface to the water table under steady-state conditions 

(WSRC-TR-2004-00106; Figure 2-7). Horizontal and vertical saturated hydraulic conductivities 

of cells with computed saturation less than 90 percent are set to 3.5x10−5

 cm/s (0.1 ft/d) to 

minimize lateral flow and thereby ensure that modeled water movement in the vadose zone is 

vertically downward (WSRC-TR-2004-00106, pg 5). 

With respect to transport properties, effective diffusion coefficients (Table 2.8-1) are assigned to 

saturated zone sediments based upon whether they are defined as sandy or clayey 

(WSRC-STI-2006-00198, pg 128). GSA/PORFLOWTM transport modeling assumes an effective 

porosity value (Table 2.8-1) for all aquifers and aquitards to account for dead-end pores that do 

not participate in radionuclide transport (WSRC-STI-2006-00198). Because the Upper Three 

Runs and Gordon aquifer material properties are similar to the lower vadose zone material 

properties, DOE converts average values for lower vadose zone bulk density 1.62 g/cm3

 and 

particle density 2.66 g/cm3

 to effective values for saturated zone bulk density (1.04 g/cm3) and 

particle density (1.39 g/cm3

) using the average total value (39 percent) and effective value (25 
126 

percent) of lower vadose zone porosity (2009 PA, page 218; WSRC-STI-2006-00198, pg 128). 

DOE also assumes these same effective material property values for the TCCZ and GCU. 

Table 2.8-1: Saturated Zone Material Properties DOE used in PORFLOWTM Modeling 

Saturated 

Sediments 

Saturated Effective 

Diffusion Coefficient 

(cm2/s) 

Effective 

Porosity 

(%) 

Effective Dry 

Bulk Density 

(g/cm3) 

Effective 

Particle Density 

(g/cm3) 

Sandy 

Clayey 

5.3x10−6 

4.0x10−6 25 1.04 1.39 

Kd values used for radionuclide transport through the sandy saturated zone are the same as 

values assigned in the PORFLOWTM vadose zone modeling for undisturbed vadose zone 

material. The entire UTR aquifer [i.e., the UTR-UZ, UTR-LZ, and tan clay confining unit] is 

assigned sandy Kd values. Kd values used for radionuclide transport through the clayey 

saturated zone are the same as values assigned in the PORFLOWTM vadose zone modeling for 

backfill. Clayey Kd values are only assigned to the GCU (SRNL-STI-2009-00115, pages 129 - 

130). Kd values are listed in Table 4.2-15 of the PA. 

DOE indicates that hydrodynamic dispersion in the local SDF/PORFLOWTM transport model is 

represented by a (i) longitudinal dispersivity of 10 m (33 ft), (ii) transverse horizontal dispersivity 

of 1 m (3 ft), and (iii) an apparent transverse vertical dispersion of 0 m (0 ft) that are 10, 1, and 0 

percent of a nominal 100 m (330 ft) plume travel distance, respectively. Although DOE 

indicates vertical dispersivity is set to 0 m, DOE also indicates that some numerical dispersion 

occurs, nonetheless (SRNL-STI-2009-00115, pg 130). It is significant to note that PORFLOWTM 

version 6.10.3 (G-TR-G-00002) used for simulation of all cases in the PA only allows 

specification of a single longitudinal and transverse dispersivity, while PORFLOWTM version 

6.30.2 used to simulate Case K in the RAI responses is able to provide separate specification of 

longitudinal and transverse dispersivities in the horizontal and vertical direction. However, the 

dispersivities for longitudinal and transverse vertical for Case K are set at values of 1 and 0.1 m 

(3.28 and 0.328 ft), respectively. Therefore, while it appears longitudinal vertical dispersivity is 

lower in Case K compared to the base case (and all other cases), the transverse vertical 

dispersivity appears to be higher than the value DOE intended to use for the base case in the 

PA. Effective diffusivities for sand and clay materials in the aquifer are assigned values of 

5.3x10-6 cm2

/s (2009 PA; Table 4.4-10). 

2.8.2.5 Model Verification, Calibration, and Validation 

PORFLOWTM acceptance testing on version 5.95.0 used to construct the GSA/PORFLOWTM 

model confirmed that the code conserves mass and satisfies Darcy’s Law 

(WSRC-TR-2004-00106, pg 21). DOE indicates in its PA that software quality assurance for the 

version of PORFLOWTM used for the local SDF/PORFLOWTM calculations (version 6.10.3) is 

covered by WSRC-SQP-A-00028 and G-TR-G-00002. 

The initial hydraulic conductivity fields of the GSA/PORFLOWTM model were modified to the 

final calibrated parameter set (e.g., Figure 2.8-6 and Figure 2.8-7) by matching model results 
127 

with measured hydraulic heads (WSRC-TR-2004-00106, pg 7). Particle tracking simulations 

were also performed to compare groundwater travel times for the GSA/FACT and 

GSA/PORFLOWTM models. The GSA/PORFLOWTM modeled travel times were generally 

longer; the maximum recharge rate was, therefore, increased from 0.46 to 0.48 m (18 to 

19 in/yr) in the GSA/PORFLOWTM model. Relative to the GSA/FACT model, modifications to 

the GSA/PORFLOWTM models that were needed to obtain similar calibration results included: 

• Increasing horizontal hydraulic conductivity in the UTR-UZ by 25 percent 

• Decreasing vertical hydraulic conductivity in the TCCZ by 50 percent 

• Increasing horizontal hydraulic conductivity in the UTR-LZ by 35 percent 

The average calibrated horizontal conductivities in the saturated UTR-UZ, UTR-LZ, and Gordon 

Aquifer are approximately 3.0, 4.0, 11.6 m/d (10, 13, and 38 ft/d), respectively. The average 

calibrated vertical conductivities for the TCCZ and the Gordon Confining Unit are 2.1x10-6 and 

3.5x10-9 cm/s (6.0x10-3 and 1.0x10-5 ft/d), respectively (2009 PA, pg 176). 

DOE evaluated the need to update the GSA/PORFLOWTM model with more recent data 

(2009 PA, pg 177). Two investigations were conducted including (i) comparisons of 

GSA/PORFLOW-generated head contours to hand-drawn contours using data from 1995, 1998, 

and 2003, and (ii) comparison of head residuals using data through 1995 and considering more 

recent data collected through 2006. The investigations suggest that the additional data 

collected since construction of the original GSA/PORFLOWTM model has not significantly 

altered DOE’s understanding of the GSA groundwater system. 

Summary statistics for hydraulic head residuals were provided in the GSA/PORFLOWTM 

documentation (WSRC-TR-2004-00106, Table 3-1); however, no calibration goals were 

specified. GSA/FACT documentation (WSRC-TR-96-0399, Vol 2 Rev. 1) indicates that 

calibration criteria should be no more stringent than the uncertainty in the data. Hydraulic head 

data was reported as having a maximum uncertainty of 1 m (3 ft) (WSRC-TR-96-0399, Vol 2 

Rev. 1). Therefore, a calibration goal of 1 m (3 ft) for the root-mean-square hydraulic head 

residual was set for the regional GSA/FACT model. WSRC-TR-96-0399, Vol. 2 Rev. 1 also 

defines a calibration goal for the maximum head residual as 5 to 10 percent of the total 

hydraulic head variation in a given aquifer. Total hydraulic head variation within the Gordon 

aquifer is given as ~24 m (~80 ft). Therefore, a maximum target head residual would be no 

greater than 1.2 to 2.4 m (4 to 8 ft) for the Gordon aquifer. Total head variation within the UTR 

aquifer is given as ~49 m (~160 ft). Therefore, a maximum target head residual would be no 

greater than 2.4 to 4.9 m (8 to 16 ft) for the UTR aquifer. While no specific calibration goals are 

set for the GSA/PORFLOWTM model, the maximum residual in the UTR-LZ of 8.2 m (27 ft) near 

H-Area in the GSA/PORFLOWTM model greatly exceeds the calibration goal indicated in the 

GSA/FACT model. WSRC-TR-2004-00106 indicates that head residuals for the GSA/PORLOW 

model are somewhat larger than those for GSA/FACT for various reasons (e.g., use of an 

artificial recharge zone near H-Area and finer vertical discretization in the GSA/FACT model) 

and that more extensive calibration efforts would likely improve the GSA/PORFLOWTM model. 
128 

Figure 2.8-6: Calibrated Horizontal Hydraulic Conductivity Assignments to the Regional 

GSA/PORFLOWTM Model (Fourmile Branch is to the South and Upper Three Runs Creek 

is to the North on the GSA [2009 PA, Figure 4.2-10, pg 176]) 

Figure 2.8-7: Calibrated Vertical Hydraulic Conductivity Assignments to the Regional 

GSA/PORFLOWTM Model (Fourmile Branch is to the South and Upper Three Runs Creek 

is to the North on the GSA [2009 PA, Figure 4.2-11, pg 177]) 

With regard to calibration in the area of interest, few calibration targets appear to be located in 

the vicinity of the SDF. Although Figure 4.2-14 in the 2009 PA provides comparisons of handcontoured water table elevations based on monitoring data versus modeled heads (pg 180), 
129 

comparisons of monitored versus modeled heads at the scale of the SDF are not provided and 

visual comparison of heads in the vicinity of SDF reveal significant differences (i.e., water table 

data reveal a greater hydraulic gradient at SDF than modeled). Text in the 2009 PA, reports a 

base case Darcy velocity of 45.7 m/yr (150 ft/yr) in the saturated zone (pg 495) for use in the 

probabilistic assessment, while the 5-year time markers (pages 274 to 276) seem to indicate 

variability across the SDF (e.g., relatively higher velocities are present towards the south of the 

SDF). For the purposes of probabilistic assessment, a normal distribution was selected with a 

mean Darcy velocity of 46 m/yr and standard deviation of 0.5 m/yr (pg 495) and a minimum 

value of 80 percent and a maximum value of 120 percent of the mean (pg 1980). 

Although no specific calibration criteria related to baseflow are provided, WSRC-TR-96-0399, 

Vol. 2 Rev. 1 also suggests that agreement between measured and modeled baseflow is 

acceptable within the uncertainty range of the data or ±20 to 50 percent. GSA/PORFLOWTM 

simulated baseflow values are provided in WSRC-TR-2004-00106 (Table 3-2, pg 15) and are 

reproduced in Table 2.8-2 below. The simulated values appear to be in good agreement given 

the level of uncertainty in the data. McQueen Branch baseflow values are slightly higher than 

the upper bound target of 150 percent or ~4.2 m3

/s (2.25 ft3/s). GSA/FACT and 

GSA/PORFLOWTM stream baseflows are similar with the largest differences occurring for Upper 

Three Runs creek and McQueen Branch (WSRC-TR-2004-00106). 

Table 2.8-2: Comparison of Estimated and GSA/PORFLOW-Modeled Baseflows 

Stream Estimated Baseflow 

m3/s (ft3/s) 

Modeled Baseflow 

m3/s (ft3/s) 

Upper Three Runs & 

Tributaries 0.52 (18.2) 0.32 (11.4) 

Fourmile Branch 0.07 (2.6) 0.11 (3.8) 

McQueen Branch 0.04 (1.5) 0.07 (2.4) 

Crouch Branch 0.05 (1.8) 0.05 (1.7) 

WSRC-TR-2004-00106 

WSRC-TR-2004-00106 (pg 25) provides additional benchmarking information, including 

comparisons of results of particle tracking simulations and tritium transport simulations 

performed for E-Area (WSRC-TR-2003-00432) with the two models. Peak concentrations 

between the models are similar, but individual runs vary on the order of ±25 percent 

(WSRC-TR-2004-00106, pg 25). WSRC-TR-2004-00106, states that particle tracking and 

transport comparisons indicate the velocity field of the two GSA models is similar. 

In the 2009 PA, DOE also compares GSA/PORFLOWTM modeled path lines with E-Area tritium 

plume distribution data and F-Area plume data in map view to show that the regional 

GSA/PORFLOWTM saturated zone model reproduces known plume trajectories. 

GSA/PORFLOWTM model comparisons to data in the vicinity of the SDF are not provided. 

Likewise, no local SDF/PORFLOWTM model comparisons to data are provided. Presumably, 

such data are either not available or too limited to validate the local SDF/PORFLOWTM model. 
130 

2.8.2.6 Point of Compliance 

The NRC guidance found in NUREG-1854 (NRC, 2007b) indicates that after the end of the 

institutional control period, the receptor evaluated to demonstrate compliance with the §61.41 

performance objective is assumed to be located at the point of maximum exposure located 

outside of the disposal site (Section 2.3.2). DOE calculates potential groundwater exposure 

concentrations at a point located 100 m (330 ft) from the SDF boundary using the local SDF 

model. However, because (i) the transport paths away from SDF sources do not occur along a 

straight line, (ii) the transport paths traverse up to three aquifers, and (iii) sources are located at 

a variety of distances from the down gradient edge of the disposal site, the actual travel 

distances to reach a 100 m (330 ft) location beyond the SDF site are, in fact, greater than 100 m 

(330 ft) for many SDF sources. Table 5.2-2 of the PA shows the approximate travel distances 

for constituents released from the SDF vaults and FDCs to the 100 m (330 ft) boundary 

(2009 PA). 

As stated above, DOE assumes the groundwater concentrations 100 m (330 ft) from the SDF 

site are the highest concentrations in the area 100 m (330 ft) or farther from the site. DOE 

indicates that this assumption is supported by Figures 5.2-5 through 5.2-7 in 2009 PA (pages 

389 and 390), which presents the plume (in plan view and cross-sectional maps) that would 

result from a continuous and conservative (no decay or sorption) tracer. Peak concentration is 

observed to decrease monotonically with travel distance from the source zone, as a result of 

hydrodynamic dispersion. DOE indicates that no physical mechanism exists to concentrate 

contamination beyond the source zone in the fully three-dimensional PORFLOWTM simulations. 

Therefore, DOE concludes the 100 m (330 ft) assessment points from the disposal site are 

adequate to capture the peak concentration that can occur at or beyond 100 m (330 ft). 

DOE calculates 100 m (330 ft) concentrations for twelve sectors (Sectors A – L) in the local SDF 

model, as shown on Figure 5.2-1 of the 2009 PA. The peak concentration values for the 100 m 

(330 ft) results are recorded for the three aquifers of concern (i.e., UTR-UZ, UTR-LZ, and 

Gordon aquifers). The concentration for each aquifer represents peak concentration in any 

vertical computational mesh within the aquifer. The vertical thicknesses of the computational 

grid are on average ~2 m (6.6 ft) in the UTR-UZ, and ~3 m (12.7 ft) in the UTR-LZ. Well screen 

averaging was not used to determine the concentrations for dose calculations because the 

typical well screen length of 6.10 m (20 ft) is similar to the thickness of the computational grid 

but conservative with respect to being lower in vertical thickness than the average well screen 

length. Dividing the results into sectors was necessary to allow the large amount of 

concentration data from PORFLOWTM to be stored and used by the GoldSim® dose calculator 

model, and to allow evaluation of variability in peak concentration for different source areas of 

the SDF. The twelve sectors are analyzed for each radionuclide and chemical to find the 

maximum groundwater concentrations 100 m (330 ft) from the SDF. DOE also calculates 1-m 

(3-ft) concentrations for four sectors (Sector 1A – 1L) for the inadvertent intruder dose 

calculations, as shown in Figure 5.2-1 of the 2009 PA. Using the sectors to determine the 

highest groundwater concentrations causes the calculated peak doses to be higher than they 

actually would be, because the peak concentrations are determined for each radionuclide 

independent of the location within the sector. 
131 

2.8.3 NRC Evaluation – Hydrology and Far-Field Transport 

2.8.3.1 NRC Evaluation of Model Construction Including Boundary Conditions 

It is commendable that DOE developed the hydraulic conductivity parameter set from field and 

laboratory characterization data (WSRC-TR-2004-00106), rather than starting with a uniform 

parameter set and adding variation as needed to match measured hydraulic heads and 

estimated stream baseflows. However, the process of migrating the GSA/FACT to the 

GSA/PORFLOWTM model could be more transparent, and as discussed in more detail below, 

calibration in the area of interest could be improved. 

With respect to source loading, DOE appears to distribute the convective flux of contaminants 

out of the near-field model for Vaults 1, 4, and the 64 FDCs to a number of source elements 

located underneath the SDF. DOE intends to flag elements located near the water table with 

centroids located within the footprint of individual SDF sources as source cells for the purposes 

of far-field modeling. With respect to FDCs, DOE appears to distribute the cumulative 

contributions of the 64 FDCs amongst 457 source cells. The minimum number of source cells 

used to model releases from any single FDC is 4 cells for FDC 20C21 and the maximum number 

of source cells used to model releases from any single FDC is 9 cells for FDC 10C. If DOE 

were to change the manner in which disposal unit releases are loaded into the far-field model by 

(i) ensuring that the same number of source cells are used for each FDC, (ii) ensuring that the 

same amount of mass is loaded into the saturated zone underneath each individual FDC, 

(iii) loading contaminant mass as a recharge concentration, or (iv) otherwise ensuring that 

source loading occurs at the water table in the far-field model, the results of the simulations 

could be significantly different. Scoping simulations performed by NRC staff using Tc-99 fluxes 

from Case K indicate that the peak sector concentrations at the 100 m boundary could be 

significantly higher for some sectors. Therefore, the manner in which contaminant fluxes are 

loaded in the far-field model should be further evaluated by DOE to ensure that the dose 

estimates are not significantly under-predicted in future PA analyses. 

With regard to grid discretization, DOE provided information during the FTF review to support 

the finite element size used in the local FTF model to ensure that the level of numerical 

dispersion is at an acceptable level (SRR-SWDA-2011-00054 Rev. 1). The FTF analysis is 

expected to be relevant to the SDF because the local FTF and SDF models have the same grid 

resolution. DOE also presented results in the FTF RAI responses that indicate additional grid 

refinement may be necessary to reduce numerical dispersion in cases of very low to no 

assumed physical dispersion. For example, if no physical dispersion is assumed, then the peak 

concentrations associated with a pulse release of a conservative tracer are shown to be a factor 

of approximately three to four times higher with a grid refined by a factor of two in each 

dimension (or a factor of 8 times more elements). Therefore, NRC staff reviewed DOE’s basis 

for the assumed level of physical dispersion in its local SDF PORFLOWTM model in the next 

 

21 Disposal unit designations used in the TER (Figure 2.8-8) correspond to designations used in the PA, which may 

not correspond to current designations.
132 

section to determine if numerical dispersion is at an acceptable level for the assumed level of 

physical dispersion. 

2.8.3.2 NRC Evaluation of Material Properties and Parameters 

As discussed above, DOE satisfactorily demonstrated that physical dispersion dominates 

numerical dispersion in the local FTF/PORFLOWTM saturated zone model and that numerical 

dispersion is at an acceptable level for the level of physical dispersion assumed 

(SRR-SWDA-2011-00054 Rev. 1, Figure RAI-FF-3.2). However, if the level of physical 

dispersion assumed in the modeling is not supported, a finer grid discretization may be needed 

to ensure numerical dispersion is at an acceptable level. The results presented in response to 

NRC RAI-FF-3 (SRR-SWDA-2011-00054 Rev. 1) also indicate that physical and numerical 

dispersion combined in the FTF/PORFLOWTM model accounts for significantly lower 

concentrations compared to modeling simulations where little to no physical dispersion is 

assumed and/or less numerical dispersion is simulated over the range of grid resolutions 

studied. Scoping simulations for SDF indicate that if no physical dispersion is assumed (only 

numerical dispersion is simulated), concentrations at the 100-meter boundary could increase by 

a factor of 2 for certain sectors for a relatively non-sorbing constituent. The impact of lower 

physical and numerical dispersion is unknown but is expected to be similar to the local 

FTF/PORFLOWTM model response (around a factor of 3 to 4). 

Hydrodynamic dispersion in the various local SDF/PORFLOWTM models discussed in the PA for 

various scenarios (or PA Cases) is represented by a longitudinal dispersivity of 10 m (33 ft) and 

transverse dispersivities of 1 m (3 ft) for horizontal transverse and 0 m (0 ft) for vertical 

transverse22, which are 10, 1, and 0 percent of a nominal 100 m (330 ft) plume travel distance, 

respectively (Table 2.8-3). Although DOE attempts to set transverse vertical dispersivity to 0 m 

in the PA Cases, DOE indicates that some numerical dispersion occurs, nonetheless 

(SRNL-STI-2009-00115, pg 130). It is significant to note that PORFLOWTM version 6.30.2 used 

for Case K, prepared after the PA was completed, does allow separate specification of 

longitudinal and transverse dispersivity in the vertical direction. For Case K, DOE specifies nonzero longitudinal and transverse vertical dispersivities of 1 m and 0.1 m for the longitudinal and 

transverse vertical dispersivities, respectively. Therefore, although DOE specifies its intent in 

the PA Cases to set transverse vertical dispersion to 0, DOE assigns a higher value of 0.1 m in 

Case K. 

In general, the literature supports significantly lower vertical dispersivities compared to 

horizontal dispersivities. Gelhar (1992) indicates that very low vertical dispersivities have been 

observed at several sites and are typically two orders of magnitude lower than longitudinal 

dispersivities. Although DOE attempted to limit vertical dispersion in Case A and other PA 

Cases, vertical dispersion may, nonetheless, be overstated in Case A and other PA cases. 

 

22 NRC staff could not confirm, prior to issuance of this TER that a PORFLOWTM command line “set w 0” noted in 

comment to “omit transverse vertical dispersion” for the SDF/PORFLOWTM PA Cases was executed in PORFLOWTM 

as intended. Based on review of the PORFLOWTM user’s manual (v. 6.12.3), the version used by NRC staff, it 

appears that while setting the velocity vector “w” will generally lower dispersion, some vertical dispersion will 

nonetheless occur. 
133 

Likewise, vertical dispersion may be overstated in Case K as a value of 1 m that is only 10 times 

less than the horizontal longitudinal dispersivity is used for vertical dispersivity in the longitudinal 

direction. 

Table 2.8-3: Dispersivities Used in DOE’s SDF/PORFLOWTM models 

Case Longitudinal1 Transverse1 Longitudinal 

(vertical)2

Transverse 

(vertical)2

Case A and all 

other PA Cases 10 m 1 m not applicable not applicable 

Case K 10 m 1 m 1 m 0.1 m 

1

 Case A (and other PA case) simulations were performed using PORFLOWTM (v 6.10.3) that only allows 

specification of a single longitudinal and transverse dispersivity that applies to both the horizontal and 

vertical directions. Case K simulations were performed using PORFLOWTM (v 6.30.2) that allows 

separate specification of longitudinal and transverse dispersivity in the vertical direction. Therefore, the 

longitudinal and transverse dispersivities specified in columns 2, 3 above apply to the horizontal direction 

for Case K. 

2

 DOE indicates in the PA that it sets vertical transverse dispersivity to zero for Case A and all other PA 

cases. NRC staff review of the PORFLOWTM input files reveals that an additional command line in 

PORFLOWTM was added in an attempt to “omit transverse vertical dispersion”, because PORFLOWTM (v 

6.10.3) does not allow direct specification of longitudinal and transverse vertical dispersivity as indicated 

in the note above. 

With regard to horizontal longitudinal dispersivity, DOE indicated during the FTF review that the 

specified dispersivity values selected from scale-dependent correlation data (Gelhar 1992) are 

supported by an optimization study for F-Area and H-Area seepage basins 

(WSRC-TR-2002-00291). Because the optimization study was based on (1) calibration to 

contaminant breakthrough data (less reliable than calibration to plume distributions), and 

(2) preferential flow pathways away from the seepage basins in the UTR-UZ are thought to exist 

that may not be reflective of flow and transport away from SDF sources, F-Area and H-Area 

seepage basin modeling results may not fully corroborate or be entirely relevant to field-scale 

dispersion at the SDF facility. Lower dispersivities [e.g., αL = 1.50 m (5 ft), αTh = 0.10 m 

(0.33 ft), αTv = 0.01 m (0.03 ft)] from the tritium optimization study at E-Area Old Burial Ground 

(WSRC-TR-96-0037), although considered less certain by DOE, may be more representative of 

dispersion along flow paths away from the SDF. In general, Gelhar (1992) supports the 

conclusion that the most reliable field data are typically associated with relatively low field-scale 

dispersivities; lower dispersivities would tend to lessen plume spread and result in higher doses 

at the point of compliance. A number of modeling studies with similar grid discretization as the 

SDF/PORFLOWTM model for the larger SRS site also support lower dispersivities based on 

calibration to three-dimensional plume data, although numerical dispersion may have been 

dominant in these exercises. In general, dispersivities derived from calibration to threedimensional plume distribution data are expected to be more reliable than calibration to 

breakthrough curve data (Gelhar, 1992). Additionally, because NRC staff is unable to conclude 

that physical dispersion is adequately supported, NRC staff is unable to conclude that horizontal 

and vertical discretization is adequate to ensure that numerical dispersion is at an acceptable 

level in the SDF/PORFLOWTM model. 
134 

In summary, DOE has not provided sufficient information to support the level of physical 

dispersion assumed in SDF/PORFLOWTM modeling. It is significant to note that changes to the 

SDF/PORFLOWTM model that may be made to address other issues identified by NRC staff as 

discussed in the model calibration section may result in changes in hydraulic gradients and 

lessen the impact of potentially excessive levels of vertical dispersion in the SDF/PORFLOWTM 

model. Evaluation of any relevant contaminant transport data in the vicinity of SDF or tracer 

experiments may provide additional support for the levels of dispersion simulated in 

SDF/PORFLOWTM models. 

NRC staff’s evaluation of Kds is discussed in the near-field section (Section 2.7.4), as the same 

Kds are used for both vadose zone and saturated zone materials. 

2.8.3.3 NRC Evaluation of Model Calibration 

With respect to GSA/PORFLOWTM model calibration, the maximum hydraulic head residual in 

the UTR-LZ of 8.2 m (27 ft) (WSRC-TR-2004-00106) greatly exceeds the GSA/FACT model 

calibration goal of no more than 5 to 10 percent of the total hydraulic head variation in the UTR 

aquifer for the maximum head residual. The UTR-LZ exhibits large residuals in the vicinity of 

H-Area (WSRC-TR-2004-00106), but the well locations having the largest residuals are not in 

the vicinity of SDF. 

In fact, very few calibration targets are located at the margins of the GSA/PORFLOWTM model 

in the vicinity of the SDF. It is not clear to what extent data collected from recent geological 

investigations or well construction at or near the SDF agree with the GSA/PORFLOWTM model 

and specific details regarding the adequacy of GSA/PORFLOWTM model calibration in the 

vicinity of the SDF are not provided in the 2009 PA Table 4.2-10, pg 181). PA data comparing 

calibration statistics over time are not particularly helpful as these statistics are biased by the 

large number of wells on the greater GSA that would dominate calibration statistics for the area 

of interest. 

Comparisons of observed to modeled heads at SDF are provided in Figure 4.2-14 of the 

2009 PA and generally reveal higher, modeled heads at the SDF and a greater horizontal 

gradient from the SDF to surface water. It is possible that lower hydraulic conductivity 

assignments in the SDF subsurface or higher hydraulic conductivity assignments nearer surface 

water might improve model calibration if the local SDF area was the focus of model calibration. 

Flach (WSRC-TR-2004-00106) suggests that more extensive calibration efforts would likely 

improve the hydraulic head residuals of the GSA/PORFLOWTM model. Nonetheless, more 

extensive calibration in the area of interest is desirable both for transparency and model 

performance. Additionally, the SDF/PORFLOWTM model does not consider capped conditions 

consistent with closure plans for the site. Given the apparent importance of the location of the 

groundwater divide and hydraulic gradient to contaminant fate and transport at SDF, it would be 

prudent for DOE to evaluate the impact of reduced recharge underneath the SDF due to the 

presence of a cover, if it has not already done so. 
135 

The impact of improved model calibration in the area of interest on PA dose predictions is not 

clear. However, lower vertical gradients at the SDF could lead to higher peak sector doses due 

to (i) less dispersion vertically, (ii) less loss of activity to the Gordon Confining Unit and Gordon 

aquifer, or (iii) greater plume overlap of SDF sources. Lower hydraulic conductivities in the area 

of interest could potentially lead to lower dilution rates of SDF plumes. The location of the SDF, 

on a groundwater divide, also significantly affects modeled plume overlap. While 64 FDC 

sources could theoretically contribute to groundwater dose at the point of maximum exposure, it 

appears that sources nearest the 100 m boundary contribute more significantly to peak sector 

doses at the SDF and that most SDF sources do not, in fact, overlap. Rather, SDF sources 

appear to spread out almost in a radial fashion, thereby minimizing plume overlap laterally 

compared to a situation where flow is primarily in the same direction. In conclusion, slight 

changes to the location of the groundwater divide or hydraulic gradients could lead to greater 

cumulative impacts and/or less aquifer dilution than assumed in the 2009 PA. Improved 

calibration in the area of interest would serve to provide additional support for the assumed level 

of dilution/dispersion of SDF sources. 

2.8.3.4 NRC Evaluation of Data and Model Uncertainty 

NRC staff attempted to evaluate the contributions of various sources to the peak dose in Case 

K. Although NRC staff used an earlier version of PORFLOWTM (v 6.12.3) that is not able to 

specify vertical dispersivities separately and the sources simulated may not contribute to the 

exact point in the UTR aquifer that the peak Case K dose in DOE’s SDF/PORFLOWTM model 

occurs from all sources, the results should provide a relative indication of individual source (or 

source group) contributions to dose at the 100 m boundary. Table 2.8-4 presents the results of 

this evaluation. Figure 2.8-8 illustrates the source locations simulated at the SDF. 

As stated above, many factors contribute to what appears to be an optimal configuration of 

sources at the SDF (i.e., minimal overlap of and cumulative impacts from SDF sources). A shift 

in the location of the groundwater divide could lead to variations in cumulative impacts from 

multiple SDF sources. A lower hydraulic conductivity (to better match the horizontal gradient) 

may lead to less dilution of SDF source plumes, or a lower vertical hydraulic gradient could lead 

to less vertical dispersion, greater plume overlap in the vertical direction, or less loss of mass to 

the GCU. Therefore, an adequate set of calibration targets and acceptable calibration statistics 

in the area of interest are needed to provide additional confidence that the level of dilution 

realized in DOE’s SDF PA model is not over-stated. 

NRC staff is also concerned with the potential impact of calcareous zones on contaminant flow 

and transport at the SDF. Many SDF sources traverse the lower zone of the UTR aquifer where 

calcareous materials are known to be more pervasive in the subsurface at SRS. The 

GSA/PORFLOWTM and SDF/PORFLOWTM models do not explicitly account for the impact of 

these zones on contaminant flow and transport (other than considering the impact of mud 

fraction on hydraulic conductivity). During the SDF review, NRC staff expressed concerns 

regarding the potential impact of these zones on contaminant flow and transport. Thayer, et al. 

(1995) indicate that contaminant data for the Burial Ground Complex (located in E-Area at the 

GSA) and the Chemical, Metals, and Pesticide Pits (located off the GSA) support the hypothesis 
136 

that contaminants may be preferentially transported within these zones. These contaminant 

data were not provided nor assessed during this evaluation. Furthermore, sorption of key 

radionuclides may be overstated in the SDF PA for calcareous materials as no site-specific Kds 

have been developed specifically for this zone and material in the UTR aquifer. However, 

because Tc and I are expected to be the key risk drivers in Case K and little to no credit is taken 

for sorption for these radionuclides, the need for additional site-specific information in this area 

may be limited. 

Table 2.8-4: Ratio of All Sources to Individual Source Peak Doses 

Case Source1 Normalized 

Concentration2

Location of 

Maximum 

Concentration 

Preparer 

Case K All 1 Sector I DOE 

Case K All 0.83 (1) Sector I NRC3

Case K 7A, B, C, D 0.58 (0.70) Sector I NRC3

Case K 7A 0.2 (0.24) Sector I NRC3

Case K 2AB 0.2 (0.24) (0.34) (4) Sector L5 NRC3

1

 Figure 2.8-8for location of sources from the 2009 PA. DOE has recently renumbered the FDCs. 

2

 Normalized to the maximum concentration in Sector I from DOE Case K. Values in parentheses are 

normalized to the maximum concentration in Sector I from a similar NRC staff PORFLOWTM simulation of 

Case K using an older version of PORFLOW. The results of the NRC simulation are 83 percent of the 

peak concentrations from DOE’s Case K simulation. 

3

 NRC simulations are performed with PORFLOWTM version 6.12.3. PORFLOWTM (v 6.12.3) does not 

allow separate specification of vertical dispersivities; therefore, dispersion is expected to be 

overestimated in the NRC simulations (concentrations are lower). 

4

 The centerline of the plumes emanating from Vaults 2A and 2B traverses compliance evaluation 

Sector L. The peak concentration from Vaults 2A and 2B is 34 percent of the peak concentration in 

Sector L from all sources. 

5 Sector L peak concentration is 57 percent of the peak concentration in Sector I. 

No formal mapping to identify calcareous zone seeps along stream valleys has been conducted. 

Nonetheless, DOE indicated in an FTF RAI resolution meeting (NRC, 2011d) that a field 

mapping activity such as this could be incorporated in the future. NRC staff support such an 

activity to evaluate the impact on the hydrogeologic system of the SDF facility of potentially 

highly porous and conductive soft zones in the UTR-LZ. DOE progress in this area will be 

evaluated during monitoring. If calcareous zone seeps are identified, tracer studies in the SDF 

UTR-LZ using innocuous tracers that are commonly used to understand preferential flow and 

transport could be conducted to better understand the effect of these zones on contaminant flow 

and transport. Results from any such tracer studies would be evaluated by NRC during the 

monitoring period. 
137 

Figure 2.8-8: Locations of Sources at the Saltstone Disposal Facility 

(2009 PA, Figure 4.2-20) 

2.8.3.5 Far-field Conclusions 

NRC staff concludes that DOE’s far-field model presents an acceptable framework to facilitate 

decision-making regarding SDF operations. However, as described in Section 2.8.3, the NRC 

staff also concludes that modeled far-field performance results in potentially overly optimistic 

levels of dilution and dispersion of SDF source plumes and limited cumulative impacts from the 

facility. Changes to key parameters (e.g., hydraulic conductivities, recharge rates reflective of 

capped conditions) and improved model calibration in the area of interest could lead to 

(i) variability in the location of the groundwater divide, (ii) changes to the hydraulic gradient, or 

(iii) changes to Darcy velocities that could result in significantly lower levels of modeled dilution 

and dispersion, which could result in larger predicted doses. Likewise, lower dispersivities or 

finer grid discretization could lead to lower levels of modeled dispersion and higher predicted 

doses. Other areas of uncertainty in the far-field model include: (i) the impact of DOE’s source 

loading approach on predicted doses and (ii) the impact of calcareous zones on contaminant 

flow and transport. The NRC staff concludes that in future PA updates, DOE could improve the 

current far-field model to reduce uncertainty in dose modeling predictions. For example: 
138 

• The calibration process could be improved and made more transparent, particularly in 

the area of interest, to provide confidence that the level of dilution in DOE’s model is not 

overstated. 

• Loading of the contaminant-flux source terms from the vaults could be evaluated to 

ensure that the dose estimates are not significantly under-predicted. 

• The appropriateness of selected dispersivities and the need for additional vertical or 

horizontal mesh refinement could be evaluated to ensure that contaminant plumes are 

not artificially dispersed in the far-field model. 

• Additional information could be collected during the monitoring period to support DOE’s 

modeling treatment of the calcareous zones in the lower portion of the UTR aquifer. 

DOE could consider additional data collection related to calcareous zone outcrop 

locations and tracer tests to provide further support for the adequacy of its modeling 

treatment of the lower zone of the UTR aquifer. Site-specific Kds may be developed for 

the lower zone of the UTR, if deemed risk-significant. 

As a result of the uncertainty in the far-field model, NRC staff will monitor these items when the 

PA is revised as part of DOE’s PA maintenance program. DOE can address this monitoring 

factor by making appropriate revisions during future PA updates. 

2.9 Air Transport of Radionuclides 

2.9.1 Modeling of Transport of Radionuclides in the Air Pathway 

The airborne and radon analysis evaluated the potential dose due to the diffusion of 

radionuclides from the waste form to the surface through the air-filled pore space in the 

overlaying material (i.e., the disposal unit roof and the engineered cover). The air pathway 

analysis was performed for the three different disposal unit designs (i.e., Vault 1, Vault 4, and 

the FDCs) and for waste forms created with DDA and ARP/MCU waste (for Vaults 1 and 4) and 

SWPF waste (for the FDCs). The radionuclides considered in the airborne pathways included 

only radionuclides that were assumed to have a measureable inventory in the disposal units at 

the time of closure and that have the potential to form a vapor phase in the waste form. These 

radionuclides included C-14, Cl-36, H-3, I-129, Sb-125, Se-79, and Sn-126. Tc-99 was also 

added to the list of radionuclides evaluated in the air transport assessment subsequent to the 

screening analysis. The flux and associated dose from Rn-222 was also modeled. 

The diffusion of the radionuclides from the waste form to the surface and the resulting flux at the 

land surface was modeled using PORFLOW. In this calculation, boundary conditions were 

assumed that forced the modeled diffusion to go upwards to the land surface. This assumption 

increases the estimated flux at the land surface because in reality diffusion will also occur in the 

downwards and sideways directions. The leaching of radionuclides out of the waste form to the 

groundwater was not included in the air pathways calculation. Advective transport due to 

fluctuations in atmospheric pressure that could cause air movement into and out of the soil (i.e., 

barometric pumping) was also not included in their analysis. 
139 

The apparent Henry’s law constants used to model the partitioning of the radionuclides between 

the liquid and gas phases were estimated using the Geochemist’s Workbench software. These 

apparent Henry’s law constants were then converted to pseudo-partitioning coefficients for use 

in the PORFLOWTM code. In the PA, DOE noted that the ionic strengths assumed in the 

estimation of the Henry’s constants were much lower than the expected ionic strength expected 

in the saltstone pore fluid because a more sophisticated model would be required to assess 

high ionic strengths and data to support these models is limited. For all of the radionuclides but 

tritium, higher ionic strengths result in higher activity coefficients and consequently higher 

partitioning of the radionuclide into the gas phase (i.e., “salting out” effect). In the response to 

comment AP-2 (SRR-CWDA-2010-00033), DOE discussed the potential effect of the increased 

ionic strength on the calculated fluxes. Based on experimental data reported in the literature, 

DOE concluded that the activity coefficients would be unlikely to increase by more than a factor 

of ten due to the higher ionic strengths. 

The PORFLOWTM calculations assumed the minimum closure cap thickness. Additionally, the 

portions of the closure layer located above the erosion barrier (i.e., the top soil layer and upper 

backfill layer) were not included in the analysis because these portions of the closure cap are 

subject to erosion over the time of the simulation. The potential reduction in gaseous flux 

caused by the presence of the geotextile fabrics, HDPE layer, and the GCL layer in the cap was 

also not included in this analysis. However, the components of the closure cap located below 

the erosion barrier were assumed to remain intact for the duration of the simulation. 

Once the fluxes of gaseous radionuclides through the cover were determined, these fluxes were 

used to estimate doses. The dose to an intruder who lives directly above the disposal site was 

calculated based on the highest peak radon flux into a basement sized control volume (500 m3

) 

that has an air exchange rate of 1 hr-1. A DCF of 1.4x10-5 mSv/Bq (5.0x10-5 mrem/pCi) was 

assumed for Rn-222 (SRR-CWDA-2010-00033). The dose from the other radionuclides to the 

intruder was based on the assumption that the conservative assumption that the intruder’s 

annual intake is equal to the emission rate from the disposal units (SRR-CWDA-2010-00033). 

With the exception of Rn-222, the doses to the member of the public were determined from the 

fluxes using dose-release factors (DRFs) calculated from effective dose equivalents generated 

using the CAP88 computer code. Receptors at 100 m from the SDF and at the site boundary 

were considered. The CAP88 computer code used site-specific meteorological information to 

determine the air concentration of the radionuclides, and then calculated doses from the 

ingestion, inhalation, air immersion, and ground shine pathways (SRNL-STI-2008-00415). The 

dose from Rn-222 to a member of the public located at 100 m from the disposal units was not 

calculated, but this dose would be lower than the dose to the intruder and is therefore bound by 

the intruder calculation. 

Using the analysis described above, the dose to the intruder from Rn-222 was estimated to be 

8.2x10-13 mSv/yr (8.2x10-11 mrem/yr) above the disposal cell with the highest peak radon flux. 

The dose to the intruder from the other gaseous radionuclides was calculated to be 

2.3x10-8 mSv/yr (2.3x10-6 mrem/yr) above Vault 1, 5.2x10-7 mSv/yr (5.2x10-5 mrem/yr) above 

Vault 4, and 3.0x10-8 mSv/yr (3.0x10-6 mrem/yr) above a FDC. The total dose to a member of 
140 

the public at 100 m from the disposal units from gaseous radionuclides other than Rn-222 was 

determined to be 8.14x10-16 mSv/yr (8.14x10-14 mrem/yr) (SRR-CWDA-2010-00033). 

A sensitivity analysis was performed by DOE for the air pathways analysis to assess the effect 

of uncertainty in the saturation conditions of the cap and the emanation rate for radon 

(SRNL-L6200-2010-00019). Based on this sensitivity analysis, it was found that the dose to the 

intruder from Rn-222 was 5x10-11 mSv/yr (5x10-9 mrem/yr) assuming the minimum moisture 

content for the cap and overlying materials. The member of the public dose assuming minimum 

moisture saturation was 1x10-15 mSv/yr (1x10-13 mrem/yr) from radionuclides other than Rn-222. 

Additionally, the calculated radon flux is directly proportional to the assumed emanation rate, so 

increasing the emanation from 0.25 to 0.7, the upper end of reported literature values, would 

only increase the flux by a factor of 2.8 (SRR-CWDA-2010-00033). 

2.9.2 NRC Evaluation – Air Transport 

DOE’s air transport analyses estimated that the potential doses from gaseous radionuclides will 

be much lower than the performance objectives. Additionally, DOE also evaluated uncertainty 

in the moisture content of the soil and the emanation of radon and found that the dose was 

small over the range of these values evaluated. DOE’s analysis included several conservative 

assumptions, including the use of the minimum cover thickness in the analysis, and the 

exclusion of the performance of the geotextile, HDPE, and GCL layers, and the exclusion of the 

layers above the erosion barrier. An additional conservatism in this calculation was that the 

leaching of radionuclides out of the waste form to the groundwater was not included in the air 

pathways assessment. This is conservative because leaching is expected to remove some 

radionuclides from the waste form over time, which would result in a smaller source term. The 

use of boundary conditions that forced the modeled diffusive flux upwards was another 

conservatism in DOE’s modeling. An additional potential conservatism in the assessment of 

Rn-222 dose is that the inventory of its ancestors, Ra-226 and Th-230, used in this analysis 

could be higher than the actual inventory due to the use of conservative assumptions in 

estimating this inventory (Section 2.2). 

An area of uncertainty in DOE’s analysis is the exclusion of advective transport due to 

fluctuations in atmospheric pressure. This type of transport can be a significant source of 

transport of gaseous material such as radon into a house located on site (Nazaroff, 1992). 

However, this type of transport can be difficult to predict and model. Additionally, there is some 

uncertainty in the effect of the assumed ionic strength on the model results. As described in the 

previous section, DOE provided some literature information about the effect of ionic strength on 

activity coefficients, but measurements have not been performed for the actual saltstone. The 

assumed range of moisture conditions is also potentially a source of uncertainty in the air 

pathway assessment. As noted above, DOE evaluated the effect of setting the moisture content 

to the lowest end of their predicted range and found that the dose was still small. However, if 

the actual saturations are lower than the assumed range, the doses could be higher than 

predicted. Even though there are some uncertainties in DOE’s air pathway analysis, the NRC 

staff concludes that the calculated doses from the air pathway were so small that the doses 
141 

from this pathway are still likely to be less than the performance objectives, even when 

accounting for uncertainty. 

2.10 Biosphere Characteristics and Dose Assessment 

DOE’s biosphere model converts calculated groundwater concentrations at the receptor location 

to public doses considering various exposure pathways (Section 2.3). This modeling involves 

numerous input parameters that describe the characteristics of the biosphere and receptor. The 

DOE biosphere model calculates the transport of radionuclides within the biosphere pathways 

that lead to human exposures. Dosimetry calculations then convert these human exposures to 

annual doses to a member of the public. Dose modeling of air releases from the Saltstone 

facility was conducted as a separate modeling activity (Section 2.9). 

2.10.1 DOE Approach to Modeling Exposure Pathways and Public Dose 

The DOE PA calculates the biosphere pathway transport and receptor dose using a 

deterministic GoldSim® biosphere model that calculates the dose results for comparison with 

performance measures. This deterministic biosphere model is separate from the probabilistic 

GoldSim® model DOE developed to evaluate the sensitivity of dose results to various input 

parameters. The DOE biosphere model converted groundwater radionuclide concentrations 

(calculated by the PORFLOWTM transport model) at various locations to all-pathway doses 

considering the biosphere exposure pathways described in Section 2.3. DOE cited Regulatory 

Guide 1.109 as the basis for the pathway models (2009 PA). Pathways described in Regulatory 

Guide 1.109 as applicable to liquid release dose calculations include potable drinking water, 

aquatic foods, shoreline deposits, and irrigated food. Irrigated food pathways described in the 

guide include produce (as a combination of non-leafy vegetables, fruit, and grains), milk, meat 

and poultry, and leafy vegetables. Plant concentrations are calculated in Regulatory Guide 

1.109 based on irrigation deposition on plant surfaces and uptake of radionuclides to plants 

from soil. The Regulatory Guide 1.109 dose calculations involve ingestion, direct radiation, and 

inhalation routes of exposure. 

For Case K, DOE ran the deterministic GoldSim® model with updated biotic transfer factors, 

additional food sources (chicken and egg), and 25-year buildup and leaching of radionuclides in 

irrigated soil. The DOE Case K model calculated the receptor dose from ingestion of meat, 

milk, and poultry-based food products by assuming the animals drink water and consume fodder 

irrigated with water from a well at the point of maximum exposure 100 m downgradient of the 

SDF. The effects of buildup and leaching were calculated using a model that modifies the 

calculated annual soil deposition of radionuclides by the years of irrigation while accounting for 

losses due to radioactive decay and leaching. Leaching was addressed by applying a root zone 

soil leaching model (Baes and Sharp, 1983) that evaluates sorption of contaminants in soil 

based on the water infiltration rate and soil properties including depth, water content, density, 

and geochemistry represented by distribution coefficients. 
142 

2.10.2 DOE Biosphere Model Input Parameter Values 

DOE’s documentation of biosphere input parameters describes the technical bases for 

bioaccumulation factors, human health and exposure parameters, and dose analysis 

parameters. Additional details on these groups of input parameters are provided in the following 

paragraphs. 

The bioaccumulation factors (also commonly known as transfer factors) used by DOE in the 

initial PA analysis (i.e., in all cases other than Cases K, K1, and K2) have been documented by 

WSRC–STI–2007-00004. The PA uses factors that represent the following transport pathways: 

soil-to-vegetable, feed-to-milk, feed-to-beef, and water-to-fish. DOE selected input parameters 

based on a review of available technical information with a preference for site-specific and more 

recent compilations. The DOE selection methodology placed priority on site-specific (Friday et 

al., 1996; Jannik, 2003) or region-specific values (if available) followed by more general 

literature compilations (Staven et al., 2003; Baes et al., 1984; and the National Council on 

Radiological Protection [NCRP], 1996). DOE selected site-specific and general literature values 

without modification, however, the geometric mean of the latest general literature value and the 

previous SRS selected value was used when the latest value differed from the prior SRS 

selected value by more than two orders of magnitude (2009 PA). 

For Case K, DOE derived updated plant bioaccumulation factors by calculating a weighted 

average of crop-specific bioaccumulation factors from the most recent International Atomic 

Energy Agency (IAEA) compilation (IAEA, 2010). DOE weighting factors were based on crop 

production data (i.e., leafy vegetables 20%, legumes 15%, tubers and roots 10%, and non-leafy 

vegetables 55%) for South Carolina. Element-specific animal product bioaccumulation factors 

for feed-to-milk, feed-to-beef, and water-to-fish were updated based on the most recent IAEA 

compilation (IAEA, 2010). Previous plant and animal product bioaccumulation factors were 

used for elements that were not included in the referenced IAEA report. Chicken and egg 

bioaccumulation factors were selected from the most recent source (IAEA, 2010) or Staven et 

al. (2003) for elements not reported by IAEA, or set to zero if no values were reported in either 

reference. 

The Case K updated soil concentration model introduced additional input parameters to the 

biosphere model to calculate the soil leaching rate constant. These additional input parameters 

included the irrigation rate (131.3 cm/yr [51.7 in/yr]) (2009 PA), precipitation and 

evapotranspiration rates (124.8 cm/yr [49.14 in/yr]; 82.7 cm/yr [32.57 in/yr]) (Jones and 

Phifer, 2008), and soil moisture content (0.039) (2009 PA), and radionuclide-specific soil 

distribution coefficients (2009 PA; Table 4.2-15). The soil distribution coefficients used in the 

calculation of the soil rate leaching constant were assumed to be the same as the Kd values in 

the vadose zone. 

DOE PA human health and exposure parameters also were documented by 

WSRC-STI-2007-00004. This group of inputs includes food and water consumption rates, the 

air inhalation rate, exposure times, and local food production and gardening related inputs. 

DOE selected values from a hierarchy of data sources with preference given to site-specific (if 
143 

available) and more recent sources of information (2009 PA). Site-specific studies included 

Hamby (1991; 1992), and regional information was obtained from a recent dose reconstruction 

study of the SRS site (CDC, 2006). When site-specific information was not available, DOE 

selected input parameter values from national or international organizations EPA (1997), 

ANL-EAIS-8, ANL-EAD-4, and NUREG/CR-5512 and assumed the information was applicable 

to SRS area practices. For Case K, the DOE selected additional chicken and egg pathway 

input parameters from a prior NRC-sponsored analysis (Simpkins et al., 2008). These inputs 

included livestock feed and water consumption rates, human consumption rates of chicken and 

egg food products, and fractions of chicken and egg food that are locally obtained. 

DOE dose analysis parameters include dose coefficients that convert human intakes of 

radionuclides and human exposure to radioactivity outside the body to dose. For the initial PA 

and for Cases K, K1, and K2, DOE selected adult internal dose coefficients from the 

International Commission on Radiological Protection, Publication 72 (ICRP, 1996). DOE 

selected inhalation dose coefficients based on the ICRP recommended lung absorption type, 

where available, and the most conservative lung absorption type for radionuclides where no 

ICRP recommendation is provided in the source document (2009 PA). DOE selected external 

dose coefficients for uniformly distributed ground surface contamination (at an infinite depth with 

no shielding and at 15 cm) and for water immersion from Federal Guidance Report No. 12 

(EPA, 1993). For Case K, DOE made corrections to the values for Pb-210 and U-232 to 

account for short half-life decay products and updated the value for Pt-193. 

2.10.3 NRC Evaluation – Biosphere Pathway and Dose Calculations 

To review the DOE PA biosphere pathway and dose calculations, the NRC staff evaluated the 

general methodology of the DOE biosphere modeling approach and the DOE selection of key 

input parameters with particular attention focused on the most recent DOE responses to NRC 

comments on the PA. 

2.10.3.1 NRC Evaluation of DOE Biosphere Modeling Approach 

The dose methodology of the DOE PA involves executing biosphere pathway and dose 

calculations with a deterministic GoldSim® model that DOE developed based on the pathway 

models in Regulatory Guide 1.109 (2009 PA). The NRC staff considers the pathway 

calculations in Regulatory Guide 1.109 to be a reasonable basis for developing the pathway and 

dose calculations. The NRC staff verified that the DOE biosphere modeling approach includes 

the Regulatory Guide 1.109 pathways, with a few exceptions. The DOE biosphere model does 

not explicitly evaluate separate pathways for leafy vegetables, fruit, or grains. DOE combines a 

subset of food products from the produce category into a single food category (i.e., vegetables). 

The NRC staff finds that the use of a single category for vegetables to be technically acceptable 

provided related input parameters are internally consistent and the analysis is complete 

(Section 2.10.3.2). Case K includes consideration of leafy and other vegetables in the 

derivation of plant transfer factors but the analysis does not provide sufficient information to 

determine whether fruit and grain transfer factors were included in the data that DOE averaged. 

The Case K analysis also includes the poultry and egg ingestion pathways, which were not 
144 

included in the initial PA analyses. Based on the preceding analysis, the staff concludes the 

biosphere pathways considered by DOE are complete. 

The dosimetry methodology applied by DOE differs from what is described in Regulatory Guide 

1.109. This is expected by the NRC staff because methods and data have been revised since 

the regulatory guide was published. The methodology used by DOE is based on adult internal 

dose coefficients from the ICRP (1996) and external (i.e., direct radiation exposure) dose 

coefficients from Federal Guidance Report No. 12 (EPA, 1993). These dose coefficients are 

acceptable to NRC staff because they incorporate the most current and appropriate scientific 

models and methodologies for an adult receptor and are consistent with Commission direction 

in SRM-SECY-01-0148 (NRC, 2002a). 

The NRC staff finds that the approach used in Case K for evaluating soil build-up from a land 

irrigation scenario in the DOE biosphere model is appropriate. While current radionuclides that 

contribute most to the calculated all-pathway PA doses involve elements that are generally 

more mobile in the environment and are less affected by soil buildup, potential future changes to 

the PA could change the radionuclides that contribute most to calculated all-pathway doses. As 

a result, NRC staff finds the DOE incorporation of the soil buildup modeling capability into Case 

K to be a useful improvement. 

2.10.3.2 NRC Evaluation of DOE Biosphere Input Parameters 

The NRC staff reviewed the parameters used in the biosphere calculations in the original PA 

and in Case K and finds the basis for the parameter selection as well as the parameters 

selected to be generally acceptable, with the exception of the items described in the following 

paragraphs. Additionally, NRC staff finds that the DOE emphasis on modeling site-specific 

pathways and selecting site-specific or site-applicable input parameters, where available, is 

generally consistent with past NRC practices that have been described, for example, for 

demonstrating compliance with NRC decommissioning regulations (NRC, 2003). 

The NRC staff asked DOE for more information about its basis for excluding bioaccumulation 

factors from the uncertainty analysis and the DOE approach to deriving bioaccumulation factors 

from available data sources (NRC, 2010b, i; B-1). DOE’s response included a comparison of 

dose calculation results using (1) the PA soil-to-vegetable bioaccumulation factors, and (2) the 

maximum values presented by DOE in Tables 4.6-1 to 4.6-4 of the 2009 PA soil-to-vegetable 

bioaccumulation factors (SRR-CWDA-2010-00033). While NRC staff agreed the absolute 

changes to dose were small, they noted in the response (NRC, 2010i) that the relative changes 

were moderate to significant. Additionally, NRC staff notes that the true maximum values for 

these factors are larger than the values used in this analysis. For example, the maximum soilto-vegetable transfer factor assumed for Tc was 5.46 in Table 4.6-1 of the 2009 PA, which is 

lower than the value assumed for Case K. Due to the wide variability in some of the available 

transfer factor data and the direct influence on pathway-specific dose results, NRC staff 

believes that it is important to use technically defensible methods to analyze and propagate 

transfer factor variability into the PA calculations, in particular, for those elements that exhibit 

high variability are or have the potential to be important contributors to the total dose results. 
145 

In the most recent RAI response (SRR-CWDA-2011-00044), DOE discussed the basis for their 

use of the geometric mean to average the prior DOE average value with a new DOE estimate in 

their calculation of the bioaccumulation factor inputs used in the original PA analysis, citing 

other past uses of a geometric mean by various practitioners. However, NRC staff’s principal 

concern with the DOE’s approach was that no basis was provided for averaging the prior DOE 

average value with a new DOE estimate. DOE’s revised Case K used biosphere parameters 

taken directly from a recent IAEA compilation that did not include any data manipulations using 

a geometric mean. The NRC staff finds the approach used for the Case K bioaccumulation 

factors to be an improvement over the past averaging approach. 

In the second RAI, NRC staff asked for more information about DOE’s basis for excluding leafy 

vegetables from the derivation of the soil-to-vegetable bioaccumulation factors (NRC, 2010i; 

B-4). The NRC staff commented that the DOE derivation of the vegetable bioaccumulation 

factor based solely on bioaccumulation factors for root vegetables (assuming limited local leafy 

vegetable productivity information) would produce a lower bioaccumulation factor than one that 

included leafy vegetable bioaccumulation factors. In the RAI response 

(SRR-CWDA-2011-00044), DOE derived updated bioaccumulation factors for Case K from 

available data by taking a weighted average of crop-specific bioaccumulation factors from the 

most recent IAEA compilation (IAEA, 2010) using crop production data (i.e., leafy vegetables 

20%, legumes 15%, tubers and roots 10%, non-leafy 55%) for the SRS region. The DOE 

approach used previous bioaccumulation factors for those elements that were not included in 

the latest IAEA report. The NRC staff finds the weighted-average approach for updating the 

vegetable bioaccumulation factors provided in the response to be better than the approach that 

treated all vegetables as root vegetables, which was used in the DOE PA. The NRC staff finds 

the approach used in Case K addresses the principal concern of comment B-4 (NRC, 2010i). 

The NRC staff finds that the DOE input parameters used in the soil buildup analysis for Case K 

were referenced to acceptable sources and had sufficient technical bases with a few 

exceptions. The irrigation rate used by DOE for calculating the leaching rate constant (which 

DOE refers to as the soil buildup rate) of 131.3 cm/yr (51.7 in/yr) may be overestimated 

because the value has not been adjusted by the DOE fraction of time vegetation is irrigated 

(0.2, 2009, PA; Table 4.6-6). Because a higher rate of irrigation also leads to a larger amount of 

leaching, a high value overestimates leaching and underestimates the buildup factor. However, 

a higher rate of irrigation also leads to larger amount of deposition of radionuclides on the soil. 

The overall soil concentration is determined by multiplying the buildup factor described above by 

the irrigation rate and the concentration in the water. It is not clear to the NRC staff if DOE used 

the high irrigation rate of 131.3 cm/yr (51.7 in/yr) in the calculation of the soil concentration, or if 

DOE used an irrigation rate that was adjusted for the fraction of time vegetation is irrigated 

(26.2 cm/yr [10.3 in/yr]). If DOE used the high irrigation rate in both places, then the net effect 

would be an overestimation of the soil concentration. However, if a high irrigation rate was used 

in estimating the leaching and a low rate was used in the calculation of the soil concentration, 

then the soil concentration and the dose would be underestimated. For example, NRC staff 

recalculation of the soil buildup factors from the DOE analysis using the adjusted irrigation rate 

of 26.2 cm/yr (10.3 in/yr) increased the buildup factor estimate (for 25 years with leaching) by 
146 

approximately 2.6 times for I-129, Ra-226, and Tc-99 relative to the current DOE comment 

response analysis. 

The NRC staff also finds that the distribution coefficients used in the buildup analysis were 

referenced to a report that selected values conservatively within the context of hydrologic 

transport modeling (WSRC-TR-2006-00004). Selecting distribution coefficients in this manner is 

non-conservative when applied to irrigation and soil sorption modeling because lower sorption 

values are typically regarded as conservative in the context of transport modeling (increases 

predicted mobility) but lower values of sorption coefficients could underestimate radionuclide 

build up for the same reason (i.e., by increasing predicted mobility). Additionally, the 

geochemical environment at the surface may differ significantly from the subsurface (e.g., 

increased organic content). The estimated buildup in soil for some elements could increase if 

further review caused distribution coefficients to be updated to higher values. For example, 

NRC staff evaluated the effect of changing the Kd for Ra from the value of 5 mL/g assumed in 

the buildup analysis to a value of 25 mL/g which was measured in sandy site soils 

(SRNL-STI-2010-00527). This change to the Kd value increased the calculated the buildup 

factor by a factor of 4.8. This increase in the buildup factor would result in a higher 

concentration in the soil and a corresponding higher concentration from the ingestion of plant 

and animal products pathways. The Kd values for Tc and I used in the buildup calculation 

(0.6 mL/g and 0 mL/g respectively) are also less than some of the measured site specific Kd

values (SRNS-STI-2008-00286, WSRC-TR-2006-00004). However, the range of Kd values 

measured for Tc and I was smaller than for Ra, so the potential effect on the calculated buildup 

factor is also smaller. Additionally, some of the higher Kd measurements for Tc and I were 

made in soil that had a high clay content, and these values might not be as applicable to the 

surface. 

In the NRC RAIs (NRC, 2010b, i) NRC staff commented that the DOE selected value for the 

drinking water ingestion rate of 337 L/yr (0.92 L/d) is inconsistent with an average member of 

the critical group definition. The water ingestion pathway is an important pathway and has 

consistently been the largest contributor to DOE PA public all-pathway calculated doses. In its 

latest response (SRR-CWDA-2011-00044) DOE provided additional description of the technical 

basis for the selected value, which is from an EPA analysis of community water survey results 

(EPA, 2004). The value chosen by DOE is the national mean per capita community water 

consumption rate for all ages. However, because the DOE receptors are defined as adults, the 

selected consumption rate value for all ages is low relative to the value reported for adults (e.g., 

401 L/yr relative to the selected DOE value of 337 L/yr). Historically, in establishing the drinking 

water standards, EPA has accepted a single value for drinking water consumption rate of 2 L/d 

as a reasonably conservative estimate. NRC has used values of approximately 1.3 L/yr and 

greater in various previous reviews (NRC, 1977). Considering the site-specific nature of the 

warmer climate at SRS, the drinking water consumption for the average member of the critical 

group could be above the EPA national average for adults of 401 L/yr. The water ingestion rate 

has a linear effect on the predicted dose from the drinking water pathway, so impacts of 

changing the parameter are therefore easy to evaluate. Because drinking water contributes 

approximately half of the dose to an off-site member of the public in most of the cases DOE 

evaluated, changing the water consumption rate to 401 L/yr would increase the total predicted 
147 

dose by approximately 10% and increasing the consumption rate to 2 L/d would increase the 

total dose by approximately 50%. For future stochastic analyses, another defensible alternative 

could be to define a drinking water consumption rate distribution that reflects the range of 

variability in the national survey data. Use of a distribution is expected to encompass the 

variability in a population of individuals with different activity levels and consumption patterns. 

NRC staff concludes that the Case K biosphere modeling included changes to the modeling 

approach and parameters that addressed most of the concerns NRC staff identified in the RAIs. 

However, NRC staff also concludes that some of the parameter values selected in Case K may 

have resulted in an underestimation of dose, such as the drinking water ingestion rate as well as 

the irrigation rate and Kd values assumed in the soil buildup calculations. As a result, NRC staff 

will monitor these items when the PA is revised as part of DOE’s PA maintenance program. 

DOE can address these PA maintenance items by making appropriate revisions and/or 

providing additional justification for these biosphere parameters when the PA for the SDF is next 

revised. 

2.11 Computer Models and Computer Codes 

2.11.1 Computer Models and Computer Codes Used by DOE 

DOE used a variety of codes to assess long-term risk from the SDF (2009 PA; Section 4.3.1). 

The 2009 PA models are integrated to provide long-term dose estimates for comparison against 

dose-based standards in 10 CFR Part 61, Subpart C (Figure 2.11-1). Primary codes used in the 

2009 PA are marked in bold, black text in Figure 2.11-1. This section provides summary 

description of the various codes used in the 2009 PA modeling including information regarding 

code capabilities, documentation, verification, and validation testing, as applicable. In addition, 

DOE’s quality assurance program, including software quality assurance, also is described. 

HELP

HELP (Hydrologic Evaluation of Landfill Performance) is a quasi-two dimensional model 

designed by the United States Army Core of Engineers under an interagency agreement with 

the United States Environmental Protection Agency with the primary objective of performing 

landfill water balance calculations. HELP version 3.07 (1997) is used to calculate infiltration 

through the proposed SDF cover. Additional information regarding the use of the HELP 

computer code to model infiltration through the SDF cover is provided in Section 2.4. Data 

inputs to HELP include weather, soil, and cover design information. Modeling outputs include 

runoff, evapotranspiration, lateral drainage, hydraulic head, water storage, and infiltration. 

HELP documentation includes a User’s Manual, engineering documentation providing source 

code information and HELP calculation methodology, and verification testing (see 

http://www.wes.army.mil/el/elmodels/helpinfo.html). Verification consists of testing the ability of 

the code to adequately predict lateral drainage and other water balance outputs against 

measured data from large-scale physical models and actual disposal cells at seven sites across 

the United States. 
148 

The HELP code has been extensively used by DOE for various SRS applications including area 

closure projects under the Comprehensive Environmental Response, Compensation and 

Liabilities Act. 
149 

Figure 2.11-1: Model Linkages in DOE’s Performance Assessment for the SDF 
150 

PORFLOW

PORFLOWTM is a commercially available multi-phase computational fluid dynamics tool 

developed by Analytic Computational Research, Inc. PORFLOWTM uses numerical methods to 

solve steady state or transient fluid flow, heat, salinity and mass transport problems in variably 

saturated, porous, or fractured media. PORFLOWTM version 5.95.0 was used to construct the 

GSA/ PORFLOWTM model described in greater detail in Section 2.8. PORFLOWTM version 

6.10.3 was used in the 2009 PA to perform unsaturated and saturated flow and transport 

simulations described in greater detail in Section 2.7.3 and 2.8 (Figure 2.11-1). PORFLOWTM 

version 6.30.2 was used to perform Case K simulations as described in RAI response 

(SRR-CWDA-2011-00044). PORFLOWTM is also used to perform vapor phase diffusion of 

contaminants from SDF sources to the ground surface for input in the air transport code CAP-88 

(Figure 2.11-1). Flow outputs from PORFLOWTM are also used in DOE’s probabilistic 

assessment executed within the GoldSim® modeling platform discussed in more detail below. 

A user’s manual and numerous validation reports are available for the PORFLOWTM computer 

code (www.acricfd.com/software/porflow/). Acceptance testing on PORFLOWTM version 5.95.0 

used to construct the GSA/ PORFLOWTM model confirmed that the code conserves mass and 

satisfies Darcy’s Law (WSRC-TR-2004-00106). DOE indicates in the 2009 PA (page 238) that 

software quality assurance for the version of PORFLOWTM used for the local SDF/PORFLOWTM 

calculations (version 6.10.3) is covered by software quality assurance plans 

(WSRC-SQP-A-00028 and G-TR-G-00002) that cover things such as installation and 

maintenance, acceptance testing, configuration control, and quality control of software models. 

DOE provides details on the data used to construct the local SDF/PORFLOWTM models and the 

overall approach to modeling contaminant flow and transport from SDF sources to various 

points of assessment (SRNL-STI-2009-00115). 

PORFLOWTM was selected for use in the 2009 PA modeling for multiple reasons including its 

capability to perform necessary project functions, extensive verification testing by the vendor 

and QA testing by site personnel, and given the familiarity of project personnel with use of the 

code. One key example of PORFLOW’s capability to meet project needs is its ability to model 

first-order decay and progeny in-growth necessary for simulation of radionuclide transport. 

GoldSim

The GoldSim® code, developed by the GoldSim® Technology Group, is a commercially available 

program described by the developer as a user-friendly and highly graphical program for carrying 

out dynamic, probabilistic simulations of complex systems to support management and 

decision-making in engineering, science, and business. GoldSim® version 9.6 was used to 

perform the probabilistic uncertainty and sensitivity analysis in support of the 2009 PA as 

described in Section 2.1.2. DOE also uses GoldSim® to calculate the dose associated with 

(i) use of contaminated groundwater by a potential receptor and (ii) direct intrusion into the 

disposal site for comparison against dose based standards in §61.41 and §61.42. Figure 2.11-1 

illustrates model linkages to GoldSim. 
151 

GoldSim® Technology Group LLC provides a user’s guide and separate data validation guide 

(GTG-2007 Vol. 1 and Vol. 2, GTG-2006b). DOE provides information about GoldSim® software 

quality assurance (G-SQA-A-00011). 

GoldSim® was selected for use by DOE due to its sufficiency in meeting project needs, input 

and output capabilities, and given its use in other similar projects (Nevada Test Site and Yucca 

Mountain). 

CAP-88

The Clean Air Act Assessment Package – 1988 (CAP-88) computer model (EPA, 1993) is used 

to estimate dose from radionuclide emissions to air. CAP-88-PC version 1.0 was used to 

assess risk to a member of the public from inhalation of radionuclides that diffuse to the ground 

surface from SDF disposal cells. CAP-88 uses a modified Gaussian plume equation to estimate 

the average dispersion of radionuclides released from up to six sources at the same release 

location with different release heights. Assessments are performed with a circular grid with a 

radius up to 80 kilometers (50 miles). CAP-88 provides estimates of annual dose to Maximally 

Exposed Individuals (MEI) considering plume and ground shine (gamma radiation), inhalation, 

and foodstuff ingestion pathways using the results of the vapor phase radionuclide diffusion to 

surface calculations output from PORFLOWTM (model linkages in Figure 2.11-1) as described in 

Section 2.9 of this TER. 

A user’s manual, which provides instruction for use of the CAP-88-PC Version 1.0 is available 

(EPA-402-B-92-001). The software quality assurance plan (SQAP) for the CAP-88 version used 

for the 2009 PA calculations is covered by Q-SQP-A-00002. 

STADIUM®

The STADIUM® code (Software for Transport and Degradation in Unsaturated Materials) is a 

proprietary code developed by SIMCO Technologies Inc. to predict short and long-term 

behavior and service life of concrete. STADIUM® was used by SIMCO Technologies Inc. as a 

subcontractor to SRS for the 2009 PA to support the calculations of the ettringite front 

movement from sulfate reactions, and thus degradation front, through the disposal unit 

concrete. The results of the degradation modeling are abstracted for use in the PORFLOWTM 

model as described in Section 2.5 of this TER (Figure 2.11-1). 

A user guide (STADIUM® User Guide), which provides information on calculation methodology 

and instructions for use, is available at http://www.mslexperts.com/slm/stadium_help/index.html. 

STADIUM® has been validated via comparison both to laboratory tests and to field conditions of 

existing structures. The comparisons are provided in the user guide and show that STADIUM®

reproduced the results for both the laboratory and field materials tests (STADIUM® User Guide). 

The work by SIMCO Technologies Inc. as reported in SRNS-STI-2008-00050 used testing of 

laboratory prepared samples of the SDF concrete formulations to input the material parameters 

into STADIUM. 
152 

Geochemist’s Workbench

Geochemist’s Workbench is a commercially available computer program developed by 

RockWare. Geochemist’s Workbench Release 6.0 was used by DOE to model the change in 

pore water composition (i.e., transition to higher Eh and pH), resulting in Kd changes in the 

saltstone waste form and disposal units as described in Section 2.6 and 2.7 of this TER. 

Geochemist’s Workbench was also used in simulations of the formation of expansive mineral 

phases in saltstone (Section 2.6) and in modeling of degradation of the GCL (Section 2.5). 

2.11.2 Probabilistic Uncertainty and Sensitivity Analysis 

DOE included an uncertainty and sensitivity analysis in its PA documentation. DOE used the 

software program GoldSim® to construct probabilistic models to perform these analyses. Output 

data is extracted from PORFLOWTM for use in the probabilistic GoldSim® models. For example, 

velocity profiles were extracted from each two-dimensional PORFLOWTM near-field model (for 

each disposal unit type and case) for use in GoldSim® for Cases A - E. Because the same 

velocity profile is used for each case for each realization, uncertainty in flow through the system 

is evaluated in a limited sense (i.e., uncertainty from case to case is assessed, but uncertainty 

in flow in a particular case is not). Furthermore as discussed in Section 2.7, flow through 

saltstone is constrained to low values in Case A and other PA cases. 

As mentioned above, the GoldSim®

 model does not independently compute flow. Flows 

through the cover system, saltstone disposal cells, unsaturated, and saturated zones are 

extracted and simplified from more complex models such as PORFLOWTM for use in GoldSim®. 

Major components of the engineered system explicitly simulated in GoldSim® for three different 

disposal cell types: Vault 1, Vault 2 (FDCs), and Vault 4, include the saltstone disposal cells, 

disposal unit walls, and disposal unit floors. The GoldSim® representation also considers 

diffusion of radionuclides to “dirt” in contact with the disposal units. Diffusion of contaminants 

between the saltstone, disposal unit walls, and surrounding soil was solved analytically, rather 

than numerically, in GoldSim. Direct transfer pathways were used in GoldSim® to transfer mass 

between cells based on the analytical solutions to the diffusion equation. After transport through 

the saltstone disposal cells or into “dirt” surrounding the disposal unit walls, radionuclides are 

transported through the SDF vadose zone located below the disposal units. Twenty cells each 

are used to represent transport through the (i) saltstone, (ii) disposal unit walls, and (iii) dirt to 

which radionuclides diffuse from the waste zone. Ten cells are used to simulate transport from 

the SDF disposal units through the vadose zone underneath the disposal units to the saturated 

zone. A set of ten waste footprint cells are used to except flux from the vadose zone and 

transport contaminants to the downgradient edge of the SDF sources in the saturated zone. 

“Near well” cells are used to transport radionuclides from the waste footprint to the 100 meter 

boundary. In the probabilistic assessment, the specific aquifer used by the receptor is 

determined stochastically based on assumed probabilities for a potential receptor completing a 

well in each aquifer. 

A plume function is used to account for plume overlap due to transverse dispersion. The plume 

function calculates a factor to be applied to centerline plume concentrations at the 100 m 
153 

boundary for each individual disposal unit to reflect plume concentrations at various reference 

locations. The center node of each of twelve sectors (sectors are illustrated in Figure 2.3-1) are 

used as reference points where the contributions of the various disposal unit plumes are 

calculated (PA, page 320). Values for plume overlap are calculated only for plumes which may 

reasonably be thought to interact. For example, the model did not simulate any plume overlap 

between the north and south sections of the SDF, nor were plumes on opposite sides of the flow 

divide assumed to coalesce. 

Inherent differences between the multi-dimensional, PORFLOWTM models and the onedimensional, analytical (GoldSim®) model make it difficult to compare modeling results. DOE 

attempts to alleviate this problem through the benchmarking process. Benchmarking was used 

to change GoldSim® modeling parameters so that certain GoldSim® intermediate results better 

matched the corresponding intermediate results from PORFLOWTM for the same system and 

scenario being simulated. DOE deemed benchmarking necessary to allow a common point 

from which comparisons between the two models could be made. Only Case A and C were 

initially benchmarked for only three radionuclides: Ra-226, I-129, and Tc-99. Additional cases 

were benchmarked subsequent to initial efforts. Benchmarking occurred at two locations in the 

model domain: (i) the flux at the interface between the unsaturated and saturated zones, and 

(ii) concentrations at the 100 m boundary. DOE also compares member of the public doses at 

the 100 m compliance boundary for each sector during benchmarking to assess how well the 

two models match with respect to the overall dose from all radionuclides and disposal units, 

rather than individual radionuclides and disposal units that might not contribute significantly to 

dose. For a detailed description of the benchmarking process see Section 5.6.2 in the PA (page 

450). 

A benchmarking factor was used to attempt to alleviate model divergence associated with 

simulation of transport of Tc-99. PORFLOWTM uses what is termed a “shrinking core model” to 

simulate oxidation of saltstone along edges and fractures with oxidation rates controlling release 

of Tc-99. The “shrinking core model” is not explicitly simulated in GoldSim. Therefore, DOE 

uses the Tc-99 Kd as a benchmarking factor to control Tc-99 release. An increase by a factor of 

500 in the oxidized Kd and a reduction by a factor of 0.3 in the reduced Kd was selected during 

benchmarking. 

Adjustments made to GoldSim® to match flux from the vadose zone to saturated groundwater 

were constrained to flow parameters and included flow factors for the following: (i) saltstone 

grout, (ii) disposal unit walls, (iii) “dirt” located adjacent to disposal unit walls, and (iv) vadose 

zone soils located underneath the disposal units. 

Saturated zone transport was complicated by the contributions of various SDF sources to the 

100 m boundary locations. Adjustments made to GoldSim® saturated zone parameters included 

a (i) plume factor to account for groundwater divide effects in Sector J, and (ii) a factor to 

account for contributions from Vaults 1 and 4 to concentrations at the 100 m boundary in 

northern sectors. 
154 

2.11.3 DOE’s QA Program for Computer Codes 

The Quality Assurance (QA) program is an integral part of SRS’s Integrated Safety 

Management System (Management Policies (MP), WSRC 1-01, Policy 4.2 Quality Assurance). 

The QA policy requires that the SRS QA program comply with DOE O 414.1C, 10 CFR 830, 

Subpart A, and the Quality Assurance Management Plan (QAMP, WSRC-RP-92-225). The 

QAMP provides for the prevention, identification, and correction of any identified errors, as well 

as establishing an evaluation process to further continuous improvement. 

General requirements for QA are described in Quality Assurance Manual, Manual 1Q, 

Procedure 2-1 Quality Assurance Program. The software quality assurance plan (SQAP) 

requirements are described in Procedure 20-1, Software Quality Assurance Manual (Manual 1Q 

20-1). DOE indicates that Procedure 20-1 fulfills the requirements of DOE Order 414.1.C and 

10 CFR 830, Subpart A. 

Conduct of Engineering and Technical Support Procedure Manual E-7, Procedure 2.60, 

Technical Reviews (E7 Manual 2.60) is the QA implementing procedure for performing technical 

reviews. Various levels of rigor are required based on the risk associated with the end use of 

the data and calculations. With respect to PA activities, DOE indicates that a design checker 

assures the technical accuracy of the design document by performing design check activities 

such as mathematical checks, if appropriate, review for correct use of inputs, including quality 

requirements, review of approach used, and reasonableness of the output. Design checkers 

must be knowledgeable in the area being reviewed and be able to perform similar designs or 

analyses as those being checked but must also have independence from the actual work being 

checked. 

DOE indicates that assurance that the input data to the various codes is verified to be accurate 

is integral to the model integration process. DOE documented the verification of the model input 

traced from source documents to modeling input and finally to appropriate sections in the PA 

(Appendix I). 

2.11.4 NRC Evaluation – Computer Models and Computer Codes 

2.11.4.1 NRC Evaluation – Models, Modeling Approach, and Model Integration 

With noted exceptions, NRC staff generally finds the codes and models used by DOE to be 

adequate for the purpose of evaluating disposal facility compliance with the performance 

objectives in 10 CFR Part 61, Subpart C. Vendor verification and validation testing exists for 

many of the codes selected and a software quality assurance program is in place to help ensure 

the integrity of model calculations (i.e., ensure that model equations are solved consistently and 

correctly). Data input quality is also addressed in DOE’s quality assurance program. Data input 

quality is essential to ensuring the quality of model outputs. However, independent peer review 

of FTF geochemical modeling identified issues with DOE’s quality assurance program 

(LA-UR-12-00079). FTF geochemical modeling is relevant to SDF as geochemical modeling 

and data inputs are similar. The independent peer review found issues with the quality of 
155 

thermodynamic data used in the calculations, as well as issues with transparency and 

traceability of documentation, and project integration issues (LA-UR-12-00079). NRC staff 

expects DOE to address the quality assurance issues identified by the independent peer review 

group, as applicable to the SDF PA modeling inputs, models, and calculations in future updates 

to its PA. 

With respect to code selection, the codes used by DOE appear to be thoughtfully considered. 

However, the use of a more sophisticated code to model infiltration through the closure cap may 

provide more realistic estimates of infiltration. Although the HELP code may be suitable for 

estimating long-term water balances, short-term events and trends may not be adequately 

represented (Bonaparte et al., 2002). Bonaparte et al. (2002) state, “the model will generally 

not be adequate for use in a predictive or simulation mode, unless calibration is performed using 

site-specific measured (not default) material properties and actual leachate generation data.” 

Since calibration data over the lifetime of the planned closure cap is unavailable, the use of an 

alternative code may provide more defensible infiltration estimates. 

With respect to model integration, the NRC staff notes a few irregularities that result from the 

use of more than one PA model to evaluate various performance measures. For example, the 

GSA/PORFLOWTM model used to calculate stream concentrations to evaluate dose impacts 

associated with recreational use of stream water has a coarser grid resolution then the 

SDF/PORFLOWTM model. Because (i) sources are loaded at the element closest to the SDF 

disposal units, (ii) the selection of source loading locations are fewer in number for the 

GSA/PORFLOWTM model given its coarser resolution (compared to the SDF/PORFLOWTM 

model), and (iii) slight variations in source loading locations can lead to variations in flow paths 

away from the SDF (given divergent flow associated with the groundwater divide at SDF), flow 

paths away from some SDF sources to GSA streams appear to differ between models (flow 

paths in Figure 4.4-12 [produced from GSA/PORFLOWTM model] compared to 4.4-19 [produced 

from SDF/PORFLOWTM model]). Other source loading issues are discussed in Section 2.8 on 

far-field modeling. For example, code limitations (e.g., inability to load flux as a recharge 

concentration) or modeling approach (e.g., use of a local SDF/PORFLOWTM model that does 

not include a recharge boundary condition) may limit options to addressing source loading 

issues; however, if deemed risk-significant, a change to DOE’s modeling approach may be 

warranted. 

The NRC staff concludes that increased transparency of modeling results and evaluation of 

intermediate modeling outputs could lead to a greater understanding of total system 

performance, as well as facilitate problem identification and resolution. For example, the 

decision to break up the cover system and model separately upper and lower cover components 

in HELP and PORFLOWTM makes it difficult to evaluate total cover system performance. In 

fact, the No Closure Cap Case addresses under-performance of most of the cover but allows 

the lower lateral drainage layer (located above all disposal units) and HDPE/GCL (located 

directly above the FDCs) to operate at the same level of performance as in Case A 

(Section 2.13). In general, barrier performance often is obscured in PA sensitivity cases due to 

the redundancy of barriers and the tendency to “switch off” some barriers or parts of barrier 

systems while others remain “on” (NRC, 2010b; PA-8). In another example, the Synergistic 
156 

Case appears to be a highly conservative sensitivity case (i.e., represents a highly degraded 

state of the entire engineered system). However, the dose impact of this case is limited given 

the limited flow of water through the saltstone matrix with the bulk of the flow going through the 

cracks in the saltstone and the disposal unit walls. Therefore, rather than representing a 

seemingly bounding case as one might surmise based on its label, the doses associated with 

the Synergistic Case appear to be limited. Evaluation and presentation of intermediate outputs 

(e.g., flow rates through saltstone) may have led to a change in the approach used to model this 

PA case and faster resolution of technical issues. 

Information regarding performance of the FDCs, which the NRC staff found to be one of the two 

most important barriers affecting engineered system performance in Case K, was not provided. 

As discussed in Section 2.7 and 2.13, the FDC disposal unit floors served to substantially 

attenuate release of Tc-99 with over 90 percent of the Tc-99 retained in the disposal unit 

concrete shortly after release from the saltstone monolith. Although the disposal unit concrete 

significantly decreases the release rates from the engineered system, DOE did not provide 

details regarding key modeling assumptions and results related to disposal unit performance for 

this very important barrier in Case K. 

2.11.4.2 NRC Evaluation - Probabilistic Uncertainty and Sensitivity Analysis 

The NRC staff has concerns with how the flow-field was “hard-wired” into the probabilistic 

GoldSim® model. As previously discussed, flows were not independently modeled in the 

probabilistic model. Deterministic PORFLOWTM flow outputs were used to parameterize the 

GoldSim® models. Flow of water into the saltstone matrix is substantially limited in most of the 

deterministic PA Cases evaluated in the probabilistic model (Section 2.7). Because flow rates in 

each case are not varied in the probabilistic analysis, the peak dose is expected to be biased 

low in the probabilistic analysis. NRC staff concludes that the probabilistic GoldSim® model 

does not appropriately consider uncertainty in dose predictions due to potential uncertainty in 

the magnitude of flow of water and oxygen into the saltstone waste form. While flow variability 

could have been considered in the probabilistic model through evaluation of alternative 

conceptual models with higher flow rates, the probabilistic model only included a single PA case 

representing higher flow rates through the saltstone matrix (i.e., Case E), and this case is 

assigned a significantly lower probability (ten percent for Case E compared to eighty-five to 

ninety-five percent for Case A). Additionally, other cases with higher flow rates through the 

saltstone matrix that the NRC staff expects to be more likely (e.g., Case K) were developed 

after PA preparation and are, therefore, not considered in the probabilistic assessment. 

With regard to benchmarking, the NRC staff concludes that documentation of the benchmarking 

process used to align the GoldSim® probabilistic modeling results with deterministic 

PORFLOWTM modeling results could have been more complete and transparent in the PA. 

However, in response to NRC comment, DOE provided a well-developed comment PA-5 

response (SRR-CWDA-2011-00044) that attempted to explain and illustrate the need for and 

effect of each benchmarking factor or set of benchmarking factors added to the GoldSim®

 model 

to improve the match between PORFLOWTM and GoldSim® modeling results. However, 

because the benchmarking process inherently assumes that PORFLOWTM modeling results are 
157 

somehow better or more valid than GoldSim® modeling results, PORFLOWTM models must first 

be shown to be well calibrated, well understood, and sensible to prevent the inadvertent transfer 

of modeling artifacts or inaccuracies from the PORFLOWTM model to the GoldSim® model. The 

NRC staff observed that the SDF/PORLOW model does not, in fact, appear to be well calibrated 

(Section 2.8). Furthermore, it is not clear that the complicated flow patterns observed in 

deterministic near-field modeling are well understood and supported, as details are lacking in 

PA documentation. Because PORFLOWTM modeling results lack transparency and validation, it 

is difficult to conclude that the multitude of adjustments made to the GoldSim® model to mimic 

the complex flow behavior of water and contaminants through the engineered and natural 

systems are beneficial or well supported. DOE appears to take the approach of significantly 

simplifying the representation of certain processes in GoldSim® and adding in adjustment 

factors to enable reproduction of the complex system behavior observed in PORFLOW. If the 

PORFLOWTM model is determined to be well calibrated and validated, this approach might be 

acceptable, but it also limits the ability of the GoldSim® model to serve as an independent check 

of the PORFLOWTM modeling results. An independent model from which comparisons can be 

made and distinct differences in system behavior can be explained can only add value to the 

overall modeling process. 

As previously discussed, DOE’s response to NRC comment PA-5, (SRR-CWDA-2011-00044) 

illustrates the effect of adding in benchmarking factors or sets of benchmarking factors in a 

similar sequence to what was done during the actual benchmarking process. In some cases, 

adjustments lead to order of magnitude or more changes in near-field GoldSim® parameters. 

While some notable improvements were made (e.g., peak dose from Ra-226 from FDCs), the 

benefit of these adjustments is not always clear. For example, the Ra-226 fluxes out of the 

unsaturated zone from Vault 1 seem to benefit little from these adjustments with fluxes differing 

by an order of magnitude between models after benchmarking. Vault 4 flux comparisons are 

generally worse following the benchmarking process with a lowering of GoldSim® fluxes out of 

the unsaturated zone. 

With regard to saturated zone benchmarking, the NRC staff recognizes improvements in the 

match between modeling results after the application of benchmarking factors; however, the 

basis for the benchmarking factors is not always intuitive. DOE used very large adjustments to 

the contributions of Vault 4 to the Sectors A - C concentrations of Ra-226: factors of 70, 100, 

and 50 that DOE indicates account for heterogeneities in the flow rates (only a single Darcy 

velocity is used in GoldSim) and possibly due to vertical mixing. NRC staff reviewed 

PORFLOWTM modeling files and noted that plumes emanating from Vault 4 actually appear to 

be located relatively low in the Upper Three Runs aquifer at the 100 m boundary. Therefore, it 

appears that the Vault 4 plumes are vertically mixed in a larger aquifer thickness compared to 

the 12 m (39 ft) assumed in GoldSim® modeling necessitating a decrease, rather than an 

increase, in the concentrations of Ra-226 in Sectors A - C in GoldSim. DOE assigns an 

average Darcy velocity of 45 m/yr in the probabilistic GoldSim® model. NRC staff agree that the 

flow field is heterogeneous across the SDF; however, particle tracks from Vault 4 (PA, page 

386) do not seem to be significantly different than average flows across the site that would 

account for the rather significant factors used to adjust the Vault 4 100 m concentrations in 

Sectors A - C, again making it difficult to explain the large factors used in this adjustment (i.e., 
158 

differences in aquifer thicknesses and velocities are on the order of factors of 2 to 4, 

respectively, and would not explain the factor of 50 to 100 increase in Ra-226 concentrations in 

Sectors A - C). DOE uses a benchmarking factor of 5 to adjust the concentrations from FDCs 

7A - D23 (Figure 2.8-8) due to minimal mixing and possible vertical dispersion effects. The NRC 

staff agrees that dilution factors near the groundwater divide near Sector I are low; however, 

contaminant plumes emanating from FDCs 7A - D are well-dispersed vertically, potentially 

leading to an offset of low dilution associated with low flow (i.e., the adjustment may have been 

greater than 5 had it not been for significant vertical mixing). DOE also uses an adjustment 

factor to account for contributions from Vault 1 to Sectors E and F due to transverse dispersion. 

Review of PORFLOWTM model outputs indicates that while small amounts of contamination 

from Vault 1 are present in Sectors E and F, the contributions to Sector F are orders of 

magnitude lower than contributions to other Sectors. DOE uses an adjustment factor of 0.5 to 

decrease 100 m well concentrations to account for additional transverse vertical dispersion. 

Given the small aquifer thickness assumed in the GoldSim® modeling (i.e., 12 m), the NRC staff 

agrees that the amount of vertical mixing is expected to be larger particularly for northern sector 

plumes and sources located further from the 100 m boundary. However, the figures showing 

the effect of the benchmarking factor on the results seemed to show doses further out of 

alignment for some key radionuclides and sectors (e.g., Ra-226 and Sectors A - C). 

Nonetheless, the overall effect of all of the adjustments seemed to improve matches between 

the two models. 

Finally, DOE indicates that an adjustment had to be made to the GoldSim® model to account for 

contributions of Vaults 1 and 4 to northern sectors due to northerly flow from these vaults when 

the plumes are transported vertically downward to the Gordon aquifer. It was not obvious to the 

NRC staff why adjustments needed to be made during the benchmarking process to account for 

flow from these vaults to the north in the Gordon aquifer, an aquifer that is not explicitly 

represented in the GoldSim® model. As discussed below, DOE assumes that Gordon aquifer 

concentrations are 5 percent of the concentrations in the Upper Three Runs aquifer and states 

in the PA that DOE believes the factor to be conservative (a factor of 600 lower concentrations 

have actually been calculated for conservative tracers in the Gordon aquifer compared to the 

Upper Three Runs aquifer based on PORFLOWTM modeling simulations). Therefore, it is not 

clear why contamination of the Gordon aquifer from Vault 1 and 4 sources was significant 

enough to warrant adjustments to the GoldSim® model in representing Gordon aquifer 

concentrations. NRC staff hypothesize that very low concentrations of key radionuclides in 

Sectors G - L that occur very early (prior to FDC failure and therefore attributable to inventory in 

Vaults 1 and 4) represent the peak concentrations in those sectors but actually occur in the 

Gordon aquifer during early simulation timeframes. Thus, inventory from Vaults 1 and 4 may be 

used in benchmarking to increase 100 m concentrations in Sectors G - L in the Upper Three 

Runs aquifer, the only aquifer simulated in GoldSim, because the concentrations would 

otherwise be zero in the Gordon aquifer (0 multiplied by 5 percent is 0). Therefore, it appears 

DOE may have allowed nominal concentrations of key radionuclides to be present in the Upper 

Three Runs aquifer to allow an even smaller amount of these key radionuclides to be present in 

 

23 Disposal unit designations used in the TER (Figure 2.8-8) correspond to designations used in the PA, which may 

not correspond to current designations. 
159 

the Gordon aquifer (5 percent of the Upper Three Runs aquifer concentration). While physically 

inconsistent with the hydrogeological conceptual model (the Upper Three Runs should not be 

contaminated from Vaults 1 and 4 in the northern sectors), the need for the adjustment factor 

can be explained. 

The NRC staff concludes that the benchmarking process could be more transparent in PA 

documentation and appears in some cases to be more like a fitting routine rather than a process 

by which physical differences between two models can be explained and addressed. In some 

cases benchmarking factors are used to compensate for lack of complexity in the GoldSim®

probabilistic modeling (e.g., shrinking core model implementation in PORFLOWTM that is not 

explicitly represented in GoldSim® modeling or differences in the dimensionality of flow fields 

between the two models). In some instances, benchmarking factors often appear at odds with 

one another and some factors appear to lack physical basis. Given the issues identified by the 

NRC staff in the probabilistic model discussed below, it is likely that some adjustments may 

have been made to the GoldSim® model to compensate for errors in the probabilistic model. 

Furthermore, as discussed in comment PA-4 (NRC, 2010i), benchmarking is limited to just three 

key radionuclides: Ra-226, Tc-99, and I-129, potentially limiting the GoldSim® model’s ability to 

accurately evaluate of contributions of other constituents with different transport behavior to 

peak dose in the probabilistic analysis. If additional isotopes that were not originally 

benchmarked are potentially significant dose contributors for the SDF, the benchmarking 

process could benefit consideration of additional radionuclides not previously considered. 

In its probabilistic assessment, DOE considers the probability of a potential receptor completing 

a well in various SDF aquifers based on regional data. Concentrations in the Gordon aquifer 

from SDF source release are expected to be significantly less than those in the Upper Three 

Runs due to attenuation in the Gordon Confining Unit and increased dilution in the Gordon 

Aquifer. Therefore, consideration of doses due to exposure of a potential receptor to various 

groundwater dependent pathways resulting from extraction of contaminated well water from the 

Gordon aquifer (in lieu of the Upper Three Runs aquifer) will significantly lower the overall peak 

of the mean dose in the 2009 PA given the relatively high probability of well completion in the 

Gordon aquifer (Table 2.11-1). The NRC staff communicated its concern with DOE’s use of 

exposure point concentrations that did not represent the point of maximum exposure in a FTF 

scoping meeting (DOE, 2008) and in the FTF TER (NRC, 2011m). Thus, while averaging doses 

based on use of well water from various aquifers and probability of well completion may provide 

valuable risk insights, use of the dose at the point of maximum exposure in the UTR aquifer is 

more appropriate for comparison against 10 CFR Part 61, Subpart C performance objectives. 

Although NRC is unable to rely on DOE’s probabilistic analysis due to issues discussed in this 

section, if DOE intends to rely on its probabilistic analysis in the future, the point of assessment 

within the Gordon Aquifer may also be risk-significant. To determine the Gordon aquifer 

concentrations for use in the probabilistic model, aquifer concentrations at the 100 m boundary 

in the Gordon aquifer were compared to concentrations at the 100 m boundary in the Upper 

Three Runs aquifer. Based on this comparison, Gordon aquifer concentrations were assigned a 

relative concentration of a factor of 20 lower than Upper Three Runs aquifer concentrations for 

use in the probabilistic assessment. However, if higher concentrations are present at points 
160 

beyond the 100 m boundary in the Gordon aquifer that were not considered in the comparison, 

then the default factor of 20 lower concentrations assigned to wells that may be completed in 

the Gordon aquifer may under-predict doses for relatively non-sorbing constituents that tend to 

drive the peak dose at SDF. 

Table 2.11-1: Aquifer Exposure Probabilities and Relative Concentrations 

Aquifer Exposure Probability Fraction of UTR-UZ 

Concentration Weighted Concentration 

UTR-UZ 0.13 1.0 0.13 

UTR-LZ 0.44 1.0 0.44 

Gordon 0.43 0.05 0.02 

Total 1.0 NA 0.59 

Data in Columns Two and Three are from the 2009 PA pages 488-489. 

In fact, it is not clear that the exposure point concentrations at the 100 m compliance boundary 

in the Gordon aquifer are the maximum concentrations at or beyond the 100 m boundary. DOE 

presents PORFLOWTM simulation results that indicate plume concentrations decrease with 

distance from the source (2009 PA, pages 389, 390). However, the figures presented by DOE 

in cross section (e.g., Figures 5.2-6 and 5.2-7) actually seem to indicate that the maximum 

concentration in the Gordon aquifer occurs beyond the 100 m compliance boundary. This 

would be especially true for sources located near the SDF boundary and further away from the 

groundwater divide. For other sources located further away from the SDF boundary or on the 

groundwater divide, vertical gradients and transport distance may be sufficient to allow transport 

of the centerline of SDF source plumes to the Gordon Confining Unit within the 100-m 

compliance boundary. 

In an FTF RAI resolution meeting held on July 21, 2011, (NRC, 2011d), DOE indicated that it 

thought a factor of 20 reduction in concentration in the Gordon Aquifer is sufficiently 

conservative for locations of higher concentration that may occur beyond the 100 m (330 ft) 

boundary at FTF. It is not clear to the NRC staff if a factor 20 reduction in concentrations of 

Upper Three Runs source concentrations at the 100 m point of compliance also is bounding for 

concentrations in the Gordon aquifer at SDF considering that Gordon aquifer concentrations 

may be higher beyond the 100 m boundary. 

In addition to the concerns described above, NRC staff found that sufficient justification was not 

provided for many of the parameter distributions included in DOE’s probabilistic GoldSim®

model. The parameters discussed below are examples of areas that NRC staff have identified 

as needing more justification is not intended to be a comprehensive list. One example where a 

stronger basis is needed is the assumed probabilities of Cases A - E. When DOE’s GoldSim®

model was run for “All Cases,” the model probabilistically selected a case among five 

cases, 2009, PA Cases A - E. In the response to NRC comment PA-11 

(SRR-CWDA-2011-00044), DOE stated that these probabilities were developed by engineers 

and scientists on the 2009 PA development team using a systematic, step-by-step approach. 
161 

However, as noted in this RAI (NRC, 2010i) NRC staff finds that the probabilities are unrealistic 

and do not reflect the current and expected future conditions of the system. 

Additionally, NRC staff does not find the ±50 percent uncertainty range assumed for Eh and pH 

transition times to adequately capture the transition time uncertainty. DOE provided additional 

justification for this range in the response to NRC comment SP-8 (SRR-CWDA-2011-00044). 

Based on the additional information, NRC considers the ±50 percent uncertainty range likely to 

capture uncertainty that could result from variability in initial saltstone mineralogy, bulk density, 

and porosity adequately. However, it appears unlikely that the ±50 percent uncertainty 

accounts for differences caused by the choice of initial mineralogy in addition to uncertainty 

caused by other factors, such as the assumed reducing capacity of saltstone, which could 

cause the Eh and pH transition pore volumes to be lower than DOE’s Case A values. 

The NRC staff also concludes that there is insufficient justification for the uncertainty 

distributions for cementitious material Kd values. In developing these uncertainty distributions, 

DOE used a statistical analysis of data on SRS sediments from a single borehole in E-Area. In 

response to an original and follow-up NRC RAI on this topic, DOE acknowledged that the 

approach is not optimal and implied that it will consider studies that may help elucidate 

uncertainty and variability in cementitious material Kd values (SRR-CWDA-2010-00033, 

SRR-CWDA-2011-00044; SP-18). The NRC staff raised the same issue during its review of the 

F-Area Tank Farm PA (SRS-REG-2007-00002 Rev. 1). As discussed in that review 

(NRC, 2011m; pages 128, 130), the NRC staff does not think it is sufficient to note that variation 

in sediment Kd values is likely to exceed variation in cementitious materials. Additionally, recent 

research at SRNL (SRNL-STI-2011-00672), found that the range of measured Kd values for the 

sorption of Cs, Sr, and I onto a variety of formulations of simulated saltstone was much larger 

than the Kd range used by DOE in stochastic models. However, the authors of this study 

suggested that the large range could be due to experimental conditions that do not prevail in the 

field and the small number of samples analyzed. 

Finally, NRC staff also noted in comment B-1 (NRC, 2010i) that the basis for excluding the biotic 

transfer factors from the uncertainty analysis was unclear. DOE stated in their RAI response 

(SRR-CWDA-2011-00044) that they believed that analyses performed by DOE 

(SRR-CWDA-2010-00033; B-1) and CNWRA (2008) demonstrated that these parameters did 

not have a significant effect on dose and that the use of uncertainty distributions for these 

parameters could result in risk dilution. NRC staff notes that while DOE’s assessment showed 

small absolute changes in the dose due to these parameters, the relative changes were large. 

Additionally, the analysis performed by CNWRA was intended to be a preliminary sensitivity 

analysis to identify the most important parameters in the CNWRA BDOSE model. Additional 

justification beyond this preliminary sensitivity analysis is needed for excluding the biotic transfer 

factors from the uncertainty analysis. Finally, since these parameters affect the magnitude of 

the peak dose, not the timing of the peak, using uncertainty distributions would not result in risk 

dilution when calculating the peak of the mean. As discussed in more detail in Section 2.10, it is 

important to use technically defensible methods to analyze and propagate transfer factor 

variability into the 2009 PA calculations due to the wide variability in the transfer factor data. 
162 

2.11.4.3 NRC Evaluation - Quality Assurance 

NRC staff has evaluated DOE’s quality assurance program and concludes that DOE has an 

adequate quality assurance program, procedures, and plan in place to ensure the quality of PA 

documentation, modeling, and calculations. Nonetheless, execution of the quality assurance 

program appears to be sub-optimal in some key areas. Significant deficiencies in probabilistic 

PA models and documentation have been identified. Specifically, NRC staff noted several 

irregularities in GoldSim® probabilistic modeling as described below. 

As an example of a potential break-down in quality controls, the NRC staff concludes that 

design checks were not effective in ensuring the reasonableness of model outputs. DOE 

provided results in its PA that showed that the peak of the mean dose for the SDF would occur 

at around 450,000 years in Sector B (Figure 5.6-39 in PA). Upon further investigation of 

GoldSim® model outputs, NRC staff concluded that the overall peak of the mean dose is 

attributable to Ra-226 and Pu-239. Given the half-life of Pu-239 (24,000 years), the NRC staff 

believes a more thorough quality check of the results would have caused DOE to question the 

timing of the peak dose. The NRC staff found that unit inventories are used in the GoldSim®

model and then inventory multipliers are applied to model outputs to calculate doses based on 

the assumed SDF inventory. The GoldSim® model inaccurately multiplies the Pu-239 

concentrations at the 100 m boundary by the Pu-239 inventory, rather than by the virtually 

negligible Cm-247 inventory, predecessor to Pu-239, responsible for the Pu-239 found at the 

100 m assessment point a few hundreds of thousands of years after closure. A careful review 

of the model output would have likely led to identification of this issue in the GoldSim®

calculations, as significant quantities of Pu-239 would not be expected at the 100 m boundary at 

the time of the peak dose (which was greater than 10 half-lives of Pu-239), unless through 

production from a predecessor radionuclide. Furthermore, peak dose from Ra-226 appears to 

be underestimated by this approach (long-term Ra-226 dose appears to be attributable to ingrowth from U-238 and U-234 with the U-234 inventory significantly greater than the Ra-226 

inventory but being multiplied by the lower Ra-226 inventory). More detailed evaluation and 

interrogation of a key PA model output, the overall peak of the mean dose, was expected based 

on DOE’s documented QA procedures. 

Furthermore, as noted in Section 2.1, the NRC staff did not fully review the DOE probabilistic 

GoldSim® model due to NRC staff concern regarding the use of outputs from deterministic 

cases (Cases A - E) which NRC staff did not consider to be appropriate inputs to the GoldSim®

model. NRC staff also had concerns about the adjustments made to the model during the 

benchmarking process. However, in performing a brief review of this model, NRC staff 

identified a number of items that seemed to be errors in the model. For example, the DOE 

GoldSim® model is designed to either run Cases A - E individually or to run “all cases” (including 

Cases A to E). When the model is run in the “all cases” mode (i.e., when the 

FixedRunCase_switch is set to false), a case is selected probabilistically for each realization 

based on DOE’s predicted probabilities for each case. The flows for Cases A - E seem to be 

selected correctly based on the case that was randomly selected. However, parameters related 

to fractures in Case C (BypassFrac, BypassWallDiff, CaseCTurnOffDiff) are not linked to the 

stochastic “Configuration” element and consequently do not change based on the case that is 
163 

probabilistically selected. Because of this, the “all cases” results do not seem to include key 

features of Case C, which likely led to the all cases probabilistic dose being underestimated. 

The flows in the unsaturated zone cells are also linked incorrectly. According to the 

documentation in DOE’s GoldSim® model, the outflow from UZCell_Out is intended to be evenly 

divided between the ten footprint cells below the disposal units. However, one ninth of the 

outflow from UZCell_Out goes to the Footprint_In cell and the remainder is split between 

UZCells 2 - 9. This results in a loop in which the contaminated water is routed back up into the 

UZ instead of going to the cells below the unsaturated zone. The effect of this contaminant 

routing is that the timing of the peak is delayed and the magnitude of the peak is decreased. 

In addition, the Vault selector for Vault 4 (data element “Vault” under \TheVaults\Vault 4) seems 

to be set incorrectly. The value for this selector is 2, which seems to correspond to an FDC. 

This selector affects the modeled vault area and the “Vault Index” (which affects the thickness of 

the unsaturated zone). 

2.11.4.4 Computer Codes and Models Conclusions 

The NRC staff finds the computer codes used by DOE in the performance assessment to be 

generally adequate for demonstrating compliance with the performance objectives. The NRC 

staff also believes that increased transparency in PA documentation and a shared 

understanding of total system performance facilitated by presentation and evaluation of 

intermediate modeling outputs could lead to more efficient resolution of key technical issues. 

NRC staff notes that some of the 2009 PA cases appear to be inconsistent with the conceptual 

model for the case. In these instances, presentation of intermediate outputs could facilitate 

identification of potential issues and provide early opportunity for adjustments, if determined to 

be necessary. 

Additionally, as discussed in Section 2.1, the NRC staff is unable to use DOE’s probabilistic 

analysis to assess facility compliance with the performance objectives in 10 CFR 61, Subpart C. 

DOE’s probabilistic model is based on a set of PA cases (Cases A - E) that NRC staff does not 

think appropriately represent the system based on currently available information. The NRC 

staff also has concerns regarding the transparency of and basis for some of the factors applied 

in the benchmarking process used to adjust the outputs of the probabilistic model to more 

closely match those of deterministic PORFLOWTM models. Improvements to the benchmarking 

process could be made in the future. Additionally, key parameters in the probabilistic 

assessment are “hard wired” (i.e., are not treated as uncertain parameters) limiting the 

evaluation of uncertainty in peak dose due to parameters such as flow rates through the 

saltstone waste form. The NRC staff also concludes that some of the uncertainty distributions 

used by DOE are not adequately supported. Finally, as described above, the NRC staff finds a 

number of possible errors in DOE’s probabilistic GoldSim® model that further limit the 

usefulness of probabilistic modeling results. 
164 

2.12 ALARA Analysis 

2.12.1 ALARA Analysis 

The method DOE used to demonstrate that doses to the off-site member of the public be 

maintained As Low As Reasonably Achievable (ALARA), as required by the 10 CFR 61, 

Subpart C performance objectives, is presented in Section 5.7 of the 2009 PA. The discussion 

that follows was provided in that section and is supplemented by DOE’s responses to NRC’s 

RAI’s (SRR-CWDA-2011-00033 and SRR-CWDA-2011-00044). 

DOE’s ALARA program is based on DOE Order 435.1-1. The NRC’s ALARA requirement, as it 

applies to §61.41, is as follows: 

“Reasonable effort should be made to maintain releases of radioactivity in effluents to the 

general environment as low as is reasonably achievable.” 

As noted in DOE’s second RAI response (SRR-CWDA-2011-00044, comment A-1) DOE’s 

guidance document for DOE Order 435.1-1 states in part: “...that the goal of the ALARA 

process is not the attainment of a particular dose level (or, in this case, level of release), but 

rather the attainment of the lowest practical dose level after taking into account social, technical, 

economic, and public policy considerations. The PA should include assessments that focus on 

alternatives for LLW disposal. ALARA is meant to provide a documented answer to the 

question: ‘Have I done all that I can reasonably do to reduce radiation doses or releases to the 

environment?’” 

In the 2009 PA, DOE stated that the goal of its ALARA process is to attain the lowest practical 

dose level after taking into account social, technical, economic, and public policy considerations. 

DOE also notes that the ALARA program at the SRS is well-documented. DOE assumed in the 

PA that using the SRS ALARA program, processes, and typical protocols to meet the ALARA 

requirement of §20.1003 would be sufficient to demonstrate that salt waste disposal at the SDF 

meets the §61.41 ALARA requirement. 

The PA states that the design of the FDCs was intended to improve the demonstration of 

ALARA at the SDF by including features that were expected to reduce releases of radioactivity 

below levels necessary to meet the dose limit. The PA lists some of these features, including 

those listed below. 
165 

• Carbon steel walls to ensure watertight containment 

• Minimum 20 cm (8 in) thick pre-cast walls of Class III sulfate resistant concrete 

• Interior coating 

• GCL in place above and below the FDCs 

• HDPE liner completely surrounding the cells 

• Thin layer of Class III sulfate resistant concrete on top to protect the GCL-HDPE during 

construction 

DOE notes in SRR-CWDA-2011-00044 that, in addition to evaluating and improving disposal 

cell design, DOE’s ALARA process also includes (1) making conservative assumptions when 

modeling, and (2) evaluating and implementing alternative salt processing processes that could 

reduce the SDF inventory. For example, DOE notes the Small Column Ion Exchange 

technology as an alternative considered for better salt processing. 

An additional example of a conservatism noted by DOE (SRR-CWDA-2011-00044) is that in the 

PA models, DOE establishes a 100-meter perimeter around the SDF that it evaluates after a 

100-year institutional control period. Since DOE owns the land, it is assumed that the site 

boundary will not change at any point during this institutional control period. DOE’s current 

access point for any off-site member of the public is approximately 8 kilometers (5 miles) from 

the SDF site boundary. DOE states that this distance is sufficient for assuming that any 

dispersion of radionuclides from the SDF would be negligible to any MOP at the site boundary. 

In Section 7.2 of the PA, DOE provides a summary of all of the assumptions they believe to be 

conservative. 

DOE also states (SRR-CWDA-2010-00033) that since the calculated dose as presented in the 

2009 PA (Case A) is well below the limits specified by the 10 CFR 61 performance objectives, 

DOE believes that a qualitative analysis (in lieu of any additional rigorous quantitative analysis) 

is reasonable for demonstrating ALARA. 

2.12.2 NRC Evaluation – ALARA Analysis 

DOE’s discussion of its ALARA analysis (2009 PA; Section 5.7) provides several examples of 

technical issues that DOE considered. In response to comments in RAI-2009-01 (NRC, 2010b) 

and RAI-2009-02 (NRC, 2010i), DOE provides further description of technical, social, economic, 

and public policy considerations taken in its ALARA analysis (SRR-CWDA-2010-00033 and 

SRR-CWDA-2011-00044). 

As discussed in more detail in Chapter 4, NRC staff believes that DOE’s radiation protection 

program limits releases to the environment and subsequent dose to members of the public while 

the site remains under DOE control. The NRC staff also agrees that while DOE maintains 

control of the site, the potential dose to a member of the public is limited due to the distance 

from SDF to the site boundary. The staff also agrees that, in the near-term, improvements 
166 

made to the disposal unit design reduce the potential release to the environment. However, the 

amount that these improvements affect long-term performance is not clear (Section 2.5). The 

NRC staff concludes that DOE has taken appropriate actions to ensure that near-term doses 

are ALARA. 

In its second response to comment A-1 (SRR-CWDA-2011-00044), DOE states that a costbenefit analysis is untimely due to ongoing operations and currently unavailable costs of new 

technology. However, NRC staff believes that changes to various aspects of the disposal 

process could be considered prior to completion of salt operations. In its second response to 

comment A-1 (SRR-CWDA-2011-00044), DOE’s consideration of introducing Small Column Ion 

Exchange technology is an example of a reasonable consideration. 

Also in its response to comment A-1 (SRR-CWDA-2011-00044), DOE states that consideration 

of many alternatives to the current disposal process was taken in DOE’s 2001 EIS for salt 

processing alternatives, DOE-EIS-0082-S2. NRC staff agrees that this reference provides a 

good analysis of potentially dose limiting alternatives and considers aspects of alternatives not 

discussed in the 2009 PA. However, the NRC staff believes that potential design changes and 

technological advances are not sufficiently considered in the DOE’s ALARA analysis. 

Alternatives that could be considered could include: 

• Evaluation of practicality of additional Tc removal, 

• Evaluation of practicality of adding a stronger Tc-reducing agent, 

• Evaluation of practicality of improving waste form quality, and 

• Evaluation of practicality of maintaining better quality control on water to cement 

ratios and curing temperatures (Section 2.6). 

The NRC staff agrees with the concept that a less detailed ALARA analysis is required when the 

predicted doses are low, however, the staff believes that only the short-term aspects of DOE’s 

ALARA demonstration are currently considered sufficient. As discussed in Section 2.1, the 

NRC staff does not believe that Case A, which DOE used as their compliance case, adequately 

represents the expected current and future conditions of the SDF. Additionally, the NRC staff 

believes that many of the assumptions DOE included in this case are non-conservative. As 

discussed in more detail in Section 2.13, the NRC staff does not agree with DOE’s statement 

that the expected dose to an off-site member of the public is well below the 10 CFR 61 

performance objectives. Because of NRC staff concerns regarding DOE demonstration of 

compliance with §61.41 (Section 2.13), the NRC staff has not made a conclusion regarding 

DOE’s long-term ALARA demonstration for this performance objective. 
167 

2.13 Protection of the General Population from Releases of Radioactivity 

2.13.1 DOE Dose Calculations 

To evaluate the potential dose to a member of the public, DOE evaluated the dose to an adult 

assumed to live and extract groundwater at the point of maximum exposure beyond a 100 m 

(330 ft) buffer zone around the SDF. The point of greatest exposure was calculated by splitting 

the SDF into 12 sectors and evaluating the dose downgradient of the SDF in each sector 

(Figure 2.3-1). In each sector, DOE used the maximum groundwater concentration from any 

depth (i.e., from the Upper Three Runs upper or lower zones or the Gordon aquifer). Peak 

concentrations typically occurred in the Upper Three Runs lower zone. Modeled peak 

radionuclide concentrations in the Upper Three Runs aquifer were significantly greater than 

modeled concentrations in the Gordon aquifer because of (1) attenuation in the Gordon 

Confining Unit, (2) limited release to the Gordon aquifer, and (3) significant dilution in the 

Gordon aquifer. Because predicted concentrations decreased with increasing distance from the 

SDF in the UTR, the point of maximum exposure in each case was 100 m (330 ft) downgradient 

of the SDF, although the sector with the maximum exposure depended on the particular case 

being evaluated. Groundwater was assumed to be used for drinking water consumption, plant 

irrigation, and watering livestock. The nearby stream water was assumed to be used for fishing 

and swimming (Section 2.3.2). For most cases, the dominant pathways were drinking water 

ingestion, fish ingestion, and vegetable ingestion. Air concentrations also were considered but 

did not contribute appreciably to predicted doses (Section 2.9). 

In Case A, which DOE considers the base case, the predicted peak dose was greatest in Sector 

B (0.014 mSv/yr [1.4 mrem/yr]) within 10,000 years of site closure and Sector I (0.031 mSv/yr 

[3.1 mrem/yr]) within 20,000 years of closure. The primary radionuclides contributing to the 

Sector B peak dose include Ra-226 (94%), and I-129 (4%), with other radionuclides contributing 

less than 1% each to the total predicted dose. For Section B, the dominant pathways for the 

10,000 year peak dose include drinking water ingestion (49%), fish ingestion (22%), vegetable 

ingestion (22%) and others (6%). The peak dose in Sector I within 20,000 years of closure was 

primarily attributable to I-129 (87%) and Ra-226 (13%), with all other nuclides contributing less 

than 1% to the total dose. Pathway contributions for the Sector I peak dose within 20,000 years 

include drinking water ingestion (45%), fish ingestion (28%), vegetable ingestion (21%) and 

others (7%). 

As previously discussed (Section 2.1), in response to an NRC request for a revised base case 

(NRC, 2010i; PA-8), DOE supplied Cases K, K1, and K2. The main differences between Case 

A and Cases K, K1, and K2, which are summarized in Table 2-1, include consideration of 

increased saltstone hydraulic conductivity, oxidation of saltstone proceeding from fractures, the 

assumption that the relative permeability in the saltstone and disposal unit concrete is always 1 

(i.e., the level of saturation does not influence hydraulic conductivity or saturation is equal to 1), 

increased hydraulic conductivity and diffusivity of disposal unit concrete, reduced inventory of 

Ra-226 and its ancestors, and updated biosphere modeling. The only differences between 

Cases K, K1, and K2 are the coefficients used to model Tc sorption in oxidized and reduced 

saltstone (i.e., Kd values for Tc-99 in saltstone). 
168 

DOE Case K peak dose results were highest for Sector I. The peak dose within 10,000 years of 

site closure was 0.13 mSv/yr (13 mrem/yr). DOE predicted that a peak dose of 0.55 mSv/yr 

(55 mrem/yr) would occur within 20,000 years of site closure. Radionuclides contributing to the 

Case K Sector I 10,000 yr peak dose include I-129 (71%), Cs-135 (28%), and Tc-99 (1.6%), 

with all other radionuclides contributing less than 1% to the total dose. Pathway contributions to 

the peak dose within 10,000 years of closure include drinking water ingestion (49%), fish 

ingestion (40%), vegetable ingestion (8%) and others (3%). Of these pathways, dose from the 

fish ingestion pathway was primarily attributable to Cs-135 while the dose from the drinking 

water and vegetable ingestion pathways was primarily due to I-129. In contrast, the Case K 

peak dose within 20,000 years is dominated by Tc-99, with all other radionuclides contributing 

less than 0.4% of the peak dose. No description of the primary dose pathways was provided for 

the peak dose within 20,000 years for Case K. DOE projected the peak dose from Case K1, 

which uses different Kd values for Tc in saltstone (i.e., 500 mL/g in reducing saltstone and 

0.8 mL/g oxidized saltstone for Case K1, as compared to 1,000 mL/g in reducing saltstone and 

10 mL/g in oxidized saltstone in Case K), to be approximately 0.09 mSv/yr (90 mrem/yr) and to 

occur at approximately 12,900 years after site closure (Figure 2.13-1). DOE projected the peak 

dose from Case K2 (500 mL/g in reducing saltstone and 0.8 mL/g oxidized saltstone) to be 

about 0.80 mSv/yr (80 mrem/yr) and, like the peak dose from Case K1, to occur at 

approximately 12,900 years after closure. 
169 

Figure 2.13-1: Doses from DOE Case K with variations in assumed Kds for Tc in oxidized 

and reducing saltstone (SRR-CWDA-2011-00044; Figure SP-19.4).24

Peak doses from Cases K, K1, and K2 are dominated by release of Tc-99 in Sector I, with the 

majority of the dose attributable to releases from FDCs 7A-D25 (Figure 2.8-8). Because the 

FDCs were all assumed to have an identical inventory in DOE’s deterministic analysis, the 

contribution from these disposal units reflects their proximity to the 100 m (330 ft) boundary and 

far-field flow effects (Section 2.8). The projected peak doses from Ra-226 in Cases K, K1, and 

K1 are smaller than the projected peak dose from Ra-226 in Case A because the inventory of 

Ra-226 and its ancestors Th-230 and U-234 were lower in Case K, K1, and K2 than they are in 

the other cases. The projected peak doses from Tc-99 are greater in Cases K, K1, and K2 than 

in Case A because Cases K, K1, and K2 include the effects of oxidation of saltstone proceeding 

from fractures and because these cases allow significantly more water to flow through the 

saltstone matrix than Case A does (Section 2.7). 

 

24 To convert mrem/yr to mSv/yr, divide values on the vertical axis by 100. 25 Disposal unit designations used in the TER correspond to designations used in the 2009 PA (as repeated in this 

TER; Figure 2.8-8), which may not correspond to current designations. 
170 

2.13.2 DOE Uncertainty and Sensitivity Analysis for Protection of the General 

Population 

In the initial PA, DOE submitted deterministic PORFLOWTM analyses of Case A, four 

“alternative disposal unit configuration” sensitivity analyses (Cases B - E), and six additional 

deterministic sensitivity analyses, which DOE called “single parameter” sensitivity cases. These 

cases are described in more detail in Section 2.1.1. In brief, Case A assumes saltstone is 

unfractured with constant hydraulic conductivity (2.0x10-9 cm/s) and diffusivity (1x10-7 cm2

/s) for 

the 20,000 year evaluation period. Case B and C both model fast flow paths through the 

saltstone, disposal unit floor, and basemat, with more flow paths in Case C than in Case B. 

Case D is similar to Case A except that the sheet drain in the FDCs and Vault 4 is modeled as a 

capillary break (i.e., an impediment to flow) instead of as a fast flow path. Case E does not 

include fast flow paths, but assumes the saltstone has a high bulk hydraulic conductivity 

(1.7x10-3 cm/s) and a moisture characteristic curve representing degraded material. In addition 

to these alternative disposal unit configuration sensitivity analyses, DOE included six additional 

sensitivity analyses in the PA. DOE referred to these as “single parameter” sensitivity analyses 

(2009 PA; Section 5.6.6), although several parameters are actually modified in two of the six 

cases. These cases include evaluations of the effects of the closure cap, increased or 

decreased sulfate attack on the Vault 1 and 4 concrete (relative to Case A), oxidation of 

disposal unit concrete, increased saltstone hydraulic conductivity, and a synergistic case that 

includes decreased cap performance, chemically and hydraulically degraded disposal unit 

concrete, water flow through fractures in saltstone, and chemical oxidation of saltstone from 

fractures (Table 2-1). 

In most of the cases in the PA, including Case A, the majority of peak doses DOE predicted to 

occur within 10,000 or 20,000 years of site closure occur in Sector B (Table 2-13), with the 

remaining peaks occurring in Sector I (Table 2-14). The majority of peak doses DOE predicted 

to occur within either 10,000 or 20,000 years in Sector B are dominated by Ra-226 

(SRR-CWDA-2010-00033; Tables PA-1.1 and PA-1.3). Exceptions include the peak dose from 

Case A within 20,000 years, which occurs in Sector I and is split almost evenly between Ra-226 

and I-129, and the peak doses in the Oxidized Concrete Case (within 10,000 years) the 10X 

Sulfate Attack Case (within 20,000 years), which occur in Sector B but are dominated by Tc-99. 

The peak dose DOE predicts within 20,000 years of closure in the “no cap” sensitivity analysis 

occurs in Sector I and is largely attributable to I-129. As previously discussed, the peak doses 

within either 10,000 or 20,000 years of closure in Cases K, K1, and K2 are attributable to Tc-99 

in Sector I. 
171 

Table 2.13-1: Sector B Doses from DOE’s Deterministic Analyses 

Case

Sector B 

Reference within 10,000 yrs within 20,000 yrs 

Dose 

(mrem/yr) 

Time 

(yrs) 

Dose 

(mrem/yr)

Time 

(yrs) 

Case A 1.4 10,000 2.9 15,080 SRR-CWDA-2009-00017 

Table 5.5-1 

Case B 3.9 1 4,180 6.0 1 15,740 SRR-CWDA-2010-00033 

Table PA-1.1 

Case C 5.1 1 7,360 5.5 1 15,420 SRR-CWDA-2010-00033 

Table PA-1.1 

Case C – PA-4 Case ~5.1 2 7,360 ~7.4 2 20,000 SRR-CWDA-2011-00044 

Figure PA-4.1 

Case D 1.3 1 9,800 2.11 16,180 SRR-CWDA-2010-00033 

Table PA-1.1 

Case E 56 1 9,400 56 1 9,400 SRR-CWDA-2010-00033 

Table PA-1.1 

No Closure Cap (with 

credit for lower lateral 

drainage layer) 

2.1 1 9,800 3.5 1 15,060 SRR-CWDA-2009-00017 

Table 5.6-17 

10x Sulfate Attack 1.7 1 10,000 4.8 1 12,140 SRR-CWDA-2009-00017 

Table 5.6-19 

No Sulfate Attack 0.8 1 10,000 2.3 1 15,600 SRR-CWDA-2009-00017 

Table 5.6-19 

Oxidized Concrete 0.4 1 860 2.5 1 15,820 SRR-CWDA-2009-00017 

Table 5.6-21 

Increased Saltstone 

Hydraulic Conductivity 5.4 1 9,860 11 1,3 15,680 SRR-CWDA-2009-00017 

Table 5.6-22 

Synergistic Case 

(as reported in PA) 4 18 1,3 6,380 28 1,3 15,760 SRR-CWDA-2009-00017 

Table 5.6-20 

Synergistic Case 

(as reported in SRRCWDA-2011-00044) 4

~23 1 ~6,500 ~36 1 ~16,000 SRR-CWDA-2011-00044 

Figure PA-9.1 

Synergistic Case - 

VP-1 ~19 not reported 

not 

reported

not 

reported 

SRR-CWDA-2010-00033

VP-1 
172 

Case

Sector B 

Reference within 10,000 yrs within 20,000 yrs 

Dose 

(mrem/yr) 

Time 

(yrs) 

Dose 

(mrem/yr)

Time 

(yrs) 

Synergistic Case - 

PA-9 ~24 2 ~8,500 ~37 2 ~16,000 SRR-CWDA-2011-00044 

Figure PA-9.1 

Synergistic Case - 

Updated PA-9 ~15 2,3 ~8,500 ~24 2 ~16,000 SRR-CWDA-2011-00044 

Figure PA-9.1 

Case K 8.8 8,710 29† 15,258 SRR-CWDA-2011-00044 

Table PA-8.11 

Case K1 not 

reported 

not 

reported NA 

Case K2 not 

reported 

not 

reported NA 

Bold values indicate the peak dose from any sector in the indicated time frame (all peaks occur either in 

Sector B or Sector I) 

To convert mrem/yr to mSv/yr, divide values in either dose column by 100.

1

 Dose only includes key radionuclides 

2

 Dose includes more radionuclides than the key radionuclides, but does not include the entire list 

considered in Case A 

3

 Value rounded to two significant digits by NRC staff 

4

 Different values were reported for the unmodified synergistic case in the PA and second RAI response 

Table 2.13-2: Sector I Dose Results from DOE’s Deterministic Analyses 

Case

Sector I

Reference within 10,000 yrs within 20,000 yrs 

Dose 

(mrem/yr) 

Time 

(yrs) 

Dose 

(mrem/yr) 

Time 

(yrs) 

Case A 0.4 10,000 3.1 15,080 SRR-CWDA-2009-00017 

Table 5.5-1 

Case B 1.6 1 4,180 1.9 1 15,740 SRR-CWDA-2010-00033 

Table PA-1.2 

Case C 1.9 1 7,360 2.2 1 15,500 SRR-CWDA-2010-00033 

Table PA-1.2 

Case C – PA-4 ~2.0 2 7,360 ~2.4 2 20,000 SRR-CWDA-2011-00044 

Figure PA-4.4 

Case D 0.3 1 9,840 0.6 1 15,580 SRR-CWDA-2010-00033 

Table PA-1.2 

Case E 14 1 9,400 19 1 15,060 SRR-CWDA-2010-00033 

Table PA-1.2 

No Closure Cap (with 

credit for lower lateral 

drainage layer) 

0.6 1

 10,000 3.6 1 15,060 SRR-CWDA-2009-00017 

Table 5.6-17 
173 

Case

Sector I

Reference within 10,000 yrs within 20,000 yrs 

Dose 

(mrem/yr) 

Time 

(yrs) 

Dose 

(mrem/yr) 

Time 

(yrs) 

10x Sulfate Attack 0.9 1 2,420 1.4 1 12,600 SRR-CWDA-2009-00017 

Table 5.6-19 

No Sulfate Attack 0.2 1 10,000 0.6 1 15,600 SRR-CWDA-2009-00017 

Table 5.6-19 

Oxidized Concrete 0.1 1 10,000 2.3 1 15,080 SRR-CWDA-2009-00017 

Table 5.6-21 

Increased Saltstone 

Hydraulic Conductivity 0.3 1

 10,000 3.9 1 15,040 SRR-CWDA-2009-00017 

Table 5.6-22 

Synergistic Case 

(as reported in PA) 4 2.1 1

 10,000 5.4 1 17,320 SRR-CWDA-2009-00017 

Table 5.6-20 

Synergistic Case 

(as reported in SRRCWDA-2011-00044) 4

~7 1

 ~9,000 ~14 1 ~16,000 SRR-CWDA-2011-00044 

Figure PA-9.4 

Synergistic Case – 

VP-1 ~3 not reported not reported not reported

SRR-CWDA-2010-00033 

Text, VP-1 

Synergistic Case - 

 PA-9 Case ~8 2

 ~9,000 ~15 2,3 ~16,000 SRR-CWDA-2011-00044 

Figure PA-9.4 

Synergistic Case - 

Updated PA-9 ~3 2

 10,000 ~5.5 2 ~15,500 SRR-CWDA-2011-00044 

Figure PA-9.4 

Case K 13† 8,720 55† 15,348 SRR-CWDA-2011-00044 

Table PA-8.11 

Case K1 ~13 ~9,000 ~90 ~12,800 SRR-CWDA-2011-00044 

Figure SP-19.4 

Case K2 ~13 ~9,000 ~80 ~13,200 SRR-CWDA-2011-00044 

Figure SP-19.4 

Bold values indicate the peak dose from any sector in the indicated time frame (all peaks occur either in 

Sector B or Sector I). 

To convert mrem/yr to mSv/yr, divide values in either dose column by 100. 

1

 Dose only includes key radionuclides

2

 Dose includes more radionuclides than the key radionuclides, but does not include the entire list 

considered in Case A 

3 Value rounded to two significant digits by NRC staff 

4 Different values were reported for the unmodified synergistic case in the PA and second RAI response 

To support its demonstration of compliance with the performance objective for protection of a 

member of the general public, DOE also submitted a probabilistic model that was intended to 

evaluate the weighted average of GoldSim® representations of Cases A through E (2009 PA; 

Table 5.6-3). However, because near-field flow results were incorporated from the deterministic 

model, DOE’s probabilistic analysis did not allow variation in parameters affecting flow in the 

near field (i.e., the saltstone, disposal units, or backfill in the unsaturated zone) within individual 

cases. 
174 

DOE’s original probabilistic analysis included sensitivity results from Case A and Case C. The 

Case A probabilistic analysis indicated the peak dose within 10,000 years was most sensitive to 

the sorption coefficient for Ra-226 in sandy soil, followed by the local fraction of vegetable 

consumption and the saturated zone thickness (2009 PA; Table 5.6-14). The Case A peak 

dose within 20,000 years of closure was sensitive to the Kd for I-129 in reducing middle aged 

concrete, vegetable production yield, and the number of pore volumes required for concrete to 

transition to middle age (2009 PA; Table 5.6-14). Results for Case C, which reflected the effect 

of water flow through fast pathways through the saltstone, floor, and basemat, showed the dose 

within 10,000 years is sensitive to the Kd for Pu in clayey soil, the unsaturated zone thickness 

below the FDCs, followed by the Kd for Pu in sandy soil. The peak dose within 20,000 years 

was sensitive to the same parameters, but more sensitive to the Kd for Pu in sandy soil than 

clayey soil (2009 PA; Table 5.6-15). 

In response to an NRC RAI (NRC, 2010i) DOE provided sensitivity results for Case E, which is 

the only case reflecting hydraulic degradation of bulk saltstone that was included in DOE’s 

probabilistic analysis. The Case E probabilistic results had a peak of the mean value within 

10,000 years of 1.32 mSv/yr (132 mrem/yr) in Sector B (SRR-CWDA-2010-00033), in 

comparison to the deterministic results for Case E, which have a projected peak of 0.56 mSv/yr 

(56 mrem/yr) within either 10,000 or 20,000 years of closure. The probabilistic results indicate 

that the Case E peak dose within 10,000 years is sensitive to the Kd for Ra in sandy soil, the 

unsaturated zone thickness, and the local fraction of vegetable consumption. Consistent with 

the sensitivity of dose to the sandy soil Kd for Ra, the peak dose in both the deterministic and 

probabilistic analyses is primarily attributable to Ra-226 (SRR-CWDA-2010-00033; PA-1 and 

PA-2). 

2.13.3 NRC Evaluation – Evaluation of Performance Assessment Results for 

Protection of the General Public 

In support of its compliance demonstration, DOE submitted a deterministic base case, several 

deterministic sensitivity analyses, and a probabilistic uncertainty and sensitivity analysis. 

Because of concerns about the design and implementation of the probabilistic model 

(Section 2.11), the NRC staff did not rely on the probabilistic model in its compliance evaluation. 

Thus the NRC evaluation of uncertainty in dose and the sensitivity of dose to input assumptions 

included three main factors: (1) an evaluation of the results of DOE’s deterministic sensitivity 

cases, (2) an evaluation of intermediate model results from DOE’s deterministic sensitivity 

analyses, and (3) independent sensitivity analyses focused on Cases K, K1, and K2. The NRC 

staff conducted independent sensitivity analyses to evaluate and attempt to isolate the effects of 

optimistic and pessimistic assumptions used in DOE’s analyses, and to investigate the effects of 

unexpected barrier performance indicated by intermediate results. In addition, the NRC staff 

conducted sensitivity analyses focused on Cases K, K1, and K2 because the NRC staff relied 

on these cases heavily in its review and these cases were supplied by DOE without 

accompanying sensitivity or uncertainty analyses. This section includes an evaluation of DOE’s 

sensitivity and uncertainty analyses, an evaluation of modeled barrier performance, a 

description of the NRC staff’s sensitivity analyses, and a comparison of the dose results to the 

performance objectives. 
175 

2.13.3.1 Consideration of DOE’s Uncertainty and Sensitivity Analyses 

Use of a set of deterministic cases, including deterministic sensitivity analyses, supported by a 

probabilistic uncertainty model, which DOE refers to as a “hybrid” approach, could provide 

useful complementary information about the uncertainty in predicted doses and the sensitivity of 

predicted dose to various input parameters. However, the utility of a model depends on its 

scope and implementation. In DOE’s 2009 SDF PA analysis, near-field flow was outside of the 

scope of the probabilistic model and was, instead, directly incorporated from deterministic 

PORFLOWTM results. As a result, the sensitivity analysis conducted with the probabilistic model 

excluded several of the near field parameters that often have a significant impact on dose (e.g., 

hydraulic conductivity in saltstone or disposal unit concrete, diffusivity in saltstone). 

The sensitivity of the probabilistic (i.e., GoldSim® model) Case A and Case E to parameters 

affecting the dose from Ra-226 (i.e., Ra sorption coefficients and fraction of local vegetable 

uptake) is consistent with the importance of Ra-226 to the peak dose in these cases. However, 

the importance of Ra to dose in these cases, like all cases included in the original PA, is likely to 

be overstated because of potentially conservative assumptions about the inventory of Ra-226 

and its ancestors that were used in these cases but revised in Case K. Thus these results are 

of interest primarily in that they indicate that the uncertainty in inventory may change the 

projected dose driver and associated parameters to which dose is most sensitive. 

Beyond the way deterministic near-field flow results were used, the NRC staff has concerns 

about other aspects of the approach taken in the probabilistic model. For example, DOE did not 

provide model support (e.g., documented expert elicitation) for the probabilities assigned to 

each of the sensitivity cases considered in the probabilistic analysis (NRC, 2010i; PA-11). In 

addition, in the probabilistic model, DOE modifies the aquifer concentrations by the probability of 

well completion in each aquifer. Although this approach is appropriate if DOE uses the 

probabilistic model solely for risk information, as discussed in the context of the review for the 

FTF (NRC, 2011m), the NRC staff believes aquifer concentrations should not be modified by the 

probability of well completion if the dose results are compared to a performance objective. For 

comparison to a performance objective, it is most appropriate to locate the point of maximum 

exposure in the highest-concentration aquifer unit that could support the groundwater 

dependent pathways evaluated in the biosphere modeling. 

In addition to concerns about the approach taken in the probabilistic model, the NRC staff has 

specific concerns about the implementation of the DOE probabilistic analysis (Section 2.11). 

These concerns include, but are not limited to, the exclusion of key features of Case C when the 

case selection is run probabilistically, the modeled flows in the unsaturated zone, the area 

assumed for Vault 4, and the methodology used to estimate the ingrowth of radionuclides. For 

example, the Case C sensitivity to Pu sorption in the environment may be due to a modeling 

artifact that overestimates the long-term ingrowth of Pu-239 (Section 2.11.4.3). Because of 

these concerns with the approach to the probabilistic modeling and the specific implementation 

of the probabilistic model, the NRC staff did not rely on DOE’s probabilistic modeling results in 

its compliance evaluation. 
176 

Deterministic PORFLOWTM Cases B through E, which DOE refers to as alternate disposal unit 

configuration cases, essentially vary assumptions about flow through saltstone and the disposal 

units. Whereas Cases B, C, and D vary assumptions about flow through potential fractures in 

the saltstone and disposal units, Case E is used to evaluate the effects of elevated flow through 

the saltstone matrix (i.e., increased hydraulic conductivity of the bulk saltstone rather than 

discrete fractures). Of these cases, the case with the largest projected peak dose (Tables 2-13 

and 2-14) is Case E (i.e., peak dose within 10,000 years of 0.56 mSv/yr [56 mrem/yr]), which 

allows the most flow through the saltstone matrix. Cases B and C, which each allow flow 

through fractures in the saltstone and the disposal unit floors, have more modestly increased 

doses compared to Case A, which represents saltsone as an uncracked monolith for the entire 

performance period. That is, the predicted peak dose within 10,000 years from Case B is 

0.039 mSv/yr [3.9 mrem/yr] and the predicted peak dose from Case C in 10,000 years is 

0.051 mSv/yr [5.1 mrem/yr]. Case D, which DOE based on the assumption that the sheet drain 

between the saltstone and disposal unit wall in Vault 4 and the FDCs would act as a capillary 

break (i.e., a barrier to flow), predicted a slightly lower peak dose within 10,000 years than Case 

A (i.e., 0.013 mSv/yr [1.3 mrem/yr]) (Tables 2-13 and 2-14). While the absolute values of the 

results of these case are difficult to interpret because they are affected by overly optimistic 

assumptions (e.g., unrealistic moisture characteristic curves (NRC, 2010b; SP-3 and SP-4), 

limited saltstone fracturing (NRC, 2010b; SP-1), and exceptional performance of the lower 

lateral drainage layer for thousands of years (NRC, 2010b)), the relative magnitude of the 

results indicates the importance of flow through the saltstone matrix to the predicted dose. 

Comparison of the results of Case E and the Increased Hydraulic Conductivity Case (2009 PA; 

Section 5.6.6.7) with the Case A results helps to illustrate the effects of flow through the 

saltstone matrix. Both Case E and the Increased Hydraulic Conductivity Case represent 

increased hydraulic conductivity of the saltstone matrix with all other aspects of the disposal 

facility performing as designed. Although saltstone in Case E has a nominal saturated hydraulic 

conductivity of 1.7x10-3 cm/s, the flow is actually limited by infiltration to a flow of approximately 

1x10-6 cm/s. The Increased Hydraulic Conductivity Case is essentially the same as Cases A 

and E, except that the saturated hydraulic conductivity of saltstone is assumed to be 

1x10-7 cm/s. In all three cases, the saturated hydraulic conductivity is assumed to remain 

constant during the analysis period (i.e., 20,000 years). These differences in saltstone hydraulic 

conductivity result in an increase in the peak dose within 10,000 years of closure from 

0.014 mSv/yr (1.4 mrem/yr) in Case A (Kh = 2x10-9 cm/s) to 0.054 mSv/yr (5.4 mrem/yr) in the 

increased hydraulic conductivity case (Kh = 1x10-7 cm/s) and 0.56 mSv/yr (56 mrem/yr) in Case 

E (flow limited by infiltration at an effective Kh of 1x10-6 cm/s). 

DOE used the Sulfate X10 Case and the No Sulfate Attack case to explore the effects of 

different degrees of hydraulic degradation of the disposal unit roofs, floors, and walls. The 

primary effect of increased sulfate attack was to shorten the time in which the disposal unit 

concrete becomes hydraulically equivalent to the backfill soil. In the 10X sulfate attack case, 

the predicted failure time for the FDC roof, floor, and walls and the predicted failure time for the 

Vault 4 walls moved from beyond 10,000 years to within 10,000 years after closure (i.e., from 

40,000 years to 5,000 years for the FDC floor, from 40,000 years to 7,000 years for the FDC 
177 

roof, from 18,000 years to 3,000 years for the FDC wall, and from 10,000 to 3,000 years for the 

Vault 4 roof). 

Predicted doses from the No Sulfate Attack and 10X Sulfate Attack cases are greatest in Sector 

B (Table 2.14). As in Case A, peak doses within 10,000 years were due primarily to Ra-226 

(85%) and I-129 (9.7%), although, unlike in Case A, the peak from the 10X Sulfate Attack case 

within 10,000 years of closure did have a non-negligible contribution from Tc-99 (3.2%). In 

further contrast to Case A, peak doses within 20,000 years in the 10X sulfate attack case have 

a significant contribution from Tc-99 (70%) with the remaining dose primarily attributable to 

Ra-226 (24%) and I-129 (4%) (SRR-CWDA-2010-00033; Table PA-1.3). DOE’s predicted peak 

doses from the 10X Sulfate Attack Case within 10,000 and 20,000 years of site closure are only 

slightly greater than the corresponding peaks in Case A (i.e., 21% greater and 55% greater, 

respectively). This result indicates that the dose is not very sensitive to the hydraulic 

performance of the disposal units when the remainder of the system performs as designed. 

Hydraulic performance of the disposal units may have a greater effect on dose if flow is not 

limited by the hydraulic performance of the cap and saltstone waste form. 

To test the effects of simultaneous degradation of multiple barriers, DOE developed the 

Synergistic Case (Table 2.1). Key features include early hydraulic degradation of the cap and 

disposal unit concrete, early oxidation of the Vault 1 and 4 walls and FDC concrete, and 

fractures through saltstone. Saltstone fractures are assumed to occur at approximately 1.5 m 

(5.0 ft) intervals26 at the time of closure and not to increase during the performance period. The 

fractures allow water flow through saltstone. The fractures also allow saltstone oxidation to 

proceed from the fracture faces. In the PA, DOE reported a predicted peak dose of 0.18 mSv/yr 

(18 mrem/yr) within 10,000 years of closure and a predicted peak of 0.28 mSv/yr (28 mrem/yr) 

within 20,000 years. In the second RAI response (SRR-CWDA-2011-00044 Figure PA-9.1), 

DOE reported different doses for the unmodified synergistic case: approximately 0.23 mSv/yr 

(23 mrem/yr) within 10,000 years and approximately 0.36 mSv/yr (36 mrem/yr) within 20,000 

years. As in Case A, peak doses within either 10,000 years or 20,000 years of closure are 

predicted to be due primarily to Ra-226 with a small contribution from I-129 

(SRR-CWDA-2010-00033; Table PA-1.3). 

In response to the NRC RAIs (NRC, 2010b, i), DOE provided additional sensitivity analyses 

based on the Synergistic Case (Table 2.1). In response to NRC concerns about the effects of 

unrealistically optimistic moisture characteristic curves, DOE recalculated the fluxes of Ra-226, 

I-129, and Tc-99 to the unsaturated zone using an assumed relative permeability of one in 

saltstone (SRR-CWDA-2010-00033, comment VP-2). The recalculation indicated that the 

Synergistic Case results were not sensitive to the difference between the moisture characteristic 

curves and a fixed relative permeability of one in saltstone. In response to an NRC concern 

about the limited set of radionuclides analyzed in the Synergistic Case (i.e., an initial inventory 

assumed for I-129, Np-237, Pu-238, Tc-99, Th-230, U-234, and U-238), DOE re-ran the 

Synergistic Case PORFLOWTM model with thirteen additional radionuclides in the initial 

 

26 Fracture spacing is described as 1.5 m (5 ft) intervals in the Review Team Report on the Performance Assessment 

for the Saltstone Disposal Facility at the Savannah River Site. However, fractures appear to occur at 0.75 m (2.5 ft) 

interval in the PORFLOWTM model file. 
178 

inventory (i.e., Am-241, Am-243, Cm-244, Cm- 245, Cs-135, Nb-93m, Pu-239, Pu-240, Pu-241, 

Pu-244, Th-229, U-233, and U-235) (SRR-CWDA-2011-00044, PA-9). The results of this 

analysis showed the additional radionuclides did not contribute significantly to dose in either 

Sector B, Sector I, or the adjacent sectors. In addition, for consistency with RAI responses 

regarding biosphere parameters, DOE also ran the Synergistic Case (including the additional 

radionuclides) with updated biosphere parameters (SRR-CWDA-2011-00044, PA-9). This 

modification resulted in slightly lower predicted peak doses within 10,000 and 20,000 years of 

closure, largely due to a reduction of the water-to-fish transfer factor for Ra-226 and an increase 

in the garden productivity factor. 

The additional sensitivity cases based on the Synergistic Case (i.e., saltstone relative 

permeability of one, additional radionuclides, and updated biosphere parameters) aid in the 

interpretation of the Synergistic Case results. However, additional factors that could lead to 

underestimates of radionuclide release make it difficult to determine if the Synergistic Case is 

likely to over- or under-predict the peak dose. For example, the assumption that the hydraulic 

conductivity of the unfractured saltstone remains constant at the value used in DOE’s Case A 

(2x10-9 cm/s) appears to be unrealistically optimistic in light of recent experimental results. 

These experimental results show hydraulic conductivity values in both laboratory samples 

prepared under conditions relevant to saltstone as well as core samples taken from emplaced 

saltstone (Section 2.6). Furthermore, even if a hydraulic conductivity of 2x10-9 cm/s were 

achieved in field-emplaced saltstone, it does not appear to be realistic that the hydraulic 

conductivity of the saltstone matrix would not degrade within 10,000 years. Although fracturing 

does reflect one degradation mode of the saltstone waste form, the result of the modeled low 

hydraulic conductivity of the saltstone matrix is to route flow almost entirely through the 

fractures. In addition, the selection of hydraulic properties similar to soil for the disposal unit 

walls, while it appears to be conservative, serves to route additional flow away from the 

saltstone (e.g., the drop in flow through the saltstone at 500 years is attributable to hydraulic 

degradation of the disposal unit walls) (Figures 2.13-1 and 2.13-2). As a result of the assumed 

low hydraulic conductivity of the saltstone and high hydraulic conductivity of the fractures and 

walls, even less water flows through the saltstone matrix in Vault 4 in the Synergistic Case than 

flows through the saltstone matrix in Case A (Figure 2.13-2), while the flows are approximately 

equal in the FDCs (Figure 2.13-3). 

Flow through the matrix is expected to be important to the predicted dose because flow through 

the matrix exposes significantly more radionuclide inventory to infiltrating water than fracture 

flow. The sensitivity of the predicted peak dose from the SDF to the hydraulic conductivity of 

the saltstone matrix is shown clearly by the results of Case E and the Increased Hydraulic 

Conductivity Case, as previously discussed in this section. In particular, the Increased 

Hydraulic Conductivity Case, which used a saturated hydraulic conductivity in saltstone similar 

to values measured for field saltstone samples (1x10-7 cm/s), has a predicted peak dose within 

10,000 years approximately 4 times greater than the predicted peak dose in Case A. 

The NRC staff understands that sensitivity analyses are designed to test the response of the 

system to various changes in input, and do not necessarily yield peak doses that are 

appropriate to compare to compliance limits. However, if major degradation modes are 
179 

considered and if dose predictions do not exceed the relevant limits, alternate cases based on 

enhanced degradation of multiple barriers can provide added confidence in a compliance 

determination. In this case, the Synergistic Case results approach, but do not exceed the 

compliance limits. However, because of the apparent sensitivity of dose to the saturated 

hydraulic conductivity of the saltstone matrix, it appears that a Synergistic Case run with a more 

realistic value of this parameter would yield peak dose predictions that exceed the relevant dose 

limits. In addition, more realistic values of the sorption coefficient for Tc in reducing and 

oxidizing saltstone and disposal unit concrete are likely to further increase the predicted release 

of Tc (Section 2.7). Thus, because of the flow-limiting effects of the low hydraulic conductivity in 

the saltstone matrix, the flow-diverting effects of the disposal unit walls, and the Tc releaselimiting effects of the high sorption coefficients chosen for Tc, the NRC staff cannot conclude 

that the results of the Synergistic Case necessarily bound the predicted dose from the SDF. 

Figure 2.13-2 Vault 4 vertical Darcy velocity through the saltstone matrix (excludes 

fractures) predicted with the DOE PORFLOWTM model for Case A, the Synergistic Case, 

and the Increased Hydraulic Conductivity Case (Data taken from the STAT.out files 

[NRC, 2010g]). 

The NRC staff concludes that the types of deterministic sensitivity analyses performed by DOE, 

including those sensitivity analyses performed in response to NRC’s RAIs, were reasonable in 

the technical areas they addressed, but were too limited in scope. In general, deterministic 

sensitivity analyses have a limited capacity to evaluate potentially unexpected effects from 

parameters that are not explicitly varied in the deterministic test. For example, there is 

uncertainty in the potential dose from Se-79 due to concerns about the sorption coefficients 

used for Se in cementitious materials and site soil (Section 2.7). Because the potential effect of 

this uncertainty has not been captured in the current sensitivity analysis, it has been identified 

as an issue to be monitored (Table A-1). These types of effects often may be captured with a 

more comprehensive probabilistic uncertainty analysis, such as the GoldSim® analysis 

1.0E-04

1.0E-03

1.0E-02

1.0E-01

1.0E+00

0 5000 10000 15000 20000

Vertical Darcy Velocity (cm/yr)

Time Since SDF Closure (years)

Increased Hydraulic Conductivity Case

Case A

Synergistic Case
180 

submitted by DOE, if correctly implemented. As previously discussed, the NRC staff was 

unable to rely on DOE’s probabilistic GoldSim® analysis as part of its compliance demonstration 

(Sections 2.11.4.2 and 2.11.4.3). 

Figure 2.13-3 FDC vertical Darcy velocity through the saltstone matrix (excludes 

fractures) predicted with the DOE PORFLOWTM model for Case A, the Synergistic Case, 

and the Increased Hydraulic Conductivity Case (data taken from the STAT.out files 

[NRC, 2010g]). 

Finally, both deterministic and probabilistic models may fail to capture conceptual model 

uncertainty if it is not specifically considered and incorporated into an analytical model. During 

the SDF review, the NRC staff identified possible alternative conceptual models in which 

significant oxidation of saltstone occurred prior to significant release of water from the disposal 

units. For example, if oxygen-bearing water infiltrates into the FDCs but is retained by the 

HDPE layer on the outside of these units, a pulsed release of Tc could occur if the HDPE 

experienced a period of rapid deterioration. Similarly, if Tc is oxidized by the infiltration of 

oxygen-bearing water from the top of a disposal unit and is re-reduced and concentrated as it 

moves downward (e.g., in the manner of a uranium roll-front), oxidation of the bottom layer of 

the disposal unit could cause a release greater than the releases considered in current 

conceptual model. Because of the potentially large doses that could be predicted with this 

conceptual model, consideration of this possibility has been identified as a PA maintenance item 

(Table A-2). 

1.0E-04

1.0E-03

1.0E-02

1.0E-01

1.0E+00

0 5000 10000 15000 20000

Vertical Darcy Velocity (cm/s)

Time Since SDF Closure (years)

Increased Hydrauilc Conductivity Case

Case A

Synergistic Case
181 

2.13.3.2 Barrier Contribution to Dose Reduction 

PA Cases

DOE’s initial sensitivity analysis demonstrates that the disposal system is designed with multiple 

barriers to radionuclide release. The primary barriers either act to limit water flow through the 

waste form or to control the water chemistry to slow radionuclide release from the waste form or 

through the disposal unit floor and walls. By design, primary barriers to water flow are (1) the 

closure cap and the engineered layers above each disposal unit (including the lower lateral 

drainage layer), (2) disposal unit roof, floor, and walls and (3) the intended low permeability of 

the saltstone waste form. DOE’s initial sensitivity analyses tested the effects of (1) the upper 

layers of the closure cap, (2) permeability of the disposal unit roof, floors, and walls, 

(3) permeability of the saltstone, and (4) the chemical condition of Vault 1 and 4 walls and floor. 

Several of the initial sensitivity analyses demonstrate that the degradation of a single barrier has 

little effect on SDF performance because of the designed performance of other barriers. For 

example, DOE’s “No Cap” sensitivity analysis tested the performance of the SDF if the upper 

layers of the closure cap (i.e., the composite hydraulic barrier, the upper lateral drainage layer27, 

and the erosion control layer) did not limit flow, resulting in infiltration of 41.8 cm/yr (16.4 in/yr) 

through these layers. The scenario also assumes that the engineered layers on top of the 

disposal units (i.e., the geotextile filter fabric and lower lateral drainage layer for Vaults 1 and 4, 

and the geotextile filter fabric, lower lateral drainage layer, HDPE geomembrane and 

geosynthetic clay liner for the FDCs) the disposal unit concrete, and the saltstone use Case A 

hydraulic performance. DOE’s analysis indicates that there is little effect of removing the credit 

for the upper layers of the closure cap if the engineered layers above each disposal unit and 

other features of the facility perform as designed (Table 2-13 and 2-14). 

The NRC staff evaluation of DOE PORFLOWTM intermediate outputs indicates the significant 

barrier to flow DOE expects to achieve with the combination of a high hydraulic conductivity 

lower lateral drainage layer and low hydraulic conductivity disposal unit roof (for Vaults 1 and 4) 

or low hydraulic conductivity HDPE layer, geosynthetic clay liner, and roof (for the FDCs). This 

contrast of low and high hydraulic conductivity layers is intended to divert water around the 

disposal units. For example, in Case A, 99.9% of the water infiltrating through the cap is 

diverted around the Vault 4 and FDC roofs up to 10,000 years after SDF closure 

(Figure 2.13-4). In the “no cap” scenario, 8,000 years after closure these layers divert 99.7% of 

the water around an FDC, while 99.98% is diverted around Vault 4. Although there is a 

significant difference in the infiltration through the upper cap layers in Case A and “no cap” 

analyses, there is not a significant difference in the predicted peak dose from each scenario 

(i.e., 0.14 mSv/yr [1.4 mrem/yr] in Case A and 0.021 mSv/yr [2.1 mrem/yr] in the “no cap” 

scenario within 10,000 years and 0.031 mSv/yr [3.1 mrem/yr] in Case A as compared to 

0.036 mSv/yr [3.6 mrem/yr] in the “no cap” case within 20,000 years). This dose result, in 

combination with the intermediate model results illustrating water diversion in the lower lateral 

 

27 Although the 2009 PA (p. 541) indicates that the “no cap” scenario models performance as if neither of the 

drainage layers is present, the DOE PORFLOWTM file used to model the scenario includes credit for the lower lateral 

drainage layer. 
182 

drainage layer, demonstrates the importance of the large contrast between the modeled 

hydraulic conductivity of the lower lateral drainage layer and the layers it overlies above each 

disposal unit (i.e., the roof for Vault 1 and 4, the HDPE, geosynthetic clay liner, and roof for the 

FDCs). 

Figure 2.13-4: Water shedding around Vault 4 in DOE Case A at 0 and 8,000 years after 

site closure (left top and bottom) as compared to an NRC interpretation of know 

conditions (top right) or potential future conditions considering uncertainty (bottom 

right) 

As discussed in the previous section, DOE developed the Synergistic Case to evaluate the 

impact of simultaneously changing multiple material parameters to account for several potential 

increased degradation mechanisms as compared to Case A. The PA describes this case as 

overly pessimistic. However, in the Synergistic Case, although the saltstone is modeled as 

fractured, the hydraulic conductivity of the saltstone matrix is the same as it is in Case A (i.e., a 

static value of 2x10-9 cm/s). Water moves primarily through fractures in the saltstone, and the 

flow through the saltstone matrix is very similar to the flow through the saltstone matrix in the 

Case A (Figures 2.13-2 and 2.13-4). Comparison with Case E and the Increased Hydraulic 

Conductivity Case suggests that a case with higher saltstone matrix hydraulic conductivity 

would predict a larger dose, presumably because water flowing through saltstone would come 

into contact with more of the saltstone inventory of radionuclides. This comparison is of interest 

particularly because DOE results suggest that the hydraulic conductivity of field-emplaced 

saltstone may be considerably greater than DOE modeled in the Synergistic Case (Section 2.6). 

In addition to these tests of various hydraulic barriers, DOE used the Oxidized Concrete Case to 

evaluate the effects of earlier oxidation of Vault 1 and 4 concrete. Specifically, the Oxidized 

Concrete case is based on the assumption that the Vault 1 and 4 floor and walls are completely 

oxidized at closure. The FDC floor and walls are modeled just as they are in Case A. In the 

Oxidized Concrete Case, the peak doses in Sectors B and I within either 10,000 or 20,000 years 
183 

of site closure decrease slightly (Tables 2-13 and 2-14). This decrease does not completely 

clarify the importance of the Vault 1 and 4 floor and walls as a chemical barrier, because it is 

attributable to a somewhat artificial difference in the Kd values for Ra in oxidizing and reducing 

saltstone (Section 2.7). The more significant conclusion from the Oxidized Concrete case in the 

context of the barrier analysis may be that complete oxidation of the Vault 1 and 4 walls is 

predicted to generate a small dose from Tc-99 at approximately 900 years after SDF closure in 

Sector B due to the assumed initial inventory in the walls (rather than the inventory in saltstone 

itself) (2009 PA; Figure 5.6-84). While the initial inventory in the Vault 1 and 4 walls is based on 

the apparently conservative assumption that the pore spaces in the walls are completely 

saturated with pore fluid containing the same concentrations of radionuclides as in saltstone, the 

timing of the peak is of interest. Because the transport in the saturated zone is expected to take 

only tens of years, the peak at 900 years is expected to be due primarily to the time needed for 

release from Tc from the oxidized disposal unit floor and walls (Kd of 0.8 mL/g) under Case A 

flow conditions during the first 1,000 years following closure (i.e., with significant cap 

performance). 

Figure 2.13-5: Modeled volumetric flow per unit thickness through Vault 4 from 5,500 to 

6,000 years after SFD closure predicted by the DOE PORFLOWTM model for Case A, 

Cases K/K1/K2, and select DOE sensitivity analyses (data taken from the STAT.out files 

[NRC, 2010g]). 

1E+01

1E+02

1E+03

1E+04

1E+05

1E+06

Sand Drain Roof Saltstone Wall Floor

Volumetric Flowrate per unit thickness 

(cm2/yr) 

Case A

Case K/K1/K2

Case E

Increased Hydraulic 

Conductivity Case
184 

Figure 2.13-6: Modeled volumetric flow per unit thickness through Vault 4 from 5,500 to 

6,000 years after SFD closure predicted by the DOE PORFLOWTM model for the 

synergistic case (data from Case A repeated from Figure 2.13-3 for comparison) (data 

taken from the STAT.out files [NRC, 2010g]). 

After release from the disposal units, the final dose-reducing barrier is dilution and dispersion in 

the saturated zone. As described in Section 2.8, the modeled far-field performance potentially 

overestimates dilution and dispersion and may significantly limit estimated cumulative impacts 

from the facility. In the southern portion of the SDF, where the peak dose occurs in Case A and 

in most of the sensitivity analyses discussed in this section, most plumes tend to spread away 

from one another both horizontally and vertically, limiting the cumulative effect of multiple 

sources. Changes to key parameters and improved local SDF/PORFLOWTM model calibration 

could result in significantly lower levels of modeled dilution and dispersion, which could result in 

larger predicted doses. For these reasons, the NRC staff concludes it is more likely that 

predicted doses would be increased, rather than reduced, by changes to the SDF far-field 

modeling. 

Cases K, K1, and K2

Because of very different assumptions about saltstone degradation and oxidation and Tc 

sorption, Cases K, K1, and K2 emphasize different barriers than the cases initially evaluated in 

the DOE PA. 

As previously discussed (Section 2.13.3.1), the magnitude and timing of the Case K peak dose 

is sensitive to assumptions about saltstone fracturing. In addition, the magnitude and timing of 

Tc release also are sensitive to assumptions about disposal unit performance. DOE’s Cases K, 

K1, and K2, display an unexpected behavior of reconcentration of Tc in the disposal unit floor 

and walls. By approximately 13,000 years after site closure over 90% of the Tc originally in the 

Saltstone 

Matrix

Synergist Case 

Saltstone 

Fractures

1E+01

1E+02

1E+03

1E+04

1E+05

1E+06

Sand Drain Roof Saltstone Wall Floor

Volumetric Flowrate per unit thickness 

(cm2/yr)

Case A

Synergistic Case
185 

saltstone has been released from the saltstone and reconcentrated in the disposal unit concrete 

(Figure 2.13-4). Because the disposal unit floor and walls have a much lower volume than the 

saltstone, this process, as modeled, would result in a 13-fold increase in the original Tc 

concentration in saltstone. This unexpected behavior occurs because the disposal unit floors 

and walls are modeled with much higher Kd values than saltstone as saltstone becomes more 

oxidized (e.g., at 10,000 years after closure, saltstone is modeled with a Kd of 0.8 mL/g as 

compared to the FDC floor at 406 mL/g, FDC walls at 388 mL/g, Vault 4 floor at 388 mL/g and 

Vault 4 walls at 228 mL/g) (Section 2.7). The NRC staff does not believe this modeled behavior 

is realistic because (1) the disposal unit concrete is expected to have about 40% of the specific 

reducing capacity of saltstone because of its lower blast furnace slag concentration; (2) the 

disposal unit floors and walls are more exposed to environmental conditions than saltstone is 

and would be expected to fracture within 10,000 years; (3) water flowing through fractures in the 

disposal unit floors and walls would be expected to create oxidized conduits in which Tc is not 

expected to be re-reduced and immobilized after release from the saltstone, and (4) water may 

flow out of the disposal unit through joints and consequently may not interact significantly with 

the disposal unit concrete. 

Figure 2.13-7 Intermediate results from DOE PORFLOWTM model for Case K, showing Tc 

release from Saltstone and re-concentration in the FDC floor and walls. At approximately 

13,000 years after closure, over 90% of the Tc inventory originally in saltstone is present 

in the FDC floor and walls. 

The two primary effects of this reconcentration in the disposal unit floor and walls are to delay 

the peak dose by approximately 5,000 years and to significantly reduce the magnitude of the 

peak dose. Peak doses are reduced because the peak dose is very sensitive to the rate at 

which Tc is released into the environment, and the modeled Tc release from the disposal units 

is much slower than the modeled Tc release from the saltstone. 

0.0

0.1

0.2

0.3

0.4

0.5

0.6

0.7

0.8

0.9

1.0

0 5000 10000 15000 20000

Tc-99 Fractional Inventory

Year

Saltstone

Disposal Unit
186 

Cases K, K1, and K2 were run with the same assumptions about far-field flow used in Case A 

and the deterministic sensitivity analyses discussed in this section. The maximum aquifer 

concentrations for Cases K, K1, and K2 occur in Sector I compared to Sector B for Case A and 

most of the other sensitivity cases evaluated in the PA. FDC sources in the northern portion of 

the SDF that contribute most to Sector I concentrations appear to be significantly affected by the 

groundwater divide that tends to spread plumes laterally in the Upper Three Runs aquifer, 

thereby limiting plume overlap (Figure 2.8-4). Additionally, the strong vertical gradient that 

appears to exist at the SDF also tends to spread contaminant plumes vertically leading to 

enhanced vertical dispersion and potential loss of contaminant mass to the Gordon Confining 

Unit. Thus, like the cases originally evaluated in the PA, dilution and dispersion in the saturated 

zone appear to be significant to performance for DOE’s Cases K, K1, and K2. The NRC staff 

expects that it is more likely that predicted doses would be increased, rather than decreased, by 

further adjustment of SDF far-field modeling. 

2.13.3.3 NRC Independent Sensitivity Analyses 

As previously discussed, Case K, K1, and K2 resolve many of the concerns the NRC staff has 

with the use of Case A as a case to compare to the performance objectives. In response to a 

DOE request to confirm that development of Case K was an acceptable approach to NRC’s 

request for a revised base case, the NRC staff agreed that development of Case K appeared to 

be an acceptable approach (NRC, 2011c). However, as discussed in Section 1.2, because of 

the complexity of the system being modeled, the NRC staff also indicated to DOE that it could 

not make any specific conclusions about the assumptions used in Case K until the staff 

reviewed DOE’s written responses to RAI-2009-02 (NRC, 2011c). Upon review of Cases K, K1, 

and K2, the NRC staff concluded that, while some of the assumptions DOE used appear to be 

conservative or realistic, others appear to be overly optimistic. Because of the combination of 

assumptions used, the NRC staff is unable to conclude whether these cases are likely to overor under-predict the potential dose from the SDF. Furthermore, unlike Cases A - E, Case K1 

was supplied to NRC with a very limited sensitivity analysis (i.e., the only sensitivity analysis 

provided was a comparison of the results of Cases K, K1, and K2, which use different Kd values 

to represent Tc sorption in saltstone). As NRC indicated to DOE during technical exchanges 

regarding the development of Case K, if Case K is to be a significant part of DOE’s compliance 

demonstration, NRC needs to understand the effects of parameter uncertainties on Case K 

results (NRC, 2011e). Thus, as described in this section, the NRC staff developed sensitivity 

analyses to better understand the potential dose from two radionuclides: (1) Ra-226, which 

dominates the dose from Case A and many of DOE’s deterministic sensitivity analyses, and 

(2) Tc-99, which dominates the predicted dose in Cases K, K1, and K2. 

Analyses related to Ra-226

In many of the deterministic analyses performed by DOE, the dose from Ra-226 contributed a 

significant portion of the peak dose (SRR-CWDA-2010-00033; Tables PA-1.1 to 1.4). Most of 

the Ra-226 dose is due to the ingrowth of Ra-226 from its parent Th-230. As noted in 

Section 2.2, the inventory of Th-230 is uncertain because it is below the analytical detection limit 

in tank waste samples taken to date (SRR-CWDA-2011-00115). In the Case K analysis, DOE 
187 

stated that it revised the inventories of Th-230 and Ra-226 to much lower values to remove 

conservative assumptions used in the initial inventory estimate. The Case K analysis also 

included changes to a number of parameters (Section 2.1), including the Ra Kd values assumed 

for reducing cementitious materials (i.e., saltstone and disposal unit concrete) and for sand and 

clay in the subsurface. As described in Section 2.7, the NRC staff believes that more 

justification is needed for the Ra Kd values assumed for the saltstone and disposal unit concrete 

in Case K. Although Ra-226 is not a primary contributor to peak dose in Cases K, K1, or K2, 

the importance of Ra-226 to the final dose depends on uncertain assumptions about the 

inventory of its ancestors, including Th-230 and U-234, in saltstone (Section 2.2). As noted in 

Section 2.7, the NRC staff believes that the revised Kd values for the sand and clay are 

appropriate because they are based on site-specific measurements. 

To understand the potential dose from Ra-226, the NRC staff used DOE’s PORFLOWTM models 

to evaluate the dose from Ra-226 due to in-growth from Th-230 for four different cases: Case A, 

Case K, Case K with edited saltstone and disposal unit concrete Kd values for Ra-226, and 

Case K with edited saltstone, disposal unit concrete, clay, and sand Kd values for Ra-226. This 

assessment was performed for Vault 4 because the Vault 4 Th-230 inventory is expected to be 

much higher than the inventory in an FDC (Table 2.13-3). The version of PORFLOWTM used by 

the NRC staff (PORFLOWTM Version 6.12.3) in performing this assessment was different than 

the version used by DOE for Cases K, K1, and K2 (PORFLOWTM Version 6.30.2), which may 

have resulted in differences in the calculated dose due to differences in the way PORFLOWTM 

implements dispersivity in these two versions. Based on a comparison of DOE’s results with the 

results of equivalent cases run in NRC’s version of PORFLOW, these differences in the 

representation of dispersivity in the two versions of PORFLOWTM are expected to cause a 

difference of less than a factor of two in the dose results.. The results of these analyses 

indicate the large difference in projected dose that can be attributed to the difference in 

inventory estimates as compared to the difference in assumed Kd values (Table 2.13-3). Peak 

dose within 20,000 years of site closure are reported because, as noted in Section 2.7, the NRC 

staff believes that there is significant uncertainty in the timing of the peak dose due to 

uncertainty in the timing of the degradation of the system. Specifically, the NRC believes that 

the peak dose for Ra-226 that DOE predicts will occur within 20,000 years of site closure in 

Case K may occur within 10,000 years of SDF closure because of the uncertainty in the timing 

of increases in saltstone hydraulic conductivity. In Case K, saltstone hydraulic conductivity is 

based on a DOE assumption that most of saltstone occurs approximately 9,000 years, which 

results in a peak dose beyond 10,000 years. DOE did not provide a basis for this assumption 

about the timing of saltstone fracturing. NRC staff analyses related to the peak dose from 

Tc-99, discussed later in this section, indicate that different fracturing models could lead to 

earlier peak releases. As a result, the NRC staff believes it is possible that the peak dose from 

Ra-226 may occur within 10,000 years of SDF closure. 

In all cases where the revised Th-230 inventory was assumed, the estimated dose was 

significantly less than 0.01 mSv/yr (1 mrem/yr). Thus, if the revised inventory of Th-230 is 

accurate, the Ra-226 dose is likely to be small regardless of the other parameter values 

assumed. The dose in the cases in which the original Th-230 inventory was used was higher, 

but the only one of these cases that had a peak dose greater than 0.25 mSv/yr (25 mrem/yr) 
188 

was the case in which both the cementitious Kd values and the subsurface Kd values were 

changed to the original Case A values. As noted in Section 2.7, the NRC staff believes that 

DOE has provided a basis for use of the site-specific clay and sand Kd values used in Case K, 

so the NRC staff does not believe that this case is applicable. The doses estimated using the 

site-specific subsurface Kd values and the Case A or Case K cementitious material Kd values 

were all less than 0.1 mSv/yr (10 mrem/yr). The NRC staff determined that, with the Kd values 

assumed in Case A or Case K, the Ra-226 peak is not likely to exceed 0.25 mSv/yr 

(25 mrem/yr). However, because the peak from Ra-226 could overlap with the peak dose from 

other radionuclides, the Ra-226 dose could still contribute significantly to the total peak dose 

from the SDF and impact the determination of whether the total dose meets the limit in the 

performance objective (i.e., 0.25 mSv/yr [25 mrem/yr]). Additionally, DOE has not presented 

site-specific information on the cementitious material Ra-226 Kd values, which was shown to 

significantly affect the predicted peak Ra-226 dose. The NRC staff concludes that site-specific 

information on the cementitious material Ra-226 Kd values is needed. Therefore, the NRC staff 

will monitor the development of site specific Kd values for Ra-226 in saltstone and the disposal 

unit concrete. 

Table 2.13-3: Estimated Peak Vault 4 Dose from Ra-226 that Ingrows from Th-230 

Vault 4 Dose to

Source Ratio 

(mrem/yr Ra-226)/

(initial Ci of 

Th-230)1

Dose Based on 

Original Th-230 

Inventory (7.5 Ci) 

(mrem/yr) 

Dose Based on 

Revised Th-230 

Inventory (0.01 

Ci) (mrem/yr) 

Case A 3 < 0.29 <2.2 < 2.9x10-3

Case K (reducing cementitious Kd value 

of 100 mL/g 3

, clay Kd of 185 mL/g, and 

sand Kd of 25 mL/g) 

0.75 5.6 7.5x10-3

Case K with a changed reducing 

cementitious Kd value of 3 mL/g (3) 1.2 9.2 1.2x10-2

Case K with a changed reducing 

cementitious Kd value of 3 mL/g (3), and 

with a changed clay Kd of 17 mL/g and 

sand Kd of 5 mL/g 

3.7 28 3.7x10-2

1

 Based on peak dose in 20,000 years. 

2 Dose to source ratio calculated based on Ra-226 dose reported in PA Figure 5.4-4. This dose includes 

contributions from FDCs and from Ra-226 from sources other than ingrowth from Tc-230 and therefore 

bounds the Case A estimated dose from Vault 4 from Ra-226 ingrown from Th-230. 

3 Only middle-age reducing conditions are provided because only middle-age conditions were predicted to 

occur during the evaluation period. The Kd values for young and old-age cementitious material changed as 

well, but were not used because these conditions did not occur. Case K uses Kd values of 100 mL/g for 

young- and middle-age reducing cementitious material and 70 mL/g for old-age reducing material. The Case 

K tests with changed reducing cementitious Kd values used 0.5 mL/g for young reducing cementitious 

material, 3 mL/g for middle-age reducing cementitious material, and 20 mL/g for old-age reducing 

cementitious material. 
189 

Analyses Related to Tc-99

NRC analyses related to Tc-99 are based on Case K1 because the NRC staff concludes the Kd

values for Tc in reducing and oxidizing grout are most realistic in this case (Section 2.7). 

However, because of apparently pessimistic assumptions about the rate of saltstone fracturing 

but unrealistically optimistic assumptions about Tc-99 retention in the disposal unit floors and 

walls (Figure 2.13-4); the NRC staff was unable to conclude whether Case K1 is likely to overor under-predict the peak dose from Tc-99. To clarify the potential dose from Tc-99 in a case 

with fractured Saltstone, the NRC staff performed two types of independent sensitivity analyses. 

First, the NRC staff used DOE’s Case K1 PORFLOWTM model to re-calculate Tc transport 

based on different assumptions about saltstone and disposal unit oxidation (but unchanged 

Case K1 flow velocities28). Second, the staff developed analytical calculations, implemented in 

Microsoft Excel®, to simulate (1) a simple average-Kd model and (2) a simple “dual-Kd” analytical 

model that simulates release of Tc from oxidized saltstone separately from release from 

reducing saltstone (NRC, 2012a, b). The analytical calculations were developed principally to 

compare results of the average-Kd model with the dual-Kd model, because this comparison 

could not be made within DOE’s PORFLOWTM model. In addition, the analytical calculations 

were used as scoping calculations to allow the effects of variations in certain model parameters 

to be evaluated quickly. 

NRC evaluated six cases with DOE’s PORFLOWTM model (Table 2.13-3). First, the NRC staff 

reran Case K1 to ensure the results matched DOE’s reported results. Then the NRC staff ran 

five sensitivity cases to evaluate the following factors: (Test 1) less sudden fracturing (i.e., 

fractures developing as a quadratic function of time rather than a logarithmic function of time); 

(Tests 2 and 3) the effects of the chemical retention of Tc in the disposal unit floors and walls; 

(Test 4) less fracturing occurring within 10,000 years of closure; and (Test 5) a lower Kd value 

for Tc in reducing saltstone. In DOE’s Case K1, the fracture rate only affects saltstone 

oxidation. Saltstone hydraulic properties degrade to soil properties at 10,000 years and are not 

explicitly based on assumptions about fracturing. Therefore, to test the importance of the 

fracture rate, the NRC staff re-evaluated the oxidation of saltstone proceeding from fractures 

with the same saltstone oxidation model and oxidation parameters used by DOE, with the 

exception that fractures were assumed to appear as a quadratic function of time rather than a 

logarithmic function of time. The NRC staff then re-calculated the weighted average Kd value 

that DOE used in Case K1 based on the same Kd-averaging process used by DOE but with the 

new projection of the amount of oxidized saltstone as a function of time. The primary output of 

the staff’s variations of DOE’s Case K1 model was the Tc flux into the unsaturated zone. Dose 

predictions were based on the ratio of Tc flux to the unsaturated zone to final dose in DOE Case 

K1. Use of this ratio is expected to provide a reasonable dose estimate in cases with slower 

oxidation (e.g., Tests 4 and 5 in Table 2.13-3). NRC scoping calculations indicate that the use 

of this ratio may over-predict dose by approximately a factor of 3 in cases with more rapid 

oxidation (e.g., Tests 2 and 3 in Table 2.13-3) because the flux-to-peak-dose ratio diminishes 

when the peak release rate is brief. 

 

28 Flow velocities used in DOE’s Cases K, K1, and K2 are identical. 
190 

Dose sensitivity to the fracture growth rate is expected to occur because, in this modeling 

approach, more Tc is released from new fractures than old fractures. Tc release from a fracture 

diminishes with time as the oxidized region near a fracture face grows and longer diffusion 

lengths into the matrix (for dissolved O2) and out of the matrix (for Tc release) slow release from 

older fractures. Thus the rate at which Tc release from a fracture slows with time is sensitive to 

assumptions about diffusivity, including the assumption that the diffusivity remains constant 

instead of increasing as saltstone degrades and the particular value of the diffusion constant. 

Although diffusion typically is modeled as proportional to the square root of time, in some cases 

diffusion in a cementitious medium has been found empirically to have a different functional 

form (Kennedy and Strenge, 1992). NRC staff calculations confirm that, in the Tc release model 

used by DOE in Cases K, K1, and K2, Tc release is dominated by new fractures. The NRC staff 

therefore expects the magnitude and timing Tc release predicted by DOE’s model to be 

sensitive to assumptions about rate at which new fracture faces develop. 

Table 2.13-4: Variations of DOE Case K1 PORFLOWTM model. 

(Highlighted values indicate values changed since the previous test.) 

Saltstone Kd (mL/g) Disposal 

Unit Kd

(mL/g) 

Fracturing 

Scheme 

Final

Fracture 

Spacing 

(m) 

Time of 

Peak 

Release 

Rate (yr) 

Dose 

Estimate1

Reduced Oxidized (mrem/yr)

DOE 

Case 

K1 

500 0.8 500 to 217 Log 0.1 12,800 90 

Test 

1 500 0.8 500 to 217 Quadratic 0.1 12,100 86 

Test 

2 500 0.8 0.8 Quadratic 0.1 8730 6802

Test 

3 500 0.8 0.8 Log 0.1 10,300 9302

Test 

4 500 0.8 0.8 Quadratic 1 19,100 25 

Test 

5 139 0.8 0.8 Quadratic 1 10,100 35 

There are 100 mSv/yr in 1 mrem/yr. 

1

 Dose estimated based on the annual fraction Tc inventory released from the near-field domain in each 

case scaled by the ratio of the peak annual fractional Tc inventory released from the near-field domain in 

DOE’s Case K1 to peak dose in Case K1. 

2

 NRC staff believes these doses are significant overestimates attributable to an artifact of DOE’s “singleporosity” average-Kd model that primarily affects cases with rapid oxidation. 

Comparison of the results of DOE’s Case K1 with the results of Test 1 (Table 2.13-4) indicates a 

very small effect of using a more gradual fracture rate. However, the true difference in the 

release rates of Tc from saltstone between DOE Case K1 and the Test 1 Case is obscured by 

the re-immobilization and gradual re-release from the disposal unit floor and walls. To evaluate 

the effect of retention in the disposal unit, the NRC staff re-evaluated DOE’s Case K1 

PORFLOWTM files using a Kd value for the disposal unit floor and walls more typical of oxidized 
191 

concrete. This analysis does not assume that the disposal floors and walls are completely 

oxidized; rather, it simulates the performance of the disposal unit floors and walls if the 

pathways along which Tc would be released from the disposal unit are oxidized. 

Comparison of DOE’s Case K1 with Test 3, which differ only in the Kd used in the disposal unit 

floor and walls, shows that the modeled Tc retention in the disposal unit floor and walls lowered 

the predicted peak doses by slightly more than an order of magnitude (Table 2.13-4). Similarly, 

comparison of Test 1 with Test 2, which also differ only in the modeled retention of Tc in the 

disposal unit floors and walls, also show a similar order-of-magnitude dose reduction 

attributable to retention in the disposal unit concrete. Re-immobilization in the disposal unit floor 

and walls decreases the peak dose because release from the disposal unit floor and walls is 

significantly more gradual than the modeled Tc release from saltstone (Figure 2.13-5). This 

result is of interest principally because it demonstrates the large effect of the assumed disposal 

unit performance in DOE’s Case K1 dose prediction. The predicted peak dose in Tests 2 and 3, 

however, are not expected to be realistic because NRC staff analytical calculations, described 

later in this section, indicate that this result is likely a significant overestimate due to an artifact 

of DOE’s average-Kd model. The artifact related to the average-Kd model appears to be most 

significant in cases with faster rates of saltstone oxidation. Therefore, the dose predictions from 

Tests 4 and 5 are expected to be less effected by the artifact. 

Comparison of Test 2 with Test 3, which differ only in the assumed rate of fracture growth, 

shows the expected decrease in peak dose if a more gradual fracture rate is used (i.e., 

quadratic instead of logarithmic fracture growth). Comparison of Test 2 and Test 3 shows a 

larger effect of the slower fracture growth rate than comparison of DOE’s Case K1 and Test 1 

because, in Tests 2 and 3, the release of Tc from saltstone is not obscured by gradual rerelease from the disposal unit floors and walls. Significantly, Test 2 also shows that if Tc is 

released through oxidized fractures in the disposal unit floor and walls, and a more gradual 

fracture rate is assumed, the predicted peak dose could occur within 10,000 years of site 

closure. 

Test 4 and Test 5 demonstrate the significant effect of the final fracture spacing. In particular, 

comparison of Test 4 with Test 2, which differ only in the final fracture spacing, shows a change 

in fracture spacing from 10 cm (4 in) to 1 m (3 ft) decreases the predicted peak dose by 

approximately a factor of 27 and delays the peak dose from approximately 8,730 to 19,100 

years after closure. In comparison to Test 4, Test 5 shows that using a Kd for reducing grout 

that is similar to values measured by DOE in tests of desorption of Tc from saltstone core 

sample in the presence of trace quantities of oxygen (i.e., 30 to 60 ppm) 

(SRNL-STI-2010-00667) causes a modest increase in dose but a significant advance in the 

timing of the peak dose (i.e., from 19,100 years to 10,100 years after closure). This result is of 

interest because it remains unclear whether saltstone could be exposed to trace quantities of 

oxygen. 

To further test the applicability of the results of the variations in DOE’s Case K1 PORFLOWTM 

model (Table 2.13-4), the staff developed analytical calculations to evaluate the behavior of 

simple average-Kd and dual-Kd models. Both models were based on the same saltstone 
192 

oxidation calculations the NRC staff used to estimate saltstone oxidation for alternate fracture 

spacings and growth rates in its variations of the DOE Case K1 PORFLOWTM model. In 

addition to oxidation proceeding from fracture faces, oxidation also was assumed to take place 

because of oxygen in water flowing through the saltstone. The fraction of oxidation proceeding 

from fracture faces was reduced to account for areas already oxidized by infiltrating water to 

avoid double-counting of the oxidation. In the simple analytical average-Kd model, the fraction 

of oxidized saltstone was used to develop a weighted-average Kd value the same way DOE 

developed the weighted average Kd value in Cases K, K1, and K2. In the dual-Kd model, the 

fraction of saltstone oxidized at each time step was used to predict release from the oxidized 

and reduced fractions separately. The amount of Tc released in each time step was tracked so 

that the oxidized areas would cease to release Tc once the inventory in the oxidized volume 

was depleted. In both models, the primary output was the annual fractional release of Tc from 

saltstone, and no credit was provided for retention of Tc in the disposal unit floors and walls. 

Comparison of the simple average-Kd and dual-Kd models indicates that the average-Kd model 

can under-predict releases in the short-term and significantly over-predict peak release rates in 

some cases (Figure 2.13-5). The NRC staff calculations indicate the average-Kd model overpredicts release rates most significantly in cases in which most of the release comes from the 

oxidized, rather than the reduced, fraction of saltstone. Thus the effect is most pronounced 

when there is relatively low water flow through saltstone to release Tc from the reduced region 

before significant oxidation occurs. Comparison of model results indicates that the dual-Kd

model allows a more gradual release of radionuclides, resulting in a smaller available inventory 

at times when the average-Kd model would predict nearly the entire initial inventory is released 

with a relatively low average sorption coefficient. This result is similar to the result of DOE’s 

comparison of an average-Kd model and a dual-Kd model (which DOE refers to as single- and 

dual-porosity models) (SRR-CWDA-2011-00114), although the NRC staff’s simplified models 

predicted a more substantial difference between the two approaches than predicted by DOE. 

The magnitude of the overprediction of peak dose that both NRC and DOE expect the averageKd model to cause is unclear. DOE’s conclusion that the average-Kd model provides a good 

representation of Tc release is based on a comparison of two GoldSim® models rather than a 

comparison involving the actual PORFLOWTM model DOE used to model Tc release. NRC staff 

concluded DOE’s GoldSim® comparison was likely to underestimate the difference between the 

models, as compared to a similar comparison performed in PORFLOW, because of certain 

implementation details in DOE’s GoldSim® models (Section 2.7). However, the DOE models 

include diffusion between the oxidized and reduced regions, which is an important process not 

included in the simplified NRC staff models. In DOE’s Case K, K1, and K2 analyses, the 

expected over-prediction of peak dose attributable to the average-Kd model is further obscured 

by the re-concentration and gradual re-release of Tc from the disposal unit floor and walls 

(Figure 2.13-7). 
193 

Figure 2.13-8 Comparison of annual fractional release from saltstone predicted by NRC’s 

simple analytical average-Kd and dual-Kd models, using DOE Case K1 fracturing, nearfield flow, and Kd values. Both models simulate the same cumulative release. 

Figure 2.13-9: Comparison of annual fractional release from saltstone and from the nearfield domain (i.e., saltstone, disposal units, and unsaturated soil) from DOE Case K1 

model output files (STAT.out files [NRC, 2010g]) with the NRC predicted saltstone 

oxidation for Case K1 fracturing. 

0.0E+0

5.0E-4

1.0E-3

1.5E-3

2.0E-3

2.5E-3

3.0E-3

3.5E-3

4.0E-3

0 5000 10000 15000 20000

Annual Fractional Release (1/yr)

Time Since Site Closure (years)

Average Kd model

Dual-Kd model

0E+00

1E-03

2E-03

3E-03

4E-03

5E-03

6E-03

7E-03

0 5000 10000 15000 20000

Fraction, 1/yr

Time since Closure (years)

PORFLOW - DOE Case K1 

release from the near-field

PORFLOW - DOE Case K1 

release from Saltstone

NRC calculation of fraction 

oxidized for Case K1 fracturing
194 

In addition to evaluating the applicability of the average-Kd model, the NRC staff’s analytical 

calculations based on the simple dual-Kd model were used to map a greater range of parameter 

variation than could be addressed efficiently with variations in the DOE Case K1 PORFLOWTM 

model. The NRC staff did not use these values for comparison to the performance objectives. 

Rather, these values were used to indicate factors to which the Case K1 dose may be sensitive 

and to develop an estimate of the uncertainty in the Case K1 value. As previously discussed, 

this effort was undertaken because the NRC staff concluded the DOE Case K1 average-Kd

model overestimated the expected annual fractional release from the saltstone waste form in a 

way that was difficult to isolate using the PORFLOWTM model. In addition, the effort was 

necessary because, unlike Cases A - E, Case K1 was supplied without an uncertainty analysis 

and with limited deterministic sensitivity analysis (i.e., the only sensitivity analysis provided was 

a comparison of the results of Cases K, K1, and K2, which used different Kd values to represent 

Tc sorption in saltstone). 

Specifically, the NRC staff used the analytical calculations to evaluate the effects of changing 

(1) saltstone fracture spacing at 10,000 years after closure, (2) the function describing the rate 

of fracture growth with time, (3) flow through saltstone between zero and 20,000 years after site 

closure, (4) sorption coefficients for Tc in oxidizing and reducing saltstone, and (5) the reducing 

capacity of saltstone. As in the variations on DOE’s Case K1 PORFLOWTM model, results were 

evaluated for final fracture spacings of 1 fracture every 10 cm (4 in) and 1 fracture every 1 m 

(3 ft). Similarly, the fracture growth rate also was alternated between logarithmic growth and 

quadratic growth with time. Three different options were evaluated with respect to flow through 

saltstone. One set of calculations used the near-field flow through saltstone from DOE’s Case 

K1 PORFLOWTM model. Another two sets of calculations assumed flows were limited by the 

hydraulic conductivity of saltstone. First, the hydraulic conductivity of saltstone was assumed to 

be 1x10-7 cm/s for the first 10,000 years after closure and to increase linearly to 1x10-6 cm/s by 

20,000 years after closure. Second, the hydraulic conductivity of saltstone was assumed to be 

1x10-8 cm/s at closure, increasing linearly to 1x10-7 cm/s by 10,000 years after closure and 

1x10-6 cm/s by 20,000 years after closure. Sorption coefficient (Kd) values in oxidizing saltstone 

were alternated between 0.8 mL/g and a nominal value of 1x10-4 mL/g to test the importance of 

sorption in the oxidized region. Sorption coefficient (Kd) values in the reduced region were 

alternated between 500 mL/g and 139 mL/g. The reducing capacity of saltstone was alternated 

between the Case A value (0.822 meq e-

/g) and the Case K1 value (0.206 meq e-/g). The 

matrix of results from varying each of these parameters is provided in Appendix B. 

In general, selecting combinations of parameters as described above yielded fractional release 

rates corresponding to doses between 0.1 and 1.7 mSv/yr (10 and 170 mrem/yr). In most 

cases, the peak release rate within 10,000 years was equivalent to the peak release rate within 

20,000 years of site closure. The largest consistent exception to that generality is cases based 

on a Case A reducing capacity (0.822 meq e-

/g) and 1 m (3 ft) final fracture spacing, which 

typically had peak doses within 20,000 years that were approximately three times greater than 

the peak doses within 10,000 years of closure because the slower oxidation resulting from less 

fracturing and more reducing capacity lead to later release peaks. 
195 

Of the factors evaluated, the factor with the largest apparent effect on the fractional release rate 

was the final fracture spacing. The calculations most likely to yield relatively low release rates 

were calculations with a final fracture spacing of 1 m (3 ft) and Case A reducing capacity 

(0.822 meq e-

/g) (See Appendix B for release rate results). In general, for cases with a 1 m 

(3 ft) fracture spacing and Case A reducing capacity, cases in which water flow was based on 

DOE’s Case K1 PORFLOWTM flow yielded slightly larger release rates than cases in which 

water flow was limited by saltstone hydraulic conductivity (i.e., limited to 1x10-7 or 1x10-8 cm/s at 

closure and 1x10-7 cm/s at 10,000 years after closure). The PORFLOWTM flow yielded larger 

peaks even though choosing an initial hydraulic conductivity of 1x10-7 cm/s allowed more water 

flow through for the first several thousand years after closure. The NRC staff hypothesizes that 

the cases based on DOE’s Case K1 flow resulted in larger peak doses because the flow 

remains low while oxidation from fracture faces proceeds and increases when the saltstone is 

relatively oxidized. 

Variations in the fracture growth rate had a modest effect in most cases; although in cases with 

a low final fracture spacing (i.e., 10 cm [4 in]) and Case K reducing capacity (i.e., 0.206 meq 

e-

/g) logarithmic fracture growth could result in approximately twice the fractional release rate of 

quadratic growth. The difference between using a Kd value of 0.8 mL/g and modeling the 

oxidizing saltstone with essential no sorption in oxidizing saltstone (i.e., Kd of 1x10-4 mL/g) was 

imperceptible in most cases and small in others. The difference between using a Kd of 

500 mL/g and 139 mL/g in the reducing grout was slightly larger but still within a factor of two in 

the cases evaluated. 

While the results of these analyses were not used as a basis for assessing compliance, 

comparison of the average-Kd and dual-Kd models supported the NRC conclusion that the tests 

run with the DOE PORFLOWTM Case K1 model modified to reflect Tc release through oxidized 

pathways in the disposal unit floors and walls (i.e., Tests 2 and 3 in Table 2.13-4) produced 

artificially large dose predictions due to an artificially large modeled pulse release from the 

waste form. In addition, comparison of the average-Kd and dual-Kd models showed that, in 

many cases, average-Kd model tended to delay the peak release rate as compared to the dualKd model. This result is of interest because variations of DOE’s PORFLOWTM Case K1 model 

(Table 2.13-4) showed that, if Tc is assumed to be released through oxidized pathways in the 

disposal unit floors and walls, the peak dose can occur within (Test 2) or around (Test 3 and 5) 

10,000 years after site closure. If the average-Kd model tends to delay the dose, it appears that 

the peak dose predicted in Tests 3 and 5, as well as Test 2, all may occur within 10,000 years of 

site closure. 

Similarly, while the results of sensitivity analyses performed with the dual-Kd model were not 

used as a basis for a compliance determination, these analyses highlighted the importance of 

the final fracture spacing and the assumed specific reducing capacity for saltstone. 

In summary, the staff conducted two types of sensitivity analyses related to DOE’s Cases K, K1, 

and K2. First, the staff used DOE’s PORFLOWTM model to test the effects of variations in 

model parameters on the peak doses from Ra-226 and Tc-99. With respect to Ra-226, the 

NRC staff concludes that site-specific information on the cementitious material Ra-226 Kd
196 

values is needed. Therefore, the NRC staff will monitor the development of site-specific Kd

values for Ra-226 in saltstone and the disposal unit concrete. With respect to Tc-99, based on 

variations of DOE’s Case K1 PORFLOWTM model, the staff concluded that that changes in 

assumptions about fracture spacing and chemical retention of Tc-99 in disposal unit concrete 

significantly affected the peak dose. The NRC staff also concluded that peak doses from Tc-99 

could occur within 10,000 years of disposal. The second type of sensitivity analysis the NRC 

staff conducted was based on simple analytical representations of a dual-Kd and average-Kd

model for Tc-99 release. Based on the results of these models, the NRC staff concluded that 

the average-Kd model tends to delay and overestimate the peak Tc-99 release in DOE’s Case K 

PORFLOWTM model. The saltstone release rates calculated with the simplified dual-Kd model 

generally supported the saltstone release rates calculated with variations of DOE’s Case K 

PORFLOWTM model. 

2.13.3.4 Comparison of Results to Performance Objectives 

The Performance Objective for Protection of the General Population from Releases of 

Radioactivity in §61.41 states: 

Concentrations of radioactive material which may be released to the 

general environment in ground water, surface water, air, soil, plants, or 

animals must not result in an annual dose exceeding an equivalent of 25 

millirems to the whole body, 75 millirems to the thyroid, and 25 millirems 

to any other organ of any member of the public. 

The NRC has evaluated compliance using a dose limit of 0.25 mSv/yr (25 mrem/yr) Total 

Effective Dose Equivalent (TEDE), consistent with the approach discussed in the Final Rule for 

Disposal of High-Level Radioactive Wastes in a Proposed Geologic Repository at Yucca 

Mountain, Nevada (66 FR 55752): 

Because each of the organs had the same limit under the older system 

even though each had a different level of radiosensitivity, it is very difficult 

to directly compare the old standards with the new standards. As noted 

in the proposed rule, the Commission considers 0.25 mSv/yr 

(25 mrem/yr) TEDE as the appropriate dose limit to compare with the 

range of potential doses represented by the older limits that had whole 

body dose limits of 0.25 mSv/yr (25 mrem/yr). 

To evaluate doses for comparison to a 0.25 mSv/yr (25 mrem/yr) TEDE limit, DOE and NRC 

have used the updated internal dose factors of the International Commission on Radiological 

Protection (ICRP) Publication 72 (ICRP, 1996), consistent with Commission direction in the Staff 

Requirements Memorandum on Processes for Revision of 10 CFR 20 Regarding Adoption of 

ICRP Recommendations on Occupational Dose Limits and Dosimetric Models and Parameters 

(SRM-SECY-01-0148). 
197 

As discussed in Section 2.3.1, compliance with the performance objectives typically is based on 

the predicted peak dose within 10,000 years of site closure. DOE predicted the peak dose to a 

member of the general public within 10,000 years of site closure for deterministic Case A, which 

DOE considers its base case, to be 0.014 mSv/yr (1.4 mrem/yr) in Sector B. DOE’s peak 

predicted dose within 20,000 years for deterministic Case A is 0.031 mSv/yr (3.1 mrem/yr) in 

Sector I. The peak of the mean dose from DOE’s “all cases” probabilistic analysis (i.e., a 

weighted average of Cases A through E) within 10,000 years of closure is approximately 

0.04 mSv/yr (4 mrem/yr) in Sector B and the all cases peak of the mean result within 20,000 

years is approximately 0.10 mSv/yr (10 mrem/yr) in Sector J (2009 PA; Figure 5.6-33). All of 

these predicted doses are below the dose limit of 0.25 mSv/yr (25 mrem/yr) specified for 

protection of the general population in §61.41. As discussed in Section 2.1, the NRC staff does 

not consider Case A to be an acceptable compliance case because (1) it is not reflective of 

current site conditions, (2) it does not account for measured ranges of parameter values or 

expected differences between laboratory and as-emplaced values, and (3) it does not 

adequately account for potential changes in site conditions during the compliance period. 

Additionally, because the probabilistic analysis incorporated near-field flow values from the 

deterministic PORFLOWTM analyses, because the subjective weights given to each alternate 

case in the probabilistic model were unsupported, and because of apparent modeling 

inconsistencies in the probabilistic analysis, the NRC staff has not used the results of DOE’s 

probabilistic analysis in its compliance evaluation. 

In DOE’s base case (Case A) and most of its sensitivity analyses, Ra-226 dominates the dose 

to a member of the general population. Based on DOE analyses and the results of NRC’s 

sensitivity analyses related to Ra-226 (Section 2.13.3.3), the NRC staff concludes that the dose 

from Ra-226 is not likely to exceed 0.25 mSv/yr (25 mrem/yr). However, if key model 

parameters vary from the values assumed in these cases, the estimated dose may increase. 

The NRC staff concludes that the Ra-226 dose is less risk significant than the Tc-99 dose. 

However, the NRC staff believes that the collection of better information on the Th-230 inventory 

or on key parameters, such as the Ra-226 Kd value in saltstone and disposal unit concrete, 

would be useful in confirming that the Ra-226 dose will not exceed the 0.25 mSv/yr 

(25 mrem/yr) dose limit consistent with the §61.41 performance objective. 

In response to NRC concerns about using Case A as the base case (NRC, 2010b, i; PA-8), 

DOE developed Cases K, K1, and K2. Because they resolve many of the specific concerns the 

NRC staff has about using Case A as the compliance case (Section 2.1), the NRC staff has 

relied heavily on these cases in its review. In its response to NRC concerns about Case A 

(SRR-CWDA-2011-00044; PA-8), DOE expressed concern that relying heavily on Case K would 

create the impression that Case A was not the most likely case. The NRC staff understands 

DOE’s desire to develop a case that provides a “best estimate” of the dose. However, because 

Case A appears to be overly optimistic compared to present day conditions (e.g., by assuming 

no fracturing of saltstone (Section 2.6), a hydraulic conductivity lower than measured values for 

field samples (Section 2.6), and Tc sorption coefficients greater than relevant measured values 

(Section 2.7)), the NRC staff believes Case A is not the most likely case and does not provide a 

best estimate of the dose from the SDF. 
198 

As previously discussed, Case K, K1, and K2 resolve many of the concerns the NRC staff has 

with the use of Case A as a case to compare to the performance objectives. However, as 

discussed in Section 1.2, because of the complexity of the system being modeled, the NRC staff 

indicated to DOE that it could not make any specific conclusions about the assumptions used in 

Case K until the staff reviewed DOE’s written responses to RAI-2009-02 (NRC, 2011c). 

Although DOE characterized Case K as a non-mechanistic, pessimistic sensitivity analysis, 

upon review of Cases K, K1, and K2, the NRC staff concluded that, while some of the 

assumptions DOE used appear to be conservative, others appear to be realistic or overly 

optimistic. One apparently realistic (as opposed to pessimistic) change is that Cases K, K1, and 

K2 use lower inventories of Ra-226 and its parent Th-230 than Case A, which lowers the peak 

dose from Ra-226. Similarly, Cases K, K1, and K2 use updated biosphere parameters that also 

lower the dose from Ra-226 and other radionuclides (SRR-CWDA-2011-00044 response PA-9). 

Other changes do not appear to be as pessimistic as DOE indicates. For example, DOE 

characterized the reduction of the specific reducing capacity of saltstone to 25% of its Case A 

value as “arbitrary” (SRR-CWDA-2011-00044 response PA-8). However, the factor was based 

on the fraction of blast furnace slag in saltstone because it is unclear why the measured specific 

reducing capacity for saltstone would be equivalent to the measured reducing capacity of pure 

blast furnace slag (Section 2.6). Similarly, DOE characterizes the initial saturated hydraulic 

conductivity of 1x10-8 cm/s used in Case K as pessimistic, but the value is actually less than 

(i.e., more favorable than) values measured for present-day as-emplaced saltstone 

(Section 2.6). Although DOE discounts the measured values in as-emplaced saltstone as an 

artifact of the sampling method, the same sampling method used on laboratory-prepared 

samples did not elevate the measured hydraulic conductivity as compared to cast samples 

(SRNL-STI-2010-00657). In summary, the NRC staff does not agree with DOE’s 

characterization of Cases K, K1, and K2 as overly pessimistic sensitivity analyses and, instead, 

believes these cases provide a more probable estimate of the peak dose from the SDF than 

Case A. 

In all three cases, DOE projected a dose peak at approximately 8,000 years due primarily to 

I-129, while a larger peak due to Tc-99 is predicted to occur at different times in Case K, K1, 

and K2 (Figure 2.13-1). In Case K, DOE predicts a 0.55 mSv/yr (55 mrem/yr) dose peak from 

Tc-99 to occur at approximately 15,300 years after site closure. However, Case K uses sorption 

coefficients (Kd values) in reduced (1,000 mL/g) and oxidized (10 mL/g) cementitious material 

that the NRC staff does not believe are consistent with DOE experimental results (Section 2.7). 

DOE predicts the dose from Case K1, which uses better-supported Kd values for Tc in the 

saltstone (i.e., 500 mL/g in reducing saltstone and 0.8 mL/g in oxidized saltstone), to be 

approximately 0.90 mSv/yr (90 mrem/yr) and to occur at approximately 12,900 years after site 

closure (Figure 2.13-1). 

As described in Section 2.13.3.2 and 2.13.3.3, the magnitude and timing of predicted peak dose 

from DOE Case K1 is sensitive to and delayed by (1) DOE assumptions about the timing and 

rate of saltstone fracture growth, (2) DOE’s use of the average-Kd model for Tc release, and 

(3) seemingly optimistic assumptions about Tc re-reduction and re-immobilization in disposal 

unit floors and walls. Because of the considerable resulting uncertainties in the timing of the 

peak Tc dose, the NRC staff does not have reasonable assurance that the maximum dose from 
199 

Case K1 (i.e., 0.9 mSv/yr [90 mrem/yr]) will occur beyond 10,000 years after site closure as the 

DOE analysis concludes. Thus, based on NRC’s use of Case K1 in its compliance evaluation 

(Section 2.1), the value of the predicted peak dose of 0.90 mSv/yr (90 mrem/yr), and the large 

uncertainty in the predicted timing of the dose, the NRC staff could not conclude it had 

reasonable assurance that the dose would not exceed the 0.25 mSv/yr (25 mrem/yr) dose limit 

specified in the performance objective within 10,000 years of site closure. 

Rather than base the compliance demonstration on DOE’s Case K1 alone, the NRC staff 

performed additional analyses to determine whether it could develop additional assurance about 

the timing or magnitude of the Case K1 peak dose. As described in Section 2.13.3.2 and 

2.13.3.3, the NRC staff believes the projected dose from Case K1 is artificially increased by 

DOE’s average-Kd approach to Tc sorption and artificially decreased by unrealistic assumptions 

about Tc retention in and gradual release from the disposal unit floors and walls. However, as 

described in Section 2.13.3.3, the NRC staff sensitivity analyses performed with DOE’s Case K1 

model, as well as NRC staff calculations with simplified models of Tc release, support a dose 

prediction within an approximate range that is between the limits established in the §61.41 

performance objective (0.25 mSv/yr [25 mrem/yr]) and the public dose limit found in §20.1301 

(i.e., 1 mSv/yr [100 mrem/yr] total effective dose equivalent). Depending on information 

developed to support assumptions about the inventory of Ra-226 and its ancestors, the 

uncertainty in the potential dose from Ra-226 and its timing could slightly increase dose 

predictions, but is not expected to significantly affect these estimates. Thus, based on current 

available information and model support, the NRC staff does not have reasonable assurance 

that salt waste disposal at the SDF meets the 10 CFR 61 Performance Objective for Protection 

of the General Population from Releases of Radioactivity. Although the NRC staff cannot 

conclude that the performance objective in §61.41 is met, the potential dose to an off-site 

member of the public from DOE’s disposal actions is still expected to be relatively low (i.e., 

approximately 1 mSv/yr [100 mrem/yr], the public dose limit29 in §20.1301). 

Furthermore, because the staff expects that any exceedance of the §61.41 limit would occur 

long after site closure, DOE may be able to develop additional information or take mitigative 

actions in the short term (e.g., add a strong chemical reducing agent to the saltstone formula 

and improve the quality assurance for saltstone grout) that could provide reasonable assurance 

that salt waste disposal at the SDF meets the 10 CFR 61 Subpart C performance objectives. 

The NRC staff has identified these issues factors it plans to use as a basis for further SDF 

monitoring (Section 6.2 and Table A-1). As described in Section 6.3, the NRC staff will review 

information DOE provides during monitoring to determine if the information addresses the 

technical concerns identified by the staff in the TER and provides reasonable assurance that 

salt waste disposal at the SDF meets the §61.41 Performance Objective. 

 

29 As indicated in the Statements of Consideration for 10 CFR 20 (56 FR 23374), the 1 mSv/yr (100 mrem/yr) value in 

the final rule represents the primary dose limit for protection of the public. The Statements of Consideration also 

indicate that the dose limit “applies only to doses from radiation and radioactive materials under the licensee's 

control.” At the SRS, there may be multiple sources of radioactivity that could contribute to dose. The 1 mSv/yr 

(100 mrem/yr) value is referenced here not as a regulatory limit, but as an indication of the relative magnitude of the 

potential dose. 
200 

3. Protection of Inadvertent Intruders 

3.1 Assessment of Inadvertent Intrusion 

To demonstrate that the performance objective in §61.42, protection of inadvertent intruders, is 

met, DOE performed an evaluation of the potential long term dose to an individual who 

inadvertently intrudes into the disposal site following the assumed institutional control period 

(i.e., 100 years after site closure) to assess compliance with the performance objective in 

§61.42. DOE considered both acute and chronic intruder scenarios and used the chronic 

intruder agriculture scenario as the basis for the compliance evaluation. In the chronic intruder 

agriculture scenario, a farmer is assumed to live on site and use water from an on-site well as a 

drinking water source as well as a source for agriculture (i.e., irrigation and livestock water). 

DOE assumes that the well will not go directly through the disposal units because the disposal 

units contain long-lasting materials which are clearly distinguishable from the surrounding soil. 

DOE noted that because the local soil is generally sandy, local well drillers do not expect to 

need to drill through high-strength geologic materials when constructing a drinking water well, 

and consequently the well driller would stop operations and move to a different location upon 

encountering engineered barriers, such as the closure cap erosion barrier or the disposal unit 

concrete roof. Although DOE does not consider the construction of a well through saltstone to 

be a credible scenario, DOE performed two sensitivity analyses to assess the potential dose 

resulting from drilling directly into a disposal unit. 

The pathways assumed in the chronic intruder agriculture calculation were the same as those 

assumed for the member of the public (Section 2.3) except that the groundwater concentrations 

were based on different well locations (on the SDF instead of outside of a 100 m boundary). 

The biosphere modeling approach and parameters used in this assessment were also the same 

as those used for the member of the public (Section 2.10). As with the dose assessment for the 

member of the public, the concentrations of radionuclides in the groundwater were determined 

using the PORFLOWTM code (Section 2.7), and the dose was calculated based on these 

concentrations using DOE’s deterministic model GoldSim®(Section 2.10). However, the 

external dose due to direct exposure to the waste form through the engineered cover and the air 

pathway dose from diffusion of radionuclides from the waste form to the surface were evaluated 

in separate calculations (SRNS-J2100-2009-00006 and Section 2.9, respectively). The doses 

from both the direct exposure and air pathways were both predicted to be negligible. As was 

true for the dose assessment for the member of the public, in the intruder assessment DOE also 

assumes that the inventory in the FDCs is evenly distributed among the 64 FDCs. 

The cases evaluated by DOE as part of the intruder assessment are summarized in Table 3-1. 

The Case A PORFLOWTM analysis was used in the initial evaluation of the dose to an 

inadvertent intruder. In this case, DOE used the highest average groundwater concentration 

within the 15 m by 15 m (50 ft by 50 ft) grid cells that encroached on the 1 m perimeter from the 

SDF (Figure 2.3-1). In the first NRC RAI (NRC, 2010b), the NRC staff expressed concern that 

the average concentration over the entire grid cell might be much less than the concentration at 

1 m from a disposal unit, especially for slow moving and relatively fast decaying radionuclides. 

In the initial DOE RAI response (SRR-CWDA-2010-00033), DOE performed an analysis of the 
201 

dose to an individual who uses water from directly below Vault 4. The concentration of 

radionuclides in this water was determined based on the flux of radionuclides entering the water 

table below Vault 4 that was calculated using the Case A PORFLOWTM model. This flux was 

converted to a concentration based on the area of the Vault 4 footprint and the Case A Darcy 

velocity of water into the water table from the vadose zone directly below the vault. In the 

second RAI response (SRR-CWDA-2011-00044), DOE performed a similar calculation for an 

FDC. However, this analysis used the updated biosphere parameters from Case K. DOE also 

provided a revised estimate of the dose to an individual using water taken from directly under 

Vault 4 based on the updated biosphere parameters. 

In the initial DOE RAI response (SRR-CWDA-2010-00033), DOE also used the Case A model 

to calculate a dose from the groundwater concentration in a 15 m by 15 m (50 ft by 50 ft) grid 

cell that encroached on the 1 m perimeter from an individual FDC. In this case, the calculated 

groundwater concentrations were multiplied by 10 to account for the maximum expected 

inventory in an individual FDC. 

DOE also provided predicted intruder doses based on the Case K PORFLOWTM model 

(SRR-CWDA-2011-00044). In this analysis, the dose was estimated based on the 

concentration of radionuclides in the groundwater beneath Vault 4 and an FDC. For the FDCs, 

the dose was based on the maximum concentration of each radionuclide in the groundwater 

under any FDC. This analysis also included the dose from stream pathways (e.g., swimming, 

fish ingestion), and used the updated Case K biosphere parameters. 

In the acute intruder sensitivity analysis, DOE assumed that a well is drilled directly through a 

disposal unit and the well cuttings are spread across a garden. An individual is then exposed to 

the cuttings while spending time in the garden (20 hours/year). This individual receives a dose 

from the ingestion and inhalation of soil containing drill cuttings, and from direct exposure to the 

soil containing drill cuttings. The inventory in the cuttings was determined based on a 

hypothetical 20 cm (8 in) core drilled directly through Vault 1, Vault 4, or an FDC. The highest 

value from the three disposal unit types was assumed for each radionuclide. DOE also 

performed a chronic intruder sensitivity analysis that used the same drilling scenario use in the 

acute intruder analysis. This scenario considered both the exposure to the drill cuttings, as well 

as the pathways considered in the chronic intruder agriculture scenario (e.g., use of 

groundwater as a drinking water source as well as a source for agriculture). 

The period of performance assumed by DOE for the inadvertent intruder for the purpose of 

demonstrating compliance with the performance objective in §61.42 was a period of 10,000 

years following the closure of the site, as was true for the assessment of compliance with the 

performance objective in §61.41 (Section 2.3). DOE also ran all transport models to at least 

20,000 years to determine peak concentrations that occur after the 10,000-year performance 

period. 

In the Chronic Intruder: 1 m from the SDF Case, 91% of dose reported by DOE as occurring 

within 10,000 years is from Ra-226. The main pathways that contribute to the dose are drinking 

water consumption (47%), vegetable consumption (37%), and fish consumption (16%). 
202 

Table 3.1-1: Description of Intruder Cases Evaluated by DOE 

Case Source Term 

Used 

updated 

biosphere 

parameters?

Reference 

Chronic Intruder: 

1 m from the SDF 

Case A 

Groundwater in grid cells that 

encroached on the 1 m 

perimeter of the SDF 

no 

SRR-CWDA-2009-00017 

(p597) 

Chronic Intruder: 

Vault 4 

Case A 

Groundwater directly under 

Vault 4 no 

SRR-CWDA-2010-00033 

(p96) 

Chronic Intruder: 

Vault 4 Case A 

Revised 

Biosphere

Case A 

Groundwater directly under 

Vault 4 yes 

SRR-CWDA-2011-00044 

(p258) 

Chronic Intruder: 

Vault 4 

Case K 

Groundwater directly under 

Vault 4 yes 

SRR-CWDA-2011-00044 

(p102) 

Chronic Intruder: 

FDC Case A 

Increased 

Concentrations 

Case A 

Groundwater in grid cells that 

encroached on the 1 m 

perimeter of the SDF. 

Calculated concentrations were 

multiplied by factor of 10 to 

account for maximum expected 

inventory. 

no 

SRR-CWDA-2010-00033 

(p96) 

Chronic Intruder: 

FDC Case A 

Revised 

Biosphere 

Case A 

Groundwater directly under a 

FDC yes 

SRR-CWDA-2011-00044 

(p258) 

Chronic Intruder: 

FDC 

Case K 

Groundwater directly under a 

FDC yes 

SRR-CWDA-2011-00044 

(p102) 

Acute Intruder: 

Drill Cuttings 

Sensitivity 

Analysis 

NA 

Drill cuttings from drilling 

through saltstone. The highest 

concentration of Vault 1, Vault 4, 

or the FDCs was assumed for 

each radionuclide. 

no 

SRR-CWDA-2009-00017 

(p604) 

Chronic Intruder: 

Drill Cuttings 

Sensitivity 

Analysis 

NA

Drill cuttings from drilling 

through saltstone. The highest 

concentration of Vault 1, Vault 4, 

or the FDCs was assumed for 

each radionuclide. 

no 

SRR-CWDA-2009-00017 

(p605) 

The doses calculated for each of these cases reported by DOE as occurring within 10,000 years and 

20,000 years are presented in Table 3-2. 
203 

Table 3.1-2: Results of Intruder Assessment Performed by DOE 

Case 

within 10,000 yrs within 20,000 yrs

Dose Reference 

(mrem/yr)1

Time 

(yrs) 

Dose 

(mrem/yr)1

Time 

(yrs) 

Chronic Intruder: 

1 m from the SDF2 1.9 10,000 7.2 15,060 SRR-CWDA-2009-00017 

(p597) 

Chronic Intruder: 

Vault 4 Case A 35 9,760 

SRR-CWDA-2010-00033 

(p96) 

Chronic Intruder: 

Vault 4 Case A 

Revised Biosphere 

48 ~9,950 ~92 ~16,500 SRR-CWDA-2011-00044 

(p258) 

Chronic Intruder: 

Vault 4 Case K 8.1 ~8,700 ~60 ~13,000 SRR-CWDA-2011-00044 (p102) 

Chronic Intruder: 

FDC Case A 

Increased 

Concentrations 

1.5 10,000 

SRR-CWDA-2010-00033 

(p96) 

Chronic Intruder: 

FDC Case A 

Revised Biosphere 

0.96 ~10,000 ~20 ~15,100 SRR-CWDA-2011-00044 

(p258) 

Chronic Intruder: 

FDC Case K 53 ~7,900 ~290 ~15,500 SRR-CWDA-2011-00044 

(p102) 

Acute Intruder: Drill 

Cuttings Sensitivity 

Analysis 

3.83

 500 3.83 500 SRR-CWDA-2009-00017 

(p604) 

Chronic Intruder: 

Drill Cuttings 

Sensitivity Analysis 

153

 500 153 500 SRR-CWDA-2009-00017 

(p605) 

1

 1 mrem/yr = 0.01 mSv/yr 

2

 This case included the dose from plumes originating from multiple disposal units 

3

 The dose prior to 500 years was not considered in this analysis because DOE expects the disposal unit 

roofs and clean cap to provide a robust barrier to intrusion directly into the saltstone waste form for at 

least 500 years after site closure. 

In the Chronic Intruder: Vault 4 Case K analysis, the dose reported as occurring within 10,000 

years is primarily due to Cs-135 (60.2%) and I-129 (34.4%), and the main pathways that 

contribute to the dose are fish consumption (63.3%), drinking water consumption (26.2%), and 

vegetable consumption (7.4%). 

In the Chronic Intruder: FDC Case K analysis, 94.9% of the dose reported as occurring within 

10,000 years is from I-129, and the main pathways that contribute to the dose are drinking water 

consumption (68.7%), and vegetable consumption (18.8%). DOE did not provide dose by 

pathway or radionuclide for the Case K doses within 20,000 years, but based on Figures PA-8.8 
204 

and PA-8.9 in SRR-CWDA-2011-00044, which show the member of the public dose, this dose is 

likely to be predominately from Tc-99. The dose calculated in the acute intruder sensitivity 

analysis was primarily due to Sn-126 and direct external exposure. 

DOE performed an uncertainty analysis using the probabilistic GoldSim®model described in 

Section 2.1. In this analysis, the 1 m concentrations were estimated from the 100 m 

GoldSim®concentrations using a multiplier of 1.6. This multiplier was based on the maximum 

ratio of the maximum Case A 1 m concentration to the maximum Case A 100 m concentration 

for any of the key radionuclides. 

3.2 NRC Evaluation – Protection of Intruders 

A specific numerical performance objective is not provided in §61.42 for intruder assessment, 

however a dose limit of 5 mSv (500 mrem) per year was described in the Draft Environmental 

Impact Statement for 10 CFR 61 for development of waste classification requirements 

(NRC, 1981), which were developed to provide for inadvertent intruder protection. Consistent 

with the review procedures in NUREG-1854 (NRC, 2007b), Chapter 5, the staff uses this 5 mSv 

(500 mrem) annual dose value in evaluating the intruder scenarios. 

The NRC staff finds that the scenarios and pathways analyzed by DOE for the assessment of 

the performance objective in §61.42 (protection of an inadvertent intruder) are appropriate 

based on the regional practices near SRS. The NRC staff finds that DOE’s revised approach of 

taking the groundwater concentration from directly below the disposal unit is preferable to the 

use of the average concentration in the 15 m by 15 m (50 ft by 50 ft) cells that begin at a 

distance of 1 m from the disposal unit, although this revised approach is conservative, because 

the average concentration over this large grid cell could be much less than the concentration at 

1 m from a disposal unit. Dilution in a grid cell could be especially problematic for slow moving 

and relatively short-lived radionuclides. This type of radionuclide was not identified as being 

significant in DOE’s analysis of the concentration of radionuclides under the disposal units, but 

they could become important if key parameter values change. As was true for the member of 

the public analysis, the NRC staff believes that Case K1 best represents the current and 

expected future conditions (Sections 2.7 and 2.13). DOE did not perform an intruder analysis 

for Case K1, but Case K was analyzed, which differed from Case K1 only in the Kd values 

selected for Tc. 

As noted in Sections 2.7 and 2.13, although NRC staff believes that Case K1 is the DOE case 

that best represents the system, this case does contain aspects that are likely conservative as 

well as aspects that are likely non-conservative. The aspects of this case that are potentially 

conservative include the assumed rapid rate of fracturing of the saltstone, the large amount of 

fracturing assumed at 10,000 years (i.e., a through-going fracture every 0.1 m), and the 

modeled faster peak rate of release caused by the use of the average-Kd approach (referred to 

by DOE as a “single porosity” model). Additionally, in the case of the intruder analysis, the 

assumed use of the groundwater directly below the disposal unit without dispersion or dilution in 

the saturated zone may be conservative. A non-conservatism in Case K1 is the modeled 

performance of the disposal unit wall and floor (Section 2.13). Also, as noted below, the NRC 
205 

staff considers DOE’s assumption regarding the timing of the fracturing (i.e., limited fracturing 

until almost 10,000 years) to be non-conservative. The NRC staff also finds that the Kd values 

used for saltstone in Case K (i.e., 1000 mL/g for reduced saltstone and 10 mL/g for oxidized 

saltstone versus 500 mL/g and 0.8 mL/g used for Case K1), which DOE used for the intruder 

analysis, is not conservative (Section 2.6). This combination of conservatisms and 

non-conservatisms clouds the conclusion of whether the outcome is likely to over- or underpredict potential doses. 

NRC staff does not find that the method used by DOE to perform the uncertainty analysis for the 

intruder (i.e., the use of a multiplier based on the ratios of the 1 m to 100 m concentration of key 

radionuclides) to be appropriate because the ratio of the 1 m to 100 m concentrations is 

probably higher than the assumed multiplier for some radionuclides. Because DOE determined 

the key radionuclides based on the radionuclides that cause the highest dose at 100 m, 

radionuclides that travel slowly enough and have short enough half-lives to decay before 

reaching 100 m were not considered in the development of the multiplier. The ratio of the 

groundwater concentration at 1 m from the SDF to the concentration at 100 m from the SDF 

would be considerably higher for slow-moving, short-lived radionuclides than for the key 

radionuclides. None of the radionuclides that contributed significantly to the deterministic 

Case A and Case K intruder doses were ones that sorb significantly (and are, therefore, slow 

moving). However, when a range of parameter values is considered in a probabilistic 

uncertainty analysis there may be some realizations in which such radionuclides are more 

significant. Additionally, as discussed in Section 2.11, NRC staff has concerns regarding the 

probabilistic GoldSim®model that was used to perform this uncertainty analysis. 

As was true for the member of the public assessment, the NRC staff is focusing on the Case K 

results in its evaluation to determine whether there is reasonable assurance that the 

performance objective in §61.42 is met. The Case K results for Vault 4 were 8.1x10-2 mSv/yr 

(8.1 mrem/yr) within 10,000 years and approximately 0.6 mSv/yr (60 mrem/yr) within 

20,000 years. The Case K dose results for an FDC were 0.53 mSv/yr (53 mrem/yr) within 

10,000 years, and approximately 2.9 mSv/yr (290 mrem/yr) in 20,000 years. As described in 

Sections 2.6 and 2.13, NRC staff does not believe that there is a basis for assuming that there 

will not be significant fracturing, and therefore oxidation, of the saltstone until nearly 10,000 

years. Therefore, as in the case of the member of the public assessment, the NRC staff 

assumes that the Case K, K1, and K2 doses DOE reported as occurring between 10,000 and 

20,000 years could occur prior to 10,000 years. 

The Case K dose from Vault 4 DOE reported as being within 10,000 years was largely due to 

Cs-135 and the fish pathway. NRC staff expects that the reported fish consumption dose from 

Cs-135 is likely an artifact of the conservative assumptions DOE made in determining the 

concentration of radionuclides in the stream as well as in the amount of fish ingested, and the 

true dose from Vault 4 is likely to be less. The dose from an FDC reported by DOE as being 

within 10,000 years was primarily due to I-129. DOE did not report the contributions from 

individual radionuclides for the peak that they predict to occur after 10,000 years, but NRC staff 

expects that this peak is likely to be primarily due to Tc-99 since this radionuclide was 

responsible for most of the dose to the member of the public during that time period. Unlike 
206 

Vault 4, the predicted FDC dose does not have as significant of a contribution from the fish 

pathway, so this dose was not affected much by the conservative fish pathway assumptions. 

The NRC staff estimated the Tc-99 dose to an intruder for Case K and Case K1 from 

PORFLOWTM files provided by DOE. Based on DOE’s PORFLOWTM files30, the maximum 

estimated concentration of Tc-99 under an FDC in Case K was 1.16x10-4 mol/m3

(3.29x10-6 mol/ft3

) under Cell 11D (Figure 2.8-8). NRC staff estimated the all pathway dose 

conversion factor (DCF) for Tc-99 assuming the Case K biosphere parameters to be 

approximately 2.23x104 (mSv/yr)/(mol/m3) [7.89x107 (mrem/yr)/(mol/ft3)] based on the Case K 

member of the public peak dose of 0.55 mSv/yr (55 mrem/yr) and the Case K peak 100 m 

concentration of 2.46x10-5 mol/m3

 (6.97x10-7 mol/ft3)

31. Based on this all pathway DCF, the 

Case K Tc-99 intruder dose is approximately 2.6 mSv/yr (260 mrem/yr). The maximum 

estimated concentration of Tc-99 under an FDC in Case K1 was 1.89x10-4 mol/m3

(5.34x10-6 mol/ft3

)

32, which, using the same pathway DCF, would be equivalent to a dose of 

approximately 4.2 mSv/yr (420 mrem/yr). 

As noted above, Case K1 includes both conservative and non-conservative assumptions. As 

part of the evaluation of the member of the public dose, the NRC staff performed a series of 

evaluations to estimate the effect of the various assumptions (e.g., extent of fracturing, rate of 

fracturing) in Case K1 on the predicted dose to the member of the public (Section 2.13). 

Because the dose to a chronic intruder is attributable to groundwater pathways (i.e., rather than 

direct exposure) and because the doses from the groundwater pathway to the member of the 

public and the intruder are driven by the release of radionuclides from the saltstone, conclusions 

from these evaluations regarding the net effect of changing particular parameter values on the 

overall release and dose are also applicable to the intruder. Additionally, because Tc does not 

sorb strongly to soils and has a long half-life, it is not expected that there would be a significant 

difference between the concentration in the aquifer directly below a disposal unit and at a 100 m 

well. 

In the member of the public sensitivity cases evaluated by NRC staff based on the DOE 

PORFLOWTM model (Table 2.13-4), it was found that the dose increased by almost an order of 

magnitude over DOE’s reported Case K1 value of 0.9 mSv/yr (90 mrem/yr) when the model 

parameters were corrected for the disposal unit performance optimism. NRC staff expects that 

this high dose increase is largely due to an artifact of the release calculated using the averageKd approach (Section 2.13). Additionally, the dose calculated by NRC using DOE’s 

PORFLOWTM model was less than the original Case K1 dose of 0.9 mSv/yr (90 mrem/yr) when 

the model parameters were corrected for the disposal unit optimism and assuming a final 

fracture spacing of 1 m instead of 0.1 m. Similarly, as documented in Appendix B, the NRC 

staff calculated fractional release rates using a simple analytical dual-Kd model that did not 

include the disposal unit optimism and corrected for the average-Kd model conservatism. The 

calculated fractional release rates (Tables B-1 and B-2) were less than or equivalent to the 

Case K1 fractional release rate from the disposal unit (2.9x10-4 yr-1) (Table B-3) if the final 

 

30 PORFLOWTM files\AquiferZ_rev1\transport_original\CaseK_rev2_spacing\All\Tc-99\STAT_V2 (NRC, 2010g) 31 PORFLOWTM files\AquiferZ_rev1\transport_original\CaseK_rev2_spacing\All\Tc-99\STAT_V2 (NRC, 2010g) 32 PORFLOWTM files\AquiferZ_rev1\transport_original\CaseK_rev2_spacing_Kd\All\Tc-99\STAT_V2 (NRC, 2010g) 
207 

fracture spacing was 1 m or if the rate of fracturing was less sudden. However, the fractional 

release rates calculated with the simple analytical dual-Kd model were higher than the Case K1 

fractional release rate from the disposal unit by a factor of approximately 2 when the original 

fracturing assumptions were used (i.e., log based fracture growth and a final fracture spacing of 

0.1 m). The results of these sensitivity analyses indicate that when the model parameters are 

only corrected for the disposal unit optimism and the average-Kd conservatism, the overall 

release, and therefore dose, would be expected to increase by a factor of approximately 2. 

However, when a lower rate and extent of fracturing is also considered, the overall predicted 

release decreases from the Case K1 value. Therefore, if the actual rate or extent of fracturing is 

less than what was assumed by DOE (i.e., log based fracture growth and a fracture spacing of 

0.1 m at 10,000 years), then the actual rate of release, and resulting dose, will be less than the 

amount calculated using the original Case K1 model. 

As noted in Section 2.13, the available model support for the expected amount and rate of 

fracturing is limited. However, the NRC staff finds that the rate of fracturing and fracture 

spacing at 10,000 years might be less than what was assumed by DOE in Case K. Additionally, 

the dilution of the radionuclides in the aquifer is likely to be greater than DOE assumed in the 

intruder calculations. Based on these considerations, the NRC staff concludes that the net 

effect of the conservatisms and non-conservatisms in Case K1 would be to lower the 

groundwater pathways dose slightly from the Case K1 dose of approximately 4.2 mSv/yr 

(420 mrem/yr). The NRC staff therefore concludes that the dose to an inadvertent intruder 

through the groundwater pathways is likely to be less than 5 mSv/yr (500 mrem/yr). 

The NRC staff agrees that, in the short term, the drilling of a well directly through saltstone 

would be unlikely considering the physical properties of the disposal units, saltstone, and 

engineered cover as compared to the natural materials in the region. However, as these 

engineered systems degrade over time, these materials might not serve as strongly as an 

engineered barrier, and these materials might not be as distinguishable from the natural 

material. In the DOE drill cuttings scenario sensitivity analyses, the doses to acute and chronic 

intruders who intrude directly into saltstone after 500 years were estimated to be 

3.8x10-2 mSv/yr (3.8 mrem/yr) and 0.15 mSv/yr (15 mrem/yr) respectively. The dose prior to 

500 years could be significantly higher due to the presence of short-lived radionuclides, but 

intrusion during this time is not likely due to the expected performance of the cover and disposal 

units. Although the NRC staff thinks the dose from the chronic drill cuttings sensitivity case is 

underestimated because it is based on the groundwater concentration determined using Case 

A, rather than the larger groundwater concentrations in Case K or K1, the results of these 

sensitivity cases are useful in that they indicate that the additional dose an intruder would 

receive from being exposed to drill cuttings containing saltstone would not be significant 

compared to the dose to an intruder from using groundwater on site. The NRC staff therefore 

concludes that the dose to an intruder who drills directly into saltstone from the drill cuttings 

would be consistent with the performance objective for protection of an inadvertent intruder. 

As was true for the member of the public analysis, the dose results estimated for the intruder 

are a function of the inventory assumed in the analysis. Section 2.2.2 describes the monitoring 

of the inventory that NRC staff has performed and the monitoring that will be performed in the 
208 

future. In this monitoring, the NRC staff will track the actual inventory disposed in Vault 1, 

Vault 4, and the individual FDCs against the inventories listed in Tables 2.2-1 and 2.2-2. If the 

inventory in any of these disposal units is greater than the inventories listed in these tables, then 

an analysis will need to be performed to evaluate the potential dose consequences. In the case 

Chronic Intruder: FDC Case A Increased Concentrations, DOE evaluated the potential dose to 

an intruder from an individual FDC containing an increased inventory of radionuclides. 

However, this analysis was done using the Case A assumptions, which NRC staff does not 

believe are realistic (Section 2.1), so the NRC staff does not believe that this analysis 

adequately assesses the potential dose from an FDC containing an increased inventory. In the 

event that an individual FDC has an inventory that is greater than the Tables 2.2-1 and 2.2-2 

inventories, the dose consequences should be analyzed using a more realistic case, such as 

Case K, or a bounding case. 

Considering the projected intruder dose based on Case K, as well as the drill cuttings sensitivity 

analyses, the NRC staff concludes the dose to an inadvertent intruder is likely to be below 

5 mSv/yr (500 mrem/yr). NRC, therefore, has reasonable assurance that the disposal plans will 

satisfy the performance objective in §61.42. 
209 

4. Protection of Individuals during Operations 

4.1 Protection of Individuals during Operations 

The performance objective in §61.43, protection of individuals during operations, states the 

following: 

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 cross-references the standards for radiation protection in 

10 CFR 20. In DOE-WD-2005-001 (DOE’s 2005 waste determination for the saltstone facility), 

DOE provided a crosswalk of the relevant DOE regulation or limit consistent with that provided 

in 10 CFR 20 to demonstrate that the DOE regulation provides an equivalent level of protection. 

As stated in Section 2.5.2 of the PA, the performance objective in §61.43 is not addressed in the 

PA and DOE does not intend to address this subject in this or any future revisions of the PA. 

Measures taken in DOE-WD-2005-001 addressed this performance objective and annual 

reports on worker dose exposure and SRS environmental compliance reports account for a 

current and continual review of this section of the performance objectives. 

In cross-referencing the regulations in 10 CFR 20, DOE only considered sections of the 

regulation containing the dose limits for the public and the workers during disposal operations. 

These sections are the following: 

• §20.1101(d), 

• §20.1201(a, e), 

• §20.1208(a), and 

• §20.1301(a, b). 

The considered dose limits listed above correspond to the dose limits in 10 CFR 835 and 

relevant DOE Orders, which establish DOE regulatory and contractual requirements for DOE 

facilities and activities. 

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 (CBU-PIT-2005-00146). 
210 

Details of the measures taken by DOE to account for the list above can be found in 

CBU-PIT-2005-00146 and WSRC-IM-2004-00008. 

4.2 NRC Evaluation – Protection of Individuals during Operations 

In the 2005 TER (NRC, 2005), the NRC concluded that, during operations, individuals are 

protected by DOE regulations which were demonstrated to provide protection comparable to 

10 CFR 20, and thus concluded that there is reasonable assurance that the performance 

objective of §61.43 for protection of individuals during operations can be met. Since the NRC’s 

2005 TER (NRC, 2005), not much has changed in DOE’s demonstration of this performance 

objective. 

As stated in the NRC monitoring plan (NRC, 2007a), to verify that DOE’s radiation protection 

program is in place for operations at the SPF and the SDF to assess compliance with §61.43, 

protection of individuals during operations, NRC staff monitors activities at the SDF in the areas 

of worker dose monitoring, groundwater and air effluent monitoring, and other topics associated 

DOE’s radiation protection program at the SDF. Since monitoring began in 2007, the NRC has 

conducted two onsite observations of the SDF that have included discussions related to this 

performance objective. 

In October 2007 (NRC, 2008a), the NRC staff incorporated a review and discussion of two 

aspects of DOE’s Radiation Protection Program (RPP): (i) worker dose monitoring and 

(ii) groundwater and air effluent monitoring. In verifying DOE’s RPP was in place for operations 

at the SDF to assess compliance with §61.43, protection of individuals during operations, NRC 

staff (i) interviewed DOE’s contractor environmental monitoring personnel, (ii) reviewed records 

and radiological control documents associated with saltstone operations, and (iii) reviewed 

associated worker dose records. The staff also focused specifically on the 2007 groundwater 

monitoring program results for three of the groundwater monitoring wells installed downgradient 

of Vault 4, and the 2007 air-effluent monitoring program for the SDF. 

In March 2008 (NRC, 2008c), the NRC staff conducted another review of the DOE’s RPP at the 

SDF. During this observation, the staff again focused on the 2007 groundwater monitoring 

program results, however, this time considering eight (instead of three, as before) groundwater 

monitoring wells installed in or near the salt waste disposal area. NRC staff also focused on the 

2007 air effluent monitoring program for the SDF; reviewed soil-sample results in the vicinity of 

Vault 4; and, observed a groundwater sampling event from a background monitoring well. 

Following the review of this topic in both of the onsite observations, NRC staff found that there is 

no conclusive indication of groundwater contamination in the vicinity of Vault 4 resulting from 

salt waste disposal operations, but stated that it will continue to monitor groundwater data. NRC 

staff also found that the air effluent sampling results for Vault 4 during filling operations indicate 

that doses to nearby workers and members of the public from air effluents were well below DOE 

regulatory limits. NRC staff also learned that personnel from SC DHEC periodically collect 

sediment samples from a nearby sedimentation basin. Also following this observation, NRC 
211 

staff began including these independent data collected by SC DHEC as part of ongoing 

monitoring activities at the SDF. 

The conclusion of both of the onsite observations was that NRC staff determined that DOE has 

an adequate RPP in place for SDF operations. Therefore, the results of the NRC’s review in its 

2005 TER (NRC, 2005), multiple onsite observations, and NRC’s review of the annual SRS 

Environmental Reports provide reasonable assurance that the §61.43 performance objective 

will be met during facility operation. NRC will continue to assess DOE’s radiation protection 

program through future monitoring activities. 
212 

5. Site Stability 

5.1 Site Stability 

The performance objective in §61.44, stability of the disposal site after closure, states the 

following: 

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. 

In Section 2.5.2 of the PA, DOE states, “§61.43 for radiation protection during operations and 

§61.44 for site stability are not subjects that will use the PA as a technical basis in future closure 

documents and are therefore not discussed.” However, DOE has developed additional 

information related to site stability which includes the effects of earthquakes, floods, and 

settlement. As the stability of the disposal facility is also important to the performance 

objectives in §61.41 and §61.42, this section includes discussion of site stability with respect to 

the performance of the disposal units, waste form, and the closure cap. Further discussion on 

the effect of site stability is included in the respective sections for the disposal units 

(Section 2.5), waste form (Section 2.6), and closure cap (Section 2.4). 

The use of grout to stabilize salt waste at the SDF is designed to provide a monolithic structure, 

minimize void space, and prevent collapse. Following the closure of the disposal units, a 

closure cap will be installed over the SDF and a 100-year period of institutional controls will 

begin. During this period, active maintenance of the SDF will include prevention of pine forest 

succession and reparation of any significant erosion. No active SDF maintenance is assumed 

to be conducted beyond the institutional control period. The closure cap will be designed to 

provide a minimum of 3 m (10 ft) of clean material above the disposal units for the protection of 

inadvertent intruders. DOE’s “SRS End State Vision” includes ownership and control of the 

entire site by the federal government in perpetuity and prohibits residential use of the site 

(2009 PA). 

SRS is located on the Atlantic Coastal Plain within the Aiken Plateau. This region has a 

relatively low seismic activity (WSRC-IM-2004-00008). The largest known earthquake in the 

vicinity of the site occurred in Charleston, SC in 1886 with a magnitude of 7.3 on the Richter 

Scale (USGS, 2011). In addition, the SDF is sited on a well-drained topographic high 

(NRC, 2005). The 10,000 yr flood level for the Upper Three Runs basin near the SDF is 

approximately 48 m (157 ft) above mean sea-level (MSL) which is significantly below the lowest 

planned elevation of a disposal unit at the SDF, which is 79 m [260 ft] MSL (2009 PA). 

Much of SRS, including the SDF, is underlain by calcareous sediment that has resulted in the 

presence of under-consolidated “soft zones” in the Santee formation (WSRC-TR-99-4083). In 

conjunction with these soft zones, layers of hardened sediment are commonly observed. These 

layers have been characterized as bridges or arches in a honeycomb-like structure that acts to 
213 

redistribute stresses. Historically, some of the soft zones have consolidated, resulting in 

depressions on the land surface. These depressions, or “sinks”, at the SRS site typically are 3 

to 5 m [10 to 15 ft] deep and, in F Area, up to 300 m [1,000 ft] long (WSRC-TR-2007-00283). A 

stereoscopic examination of aerial photography of the SRS illustrated the potential presence of 

several basins and sinks located in and near the SDF (USACE, 1952; Figure A8). However, a 

subsequent investigation in 1986 stated that no evidence of ground subsidence was observed in 

the vicinity of where SDF is now located (87814-PT1). 

In 2006 and 2009, geotechnical investigations beneath FDC 2A/2B and 7A - D33 project sites 

indicated that the top of the Santee formation was approximately 24 m to 34 m Below Ground 

Surface (BGS) (80 ft to 110 ft BGS) with soft zones observed beneath the FDC 2A/2B site 

between approximately 30 m to 37 m BGS (100 ft to 120 ft BGS) (K-ESR-Z-00001; 

K-ESR-Z-00002). The sediment in this formation was described as being calcareous fine- to 

medium-grained sand with some clay and silt in addition to occasional fragments of shells and 

limestone. In response to NRC comment PA-14, DOE stated that the average thickness of the 

soft zones at the SDF is generally a few feet, although a soft zone approximately 4.3 m (14 ft) 

thick was identified beneath the FDC 2A/2B site (SRR-CWDA-2011-00044). DOE’s postulated 

maximum lateral dimension of these features is 3 m to 6 m (10 ft to 20 ft) 

(SRR-CWDA-2011-00044). The stability of these soft zones has been investigated since the 

1950s (USACE, 1952) with the understanding and treatment of these features evolving since 

preliminary evaluations (WSRC-TR-99-4083). Early approaches to stabilizing structures built in 

these regions included subsurface grouting, however it was determined that these grouting 

campaigns provided limited benefit in mitigating the potential settlement from the soft zones 

(WSRC-TR-99-4083). To evaluate the effect of consolidation of the subsurface layers beneath 

the SDF, DOE has conducted geotechnical and structural integrity analyses. 

DOE conducted a structural integrity analysis to evaluate the effects of static and dynamic 

settlement on Vault 4 under closure conditions, which included disposal unit loading and 4.0 m 

(13 ft) of closure cap material (T-CLC-Z-00006). Static settlement displacements were based 

on soil properties and actual settlement measurements from the Defense Waste Processing 

Facility (DWPF). DOE considers this data to be applicable to Vault 4 because of similarities in 

the soil profiles and the close proximity of the DWPF to the SDF. The maximum displacement 

for Vault 4 was calculated to be 19 cm (7.3 in). Dynamic settlement estimates relied on sitespecific geotechnical data which indicated that no soft zone settlement is expected because the 

Santee Formation is very dense beneath Vault 4. Settlement due to liquefaction from 

earthquakes with return periods of 2,500 yrs and 10,000 yrs ranged from 0.8 cm to 2.5 cm 

(0.25 in to 1.0 in) and 3.8 cm to 10.2 cm (1.5 in to 4.0 in), respectively. Mean dynamic 

settlement values of 1.9 cm (0.75 in) and 7.0 cm (2.75 in) were used in the structural analysis 

for earthquake return periods of 2,500 yrs and 10,000 yrs, respectively. The results of the 

analysis predicted disposal unit cracking to occur at construction joints, which are located on 

9 m (30 ft) centers in the base slab and walls. Cracking in the much weaker saltstone monolith 

was predicted to follow the joint-initiated disposal unit cracking. 

 

33 Disposal unit designations used in the TER correspond to designations used in the 2009 PA (Figure 2.8-8), which 

may not correspond to current designations. 
214 

Table 5.1-1: Summary of predicted static settlements for the FDC 2A/2B project site 

(adapted from K-ESR-Z-00001; Section 5.2.4)

Phase 

Minimum 

Settlement 

(in) 

Maximum 

Settlement 

(in) 

Average 

Settlement 

(in) 

Immediately after 

operations completed 2 17 4 

Immediately after closure 

cap completed 5 13 7 

30 years after closure 

cap completed 6 18 9 

2.5 cm are in one inch. 

DOE conducted geotechnical investigations for FDC 2A/2B and 7A – D (Figure 2.8-8) to 

evaluate the settlement of the subsurface layers due to static loading (i.e., the presence of 

disposal units, saltstone grout, and the closure cap) and the dynamic settlement due to potential 

liquefaction from earthquakes and the settlement due to the compression of the soft zones 

(K-ESR-Z-00001; K-ESR-Z-00002). Based on the similarities of the DWPF to the SDF, DOE 

estimated the total static settlement for the FDC project sites (Table 5.1-1 and 5.1-2). 

Differential settlement at the various phases was estimated to be of the same magnitude as the 

average uniform settlement in Table 5.1-2 (K-ESR-Z-00002). Estimated dynamic settlement for 

FDCs 2A/2B due to liquefaction from a design basis earthquake with a return period of 2500 yrs 

and a peak ground acceleration (PGA) of 0.21 g ranged from 0.0 cm to 5.8 cm (0.0 in to 2.3 in). 

The maximum surface settlement resulting from the compression of the 4.3 m (14 ft) thick soft 

zone was calculated to be approximately 1 cm (0.5 in) (K-ESR-Z-00001). The estimated 

dynamic settlement for FDCs 7A - D due to liquefaction from an earthquake with a PGA of 

0.20 g ranged from 0.8 cm to 5.1 cm (0.3 in to 2 in). Although no significant soft zones were 

identified beneath the FDCs 7A - D sites, the soft zone configuration from the FDC 2A/2B site 

was used in a conservative estimate of potential settlement from compression of soft zones, 

which similarly resulted in a maximum estimated surface settlement of 1 cm (0.5 in) 

(K-ESR-Z-00002). In response to NRC comment VP-1, DOE stated that seismic events and 

differential settlement are incorporated into the design calculations for the FDCs 

(SRR-CWDA-2010-00033). 
215 

Table 5.1-2: Summary of predicted static settlements for the FDC 7A – D34 project site 

(adapted from K-ESR-Z-00002; Table 5)

Phase 

Minimum 

Settlement 

(in) 

Maximum 

Settlement 

(in) 

Average 

Settlement 

(in) 

Immediately after 

operations completed 2 16 4 

Immediately after closure 

cap completed 3 13 9 

30 years after closure 

cap completed 5 18 11 

2.5 cm are in one inch. 

5.2 NRC Evaluation – Site Stability 

The NRC staff agrees that stabilizing salt waste with grout and completely filling the disposal 

units will provide a monolithic structure, minimize void space, and prevent collapse and 

differential settlement that could occur due to consolidation of the waste. Thus, the main 

processes evaluated with respect to site stability at the SDF are earthquakes, floods, erosion, 

and settlement. The NRC staff determined that the dynamic settlement expected to be caused 

by an earthquake with a 10,000 yr return period is unlikely to cause significant disruption to the 

SDF. Similarly, the staff concludes that floods are unlikely to disrupt the SDF because the 

10,000 yr flood level for the Upper Three Runs basin near the SDF is significantly below the 

lowest planned elevation of a disposal unit at the SDF. As discussed in Section 2.4, NRC staff 

concluded that the closure cap can provide adequate long-term erosion protection. 

With respect to settlement due to static loading, the NRC staff concluded DOE’s approach to the 

geotechnical investigations at the SDF documented in K-ESR-Z-00001 and K-ESR-Z-00002 is 

reasonable, however, it is not clear that the conclusions of these investigations are consistent 

with the assumptions in the PA. NRC staff requested additional information in IEC-5 regarding 

the assumption that settlement due to static loading will only be a few inches and will be 

uniformly distributed over the closure cap (NRC, 2010b; RAI-2009-01). In response, DOE 

stated that the maximum static settlement estimated for the F-Area closure cap is estimated to 

be 5.6 cm (2.2 in) at 10,000 yrs post closure (SRR-CWDA-2010-00033). As the estimated 

pressure from the planned F-Area closure cap is more than twice the pressure of the estimated 

SDF closure cap, DOE concluded that a range of 5 cm to 8 cm (2 in to 3 in) is conservative. In 

addition, DOE stated that the assumption of uniform settlement was supported by the differential 

settlement at FTF of less than 0.3 cm (0.1 in) for any given tank. However, the predicted 

settlement for FDCs 2A/2B and 7A – D (Figure 2.8-8) in SDF-specific calculations (Table 5.1-1 

and 5.1-2) exceed these values, and also exceed the values assumed in Saltstone Disposal 

Facility Closure Cap Concept and Infiltration Estimates (WSRC-STI-2008-00244). Because the 

 

34 DOE has changed the numbering of the FDCs since the PA was published. These FDCs are now referred to as 

2A/2B, 3A/3B, and 5A/5B.
216 

magnitude of settlement (uniform and differential) assumed in the performance assessment 

(Section 6.1 of WSRC-STI-2008-00244) was significantly less than the settlement predicted in 

recent investigations, the NRC staff finds that DOE has not shown whether settlement due to 

static loading may disrupt the performance of the synthetic materials and drainage layers. In 

addition, relevant American Concrete Institute (ACI) standard (ACI-372) for the FDCs suggests 

the maximum recommended limits for uniform and differential settlement are 15 cm (6 in) and 

5.3 cm (2.1 in), respectively, which are also exceeded by the predicted settlement for FDCs 

2A/2B and 7A – D. Because neither the standard nor the DOE analysis addresses the 

consequences of exceeding these criteria, the risk-significance of this finding is unclear. 

Accordingly, the NRC staff will monitor the development of information related to settlement of 

the SDF due to static loading. 

The NRC staff also requested additional information on the implications of calcareous material 

in the subsurface at SRS and how these features were addressed in the PA (NRC, 2010i; 

PA-14). With respect to site stability, DOE referenced a geotechnical evaluation report 

(K-ESR-Z-00002), which considered the effects of consolidation of the soft zones at the SDF 

(SRR-CWDA-2011-00044). However, DOE’s analyses of the potential settlement due to 

consolidation of these zones do not account for the potential removal of subsurface material 

which has resulted in subsidence observed at SRS. There is evidence of sinks near the SDF 

that are much more significant than DOE’s analysis of consolidation of the current soft zones 

suggests. It is not clear whether the process of sinkhole formation will continue. In response to 

NRC comment PA-14, DOE discussed the hypothesis that soft zone formation occurred by the 

dissolution of carbonate material by meteoric water when the Santee Formation was located in 

the vadose zone (SRR-CWDA-2011-00044). The potential for mass removal of carbonate 

material leading to subsidence within 100 years was dismissed due to the Santee formation 

currently being located beneath the water table, in a relatively stable geochemical environment 

(WSRC-TR-99-4083). Due to the location of the soft zones at the SDF (30 – 37 m BGS [100 – 

120 ft BGS]), the NRC staff does not expect the closure cap to result in the lowering of the water 

table to below the calcareous sediment, which would likely accelerate the dissolution of this 

material. However, evidence of dissolution based on elevated bicarbonate ion concentrations 

and relatively high pH values for groundwater samples collected in or near the Santee formation 

(WSRC-RP-92-450), demonstrates an ongoing evolution of the subsurface (USACE, 1952). 

Although dissolution of the calcareous sediment in the saturated zone is likely to be a very slow 

process, it has not been demonstrated that dissolution is insignificant with respect to site 

stability over the course of a 10,000 yr performance period. For example, subsidence beneath 

the SDF could result in fracturing in the saltstone grout and disposal units and increased 

localized infiltration because of run in from the surrounding area. Accordingly, the NRC staff will 

monitor the development of information related to the potential for sink development at the SDF. 

The NRC staff also will monitor the development of model support for the assumption that 

reasonably predicted future dissolution of calcareous sediment is insignificant to site 

performance. 

The disposal of salt waste as solidified saltstone in combination with the erosion controls DOE 

has included in its closure cap design provides reasonable assurance that salt waste disposal at 

the SDF will meet the site stability performance objective (§61.44). The use of saltstone grout 
217 

will limit void space in the disposal units and therefore the adverse effects of differential 

settlement of the waste form will be avoided (e.g., collapse of the disposal unit). However, the 

NRC staff has identified concerns related to site stability that could affect the long-term 

performance of the SDF with respect to §61.41. Based on the uncertainty in the potential risk 

associated with settlement due to static loading and the calcareous zones, these aspects of site 

stability will be included in NRC's revised monitoring plan. 
218 

6. Conclusions 

The NRC staff reviewed the 2009 Performance Assessment for the Saltstone Disposal Facility 

at the Savannah River Site, (2009 PA) dated October 2009 (SRR-CWDA-2009-000017), and 

associated documentation. The 2009 PA is an update to DOE’s January 31, 2006 PA 

performed in support of the Section 3116 Determination, Salt Waste Disposal, Savannah River 

Site (DOE-WD-2005-001 Rev. 0). In its December 2005 TER (NRC, 2005) documenting NRC’s 

review of the PA in support of DOE’s draft Section 3116 determination for the SDF 

(DOE-WD-2005-001 Rev. 1), the NRC staff concluded that it had reasonable assurance that salt 

waste disposal at the SDF would meet the performance objectives of 10 CFR 61 provided 

certain assumptions in DOE’s analyses were verified during monitoring. During the current 

review, the NRC staff carefully evaluated information related to these assumptions (i.e., 

information regarding saltstone oxidation, saltstone and disposal unit hydraulic conductivity, 

field-scale properties of as-emplaced saltstone, saltstone fracturing, numerical modeling of flow 

through fractures, radionuclide concentrations, moisture characteristic curves, and erosion 

control), as well as new factors of importance to the modified disposal plans and revised 

conceptual model. 

The NRC staff concludes it has reasonable assurance that waste disposal at the SDF meets the 

10 CFR 61 performance objectives for protection of individuals against intrusion (Chapter 3), 

protection of individuals during operations (Chapter 4), and site stability (Chapter 5). However, 

the staff no longer has reasonable assurance that DOE’s disposal activities at the SDF meet the 

performance objective for protection of the general population from releases of radioactivity35. 

Based on NRC’s review of the DOE PA and its own independent analyses, the NRC staff 

concludes that predicted doses in many of the scenarios the NRC staff considers reasonable fall 

within a range of approximately 0.25 mSv/yr (25 mrem/yr) [the limit established in the §61.41 

performance objective] to approximately 1 mSv/yr (100 mrem/yr) [the public dose limit found in 

§20.1301]. Thus, although the NRC staff cannot conclude that the performance objective in 

§61.41 is met, the potential dose to an off-site member of the public from DOE’s disposal 

actions is still expected to be relatively low. Furthermore, the staff expects that any exceedance 

of the §61.41 limit would occur long after site closure. The NRC staff has identified additional 

information DOE could develop and mitigative actions DOE could take in the short term that 

might provide reasonable assurance that salt waste disposal at the SDF meets the 10 CFR 61 

Subpart C performance objectives. This information and these actions are described as part of 

the Key Monitoring Factors NRC plans to use as a basis for further SDF monitoring (Section 6.2 

and Table A-1). As described in Section 6.3, the NRC staff will review information DOE 

provides during monitoring and determine if the information addresses the technical concerns 

identified by the staff in the TER. 

 

35 As discussed further in Section 2.13, NRC has evaluated compliance using a dose limit of 0.25 mSv/yr (25 mrem/yr) 

Total Effective Dose Equivalent (TEDE) consistent with the approach discussed in the Final Rule for Disposal of 

High-Level Radioactive Wastes in a Proposed Geologic Repository at Yucca Mountain, Nevada 66 FR 55752. DOE 

and NRC have used the updated internal dose factors of ICRP Publication 72 (ICRP, 1996), consistent with 

Commission direction in SRM-SECY-01-0148 (NRC, 2002). 
219 

6.1 Compliance with the Performance Objectives of 10 CFR Part 61, Subpart C 

6.1.1 Protection of the General Population from Releases of Radioactivity 

In the PA and in subsequent RAI responses (SRR-CWDA-2011-00044, PA-8), DOE indicated 

that it believes Case A is an appropriate base case because it reflects DOE’s expected site 

performance. Based on this deterministic Case A analysis DOE’s expected peak dose to a 

member of the general population from the SDF is 0.014 mSv/yr (1.4 mrem/yr) within 10,000 

years of site closure and 0.31 mSv/yr (3.1 mrem/yr) within 20,000 years of closure. As 

described in this TER (Sections 2.1 and 2.13), the NRC staff does not find Case A to be an 

appropriate case to compare to the performance objectives because it does not accurately 

reflect current site conditions, does not account for the full ranges of measured values of key 

parameters or expected differences between laboratory and as-emplaced values, and does not 

appropriately account for potential changes in parameter values with time. 

In response to NRC concerns about Case A, DOE supplied Cases K, K1, and K2. The main 

differences between these cases and Case A include (1) consideration of increased saltstone 

hydraulic conductivity, (2) the assumption of oxidation of saltstone proceeding from fractures for 

the purpose of modeling Tc release, (3) the assumption that the relative permeability in the 

saltstone and disposal unit concrete is always 1, (4) increased degradation of disposal unit 

concrete, (5) reduced inventory of Ra-226 and its ancestors Th-230, U-234 and Pu-238, and 

(6) updated biosphere modeling (Table 2-1). Cases K, K1, and K2 differ from one another only 

in the Kd values used to represent Tc sorption in oxidizing and reducing saltstone and disposal 

unit concrete. Of these three cases, the NRC staff concludes Case K1 uses the best-supported 

Kd values. The NRC staff disagrees with DOE’s characterization of Cases K, K1, and K2 as 

overly-pessimistic sensitivity analyses (Section 2.13.3.4). Because it resolves many of the 

concerns NRC identified with Case A, the NRC staff finds Case K1 is more appropriate for 

determining whether salt waste disposal at the SDF meets the performance objectives than 

Case A is. Therefore, the NRC staff has relied heavily on Case K1 in its review. 

DOE predicts Case K1 will lead to a dose to an off-site member of the general public of 

approximately 0.9 mSv/yr (90 mrem/yr) at approximately 12,900 years after site closure. 

However, the timing of this peak dose is very sensitive to, and delayed by the following factors: 

(1) the assumed timing and rate of saltstone fracture growth, 

(2) use of an average Kd value to track Tc release from saltstone instead of tracing 

release from oxidized and reduced regions separately, and 

(3) seemingly overly-optimistic assumptions about Tc re-reduction and reimmobilization in disposal unit floors and walls (see Sections 2.13.3.2 and 

2.13.3.3). 

Because of the large uncertainty in the predicted timing of the 0.9 mSv/yr (90 mrem/yr) peak 

dose from Tc-99 and the expectation that the predicted peak in DOE’s Case K1 model was 

delayed by unsupported assumptions, the NRC staff could not conclude it had reasonable 
220 

assurance that the dose would meet the 0.25 mSv/yr (25 mrem/yr) limit for 10,000 years after 

site closure based on DOE’s Case K1 results. 

Rather than base the compliance demonstration on DOE’s Case K1 alone, the NRC staff 

performed additional analyses to determine whether it could reduce the uncertainty in the timing 

or magnitude of the Case K1 peak dose. However, as explained in Section 2.13.3.3, NRC 

modifications of DOE’s Case K1 PORFLOWTM model that were expected to lower the predicted 

dose led to predicted Tc-99 peak doses greater than 0.25 mSv/yr (25 mrem/yr) at approximately 

10,000 years after site closure (Figure 2.13-1). Most of these predicted doses fell within a range 

of approximately 0.25 mSv/yr (25 mrem/yr) (the §61.41 performance objective) to approximately 

1 mSv/yr (100 mrem/yr) total effective dose equivalent (the public dose limit found in §20.1301. 

As described further in Section 2.13.3.3, NRC evaluations of a simplified dual-Kd release model, 

which the NRC staff developed to avoid an artifact of DOE’s average-Kd model that increases 

the dose in DOE’s Case K1,also led to annual fractional release rates for Tc-99 similar to DOE’s 

results (Appendix B). 

Because the staff expects that any exceedance of the §61.41 limit would occur long after site 

closure, the staff has not identified an immediate safety concern at the SDF. The NRC staff has 

identified additional information DOE could develop and mitigative actions DOE could take in 

the short term that might provide reasonable assurance that salt waste disposal at the SDF 

meets the 10 CFR 61 Subpart C performance objectives. The NRC staff has described this 

information in its Key Factors it plans to use as a basis for further SDF monitoring (Section 6.1.3 

and Table A-1). 

6.1.2 Protection of Individuals from Inadvertent Intrusion 

As described in Chapter 3, the NRC staff finds that the scenarios and pathways analyzed by 

DOE for the assessment of the performance objective in §61.42 (protection of an inadvertent 

intruder) are appropriate based on the regional practices near SRS. Specifically, DOE expects 

an individual is unlikely to inadvertently intrude into a disposal unit (e.g., by drilling a well) 

because the cementitious disposal unit and waste form, even at long times after closure, would 

be substantially different than and more difficult to disturb than the natural soil in the region. 

Therefore, DOE assumed an inadvertent intruder would drill a well near, rather than through, a 

disposal unit. In a sensitivity analysis, DOE also evaluated the dose to an individual who drills 

into a disposal unit and spreads drill cuttings on the land surface. The NRC staff finds this 

approach to be reasonable. 

DOE evaluated the doses to an acute and chronic intruder and found the chronic intruder dose 

to be greater. DOE estimated a peak chronic intruder dose within 10,000 years based on 

Case A to be 0.019 mSv/yr (1.9 mrem/yr) and a peak chronic intruder dose based on Case K 

within 20,000 years36 to be 2.9 mSv (290 mrem/yr). As explained in more detail in Chapter 3, 

the NRC staff estimated a peak chronic intruder dose based on DOE’s Case K1 to be 

 

36 As discussed in Section 2.13, because of the uncertainty in the timing of fracturing and the resulting uncertainty in 

radionuclide release from Cases K, K1, and K2, the NRC staff has determined that doses DOE predicts to occur 

within 20,000 years after site closure may occur within 10,000 years of closure. 
221 

approximately 4.2 mSv/yr (420 mrem/yr). Because this dose is dominated by groundwater 

pathways, it is subject to the same optimisms and pessimisms described for the Case K1 dose 

to an off-site member of the public (Section 2.13.3). Specifically, it appears to be artificially 

increased by DOE’s average-Kd model but unjustifiably diminished by unsupported credit for 

chemical retention of Tc in the disposal unit concrete. In Case K1, these factors appear to have 

similar effects on dose (Section 2.13.3.3). In addition to the assumptions affecting the dose to 

an off-site member of the public, the chronic intruder dose incorporates apparently conservative 

assumptions. Specifically, the NRC staff concludes that the Case K1 chronic intruder dose 

predicted by DOE is likely to be conservative (i.e., overestimate the potential intruder dose) for 

the following reasons: (1) the rate of fracturing and final fracture spacing may be less than 

assumed by DOE in Case K1, which would lower expected groundwater concentrations 

(Section 2.13.3.3) and, (2) radionuclides are likely to encounter more dilution in the aquifer than 

assumed by DOE in the intruder analysis. The NRC staff therefore concludes that the dose to 

an inadvertent intruder through the groundwater pathways is likely to be less than 5 mSv/yr 

(500 mrem/yr), provided that key assumptions in the Case K1 analysis are true. These key 

assumptions are identified as monitoring issues (Section 6.2 and Table A-1). 

In the DOE drill cuttings scenario sensitivity analyses, the doses to acute and chronic intruders 

who intrude directly into saltstone after 500 years were estimated to be 0.038 mSv/yr 

(3.8 mrem/yr) and 0.15 mSv/yr (15 mrem/yr) respectively. The dose prior to 500 years could be 

significantly higher due to the presence of short lived radionuclides, but intrusion during this time 

is not likely due to the expected performance of the erosion barrier layer in the closure cover 

that acts as an intruder deterrent and robust disposal unit roofs and clean cap layers. The 

results of these sensitivity cases indicate that the additional dose an intruder would receive from 

being exposed to drill cuttings containing saltstone would not be significant compared to the 

dose to an intruder from using groundwater on site, which the NRC staff expects to meet the 

5 mSv/yr (500 mrem/yr) dose limit specified in §61.42. Therefore, the NRC staff concludes that 

the dose to an intruder who drills directly into saltstone meets the performance objective for 

protection of an inadvertent intruder. 

6.1.3 Protection of Individuals during Operations 

In the 2005 TER (NRC, 2005), the NRC concluded that, during operations, individuals are 

protected by DOE regulations that provide protection comparable to 10 CFR 20. Since that 

time, the NRC staff has conducted two onsite observations of the SDF related to this 

performance objective. NRC staff reviewed aspects of DOE’s Radiation Protection Program 

(RPP) related to (1) worker dose monitoring and (2) groundwater and air effluent monitoring. 

The conclusion of both of the onsite observations was that DOE has an adequate RPP in place 

for SDF operations. Based on the results of the NRC’s review in its 2005 TER (NRC, 2005), 

onsite observations, and NRC’s review of the annual SRS Environmental Reports the NRC staff 

has reasonable assurance that DOE’s disposal actions at the SDF meet the §61.43 

performance objective for protection of individuals during operations. NRC will continue to 

assess DOE’s radiation protection program through future monitoring activities. 
222 

6.1.4 Stability of the Disposal Site after Closure 

The NRC Staff concludes the saltstone waste form will provide a monolithic structure, minimize 

void space, and prevent collapse and differential settlement that could occur due to 

consolidation of the waste. Thus, the main processes evaluated with respect to site stability at 

the SDF are earthquakes, floods, erosion, and settlement. NRC staff evaluated the dynamic 

settlement expected to be caused by an earthquake with a 10,000 year return period and found 

it was unlikely to cause significant disruption to the SDF (Section 5.1). Similarly, the staff 

determined that floods are unlikely to disrupt the SDF because the 10,000 year flood level for 

the Upper Three Runs basin near the SDF is significantly below the lowest planned elevation of 

a disposal unit at the SDF (Section 5.1). Although the design for the infiltration and erosion 

control cover will be made final closer to the time of site closure, the NRC staff concluded that, if 

implemented as preliminarily designed, the cap will include enough depth and a sufficiently 

robust erosion control layer to prevent exposure of the saltstone waste. 

As compared to the potential effects of earthquakes, floods, and erosion, there is more 

uncertainty in the potential effects of static settlement due to loading of the subsurface layers 

and settlement due to calcareous zones present or potentially developing under the SDF. With 

respect to static settlement, recent studies predict greater static settlement in the SDF than 

addressed by DOE in the PA (Chapter 5). Because updated modeling may be needed to 

account for the new information in the PA, this item has been identified as a monitoring factor. 

Much of SRS, including the SDF, is underlain by calcareous sediment in the Santee formation 

resulting in the presence of “soft zones.” Historically, some of these zones have consolidated, 

resulting in depressions on the land surface. An evaluation of aerial photography of the SRS 

illustrated the potential presence of several basins and sinks located in and near the SDF. 

Although the potential for mass removal of carbonate material leading to subsidence within 100 

years was dismissed due to the Santee formation being located beneath the water table, it is 

unclear whether there could be a change in the water table elevation leading to additional mass 

removal within 10,000 years of site closure. The potential for sinks to form in the future cannot 

be ruled out because of these site features: (2) the presence of calcareous material in the SDF 

subsurface, (2) groundwater samples indicating ongoing dissolution of calcareous material near 

the SDF, and (3) evidence of sinks in the surrounding area. 

To support its analysis of site stability, DOE noted that the average thickness of a soft zone at 

the SDF is generally a few feet, with a postulated maximum lateral dimension of about 3 m to 

6 m (10 ft to 20 ft). DOE concluded that consolidation of such a soft zone would cause a 

maximum differential settlement of approximately 1 cm (0.5 in), which DOE concluded would 

have minimal effects on the stability of a disposal unit (Chapter 5). In contrast, observed sinks 

at the SRS site are typically 3 m to 5 m (10 ft - 15 ft) deep and in F-Area, up to 300 m (1,000 ft) 

long. A sink of this magnitude, if it were to occur at the SDF, could result in fracturing in the 

saltstone grout and disposal units and increased localized infiltration because of run in from the 

surrounding area. Although sinks have been identified in the surrounding area, the time over 

which the sinks developed is unclear. Thus the likelihood that a sink could develop at the SDF 

within 10,000 years after disposal also is unclear. Because of the significant uncertainty in the 
223 

potential for settlement due to consolidation of soft zones under the SDF, the development of 

additional information about the potential effects of calcareous zones on site stability has been 

identified as a key factor for SDF monitoring (Section 6.2 and Table A-1). 

The disposal of salt waste as solidified saltstone in combination with the erosion controls DOE 

has included in its closure cap design provides reasonable assurance that salt waste disposal at 

the SDF will meet the site stability performance objective (§61.44). The use of saltstone grout 

will limit void space in the disposal units and therefore the adverse effects of differential 

settlement of the waste form will be avoided (e.g., collapse of the disposal unit). However, the 

NRC staff has identified concerns related to site stability that could affect the long-term 

performance of the SDF with respect to §61.41. Based on the uncertainty in the potential risk 

associated with settlement due to static loading and the calcareous zones, these aspects of site 

stability will be included in NRC's revised monitoring plan. 

6.2 Key Monitoring Factors 

Based on its review of DOE’s updated PA (SRR-CWDA-2009-000017) and associated 

documentation, as well as its independent analyses, the NRC staff has identified factors that are 

important to assessing whether DOE’s disposal actions meet the performance objectives of 

10 CFR 61, Subpart C. Many, though not all, of these factors are similar to factors identified in 

the NRC staff’s 2005 TER for salt waste disposal at the SDF (NRC, 2005), which the staff used 

as the basis of its current monitoring plan (NRC, 2007). Additionally, DOE has identified areas 

of ongoing and future work that are similar to many of these factors (PA Section 8.2 and 

SRR-CWDA-2011-00052). Key areas of interest continue to be waste form and disposal unit 

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

radiological inventory. In addition, additional monitoring factors related to site stability will be 

part of monitoring as a result of this review. The factors developed during this review will serve 

as the basis for a revised monitoring plan. A summary of the monitoring factors for the key 

areas of the PA is provided below. A more extensive list of the monitoring factors is provided in 

Appendix A. 

Inventory: NRC will monitor the inventory in Vault 1, Vault 4, 

and in each FDC in comparison to the values in 

Table 2-2 (with the exception of Ra-226 and 

Th-230). NRC will monitor the inventory of Ra-226 

and Th-230 as compared to the values in Table 2-3. 

NRC will also monitor the methods DOE uses to 

assess radionuclide inventories.

Infiltration and Erosion Control: NRC will monitor issues related to long-term closure 

cap performance, including any changes to the 

design or planned implementation of the closure 

cover, as its design develops and is made final 

closer to the time of site closure. NRC will monitor 

the development of model support for the 
224 

performance of the lower lateral drainage layer. 

NRC will monitor updates to DOE’s assessment of 

the effects of settlement on the long-term 

performance of the composite layer in the closure 

cap, the composite layers overlying the FDC roofs, 

and the lateral drainage layers. 

Waste Form Hydraulic Performance: NRC will monitor the hydraulic conductivity in fieldemplaced saltstone samples, the potential variability 

of field-emplaced saltstone properties and DOE’s 

quality assessment program for grout. NRC also will 

monitor the applicability of data measured using 

laboratory-produced samples to field-emplaced 

saltstone and the effect of the curing temperature 

profile on the hydraulic properties of field-emplaced 

saltstone. 

Waste Form Physical Degradation: NRC will monitor the development of model support 

for assumptions about saltstone fracturing. 

Waste Form Chemical Performance: NRC will monitor the radionuclide release from fieldemplaced saltstone, the chemical reduction of Tc in 

saltstone, the reducing capacity of saltstone and the 

expected evolution of redox conditions in saltstone. 

NRC also will monitor the sorption of Ra-226, Se-79, 

and Sr-90 in saltstone and the potential for an initial 

short-term release from saltstone (i.e., a rinserelease). 

Disposal Unit Performance: NRC will monitor the sorption of Tc-99, Ra-226, and 

Se-79 in disposal unit concrete. NRC will also 

monitor the development of model support for the 

long-term physical integrity of cementious materials 

forming most of the disposal unit structure as well as 

non-cementitious materials (e.g., epoxy, and 

neoprene seals) used in disposal unit joints. 

Subsurface Kd Values: NRC will monitor the Kd value assumed for Se-79 in 

sand and clay. 

Environmental Monitoring: NRC will monitor results from leak detectors 

installed under select FDCs, and NRC will continue 

to review groundwater monitoring data collected 

near the SDF. 
225 

Radiation Protection Program: NRC will continue to monitor annual worker dose 

reports and the results of DOE’s air monitoring 

program. 

Site Stability: NRC will monitor the development of model support 

for the potential settlement of the SDF due to the 

consolidation of subsurface layers and potential 

formation of sinks under the SDF. 

In addition to the monitoring factors, NRC staff identified a number of items for review when the 

PA is revised as part of DOE’s PA maintenance program (Table A-2). These items include 

aspects of model development (e.g., consideration of alternate conceptual models, evaluation of 

intermediate results), the diffusivity in saltstone, model support for the use of moisture 

characteristic curves, parameters related to the far-field flow model, and biosphere parameters. 

NRC staff considers these items to be potentially important to the predicted dose, though these 

items are not as risk-significant as the key monitoring factors. 

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, factors to be monitored 

may change as DOE implements its disposal plans. NRC staff will revise its plan to monitor 

DOE’s disposal actions at the SDF in coordination with SC DHEC. This revised plan will 

present the details of NRC’s planned future monitoring activities. 

6.3 Path Forward 

Section 3116(b) of the NDAA requires the NRC to monitor certain disposal actions taken by the 

DOE for the purpose of assessing compliance with the performance objectives set out in 

10 CFR Part 61, Subpart C. NRC monitoring activities of the Saltstone facility began following 

the 2007 completion of the U.S. Nuclear Regulatory Commission Plan for Monitoring the U.S. 

Department of Energy Salt Waste Disposal at the Savannah River Site in Accordance with the 

National Defense Authorization Act for Fiscal Year 2005, and have included a range of onsite 

observations, technical reviews, and meetings with DOE related to assessing DOE’s 

compliance with the 10 CFR Part 61, Subpart C performance objectives. The NDAA also 

requires that NRC report any noncompliance to Congress, the State, and the DOE as soon as 

practicable after discovery of the noncompliant conditions and states that NRC's monitoring is 

subject to judicial review. 

As stated in NUREG-1854, there are three primary reasons that DOE disposal actions could be 

found noncompliant: (1) if there are sufficient indications that the requirements of the 

performance objectives are currently not being met, (2) if there are sufficient indications that 

there is no longer reasonable assurance that the dose limits specified in the performance 

objectives will be met in the future, or (3) if key aspects relied upon to demonstrate compliance 

with one or more performance objectives are no longer supported due to the lack of supporting 

information obtained during the monitoring period. As documented and explained in the TER, 
226 

the NRC is not stating that releases have occurred from the disposal facility that could lead to 

annual doses that exceed the limits established in §61.41 (item 1 listed above). The NRC staff 

is, however, concerned that (i) information collected during the monitoring period (e.g., hydraulic 

conductivity assessments, Tc sorption measurements) does not support DOE’s compliance 

demonstration and (ii) sufficient information has not been provided to support many key 

modeling assumptions relied on for performance (items 2 and 3 above). 

Per the results of the NRC staff’s review, and in accordance with the reasons listed above for 

identifying reasons for noncompliance, the NRC is sending a letter of concern to both DOE and 

the SC DHEC so that SC DHEC is kept informed and has an opportunity to provide input and 

comments, and to provide DOE with an opportunity to furnish information that demonstrates its 

disposal actions are in compliance with the performance objectives. For example, DOE may 

present new or additional information or make design changes that would enable NRC to 

conclude with reasonable assurance that salt waste disposal at the SDF meets the 10 CFR 61 

performance objectives (Appendix A). If the staff determines that, based on the information 

provided, there is a sufficient basis to conclude that the performance objectives are met, NRC 

will send a notification of resolution letter. This letter and the potential resolution letter will be 

made publicly available on NRC's website as they formally document NRC’s concern and its 

resolution. 

If, after having reviewed any additional information received from DOE, the staff determines that 

it still cannot conclude there is reasonable assurance the performance objectives will be met, 

NRC will issue a noncompliance notification letter to the DOE in accordance with the NDAA. 

Also in accordance with the NDAA, the NRC is required to inform DOE, the covered State, and 

Congress if it considers any of DOE's waste disposal actions to be noncompliant with the 

performance objectives of 10 CFR 61, Subpart C. 
227 

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245 

CONTRIBUTORS 

NRC Staff 

Project Manager for Saltstone Monitoring and the Saltstone PA Review TER

Nishka J. Devaser 

Project Manager 

M.S., Nuclear Engineering, University of Tennessee 

B.S., Nuclear and Radiological Engineering, University of Tennessee 

Lead Technical Reviewer

A. Christianne Ridge 

Senior Systems Performance Analyst 

Ph.D., Environmental Engineering, University of California, Berkeley 

M.S., Environmental Engineering, Cornell University 

B.A., Physics, Drew University 

Technical Reviewers

George W. Alexander 

Systems Performance Analyst 

Ph.D., Energy & Geo-Environmental Engineering, The Pennsylvania State University 

M.S., Energy & Geo-Environmental Engineering, The Pennsylvania State University 

B.S., Geo-Environmental Engineering, The Pennsylvania State University 

Cynthia S. Barr 

Senior Systems Performance Analyst 

M.S., Environmental Systems Engineering, Clemson University 

B.S., Mathematics, College of Charleston 

B.A., Political Science (Pre-Law), College of Charleston 

Karen E. Pinkston 

Systems Performance Analyst 

Ph.D., Environmental Engineering, University of California, Berkeley 

M.S., Environmental Engineering, University of California, Berkeley 

B.S., Civil Engineering, Carnegie Mellon University 
246 

Center for Nuclear Waste Regulatory Analyses Staff 

Biswajit Dasgupta 

Center for Nuclear Waste Regulatory Analyses 

Staff Engineer 

Ph.D., Structural Engineering, University of Minnesota 

M.Tech., Structural Engineering, Indian Institute of Technology, Bombay, India 

B. Tech., Civil Engineering, Indian Institute of Technology, Kharagpur, India 

Patrick A. LaPlante 

Center for Nuclear Waste Regulatory Analyses 

Staff Scientist 

M.S., Biostatistics and Epidemiology, Georgetown University 

B.S., Environmental Studies, Western Washington University 

Roberto T. Pabalan 

Center for Nuclear Waste Regulatory Analyses 

Institute Scientist 

Ph.D., Geochemistry and Mineralogy, The Pennsylvania State University 

B.S., Geology, University of the Philippines 

David A. Pickett 

Center for Nuclear Waste Regulatory Analyses 

Senior Program Manager 

Ph.D., Geology, California Institute of Technology 

M.S., Geology, California Institute of Technology 

B.A., Geology, Rice University 

Pavan K. Shukla 

Center for Nuclear Waste Regulatory Analyses 

Senior Research Engineer 

Ph.D., Chemical Engineering, University of Florida, Gainesville, Florida 

M.S., Chemical Engineering, Indian Institute of Technology, Kanpur, India 

B.S., Chemical Engineering, Indian Institute of Technology, Kanpur, India 

Gary R. Walter 

Geosciences and Engineering Division 

Manager, Department of Earth, Material, and Planetary Sciences 

Ph.D., Hydrology, University of Arizona 

M.A., Geology, University of Missouri, Columbia
247 

Appendix A Monitoring Factors 
248 

Table A-1: Key Monitoring Factors 

Monitoring 

Topic 

Performance 

Objective Factor Basis 

Inventory §61.41 and 

§61.42 

NRC will monitor the inventory in Vault 1, Vault 4, and in each FDC in comparison to the values in Table 2.2-1, with the exception of Ra-226 and Th-230. NRC will monitor the inventory of Ra-226 and Th-230 as compared to the values in Table 2.2-2. 

The distribution of inventory among the disposal be important in determining potential doses to bosite member of the public and an inadvertent intruBecause a single FDC can dominate the dose to inadvertent intruder, and because the projected indose was a significant fraction of the 5 mSv/yr (500 mrem/yr) limit, each FDC inventory must be with PA assumptions (2009 PA) for NRC to retain reasonable assurance disposal meets the performobjective for intruder protection. Because a smalof FDCs can dominate the dose to an off-site methe public, the inventory in individual FDCs also ito protection of the general public. Although certdisposal units dominate the dose projected in DOupdated performance assessment, the location odose and degree of plume overlap can change wchanging assumptions about far-field transport. 

Inventory §61.41 and 

§61.42 

NRC will monitor the methods DOE uses to assess radionuclide inventories. 

Because of the considerable uncertainty in inventestimates and the importance of radionuclide invedose to an off-site member of the public as well ainadvertent intruder, NRC will monitor the methoduses to assess radionuclide inventories. NRC wiradionuclides that are currently identified as risk sas well as relevant ancestors (e.g., Tc-99, Ra-226I-129) but also will consider radionuclides that cobecome more risk significant if the inventory incresignificantly or if modeling assumptions change (eSe-79). In particular, NRC will monitor assumptiothe inventory of Th-230 in tanks that are known to Th-230 bearing waste and any methods used to ethe concentration of Th-230 when Th-230 is below detection limit. 
249 

Monitoring 

Topic 

Performance 

Objective Factor Basis 

Infiltration and 

Erosion Control

§61.41 and 

§61.42 

NRC will monitor updates to DOE’s assessment of the effects of settlement on the long-term performance of the composite layer in the closure cap and overlying the FDC roofs. 

Recent studies predict greater settlement in the Saddressed by DOE in the PA. Because settlemepotential to disrupt the composite layer in the closand the lateral drainage layers that play a key roldiverting water around the FDCs, NRC will monitDOE evaluations of the expected effects of settlecomposite layers in the closure cap and overlying roofs. 

Infiltration and 

Erosion Control

§61.41, 

§61.42, and 

§61.44 

NRC will monitor issues related to long-term closure cap performance. 

DOE expects the SDF closure cap to limit infiltraterosion for thousands of years after site closure. the cap design will be made final closer to the timclosure, NRC will monitor whether any design chaspecific aspects of the cap implementation (e.g., erosion barrier source material) are expected to apredicted cap performance assumed in the PA. Bclosure cap installation, additional information is nabout the potential effects of head build up above composite layer on slope stability, the potential foformation due to the cumulative effects of smallerfrequent flood events, and the predicted hydraulic conductivity of the foundation layer. Additional isbe identified as the cap design develops. 

Waste Form 

Hydraulic 

Performance 

§61.41 and 

§61.42 

NRC will monitor the measured values of hydraulic conductivity in field-emplaced saltstone samples, values from other representative studies, and the appropriateness of any new sampling technique. 

The hydraulic performance of field-emplaced saltcritical to adequate SDF performance. Recently measured saturated hydraulic conductivity values saltstone in a range greater than assumed in the Because of DOE concerns about the sampling teDOE is developing new methods to sample field-saltstone. This issue previously has been trackeof Open Issue 2007-1. 
250 

Monitoring 

Topic 

Performance 

Objective Factor Basis 

Waste Form 

Hydraulic 

Performance 

§61.41 and 

§61.42 

NRC will monitor the potential variability of asemplaced saltstone properties and DOE’s quality assessment program for grout. 

Variations in the composition of saltstone grout pthe SPF and emplaced in the disposal units (e.g.in the water-to-cementitious material ratio, aluminconcentration, presence of admixtures) may affecproperties of saltstone grout. Because of the sendose predictions to saltstone hydraulic propertiesassurance for the emplaced saltstone waste form important for building reasonable assurance in SDperformance. This factor includes both determinipotential variability in the hydraulic properties due variations in the saltstone composition and ensurfactors determined to significantly affect the hydraproperties are well controlled in the production of This factor is related to issues previously tracked Issue 2007-1 and Open Issue 2007-2. 

Waste Form 

Hydraulic 

Performance 

§61.41 and 

§61.42 

NRC will monitor the applicability of data on the hydraulic properties measured using laboratory-produced samples to ensure that it adequately reflects the properties of fieldemplaced saltstone. 

Because laboratory-produced samples have beea basis for the value of the saltstone saturated hyconductivity and diffusivity used in the PA, the apof the measurements made using laboratory-prodsamples to field conditions could affect dose predFor example, it is important to consider the potenof scale, temperature, the presence of admixturereducing conditions in the lab sample when estimproperties of as-emplaced saltstone from laboratoproduced samples. 
251 

Monitoring 

Topic 

Performance 

Objective Factor Basis 

Waste Form 

Hydraulic 

Performance 

§61.41 and 

§61.42 

NRC will monitor the effect of the curing temperature profile on the hydraulic properties of asemplaced saltstone. 

DOE studies have demonstrated the potential imof the curing temperature profile to the saturated conductivity of saltstone. Saltstone hydraulic conin turn, significantly affects dose predictions. It is important to understand the potential effect of curtemperature on the hydraulic properties and to vethe cure temperatures in the saltstone monolith wresult in a waste form that has higher hydraulic conductivities. 

Waste Form 

Physical 

Degradation 

§61.41 and 

§61.42 

NRC will monitor the development of model support for assumptions about saltstone fracturing.

DOE evaluations of alternative cases and NRC independent analyses indicate the sensitivity of dprojections to the assumed rate and extent of salfracturing because of effects on saltstone oxidatioas water flow through the disposal system. 

Waste Form 

Chemical 

Performance 

§61.41 and 

§61.42 

NRC will monitor whether measured radionuclide release from samples of field-emplaced saltstone are consistent with assumptions used in the PA. 

Because of the importance of radionuclide releasthe projected dose, it will be important to determithe release of radionuclides from field-emplaced is consistent with assumptions in the PA. This isspreviously has been tracked as part of Open IssuAdditionally, leaching experiments conducted to dreflected the bulk constituents of saltstone and siwaste, but have not included admixtures used in production of saltstone at the SPF. Because certadmixtures, such as such as the anti-foam agent Phosphate (TBP) may form chemical complexes radionuclides that limit radionuclide sorption or insolubility, it will be important to determine whetheradionuclide leaching from samples containing thadmixtures is consistent with assumptions in the 
252 

Monitoring 

Topic 

Performance 

Objective Factor Basis 

Waste Form 

Chemical 

Performance 

§61.41 and 

§61.42 

NRC will monitor the development of a robust demonstration that saltstone will chemically reduce Tc(VII) to Tc(IV) to the extent assumed in DOE’s PA model under the range of conditions to which saltstone is expected to be subjected during the compliance period. 

Studies DOE has relied on to demonstrate Tc retsaltstone have included experimental artifacts [i.epresence of sodium thiosulfate or H2(g)] that have 

difficult to interpret the results of these experimenon these experiments, it is unclear whether saltstcan reduce Tc and maintain Tc in a reduced stateFurthermore, recent DOE studies have shown the sensitivity of Tc retention in saltstone to trace quaoxygen. The peak dose to an off-site member of and the inadvertent intruder is sensitive to Tc relewhich in turn is sensitive to Tc redox state. For threasons, there must be a robust demonstration oof saltstone to reduce Tc and maintain Tc in a redstate in environmental conditions similar to the exenvironmental conditions of the emplaced waste be important to demonstrating there is reasonablassurance that SDF disposal meets the performaobjectives for protection of a member of the genepopulation. The reduction of Tc in saltstone prevbeen tracked as Open Issue 2009-1. 

Waste Form 

Chemical 

Performance 

§61.41 and 

§61.42 

NRC staff will monitor the development of additional information regarding the initial reducing capacity of saltstone as compared to the value assumed in the PA (i.e., 0.82 meq e-/g) 

and the expected evolution of redox conditions over time. 

The DOE Case K model and NRC sensitivity anademonstrate the importance of saltstone reducing to the projected Tc release rate. However, it is uthe measured value of the specific reducing capasaltstone, which contains only 25% blast furnace equivalent to the measured reducing capacity of furnace slag. Additionally, there is uncertainty in 

transition times assumed in the PA, which affects predicted release of redox sensitive radionuclides than Tc (because they were modeled with a porestep-change release model). 
253 

Monitoring 

Topic 

Performance 

Objective Factor Basis 

Waste Form 

Chemical 

Performance 

§61.41 and 

§61.42 

NRC will monitor measurements of Ra-226, Se-79, and Sr-90 sorption in or leaching from saltstone. 

Although the dose from Ra-226 dominates the doCase A and most of the DOE sensitivity analysesrelied on measured values of Sr sorption to estimsorption in oxidizing saltstone in the PA. If Ra-22continues to make a significant contribution to DOprojected does from the SDF, it will be important the uncertainty in Ra sorption by collecting elemeinformation about Ra sorption in and release from saltstone. Additionally, as discussed in Section 2.7, the NRC does not believe that there is an adequate basis f

values DOE uses to represent Se or Sr sorption isaltstone. Lower saltstone Kd values could result 

dose contributions to an off-site member of the puSe-79) or an inadvertent intruder (from Sr-90). 

Waste Form 

Chemical 

Performance 

§61.41 and 

§61.42 

The NRC staff will monitor the development of support for the assumption that shortterm rinse release of radionuclide’s from saltstone seen in laboratory experiments will not significantly affect projected peak doses from groundwater pathways at the SDF. 

Studies of Tc release from saltstone samples oftedemonstrate an initial relatively rapid release of Tcharacterized as a “rinse-release” phenomenon aexcluded from calculated release rates. If water excluded from the SDF for extended periods afteclosure, this rinse-release, if applicable to as-empsaltstone, would not occur until well after the shesystem is grouted and closed. 
254 

Monitoring 

Topic 

Performance 

Objective Factor Basis 

Disposal Unit 

Performance 

§61.41 and 

§61.42 

NRC will monitor the development of information about Ra and Se sorption in disposal unit concrete. 

Although the dose from Ra-226 dominates the prdose in Case A and most of the DOE sensitivity aDOE relied on measured values of Sr sorption as surrogate for Ra sorption in the PA. If Ra-226 comake a significant contribution to DOE’s projectefrom the SDF, it will be important to reduce the unin Ra sorption by collecting radionuclide-specific information about Ra sorption in disposal unit conAdditionally, the NRC staff does not believe that tadequate basis for the Kd value DOE used in the 

estimate Se sorption in disposal unit concrete. 

Disposal Unit 

Performance 

§61.41 and 

§61.42 

NRC will monitor the development of information about Tc sorption in disposal unit concrete. 

NRC analyses of intermediate results from DOE’s PORFLOWTM model demonstrate the importance retention in disposal unit concrete to the timing anmagnitude of the dose from fractured saltstone. 

Disposal Unit 

Performance 

§61.41 and 

§61.42 

NRC will monitor the development of model support for the performance of the lower lateral drainage layer, which depends on the performance of the disposal unit roofs and the HDPE/GCL layers above the FDCs. 

In most of DOE’s modeled cases, the lower lateralayer above each disposal unit diverts nearly all oinfiltrating water around the disposal units. This mdiversion of infiltrating water is due to the large dithe hydraulic conductivity between the lower laterdrainage layer and the disposal unit roofs (for Va4) and the HDPE/GCL layer (for the FDCs). Becincrease in the amount of infiltrating water will incamount of leaching from and the rate of degradatwaste form, an increase in the amount of infiltratiwould likely result in a higher dose. Therefore, thmonitor model support for the long-term performadisposal unit roofs and HDPE/GCL layers overlyiFDCs. 
255 

Monitoring 

Topic 

Performance 

Objective Factor Basis 

Disposal Unit 

Performance 

§61.41 and 

§61.42 

NRC will monitor the development of model support regarding the long-term fracturing of disposal unit concrete. 

NRC analyses of intermediate results from DOE’s PORFLOWTM model demonstrate the importance retention in disposal unit concrete to the timing anmagnitude of the dose from fractured saltstone. Radionuclides that flow through fast-pathways crfractures in the disposal unit floors or walls are noexpected to experience as much sorption as radimoving through an unfractured cementitious matr

Disposal Unit 

Performance 

§61.41 and 

§61.42 

NRC will monitor the development of model support for the long-term physical integrity of noncementitious materials (e.g., epoxy, and neoprene seals) used in disposal unit joints. 

There is uncertainty associated with the performanovel components in the design, as a result of a linformation on long-term performance of novel co(e.g., epoxy, and neoprene seals). DOE may neereview its experience base with these materials in facilities, or, for example, perform accelerated tesobtain long-term performance data. NRC analyses of intermediate results from DOE’s PORFLOWTM model demonstrate the importance retention in disposal unit concrete to the timing anmagnitude of the dose from fractured saltstone. radionuclides flow through fast-pathways created degradation of joint material instead of through thunit cementitious material, radionuclides may not effectively retained in the disposal unit concrete. 

Subsurface Kd

values 

§61.41 and 

§61.42 

NRC will monitor the Kd

value assumed for Se in sand and clay soils. 

As discussed in Section 2.7, the NRC staff does that the Kd value assumed in the PA for Se for sa

clay was adequately supported. 

Environmental 

Monitoring 

§61.41 and 

§61.42 

NRC will monitor results from leak detectors installed under select FDCs. 

DOE’s Consent Order of Dismissal with the SC Drequires DOE to install leak detection on Cell 3A fifth cell constructed thereafter. Monitoring informthese leak detectors will ensure NRC staff is awaearly hydraulic failure of the FDCs. 
256 

Monitoring 

Topic 

Performance 

Objective Factor Basis 

Environmental 

Monitoring 

§61.41 NRC will continue to 

review groundwater monitoring data collected near the SDF. 

Reviewing groundwater monitoring data will help NRC staff is aware of any early release of radionfrom saltstone. It may also provide staff with otheindicators of SDF performance, such as unexpecplumes of nitrate or increased alkalinity. 

Radiation 

Protection 

Program 

§61.43 NRC will continue to 

monitor annual worker dose reports. 

NRC has previously determined that DOE has an program in place to protect individuals during opeAs part of its monitoring responsibilities under the NRC will continue to monitor annual worker dose 

Radiation 

Protection 

Program 

§61.43 NRC will continue to 

monitor results of DOE’s air monitoring system. 

Releases in the air pathway during operations cacontribute to worker dose during operations. 

Site Stability §61.41, 

§61.42, and 

§61.44 

NRC will monitor the development of modeling support for the potential for settlement of the SDF due to consolidation of subsurface layers and the potential formation of sinks under the SDF. 

The potential for differential settlement due to conof soft zones or the formation of sinks under the Simportant to the assessment of site stability becaappears formation of a sink at the SDF could causignificant saltstone fracturing and disruption to thinfiltration controls and disposal units. 

Site Stability §61.41, 

§61.42, and 

§61.44 

NRC will monitor updates to DOE modeling of the effects of static settlement on site stability. 

Recent studies predict greater static settlement in than addressed by DOE in the PA. Updated modnecessary to support DOE’s conclusion that statisettlement will not adversely affect the performanclosure cap, disposal units, or saltstone grout. 
257 

Table A-2: PA Maintenance Items 

Topic Performance Objective Factor Basis 

Modeling §61.41 and 

§61.42 

NRC staff will continue to monitor any changes to the DOE PA for SDF, including the implementation of the conceptual model, consistency of intermediate model results with the conceptual model, quality assurance of models and codes used, and the appropriate use of probabilistic factors, when used. 

NRC’s review of the updated 2009 PA for as documented in this TER, indicates the importance of these PA factors in the NRC development of reasonable assurance thadisposal at the SDF meets the performanobjectives for protection of the general poand protection of individuals against inadvintrusion. 

Modeling §61.41 and 

§61.42 

NRC staff will monitor the defensibility of DOE’s conceptual models for releases of radionuclides from the SDF and potential exposures of offsite members of the public and potential inadvertent intruders. 

Conceptual model uncertainty is difficult tin dose models but can dominate the uncthe dose predictions. For example, an altconceptual model in which saltstone oxidilong period of time in which little or no wainto the waste and then is suddenly exposincreased water flow (e.g., though HDPE could generate a much larger peak dose more gradual failure. Because of the poteimportance of alternate conceptual modepredictions, NRC staff will monitor DOE’s consideration of alternate conceptual modfuture PA development. 
258 

Topic Performance Objective Factor Basis 

Waste Form §61.41 and 

§61.42 

NRC will monitor measurements of intrinsic diffusivity in degraded saltstone and model support for assumptions about intrinsic diffusivity in saltstone used in future PA revisions. 

DOE Case K results and NRC analyses demonstrate the sensitivity of the magnitutiming of the dose from Tc-99 to the rate osaltstone oxidation. In DOE Case K, the of an oxidation front is modeled as proceefunction of the square root of time, which progression of oxidation from older fractuHowever, other functional relationships arif saltstone degrades and the diffusivity inwith time. 

Waste Form §61.41 and 

§61.42 

NRC will monitor model support for Kd values used to represent 

sorption of radionuclides in saltstone in future PA revisions. 

Based on DOE and NRC sensitivity analyCase A and Case K, saltstone Kd values a

expected to affect predicted doses significBecause Kd values can affect which radio

are the primary dose contributors, the NRmonitor changes in Kd values in saltstone

NRC staff will monitor model support for sKd values for radionuclides that are risk-s

based on the Kd values and the uncertain

values. 
259 

Topic Performance Objective Factor Basis 

Near Field 

Flow and 

Transport 

§61.41 and 

§61.42 

NRC will monitor the development of model support for any moisture characteristic curves used to modify the permeability of saltstone or disposal unit concrete in a revised PA. 

DOE’s Case A and the sensitivity analysein the PA relied on moisture characteristic that reduced water flow through saltstoneand disposal unit concrete more than expbased on comparison to curves published similar materials. DOE’s Case K analysis take credit for decreasing permeability witdecreasing saturation (as described by mcharacteristic curves) and instead assumerelative permeability always is 1. Becaussensitivity of projected doses to the modepermeability of saltstone, the developmenappropriate model support for any moistucharacteristic curves used in a revised baexpected to be important to the developmreasonable assurance that waste disposathe performance objectives. 

Near Field 

Flow and 

Transport 

§61.41 and 

§61.42 

NRC will monitor model support for Kd values used to represent 

sorption of radionuclides in disposal unit concrete. 

As discussed in NRC’s barrier analysis (Section 2.13.3), the Kd values used in dis

concrete can have a significant effect on tmodeled release rate of radionuclides into field environment. Release rates, in turn, directly affect the predicted dose. 

Far-Field Flow §61.41 NRC will monitor the adequacy 

of DOE’s far-field model calibration. 

The calibration process could be improvemade more transparent, particularly in the near to the SDF, to provide confidence thlevel of dilution and dispersion in DOE’s Stransport models is not overstated. 
260 

Topic Performance Objective Factor Basis 

Far-Field Flow §61.41 NRC will monitor DOE’s source 

loading approach to ensure that the dose estimates are not significantly under-predicted. 

If DOE were to change how disposal unit are loaded into the far-field model to ensu(i) the same amount of mass is loaded intsaturated zone underneath each individuaand (ii) that source loading occurs at the win the far-field model, the results of the simcould be significantly different. Scoping sperformed by NRC staff using Tc-99 fluxeCase K indicate that the peak sector concat the 100 m boundary could be significanfor some sectors. 

Far-Field Flow §61.41 NRC will monitor the 

appropriateness of selected dispersivities and the need for additional vertical or horizontal mesh refinement to ensure that contaminant plumes are not artificially dispersed in the farfield model. 

DOE presented results in the FTF RAI res(SRR-CWDA-2009-00054) that indicate agrid refinement may be necessary to redunumerical dispersion in cases of very low assumed physical dispersion; this analysiexpected to be relevant to the SDF becaulocal FTF and SDF models have the samresolution. For example, if no physical disassumed, then the peak concentrations awith a pulse release of a conservative tracshown to be a factor of approximately thretimes higher with a grid refined by a factoeach dimension (or a factor of 8 times moelements). 
261 

Topic Performance Objective Factor Basis 

Far-Field Flow §61.41 NRC will monitor DOE’s efforts 

to collect data and information to evaluate the potential impact of calcareous zones on contaminant flow and transport at the SDF. 

Many SDF sources traverse the lower zonUTR aquifer where calcareous materials apervasive in the subsurface at SRS. Somevidence exists that contaminants in BuriaComplex (located in E-Area at the GSA) aChemical, Metals, and Pesticide Pits (locathe GSA) may be preferentially transportethese zones. In an FTF RAI resolution m(NRC, 2011x) DOE indicated that a field mactivity could be performed. If calcareous seeps are identified, tracer studies in the UTR-LZ using innocuous tracers that are used to understand preferential flow and tcould be conducted to better understand of these zones on contaminant flow and t

Far-Field Flow §61.41 NRC will monitor model support 

for Kd values used to represent 

sorption of radionuclides in site soil. 

Sorption in soil can significantly affect dosshort- to moderate- half-life radionuclides retained long enough to experience signifdecay. Thus NRC will monitor support fo

for risk-significant radionuclides or radionthat may become risk-significant based ochoice of Kd values. 
262 

Topic Performance Objective Factor Basis 

Biosphere §61.41 and 

§61.42 

NRC will monitor whether Kd

values for key radionuclides in surface soil are expected to significantly increase predicted radionuclide build-up in biosphere soil. 

Distribution coefficients used in the build-analysis were based on values selected conservatively in the context of hydrologic modeling (i.e., values were purposefully blow). This selection process is non-consewhen applied to irrigation and soil sorptiomodeling because lower sorption values cunderestimate radionuclide build-up. NRC scoping calculations show using site-spec

measurements could increase estimated by approximately a factor of 2 to 5. This iwould result in a higher predicted soil conand a higher projected dose from ingestioand animal products. 

Biosphere §61.41 and 

§61.42 

NRC will monitor consumption factors and uncertainty distributions for transfer factors. 

The assumed consumption of drinking wa2009 PA is approximately a factor of 2 lesEPA recommended assumption of 2 L/d asupported by site-specific information. Trfactors can have considerable uncertaintyshould be evaluated in any future probabmodel. 
263 

Appendix A: References

10 CFR 20, Federal Register, Standards for Protection against Radiation. Code of Federal 

Regulations, Office of the Federal Register. Washington, DC. January 1, 2010. 

10 CFR 61, Federal Register, Licensing Requirements for Land Disposal of Radioactive Waste. 

Code of Federal Regulations, Office of the Federal Register. Washington, DC. 

January 1, 2001. 

DOE-WD-2005-001 Rev. 1, DOE, Basis for Section 3116 Determination Salt Waste Disposal at 

the Savannah River Site. DOE. Washington, DC. January, 2006. ML102850319. 

SRR-CWDA-2009-00017, SRR, Performance Assessment for the Saltstone Disposal Facility at 

the Savannah River Site. Savannah River Site. Aiken, SC. October 29, 2009. ML101590008. 

SRR-CWDA-2010-00033, SRR, Comment Response Matrix for NRC - Requests for Additional 

Information on the Saltstone Disposal Facility Performance Assessment. Savannah River Site. 

Aiken, SC. July 23, 2010. ML102090664. 

SRR-CWDA-2011-00044 Rev. 1, SRR, Comment Response Matrix for NRC - Second Request 

for Additional Information on the Saltstone Disposal Facility Performance Assessment. 

Savannah River Site. Aiken, SC. August 26, 2011. ML113320303. 

SRR-CWDA-2011-00054 Rev. 1, SRR, Comment Response Matrix for NRC Comments on the 

Draft Basis for Section 3116 Determination and Associated SRR-CWDA-2009-00045, F-Tank 

Farm Waste Tank Inventory for use in Performance Assessment Modeling. Savannah River 

Site. Aiken, SC. December, 2009. ML100550169. 

U.S. Nuclear Regulatory Commission (NRC), Technical Evaluation Report for Draft Waste 

Determination for Salt Waste Disposal, Letter from L. Camper to C. Anderson, DOE. NRC. 

Washington, DC. December 28, 2005. ML053010225. 

U.S. Nuclear Regulatory Commission (NRC), U.S. Nuclear Regulatory Commission Plan for 

Monitoring the U.S. Department of Energy Salt Waste Disposal at the Savannah River Site in 

Accordance with the National Defense Authorization Act for Fiscal Year 2005. NRC. 

Washington, DC. May 3, 2007. ML070730363. 

U.S. Nuclear Regulatory Commission (NRC), Technical Evaluation Report for the Draft Waste 

Determination for the Savannah River Site F-Area Tank Farm, Letter from L. Camper to F. 

Marcinowski, DOE. NRC. Washington, DC. October 27, 2011. ML112371751. 
264 

Appendix B Results of NRC Staff Sensitivity Analyses Performed with a Simple 

Analytical Dual-Kd Model 

DOE and NRC observed that, compared to a model that tracks release from oxidized and 

reduced grout separately, the average-Kd model DOE used to estimate Tc release predicts less 

release at early times and more release when saltstone nears full oxidation 

(SRR-CWDA-2011-00114, TER Sections 2.7 and 2.13). Because of this behavior, the averageKd model appears to significantly over-estimate Tc peak fractional release rates from saltstone 

under certain conditions (Section 2.13). Because this apparent overestimate was difficult to 

avoid using the DOE PORFLOWTM model, the NRC staff could not conduct all of the necessary 

sensitivity analyses for Case K1 by modifying DOE’s PORFLOWTM model. Thus, the NRC staff 

developed a simple analytical dual-Kd model to conduct certain sensitivity analyses based on 

Case K1 (NRC, 2012a, b). These NRC sensitivity analyses were necessary in part because, 

unlike Cases A - E, DOE submitted Case K with only a limited deterministic sensitivity analysis 

(i.e., evaluation of certain Tc Kd values by comparison of Cases K, K1,and K2). As discussed in 

Section 2.13, the NRC staff did not use the results of the NRC model for comparison to the 

performance objectives. Rather, the NRC staff used the model to identify risk-significant 

assumptions and to develop an estimate of the uncertainty in the Case K1 value. 

Sensitivity and uncertainty were evaluated only with respect to source-term release. 

Specifically, the staff evaluated the effects of fracture spacing, fracture growth rate, Kd values for 

Tc in oxidizing and reducing saltstone, reducing capacity, and flow assumptions. Assumptions 

about water flow were evaluated with cases in which the flow thorough saltstone was assumed 

to be limited by the hydraulic conductivity of saltstone (Table B-1) and cases in which flow was 

based on the flow used in the DOE Case K1 PORFLOWTM model (Table B-2). Results were 

evaluated in terms of the peak annual fractional release rate of Tc from saltstone. The annual 

fractional release rate is the activity of Tc that leaves saltstone in a year divided by the initial Tc 

inventory. The annual fractional release rate is not a direct indication of dose, because releases 

of Tc from saltstone are attenuated by transport through the disposal unit concrete, site soils, 

and biosphere. However, in general, higher annual fractional release rates lead to higher doses 

to off-site members of the public37. 

Model sensitivities were evaluated by comparing the predicted peak fractional release rate in 

various test cases. For example, in general, cases with a fracture spacing of 1.0 m (3.3 ft) at 

10,000 years have lower release rates than cases with a fracture spacing of 10 cm (4 in) at 

10,000 years (i.e., compare tests 1 though 8 to tests 9 through 16 in Table B-1 and tests 17 

though 20 to tests 21 through 24 in Table B-2). Similarly, cases in which fractures develop as a 

logarithmic function of time have greater peak release rates than cases with the same flow and 

sorption parameters in which fractures develop as a quadratic function of time (e.g., compare 

test 1 with 3; test 2 with 4; or test 3 with 5 in Table B-1). 

 

37 Because the projected dose to an individual inadvertently intruding on the SDF is expected to be dominated by 

groundwater pathways instead of direct exposure (Chapter 3), higher fractional release rates also would lead to 

higher projected doses to an inadvertent intruder at the SDF. 
265 

Model sensitivities to flow were evaluated by comparing the effects of using different initial 

saltstone hydraulic conductivities (Table B-1) and by using the flow rates DOE used in the 

PORFLOWTM model in Case K1, which account for effects of the closure cap (Table B-2). In 

general, for the flow rates that were evaluated, cases with a constant flow have lower release 

rates than cases with increasing flow (if fracturing and sorption are not changed). This effect 

can be seen with two different types of comparisons. First, because the tests shown in 

Table B-1 all use a hydraulic conductivity of 1x10-7 cm/s at 10,000 years, cases with an initial 

hydraulic conductivity of 1x10-7 cm/s are modeled with a constant flow for the first 10,000 years 

after closure38. In contrast, cases with an initial hydraulic conductivity of 1x10-8 cm/s are 

modeled with an increasing flow. By comparing results for the first 10,000 years for cases that 

differ only in the initial flow rate (e.g., tests 1 and 2; tests 3 and 4; tests 5 and 6 in Table B-1), it 

is possible to compare otherwise equivalent cases that have either constant or increasing flow. 

Second, a similar effect is seen by comparing the constant flow cases (i.e., odd-numbered tests 

in Table B-1) with the results of cases based on the DOE Case K1 PORFLOWTM model 

(Table B-2). In each case, tests with comparable fracture growth, final fracture spacing, and 

sorption coefficients lead to a greater peak annual fractional release rate within 10,000 years if 

flow is based on DOE Case K1 PORFLOWTM model, which increases with time. The result that 

peak predicted release rate is greater if the initial hydraulic conductivity is lower may appear to 

be counterintuitive. However, this observation is consistent with the general trend that a low 

level of release at early times leads lower peak release rates than cases in which saltstone 

oxidizes during a period of low water flow and is subject to higher water flow only when 

saltstone is more significantly oxidized. While this discussion of observations is not exhaustive, 

it provides examples of how the NRC staff used the results of the simple analytical model. 

The model includes oxidation proceeding from fracture faces due to gas-phase transport of 

oxygen into the fractures, as well as matrix oxidation attributable to oxygenated water flowing 

through saltstone. In contrast, the DOE Case K, K1, and K2 models include oxidation 

proceeding from fracture faces, but not oxidation of the matrix from inflowing water. In the NRC 

simple analytical model, the fraction of oxidation proceeding from fracture faces was reduced to 

account for areas already oxidized by infiltrating water to avoid double-counting. The fraction of 

saltstone oxidized at each time step was used to predict release from the oxidized and reduced 

saltstone fractions separately. Tc release was based on the assumption that Tc comes to 

equilibrium within the oxidized and reduced regions separately based on the Kd for reduced or 

oxidized saltstone, as appropriate. Diffusion between the reduced and oxidized regions was 

neglected. The amount of Tc released from both the oxidized and reduced regions was tracked 

separately for each time step. In particular, inventory in the oxidized region in each time step 

was tracked so that the oxidized areas would cease to release Tc once the inventory in the 

oxidized volume was depleted). The inventory of Tc in the reduced region was tracked as well; 

however, the inventory in the reduced region typically was not exhausted until the inventory in 

the entire monolith was exhausted. Details of the analytical model are provided in a separate 

document (NRC, 2012b). 

 

38 In all cases in Table B-1, the modeled flow through saltstone increased from 1x10-7 cm/s to 1x10-6 cm/s from 

10,000 to 20,000 years after closure. 
266 

The simple analytical model estimates the release from saltstone, and does not account for 

attenuation (i.e., lowering of the peak release rate) in the disposal unit floor and walls. The 

amount of attenuation the disposal unit provides depends in part39 on the magnitude of the peak 

release from saltstone and on the Kd values used to model Tc transport in the disposal unit 

concrete. Comparison of peak fractional release rates from saltstone and from the disposal unit 

in DOE’s PORFLOWTM model for Cases K, K1, and K2 illustrates the attenuation in the disposal 

unit (Table B-3). For example, the effect of the magnitude of the peak release from saltstone is 

demonstrated by comparison of Cases K and K1. Although Case K uses larger disposal unit Kd

values for Tc than Case K1, (i.e., decreasing from 1,000 mL/g to approximately 420 mL/g over 

20,000 years in Case K and from 500 mL/g to approximately 210 mL/g in 20,000 years in Case 

K1), more attenuation occurs in K1 because it has a larger peak saltstone release rate. The 

effect of the disposal unit Kd values is seen by comparison of Case K1 and a modified version of 

K1 in which the Kd value used in the disposal unit concrete is changed to a value representative 

of oxidized concrete (i.e., 0.8 mL/g). Although both cases have the same peak saltstone 

release rate, there is more than an order of magnitude more attenuation in the disposal unit 

concrete in Case K1 than the modified Case K1 because of the difference in disposal unit Kd

values (i.e., factor of 22 attenuation in Case K1 and factor of 1.1 in the modified Case K1). In 

Cases K, K1, and K2, the peak release rate from the unsaturated soil is similar to the peak 

release from the disposal unit, indicating that little additional attenuation of Tc release takes 

place in the unsaturated soil. In the modified version of Case K1, little attenuation occurs in the 

disposal unit concrete but more attenuation is seen in the unsaturated soil. More attenuation 

occurs in the unsaturated zone in the modified version of Case K1 than in Cases K, K1, and K2, 

presumably because the modified version of Case K1 has a higher release rate from the 

disposal unit into the unsaturated soil than the other cases do. 

 

39 The amount of attenuation also depends in part on the shape of the peak (i.e., magnitude as a function of time). 
267 

Table B-1: Peak annual fractional release rates estimated with a simple analytical dual-Kd model (NRC, 

in which flow through saltstone is limited by the saltstone hydraulic conductivity 

Input Values1 Output Values2

Test 

Number 

Fracture 

Growth Rate 

(function of 

time) 

Hydraulic 

Conductivity at 

time of closure 

(cm/s) 

Sorption 

Coefficient 

(Kd) Reducing 

Saltstone (mL/g) 

Case K Saltstone 

Reducing capacity 

(0.206 meq e-/g) 

Ca

Re

(0

In 10,000 

years 

In 20,000 

years 

In 10

yea

Final Fracture Spacing of 10 cm (3.9 in)

1 Logarithmic 1x10-7 500 4.8x10-4 4.8x10-4 3.1x2 Logarithmic 1x10-8 500 5.6x10-4 5.6x10-4 3.1x3 Quadratic 1x10-7 500 2.7x10-4 2.7x10-4 2.9x4 Quadratic 1x10-8 500 2.8x10-4 2.8x10-4 2.9x5 Logarithmic 1x10-7 139 3.8x10-4 3.8x10-4 3.3x6 Logarithmic 1x10-8 139 5.1x10-4 5.1x10-4 3.3x7 Quadratic 1x10-7 139 2.4x10-4 2.4x10-4 2.9x8 Quadratic 1x10-8 139 2.8x10-4 2.8x10-4 2.9x

Final Fracture Spacing of 1.0 m (3.3 ft)

9 Logarithmic 1x10-7 500 8.3x10-5 1.6x10-4 4.1x

10 Logarithmic 1x10-8 500 8.4x10-5 1.7x10-4 4.1x11 Quadratic 1x10-7 500 6.7x10-5 1.6x10-4 4.2x12 Quadratic 1x10-8 500 6.8x10-5 1.7x10-4 4.1x13 Logarithmic 1x10-7 139 1.0x10-4 1.6x10-4 6.5x14 Logarithmic 1x10-8 139 1.0x10-4 1.9x10-4 6.6x15 Quadratic 1x10-7 139 8.3x10-5 1.5x10-4 6.5x16 Quadratic 1x10-8 139 8.6x10-5 1.9x10-4 6.3x

1All cases use an oxidizing saltstone Kd of 0.8 mL/g, a hydraulic conductivity of 1x10-7 cm/s at 10,000 years and 1x10-6 cm/

closure 

2Peak Annual Fractional Release (year-1) 
268 

Table B-2: Peak annual fractional release rates estimated with a simple analytical dual-Kd

model (NRC, 2012a, b) for cases in which flow is based on DOE PORFLOWTM Case K1 

flow through saltstone 

Input Values1 Output Values2

Test 

Number 

Fracture 

Growth Rate 

(function of 

time) 

Sorption 

Coefficient (Kd) 

Reducing 

Saltstone (mL/g) 

within 10,000 

years of closure 

within 20,000 

years of closure

Final Fracture Spacing of 10 cm (3.9 in)

17 Logarithmic 500 5.6x10-4 5.6x10-4

18 Quadratic 500 3.1x10-4 3.1x10-4

19 Logarithmic 139 5.6x10-4 5.6x10-4

20 Quadratic 139 3.1x10-4 3.1x10-4

Final Fracture Spacing of 1.0 m (3.3 ft)

21 Logarithmic 500 1.6x10-4 1.6x10-4

22 Quadratic 500 1.4x10-4 1.4x10-4

23 Logarithmic 139 2.2x10-4 2.2x10-4

24 Quadratic 139 2.0x10-4 2.0x10-4

1

All cases use an oxidizing saltstone Kd of 0.8 mL/g and Case K Saltstone Reducing capacity 

(0.206 meq e-

/g) 2

Peak Annual Fractional Release Rate (year-1) 

Table B-3: Peak annual fractional release rates predicted with DOE’s Case K, K1, and K2 

PORFLOWTM models and an NRC-modified version of Case K1 

Release Peak Annual Fractional Release Rate (year-1)

Case K Case K1 Case K2 Modified1 K1

from Saltstone 1.3x10-3 6.4x10-3 1.2x10-3 6.4x10-3

from Disposal Unit 1.8x10-4 2.9x10-4 2.6x10-4 5.7x10-3

from Unsaturated Zone (UZ) 1.8x10-4 2.9x10-4 2.6x10-4 3.0x10-3

Attenuation Ratio of Saltstone Release to Disposal Unit or 

UZ Release

In Disposal Unit 7.0 22 4.5 1.1 

In Disposal Unit and Unsaturated Zone 7.1 22 4.6 2.1 

data for Cases K, K1 and K2 taken from DOE STAT.out files [NRC, 2010] 

1

 Case identical to K1 except that the Kd for Tc in disposal unit concrete represents oxidized concrete 

conditions (i.e., 0.8 mL/g) 
269 

Appendix B: References

10 CFR 61, Federal Register, Licensing Requirements for Land Disposal of Radioactive Waste. 

Code of Federal Regulations, Office of the Federal Register. Washington, DC. 

January 1, 2001. 

SRR-CWDA-2011-00114, SRR, Comparison of a Single-Porosity Shrinking Core Model with a 

Dual-Porosity Fractured Medium Model. Savannah River Site. Aiken, SC. June 28, 2011. 

ML113320431. 

U.S. Nuclear Regulatory Commission (NRC), Index for Saltstone Performance Assessment 

PORFLOWTM and GoldSim®Model Files Provided by DOE in 2010. NRC. Washington, DC. 

November 2, 2010g. ML103060015. 

U.S. Nuclear Regulatory Commission (NRC), Review of Staff Release Rates Calculations for 

Saltstone. NRC. Washington, DC. April 2, 2012a. ML120880224. 

U.S. Nuclear Regulatory Commission (NRC), Description of NRC Staff Simple Analytical 

Average-Kd and Dual-Kd Models. NRC. Washington, DC. May 4, 2012b. ML121180114.