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DOE OSTI2023

A Fission-like Neutron Spectrum Shaping Assembly based on D-T Generator

For nearly five decades, 252 Cf sources have been used for testing and calibration over a wide range of neutron detection devices used for nuclear safety and radiation protection. However, the vastly increased cost of 252 Cf sources, its short half-life, and concerns about future shortage have prompted nuclear facilities and radiological calibration laboratories that rely on high-intensity neutron sources to seek alternatives. While there appear to be no other radionuclide neutron sources that would match the 252 Cf performance characteristics, all of them, including 252 Cf, present similar concerns in terms of hazards, safety, and security. As a possible solution to this issue, the Pacific Northwest National Laboratory (PNNL) developed a low-risk, cost-efficient alternative to produce fission-like spectra based on commercially available deuterium-tritium (D-T) neutron generators. The alternative of producing a fission-like neutron spectrum without any fissionable materials or beryllium would enable the nuclear facilities across the entire complex to eliminate their reliance on high intensity 252 Cf sources for nuclear safety and radiation protection applications, substantially reducing risks and associated costs. This report describes efforts of designing the spectrum shaping assembly, its construction and characterization of the neutron field produced. Performance of this newly developed D-T generator based neutron field is demonstrated via calibration of a criticality detector, neutron survey instruments, and personal dosimeters verified against current standardized fields produced by 252 Cf spontaneous fission sources.

42 ENGINEERING↗
DOE OSTI2023

Update of Receptor Locations for INL National Emissions Standard for Hazardous Air Pollutants (NESHAP) Assessments

This report documents the selection of updated public receptor locations for use in conducting regulatory assessments of radioactive air emissions for Idaho National Laboratory (INL) facilities. The basis for this receptor location analysis is found in 40 CFR 61 Subpart H, “National Emissions Standards for Emissions of Radionuclide Other than Radon from Department of Energy Facilities,” where the requirements for identifying National Emission Standards for Hazardous Air Pollutants (NESHAP) receptor locations are outlined. Receptor locations (residences, schools, businesses, or farm operations) were identified using high-resolution satellite imagery. This assessment identified 31 unique receptor locations for all INL Site facilities, and 30 receptor locations for INL in-town facilities that have the potential to emit radioactive materials. Identification of receptors for each in-town facility is new to this report.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2023

LLNL SFA OBER FY23 Program Management and Performance Report: BioGeoChemistry at Interfaces

The focus of the BioGeoChemistry at Interfaces SFA has been to identify and quantify the biogeochemical processes and the underlying mechanisms that control actinide mobility in an effort to reliably predict and control the cycling and migration of actinides in the environment. The research approach has included: (1) Field Studies that capture actinide behavior on the timescale of decades (Research Thrust 1), and (2) Fundamental Laboratory Studies that isolate specific biogeochemical processes observed in the field (Research Thrust 2). These Research Thrusts are underpinned by the unique capabilities and staff expertise at Lawrence Livermore National Laboratory, allowing the BioGeoChemistry at Interfaces SFA to advance our understanding of actinide migration behavior in the environment, and serve as a resource for environmental radiochemistry research internationally (Figure 1). In FY23, our SFA research focused on transient redox gradients across stratified waters, sediment-water interfaces, and mineral-water interfaces to address processes controlling cycling of redox-sensitive metals. Nevertheless, Research Thrusts 1 and 2 are guided by the following broad central hypotheses that were developed during our last program review held at the end of FY18: Thrust 1 Hypothesis: Biogeochemical processes occurring on the timescale of years to decades lead to greater actinide recalcitrance in sediments and limits their migration in surface and groundwater. Thrust 2 Hypothesis: Long-term biogeochemical processes include mineral and surface alteration, which leads to stabilization of actinide surface associations or incorporation into mineral precipitates. Our strategic goal is to use the knowledge gained from our Science Plan to advance our understanding of the behavior of actinides and other radionuclides (e.g. Cs) and provide DOE with the scientific basis for remediation and long-term stewardship of DOE’s legacy sites. More broadly, we are improving our understanding of transport phenomena in environmental systems sciences with a particular emphasis on environmentally relevant (long-term) timescales. While we retained some of our historical focus on actinide biogeochemistry this past fiscal year, biogeochemical processes occurring at unique Test Bed locations associated with this SFA provide fundamental information on abiotic and biotic redox processes that control the cycling of redox sensitive metals under dynamic and transient conditions. Furthermore, in collaboration with SFA teams at Argonne National Laboratory, our SFA has begun to transition away from the current research focus and develop a new research program in terrestrial wetland systems. The Terrestrial Wetland Function and Resilience SFA program plan will be delivered to the Environmental System Science (ESS) program within BER at the end of FY23.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
DOE OSTI2022

Incorporating Heterogeneity into 3D Geologic Models for the Alluvial Basin Reference Case

The unsaturated zone alluvium reference case is one of several geologic systems under consideration by the U.S. Department of Energy Office of Nuclear Energy for hosting repositories for spent nuclear fuel and associated waste (Sevougian et al., 2019). As noted by Mariner et al. (2018), the generic alluvial basin offers positive attributes that merit its consideration as a reference case by the Spent Fuel and Waste Science and Technology (SFWST) campaign. Hundreds of alluvial basins and sub-basins are scattered across the arid western United States (Figure 1-1). Precipitation and infiltration rates are relatively low with high evapotranspiration, resulting in deep regional water tables. Deep regional water tables allow for significant vertical separation between proposed repository depths and groundwater and thus longer transport paths from a hypothetical repository to an aquifer. Accumulations of alluvial sediments within these basins are typically on the order of hundreds of meters, and locally may exceed 1,000 m in thickness, as is the case for the Deming sub-basin in southern New Mexico. In alluvial basins, alluvium creates a thick geologic host medium that serves as the natural barrier system (NBS) in the conceptual model framework of a geologic disposal system and isolates the waste packages from receptors in the biosphere. Further, alluvial basin fill is typically comprised of stacked playa and lacustrine deposits along the basin axis (Perry et al., 2018). Characterized by low permeability, these layers protect the biosphere above the repository and the groundwater resources below the repository by limiting flow and transport of radionuclides.

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DOE OSTI2023

Harnessing Uncertainty through Functional Data Analysis in Gas Breakthrough Data

Detecting subsurface explosions from radionuclide gas migration through rock fractures is an effective way to identify nuclear activity. Los Alamos National Laboratory (LANL) has developed simulation methods, based on data from the 1962 Hardhat underground nuclear test, to predict gas breakthrough times at the surface. However, these methods rely on an imperfect understanding of the relationship between rock damage and fracture permeability. Our clinic project studies methods for predicting breakthrough curves that characterize total mass produced as a function of time, as well as quantifying the uncertainty associated with these predictions. The model that is currently employed to relate damage to permeability uses an empirically motivated power-law expression, with a range of parameter values that are compatible with the experimental Hardhat data. We develop emulators, built from functional data analysis techniques and trained on simulation data, that rapidly predict the gas breakthrough curve given a damage field and given the parameter values of the power-law equation. Using Bayesian regression, we address the problem of uncertainty quantification in our emulators. Finally, in order to test the robustness of the model, we further validate it on a damage field representing different physical conditions.

58 GEOSCIENCES↗
DOE OSTI2023

Argonne National Laboratory Site Environmental Report for Calendar Year 2022

This report discusses the status and the accomplishments of the environmental protection program at Argonne National Laboratory for calendar year 2022. The status of Argonne environmental protection activities with respect to compliance with the various laws and regulations is discussed, along with environmental management, sustainability efforts, environmental corrective actions, and habitat restoration. To evaluate the effects of Argonne operations on the environment, samples of environmental media collected on the site, at the site boundary, and off the Argonne site were analyzed and compared with applicable guidelines and standards. A variety of radionuclides were measured in air, surface water, groundwater, and bottom sediment samples. In addition, chemical constituents in surface water, groundwater, and wastewater were analyzed. External penetrating radiation doses were measured, and the potential for radiation exposure to off site population groups was estimated. Results are interpreted with respect to the origin of the radioactive and chemical substances (i.e., natural, Argonne, and other) and are compared with applicable standards intended to protect human health and the environment. A U.S. Department of Energy (DOE) dose calculation methodology, based on International Commission on Radiological Protection (ICRP) recommendations and the U.S. Environmental Protection Agency’s (EPA) CAP 88 Version 4.1 computer code, was used in preparing this report.

54 ENVIRONMENTAL SCIENCES↗
DOE OSTI2022

Updated Economic Model for Estimation of GDP Losses in the MACCS Offsite Consequence Analysis Code RDEIM Model Report for MACCS v4.2

This report updates the Regional Disruption Economic Impact Model (RDEIM) GDP-based model described in Bixler et al. (2020) used in the MACCS accident consequence analysis code. MACCS is the U.S. Nuclear Regulatory Commission (NRC) used to perform probabilistic health and economic consequence assessments for atmospheric releases of radionuclides. It is also used by international organizations, both reactor owners and regulators. It is intended and most commonly used for hypothetical accidents that could potentially occur in the future rather than to evaluate past accidents or to provide emergency response during an ongoing accident. It is designed to support probabilistic risk and consequence analyses and is used by the NRC, U.S. nuclear licensees, the Department of Energy, and international vendors, licensees, and regulators. The update of the RDEIM model in version 4.2 expresses the national recovery calculation explicitly, rather than implicitly as in the previous version. The calculation of the total national GDP losses remains unchanged. However, anticipated gains from recovery are now allocated across all the GDP loss types – direct, indirect, and induced – whereas in version 4.1, all recovery gains were accounted for in the indirect loss type. To achieve this, we’ve introduced new methodology to streamline and simplify the calculation of all types of losses and recovery. In addition, RDEIM includes other kinds of losses, including tangible wealth. This includes loss of tangible assets (e.g., depreciation) and accident expenditures (e.g., decontamination). This document describes the updated RDEIM economic model and provides examples of loss and recovery calculation, results analysis, and presentation. Changes to the tangible cost calculation and accident expenditures are described in section 2.2. The updates to the RDEIM input-output (I-O) model are not expected to affect the final benchmark results Bixler et al. (2020), as the RDEIM calculation for the total national GDP losses remains unchanged. The reader is referred to the MACCS revision history for other cost modelling changes since version 4.0 that may affect the benchmark. RDEIM has its roots in a code developed by Sandia National Laboratories for the Department of Homeland Security to estimate short-term losses from natural and manmade accidents, called the Regional Economic Accounting analysis tool (REAcct). This model was adapted and modified for MACCS. It is based on I-O theory, which is widely used in economic modeling. It accounts for direct losses to a disrupted region affected by an accident, indirect losses to the national economy due to disruption of the supply chain, and induced losses from reduced spending by displaced workers. RDEIM differs from REAcct in in its treatment and estimation of indirect loss multipliers, elimination of double-counting associated with inter-industry trade in the affected area, and that it is intended to be used for extended periods that can occur from a major nuclear reactor accident, such as the one that occurred at the Fukushima Daiichi site in Japan. Most input-output models do not account for economic adaptation and recovery, and in this regard RDEIM differs from its parent, REAcct, because it allows for a user-definable national recovery period. Implementation of a recovery period was one of several recommendations made by an independent peer review panel to ensure that RDEIM is state-of-practice. For this and several other reasons, RDEIM differs from REAcct.

29 ENERGY PLANNING, POLICY, AND ECONOMY↗
DOE OSTI2022

Characterization of Contaminants and Mobility in WMA A-AX Boreholes D0006 and D0008

This study was initiated to determine if subsurface contaminants in the A-AX vadose zone at the Hanford site (boreholes D0006 and D0008) caused corrosion of nearby stainless steel well casings (299-E-24-19, 299-E25-46, and 299-E25-236) and, based on the type and distribution of contaminants, determine if contaminants are from one or more sources including a) the A-104 and/or A-105 tanks, b) the 242A evaporator, and c) the 200-E-286 ditch. A tiered approach was used for characterizing contaminants in vadose zone sediments starting with mobile and total contaminant concentrations (Tier 1), characterizing contaminant speciation (Tier 2), and measuring contaminant leaching (Tier 3). Total and mobile contaminant measurements at seven depths showed elevated sulfate and nitrate in all samples (130 to 280 ft depth), elevated chloride in five depths, and no elevated radionuclides except I-129 in D0006 at 206 to 208 ft depth. If these contaminants were initially acidic, stainless-steel corrosion would occur. However, because sediments likely have a high acid neutralization capacity, an acidic spill would have been neutralized within tens of feet, so casing corrosion would likely occur in shallow sediments. GEL labs data also shows low (but above natural) chromate from shallow to 170 ft depth, then higher chromate to 280 ft, suggesting a surface Cr spill or casing corrosion. GEL labs data also showed evidence of more than one plume, with deep nitrate, sulfate, and chloride migration, but only shallow tritium migration. Tritium should migrate nearly unretarded in the subsurface like nitrate and chloride. Elevated aqueous cations and anions in the D0006 206 to 208' depth pore water (SO 4 2- , NO 3 - , Na + , K + , Ca 2+ , Mg 2+ , Si) could be the result of acid disposal and calcite dissolution at shallow depth, with deeper migration of SO 4 2- , NO 3 - from acids and Ca 2+ and Mg 2+ from calcite. Elevated Si (160 mg/L) indicates dissolution of clays or other silicates, which can also occur in acids. Acid extractable metals did show elevated Cr (but not Fe, Ni, Mo, and Mn) above natural levels, which may indicate stainless-steel well corrosion or transport of metals from a shallow spill. The elevated I-129 in D0006 at 206 to 208' depth was low with 0.34 μg/g in pore water and 1.5 μg/g total extractable I-129 in the sediment, which indicates significant fraction of I-129 was bound in one or more precipitates. Although total I-129 was below detection limits in most extractions due to high molybdate interference, I-127 analysis provided some insight into potential precipitate phases. Total I-127 was mainly extracted by acetate or acetic acid (67%), which may be iodine in calcite. Inorganic carbon measurement identified 1.65% calcite, and total organic carbon (TOC) was below detection limits, so iodine was not associated with organic matter. Nearly all the I-127 was present as iodide. In 1-D leach experiments, I-127 as iodide leached from the sediment within 2 pore volumes. Due to high and dynamic molybdate leaching from the sediment and interference with I-129 analysis, there were few quantifiable I-129 measurements. The calculated molybdate release rates from the sediment suggested Mo was being released from the same precipitate phase, such as molybdate incorporated into calcite. Overall, because there was little radioactive contamination measured in D0006 and D0008 boreholes, potential releases from A-104 and A-105 did not reach this location south of the A tank farm. Elevated cations and anions found in D0006 and D0008 at depth as the result of acidic disposal could cause shallow stainless steel casing corrosion. The nearby 242A evaporator generally contained alkaline waste and tritium, but the nearby 200-E-286 ditch generally contained high Cl - liquid effluent, so these are unlikely to be acidic sources.

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DOE OSTI2023

Characterization Results for the March 2023 Tank Farm 2H Evaporator Overhead Sample

On an annual basis, Savannah River Mission Completion (SRMC) provides 2H and 3H evaporator overhead samples to Savannah River National Lab (SRNL) to be analyzed per Section 5.2 of the Effluent Treatment Project (ETP) Waste Compliance Plan (WCP) and the Waste Acceptance Criteria (WAC). This report presents characterization results for the March 2023 2H evaporator overhead sample. The sample was clear and colorless with no visible solids. The results provide measurements for cesium-137 ( 137 Cs), strontium-90 ( 90 Sr), and iodine-129 ( 129 I) with the radionuclide concentration limits specified by the WAC.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2023

Cation Exchange Capacity, Anion Exchange Capacity, and Mineralogy of F-Area Aquifer Sediments

Strontium-90 (Sr-90) is a contaminant of concern in groundwater and surface water at both F-Area and H-Area Seepage Basins. This contaminant was disposed of, along with other heavy metals and radionuclides, into a series of unlined seepage basins from 1955 until 1988. The acidity of the wastewater increased Sr-90 mobility from the basin soil through the vadose zone and into the Upper Aquifer Zone (UAZ), creating a groundwater plume that discharges into wetlands areas and a local stream called Fourmile Branch.

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DOE OSTI2023

Preliminary Enhanced LAW Glass Formulation Algorithm

This report summarizes the Preliminary Enhanced Low-Activity Waste Glass Formulation Algorithm (GFA), its background information, and the calculations it performs. The Preliminary Enhanced Low-Activity Waste GFA is a tool developed in MATLAB to formulate glass for a given waste composition while attempting to maximize waste loading. It is intended for use at the Waste Treatment and Immobilization Plant, where nuclear waste will be vitrified into glass. The formulated glass is required to satisfy several processing and product quality constraints. In addition, calculations must account for associated uncertainties in constraint prediction and measurement. The GFA adheres to Pacific Northwest National Laboratory nuclear quality assurance procedures and has been validated and verified. In this second revision to the report, two constraints: the lithium and zirconia target constraints were amended to Table 3.1 and programmed into the algorithm. Additionally, three sections were added. Section 3.2.1 includes the Enhanced LAW Correlation Rules, an addition to the glass optimality criteria. Section 4.0 discusses a stepped process approach taken to ensure the algorithm can adequately react to differences between its recommended output and the actual measurements and transfers made during plant operation, and calculation of specific radionuclide concentrations. Section 5.2 describes changes to the user interaction with the algorithm program including the addition of a graphical user interface and compilation of the software.

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DOE OSTI2023

Solidification of REDC Organics For Disposal As Solid Waste

The Radiochemical Engineering Development Center (REDC) generates various liquid organic wastes from processing irradiated targets to recover heavy elements. In the past, these organics were discharged to the Oak Ridge National Laboratory (ORNL) liquid low-level waste (LLLW) system, along with the aqueous waste. Because of a reduction in aqueous LLLW from other ORNL generators and an increase in the radionuclide concentration, particularly 238 Pu, in the REDC organic waste stream, the organics can no longer be discharged to the LLLW system. The plan is to solidify the liquid organic waste for disposal as solid waste. Solidifying the organic liquids using PM-199 Organoclay ® would produce solid waste forms with no free liquid, which should qualify for disposal at the Waste Isolation Pilot Plant. Granular Organoclay can be added to the organic waste solutions until there is a dry layer on top of the liquid and then allowed to cure for a few hours to produce a solid wasteform.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2020

Closure Report for the Project Shoal Area: Subsurface Corrective Action Unit 447 Shoal,Nevada, Site

The U.S. Department of Energy (DOE) Office of Legacy Management (LM) has prepared this Closure Report for subsurface Corrective Action Unit (CAU) 447 at the Shoal, Nevada, Site, which was formerly called the Project Shoal Area. The Shoal site was the location of an underground nuclear test in 1963 that resulted in residual radionuclide contamination at the detonation depth of 1211 feet (ft) (DOE 2015a). The environmental restoration process and Corrective Action strategy are conducted in accordance with the Federal Facility Agreement and Consent Order (FFACO) (FFACO 1996, as amended) and all applicable Nevada Division of Environmental Protection (NDEP) statues and regulations. Responsibility for the environmental site restoration was transferred from the DOE, National Nuclear Security Administration, Nevada Field Office, to LM on October 1, 2006. The Shoal site is south of U.S. Highway 50, approximately 30 miles southeast of Fallon, in Churchill County, Nevada (Figure 1). The site was selected for underground nuclear testing in 1961, as part of the Vela Uniform Program. Vela was a research and development program directed toward locating, detecting, and identifying underground and high-altitude nuclear detonations. The Vela Uniform program was specific to underground detonations (AEC 1964). The objective of Project Shoal was to detonate a nuclear device underground in an active seismic area to improve the United States’ ability to detect, identify, and locate underground nuclear detonations. The underground nuclear test at Shoal was a joint effort between the U.S. Department of Defense (DOD) and the U.S. Atomic Energy Commission (AEC), which was the predecessor agency to DOE. Figure 1 is a map showing the site location.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

Non-LWR Fuel Cycle Scenarios for SCALE and MELCOR Modeling Capability Demonstration

To assess the modeling and simulation capabilities for radionuclide characterization, criticality, and shielding in the nuclear fuel cycle of non-light–water reactor (LWR) technologies, the US Nuclear Regulatory Commission (NRC) initiated a collaborative project between the NRC, Sandia National Laboratories (SNL), and Oak Ridge National Laboratory (ORNL) with the goal to demonstrate capabilities of MELCOR and SCALE to calculate accident scenarios in all stages of the nuclear fuel cycle for relevant non-LWRs. The first project task was to develop representative nuclear fuel cycles and identify potential hazards and accident scenarios in the individual fuel cycle stages based on publicly available information. Because the nuclear fuel cycle is not established for any non-LWR concept, many assumptions were made, and it is anticipated that the details of the fuel cycles will eventually look different.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

Chemical Characterization Results for Tank 40 Sludge Batch 10 Waste Acceptance Product Specifications (WAPS) Sample

A 3-L sample (HTF-40-23-24) obtained from Tank 40 following transfer of Tank 51 to achieve the final Sludge Batch 10 processing composition was received by Savannah River National Laboratory (SRNL). SRNL was tasked with characterization of the sample in accordance with requirements for reporting the Waste Acceptance Product Specifications. A representative subsample, approximately 500 mL, was prepared by continuously agitating the 3-L sample via a mixing blade inserted into the bottle. This subsample was then used for the chemical and physical characterization reported here. This characterization includes the following: (1) Supernate and slurry density; (2) Weight percent solids; (3) Mercury Analysis (supernate and slurry); (4) Anions; (5) Total Organic/Inorganic Carbon; (6) Free Hydroxide; (7) Other Base; (8) Elementals; (9) Volatile Organic Analysis and Semi-volatile Organic Analysis; and (10) Fissile Radionuclides and Uranium Enrichment. The results of this physical and chemical characterization of the sample are documented in this report.

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DOE OSTI2024

Corrective Action Decision Document/Corrective Action Plan for Corrective Action Units 101 and 102: Central and Western Pahute Mesa Nevada National Security Site, Nevada (Rev. 1)

This corrective action decision document (CADD)/corrective action plan (CAP) has been prepared for the Central and Western Pahute Mesa (PM) Corrective Action Units (CAUs) 101 and 102. Corrective action is needed because underground radioactive contamination exists in these areas resulting from historic underground nuclear testing. The PM CAUs are located in the northwestern portion of the Nevada National Security Site and comprise 82 corrective action sites. A total of 85 underground nuclear detonations took place within the area covered by the PM CAUs between 1965 and 1992 and resulted in the release of radionuclides (RNs) in the subsurface in the vicinity of the test cavities.

54 ENVIRONMENTAL SCIENCES↗
DOE OSTI2023

Design of Instrumentation for Noble Gas Transport in LSTL Needed for Model Development

Development of molten salt reactors (MSRs) requires consideration of several physical phenomena that are different from the light water reactor (LWR) experience. While small-scale tests have been performed to understand the behavior of molten salts, the scale-up to larger experiments is an important step in the development of MSR designs. These small-scale experiments have given information on the thermodynamic state of the heat transport fluid and its constituents, thermophysical properties, convective flow, and corrosion. As detailed models have been developed, using these models to simulate larger scale experiments must incorporate physics that becomes important during scale-up, temperature inhomogeneities, complex flow behavior, and two-phase flow. The purpose of the work package has been to link experimental and modeling efforts in the support of licensing development to design and evaluate technologies to mitigate radionuclide release from MSRs. There were two main activities in this work package—one investigating fission product transport into the cover gas and the other on fission gas transport in the molten salt.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2023

Sludge Batch 11 Assembly: Tank 35

Savannah River Mission Completion Nuclear Safety and Engineering Integration (SRMC-E) has requested that Savannah River National Laboratory (SRNL) perform Tank 35 characterization analyses in support of Sludge Batch 11 (SB11) assembly. Two Tank 35 samples were delivered to SRNL and composited into a single sample in April 2023. The composite sample was analyzed for the following: density, weight percent solids, chemical composition, radionuclides, supernate corrosion control tests, and x-ray diffraction for burkeite, gibbsite, and boehmite.

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