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DOE OSTI2020

DECOVALEX-2019 (Task B Final Report)

The DECOVALEX Project is an on-going international research collaboration, established in 1992, to advance the understanding and modeling of coupled Thermal (T), Hydrological (H), Mechanical (M) and Chemical (C) processes in geological in geological systems. DECOVALEX was initially motivated by the recognition that prediction of these coupled effects is an essential part of the performance and safety assessment of geologic disposal systems for radioactive waste and spent nuclear fuel. This document is the final report of Task B which was proposed and coordinated by the Swiss Federal Nuclear Safety Inspectorate (ENSI) and Lawrence Berkeley National Laboratory (LBNL), presenting the technical definitions of the problems studied, approaches applied, achievements made and outstanding issues for future research. Task B focused mainly on modelling of fault activation experiments (FS experiment) recently performed at the Mont Terri underground research laboratory, Switzerland. The FS experiment explores the coupling between fault reactivation in a clay host rock and the potential-enhanced fluid displacement through a previously low-permeability formation. Based on this experiment, the DECOVALEX-2019 Task B addresses how the change in permeability induced by the fault activation and the resulting fluid flow within the fault can be simulated including the dependence on its mechanical behavior. This will support the understanding of the processes during fault activation itself and it can help to determine consequences for contaminant transport due to potentially created permeable flow paths in otherwise low permeability argillaceous rocks.

58 GEOSCIENCES↗

DOE OSTI2020

DECOVALEX-2019 (Task E Final Report)

The DECOVALEX Project is an on-going international research collaboration, established in 1992, to advance the understanding and modeling of coupled Thermal (T), Hydrological (H), Mechanical (M) and Chemical (C) processes in geological in geological systems. DECOVALEX was initially motivated by the recognition that prediction of these coupled effects is an essential part of the performance and safety assessment of geologic disposal systems for radioactive waste and spent nuclear fuel. Later it was realized that these processes also play a critical role in other subsurface engineering activities, such as subsurface CO 2 storage, enhanced geothermal systems, and unconventional oil and gas production through hydraulic fracturing. Research teams from many countries (e.g., Canada, China, Czech Republic, Finland, France, Germany, Japan, Republic of Korea, Spain, Sweden, Switzerland, Taiwan, United Kingdom, and the United States) various institutions have participated in the DECOVALEX Project over the years, providing a wide range of perspectives and solutions to these complex problems. These institutions represent radioactive waste management organizations, national research institutes, regulatory agencies, universities, as well as industry and consulting groups. This document is the final report of Task E which was proposed and coordinated by Andra, the National Radioactive Waste Management Agency in France, presenting the technical definitions of the problems studied, approaches applied, achievements made and outstanding issues for future research. The purpose of Task E of the DECOVALEX-2019 project is to investigate upscaling THM modelling from small-scale experiments (some cubic meters) to full-scale experiments (some ten cubic meters) and finally to the scale of the waste repository (cubic kilometers). To achieve this aim, the data of two in-situ heating experiments performed by Andra (the French National Radioactive Waste Management Agency) in the Meuse/Haute-Marne Underground Research Laboratory (MHM URL) have formed the basis for the understanding of the THM behavior of the COx at different scales. The first experiment provided the reference values of the THM parameters by means of a calibration exercise and they were used for a blind prediction and an interpretative analysis of the second one.

58 GEOSCIENCES↗

DOE OSTI2022

Hanford Site Composite Analysis: LLBG-200-W B Vadose Zone Model

The objectives of the vadose modeling for the updated Hanford Site Composite Analysis (CA) are to simulate the flow and transport of water and radionuclide releases from the surface to the water table and to provide radionuclide transfer rates to the CA saturated zone model (CP-57037, Model Package Report: Plateau to River Groundwater Model, Version 8.3). Water additions include natural recharge and water discharged to the ground as a result of industrial processes associated with Hanford Site operations. Contaminant sources include radionuclides in water discharged to the ground during operations and radionuclides disposed “dry” in solid waste burial grounds or other means. The following 16 radionuclides were selected for this modeling effort: carbon-14 (C-14), chlorine-36 (Cl-36), tritium (H-3), iodine-129 (I-129), neptunium-237 (Np-237), rhenium-187 (Re-187), strontium-90 (Sr-90), technetium-99 (Tc-99), uranium-232 (U-232), uranium-233 (U-233), uranium-234 (U-234), uranium-235 (U-235), uranium-236 (U-236), uranium-238 (U-238), radium-226 (Ra-226), and thorium-230 (Th-230). The simulation time starts in 1943 and ends at 12070, which is 10,000 years after assumed Hanford Site closure in 2070. The parallel version of the Subsurface Transport Over Multiple Phases (STOMP 1 ) simulator, officially named the exascale Subsurface Transport Over Multiple Phases (eSTOMP), is used to simulate flow and transport for the vadose models. The documentation for the STOMP code is comprehensive. The theoretical and numerical approaches applied in the STOMP code are documented in a published theory guide (PNNL-12030, STOMP Subsurface Transport Over Multiple Phases Version 2.0 Theory Guide). The code has undergone a rigorous verification procedure against analytical solutions, laboratory-scale experiments, and field-scale demonstrations. The application guide (PNNL-11216, STOMP Subsurface Transport Over Multiple Phases Application Guide) provides instructive examples in the application of the code to classical groundwater problems. The user’s guide (PNNL-15782, STOMP: Subsurface Transport Over Multiple Phases Version 4.0: User’s Guide) describes the general use, input file formatting, compilation, and execution of the code. The primary output of the vadose zone modeling is radionuclide transfer rates to the groundwater for input into the saturated zone model. The rates will be summed over the 100 by 100 m saturated zone model grid cells that fall within the vadose zone model source domain.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2022

Exceedance Response Action (ERA) Level 2 (Technical Report)

54 ENVIRONMENTAL SCIENCES↗

DOE OSTI2022

Vadose Zone Model for B-63 Area for Composite Analysis (Rev. 1)

The objectives of the vadose modeling for the updated Hanford Site composite analysis (CA) are to simulate the flow and transport of water and radionuclide releases from the surface to the water table and to provide radionuclide transfer rates for the plateau to river (P2R) model, version 8.3 (CP-57037, Model Package Report: Plateau to River Groundwater Model, Version 8.3). Water additions include natural recharge and water discharged to the ground as a result of industrial processes associated with Hanford Site operations. Contaminant sources include radionuclides in water discharged to the ground during operations and radionuclides disposed “dry” in solid waste burial grounds or other means. The following 16 radionuclides were selected for this modeling effort; carbon-14 (C-14), chlorine-36 (Cl-36), tritium (H-3), iodine-129 (I-129), neptunium-237 (Np-237), rhenium-187 (Re-187), strontium-90 (Sr-90), technetium-99 (Tc-99), uranium-232 (U-232), uranium-233 (U-233), uranium-234 (U-234), uranium-235 (U-235), uranium-236 (U-236), uranium-238 (U-238), radium-226 (Ra-226), and thorium-230 (Th-230). The simulation time starts in 1943 and ends at 12070, which is 10,000 years after assumed Hanford Site closure in 2070.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2022

Vadose Zone Model for LLBG-200W A for Composite Analysis

The objectives of the vadose modeling for the updated Hanford Site composite analysis (CA) are to simulate the flow and transport of water and radionuclide releases from the surface to the water table and to provide radionuclide transfer rates for the plateau to river (P2R) model, version 8.3 (CP-57037, Model Package Report: Plateau to River Groundwater Model, Version 8.3). Water additions include natural recharge and water discharged to the ground as a result of industrial processes associated with Hanford Site operations. Contaminant sources include radionuclides in water discharged to the ground during operations and radionuclides disposed “dry” in solid waste burial grounds or other means. The following 16 radionuclides were selected for this modeling effort: carbon-14 (C-14), chlorine-36 (Cl-36), tritium (H-3), iodine-129 (I-129), neptunium-237 (Np-237), rhenium-187 (Re-187), strontium-90 (Sr-90), technetium-99 (Tc-99), uranium-232 (U-232), uranium-233 (U-233), uranium-234 (U-234), uranium-235 (U-235), uranium-236 (U-236), uranium-238 (U-238), radium-226 (Ra-226), and thorium-230 (Th-230). The simulation time starts in 1943 and ends at 12070, which is 10,000 years after assumed Hanford Site closure in 2070.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2022

Vadose Zone Model for LLBG-200W A for Composite Analysis

The objectives of the vadose modeling for the updated Hanford Site composite analysis (CA) are to simulate the flow and transport of water and radionuclide releases from the surface to the water table and to provide radionuclide transfer rates for the plateau to river (P2R) model, version 8.3 (CP-57037, Model Package Report: Plateau to River Groundwater Model, Version 8.3). Water additions include natural recharge and water discharged to the ground as a results of industrial processes associated with Hanford Site operations. Contaminant sources include radionuclides in water discharged to the ground during operations and radionuclides disposed "dry" in solid waste burial grounds or other means. The following 16 radionuclides were selected for this modeling effort: carbon-14 (C-14), chlorine-36 (Cl-36), tritium (H-3), iodine-129 (I-129), neptunium-237 (Np-237), rhenium-187 (Re-187), strontium-90 (Sr-90), technetium-99 (tc-99), uranium-232 (U-232), uranium-233 (U-233), uranium-234 (U-234), uranium-235 (U-235), uranium-236 (U-236), uranium-238 (U-238), radium-226 (Ra-226), and thorium-230 (Th-230). The simulation time starts in 1943 and ends at 12070, which is 10,000 years after assumed Hanford Site closure in 2070.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2023

Packaging Capacity Calculation: Pu Oxide Packaging Options with the 2 Quart SAVY Container

This calculation supports programmatic efforts to Dilute and Dispose Pu. Metal is to be oxidized at LANL, then transported to SRS for dilution at SRS. Ultimately, the diluted oxide will be disposed of at the Waste Isolation Pilot Plant (WIPP). Currently 3013 containers are used to package the oxide, which are placed in to 9975 Type B shipping containers. The Oxide Packaging Technology Maturation Plan determined that a transition to the used of 2-qt SAVY containers placed in 9977 would increase efficiency in shipping and handling and increase programmatic flexibility. The 2-quart SAVY container meets DOE M 441.1 and TA-55 Documented Safety Analysis requirements for handling and storage of Pu oxide at LANL.

36 MATERIALS SCIENCE↗

DOE OSTI2022

Vadose Zone Model for B-63 Area for Composite Analysis (Rev. 2)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2023

Shear Assisted Processing and Extrusion (ShAPE) of Plastics: Recycling and Remolding

Polyethylene (PE) and polypropylene (PP) are often disposed as mixed plastic wastes. The challenges in recycling and upcycling these mixed polyolefin wastes lie in the difficulty in separating individual constituents in a cost-effective and scalable manner. Direct recycling the mixed PE and PP wastes in conventional melt-phase extruders typically result in a product with poor properties and low added value, because of immiscibility, phase separation, and lack of crystallinity. Friction extrusion (FE), a solid phase processing technique that has successfully extruded metal matrix composites with desired end products, has never been utilized to address the issue of recycling mixed plastic wastes. In this study, FE was performed on single-stream low-density polyethylene (LDPE), single-stream PP and mixed-stream LDPE+PP. Consolidated filaments of 2.5 mm diameter were extruded from different precursors. The thermal, infrared, and microscopic properties of extruded filaments were measured to evaluate the effects of FE process on the structure of recycled polymers. Meanwhile, the energy efficiency of FE was estimated based on extrusion rates and compared with conventional melt extrusion processes.

32 ENERGY CONSERVATION, CONSUMPTION, AND UTILIZATI↗

DOE OSTI2023

Exceedance Response Action (ERA) Level 2 Technical Report Update

The Sandia National Laboratories, California (SNL/CA) site comprises approximately 410 acres and is located in the eastern portion of Livermore, Alameda County, California. The property is owned by the United States Department of Energy/National Nuclear Security Administration (DOE/NNSA) and is being managed and operated by National Technology & Engineering Solutions of Sandia, LLC (NTESS). Operations at the SNL/CA facility consist of DOE statutory responsibilities for nuclear weapon research and design, development of energy technologies, and basic scientific research. Specific industrial activities occur in discrete buildings and include electroplating or anodizing, machine shop, permitted hazardous waste treatment, storage and disposal facility (TSDF), and a scrap yard. This report provides an update to the ERA Level 2 Technical Report Update previously submitted in June 2022 (2022 ERA L2 Technical Report Update) for iron, aluminum, and pH. In accordance with the California’s General Permit for Discharges of Storm Water Associated with Industrial Activities (IGP) Section XII.D.3.c, an ERA Level 2 Technical Report update is to be submitted with the annual report when a facility utilizing an Industrial Activity Best Management Practice (BMP) Demonstration (IABMPD) experiences additional exceedances of the numeric action levels.

54 ENVIRONMENTAL SCIENCES↗

DOE OSTI2025

Nuclear Material Accounting and Control (NMAC) for a Commercial Fuel Debundling Facility in the United States

The concept of fuel debundling proposes that all non-fuel components of a fuel assembly be removed prior to beginning the recycling process. It was introduced to reduce the quantity of and cost of disposing high level radioactive waste during recycling in the United States. However, nuclear material accountancy and control programs could also benefit from fuel debundling due to a reduction in self-shielding material for individual fuel rods compared to a fuel assembly. Current NRC regulations and guidance point to an independent fuel debundling facility with a material-possession-only license as the best option for implementing this concept into a used fuel recycling scheme in the United States. This allows the facility to comply only with simple NMAC requirements under Subpart B of 10 CFR Part 74 without the need to submit an NMAC plan to the NRC, exempting it from more tedious and complex NMAC requirements. However, a recycling facility may desire a Pu mass estimate from measurements of each individual fuel rod to improve accountancy into the dissolver. Many technologies and methodologies were considered for performing NMAC in an independent fuel debundling facility in the United States. Few were able to meet anticipated throughput demands of up to 3,300 rods/day and achieve measurement uncertainties suitable for input accountancy for a recycling facility. The best option was a combination of passive gamma-ray spectroscopy and neutron coincidence counting to meet all requirements for the proposed Fuel Debundling Detector (FUDD). Follow-on work will include a full uncertainty analysis to inform what assumptions can be made to meet input accountancy objectives before beginning experimental activities.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2022

A computational pipeline to generate a synthetic dataset of metal ion sorption to oxides for AI/ML exploration

The charged mineral/electrolyte interfaces are ubiquitous in the surface and subsurface–including the surroundings of the geological disposal sites for radioactive waste. Therefore, understanding how ions interact with charged surfaces is critically important for predicting radionuclide mobility in the case of waste leakage. At present, the Surface Complexation Models (SCMs) are the most successful thermodynamic frameworks to describe ion retention by mineral surfaces. SCMs are interfacial speciation models that account for the effect of the electric field generated by charged surfaces on sorption equilibria. These models have been successfully used to analyze and interpret a broad range of experimental observations including potentiometric and electrokinetic titrations or spectroscopy. Unfortunately, many of the current procedures to solve and fit SCM to experimental data are not optimal, which leads to a non-transferable or non-unique description of interfacial electrostatics and consequently of the strength and extent of ion retention by mineral surfaces. Recent developments in Artificial Intelligence (AI) offer a new avenue to replace SCM solvers and fitting algorithms with trained AI surrogates. Unfortunately, there is a lack of a standardized dataset covering a wide range of SCM parameter values available for AI exploration and training–a gap filled by this study. Here, we described the computational pipeline to generate synthetic SCM data and discussed approaches to transform this dataset into AI-learnable input. First, we used this pipeline to generate a synthetic dataset of electrostatic properties for a broad range of the prototypical oxide/electrolyte interfaces. The next step is to extend this dataset to include complex radionuclide sorption and complexation, and finally, to provide trained AI architectures able to infer SCMs parameter values rapidly from experimental data. Here, we illustrated the AI-surrogate development using the ensemble learning algorithms, such as Random Forest and Gradient Boosting. These surrogate models allow a rapid prediction of the SCM model parameters, do not rely on an initial guess, and guarantee convergence in all cases.

Li, Chunhui↗

DOE OSTI2021

International collaboration in disposal research: comparative modeling of coupled processes in the DECOVALEX project

Abstract. This presentation gives an overview of an international research collaboration for advancing the understanding and modeling of coupled thermo-hydro-mechanical-chemical (THMC) processes in geological systems. The creation of the international DECOVALEX project, now running for more than 25 years, was motivated by the recognition that prediction of these coupled effects is an essential part of the performance and safety assessment of geologic disposal systems for radioactive waste and spent nuclear fuel. DECOVALEX emphasizes joint analysis and comparative modeling of state-of-the-art field and laboratory experiments, across a range of host rock options and repository designs. Participating research teams are from radioactive waste management organizations, national research institutes, regulatory agencies, universities, and consulting groups, providing a wide range of perspectives and solutions to these complex problems. The presentation provides examples of the research contributions made collectively in past DECOVALEX tasks and also touches on the unique modeling challenges tackled in the ongoing project phase, referred to as DECOVALEX-2023. The current phase comprises 17 partner organizations, about 50 modeling teams, and 7 modeling tasks, which cover a broad portfolio from fundamental studies on gas migration to full-scale in situ heater experiments in different host rocks to performance assessment studies. Together, these examples illustrate that the insight and scientific knowledge gained within the DECOVALEX project would not have been possible if one group had studied these complex THMC modeling challenges alone rather than within a truly collaborative setting.

Birkholzer, Jens T.↗

NASA NTRS1978

A review of NASA-sponsored technology assessment projects

Recent technology assessment studies sponsored by NASA are reviewed, and a summary of the technical results as well as a critique of the methodologies are presented. The reviews include Assessment of Lighter-Than-Air Technology, Technology Assessment of Portable Energy RDT&P, Technology Assessment of Future Intercity Passenger Transportation Systems, and Technology Assessment of Space Disposal of Radioactive Nuclear Waste. The use of workshops has been introduced as a unique element of some of these assessments. Also included in this report is a brief synopsis of a method of quantifying opinions obtained through such group interactions. Representative of the current technology assessments, these studies cover a broad range of socio-political factors and issues in greater depth than previously considered in NASA sponsored studies. In addition to the lessons learned through the conduct of these studies, a few suggestions for improving the effectiveness of future technology assessments are provided.

Mascy, A. C.↗

NASA NTRS1977

Quantitative mapping of suspended solids in wastewater sludge plumes in the New York Bight apex

The purpose of this investigation was to apply the previously reported methodology to remotely sensed data that were collected over wastewater sludge plumes in the New York Bight apex on September 22, 1975. Spectral signatures were also determined during this study. These signatures may be useful in the specific identification of sludge plumes, as opposed to other plumes such as those created by the disposal of industrial acid wastes.

Johnson, R. W.↗

NASA NTRS1990

Feeder System For Particle-Size Analyzer

Feeder system meters precise stream of powder into precise flow of gas. Used to feed light-scattering particle-size analyzer that determines distribution of sizes of particles in powder. Dry analysis of powder takes about one-third time of conventional wet analysis and consumes less powder. Feed rate more precisely controllable, leading to more precision in analysis. In dry analysis, no need to dispose of hazardous liquid waste.

Ramsey, Keith E.↗

DOE OSTI2020

Prospects for Permanent Disposal of Spent Nuclear Fuel and High-Level Radioactive Waste in the US.

Abstract not provided.

Swift, Peter N.↗