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DOE OSTI2025

DECOVALEX-2023: An international collaboration for advancing the understanding and modeling of coupled thermo-hydro-mechanical-chemical (THMC) processes in geological systems

The DECOVALEX initiative is an international research collaboration (www.decovalex.org), initiated in 1992, for advancing the understanding and modeling of coupled thermo-hydro-mechanical-chemical (THMC) processes in geological systems. DECOVALEX stands for “DEvelopment of COupled Models and VALidation against EXperiments”. The creation of this international initiative 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 the complex perturbations and coupled processes in geologic repositories and how these impact long-term performance predictions. The most recent phase of the DECOVALEX Project, here referred to as DECOVALEX-2023, started in early 2020 and ended in late 2023. More than fifty research teams associated with 17 international DECOVALEX partner organizations participated in the comparative evaluation of eight modeling tasks covering a wide range of spatial and temporal scales, geological formations, and coupled processes. This Virtual Special Issue on DECOVALEX-2023 provides an in-depth overview of these collaborative research efforts and how these have advanced the state-of-the-art of understanding and modeling coupled THMC processes. While primarily focused on radioactive waste, much of the work included here has wider application to many geoengineering topics.

Coupled processes↗
DOE OSTI2017

Preliminary Concept of Operations for the Spent Fuel Management System+ - 17134

The Nuclear Fuels Storage and Transportation Planning Project (NFST) within the US Department of Energy's Office of Nuclear Energy is tasked with identifying, planning, and conducting activities for developing interim storage and transportation capabilities in support of an integrated waste management system (IWMS). The IWMS is expected to transport Spent Nuclear Fuel (SNF) from a reactor and deliver it to a repository in a suitable container, with the possibility of interim storage. The system will be composed of multiple subsystems including possible interim storage facilities (ISFs), one or more repositories, facilities to package and/or repackage SNF, and transportation. The project team is analyzing options for an IWMS to identify viable options. To support analysis, the project team has developed a concept of operations framework for SNF related components of the potential integrated system and the interdependencies between those components. A goal of this work is to aid system analysts in developing consistent models across the project. The concept of operations framework will be updated periodically as new developments emerge. At a high level, SNF is expected to be transported from reactor sites to a repository. Initially, SNF is unloaded from reactors and placed in spent fuel pools (SFPs) for wet storage at utility sites. To maintain space in SFPs, SNF is placed in containers at reactor sites for storage and/or transportation only after the SNF has cooled enough to satisfy loading limits. Once transportation requirements are met, the SNF may be transported to an ISF, where it is stored until a repository is developed, or it is transported directly to a repository if one is available.While the high level operation of the system is straightforward, analysts must evaluate numerous alternatives, such as the number of ISFs (if any), ISF design, the stage at which SNF repackaging occurs (if needed), repackaging technology, types of containers used, repository design, component sizing, and timing of events. These alternatives arise due to technological, economic, or policy considerations. This paper provides an overview of the potential alternatives identified in the concept of operations framework at a conceptual level. (authors)

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DOE OSTI2017

Proceedings of the 7th US/German Workshop on Salt Repository Research, Design, and Operation.

The 7th US/German Workshop on Salt Repository Research, Design, and Operation was held in Washington, DC on September 7-9, 2016. Over fifty participants representing governmental agencies, internationally recognized salt research groups, universities, and private companies helped advance the technical basis for salt disposal of radioactive waste. Representatives from several United States federal agencies were able to attend, including the Department of Energy´s Office of Environmental Management and Office of Nuclear Energy, the Environmental Protection Agency, the Nuclear Regulatory Commission, and the Nuclear Waste Technical Review Board. A similar representation from the German ministries showcased the covenant established in a Memorandum of Understanding executed between the United States and Germany in 2011. The US/German workshops´ results and activities also contribute significantly to the Nuclear Energy Agency Salt Club repository research agenda.

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DOE OSTI2017

Proceedings of the 7th US/German Workshop on Salt Repository Research, Design, and Operation

The 7th US/German Workshop on Salt Repository Research, Design, and Operation was held in Washington, DC on September 7-9, 2016. Over fifty participants representing governmental agencies, internationally recognized salt research groups, universities, and private companies helped advance the technical basis for salt disposal of radioactive waste. Representatives from several United States federal agencies were able to attend, including the Department of Energy's Office of Environmental Management and Office of Nuclear Energy, the Environmental Protection Agency, the Nuclear Regulatory Commission, and the Nuclear Waste Technical Review Board. A similar representation from the German ministries showcased the covenant established in a Memorandum of Understanding executed between the United States and Germany in 2011. The US/German workshops' results and activities also contribute significantly to the Nuclear Energy Agency Salt Club repository research agenda.

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DOE OSTI2017

Treatment of Radioactive Sludge From Waste Water Treatment for Weight and Volume Reduction with FreezeTec-Technique - 17010

At the Studsvik site in Sweden, Studsvik Nuclear AB operates a water treatment facility that treat all the waste water from the nuclear facilities at the site. The facility contains several pools for waste water treatment. The standard procedure for treating the water consist of adding FeCl{sub 3} to precipitate any particulates. Historically this waste has been mixed with concrete and grouted into barrels which will increase the volume needed for disposal in final repository. The sludge that is generated from the water treatment facility has been subject to treatment with the Freeze-Tec method where the sludge is freeze dewatered and after thawing generating a solid phase with high dry substance and a clear water phase. The radioactive particles/nuclides tend to stay in the solid phase while the water phase has very low concentration of radioactive particles. 99.9% of the Co-60 and Cs-137 activity end up in the solids and a dry solid with smaller volume and weight is obtained after treatment. (authors)

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DOE OSTI2020

Preliminary Reporting of Thermophysical Property Measurements for the Ghareb Formation

Accurate knowledge of thermophysical properties of rock is vital to develop meaningful models of high level nuclear waste emplacement scenarios. The Israel Atomic Energy Commission is considering storing high level nuclear waste in the Ghareb formation, a porous kerogen bearing chalk. Sandia is supporting this effort with an evolving lab- based geomechanics testing program. We have completed measurements of thermal properties up to 275C and room temperature hydrostatic compaction measurements. We report thermal conductivity, thermal diffusivity, specific heat, and mass loss from our thermal measurements, and we report bulk moduli and porosity loss from our compaction measurements. These values are crucial for the numerical models to simulate heat transfer and formation compressibility around a heat generating repository.

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DOE OSTI2020

Analysis Of The Sludge Batch 7b (Macrobatch 9) DWPF Pour Stream Glass Sample (Rev. 1)

The Defense Waste Processing Facility (DWPF) began processing Sludge Batch 7b (SB7b), also referred to as Macrobatch 9 (MB9), in January 2012. SB7b is a blend of the heel of Tank 40 from Sludge Batch 7a (SB7a) and the SB7b material that was transferred to Tank 40 from Tank 51. SB7b was processed using Frit 418. During processing of each sludge batch, the DWPF is required to take at least one glass sample to meet the objectives of the Glass Product Control Program (GPCP), which is governed by the DWPF Waste Form Compliance Plan, and to complete the necessary Production Records so that the final glass product may be disposed of at a Federal Repository. Two pour stream glass samples were collected while processing SB7b. The samples were transferred to the Savannah River National Laboratory (SRNL) where one was analyzed and the other was archived. The following conclusions were drawn from the analytical results provided in this report: The sum of oxides for the official SB7b pour stream glass is within the Product Composition Control System (PCCS) limits (95-105 wt%); The average calculated Waste Dilution Factor (WDF) for SB7b is 2.3. In general, the measured radionuclide content of the official SB7b pour stream glass is in good agreement with the calculated values from the Tank 40 dried sludge results from the SB7b Waste Acceptance Program Specification (WAPS) sample; As in previous pour stream samples, ruthenium and rhodium inclusions were detected by Scanning Electron Microscopy-Electron Dispersive Spectroscopy (SEM-EDS) in the SB7b pour stream sample; The Product Consistency Test (PCT) results indicate that the official SB7b pour stream glass meets the waste acceptance criteria for durability with a normalized boron release of 0.8 g/L, which is an order of magnitude less than the Environmental Assessment (EA) glass; The measured density of the SB7b pour stream glass was 2.70 g/cm 3 ; The Fe 2+ /ΣFe ratio of the SB7b pour stream samples was 0.07.

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DOE OSTI2022

Bioenergy Feedstock Library Annual Summary Report

The Bioenergy Feedstock Library (BFL), part of the Biomass Feedstock National User Facility (BFNUF) located at INL, is a physical sample repository and a web-accessible electronic database. The BFL stores physical and chemical characteristics of biomass and waste carbon sources for energy use as well as samples generated from across U.S. Department of Energy (DOE) Bioenergy Technologies Office (BETO) funded projects. The objective of this Bioenergy Feedstock Library Annual Summary Report for 2022 is to focus on the (1) publicly available analytical data and equipment tracked through the BFNUF, (2) physical samples available for request, (3) sample and data archival progress from recent BETO-funded projects, and (4) publicly available data sets created upon request from BETO, INL projects, or outside entities. This report highlights key statistics from FY22 and available data and information important for INL, BFL users, academics, and industry.

09 BIOMASS FUELS↗
DOE OSTI2023

Bioenergy Feedstock Library 2023 Annual Summary Report

The Bioenergy Feedstock Library (BFL), part of the Biomass Feedstock National User Facility (BFNUF) located at INL, is a physical sample repository and a web-accessible electronic database. The BFL stores physical and chemical characteristics of biomass and waste carbon sources for energy use as well as samples generated from across U.S. Department of Energy (DOE) Bioenergy Technologies Office (BETO) and U.S. Department of Agriculture funded projects. The objective of this Bioenergy Feedstock Library Annual Summary Report for 2023 is to focus on the updates to (1) publicly available analytical data and equipment tracked through the BFNUF, (2) physical samples available for request, (3) sample and data archival progress from recent BETO-funded projects, and (4) publicly available datasets created upon request from BETO, INL projects, or outside entities compared to the 2022 Bioenergy Feedstock Library Annual Summary Report. This report highlights key statistics and available data and information important for INL, BFL users, academics, and industry.

09 BIOMASS FUELS↗
DOE OSTI2024

Bioenergy Feedstock Library Annual Summary Report 2024

The Bioenergy Feedstock Library (BFL), part of the Biomass Feedstock National User Facility (BFNUF) located at Idaho National Laboratory (INL), is a physical sample repository and a web-accessible electronic database. The BFL stores physical and chemical characteristics of biomass and waste carbon sources for energy use, as well as samples generated from U.S. Department of Energy (DOE) Bioenergy Technologies Office (BETO) and U.S. Department of Agriculture-funded projects. The objective of this Bioenergy Feedstock Library Annual Summary Report for 2024, similar to the 2023 Annual Summary Report , is to focus on the updates to: (1) publicly available analytical data and equipment tracked through the BFNUF, (2) significant increases in the physical samples available for request, (3) sample and data archival progress from recent BETO-funded projects, and (4) publicly available data sets created upon request from BETO, INL projects, or outside entities compared to the previous annual summary reports. This report highlights key statistics and available data and information important for INL, BFL users, academics, and industry.

09 BIOMASS FUELS↗
DOE OSTI2025

Effects of Borate and Organics on U(VI) Solubility in WIPP Brine

This report provides an update to a previously issued report on the solubility of uranium (VI) at different borate concentrations and in the presence of organics. The solubility of uranium (VI) in the Waste Isolation Pilot Plant (WIPP)-relevant brine was determined to support ongoing WIPP recertification activities (CRA-2026). This research was performed by the Los Alamos National Laboratory-Carlsbad Operations (LANL-CO) Actinide Chemistry and Repository Science Program (ACRSP).

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DOE OSTI2019

Proceedings of the 10th US/German Workshop on Salt Repository Research Design and Operation. Rev. 1

The 10th US/German Workshop on Salt Repository Research, Design, and Operation was hosted by RESPEC and South Dakota School of Mines & Technology, both located in the Black Hills of South Dakota. Over 60 registered participants representing Germany, United States, the Netherlands, and the United Kingdom availed themselves to excellent facilities on the School of Mines campus. As the 10th annual workshop, this occasion is a milestone of the modern era of collaboration between the US and Germany, which has extended to other countries with potential for radioactive waste disposal in salt. Thrust areas covered in the annual workshops are selected by participants and typically include elements of continuity from year to year. This year's themes included siting, modeling challenges, seal systems and materials, operational safety, and special topics. Two major breakout sessions addressed test sample conditioning and natural closure of salt openings. When the new-generation US/German workshops were conceived, one goal was to identify challenging issues related to salt repository sciences and then conduct open discussions in special breakout sessions. These timely sessions achieve the workshop paradigm and provide in-depth dialogue on important salt repository considerations. In many respects, general themes of these salt repository workshops reflect advances in the scientific basis for nuclear-waste disposal in salt formations and develop naturally as a consequence of unremitting attention largely as a result of the workshop commitment. Technical capabilities in the laboratory and field continue to improve in concert with accumulating experience. Because we are working together, mechanical deformation at the micro-scale can be interpreted at a large scale, which is fundamental to predictive modeling of salt repository evolution. This document records the Proceedings of the 2019 gathering of salt repository nations and has been compiled by RESPEC, demonstrating an adopted protocol in which the host organization creates that year's Proceedings. To assist with organization and compilation, individual chapters are summarized by volunteer subject-matter experts from the participatory audience; these contributors are recognized in the Acknowledgements. All formal presentations are included in this document, thus providing a resource for referencing and many excellent photographic images. Appendices include the agenda, list of participants, abstracts, and presentations. A primary purpose for recording workshop activities is to create and sustain an accessible record of salt repository research. These archives also illustrate transparent development of research areas each year.

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DOE OSTI2018

The High Capacity Type B Waste Transport Cask to Serve Dismantling Project - 18344

The United States have already announced around 20 reactor shutdown and many more in the next 10 year. Financial constraints on plant operators and the low cost of fossil energy contribute to accelerate the shutdown of additional reactors. The industry has been gaining experience with reactor dismantling and is continuously trying to improve the cost efficiency of the upcoming projects. To do so, larger and more capacitive transport casks are needed to transport increasing volume of Low Level Waste (LLW) at a more attractive cost. Waste disposal is also a major cost element in a dismantling project. Consequently, the capacity to segment, package and transport the waste in the most optimized configuration will result in significant cost savings. While for LLW, the constraints for package requirements are limited, activated waste resulting mainly from reactor vessel and vessel internals segmentation require additional shielding to meet disposal and transportation regulations. The key element to succeed is the availability on the market of a licensed type B transport package with a larger payload that allows at the same time an optimization of the waste segmentation. AREVA TN has designed and licensed versatile high capacity type B casks for the transport of spent fuel canisters as part of the NUHOMS{sup R} system that can also be used to transport radioactive waste. While it will take a few more years to transport spent fuel to a consolidated interim storage facility and even more to a repository, the type B casks can be used to transport activated LLW from dismantled reactors to disposal sites. A first of a kind project will be performed with AREVA's partner NorthStar at Vermont Yankee Nuclear Power Plant. AREVA's team will perform the reactor segmentation and package the material into an AREVA's large capacity type B cask. This unmatched cask capacity is expected to significantly decrease the LLW disposal costs, as well as the onsite operations costs, while optimizing the schedule. The package will be transported by rail to waste disposal facilities. This paper will present the main characteristics of the package, provide an update of the overall schedule, and explain the innovations that contribute to simpler and more cost effective transport and operation. Finally, it will emphasize the benefits of using larger type B casks for immediate and future transportation needs. (authors)

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DOE OSTI2025

Effect of pore fluid chemistry on the mechanical behavior of a divalent compacted bentonite, an experimental and constitutive study

Ongoing research in isolating high-level nuclear waste and spent fuel has highlighted compacted bentonite as a suitable material for engineered barrier systems in deep geological repositories due to its extraordinary swelling and retention properties. This research focuses on the chemo-mechanical behavior of compacted bentonite exposed to different pore fluids with different concentrations and loading conditions. The study involves swelling pressure and compressibility experiments along with mineralogy analysis employing X-ray diffraction (XRD) and Cation exchange. The tests were conducted on BCV (a Mg/Ca- bentonite) compacted at a dry density of 1.48 ± .02 Mg/m 3 . An advanced chemical-mechanical constitutive model for unsaturated highly expansive clays was adopted to simulate the material response and better understand its behavior. The model is able to account for the main phenomena at both macro and microstructural levels and the interactions between them. The model successfully replicated experimental observations. The XRD analyses support the macroscopic observation, indicating that salinity impacts crystalline swelling as demonstrated by the reduction of basal spacing from 19.27 Å to 15.68 Å when the osmotic suction increases from 0 MPa to 33 MPa. The results suggested that the osmotic pressure generated by the concentration in the pore fluids promotes a reduction in swelling pressures, swelling strains, and crystalline swelling of clay minerals. Also, it affects the pre-consolidation stress and the compressibility of the compacted samples. In conclusion, it was also observed that both solution type and solution concentration impact the clay swelling pressure.

Chemo-mechanical constitutive model↗
DOE OSTI2025

Demonstration Test Report: Vacuum Thermal Desorption (VTD) Testing to Support West Area Waste Feed Test 2

Approximately 54 million gallons of radioactive and hazardous waste are stored in underground storage tanks at the U.S. Department of Energy’s (DOE’s) Hanford Site in southeastern Washington State. The Hanford Waste Treatment and Immobilization Plant (WTP) is being constructed to immobilize the waste in glass waste forms. The baseline configuration of the WTP includes a Pretreatment Facility to separate the wastes into a small volume of high-level waste (HLW) containing most of the radioactivity and a larger volume of low-activity waste (LAW) containing most of the non-radioactive chemicals. The HLW will be converted to glass in the HLW vitrification facility for ultimate disposal at a federal repository.

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DOE OSTI2022

Geologic Disposal Safety Assessment (GDSA) Biosphere Model Development

The Spent Fuel and Waste Science and Technology Campaign of the U.S. Department of Energy Office of Nuclear Energy, Office of Spent Fuel and Waste Disposition is conducting research and development on geologic disposal of spent nuclear fuel and high-level nuclear waste. This work includes the Geologic Disposal Safety Assessment (GDSA) program which is charged with development of generic deep geologic repository concepts and system performance assessment models. One part of the GDSA framework is the development of a biosphere model capable of assessing doses to potential receptors exposed to radionuclides released from geologic disposal sites. As part of the GDSA framework, a biosphere model compatible with the PFLOTRAN massively parallel subsurface flow and reactive transport code is under development. The PFLOTRAN model provides the radionuclide source term for the biosphere model. The GDSA Biosphere model then assesses the potential movement of radionuclides through the surface biosphere and the subsequent exposure to a human receptor living in the biosphere. The biosphere model includes pathways originating from the groundwater as well as pathways originating from surface water bodies that have a water exchange with a contaminated groundwater body. The pathways for human exposure include consumption of drinking water, irrigated crops, meat animals, aquatic vegetation, and animals, etc.; external exposure from irrigated ground surfaces, surface water bodies, recreational activities, etc.; and inadvertent exposures such as ingestion of contaminated soils or shower water, etc. The GDSA Biosphere model was designed to be flexible and generic in order to accommodate a variety of different sites and climate states. This presentation will present the on the purpose, design, and development progress of the GDSA Biosphere Model.

GDSA, biosphere, repository↗
DOE OSTI2020

Critical Design Elements for an On-Site Mercury Concentration and Holding Facility - 20175

Administrative changes and funding issues have delayed the Department of Energy (DOE) from establishing and commissioning a National Repository for mercury as required under the Mercury Export Ban Act (MEBA) of 2008. In the interim, a few hazardous waste Treatment, Storage, and Disposal Facilities (TSDFs) have been authorized to accept mercury classified as a hazardous waste as under Subtitle C of the Resource Conservation and Recovery Act (RCRA). Waste generators are charged (either by weight or by volume) for the waste they manifest to these TSDFs. In many cases, mercury and mercury compounds have chemically sorbed to (or physically lodged into) the interstitial spaces of a substrate material. This almost always causes the amount of waste generated to be unnecessarily large, due to the extra volume (or weight) of the accompanying, non-mercury materials. Depending on a multitude of factors, it could be financially beneficial for waste generators to take greater control of their mercury affairs by separating these substrate materials from the mercury and mercury compounds that have become associated with them. Building and operating an on-site mercury Concentration and Holding Facility (CHF) could help relieve a waste generator's overall cost burden. The purpose of such CHFs is not to treat or dispose of mercury, but to recover and concentrate mercury from bulky mercury-laden materials to the extent that the host substrate to which the mercury may have previously been sorbed could pass a Toxicity Characteristic Leaching Procedure (TCLP) test and either be disposed of at a conventional landfill, or be disposed of at a hazardous waste landfill, but absent additional restrictions specifically for mercury. A CHF represents a way for waste generators to save money by separating mercury from common host materials, concentrating it, and then safely storing it until transportation to an authorized TSDF takes place This paper focuses on the critical design elements waste generators should consider when designing and operating a mercury CHF. (authors)

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DOE OSTI2020

M4SF-20LL010301042: Updated Thermodynamic Database for Use in Generic Disposal System Assessment

This progress report (Level 4 Milestone Number M4SF-20LL010301042) summarizes research conducted at Lawrence Livermore National Laboratory (LLNL) within the Argillite Activity Number SF-20LL010301041. LLNL is leading efforts in the development of thermodynamic databases in support of the Spent Fuel and Waste Science Technology (SFWST) program. Thermodynamic models provide the basis for understanding the stability of solid phases and speciation of aqueous species and modeling the evolution of repository conditions. The LLNL effort is being performed in coordination with other US database development efforts. The effort includes a review of available thermochemical databases and a path forward for database integration. International coordination with the NEA-TDB is supported through crystalline international work package.

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