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At least 793 records · Page 44

Charge-Coupled Substituted Garnets (Y 3-x Ca 0.5x M 0.5x )Fe 5 O 12 (M = Ce, Th): Structure and Stability as Crystalline Nuclear Waste Forms

The garnet structure has been proposed as a potential crystalline nuclear waste form for accommodation of actinide elements, especially uranium (U). In this study, yttrium iron garnet (YIG) as a model garnet host was studied for the incorporation of U analogs, cerium (Ce) and thorium (Th), incorporated by a charge-coupled substitution with calcium (Ca) for yttrium (Y) in YIG, namely, 2Y 3+ = Ca 2+ + M 4+ , where M 4+ = Ce 4+ or Th 4+ . Single-phase garnets Y 3–x Ca 0.5x M 0.5x Fe 5 O 12 (x = 0.1–0.7) were synthesized by the citrate–nitrate combustion method. Ce was confirmed to be tetravalent by X-ray absorption spectroscopy and X-ray photoelectron spectroscopy. X-ray diffraction and 57 Fe–Mössbauer spectroscopy indicated that M 4+ and Ca 2+ cations are restricted to the c site, and the local environments of both the tetrahedral and the octahedral Fe 3+ are systematically affected by the extent of substitution. The charge-coupled substitution has advantages in incorporating Ce/Th and in stabilizing the substituted phases compared to a single substitution strategy. Enthalpies of formation of garnets were obtained by high temperature oxide melt solution calorimetry, and the enthalpies of substitution of Ce and Th were determined. The thermodynamic analysis demonstrates that the substituted garnets are entropically rather than energetically stabilized. This suggests that such garnets may form and persist in repositories at high temperature but might decompose near room temperature.

Guo, Xiaofeng↗

Conceptual Design of the Package Handling System for the Deep Borehole Engineering Demonstration - 17064

Deep borehole disposal involves drilling one or more boreholes into crystalline basement rock to a depth of about 5 km, emplacing waste packages in the lower 2 km of the borehole, and sealing and plugging the upper 3 km. These depths are several times deeper than for typical mined repositories. A deep borehole field test (DBFT) is proposed to confirm the safety and feasibility of the deep borehole concept. The primary goals are to demonstrate the construction and characterization of deep boreholes, demonstrate the equipment and operations used for safe waste handling and emplacement downhole, study geologic controls on waste form stability and isolation, and evaluate the overall safety and practicality of the concept. The test will be conducted using simulated waste packages; no radioactive wastes will be handled at the deep borehole field test site. A key component of the DBFT is an engineering demonstration of the equipment and operations needed to transport waste packages to the site, move packages between the transportation and emplacement systems, and lower or retrieve packages into or out of the borehole. This paper describes the conceptualization of test package handling, from transportation and receipt through the interface with the borehole. It also describes the use of an integrated test facility to check out equipment prior to deployment at the DBFT site, and performance of the engineering demonstration at the DBFT site. (authors)

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Delayed Gamma-ray Spectroscopy for Non-Destructive Assay of Nuclear Materials

This project addresses the need for improved non-destructive assay techniques for quantifying the actinide composition of spent nuclear fuel and for the independent verification of declared quantities of special nuclear materials at key stages of the fuel cycle. High-energy delayed gamma-ray spectroscopy following neutron irradiation is a potential technique for directly assaying spent fuel assemblies and achieving the safeguards goal of quantifying nuclear material inventories for spent fuel handling, interim storage, reprocessing facilities, repository sites, and final disposal. Other potential applications include determination of MOX fuel composition, characterization of nuclear waste packages, and challenges in homeland security and arms control verification.

98 NUCLEAR DISARMAMENT, SAFEGUARDS, AND PHYSICAL P↗

Delayed Gamma-Ray Spectroscopy for Non-Destructive Assay of Nuclear Materials

High-­energy, beta-delayed gamma-­ray spectroscopy is a potential, non-­destructive assay techniques for the independent verification of declared quantities of special nuclear materials at key stages of the fuel cycle and for directly assaying nuclear material inventories for spent fuel handling, interim storage, reprocessing facilities, repository sites, and final disposal. Other potential applications include determination of MOX fuel composition, characterization of nuclear waste packages, and challenges in homeland security and arms control verification. Experimental measurements were performed to evaluate fission fragment yields, to test methods for determining isotopic fractions, and to benchmark the modeling code package. Experimental measurement campaigns were carried out at the IAC using a photo-­neutron source and at OSU using a thermal neutron beam from the TRIGA reactor to characterize the emission of high-­energy delayed gamma rays from 235 U, 239 Pu, and 241 Pu targets following neutron induced fission. Data were collected for pure and combined targets for several irradiation/spectroscopy cycle times ranging from 10/10 seconds to 15/30 minutes.The delayed gamma-ray signature of 241 Pu, a significant fissile constituent in spent fuel, was measured and compared to 239 Pu. The 241 Pu/ 239 Pu ratios varied between 0.5 and 1.2 for ten prominent lines in the 2700-­3600 keV energy range. Such significant differences in relative peak intensities make it possible to determine relative fractions of these isotopes in a mixed sample. A method for determining fission product yields by fitting the energy and time dependence of the delayed gamma-­ray emission was developed and demonstrated on a limited 235 U data set. De-­convolution methods for determining fissile fractions were developed and tested on the experimental data. The use of high count-­rate LaBr 3 detectors was investigated as a potential alternative to HPGe detectors. Modeling capabilities were added to an existing framework and codes were adapted as needed for analyzing experiments and assessing application-­specific assay concepts. A de-­convolution analysis of the delayed gamma-­ray response spectra modeled for spent fuel assemblies was performed using the same method that was applied to the experimental spectra.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗

Past Approaches for Spent Nuclear Fuel, High-Level, and Transuranic Waste Disposal in the United States—Part 1: Safety Criteria and Treatment of Uncertainty

The United States (US), with its 50-year experience in developing deep geologic disposal for transuranic waste, spent nuclear fuel (SNF), and high-level radioactive waste (HLW), has much to share with other countries. Yet, other countries are better able to translate the US experience and corresponding policy decisions into solutions sensible for their country when they understand the compliance requirements in US laws and regulations. This paper presents past approaches in the generic and site-specific standards of the US Environmental Protection Agency (EPA) and implementing regulations of the US Nuclear Regulatory Commission (NRC) using the framework provided by (1) key questions of the Blue Ribbon Commission on America’s Nuclear Future, and (2) international consensus standards. Both the 1985 EPA generic standards, as updated in 1993 and applied at the operating Waste Isolation Pilot Plant in southern New Mexico for transuranic waste from atomic energy defense activities, and the EPA 2008 site-specific standards and implementing regulations, as applied at the proposed Yucca Mountain repository in southern Nevada for SNF and HLW, adopt the strategy of using quantitative, probabilistic analysis to assess performance and compliance.

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Annual Transuranic Waste Inventory Report – 2015 (Data Cutoff Date 12/31/2014)

The purpose of this Annual Transuranic Waste Inventory Report – 2015 (ATWIR-2015) is to document the updated inventory estimate of transuranic (TRU) waste reported by the TRU waste generator sites as of December 31, 2014. This updated inventory information is provided to the U.S. Department of Energy (DOE) Carlsbad Field Office (CBFO) and is available to the DOE TRU waste complex, Waste Isolation Pilot Plant (WIPP) stakeholders, and regulators. The TRU waste inventory information is used for strategic planning, supports DOE/CBFO’s input into documents (e.g., WIPP Documented Safety Analysis and National Environmental Policy Act analyses), the development of new containers or shipping packages, planned change requests, and other design changes that may take place in the repository. Furthermore, if requested by DOE/CBFO, this report will provide the basis for the Performance Assessment Inventory Report for performance assessment modeling. This report focuses on the TRU waste stored or projected to be generated (up through calendar year [CY] 2033) at the TRU waste generator sites.

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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↗

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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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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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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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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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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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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Annual Transuranic Waste Inventory Report – 2016 (Data Cutoff Date 12/31/2015)

The purpose of this Annual Transuranic Waste Inventory Report – 2016 (ATWIR-2016), using a cutoff date of December 31, 2015, is to document the updated inventory estimate of transuranic (TRU) waste reported by the TRU waste generator sites in the Annual Transuranic Waste Inventory Report – 2015 (ATWIR-2015) (DOE 2015a) that had a cutoff date of December 31, 2014. This updated inventory information is provided to the U.S. Department of Energy (DOE) Carlsbad Field Office (CBFO) and is available to the DOE TRU waste complex, Waste Isolation Pilot Plant (WIPP) stakeholders, and regulators. The TRU waste inventory information is used for strategic planning, and supports DOE/CBFO’s input into documents (e.g., WIPP Documented Safety Analysis and National Environmental Policy Act analyses), the development of new containers or shipping packages, planned change requests, and other design changes that may take place in the repository. Furthermore, if requested by DOE/CBFO, this report will provide the basis for the Performance Assessment Inventory Report for performance assessment (PA) modeling. This report focuses on the TRU waste stored or projected to be generated (up through calendar year [CY] 2050) at the TRU waste generator sites.

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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↗

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.

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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↗

Actinide Solubility and Speciation in the WIPP [PowerPoint]

The presentation begins with the role and need for nuclear repositories (overall concept, international updates (Sweden, Finland, France, China), US approach and current status), then moves on to the WIPP TRU repository concept (design, current status--safety incidents of February 5 and 14, 2014, path forward), and finally considers the WIPP safety case: dissolved actinide concentrations (overall approach, oxidation state distribution and redox control, solubility of actinides, colloidal contribution and microbial effects). The following conclusions are set forth: (1) International programs are moving forward, but at a very slow and somewhat sporadic pace. (2) In the United States, the Salt repository concept, from the perspective of the long-term safety case, remains a viable option for nuclear waste management despite the current operational issues/concerns. (3) Current model/PA prediction (WIPP example) are built on redundant conservatisms. These conservatisms are being addressed in the ongoing and future research to fill existing data gaps--redox control of plutonium by Fe(0, II), thorium (analog) solubility studies in simulated brine, contribution of intrinsic and biocolloids to the mobile concentration, and clarification of microbial ecology and effects.

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