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At least 505 records · Page 28

Halide sublattice dynamics drive Li-ion transport in antiperovskites

Here, in this work, we resolve how proton dynamics and halide mixing enhance or impede ionic conduction in protonated lithium antiperovskites (pLiAP) at compositions near the eutectic points of the halide salts. As a material class, pLiAPs of the form Li 3-x OH x X, (X = Cl, Br) show vast compositional design freedom; however, the resulting properties are susceptible to synthesis and processing methodologies. Proton incorporation and halide mixing stabilize the perovskite cubic phase at low temperatures (<50 °C) and using halide mixtures near the eutectic points (~250 to 300 °C) offer possibilities of lower temperature and faster synthesis and processing conditions (<1 h). Mixed-halide compositions such as Li 2 OHCl 0.37 Br 0.63 lead to a 30-fold improvement in room temperature ionic conductivity of a single halide structure, 1.5 × 10 -6 vs. 4.9 × 10 -8 S cm -1 (Li 2 OHCl). We combine infrared spectroscopy and nuclear magnetic resonance with first-principles density functional theory calculations to deconvolute halide mixing effects from local proton dynamics on Li-ion transport. In contrast to what has been supposed, our findings suggest that the halide sublattice dynamics, besides the OH rotation, correlate strongly with the fast-ion conduction at high temperatures.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗

Workshop to Plan R&D Support of Fuel/Basket Modification for Direct Disposal of Future DPCs

By 2030 about half of all spent nuclear fuel (SNF) arising from the current fleet of commercial power plants will be in dual-purpose canisters (DPCs), which are designed for storage and transportation but not for disposal. As an alternative to complete repackaging of the fuel for disposal, considerable cost savings and lower worker dose could be realized by directly disposing of this SNF in DPCs. The principal technical consideration is criticality control in a geologic repository, because the DPCs are large and depend on neutron absorbing basket components for criticality control. Neutron absorbing materials are generally aluminum-based, and under disposal conditions can degrade after a few hundred years contact with ground water. Simple modifications to the SNF assemblies or the DPC baskets could help to achieve direct disposal, and this is one of the approaches being studied to address the possibility of disposal criticality (SNL 2020a). Five fuel/basket modification concepts have been proposed (SNL 2020b) and a virtual workshop was conducted to solicit review and feedback on these concepts. The proposed solutions are: 1) zone loading of DPCs to limit reactivity, 2) replacing absorber plates with advanced neutron absorbing (ANA) material, 3) adding disposal control rods to pressurized water reactor (PWR) assemblies, 4) rechanneling boiling water reactor (BWR) assemblies with ANA material, and 5) basket insert plates (chevron inserts) made from ANA material. The presentations from the workshop are provided in this report, and the workshop discussions are summarized. This information includes prioritization of the proposed fuel/basket modification solutions, and prioritization of the associated model development, validation testing, and quality assurance activities. Information documented in this report will help to steer research and development efforts at Sandia National Laboratories, Oak Ridge National Laboratory, and Idaho National Laboratory that support the U.S. Department of Energy, Office of Nuclear Energy, Spent Fuel and Waste Science and Technology program

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

Physical Sciences Vistas (Issue 3 2020)

In this issue, highlights of our outstanding R&D supporting the Laboratory’s nuclear security mission include the following. 1. The essential nuclear materials science contributions of MST’s Materials Properties Team to a range of Lab missions, operations, and initiatives. 2. Work in the Sigma Complex supporting NNSA’s Advanced Manufacturing Development milestones by members of Sigma Division and colleagues across the Laboratory and other NNSA sites. 3. Greg Dale’s role in the Lab’s molybdenum program and today as a technical lead on the Scorpius project. 4. Measurements that reveal the role thermal interfaces play in dynamic compression experiments. 5. The first-ever synthesis of an actinide framework, which offers opportunities for understanding these structures as potential radioactive waste forms and provides new models of actinide species transport in the environment. This issue also showcases, as part of our commitment to simultaneous excellence in mission operations and community relations, the successful high-hazard repair of the LANSCE accelerator, which resulted in a large team Laboratory Distinguished Performance Award, and a virtual Summer Physics Camp for Young Women that brought together participants from around the world, even as COVID kept them socially distant in their homes.

36 MATERIALS SCIENCE↗

Update of the Nuclear Criticality Slide Rule Calculations: Plutonium systems – Delayed Fission Gamma

IRSN (France), LLNL (USA) and ORNL (USA) began a long-term collaboration effort in 2015 to update the nuclear criticality Slide Rule for the emergency response to a nuclear criticality accident. The Slide Rule permits the estimation of neutron and gamma dose rates and integrated doses based upon estimated fission yields, as a function of distance from the fission source, and time after criticality accidents for different critical systems. This paper presents results from the fourth phase of the current update of the Slide Rule project, in which delayed fission-product gamma (DFG) dose rates of unreflected plutonium critical systems were compared by several modern 3D radiation transport codes (MCNP, COG, SCALE), using updated flux-to-dose conversion factors. Dose rates are calculated for fissile material at five moderation ratios (H/Pu), at 1 m above the ground as a function of distance (between 30 cm and 1,200 m) from the external surface of the source to the center of the detector, and for periods between 1 s and 1,000 min after the critical instantaneous event. Further efforts have been devoted to the determination of the delayed gamma source, by comparing the time-dependent energy spectra obtained from several methods. Overall, DFG dose rates calculated by each participant led to consistent results. Extra effort is under way to identify the cause of the remaining differences, by comparing precisely the gamma source, and particularly nuclides inventories.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗

Mark-18A Cold Runs

The Savannah River National Laboratory (SRNL) is tasked by the National Nuclear Security Administration (NNSA) to recover highly valued isotopes from irradiated Mark-18A (Mk-18A) targets. The Savannah River Site (SRS) has sixty-five Mk-18A targets available for the recovery of the high valued materials. The sixty-five Mk-18A targets are currently stored in the L-Area Basin and will be removed one at a time and individually transported to SRNL. Upon receipt at SRNL, the Mk-18A target material will be removed from the confinement, dissolved, chemically separated, and calcined to a stable oxide. The flowsheet is designed to recover the plutonium as well as the trivalent actinides. The remaining unrecovered material will be discarded to the high activity drain (HAD) system in SRNL. A specially designed cask was procured for transport of the targets from L-Area to SRNL. Once received at SRNL, the targets will be loaded into the back of Cell 7 and resized as they enter the cell. The resized targets (1/4 lengths) will then be processed one at a time through the following processes: caustic dissolution and filtration; acidic dissolution and filtration, Reillex anion exchange, diglycolamide (DGA) cation exchange; and DGA calcination. This processing will result in two product streams. The first is an aqueous plutonium solution which will be removed from the shielded cells and taken to a glovebox for further purification and conversion to an oxide. The second is a calcined oxide containing the Am and Cm as well as other lanthanide fission products which will be removed from the shielded cells using a bagless transfer system. Both materials will be packaged for shipment to Oak Ridge National Laboratory (ORNL). All equipment to carry out this process was designed, procured or fabricated, and installed in a mock-up facility (716-4A) at SRS to allow for simulation testing in a non-radioactive area. This equipment was then dismantled and transferred from 716-4A to 773-A and installed in the SRNL Shielded Cells. After installation in the shielded cells facility testing was performed using water followed by surrogates and cold chemicals. Issues were identified during these evaluations, including equipment issues as well as technical challenges. Many of the issues were rectified during performance of the cold runs, and the remaining have a resolution identified. Table ES-1 provides a summary of all issues identified during the cold run operations, as well as the status and identified resolutions to outstanding issues.

07 ISOTOPE AND RADIATION SOURCES↗

U.S. High Temperature Materials Highlights

U.S. GIF VHTR work is continuing on graphite qualification, Alloy 617 regulatory issues beyond the Code space, Alloy 800H weldments, and ASME Codes and Standards R&D is still considering both pebble bed and prismatic and steam generator and heat exchanger U.S. DOE Advanced Reactor Demonstration Program (ARDP) Two U.S.-based teams were selected to demonstrate advanced nuclear reactors in the United States that can be operational by 2027 One of the teams is X-energy (Rockville, MD) which will demonstrate a modular gas-cooled reactor design (Xe-100) with four 80 MWe, TRISO fuel, pebble bed reactors A number of U.S.-based teams were selected to design and develop safe and affordable reactor technologies that can be licensed and deployed over the next 10 to 14 years (Risk Reduction) One of the teams is BWXT Advanced Technologies, LLC which will develop a commercially viable transportable microreactor with the design focused on using TRISO fuel particles and silicon carbide (SiC) matrix A number of U.S.-based teams were selected to assist the progression of advanced reactor designs in their earliest phases (Advanced Reactor Concepts-20) One of the teams is Massachusetts Institute of Technology which will mature the Modular Integrated Gas-Cooled High Temperature Reactor (MIGHTR) concept with a horizontal compact design from a pre-conceptual stage to a conceptual stage to support commercialization

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE Standard Canister Loading Methodology Poster

Over 250 different types of spent nuclear fuel (SNF) are managed by the U.S. Department of Energy (DOE) at the Idaho National Laboratory (INL). The current configuration of SNF at INL is not transportable off-site. In the Yucca Mountain Repository License Application, the SNF was to be loaded into DOE Standard Canisters which are capable of storage, transportation, and eventual disposal using the proper overpacks. The DOE Standard Canister will allow INL to place the SNF in a Road-Ready configuration in preparation for transportation off-site. The focus of this work is development of a general loading methodology for the DOE Standard Canister. This methodology is to be general enough that it will apply to all SNF to be loaded into the DOE Standard Canister. The main sections of the methodology to be outlined are prerequisites before loading (including any issues with materials such as inserts or baskets), loading of SNF, and canister closure and quality assurance (including welding and inspection). Each subsection will analyze and provide the bases for each step.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Antiferromagnetic Ordering in Quasi-One-Dimensional FeBi 4 S 7

Here, we report a detailed study of the synthesis, composition, magnetic structure, and transport properties of a quasi-one-dimensional antiferromagnet FeBi 4 S 7 that contains chains of edge-sharing FeS 6 octahedra. High-resolution powder X-ray diffraction (PXRD) analysis, aided by variation of synthetic conditions, suggests that the true formula of the material is Fe 1.2 Bi 3.8 S 7 , due to the minor substitution of Fe into Bi sites. This finding is in agreement with crystal structure refinement from neutron powder diffraction data as well as with the small band gap of 0.23 eV determined from electrical transport measurements. Analysis of the neutron diffraction pattern collected below the antiferromagnetic ordering temperature of 64 K revealed ferromagnetic coupling between the Fe moments in the chains of FeS 6 octahedra. The overall ordering, however, is antiferromagnetic due to the antiparallel arrangement of moments on neighboring chains. The collinear spin arrangement is described by a k-vector (1, 0, 1/2), which indicates doubling of the unit cell in the c direction and the loss of the C-centering translation as compared to the nuclear cell. The ferromagnetic nature of the sulfide-bridged chains of Fe 2+ ions in FeBi 4 S 7 , in contrast to the antiferromagnetic coupling between Fe moments in compounds with similar structural fragments, can be justified by the analysis of metric parameters that characterize the Fe–S bonding in these materials.

36 MATERIALS SCIENCE↗

3D TRISO particle-explicit compact meshing

The TRI-structural ISOtropic (TRISO) layered fuel particle is a robust nuclear fuel form offering enhanced safety and performance for advanced reactor concepts, including high-temperature gas-cooled reactors and other Generation IV designs. These poppy-seed-sized particles are embedded in a graphite matrix to form fuel elements that must withstand elevated temperatures and high burn-up levels. The heterogeneous nature of these fuel elements — comprising thousands of randomly distributed TRISO particles — produces complex stress fields and thermal gradients that one- and two-dimensional models cannot accurately capture. While three-dimensional modeling has improved predictions of dimensional changes, internal pressure buildup, and fission product transport under irradiation, current approaches rely on homogenized material properties that are known to have considerable divergence from experimental observations. This work presents a methodology for optimized random packing of TRISO fuel compacts and full three-dimensional mesh generation within the BISON fuel performance code, with each particle coating layer individually discretized. The resulting mesh was demonstrated through heat conduction simulations under representative in-reactor operating conditions, showing strong agreement with expected behavior. This capability enables detailed analysis of particle-to-particle interactions, matrix cracking mechanisms, and the statistical distribution of coating layer failures — all of which directly govern fuel performance and safety margins.

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Nuclear Software Validation for 7% Enriched UO 2 Fuel Lattices

This analysis performs validation for the Virtual Environment for Reaction Application’s (VERA’s) neutronics codes MPACT and Shift against measured data from the Seven Percent Critical Experiment (7uPCX) at Sandia National Laboratory. The benchmarking supports the future application of VERA for analysis of high-assay low-enriched uranium (HALEU) fuel in commercial pressurized water reactors (PWRs). VERA demonstrates very good agreement with measured data, and the deterministic code MPACT also agrees very well with higher-fidelity stochastic methods KENO-VI and Shift in both total neutron flux and fission rate distribution comparisons. Despite some significant differences between the 7uPCX and existing commercial PWRs that challenge the neutron cross section data and transport solver methods used by MPACT, all of the results are excellent. The agreement increases the confidence in the applicability and accuracy of VERA for reactor analysis of HALEU fuels in commercial reactors.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Graphite Oxidation Behavior: NRC Graphite Behavior Model

Graphite microstructure and oxidation Graphite crystal structure and basal planes Reactive surface area sites Zig-zag and Arm-chair Pore microstructure effects Interconnectivity Reactivity and oxygen transport Oxygen-graphitereaction Not C-O reaction! Temperature effects Oxygen diffusion to graphite Diffusion into the microstructure High temperature vs low temperature diffusion Oxidation: It’s a gradient of factors

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

$\mathrm{REX}$: An analytical tool for reactor operating envelope expansion through fuel-clad thermo-mechanics

Here REX, an analytical reactor design tool for the identification and expansion of reactor operating envelopes, is introduced and detailed. Written in MATLAB, REX takes core geometries and axially discretized detector and depletion output from the Serpent neutron transport code and performs full-core, pin-specific thermal-hydraulic and thermo-mechanical calculations. REX identifies the operating envelopes of advanced, solid-fuel nuclear reactors on the basis of their fuel-cladding, fuel-gas, and coolant-cladding interfacial thermo-mechanics. It iteratively pushes coolant inlet temperatures and modifies fuel pin geometries to induce mechanical failure in their original cladding materials at the fullest extent of their fuel cycle lengths and permissible coolant inlet temperatures. It then attempts to expand their operating envelopes by determining the mechanical responses of alternative cladding materials under the same geometric conditions. The result of the REX calculation sequence is a set of 5-D variables describing the temperatures, pressures, geometries, and mechanics of the core as functions of assembly, fuel pin, axial zone, depletion step, and coolant inlet temperature for each candidate cladding material with the limiting fuel pin geometry for the original material.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

Microreactor Agile Nonnuclear Experimental Testbed Test Plan

Microreactors are an attractive technology option for kick-starting nuclear innovation if they can be operated at high temperature, yielding high power conversion thermal efficiencies comparable or better than in commercial light water reactors. Microreactors are currently the smallest variation of Small Modular Reactors (SMRs). SMRs are “newer generation reactors designed to generate electric power up to 300 MWe and whose components and systems can be shop-fabricated and then transported as modules to the sites for installation as demand arises.” (IAEA, 2016). Vendors are developing microreactor designs to provide an affordable, potentially mobile source of electricity - see Fig. 1 for an example of a microreactor on a semi-truck. Various microreactor designs are possible including heat pipe- and gascooled options, which are the focus of the nonnuclear testing described in this document. In heat pipe microreactors, high-temperature heat pipes using liquid sodium or potassium working fluid transport fission heat from the core to a heat removal section which in turn transfers heat to the power conversion system working fluid. In a gas-cooled design, He or other gas will flow through a solid monolith of material and transfer heat as the temperature of the gas increases through a heat exchanger to a power conversion unit. The logistics of all these processes will be examined and tested through a series of articles at the nonnuclear test bed at Idaho National Laboratory (INL), the Microreactor Agile Nonnuclear Experiment Testbed (MAGNET) facility. Microreactors designed to produce power of 0.1-20 MWt offer the potential for more affordable nuclear energy for a range of applications. In a heat pipe microreactor, heat pipes, fuel rods, and/or moderator are intermixed in the reactor core assembly. Heat pipes extend from the core region into the heat removal section where the power conversion unit working fluid flows through holes or channels, transferring heat from the heat pipes to the working fluid. In a gas-cooled microreactor, gas flows through the solid monolith region and up into the heat exchanger region, transferring heat to the working fluid. For initial testing, the heat removal working fluid can be a low pressure gas for testing that addresses thermal stresses. In the final application, heat addition to the power conversion working fluid typically occurs at high pressure, supporting operation of an air-Brayton, supercritical CO 2 (SCO2), or He-recuperated Brayton cycle. Various stages of the steps above will be demonstrated through the tests described in this report.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

Characterising Ion-Irradiated FeCr: Hardness, Thermal Diffusivity and Lattice Strain

Ion-irradiated FeCr alloys are useful for understanding and predicting neutron damage in the structural steels of future nuclear reactors. Previous studies have largely focused on the structure of irradiation induced defects, probed by transmission electron microscopy (TEM), as well as changes in mechanical properties. Across these studies, a wide range of irradiation conditions has been employed on samples with different processing histories, which complicates the analysis of the relationship between defect structures and material properties. Furthermore, key properties, such as irradiation-induced changes in thermal transport and lattice strain, are little explored. Here we present a systematic study of Fe3Cr, Fe5Cr and Fe10Cr binary alloys implanted with 20 MeV Fe3+ ions to nominal doses of 0.01 dpa and 0.1 dpa at room temperature. Nanoindentation, transient grating spectroscopy (TGS) and X-ray micro-beam Laue diffraction were used to study the changes in hardness, thermal diffusivity and strain in the material as a function of damage and Cr content. Our results suggest that Cr leads to an increased retention of irradiation-induced defects, causing substantial changes in hardness and lattice strain. However, thermal diffusivity varies little with increasing damage and instead degrades significantly with increasing Cr content in the material. We find significant lattice strains even in samples exposed to a nominal displacement damage of 0.01 dpa. The defect density predicted from the lattice strain measurements is significantly higher than that observed in previous TEM studies, suggesting that TEM may not fully capture the irradiation-induced defect population. (C) 2020 Acta Materialia Inc. Published by Elsevier Ltd. All rights reserved.

FeCr alloys↗

Physics of the Serrano Sn Transport Package

Serrano is a software package for simulating photon transport in moving material. It presently supports 1-D, unstructured 2-D for a wide variety of element types, and 3-D for tetrahedra and hexahedra. The package is designed to be a component of larger codes, such as radiation hydrodynamics codes. Only the radiation transport equations are solved by Serrano itself.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗

Nuclear Security Interface Considerations for Regaining Control of Discovered Nuclear and Radioactive Materials

Recognizing that regaining control of orphan radioactive and nuclear material is challenging, the international community has developed requirements and guidance to support countries in addressing the problem. However, limited guidance is available on the practical nuclear security related issues that arise when nuclear or other radioactive material out of regulatory control (MORC) is encountered and efforts are undertaken to regain control. For example, the IAEA Specific Safety Guide No. SSG-19, “National Strategy for Regaining Control over Orphan Sources and Improving Control over Vulnerable Sources”, provides recommendations on a methodology for establishing a national strategy for regaining control of orphan sources. However, except for Pu239 in radioactive sources, nuclear material is outside the scope of this safety guide, and practical safety and security measures below the level of national strategy are not discussed adequately.This paper discusses the interfaces between the stakeholder groups and the responsibilities of each when encountering common scenarios that deal with MORC transitioning to regulatory control. The stakeholders include but are not limited to individuals who encounter MORC, and providers of formal and informal transport systems, and traditional and ad hoc storage solutions. This paper explores practical legal, financial, and institutional issues that hinder implementation of required safety and security practices. This paper also offers strategies that can be implemented on the individual, organizational, national, and international level to regain control of MORC.

Shannon, Michael↗

Efficiency Studies of Fast Neutron Tracking Using MCNP

Fast neutron identification and spectroscopy is of great interest to nuclear physics experiments. Using the neutron elastic scattering, the fast neutron momentum can be measured. Wang and Morris introduced the theoretical concept that the initial fast neutron momentum can be derived from up to three consecutive elastic collisions between the neutron and the target, including the information of two consecutive recoil ion tracks and the vertex position of the third collision or two consecutive elastic collisions with the timing information. Here, we also include the additional possibility of measuring the deposited energies from the recoil ions. In this paper, we simulate the neutron elastic scattering using the Monte Carlo N-Particle Transport Code (MCNP) and study the corresponding neutron detection and tracking efficiency. The corresponding efficiency and the scattering distances are simulated with different target materials, especially natural silicon (92.23%28Si, 4.67%29Si, and 3.1%30Si) and helium-4 (4He). The timing of collision and the recoil ion energy are also investigated, which are important characters for the detector design. We also calculate the ion traveling range for different energies using the software, “The Stopping and Range of Ions in Matter (SRIM)”, showing that the ion track can be most conveniently observed in 4He unless sub-micron spatial resolution can be obtained in silicon.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗

Activity gradient driven mass transport in molten fluoride salt medium

Abstract The molten salt-cooled reactor concept has garnered significant interest and one of the current challenges limiting the deployment of these reactor concepts is the complex corrosion phenomenon observed in molten salt environments. One of these phenomena is activity gradient mass transport, which has been shown to affect dissimilar materials submerged in the same salt medium even when best efforts have been made to electrically isolate dissimilar materials from one another. This mechanism while shown experimentally, has not been predictively studied through a modeling approach. In this study, activity gradients in several 316L-X materials systems have been modeled and the mass transport predicted by the model has been confirmed through static isothermal corrosion testing in a molten fluoride salt medium.

36 MATERIALS SCIENCE↗