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DOE OSTI2015

GRAPhEME: a setup to measure (n, xn γ) reaction cross sections

Most of nuclear reactor developments are using evaluated data base for numerical simulations. However, the considered databases present still large uncertainties and disagreements. To improve their level of precision, new measurements are needed, in particular for (n, xn) reactions, which are of great importance as they modify the neutron spectrum, the neutron population, and produce radioactive species. The IPHC group started an experimental program to measure (n, xn gamma) reaction cross sections using prompt gamma spectroscopy and neutron energy determination by time of flight. Measurements of (n, xn gamma) cross section have been performed for {sup 235,238}U, {sup 232}Th, {sup nat,182,183,184,186}W, {sup nat}Zr. The experimental setup is installed at the neutron beam at GELINA (Geel, Belgium). The setup has recently been upgraded with the addition of a highly segmented 36 pixels planar HPGe detector. Significant efforts have been made to reduce radiation background and electromagnetic perturbations. The setup is equipped with a high rate digital acquisition system. The analysis of the segmented detector data requires a specific procedure to account for cross signals between pixels. An overall attention is paid to the precision of the measurement. The setup characteristic and the analysis procedure will be presented along with the acquisition and analysis challenges. Examples of results and their impact on models will be discussed. (authors)

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗

DOE OSTI2025

Unraveling Impurity-Dependent Morphological and Chemical Evolution of Ni–20Cr Alloy in Eutectic LiCl–KCl Molten Salt

Understanding the interfacial evolution of alloys in molten salt with different amounts of water (H 2 O) and oxygen (O 2 ) impurities is significant for applications in many fields, including concentrated solar power, molten salt reactors, and applications in pyrochemical reprocessing and electrorefining. Additionally, the impurity-driven corrosion mechanisms that lead to various morphological and chemical evolution characteristics at the interfaces of structural alloys and molten salts are not fully understood. In the present work, the three-dimensional (3D) morphological evolution of Ni-20Cr microwires in LiCl-KCl was studied at 500 °C under different moisture and oxygen conditions using in situ synchrotron transmission X-ray microscopy (TXM) and scanning transmission electron microscopy (STEM) techniques. No significant morphological changes were observed in Ni-20Cr microwires under vacuum conditions. However, the wires exhibited distinct morphological evolutions when exposed to molten salt containing H 2 O alone, as well as when both H 2 O and O 2 were present. Furthermore, Cr 2 O 3 precipitates were observed in the molten salt during corrosion with only H 2 O present, while Cr 6+ species were identified in the salt when O 2 was added. Further, these findings are crucial for understanding the corrosion mechanisms of molten salt with different amounts of H 2 O and O 2 contamination, providing insights for developing corrosion mitigation methods and improving the stability of containment alloys in molten salt applications.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗

DOE OSTI2024

Overview of ASDEX upgrade results in view of ITER and DEMO

Experiments on ASDEX Upgrade (AUG) in 2021 and 2022 have addressed a number of critical issues for ITER and EU DEMO. A major objective of the AUG programme is to shed light on the underlying physics of confinement, stability, and plasma exhaust in order to allow reliable extrapolation of results obtained on present day machines to these reactor-grade devices. Concerning pedestal physics, the mitigation of edge localised modes (ELMs) using resonant magnetic perturbations (RMPs) was found to be consistent with a reduction of the linear peeling-ballooning stability threshold due to the helical deformation of the plasma. Conversely, ELM suppression by RMPs is ascribed to an increased pedestal transport that keeps the plasma away from this boundary. Candidates for this increased transport are locally enhanced turbulence and a locked magnetic island in the pedestal. The enhanced D-alpha (EDA) and quasi-continuous exhaust (QCE) regimes have been established as promising ELM-free scenarios. Here, the pressure gradient at the foot of the H-mode pedestal is reduced by a quasi-coherent mode, consistent with violation of the high-n ballooning mode stability limit there. This is suggestive that the EDA and QCE regimes have a common underlying physics origin. In the area of transport physics, full radius models for both L- and H-modes have been developed. These models predict energy confinement in AUG better than the commonly used global scaling laws, representing a large step towards the goal of predictive capability. A new momentum transport analysis framework has been developed that provides access to the intrinsic torque in the plasma core. In the field of exhaust, the X-Point Radiator (XPR), a cold and dense plasma region on closed flux surfaces close to the X-point, was described by an analytical model that provides an understanding of its formation as well as its stability, i.e., the conditions under which it transitions into a deleterious MARFE with the potential to result in a disruptive termination. With the XPR close to the divertor target, a new detached divertor concept, the compact radiative divertor, was developed. Here, the exhaust power is radiated before reaching the target, allowing close proximity of the X-point to the target. No limitations by the shallow field line angle due to the large flux expansion were observed, and sufficient compression of neutral density was demonstrated. With respect to the pumping of non-recycling impurities, the divertor enrichment was found to mainly depend on the ionisation energy of the impurity under consideration. In the area of MHD physics, analysis of the hot plasma core motion in sawtooth crashes showed good agreement with nonlinear 2-fluid simulations. This indicates that the fast reconnection observed in these events is adequately described including the pressure gradient and the electron inertia in the parallel Ohm’s law. Concerning disruption physics, a shattered pellet injection system was installed in collaboration with the ITER International Organisation. Thanks to the ability to vary the shard size distribution independently of the injection velocity, as well as its impurity admixture, it was possible to tailor the current quench rate, which is an important requirement for future large devices such as ITER. Progress was also made modelling the force reduction of VDEs induced by massive gas injection on AUG. The H-mode density limit was characterised in terms of safe operational space with a newly developed active feedback control method that allowed the stability boundary to be probed several times within a single discharge without inducing a disruptive termination. Regarding integrated operation scenarios, the role of density peaking in the confinement of the ITER baseline scenario (high plasma current) was clarified. The usual energy confinement scaling ITER98(p,y) does not capture this effect, but the more recent H20 scaling does, highlighting again the importance of developing adequate physics based models. Advanced tokamak scenarios, aiming at large non-inductive current fraction due to non-standard profiles of the safety factor in combination with high normalised plasma pressure were studied with a focus on their access conditions. A method to guide the approach of the targeted safety factor profiles was developed, and the conditions for achieving good confinement were clarified. Based on this, two types of advanced scenarios (‘hybrid’ and ‘elevated’

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗

DOE OSTI2025

Calibration of RAFM Micromechanical Model for Creep Using Bayesian Optimization for Functional Output

A Bayesian optimization procedure is presented for calibrating a multimechanism micromechanical model for creep to experimental data of F82H steel. Reduced activation ferritic martensitic (RAFM) steels based on Fe(8–9)%Cr are the most promising candidates for some fusion reactor structures. Although there are indications that RAFM steel could be viable for fusion applications at temperatures up to 600°C, the maximum operating temperature will be determined by the creep properties of the structural material and the breeder material compatibility with the structural material. Due to the relative paucity of available creep data on F82H steel compared to other alloys such as Grade 91 steel, micromechanical models are sought for simulating creep based on relevant deformation mechanisms. As a point of departure, this work recalibrates a model form that was previously proposed for Grade 91 steel to match creep curves for F82H steel. Due to the large number of parameters (9) and cost of the nonlinear simulations, an automated approach for tuning the parameters is pursued using a recently developed Bayesian optimization for functional output (BOFO) framework (Huang et al., 2021, “Bayesian optimization of functional output in inverse problems,” Optim. Eng., 22, pp. 2553–2574). Incorporating extensions such as batch sequencing and weighted experimental load cases into BOFO, a reasonably small error between experimental and simulated creep curves at two load levels is achieved in a reasonable number of iterations. In conclusion, validation with an additional creep curve provides confidence in the fitted parameters obtained from the automated calibration procedure to describe the creep behavior of F82H steel.

42 ENGINEERING↗

DOE OSTI2016

Grizzly: A Multi-scale and Multi-Physics Tool to Model Aging of Nuclear Power Plant Components

Grizzly is a multi-scale multi-physics tool being developed at Idaho National Laboratory (INL) as part of the US Department of Energy's Light Water Reactor Sustainability program to provide improved safety assessments of systems, components, and structures in nuclear power plants subjected to age-related degradation. Its goal is to provide an improved scientific basis for decisions surrounding license renewal, which would permit operation of commercial nuclear power plants beyond 60 years. Grizzly is based on INL's MOOSE framework, which enables multi-physics simulations in a massively parallel computing environment. The reactor pressure vessel (RPV) has been chosen as the initial application for Grizzly. The primary concern for the RPV is that it becomes increasingly susceptible to fracture initiation at the sites of pre-existing flaws over time. This is because exposure to irradiation and high temperatures embrittles the steel. Assessing the safety of an RPV under transient conditions involves simulations at multiple scales. At the engineering scale, strongly coupled equations of heat conduction and solid mechanics are solved to simulate the global response of the RPV to accident conditions. Sub-models are used to represent regions with pre-existing flaws and the stress intensity factors are evaluated from the domain integrals. Crack propagation is then determined using a model that provides the temperature and neutron fluence dependent fracture toughness of the RPV steel. The ability of Grizzly to provide accurate engineering scale estimates of RPV safety beyond 60 years of operation strongly relies on the predictability of the embrittlement model. Multiple models for embrittlement have been proposed, and the physically based empirical model is implemented in Grizzly. However, the formation of late blooming phases greatly reduces the confidence on the extrapolated values obtained from embrittlement model, which is further exacerbated by the absence of real field data for validation. Experimentally validated lower length scale models can significantly improve the predictability of the embrittlement model beyond the calibrated range and the development of such models is another ongoing effort in Grizzly. The modeling activity in Grizzly spans across different length scales, and can be divided into: (i) Irradiation induced microstructure evolution at the atomistic and single crystal length scale using molecular dynamics, atomistic kinetic Monte-Carlo, phase-field and rate theory; (ii) Effect of the irradiation induced defects on the flow stress and fracture processes at the polycrystalline scale using crystal plasticity. Subsequently, the prediction from these models can be used to obtain irradiation induced ductile-to-brittle transition temperature (DBTT) shifts from specimen-scale simulations; (iii) Global thermo-mechanical response of the RPV under normal and accident scenarios, and probabilistic risk analyses at the engineering scale using the DBTT shifts obtained from the lower length scale models. The item (i) has been discussed in a previous report. In this extended abstract an overview of items (ii) and (iii) is provided. The extended abstract summarizes the current status of the modeling effort under Grizzly. At the engineering scale, reduced order model development is underway using the statistical tool RAVEN. This will allow probabilistic analysis considering a wide range of flaw topologies in a computationally efficient manner. Coupling the irradiation induced defect growth predictions to crystal plasticity is also in progress. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2022

The Status and Ambitions of the US Heavy Element Program

The aim of this whitepaper is to highlight the current capabilities and priorities of the US Heavy Element community and to provide the framework for a coordinated advancement of nuclear science from these studies. This is an organized effort to reflect on what has been achieved in the field given the recommendations and initiatives of the 2015 NSAC Long Range Plan, and on what can be realized in the next decade given current and possibly expanded investments. Current investments have positioned the US community to be among the world leaders in studies of the nuclear and chemical properties of the heaviest elements. These include studies of reaction mechanisms, moving us ever closer to the “island of stability”, in spectroscopy, allowing us to better understand nuclear structure at these extreme proton numbers, in chemical behavior, looking to determine how these elements should be placed on the Periodic Table, in performing the first measurements where isotopes are directly identified by their mass numbers, and in laying the foundation for a potential US-led new element discovery experiment. At present, the highest priority of the US Heavy Element community is to capitalize on the current investments by supporting the operations of US facilities at optimal values. These facilities include the Argonne Tandem Linac Accelerator System at Argonne National Laboratory and other Department of Energy facilities such as the 88-Inch Cyclotron at Lawrence Berkeley National Laboratory, which has a dedicated superheavy element program, as well as university laboratories, including Texas A&M University. The High Flux Isotope Reactor at Oak Ridge National Laboratory is crucial to providing the radioactive isotopes required for heavy element science targets. This facility should be supported to provide the actinide materials that are essential for US-based science. Production of stable, rare isotopes for beam material, including 48 Ca, 50 Ti, 54 Cr and 58 Fe, at the Stable Isotope Production and Research Center is critical to continued research in heavy element science and should be a priority. The continued development of targets for heavy element science and retaining US-based expertise is critical for the heavy element community. This is an area that is currently under pressure. For example, the target laboratory at Argonne National Laboratory serves a broad community and is currently under threat due to loss of critical personnel. The skills needed to make targets for nuclear science and develop new targetry methods need to be supported long term at Argonne, Oak Ridge and Lawrence Livermore National Laboratories as well as maintaining the programs at Oregon State University and San José State University as vital pipelines for training students. Advances in theory are the foundation to understand how nuclei behave and to predict those behaviors in new circumstances. Progress in these studies will necessitate continued and new investment and access to high-performance computing. The future health of the heavy element field is dependent on the continuous support of talented early-career professionals at all levels. It is critical that opportunities continue to be created for the next generation to become established in heavy element research so that we can ensure the field is attracting and retaining the best minds for continued success. It is also clear that to ensure diversity of ideas, perspectives and techniques, we need to recruit diverse personnel that are trained at the best facilities. The heavy element community is in support of continued investment to programs with initiatives in diversity, equity, and inclusion. Looking to the next decade of research, support needs to maintain and grow US leadership in heavy element science. Specifically, new investments in state-of-the-art instrumentation will be essential to scientific development of the field and in expanding scientific knowledge. Advances in the next generation of electron cyclotron resonance ion sources, multi-reflection time-of-flight devices, laser spectroscopy, trapping methods and next generation alpha and gamma spectroscopy systems should be prioritized.

38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗

DOE OSTI2017

Development of the Advanced D1S for Shutdown Dose Rate Calculations in Fusion Reactors

The assessment of the shutdown dose rate due to neutron activation is fundamental for shielding design, materials selection, licensing and planning of maintenance operations in high performances fusion devices. The shutdown dose rate calculations require the combined use of radiation transports and inventory codes. The typical approach consists in a two steps procedure. A pure neutron transport simulation is performed in a first step to calculate the spatial distribution of neutron spectra. The results are used as input for an activation calculation that provides the decay gamma source used in a second step for a pure gamma transport simulation (rigorous-two step, R2S). The Direct 1-Step (D1S) approach has been developed more than ten years ago by ENEA and the ITER team: this technique relies on a coupled neutron-decay gamma transport in a single simulation. It is based on the use of a modified version of the MCNP5 Monte Carlo code with specially tailored nuclear cross-section data for transport and FISPACT inventory code for temporal behaviour. The 'Advanced-D1S' is an improved version of D1S in which additional computational features have been introduced, such as the mesh tally capability and continuous time evolution assessment of the dose rate in single run dose. It has been validated through measurements at JET tokamak and comparison with R2S codes and a future benchmark experiment during next high performance deuterium-tritium campaign is in preparation at JET. It has also been extensively used for ITER calculations and recently applied to DEMO assessment. Present limitations and in progress code improvement strategy are described and discussed in this paper.

61 RADIATION PROTECTION AND DOSIMETRY↗

DOE OSTI2017

Development of Bulk Nanocrystalline Tungsten Alloys for Fusion Reactor Structures

This project developed a technology for manufacturing bulk ultrafine tungsten materials that are at or near full density for fusion reactor structural applications, aiming to improve ductility and toughness of tungsten before and after irradiation. The project involved the development of fabrication processes for making bulk ultrafine grained W, the development of new alloys of ultrafine grained W and evaluations of properties of these specific materials. The goal of this fabrication process is to produce fully dense bulk W with ultrafine grain sizes, with uniform distributions of grain size and additives. To date there is no known process that could be used to make ultrafine grained tungsten in a fully dense state and in a cost-acceptable fashion. The specific technology described in this proposal for making ultrafine grained tungsten involves a suite of nano-particle processing and sintering techniques. The program also developed new alloys of ultrafine grained W, e.g. W-(Ta,V,Ti)-TiC alloys to improve ductility and toughness before and after irradiation. By completing this project, we achieved the following objectives: • Demonstrated experimentally the feasibility of producing bulk ultrafine grained tungsten alloys (at or near 100% dense, <1000 nm grain size) using the proposed process • Demonstrated the proposed ultrafine grained W alloys, namely, W-(Ta, V, Ti)-TiC, can indeed be made using the proposed process • Demonstrated that the properties of nano tungsten alloys meet the requirements for fusion reactor applications. The overall goal was to harness the potential of ultrafine grained W produced using the proposed processes as the core structural materials for future fusion reactors. The project was very successful overall, meeting all milestones and surpassing project goals in terms of process development and material’s blistering resistance properties. A novel process similar to the conventional press-and-sinter powder metallurgy method was developed for producing ultrafine grain tungsten from nanosize tungsten powders. Grain growth was significantly controlled during sintering by certain alloy compositions, particularly Ti, and most compositions sintered to maximum densification. To optimize this process, the effect of processing parameters on the densification and grain growth of nano-W powders was investigated. Near-fully densified tungsten was obtained at sintering temperatures between 1100 and 1300 °C, and both Ar and H2 sintering atmospheres were investigated. The Ar sintering atmosphere was determined to more favorably promote densification and minimize grain growth. The nanosized tungsten powder compacts were subjected to reduction in H2 as a part of the sintering cycle. The reduction temperature was found to have significant effects on the sintering of nano-W powder, primarily as a result of grain coarsening, which was seen at temperatures as low as 700 °C. In an effort to inhibit grain growth, the effect of Ti-based additives on the densification and grain growth of nano-W powders was investigated in this project. The addition of 1 wt.% Ti into tungsten led to more than a 63% decrease in average grain size of sintered samples at comparable density levels. Compared to conventional high temperature sintering, a lower temperature sintering cycle for a longer hold time resulted in both near-full density and fine grain size. The roles of the Ti additives include not only the inhibition of grain growth, but also the potential absorption of oxygen from W particles. The project has resulted in the publication; thus far, of six peer reviewed journal articles and seven conference presentations, as well as a master’s thesis. Two additional journal articles are currently in preparation. Presentations and articles were a particular focus of the second half of the project, once significant experimentation had been performed and analyzed. As part of our efforts to disseminate information of our results, the W research teams with Prof. Fang had a strong presence at multiple international conferences during 2015 and 2016. Several research groups in the US are now performing experiments using the ultrafine grained W materials.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗

DOE OSTI2016

A Methodology for Loading the Advanced Test Reactor Driver Core for Experiment Analysis

The Advanced Test Reactor (ATR) at the Idaho National Laboratory (INL) is one of the premier irradiation facilities in the world. The ATR hosts experiments from a variety of laboratories, companies, and disciplines. Because of its flexibility in providing customized irradiation conditions to experiments, such as coolant chemistry, temperature, and reactor power, analysis for experiments, both from a programmatic and safety perspective, can be a challenging endeavor. Because of these experiment variables, there is always an effort to improve upon the existing analysis methodology in order to provide the sponsor of any given experiment with the most accurate predictions possible. The paper discusses the effort to improve upon the methodology for loading the ATR driver core for experiment analysis. Experiment analysis at INL is generally broken up into three separate disciplines: neutronics, thermal-hydraulics, and structural. The focus of this paper is neutronics analysis. After the experiment has a preliminary design, the neutronics analyst will create a model in the Monte Carlo N-Particle (MCNP) transport code to calculate fuel and material heating rates from MCNP tallies and provide those to the thermal-hydraulics analyst, and calculate reactivity worths, perturbation on the axial fission profile of the ATR driver core by the experiment, void worths, and experiment backup worths. Also, there are usually several programmatic parameters that the experiment sponsor requests that the neutronics analyst calculates. These can include heat generation rates, burnup and displacements per atom (DPA). In order to accurately analyze these parameters, an important input into the physics analyst's MCNP model is the loading of the ATR driver core. The ATR driver core consists of 40, 19-plate highly-enriched uranium (HEU) fuel elements, arranged in a serpentine shape, as pictured in Figure 1. The ATR driver core is responsible for the vast majority of the thermal power of the reactor and the nuclear interactions with the experiments being irradiated in ATR. A unique feature of ATR is the 'power tilt' across the reactor. A power tilt means that the Northwest, Northeast, Southwest, and Southeast Lobes of ATR frequently operate at different powers, providing different irradiation conditions for the experiments in and around the lobes of the reactor. In order to achieve lobe power splits, different fuel elements for each of the 40 positions are chosen by ATR Reactor Engineering so that lobe power splits can be maintained during an operating cycle. A typical fuel loading for an ATR cycle consists of 14 to 18 new fuel elements and 2 to 26 depleted fuel elements. These loadings are typically chosen from three different types of fuel elements. All three ATR fuel element types share a common design, spelled out in Chapter 4 of the ATR Safety Analysis Report (SAR). The fuel element meat is HEU UAlx dispersion fuel sandwiched between aluminum cladding with a length of 121.92 cm (4 ft), with 19 plates per fuel element. The three variations of this common design are designated by the first two letters in their serial numbers: XA, YA, and NB. XA elements are most commonly used. These elements have nominally 1075 g of uranium-235 and 0.66 g of boron-10. The NB elements have the same uranium-235 loading as the XA elements, but no boron in any plates. These elements are used when excess reactivity is needed. The YA elements have the same loading as the XA elements, except that the outermost plate (Plate 19) has no boron or uranium. These elements are used to mitigate aged reflector damage. In order to accurately model the ATR driver core, all three types of fuel elements must be accounted for in the MCNP model. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2015

Fission Product Sorptivity in Graphite

Both adsorption and absorption (sorption) of fission product (FP) gases on/into graphite are issues of interest in very high temperature reactors (VHTRs). In the original proposal, we proposed to use packed beds of graphite particles to measure sorption at a variety of temperatures and to use an electrodynamic balance (EDB) to measure sorption onto single graphite particles (a few μm in diameter) at room temperature. The use of packed beds at elevated temperature is not an issue. However, the TPOC requested revision of this initial proposal to included single particle measurements at elevated temperatures up to 1100 °C. To accommodate the desire of NEUP to extend the single particle EDB measurements to elevated temperatures it was necessary to significantly revise the plan and the budget. These revisions were approved. In the EDB method, we levitate a single graphite particle (the size, surface characteristics, morphology, purity, and composition of the particle can be varied) or agglomerate in the balance and measure the sorption of species by observing the changes in mass. This process involves the use of an electron stepping technique to measure the total charge on a particle which, in conjunction with the measured suspension voltages for the particle, allows for determinations of mass and, hence, of mass changes which then correspond to measurements of sorption. Accommodating elevated temperatures with this type of system required a significant system redesign and required additional time that ultimately was not available. These constraints also meant that the grant had to focus on fewer species as a result. Overall, the extension of the original proposed single particle work to elevated temperatures added greatly to the complexity of the proposed project and added greatly to the time that would eventually be required as well. This means that the bulk of the experimental progress was made using the packed bed sorption systems. Only being able to recruit one graduate student meant that data acquisition with the packed bed systems ended up competing for the graduate student’s available time with the electrodynamic balance redesign and assembly portions of the project. This competition for available time was eventually mitigated to some extent by the later recruitment of an undergraduate student to help with data collection using the packed bed system. It was only the recruitment of the second student that allowed the single particle balance design and construction efforts to proceed as far as they did during the project period. It should be added that some significant time was also spent by the graduate student cataloging previous work involving graphite. This eventually resulted in a review paper being submitted and accepted. Our specific revised objectives in this project were as follows: Experimentally obtain isotherms of Iodine for reactor grade IG-110 samples of graphite particles over a range of temperatures and pressures using an EDB and a temperature controlled EDB; Experimentally obtain isotherms of Iodine for reactor grade IG-110 samples of graphite particles over a range of temperatures and pressures using a packed column bed apparatus; Explore the effect that charge has on the adsorption isotherms of iodine by varying the charges on and the voltages used to suspend the microscopic particles in the EDB; and To interpret these results in terms of the existing models (Langmuir, BET, Freundlich, and others) which we will modify as necessary to include charge related effects.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗

DOE OSTI2023

POWTEX visits POWGEN

The high-intensity time-of-flight (TOF) neutron diffractometer POWTEX for powder and texture analysis is currently being built prior to operation in the eastern guide hall of the research reactor FRM II at Garching close to Munich, Germany. Because of the world-wide 3 He crisis in 2009, the authors promptly initiated the development of 3 He-free detector alternatives that are tailor-made for the requirements of large-area diffractometers. Herein is reported the 2017 enterprise to operate one mounting unit of the final POWTEX detector on the neutron powder diffractometer POWGEN at the Spallation Neutron Source located at Oak Ridge National Laboratory, USA. As a result, presented here are the first angular- and wavelength-dependent data from the POWTEX detector, unfortunately damaged by a 50 g shock but still operating, as well as the efforts made both to characterize the transport damage and to successfully recalibrate the voxel positions in order to yield nonetheless reliable measurements. Also described is the current data reduction process using the PowderReduceP2D algorithm implemented in Mantid [Arnold et al. (2014). Nucl. Instrum. Methods Phys. Res. A , 764 , 156–166]. The final part of the data treatment chain, namely a novel multi-dimensional refinement using a modified version of the GSAS-II software suite [Toby & Von Dreele (2013). J. Appl. Cryst. 46 , 544–549], is compared with a standard data treatment of the same event data conventionally reduced as TOF diffraction patterns and refined with the unmodified version of GSAS-II . This involves both determining the instrumental resolution parameters using POWGEN's powdered diamond standard sample and the refinement of a friendly-user sample, BaZn(NCN) 2 . Although each structural parameter on its own looks similar upon comparing the conventional (1D) and multi-dimensional (2D) treatments, also in terms of precision, a closer view shows small but possibly significant differences. For example, the somewhat suspicious proximity of the a and b lattice parameters of BaZn(NCN) 2 crystallizing in Pbca as resulting from the 1D refinement (0.008 Å) is five times less pronounced in the 2D refinement (0.038 Å). Similar features are found when comparing bond lengths and bond angles, e.g. the two N—C—N units are less differently bent in the 1D results (173 and 175°) than in the 2D results (167 and 173°). The results are of importance not only for POWTEX but also for other neutron TOF diffractometers with large-area detectors, like POWGEN at the SNS or the future DREAM beamline at the European Spallation Source.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗

DOE OSTI2025

Accelerated Irradiation Testing and Post-Irradiation Characterization: U.S.-Based Capabilities for Advanced Nuclear Systems and Radioisotope Production

Irradiation experiments and post-irradiation examinations, together referred to as irradiation testing (IRT), are prerequisites for nuclear fuel and material qualification for the deployment of new and advanced reactors, as well as radioisotope production, thereby ensuring regulatory compliance. Qualified research and test reactors (RTRs) and testing facilities are essential to enable IRT to verify performance and safety under prototypical reactor conditions. In the past, qualification of new fuels or structural materials required about 20 years. Synergist strategies, advanced tools, and qualified methods are needed to greatly reduce this timeframe of IRT and radioisotope production. This study, termed accelerated-IRT, focuses on identifying gaps and leveraging U.S.-based RTRs and material testing capabilities, leveraging the preliminary evaluation and qualification of selected RTRs to provide a generic as well as specific-case solution paths forward, ensuring adherence to stringent regulatory standards. Furthermore, IRT and radioisotope production utilizing qualified RTRs necessarily includes modeling and simulation to support the design (i.e. neutronics, thermal, and structural aspects) and manufacturing of irradiation test specimens, vehicles, capsules, apparatuses, and flow loops. In addition, IRT can be improved by applying advanced manufacturing techniques and in-pile sensors and instrumentation, as discussed in this study.

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2023

Time dependent model for non-inductive ECH X-I current ramp-up for SHPD tokamak facility

Non-inductive (NI) plasma current start-up and ramp-up is an important research topic for spherical tokamak (ST) based reactors and fusion pilot plant (FPP). For a compact FPP, the OH flux availability is highly restricted due to its compact geometry. Efficient fundamental extraordinary mode (X-I) electron cyclotron heating (ECH) current start-up and ramp-up regime was identified for a reactor-like high toroidal magnetic field range which has more than a hundred times higher current drive efficiency compared to more conventional ECH methods for the relevant start-up temperature range. High current drive efficiency is possible due to the strong X-I fundamental ECH interaction only with unidirectional passing electrons constrained by the wave accessibility conditions. Here, we extend the X-I electron cyclotron current drive (ECCD) investigation to a time dependent model to simulate the non-inductive current ramp-up to 10 MA for the Sustained High-Power Density (SHPD) facility. As the X-I ECCD driven current I EC rises, due to the back EMF driven negative current, the net plasma current Ip rises more slowly with the current resistive time scale. For tokamak confinement time (both L-mode and H-Mode) which tends to rise with I p , a positive feedback results and even with constant applied ECH power, T e0 , I EC , and I p can continue to rise to very high values. However, in a realistic situation, the T e0 rise should saturate due to a number of factors such as enhanced core radiation and increased power loss at high temperature. To simulate this effect, we adopt a maximum T e0 model which would limit the temperature rise to certain T e0 . With this model, we investigated the current ramp-up for various ECH power levels and the maximum T e0 of 15, 20, and 25 keV. We find that while the power required is reduced with increasing T e0 limit due to increased current drive efficiency, the time to reach 10 MA tends to go up due to the reduced plasma resistivity for the higher T e0 limit. We also find for a given Te0 limit, the time to reach 10 MA tends to be reduced by increasing the applied ECH power by over driving the current ramp-up where I EC is driven at significantly higher level than 10 MA. While the current ramp-up time may not be an issue for the steady-state reactor systems, if it is desirable to minimize the current ramp-up time, it is prudent to have a sufficient ECH power for current over-drive and have some T e0 limiting tools such as impurity seeding for enhanced radiation. In conclusion, a well-controlled NI ECH start-up also has a potential of improving the tokamak start-up reliability and avoid run-away electrons while the NI off-axis current drive could enhance MHD stability and plasma performance improvements.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗

DOE OSTI2026

Machine-Learning-Driven Discovery of Water Splitting BaFe 2 O 4 and Human-in-the-Loop Improvement via Al-Substitution for Increased Thermal Stability

Thermochemical hydrogen (TCH) production offers a promising method for converting thermal energy into hydrogen fuel through heat-driven redox cycles of metal oxides. Here, in this work a defect graph neural network (dGNN) was used to predict oxygen vacancy formation energies ΔH V O combined with Materials Project predictions of oxygen chemical potential stability to screen candidate oxides via high-throughput database analysis. BaFe 2 O 4 was identified as a promising material for experimental validation based on its predicted ΔH V O , oxygen chemical potential stability range, and potential for tunable substitutions to improve thermal properties. Experimental validation using thermogravimetric analysis (TGA), stagnation flow reactor (SFR), X-ray diffraction (XRD), and electron microscopy confirmed positive water-splitting behavior but also revealed limitations in thermal stability under aggressive reduction conditions. To address this, a human-in-the-loop modification strategy was employed introducing Al substitution in BaFe 2–x Al x O 4 ; this modification improves thermal stability, alters the crystal structure and enhances overall performance. These results demonstrate a combined computational and experimental workflow in which machine learning accelerates identification of promising candidates, while targeted experimental design enables optimization of functional performance. This approach advances the development of robust, cost-effective TCH materials and highlights the importance of integrating data-driven discovery with human-guided materials design in paving the way for scalable hydrogen production technologies.

organic↗

DOE OSTI2022

Disruption prediction with artificial intelligence techniques in tokamak plasmas

In nuclear fusion reactors, plasmas are heated to very high temperatures of more than 100 million kelvin and, in so-called tokamaks, they are confined by magnetic fields in the shape of a torus. Light nuclei, such as deuterium and tritium, undergo a fusion reaction that releases energy, making fusion a promising option for a sustainable and clean energy source. Tokamak plasmas, however, are prone to disruptions as a result of a sudden collapse of the system terminating the fusion reactions. As disruptions lead to an abrupt loss of confinement, they can cause irreversible damage to present-day fusion devices and are expected to have a more devastating effect in future devices. Disruptions expected in the next-generation tokamak, ITER, for example, could cause electromagnetic forces larger than the weight of an Airbus A380. Furthermore, the thermal loads in such an event could exceed the melting threshold of the most resistant state-of-the-art materials by more than an order of magnitude. In this work, to prevent disruptions or at least mitigate their detrimental effects, empirical models obtained with artificial intelligence methods, of which an overview is given here, are commonly employed to predict their occurrence—and ideally give enough time to introduce counteracting measures.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗

DOE OSTI2023

Calibration strategy of the PROSPECT-II detector with external and intrinsic sources

This paper presents an energy calibration scheme for an upgraded reactor antineutrino detector for the Precision Reactor Oscillation and Spectrum Experiment (PROSPECT). The PROSPECT collaboration is preparing an upgraded detector, PROSPECT-II (P-II), to advance capabilities for the investigation of fundamental neutrino physics, fission processes and associated reactor neutrino flux, and nuclear security applications. P-II will expand the statistical power of the original PROSPECT (P-I) dataset by at least an order of magnitude. The new design builds upon previous P-I design and focuses on improving the detector robustness and long-term stability to enable multi-year operation at one or more sites. The new design optimizes the fiducial volume by elimination of dead space previously occupied by internal calibration channels, which in turn necessitates the external deployment. In this paper, we describe a calibration strategy for P-II. Here, the expected performance of externally deployed calibration sources is evaluated using P-I data and a well-benchmarked simulation package by varying detector segmentation configurations in the analysis. The proposed external calibration scheme delivers a compatible energy scale model and achieves comparable performance with the inclusion of an additional AmBe neutron source, in comparison to the previous internal arrangement. Most importantly, the estimated uncertainty contribution from the external energy scale calibration model meets the precision requirements of the P-II experiment.

47 OTHER INSTRUMENTATION↗

DOE OSTI2022

Specifications of EBR-II Neutron Radiography Method Description and Digitization Approach

All neutron radiography (NRAD) images of fuel pins in Argonne’s collection were originally generated using the NRAD imaging facility established in the Hot Fuel Examination Facility (HFEF) at Idaho National Laboratory (INL). The NRAD reactor facility was built in 1977 and has been operating since. The reactor is a TRIGA-type reactor operating at a power level of 250 kWth to provide a neutron source for radiography imaging. The reactor is equipped with two beam tubes (i.e., east beam tube and north beam tube) to guide the neutron beams to two radiography stations. The east radiography station is directly under the HFEF main cell and is dedicated for specimens already in the HFEF hot cell. The north radiography station is outside of the main HFEF hot cell and allows NRAD imaging of non-irradiated items. The NRAD images of EBR-II irradiated metallic fuel pins were taken in the east radiography station. Thermal neutrons have the capability to transmit through most materials and are ideal for NRAD imaging. However, because of their high thermal neutron absorption cross-section, fissile materials (e.g., highly-enriched nuclear fuels) may not be as transmissible to thermal neutrons. This is also the case for oversize specimens with extraneous thickness. Epithermal neutron imaging is therefore used as a complement to thermal neutron NRAD imaging. At HFEF’s NRAD facility, both thermal and epithermal neutrons can be used for NRAD imaging. Irradiated nuclear fuels emit high levels of γ radiation that can easily darken X-ray films, so direct exposure NRAD cannot be used to image them. Instead, an indirect NRAD imaging method was developed at HFEF’s NRAD facility. In this method, foils made of materials that can be activated by neutrons (i.e., with large neutron absorption cross section) are used to collect transmitted neutron signals. Then the activated foils are then placed against X-ray films and enclosed in a vacuum cassette so that the γ decay from the activated foils can produce images on the X-ray films. Then, general X-ray film processing procedures are used to digitize and store the images. By using different foil materials, different energy neutrons can be used for NRAD imaging. At the HFEF NRAD station, two types of films are commonly used: dysprosium (Dy) foils with thickness of 130 microns are used to capture thermal neutron signal, while indium (In) foils with thickness of 130 microns are used to capture epithermal neutron signal. A cadmium or gadolinium foil is put before the indium foil to work as a thermal neutron filter. The thermal and epithermal NRAD images can be taken simultaneously by using a Dy/Cd/In sandwiched foil combination. The typical NRAD exposure time is approximately 20 minutes. Then the exposed foils are transferred to film vacuum cassettes. The vacuum ensures that there is no gap between the foil and the film. The foil-to-film exposure time is at least three half-lives of the corresponding radioisotopes, which are 3 hours for In and 7.5 hours for Dy, respectively. Exposed films are processed using an automatic film processor to produce completed NRAD images.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2016

Surrogate Post-Detonation Urban Debris (SPUD): Standard Reference Materials 4600 and 4601

If a terrorist nuclear explosion were to occur in the United States, high-quality (accurate and precise) legally defensible nuclear forensics measurements will be required to assure the certainty and confidence needed, for conclusions drawn from these data will be coupled with law enforcement and intelligence information for the purposes of attribution and retribution. Standard Reference Materials (SRMs)/Certified Reference Materials (CRMs) establish the traceability, accuracy, and precision of nuclear forensic measurements and provide the ability to benchmark measurement methods essential to meeting the requirements of legal scrutiny. Currently there is a critical shortage of key SRMs/CRMs required for accurate and precise measurements that are fundamental to nuclear measurements supporting pre- and post-detonation nuclear forensics. Supplies either do not exist, are not good enough, are in short supply, and/or are very expensive. The radionuclide measurement techniques and data quality objectives for pre- and post-detonation nuclear forensics are quite different than the applications for which the current suite of available SRMs/CRMs was developed. Specifically, 1) modern mass spectrometry methods have largely replaced traditional activity-based measurement methods for quantifying small quantities of long-lived radionuclides and most trace radionuclide SRMs/CRMs are certified for activity and are thus not adequate for current applications, 2) modern instrumental methods are capable of higher precision than historic measurement techniques, and in some cases exceed the precision of the available SRMs/CRMs, and 3) the quality assurance requirements for pre- and post-detonation nuclear forensics are such that the existing cadre of SRMs/CRMs does not adequately support the needs for method validation. There are numerous governmental stakeholders both within the US (DOE, DoD) and internationally (IAEA, JRC-ITU) whose missions are either directly involved with nuclear forensics measurements or rely on the forensic data obtained to make the exacting policy decisions (DOJ (FBI), DHS, CIA, DIA, DoD) associated with disrupting the production of illicit nuclear materials, interdicted nuclear materials or a nuclear/radiological detonation event. Starting in 2008 the DHS NTNFC, in collaboration with NBL and NIST, took on the production of suitable SRMs/CRMs for the pre-detonation nuclear forensics community. Two CRMs have been released since the start of this effort and two more are slated for completion in 2016. More recently, NIST, as a part of its mission as the US National Metrology Institute (NMI) to deliver standards (providing traceability) and foster homeland security, has launched the first certification campaign of a measurement traceable post nuclear detonation SRM. NIST, in cooperation with partner labs (AFIT (DoD), FBI (DOJ), and NPL) and with support from the FBI, has developed two Surrogate Post-Detonation Urban Debris (SPUD) Standard Reference Materials (SRMs) - SRM 4600 and 4601. These SRMs have been developed to simulate the expected debris mixture from the detonation of an Improvised Nuclear Device (IND) in an urban environment, which is capable of producing fresh fission as well as activation products. The elemental composition of the vitrified debris materials is a composite of minerals and metals characteristic of the concrete and steel, the building materials that would be exposed to a nuclear detonation and whose resulting activation products would be found in rubble samples collected for forensics evidence and attribution. Additionally, because post-detonation samples would also contain short-lived fission products, the debris materials have been doped with Uranium (U). To create the nuclear fission products, samples of the SPUD materials can be placed in a well-characterized nuclear reactor and exposed to a set neutron flux and energy, causing the encased U atoms to split and yield short-lived fission products. Because a known amount of U is added to the SPUD materials it is feasible to determine the concentration of the resulting fission products. The subsequent irradiated SPUD glass would be chemically similar to samples that would be generated in an actual detonation, and thus can serve as a standardized reference material for nuclear forensic post detonation measurements. (authors)

61 RADIATION PROTECTION AND DOSIMETRY↗