Reactor Fleet Research Articles

Simulating self-powered neutron detector responses to infer burnup-induced power distribution perturbations in next-generation light water reactors

Understanding how 3D power distribution will be monitored throughout reactor core volumetric space in next-generation nuclear power reactors is crucial to the design, deployment, and licensing of these reactors. Although numerous techniques exist for 3D power distribution monitoring based on the response of both in situ and ex situ sensors currently implemented or proposed for use in the US reactor fleet, crucial details about these techniques are often unclear. The publicly available documentation does not include information such as how well these techniques are characterized and optimized in their implementations and the levels of uncertainty in the inferred 3D power distribution. The work described herein investigated a recently developed 3D power distribution inferencing method as applied to two next-generation reactor simulations: (1) the NuScale small modular reactor design and (2) the Westinghouse AP1000 design, both of which contain in-core strings of vanadium self-powered neutron detectors (SPNDs). This investigation considered a range of SPND string sensor densities, as well as a range of 3D power distribution axial segment sizes. SPND response simulation is informed by neutron flux calculations in representative homogenized cores. For the different sensor densities and power distribution axial segment sizes in these simulations, the average solution error, solver iterations, and run time were tracked to parameterize the sensor-core configuration.

Guiding Transformation: The Role of a Design Philosophy

Unlike renewables such as wind and photovoltaics, nuclear power is a carbon-free source of energy that offers reliable, dispatchable baseload energy. This unique characteristic makes nuclear energy an important component of the U.S. mix of carbon-free energy, and thus, a major contributor to achieving the goals laid out in the Paris Agreement. However, the current fleet of nuclear power plants are being outpriced by other energy sources, such as natural gas. One contributor to the high cost of nuclear is the outdated concept of operations. The current fleet of nuclear reactors employs the same concept of operations they started with over half a century ago. Idaho National Laboratory’s (INL’s) Human Factors Engineering (HFE) team under the U.S. Department of Energy’s Light Water Reactor Sustainability Program is engaged in helping the nuclear fleet modernize their control rooms. The goal is to transform nuclear power plant operators’ perceptions to improve efficient and safe plant operation. A major component to successfully transforming an aged control room with advanced technology is to use a design philosophy that guides the modernization effort. This paper discusses design philosophy and the role it plays. Also discussed is the initial approach to design philosophy and adherence to safety and regulatory requirements. Last, a brief discussion of how INL’s HFE team plans to implement a design philosophy that can be used industry wide.

CTF: A modernized, production-level, thermal hydraulic solver for the solution of industry-relevant challenge problems in pressurized water reactors

CTF is a thermal hydraulic (T/H) subchannel tool that has been extensively developed over the past ten years as part of the Consortium for Advanced Simulation of Light Water Reactors (CASL) program. The code was selected early in the CASL program for support of high-impact challenge problems that were found to be relevant to the nuclear industry and its currently operating fleet of pressurized water reactors (PWRs), including issues such as departure from nucleate boiling (DNB), crud-induced power shifts (CIPSs), and reactivity-insertion accidents (RIAs). By incorporating CTF into the multiphysics Virtual Environment for Reactor Application (VERA) core simulator software developed by CASL, CTF has become the primary means of providing fluid and fuel thermal feedback, as well as T/H figure-of-merits (FOMs) in large-scale reactor simulations. With the goal of solving industry challenge problems, CASL placed great emphasis on developing high-quality, high-performance, validated software tools that offer higher fidelity than what is currently possible with current industry methods. In support of this effort, CTF was developed from a research tool into an nuclear quality assurance (NQA-1)–compliant, production-level software tool that is capable of addressing the stated challenge problems and goals of CASL. This paper presents a review of the major technological achievements that were realized in developing CTF over the past decade of the CASL program and presents an overview of the code solution approach and closure models.

Condition assessment of cable insulation materials in advanced reactor environments

While existing cable insulation materials are sufficient for use in the current reactor fleet, environmental conditions inside some advanced reactors, such as small modular reactors, will be substantially harsher. The work herein was conducted to assess the performance and overall survivability of several high-temperature, radiation resistant insulation polymers in a simulated small modular reactor environment. The materials evaluated under this research include polyether ether ketone, polyimide, and silicone rubber. These polymers were subjected to 1000 h of environmental stress exposure under two different conditions: 1) high temperature (i.e., 250 °C) in atmospheric pressure and 2) high temperature in a low-pressure vacuum (i.e., less than 2 psia). During the environmental exposure process, each insulation was periodically tested using several materials characterization techniques including Fourier transform infrared spectroscopy, oxidation induction time, and oxidation induction temperature. The objective of this testing was to assess and monitor changes that occurred in the thermal and chemical properties of the materials during the environmental exposures. The overall survivability of each insulation was assessed based on these property changes. The results of this research revealed that the polyimide and polyether ether ketone insulations had significantly better thermal and chemical stability than the silicone rubber during both the high-temperature atmospheric and vacuum exposures. Both polyimide and polyether ether ketone experienced minimal reductions in their chemical and thermal properties over the course of 1000 h in the simulated environments. Based on this research, polyimide and polyether ether ketone show promise as potential base materials for small modular reactor cable insulation.

Design space exploration for salt-cooled reactor system: Part I – Thermal hydraulic design

Fluoride-salt-cooled High-temperature Reactors (FHRs) are a variant of nuclear reactor offering high power densities and operating temperatures. This paper proposes a new design stream of FHR, leveraging developments and experience from the British Advanced Gas-cooled Reactor (AGR) fleet, the only commercial fleet of high-temperature reactors to-date. The design space for the proposed AGR-like FHR is explored using a thermal-hydraulic simulation and optimisation package, seeking to maximise the reactor’s economic performance. Starting from a salt-cooled AGR assembly and applying reasonable thermal-hydraulic constraints, this exploration includes varying the assembly fuel type and geometry, coolant salt, as well as design of the pumps and heat exchangers. Some of the resulting designs promise substantially higher power densities than AGRs and improved economics.

Silicon Carbide Neutron Detectors for Harsh Nuclear Environments: A Review of the State of the Art

Silicon carbide (SiC) semiconductor is an ideal material for solid-state nuclear radiation detectors to be used in high-temperature, high-radiation environments. Such harsh environments are typically encountered in nuclear reactor measurement locations as well as high-level radioactive waste and/or “hot” dismantling–decommissioning operations. In the present fleet of commercial nuclear reactors, temperatures in excess of 300 °C are often encountered, and temperatures up to 800 °C are anticipated in advanced reactor designs. The wide bandgap for SiC (3.27 eV) compared to more widely used semiconductors such as silicon (1.12 eV at room temperature) has allowed low-noise measurements to be carried out at temperatures up to 700 °C. The concentration of thermally induced charge carriers in SiC at 700 °C is about four orders of magnitude less than that of silicon at room temperature. Furthermore, SiC radiation detectors have been demonstrated to be much more resistant to the effects of radiation-induced damage than more conventional semiconductors, such as silicon, germanium, or cadmium zinc telluride (CZT), and have been demonstrated to be operational after extremely high gamma-ray, neutron, and charged-particle doses. The purpose of the present review is to provide an updated state of the art for SiC neutron detectors and to explore their applications in harsh high-temperature, high-radiation nuclear reactor applications. Conclusions related to the current state-of-the-art of SiC neutron detectors will be presented, and specific ideal applications will be discussed.

Creep deformation phenomena in near-surface carburised layers of 316H stainless steels

The local creep-induced phenomena occurring within the near-surface carburised layer in two pre-carburised type 316H austenitic stainless steels, containing 0.98 or 1.52 wt% Mn, was assessed at 550 °C. The pre-carburisation exposed both steel grades at either 550 °C (8000 h) or 600 °C (3000 h) to simulated CO2-rich gas environment of UK Advanced Gas-cooled Reactors. The local properties within the carburised layer causes a reduction in minimum creep rate as compared to the non-carburised condition. In both steels carburised at 600 °C, ferrite is observed to form in the most heavily carburised regions, including slip bands, during the initial plastic deformation to the target applied stress. The ferrite area fraction further increases with creep strain. Within the ferritic areas, new fine grains are observed in the creep tested specimens and the fraction of high angle grain boundaries is >60%. Those fine grains have a lower degree of lattice strain compared to the surrounding ferrite area and especially to the austenite matrix. The lattice strain localises within the carburised layer, and the formation of ferrite from the strained austenite, and potentially a dynamic recrystallization within the ferrite regions, pinpoints to a dominant strain relief-driven mechanism. Lattice strain localises primarily in the pre-carburised austenite grains and its carbide-decorated boundaries, and induces the continuous formation and refinement of ferrite grains, and the development of intergranular cracks that evolve further into the bulk structure. These results emphasise the necessity to account for environmental-creep interactions at elevated temperatures in structural assessment procedures, in the current gas-cooled fission reactor fleet and beyond into (very) high-temperature Gen IV and advanced modular reactor designs.

Hydrogen Isotopes Permeation in Clean or Unoxidized FeCrAl Alloys: A Review

The US Department of Energy is working with fuel vendors to develop accident tolerant fuels (ATF) for the current fleet of light water reactors (LWRs). The ATF should be more resilient to loss of coolant accident scenarios and help extending the life of the operating LWRs. One of the proposed ATF concepts is to use iron-chromium-aluminum (FeCrAl) alloys for the cladding of the fuel. A concern in using ferritic FeCrAl is that this type of cladding may result in an increase in the concentration of tritium in the coolant. The objective of the current critical review is to collect and assess information from the literature regarding diffusion or permeation of hydrogen (H) and its isotopes deuterium (D) and Tritium (T) across industrial alloys (including FeCrAl) used or intended for the nuclear industry. Over a hundred years of data reviewed shows that the solubility of hydrogen in ferritic alloys is lower than in austenitic alloys but hydrogen permeates faster through a ferritic material than through austenitic materials. The tritium permeation rates in FeCrAl alloys are between those in austenitic stainless steels and in ferritic FeCr steels. The activation energy for hydrogen permeation is approximately 30 pct higher in the austenitic alloys compared with the ferritic (typically ∼ 50 kJ/mol in ferritic vs. ∼ 65 kJ/mol in the austenitic). None of the major elements in FeCrAl alloys react with hydrogen to form detrimental hydride phases. The effect of surface oxides on FeCrAl delaying hydrogen entrance into FeCrAl alloy is not part of this review.

A new tool for the simulation of different nuclear fleets at equilibrium

Scenario simulations are the main tool for studying the impact of a nuclear reactor fleet on the related fuel cycle facilities. This equilibrium preliminary study aims to present the functionalities of a new tool and to show the wide variety of reactors/cycles/strategies that can be studied in steady state conditions and validated with more details thanks to dynamic code. Different types of scenario simulation tools have been developed at CEA over the years, this study focuses on dynamic and equilibrium codes. Dynamic fuel cycle simulation code models the ingoing and outgoing material flow in all the facilities of a nuclear reactor fleet and their evolutions through the different nuclear processes over a given period of time. Equilibrium fuel cycle simulation code models advanced nuclear fuel cycles in equilibrium conditions, i.e. in conditions which stabilize selected nuclear inventories such as spent nuclear fuel constituents, plutonium or some minor actinides. The principle of this work is to analyze different nuclear reactors (PWR, AMR) and several fuel types (UOX, MOX, ERU, MIX) to simulate advanced nuclear fleet with partial and fully plutonium and uranium multi-recycling strategies at equilibrium. At this first stage, selected results are compared with COSI6 simulations in order to evaluate the precision of this new tool, showing a significant general agreement.

Comparative study of a stationary two-component nuclear energy system with light water and fast breeder reactors versus one-component one with self-sufficient non-breeding fast neutron reactors

Russia is currently in its initial phase of transitioning to a two-component nuclear power structure, consisting of nuclear power plants with thermal and fast reactors operating in a closed nuclear fuel cycle. Alternative approaches to closing the nuclear fuel cycle often have different assumptions regarding the role of fast reactors in this new structure. They can either imply that fast reactors are self-sufficient and are economically viable in such a way, that they can completely replace existing thermal reactor technology, or that they are used for breeding fuel for the whole nuclear energy system and are essentially complementary to VVER reactors in a permanent two component symbiotic structure. The results of this study show that nuclear power scenarios favoring the development of a large-scale fast reactor fleet with a breeding ratio ∼1 without any additional VVER type component in the nuclear power structure is optimal from an economic standpoint.

Evaluation of thorium-based nuclear fuel breeding performance of a fast neutron irradiator based on a low-aspect ratio tokamak

Fast neutron irradiation has the capability of transforming fertile material, such as 238U or 232Th, into fissile material that can be used to manufacture fuel for conventional thermal light-water fission reactors. This would help both expand and extend the lifetime of nuclear power, which is currently constrained to natural uranium resources, which could only last between 50 and 150 years depending on the growth of the nuclear reactor fleet. In this paper, a design for a fast neutron source based on a low-aspect ratio tokamak is evaluated, based on two figures of merit: the tritium breeding ratio (TBR), a measure of tritium self-sufficiency, and the reactor support ratio (RSR), related to the speed of fissile material breeding. The system has neutron multiplying regions to ensure there are enough neutrons to breed both tritium and fissile material; it also has tritium breeding regions that contain 6Li which, when bombarded with neutrons, produce tritium. The MCNP code is used to perform neutronics simulations and obtain values for energy-resolved neutron fluxes for the relevant regions of the system, and the ORIGEN code is used to perform kinetic calculations for nuclear reactions between the neutrons and the materials within the regions. Simulations show that for the case of a spherical tokamak with aspect ratio of 1.8 at 120 MW of neutron power, the tritium breeding ratio can have values ranging 0.85 and 1.25, and with that power level the system can bring 336 assemblies to 5% 233U enrichment in irradiation times between 24 and 65 months. The values depend on both the selection of materials and key geometry parameters, such as the thickness of container walls, as well as the volumes for both neutron multiplying and tritium breeding regions. FLiBe was found to give the best performance as neutron multiplier, since it contributes to the tritium self-sufficiency as well; from the tritium breeding perspective, lithium oxide was found to be the best choice.

Evaluation of local stress state due to grain-boundary sliding during creep within a crystal plasticity finite element multi-scale framework

Previous studies demonstrate that grain-boundary sliding could accelerate creep rate and give rise to large internal stresses that can lead to damage development, e.g. formation of wedge cracks. The present study provides more insight into the effects of grain-boundary sliding (GBS) on the deformation behaviour of realistic polycrystalline aggregates during creep, through the development of a computational framework which combines: i) the use of interface elements for sliding at grain boundaries, and ii) special triple point (in 2D) or triple line (in 3D) elements to prevent artificial dilation at these locations in the microstructure with iii) a physically-based crystal plasticity constitutive model for time-dependent inelastic deformation of the individual grains. Experimental data at various scales is used to calibrate the framework and compare with model predictions. We pay particular consideration to effects of grain boundary sliding during creep of Type 316 stainless steel, which is used extensively in structural components of the UK fleet of Advanced Gas Cooled Nuclear Reactors (AGRs). It is found that the anisotropic deformation of the grains and the mismatch in crystallographic orientation between neighbouring grains play a significant role in determining the extent of sliding on a given boundary. Their effect on the development of stress within the grains, particularly at triple grain junctions, and the increase in axial stress along transverse boundaries are quantified. The article demonstrates that the magnitude of the stress along the facets is highly-dependent on the crystallographic orientations of the neighbouring grains and the relative amount of sliding. Boundaries, transverse to the applied load tend to carry higher normal stresses of the order of 100-180 MPa, in cases where the neighbouring grains consist of plastically-harder crystallographic orientations. An approach to estimate the amount of sliding as a function of GB particle morphology and diffusive accommodation is developed and employed. The possible implications of these findings on the initiation of creep damage are commented on.

Assessment of the Radiotoxicity of Spent Nuclear Fuel from a Fleet of PWR Reactors

The paper presents the methodology for the estimation of the long-term actinides radiotoxicity and isotopic composition of spent nuclear fuel from a fleet of Pressurized Water Reactors (PWR). The methodology was developed using three independent numerical tools: the Spent Fuel Isotopic Composition database, the Nuclear Fuel Cycle Simulation System and the Monte Carlo Continuous Energy Burnup Code. The validation of spent fuel isotopic compositions obtained in the numerical modeling was performed using the available experimental data. A nuclear power embarking country benchmark was implemented for the verification and testing of the methodology. The obtained radiotoxicity reaches the reference levels at about 1.3 × 105 years, which is common for the PWR spent nuclear fuel. The presented methodology may be incorporated into a more versatile numerical tool for the modeling of hybrid energy systems.

Microstructure and microchemistry of laser welds of irradiated austenitic steels

This article investigates the integrity of laser welds on neutron irradiated, He-containing steels. Life extension of the current fleet of light water reactors could necessitate repair of cracks on irreplaceable internal components, but heat input of weld repairs exacerbates the problem by initiating He-induced cracking. Laser welding is a promising low-heat-input technology thought to limit the extent of He-induced cracking. In this study, we produce laser welds in a hot cell on AISI 304L stainless steel plates previously irradiated in the Experimental Breeder Reactor (EBR)-II. We select a systematic set of three specimens spanning fluences ~1–28 displacements per atom (dpa) at ~415–430 °C and He concentrations ~0.2–8 atomic parts per million (appm) amounting to ~0.2–2.8% swelling. He-induced cracking is observed only in specimens containing ≥3 appm He. Laser welding nearly eliminates all irradiation-induced cavities and reduces the dislocation loop number density, similar to conventional post-irradiation annealing. Microchemically, laser welding induces Cr-rich precipitation and suppresses grain boundary radiation-induced segregation. The mechanism of He-induced cracking is discussed in the context of these microchemical changes. The weld heat input is calculated and suggests that further refinement of laser welding parameters may improve the cracking resistance for higher dose and He conditions.

Automated Object Detection for Visual Inspection of Nuclear Reactor Cores

Remote visual inspection is a common approach to understanding the health of key components and substructures within nuclear power plants, particularly in difficult to access and high dosage areas. Interpretation of inspection footage is a manually intensive procedure and challenges arise in localizing and dimensioning defects directly from a video feed, which may be subject to uncertainty from a range of sources such as lens distortion, nonuniform lighting, and lack of depth from a monocular camera system. A common approach to addressing these issues is to develop a scaling factor based on identifying a reference object of known dimensions in the image and using this to size regions of interest. Manual, accurate identification of these reference objects is onerous, time consuming, and prone to variation across different human experts, therefore, robust identification of suitable reference objects in an automated, reliable, and repeatable manner is of significant value. In this paper we evaluate two approaches for the automated detection of reference objects in the inspection of graphite cores in the United Kingdom’s fleet of advanced gas-cooled reactors (AGRs). The first method is a multistep approach using tools from mathematical morphology. The approach uses a genetic algorithm to “grow” suitable structuring elements, refine the order of operations, and remove operations proposed by the human designer that have a negative impact on performance. The second approach uses semantic segmentation, a technique which is normally applied to scene labeling in computer vision applications, applied to produce a binary mask, separating the reference object from the background. We show that this second method performs significantly better than the mathematical morphology approach when applied to the identification of brick interface keyways in AGR inspection images. Though improved in terms of accuracy, it is recognized that a greater initial effort is required to train the approach, and as it utilizes black-box neural network approaches, the greater transparency offered by the mathematical morphology approach is lost. While explicability of techniques is often a highly desirable characteristic of automated analysis techniques applied to health assessment within nuclear power plants, the results of the reference object detection can be made explicit to the end user, ensuring that the human analyst is retained within the decision-making process thus mitigating the need for transparency.

Ageing and air leakage assessment of a nuclear reactor containment mock-up: VERCORS 2nd benchmark

Electricité de France (EDF) operates a large fleet of nuclear reactors and is responsible for demonstrating the safety of facilities, including concrete containment buildings (CCB), which are non-replaceable components. The leak-tightness of CCBs is assessed every 10 years during integrated leak-rate tests (IRLT). For double-wall containments, which have no metallic liners, the leak-tightness is strongly influenced by the degree of cracking of concrete and opening of the cracks, which mostly depends on (a) the prestress decrease due to the delayed strains of concrete and to a lesser extent due to relaxation of tendons steel, and (b) the saturation degree of the Powered by Editorial Manager® and ProduXion Manager® from Aries Systems Corporation concrete wall. Therefore, to optimize the maintenance programs on CCBs, it is important to predict the evolution of drying, creep and shrinkage strains of concrete to be able to correctly assess the pre-stress losses, and finally the air leak-tightness at a structural level during pressure tests or under accidental loadings.To improve our understanding and identify the best modelling practices on this issue, a large experimental program called VERCORS was launched in 2014. VERCORS is a 1/3 mock-up of a 1300 MWe nuclear reactor CCB. It has been widely instrumented, and its concrete thoroughly characterized. A specific attention has been paid to ensure it is consistent with real CBBs features in EDF's nuclear fleet. To complement its internal R&D efforts, EDF decided to associate external partners to this program. One of the means for this is the organization of benchmarks, where all teams are given data and information about the mock-up and are asked to quantitatively predict its behaviour. The present paper reports the organization and findings of the 2nd benchmark which was organized in 2018 and gathered several international teams around the same objective: improve the confidence in the modelling of structural behaviour as well as the leak-tightness of concrete in containment walls under pressure test loading.The benchmark has shown once again that predicting the mechanical and leakage behaviour of containment buildings is a difficult task. The benchmark also yielded interesting information about the possibility to use spatially reduced models to predict the mechanical behaviour and leakage and underlined the fact that more research must be done to better predict the localization of cracks and leakage.Some lessons have been learnt for the next benchmark: EDF will ask to clarify further the calibration methods, will give more data (including drying, creep and shrinkage at different temperatures and moisture measurements in the mock-up), and will help the participants using local leakage data by projecting the raw measurements on a regular grid, so that the local leakage models can be improved.

From “Inherently Safe” to “Proliferation Resistant”: New Perspectives on Reactor Designs

A recent American “mini-series” on Chernobyl, widely watched across the world, presented viewers with the concluding finding that this massive accident had occurred because the reactor design had inherent flaws; flaws that were known but not previously fixed because it was “cheaper” that way. The reactor design in question is the RBMK, and this paper will argue that this design was far from “cheap,” neither then nor now, and that its adoption as the second standard design for the Soviet Union’s nuclear power reactor fleet was based on much more than economic considerations. With the benefit of hindsight, it is easy to forget that reactor designs are always chosen for a multitude of reasons and never solely based on their technical or economic merits. Based on archival research, interviews, and industry publications, I show that approving and building RBMK reactors made good sense at the time, despite later claims to the contrary. Then I take the examples of a small modular reactor (SMR), the proposed NuScale Power Module, and a fast neutron reactor, TerraPower’s proposed Traveling Wave Reactor, to argue that we witness comparable negotiations today, as new designs for reactors (1) attempt to fit into existing safety and regulatory frameworks, (2) navigate security and nonproliferation concerns, and (3) embody visions of a specific sociotechnical order. I conclude that technical designs never occur in a socioeconomic, political, or cultural vacuum; instead, they are developed by people steeped in social norms, regulatory concerns, and economic expectations of a specific time and place. In the spirit of making this point relevant to practitioners, I will suggest ways of making these implicit frameworks visible, to actively and consciously start tweaking them, while staying aware of the implications that technical choices may have on our social expectations and vice versa.

Assessing the Applicability of the Structure-Based Turbulence Resolution Approach to Nuclear Safety-Related Issues

The accuracy of computational fluid dynamics (CFD) predictions plays a fundamental role in supporting the operation of the current nuclear reactor fleet, and even more importantly the licensing of advanced high-efficiency reactor concepts, where local temperature oscillations driven by thermal striping, cycling and stratification can limit the structural performance of vessels and components. The complexity of the geometrical configurations, coupled to the long operational transients, inhibits the adoption of large eddy simulation (LES) methods, mandating the acceptance of the more efficient Reynolds-averaged Navier-Stokes (RANS)-based models, even though they are unable to provide a complete physical description of the flow in regions dominated by complex unsteady coherent structures. A new strategy has been proposed and demonstrated at Massachusetts Institute of Technology (MIT) toward the enhancement of unsteady Reynolds-averaged Navier-Stokes (URANS) predictions, using local resolution of coherent turbulence, to provide higher fidelity modeling in support of safety-related issues. In this paper, a comprehensive assessment of the recently proposed Structure-based (STRUCT-ε) turbulence model is presented, starting from fundamental validation of the model capabilities and later focusing on a representative safety-relevant application, i.e., thermal mixing in a T-junction. Solutions of STRUCT-ε, the widely used Realizable k−ε model (RKE) and Large Eddy Simulation with Wall-Adapting Local Eddy-viscosity subgrid scale closure (LES-WALE) are compared against the experimental data. Both the velocity and temperature fields predicted by the STRUCT-ε model are in close agreement with the high-fidelity data from the experiments and reference LES solutions, across all validation cases. The approach demonstrates the potential to address the accuracy requirements for application to nuclear safety-related issues, by resolving the turbulent flow structures, while the computational efficiency provides the ability to perform consistent uncertainty quantification.

High-throughput ion irradiation of additively manufactured compositionally complex alloys

Several advanced nuclear reactor designs promise efficiency and safety improvements over the current reactor fleet but are limited by the current set of ASME code-qualified materials. Novel alloys including high-entropy alloys (HEAs), and more broadly compositionally complex alloys (CCAs), have shown promising irradiation-tolerance. However, the vast range of alloy compositions adds to an already time-consuming alloy development process. In this study, to accelerate the development of novel alloys for nuclear applications, a high-throughput (HTP) methodology has been employed. Additive manufacturing has been used to produce a compositional array of unary, binary, ternary, and quaternary alloys, including several CCAs, which span the Cr-Fe-Mn-Ni composition space. The compositional array was homogenized at 1000°C for 24 hours and each sample was irradiated using 4-MeV Ni2+ ions at room temperature to a peak damage of 50 dpa, as estimated using SRIM, at the University of Wisconsin Ion Beam Laboratory. A custom XY stage was built to accommodate the large compositional array and half of each sample was masked during irradiation enabling both the irradiated and unirradiated properties of each alloy to be characterized side-by-side. Each alloy was characterized using X-ray fluorescence (XRF), X-ray diffraction (XRD), scanning electron microscopy (SEM), energy dispersive spectroscopy (EDS), and nanoindentation. CALPHAD simulations spanning the entire Cr-Fe-Mn-Ni composition space at 1000°C were performed to compare predicted equilibrium phases with phases identified experimentally from the unirradiated regions of each alloy. Nanoindentation measurements indicate radiation-induced hardening ranging from ~1-1.5 GPa in each of Cr-Fe-Mn-Ni CCAs, which is relatively insensitive to modest changes in alloy composition and comparable to hardening observed in neutron irradiated Cr-Fe-Mn-Ni CCAs in the literature. Overall, a substantial time savings was realized by employing HTP synthesis, irradiation, and characterization in this study compared to conventional techniques, the implications of which are discussed.

Enrichment dynamics for advanced reactor HALEU support

Transitioning to High Assay Low Enriched Uranium-fueled reactors will alter the material requirements of the current nuclear fuel cycle, in terms of the mass of enriched uranium and Separative Work Unit capacity. This work simulates multiple fuel cycle scenarios using Cyclus to compare how the type of the advanced reactor deployed and the energy growth demand affect the material requirements of the transition to High Assay Low Enriched Uranium-fueled reactors. Fuel cycle scenarios considered include the current fleet of Light Water Reactors in the U.S. as well as a no-growth and a 1% growth transition to either the Ultra Safe Nuclear Corporation Micro Modular Reactor or the X-energy Xe-100 reactor from the current fleet of U.S. Light Water Reactors. This work explored parameters of interest including the number of advanced reactors deployed, the mass of enriched uranium sent to the reactors, and the Separative Work Unit capacity required to enrich natural uranium for the reactors. Deploying Micro Modular Reactors requires a higher average mass and Separative Work Unit capacity than deploying Xe-100 reactors, and a lower enriched uranium mass and a higher Separative Work Unity capacity than required to fuel Light Water Reactors before the transition. Fueling Xe-100 reactors requires less enriched uranium and Separative Work Unit capacity than fueling Light Water Reactors before the transition.