Future Reactors Research Articles

Mitigation of divertor edge localised mode power loading by impurity seeding

One of the major challenges for the design of future thermonuclear reactors is the problem of power exhaust—the removal of heat fluxes deposited by plasma particles onto the plasma-facing components (PFCs) of the reactor wall. In order for the reactor to work efficiently, the power loading of the PFCs has to stay within their material limits. A substantial part of these heat fluxes can be deposited transiently during the impact of edge localised modes (ELMs), which typically accompany the high confinement mode, a regime foreseen for tokamak ITER and next-step devices. One of the possible ways to mitigate the deposition of localised heat fluxes during ELMs is injection of impurities, which could similarly to inter-ELM detachment dissipate part of the energy carried by plasma particles, the so-called ELM buffering effect. In this contribution, we report on experimental observations in impurity seeded discharges in ASDEX Upgrade, where injection of argon is capable of reducing the ELM energy by up to 80 (60 without degradation of confinement). A simple model of ELM cooling is in some cases capable of providing quantitative prediction of this effect. The ELM peak energy fluence was reduced by a factor 8 without a degradation of the pedestal pressure. Should such mitigation be achieved in ITER, the resulting power loading would satisfy the material limits of divertor tungsten monoblocks (Eich et al 2017 Nucl. Mater. Energy 12 84–90) and as such avoid the risk of their melting. The most favourable results in terms of confinement and divertor heat flux mitigation were achieved by use of a mixture of argon and nitrogen, where the later impurity helped to improve the confinement. The ELM frequency was identified as a scaling factor for in discharges with impurity seeding, suggesting that high frequency ELMs are favourable for future devices.

Spectrometric performance of SiC radiation detectors at high temperature

In this work, we have investigated the performance of a 4H–SiC radiation sensor in the temperature range from 25 °C to 450 °C to explore its compatibility as detector of fast ion losses in plasma diagnostic of future nuclear fusion reactors. To simulate the escape of fusion-born alpha particles in D-T (deuterium-tritium) fusion plasmas, spectroscopic measurements were carried out in a vacuum chamber by irradiating the detector with a 3.5 MeV alpha beam from a Tandem accelerator. The detector was found to have an energy resolution ≤2% over the entire temperature range analyzed. Relevantly, the excellent spectrometric capabilities of the device have allowed us to calculate from experimental data, with unprecedented accuracy, the average energy required to create a single electron-hole pair in 4H–SiC as a function of temperature.

Future nuclear fission reactors – uncertainties, the effect of parameter choice and an application to small modular reactor concepts

Abstract. Today, steep declines in the generation costs of renewable energy systems, particularly solar photovoltaic and wind, combined with a recent spur in storage and flexible technologies mainly driven by batteries and renewable-energy-generated hydrogen are driving a paradigm shift in energy systems across the world, and renewable energy dominates investments in installed electricity generation systems (Ram et al., 2022). In contrast, in 2022 USD 4.8 billion (ca. 21 %) of the total spending in research and development expenditures by countries within the Organisation for Economic Co-operation and Development (OECD) was still spent on nuclear development alongside the current discussion in recent literature about the role of new nuclear developments in combating the recent energy crisis and climate change (Asuega et al., 2023; Ingersoll et.al., 2020; IEA, 2023). Since predicting future costs of technologies not yet developed is a complex exercise, we start this paper with a comparison of two different approaches from production theory and show that they have a theoretically equal structure (Steigerwald et al., 2023). This we then apply to estimate a range of potential production costs for 15 so-called small modular reactor (SMR) concepts for which sufficient data are available. These include light-water, high-temperature and fast reactors. In a third step, we apply the Monte Carlo method to benchmark the cost projections assumed by the manufacturers by varying production parameters in simulations of the net present value (NPV) and the levelized cost of electricity (LCOE). Besides this, we test whether the differences between the manufacturer estimates and ours differ between technology families of SMR concepts and apply a sensitivity analysis. This contributes to an intensifying debate in the literature on the economics and finance of SMR concepts (Steigerwald et al., 2023). In a final step we compared our results with current costs of renewable energy technologies and could conclude that none of the tested concepts is able to compete economically with existing renewable technologies, not even when taking their variability and necessary system integration costs into account. We conclude that any technology foresight has to take the case specifics into account as much as possible, including technological and institutional specifics; this also holds for SMR concepts.

Simulator Feature Framework: Requirements to Support Training, Research, and Education

There is an expansion of nuclear power plant control room simulator use beyond traditional operator training. Modern simulator designs require more robust capabilities to serve the diverse needs of multiple user groups including researchers and educators. A common framework for evaluating simulator features to support training, research, and education is critical to ensure future simulators enable research for immediate and future plant modernization and advanced reactor deployment needs. An initial framework comprised of 8 feature categories was developed by reviewing published simulator-based research and analyzing simulator features against research objectives and results. Future research will focus on evaluating the suitability of this framework in characterizing and differentiating simulators across training, research, and education use cases.

Improved and Innovative Accident-Tolerant Nuclear Fuel Materials Considered for Retrofitting Light Water Reactors—A Review

Since 2011, there has been an international effort to evaluate the behavior of newer fuel rod materials for the retrofitting of existing light water reactors (LWR). These materials include concepts for the cladding of the fuel and for the fuel itself. The materials can be broadly categorized into evolutionary or improved existing materials and revolutionary or innovative materials. The purpose of the newer materials or accident-tolerant fuels (ATF) is to make the LWRs more resistant to loss-of-coolant accidents and thus increase their operation safety. The benefits and detriments of the three main concepts for the cladding are discussed. These include (i) coatings for existing zirconium alloys; (ii) monolithic iron–chromium–aluminum alloys; and (iii) composites based on silicon carbide. The use of ATF materials may help extend the life of currently operating LWRs, while also being a link to material development for future commercial reactors.

Dynamic edge transport investigations at ASDEX Upgrade using gas puff modulation

Gas puff modulation experiments are performed at ASDEX Upgrade in L-mode, EDA H-mode and quasi-continuous exhaust discharges. Plasma density and temperatures are measured and their temporal development is analyzed simultaneously, revealing that both heat and particle transport are strongly influenced by the modulation. As a consequence, the particle transport coefficients are underdetermined. In the transport modelling, the pedestal cannot be treated as a single region, but the pedestal foot must be allowed to increase its transport with gas puff modulation independently. The analysis of the temporal behaviours of the heat and particle diffusivities shows that they are strongly correlated. Considering the heat diffusivity as a proxy for the particle diffusivity, allows interpretation of the density evolution: a pinch is not required for any of the discharges. An analysis with the gyrokinetic turbulence code GENE identifies dominant instabilities and reproduces several experimentally found trends. Despite all uncertainties concerning particle transport, one can expect a future reactor featuring a weak edge density gradient even with purely diffusive transport.

Wall cratering upon high velocity normal dust impact

Dust-wall high speed impacts, triggered by the termination of runaway electrons on plasma facing components, constitute a source of erosion. Normal high velocity mechanical impacts of tungsten dust on bulk tungsten plates are reproduced in a controlled manner by light gas gun shooting systems. Post-mortem surface analysis revealed that three erosion regimes are realized; plastic deformation, bonding and partial disintegration. The large impact statistics allowed the extraction of reliable empirical damage laws in the latter regime, which can be employed for erosion estimates in future reactors.

Development of new treatment of fuel isotope vector in the core degradation accident analysis of fast breading reactor

ABSTRACT In a design of future Sodium-cooled Fast Reactor (SFR), there is a demand for evaluation of event sequences and consequences of core disruptive accidents. Future SFRs include a unique core design with axially or horizontally heterogeneous core arrangement having complex fuel isotope distribution. A new model to flexibly represent fuel isotope distribution, called as the Pu-vector model, has been developed in this study for inclusion in the SIMMER-III and SIMMER-IV codes (simply called as SIMMER). The model calculates movement of individual fuel isotopes, assuming they always accompany the convecting fuel in the fluid-dynamics model. The accuracy of the Pu-vector model was confirmed by comparing with the standard Monte Carlo static neutronics calculation. The new model can improve some of the limitations in the current SIMMER code, in which the fuel isotopes are represented only by two groups, fertile and fissile fuels. Assignment of a number of fuel isotopes to the two groups requires a detailed examination of different combinations of fuel isotopes to determine an optimized combination. The Pu-vector model can eliminate this complicated procedure to be performed prior to a SIMMER analysis, and more importantly provides accurate spatial distribution of nuclide isotopes and thus will improve the applicability of SIMMER to the analyses of future large heterogeneous reactors.

Synthesis and thermophysical property determination of NaCl-PuCl3 salts

Currently, a knowledge gap exists in the available data and understanding of thermophysical properties relating to fresh fuel salts, especially those containing plutonium. These data are necessary for designing, constructing, and licensing future commercial molten-salt reactors. Thermophysical properties, such as melting temperature, salt stability, density, and heat capacity were ascertained using NaCl-PuCl3 (36 mol% PuCl3) and a more sodium rich composition containing 25 mol% PuCl3. The NaCl-PuCl3 salt mixture was synthesized for this study and contained 63.4 mol% NaCl, 36.3 mol% PuCl3 and was 99.7% pure. Upon heating, the NaCl–PuCl3 eutectic was stable at temperatures up to 800 °C. The onset of melting occurred at 451 ± 3 °C, and the enthalpy of fusion was determined to be 23.0 ± 1.4 kJ/mol. Heat capacity measurements in the liquid phase ranged from 107.7 to 91.3 J/mol.K, with an average value of 104.6 ± 11.4 J/mol.K between 500 and 650 °C. Three independent trials of the molten NaCl-PuCl3 salt found the density to be ρ(T) = 3.8589 – 9.5342·10-4 T(°C). In addition, ab initio molecular dynamic simulations to calculate density and heat capacity values are included.

Determination of the fracture toughness of a 9Cr ODS steel fuel cladding using the pin loading Tension test

A Pin Loading Tension (PLT) test was developed to evaluate the fracture behavior of thin-walled tubes. The fracture resistance curve (J-R curve) was determined from the load-CMOD curve using the single specimen technique based on the elastic unloading method following the procedure described in the ASTM E1820-22e1 standard [1]. However, the standard does not provide the necessary functions, geometric function f(a/W) and plastic factor ηpl, to apply the methodology to the PLT test. In this work, the geometric functions needed to evaluate the crack length from the unloading compliance, which allows to highlight effects not seen in the past and the elastic and plastic parts of the J-integral were determined using finite element analysis. Special care was taken to consider the effects of contact and friction between the test specimen and the loading device. The methodology was then applied to experimentally study the crack growth resistance of a 9Cr ODS (Oxide Dispersion Strengthened) steel tube, which is a candidate material for fuel claddings of future fast-neutron reactors.

Disruption prediction for future tokamaks using parameter-based transfer learning

Tokamaks are the most promising way for nuclear fusion reactors. Disruption in tokamaks is a violent event that terminates a confined plasma and causes unacceptable damage to the device. Machine learning models have been widely used to predict incoming disruptions. However, future reactors, with much higher stored energy, cannot provide enough unmitigated disruption data at high performance to train the predictor before damaging themselves. Here we apply a deep parameter-based transfer learning method in disruption prediction. We train a model on the J-TEXT tokamak and transfer it, with only 20 discharges, to EAST, which has a large difference in size, operation regime, and configuration with respect to J-TEXT. Results demonstrate that the transfer learning method reaches a similar performance to the model trained directly with EAST using about 1900 discharge. Our results suggest that the proposed method can tackle the challenge in predicting disruptions for future tokamaks like ITER with knowledge learned from existing tokamaks.

Design Analysis of a Helium Xenon-Printed Circuit Heat Exchanger for a Closed Brayton Cycle Microtransport Reactor

Microtransport with small size and a wide range of applications is very attractive for the utilization of future reactor modularization. A microreactor uses a closed Brayton cycle (CBC) to achieve high conversion efficiencies at low specific mass. The recuperator is one of the key components of CBC system which recovers heat exhausted from the turbine. A printed circuit heat exchanger (PCHE) is currently the preferred type of recuperator for the closed Brayton cycle due to its high heat transfer efficiency, high compactness, and high pressure and temperature resistance. This work intends to analyze different geometry configurations of a zigzag PCHE to increase its heat transfer efficiency while reducing its weight and size. The effect of geometric structure parameters such as channel diameter, zigzag pitch length, and zigzag angle on the Nusselt number and Fanning friction factor is investigated using a zigzag PCHE unit model. Based on the numerical simulation, the principle of least squares is employed to carry out a nonlinear fitting of the flow and heat transfer criterion correlation equations. Besides, the maximum value of the Nusselt number and minimum value of the Fanning friction factor are optimized as two conflicting objective functions using the nondominated sorting genetic algorithm-II (NSGA-II), with which a set of optimal solutions is obtained. Meanwhile, the shortest normalized distance is used to determine the compromised solution on the Pareto optimal points, and the independent variable’s sensitivity analysis is performed. Finally, a multiobjective optimization analysis is conducted for the PCHE to achieve lightweight and the high heat transfer efficiency design requirements of SIMONS.

Molecular dynamics simulation on the dissolution and diffusion characteristics of FeCrAl alloy in liquid LBE

Lead-bismuth eutectic (LBE) corrosion-resistant materials are crucial for the development of future fourth-generation nuclear reactors, and the dissolution and diffusion of material components are important processes in LBE corrosion-resistant materials. FeCrAl alloy is an important structural material for fourth-generation reactors, and it is essential to study its dissolution and diffusion processes in liquid LBE. Molecular dynamics simulations (MD) were used to investigate the mutual diffusion process between the FeCrAl alloy and liquid LBE. The dissolution and diffusion abilities of Fe, Cr, and Al atoms in liquid LBE were evaluated using mean-square displacement (MSD) and diffusion activation energy (DAE). Fe atoms exhibited the highest degree of diffusion, while Al atoms showed the lowest degree of diffusion, resulting in a distinct difference in the elemental distribution at the solid–liquid interface (SLI). There was a significant difference in the depth of penetration of Pb and Bi atoms into the FeCrAl alloy matrix, with Bi atoms penetrating deeper. Furthermore, there is a thermal vibration process during the invasion of Pb and Bi atoms into the FeCrAl alloy matrix, which promotes the invasion process of Pb and Bi atoms. In addition, an increase in temperature exacerbates the diffusion of Fe, Cr, and Al atoms in liquid LBE and the process of Pb and Bi atom invasion into the FeCrAl alloy matrix. Our research may contribute to the design of future nuclear-grade FeCrAl alloy and the development of next-generation reactors.

Development of pebble-based extruded carbon rods for extreme plasma heat flux environments

This work presents first experiments toward the development of continuously renewable (extrudable) pebble-based carbon rods for use as plasma-facing components in extreme steady-state plasma flux environments. The primary envisioned application of this work is a first wall that can survive long-term in future magnetic fusion power reactors while also improving recovery of the reactor fuel (tritium and deuterium atoms). Bench tests applying extreme steady-state front-surface heat loads of up to 50 MW/m2 are presented. Continuous pebble rod front-surface recession and intact pebble recovery are successfully demonstrated, at a rate of order 0.2 cm/s. Numerical simulations of the pebble rod front-surface recession are able to match observations reasonably well, indicating that the recession mechanism can be understood as occurring due to pebble thermal expansion and resulting shock and cracking of the inter-pebble binder. Tests of the pebble rod extrusion demonstrate that friction between the rods and the stainless steel extrusion channel is tolerably low (<50 N for the expected channel length) over a wide range of temperatures. Front-surface outgassing rates below 1000 Torr L/s/m2 are achieved, believed to be sufficiently low for use in magnetic fusion reactors. Initial parametric scans over pebble rod size and binder fraction to vary front-surface recession rates are presented.

Overview of multiphysics R&D activities at the CEA/IRESNE institute

The major mission of the IRESNE (Research Institute on Nuclear Systems for low carbon Energy production) institute of the DES (Energy Division) at CEA (French Alternative Energies and Atomic Energy Commission) is to study current and future nuclear reactors integrated into low-carbon energy systems. To fulfill this mission, the IRESNE’s main activities rely on the development and use of advanced physical models, multiphysics tools including efficient coupling schemes, and dedicated experimentation. Approaches to nuclear systems using multiphysics simulations are now required, in order to better understand the complex intrinsic features of the coupled phenomena.Hence, IRESNE addresses eight different multiphysics research fields: fuel fabrication, fuel behavior under irradiation, fuel assembly integrity under fluid–structure interaction, contamination of fluid circuits, sodium technology, behavior of reactors, corium and fission products, and radionuclide transfer to environment. A general overview of IRESNE’s current multiphysics activities in these eight fields is provided in this paper. It focuses on experimentation, modeling, simulation and tool qualification, corresponding to the Verification Validation and Uncertainty Quantification process. Moreover, IRESNE’s key challenges to be met in the next 5 to 10 years, to improve and validate multiphysics Modeling and Simulation, are also discussed in this contribution. The ultimate goal is to get closer to the development of a digital reactor, core and fuel, still progressing on applied mathematics, high-performance computing, and understanding and modeling of physical mechanisms.

Particle pump-out induced by trapped electron mode turbulence in electron cyclotron heated plasmas on XuanLong-50 spherical torus

Particle pump-out effects induced by low-frequency (<200 kHz) density fluctuations were observed in solely electron cyclotron wave (ECW)-heated plasmas on the spherical torus XuanLong-50 (EXL-50) without a central solenoid. The intensity of the relative density fluctuations increases with increasing ECW power and decays when the ECW is turned off while sustaining the plasma current. The electron densities are maintained relatively high and steady when the density fluctuations are completely absent, indicating that the outward transport of electrons is dominated by the particle pump-out effect of the ECW. The density fluctuations are modulated by a supersonic molecular beam injection pulse and the modulation amplitude decreases with increasing electron density at the same ECW injection power and decreasing ECW power at the same electron density, respectively. Analysis revealed that a critical value of electron temperature gradient (ETG) triggers the density fluctuations, and the intensity of the relative density fluctuations is positively correlated with the ETG and approximately inversely proportional to the effective collision frequency. With plasma parameters similar to those of EXL-50 experiments, the HD7 code simulations demonstrate that trapped electron mode (TEM) turbulence can be excited by ETG higher than the critical value observed in the experiment. In addition, the dependence of the mode growth rate (supposed to be proportional to the saturation level of fluctuations in quasi-linear theory) and the measured intensity of the density fluctuations is comparable. The simulated outward particle flux integrated over the poloidal wave number spectrum is significant and proportional to ETG. These observations demonstrate that the density fluctuation is TEM turbulence, which is driven by ETG and induces particle pump-out when the electron density/effective electron collision frequency is low. The potential relevance of this work with the controls of plasma profiles, impurities, helium ash, and heat transport in future reactors of similar low effective collision frequency is also discussed.

Structure, stability and defect energetics of interfaces formed between conventional and transformed phases in Cu–Nb layered nanocomposite

Layered nanocomposite material having fcc-bcc interface with Kurdjumov-Sachs interface orientation relation has shown great potential as radiation resistant structural material for future fusion energy reactors. The superior radiation resistant properties of this material are attributed to it’s special fcc-bcc interface structure. In this study we have reported a stable interface between conventional bcc phase of Nb and transformed bcc phase of Cu. This bcc-bcc interface is found to be stable from both strain-energy and dynamical stability analysis. We have also shown that the bcc-bcc interface has different defect energetics behaviour compared to previously reported fcc-bcc interface which has a negative impact on the self annihilation property of the material against radiation induced defects. These aspects should be carefully considered in the future design of robust layered material for extreme radiation environment.

Fractographic studies and tensile strength of MPG-6 graphite based on uncalcined coke, neutron-irradiated at a high temperature

This paper presents the results of studies of the mechanical properties and fracture characteristics of MPG-6 graphite in the unirradiated state and neutron-irradiated at high temperature. MPG-6 is considered as a material for future high-temperature gas-cooled reactors (HTGR). In particular, MPG-6 is considered as one of the candidate materials for prismatic fuel blocks of the Russian experimental hydrogen-producing nuclear power plant with the VTGR high-temperature reactor currently under design.It is shown that MPG-6 graphite, based on uncalcined petroleum coke, which causes a small difference in the size of crystallites in areas with different preferred orientations of crystallites, has higher radiation stability of mechanical properties than GR-280 graphite based on calcined coke.

Superior strength-ductility synergy of high potassium-doped tungsten rods with large swaging deformation

The high ductile-to-brittle transition temperature (DBTT) is a key factor limiting the use of tungsten (W) - based materials in future nuclear fusion reactors. To overcome this challenge, and to obtain W alloys with improved mechanical properties without sacrificing substantial thermal conductivity, a large-sized W rod with a total weight of 42 kg and with 96 ppm doped potassium (K) was fabricated through a powder metallurgy mass production route. Then high-temperature swaging processes were conducted to produce a rod with large plastic deformation of 94.7% (log. Strain of 2.95). The swaged rod showed good low-temperature ductility and high strength simultaneously. At 50 °C, the AKS-W rod fractured in brittle cleavage mode. The fracture changed to a mixture of brittle and ductile at 100 °C, indicating a DBTT of 50 °C - 100 °C. This value was lower than many other reported in pure W and K-doped alloys. The swaged W rod exhibited a high strength of larger than 1.0 GPa up to 100 °C, with good conductivity at low temperatures. The maximum total elongation of 30.24% was obtained at 200 °C. The superior strength-ductility synergy of AKS-W rod was observed to be attributed to its unique microstructure and K bubbles morphology, which evolved during severe hot plastic deformation.

Synthesis and characterization of 9Cr ODS F/M steel with optimized Y2O3 and Ti content and its comparison with P91 steel

Oxide dispersion strengthened (ODS) steels are considered as potential candidates for future Gen-IV reactors core structural applications. Ti is added to this alloy to refine the dispersoid and facilitates formation of Y-Ti-O complex oxides. Optimisation of Y2O3 and Ti content in these ODS steels governs size and stoichiometry of the dispersoids. Equal amount of both Ti and Y2O3 content resulted in formation of finer and stable Y-Ti-O dispersoids. In this study two varieties of ODS alloys are synthesized: one with 0.35 wt% Ti and Y2O3 while another with 0.2 wt% Ti and Y2O3 by mechanical milling followed by SPS. The microstructure and microtexture of these two varieties of ODS steels were compared with modified 9Cr1Mo (P91) steel. The creep-rupture studies reveal that the ODS steels have better creep-rupture life as compared to P91 steel, and the steel with 0.35 wt% Ti and Y2O3, possess superior mechanical properties.