Reactor Components Research Articles

Prospects for melt refining of radioactive waste from fusion reactors

Fusion has the prospect of producing only low-level radioactive waste, but the volume of this waste is potentially very large. A possible means of reducing the volume of this waste is to recycle the activated materials, e.g. within the fusion industry. In the process of steelmaking, certain impurities are removed via accumulation in slag or fumes produced during the process, and this has been successfully employed as a means of clearing surface-contaminated LWR steels (e.g. steam generators). In this work, we simulate the process applied to neutron-activated fusion reactor components including the blanket, divertors, and shielding.The process is based on a well known partitioning between melt, slag, dust, and gas that is based on the oxidation potential of the elements and their volatility. We model this partitioning of activated materials and use FISPACT-II to simulate their activation during operation; the processes are repeated cyclically to simulate the long-term accumulation of activation products in the steel and waste products. The principal findings are that most of the induced activity remains in the metal rather than accumulating in the slag, and that while the slag remains a low-level waste under the assumptions employed, its volume actually exceeds that of the recycled material, and it therefore did not constitute a successful waste volume reduction technique.

Predictive atmospheric dispersion and deposition characteristics of activated tungsten dust

Tungsten is considered to be the most promising material for plasma-facing components of fusion reactors in addition to other industrial and military uses. Activated tungsten dust produced by plasma-wall interaction and released under accidental conditions would be the emerging environmental pollutant with both chemical and radiation toxicity. For the first time, atmospheric dispersion and deposition characteristics were investigated for tungsten dust with various particle size spectrum similar as generated in fusion reactors. Euler-Lagrange approach was adopted to describe parcels motion and deposition behaviors. Proper numbers of parcels were determined for the atmospheric dispersion and various parcel diameters scale. Near-surface concentration distribution and dry deposition velocity were predicted for tungsten dust with both single size dust and realistic particle size spectrum dust. Influences of chemical forms and release factors on atmospheric dispersion and deposition of tungsten dust were preliminary identified at last. It is indicated that 150 million parcels are needed for 10 μm dust under 5 km scale. Dry deposition velocity of tungsten dust is up to 11.3 cm s−1 and increase along with particle diameter because of its high density and gravitational deposition effect. The critical diameter judging whether gravitational deposition could appear for tungsten dust is 1 µm. Deposited particle size spectrum would be completely different along the downwind distance which would result in differences in subsequent behaviors in other environmental compartments.

Unstable combustion behavior of syngas/air mixture with different components in a narrow gap disk reactor

In this paper, the combustion behavior of the syngas/air mixture in a narrow gap disk reactor were explored. Besides, the influence of the hydrogen volume fraction and the mixture equivalence ratio in syngas on the property of propagating flames were investigated in the narrow gap disk reactor under the gap thicknesses of 8 mm, 6 mm, 4 mm, 3 mm, and 2 mm, respectively. The focus was placed on the behavior of the flame propagation under the gap thickness of 2 mm. The research results demonstrated that the increase of the hydrogen volume fraction from 5 % to 30 % exerted a significant impact on the flame propagation velocity (FPV) and the pressure in the tube. Moreover, the equivalence ratio of this mixture could markedly affect both the propagation characteristics and morphology of flames. Strong flame instabilities accompanied by pressure oscillations were observed in the reactor at a 2 mm gap thickness during the later stages of flame propagation. Notably, there were significant differences in unstable combustion behavior between different syngas components. Moreover, the combined action of heat losses and compression waves leads to significant flame instability. The insights gained from this study have practical implications in various production applications. For example, optimizing hydrogen volume fraction and mixture equivalence ratios can enhance combustion efficiency and reduce pressure fluctuations in industrial reactors, leading to improved energy utilization and reduced environmental emissions.

Study of the impact of hydrides on the fracture toughness of Zr-2.5Nb alloy

Hydrogen absorption and hydride formation in zirconium (Zr) alloys is a well-known issue and has been extensively studied in the past. Hydrogen has an impact not only on the strength but also on the fracture toughness of Zr alloys. As various Zr alloys are used for manufacturing the core components of nuclear reactors, such as nuclear fuel claddings, fuel channels, fuel assemblies, fixtures, and etc., their structural integrity has a great impact on the safety and reliability of the whole nuclear power plant. Furthermore, the hydrogen absorption ageing mechanism retains its significance not just during reactor operation but also throughout decommissioning and storage of spent nuclear fuel, especially when sufficient levels of absorbed hydrogen concentrations are reached, potentially leading to a delayed hydride cracking mechanism. Assessing the impact of hydrogen absorption on the strength of zirconium alloys typically involves time-consuming, costly, and sample-intensive experimental studies. Consequently, there is a demand for alternative research approaches. This article introduces the methodology to evaluate the fracture toughness of the Zr-2.5Nb alloy with varying hydrogen concentrations.

Welding with Pressure for Making Nuclear Reactor Components

Welding with Pressure for Making Nuclear Reactor Components

Study of irradiation-induced vacancy-type defects in tungsten using positron annihilation spectroscopy

Tungsten has been considered as a potential candidate material for the application of plasma-facing components in magnetic confinement fusion reactors. Fusion neutrons introduce severe displacement damage into the first-wall materials along with a high amount of helium which is detrimental to the desirable material properties. Ion beam implantation combined with positron annihilation spectroscopy is a well-adopted method to investigate the irradiation damage in materials without any induced radioactivity. In this study, two separate tungsten samples were individually irradiated with 9.1 MeV Au5+ ions to a dose of 2.5 × 1014 ions/cm2 and 130 keV He ions to a dose of 1 × 1016 ions/cm2 with the intention to investigate the evolution of vacancy-type defects in the absence and presence of helium. The irradiated samples were subjected to isochronal annealing and characterized using positron annihilation Doppler broadening spectroscopy. The evolution of vacancy-type defects during different stages of annealing was investigated using defect-sensitive positron S-parameter. An increase in S-parameter due to the presence of vacancy-type defects was observed upon irradiation in both samples. The S-parameter variations with respect to the temperature showed different stages of vacancy annealing corresponding to stage III from 473 K to 673 K where the migration and clustering of mono-vacancies occurs, stage IV from 673 K to 973 K due to the growth of nano-voids and stage V from 973 K to 1273 K due to the dissociation of nano-voids. Complete recovery of irradiation-induced defects occurred in both samples at 1273 K. Both the samples showed similar vacancy annealing stages and no signature on the formation and growth of He-vacancy clusters or He bubbles was observed in the present study.

Numerical and experimental development of cupronickel filler brazed joints for divertor and first wall components in DEMO fusion reactor

The brazeability of a cupronickel commercial alloy (Cu10Ni) was evaluated for its use as a filler alloy for high-temperature joining of tungsten to the reduced activation ferritic/martensitic steel EUROFER 97 (W-E) and between tungsten base materials (W-W) for its application at the first wall and divertor of future fusion reactors. In addition, given the importance of the residual stresses in these heterogenous joints, a study of the brazing conditions and the impact of the selected filler has been conducted using numerical software to understand its impact on the quality of the joint.Two thermal cycles were evaluated (1165 °C and 1190 °C) and selected based on the thermal characterization of the filler alloy. The microstructural examination revealed that, in W-E joints, nickel acts as an activator element, reacting and forming interfacial layers at the EUROFER 97 - Cu10Ni interface. In the case of the W-W joints, a lower level of diffusion phenomenon and metallurgical interaction between Cu10Ni and base materials were observed. The hardness profile indicated that the hardening process of EUROFER 97 was associated with the formation of untempered martensite. On the other hand, tungsten kept the received hardness. The mechanical characterization by shear test reported similar values between both types of joints carried out at 1190 °C but different when the temperature was increased (1165 °C), associated with the brittle character of tungsten and its lower metallurgical interaction.The numerical analysis of the brazing process carried out with ANSYS software shows that residual stresses are accumulated mainly at the interfaces. The information provided by the simulation shows, for a 50 µm filler thickness, the importance of mitigating the residual stress by selecting a filler with an intermediate Coefficient of Thermal Expansion (CTE) that alleviates mechanical stresses relative to the base materials.

Microstructure and Magnetism of Heavily Helium-Ion Irradiated Epitaxial Iron Films

This study reports on the microstructure and magnetism of pure iron irradiated with high doses of helium ions. Iron alloys are important structural materials used as components in fusion reactors, and a comprehensive database of their various properties has been developed. But little has been investigated on magnetic properties, in particular, the effects of high doses and helium cavities are lacking. Single-crystal iron films, with a thickness of 200 nm, were prepared using the ultra-high vacuum evaporation method. These films were then irradiated with 30 keV He+ ions at room temperature up to a dose of 18 dpa. X-ray diffraction measurements and cross-sectional transmission electron microscope observations revealed significant microstructural changes, including a large lattice expansion perpendicular to the film plane and the formation of high-density cavities after irradiation. However, the saturation magnetization and the shape of the magnetization curve showed almost no change, indicating the robustness of the magnetic properties of iron.

Experimental investigation and constitutive material modelling of low cycle fatigue of EUROFER97 for fusion applications

EUROFER97 is one of the candidate materials for structural components of fusion nuclear reactors (i.e. ITER and DEMO). Due to the plasma instabilities and cyclic operating conditions, these structural components are expected to face asymmetric thermomechanical fatigue loading. In this work, existing models based on the Chaboche viscoplasticity model have been modified in order to represent the fatigue behaviour of EUROFER97 under different strain ratios and ranges. The modified model has been parameterized based on experimental data obtained from fatigue tests of EUROFER97, in which the strain range varied from 0.6 % to 1.5 % and the strain ratio was either -1 or 0 at room temperature or 350 °C. The incorporation of the strain memory effect into the isotropic hardening due to the observed non-Masing behaviour of EUROFER97 enabled the use of unified parameters for every temperature that could cover all the different testing conditions. In order to validate the current model, the simulated results are compared to strain controlled low cycle fatigue test data performed at room temperature and 350 °C, which were not used for the parameterization. A good match between the simulation and experimental results proves the predictive capability of the proposed modified model.

Spatial Computer Model of the UCl<sub>3</sub>–NaCl–MgCl<sub>2</sub>–PuCl<sub>3</sub> Isobaric Phase Diagram

A four-dimensional (4D, in concentration–temperature coordinates) computer model of the isobaric phase diagram of uranium, sodium, magnesium, and plutonium chlorides, as well as four three-dimensional (3D) computer models of the phase diagrams of the ternary systems forming it, has been constructed. The technology of assembling a 4D model of 46 hypersurfaces and 17 phase regions was used in the design. The obtained 4D model of the UCl3–NaCl–MgCl2–PuCl3 phase diagram makes it possible to visualize a four-dimensional object as a whole (with all its hypersurfaces and phase regions) by any arbitrarily given 2D and 3D sections, as well as it is able to reproduce published (experimental or thermodynamically calculated) 2D sections. The scope of application of the results of the work is the development of materials for fuel components of fourth-generation molten salt reactors and pyrochemical recycling of spent fuel rods. For the first time, a comprehensive, complete description of phase diagrams composed of uranium, plutonium, sodium, and magnesium chlorides has been obtained.

Recrystallization kinetics of larger yttria particle dispersion tungsten alloy warm rolled to 50% thickness reduction

Tungsten-based materials are considered one of the candidate materials for the key components (such as the first wall and diverter) of future thermonuclear fusion reactors because of their excellent properties. At high operation temperatures, restoration processes such as recovery, recrystallization and grain growth will unavoidably embrittle ductile tungsten achieved by rolling. Dispersed second-phase particles will affect the recrystallization behavior. The thermal stability of a yttria dispersion-strengthened tungsten plate warm-rolled to 50% thickness reduction is investigated. Isochronous annealing is used to predict its recrystallization temperature range and isothermal annealing is used to study its recrystallization kinetics. The tungsten alloy in the present study has a faster recrystallization kinetics and a smaller activation energy compared to that of W-2 vol% Y2O3 alloy with small Y2O3 particles, indicating that the larger Y2O3 particles have little effect on impeding grain boundaries during recrystallization.

Sampling, characterization, method validation, and lessons learned in analysis of highly activated stainless steel from reactor decommissioning

Characterisation of activated reactor components is a multifaceted effort, which include both experimental work and activation calculations which together form a comprehensive characterisation approach. However, both the experimental work and activation calculations have their constraints and uncertainties. This work presents the characterisation of a highly activated stainless steel pipe including sampling planning, sampling, sub-sampling, elemental and radionuclide analyses, and activation calculations. 55Fe, 63Ni and 60Co analyses were carried out in a trilateral intercomparison exercise and the results were compared with activation calculations. The results showed excellent alignment between experimental and activation calculation results.

Simulation of noble metal particle growth and removal in the molten salt fast reactor

In the Molten Salt Fast Reactor (MSFR), fuel is dissolved in a liquid salt, causing the formation and decay of fission products, such as noble metals, to be inside the salt. Noble metals have very low solubility and do not dissolve in the salt. They are known to agglomorate into particles which, in turn, diffuse, drift or sediment to interfaces such as walls, free surfaces and bubbles, where they deposit. Leveraging this in the MSFR, online or offline helium bubbling is foreseen to act as a means to extract noble metal particles from the salt. This can potentially limit strong deposition of particles onto walls (i.e., plating), preventing dangerous decay heat hotspots in the heat exchangers or other primary circuit components. In this paper, we investigate the efficiency of particle removal by helium bubbles in the MSFR using Computational Fluid Dynamics (CFD). Theoretical estimates of equilibrium particle size distributions are used as initial condition, based on previous work. CFD calculations are performed in a simplified MSFR geometry. Bubble expansion, turbulent bubble coalescence and turbulent break-up are accounted for using the poly-disperse Log-normal Method of Moments (LogMoM) model, embedded in the two-fluid framework in OpenFOAM. The CFD analyses presented in this paper, in conjunction with the developed theory of noble metal particle formation, growth and capture, will help in the understanding and optimization of the fuel cycle in the MSFR or any other molten salt reactor design. The precise knowledge of how noble metal particle populations in the salt can be controlled, and how plating of noble metal particles onto primary circuit reactor components can be reduced, contributes to safe molten salt reactor operation.

Microstructure and electrochemical behaviour of laser clad stainless steel 410 substrate with stainless steel 420 particles

Abstract Stainless steel (SS) 410 is widely used in many engineering applications such as turbines, nuclear reactor components and seamless pipes due to its superior mechanical properties i.e., high mechanical strength, creep resistance and ductility. Corrosion resistance is an important property which is required for these components to reduce the material degradation from the surface. In the present study, an investigation has been made to enhance the electrochemical behavior of SS410 by depositing SS420 particles using the laser cladding technique. The hardness measured through nano-indentation tests and microstructure of the resulting surface were analysed. Also, electrochemical studies were performed on laser clad specimens with different durations such as 0 h, 8 h, 14 h and 36 h. The improvements in corrosion resistance were analysed using polarization curves and electrochemical impedance spectroscopy (EIS). Further, the morphology and roughness of the corroded surface were analysed. The results showed that a closely packed acicular structure was formed through cladding, which improved the nanohardness compared to the base material. In addition, the clad specimen with 14 h test duration exhibited excellent corrosion resistance amongst the other specimens. The morphological analysis of the corroded specimen revealed the formation of oxides which is indentified to be a major reason for improved corrosion resistance.

Hydrogen Diffusion in Nickel Superalloys: Electrochemical Permeation Study and Computational AI Predictive Modeling.

Ni-based superalloys are materials utilized in high-performance services that demand excellent corrosion resistance and mechanical properties. Its usages can include fuel storage, gas turbines, petrochemistry, and nuclear reactor components, among others. On the other hand, hydrogen (H), in contact with metallic materials, can cause a phenomenon known as hydrogen embrittlement (HE), and its study related to the superalloys is fundamental. This is related to the analysis of the solubility, diffusivity, and permeability of H and its interaction with the bulk, second-phase particles, grain boundaries, precipitates, and dislocation networks. The aim of this work was mainly to study the effect of chromium (Cr) content on H diffusivity in Ni-based superalloys; additionally, the development of predictive models using artificial intelligence. For this purpose, the permeability test was employed based on the double cell experiment proposed by Devanathan-Stachurski, obtaining the effective diffusion coefficient (Deff), steady-state flux (Jss), and the trap density (NT) for the commercial and experimentally designed and manufactured Ni-based superalloys. The material was characterized with energy-dispersed X-ray spectroscopy (EDS), atomic absorption, CHNS/O chemical analysis, X-ray diffraction (XRD), brightfield optical microscopy (OM), and scanning electron microscopy (SEM). On the other hand, predictive models were developed employing artificial neural networks (ANNs) using experimental results as a database. Furthermore, the relative importance of the main parameters related to the H diffusion was calculated. The Deff, Jss, and NT achieved showed relatively higher values considering those reported for Ni alloys and were found in the following orders of magnitude: [1 × 10-8, 1 × 10-11 m2/s], [1 × 10-5, 9 × 10-7 mol/cm2s], and [7 × 1025 traps/m3], respectively. Regarding the predictive models, linear correlation coefficients of 0.96 and 0.80 were reached, corresponding to the Deff and Jss. Due to the results obtained, it was suitable to dismiss the effect of Cr in solid solution on the H diffusion. Finally, the predictive models developed can be considered for the estimation of Deff and Jss as functions of the characterized features.

Investigating helium ion irradiation resistance in additively manufactured austenitic stainless steels

Additive manufacturing (AM) enables the rapid fabrication of complex shapes using engineering materials such as austenitic stainless steels and can imbue them with high irradiation resistance for use in reactor components. This is attributed to their refined grain structure and high grain boundary area. In this study, austenitic stainless steels (type 304) fabricated via direct energy deposition (DED) and powder bed fusion (PBF) techniques were irradiated with 5 MeV He ions to an approximate dose of 0.6 dpa at 300°C; subsequently, they were characterized through electron microscopy and the micro tensile testing. The results revealed that austenitic stainless steels manufactured using AM methods exhibited outstanding mechanical performance. The high performance of austenitic stainless steels fabricated through the DED technique can be attributed to their high tensile strength and excellent ductility elongation. This excellent performance is believed to be caused by the low stacking fault energy and the corresponding martensite formation during deformation. In particular, it was found that better mechanical properties were maintained even after helium irradiation, which is an important result obtained from the micro-tensile test. Even a small variation in the chemical composition and sub-microstructure of AM materials could result in improved irradiation tolerances.

Effects of grain boundaries on the evolution of radiation-induced bubbles in polycrystalline tungsten: A phase-field simulation

Tungsten (W) is widely recognized as a promising material for plasma-facing components in nuclear fusion reactors, and the evolution of bubbles in polycrystalline tungsten significantly impacts its properties. This study developed a phase-field model considering elastic inhomogeneity to investigate the evolution of radiation-induced bubbles. The Gibbs free energy of helium bubbles was derived using the Van der Waals equation. An explicit nucleation algorithm accounting for the position of grain boundaries (GBs) was coupled to the model. Furthermore, the model also considered the inequality between the internal pressure of the bubble and the surface tension. The growth kinetics of a single bubble with elastic heterogeneity were examined, and based on these findings, the evolution of bubbles under irradiation was simulated, yielding results consistent with experimental observations. Furthermore, the influence of GBs on radiation-induced bubbles was investigated using the developed model. The simulation results demonstrated that both GBs, as preferential nucleation regions and as sinks for vacancy absorption, had an impact on bubble nucleation and growth. The superior radiation resistance observed in nanocrystalline tungsten was dominantly attributed to the absorption of vacancies by GBs.

Hydrogen isotope permeation through interfaces and permeability of tungsten layers

The estimation of the hydrogen isotope flux through a fusion reactor wall component is important for the material selection and in order to guarantee a safe and economical reactor operation. Since the permeation flux through a component cannot be measured directly, due to the large size of such a component, the deuterium permeability of various fusion materials were investigated in the last years. In order to investigate, if an interface, meaning a change of materials in a component, has an influence on the permeation flux, combined material systems in laboratory scale are studied. The combined materials systems are produced by applying a thin W or Cu layer on a polished bulk steel or CuCrZr substrates by magnetron sputter deposition. In a previous study (Houben et al., 2022), the combined material system of Cu coated steel was investigated. The conclusion was that the interface has a minor influence on the permeation flux compared to the large influence of the layer microstructure. In order to investigate, if the minor influence of the interface is valid in general, the systems W coated steel and CuCrZr are studied in this publication. By heating the substrate during W sputter deposition, a crack-free W layer is produced and crack propagation in the W layer at elevated temperature is prevented. The W layer permeabilities are obtained for both W coated substrates and are similar, but compared to the W bulk permeability from literature the W layer permeabilities are several orders of magnitude larger.The main conclusion from these studies is that in all investigated combined material systems the influence of the interface on the permeation flux is minor compared to the large influence of the microstructure. Therefore, for a reliable estimation of the permeation flux through a fusion reactor component it is crucial to characterize the applied materials. Especial for coatings the measurement of the layer permeability is important, since the layer permeability of a material can be very different compared to the permeability of a bulk material.

Irradiation of ultrasonic sensors and adhesive couplants for application in light water reactor primary loop piping and components

The Electric Power Research Institute (EPRI) Nuclear Sector and US Department of Energy Light Water Reactor Sustainability Program are committed to engaging in research and development endeavors to address materials aging issues specific to long term operation of light water power reactors. To this effect, EPRI launched an industry initiative to develop nondestructive evaluation systems for online monitoring of existing cracks in light water reactor primary coolant loop piping and components. One of the goals of this initiative is to develop a sensor system (or systems) that can determine nondestructively if cracks are growing or arrested and, in the case of the former, to characterize their growth rates. A missing component of this initiative is an experimental assessment of how sensors and adhesive couplants will perform in service when exposed to chronic energetic neutron radiation, particularly at the primary coolant loop hot and cold leg dissimilar metal welds, which join the primary loop piping to the reactor pressure vessel and reside in the vicinity of the reactor core. The objective of this experimental study was to determine how ultrasonic transducers and adhesive couplants perform when exposed to irradiation in a test reactor to simulate and accelerate in-service exposure. To achieve this objective, the signal stability of piezoelectric transducers and performance of adhesive couplants as a function of accumulated fast neutron fluence were characterized by collecting ultrasonic data in-situ during irradiation. Of particular interest were the ultrasonic signal quality and time decay of the amplitude of acoustic reflections as a function of fast neutron fluence. The results of the study showed that, of the 8 transducer/substrate sample assemblies tested, only 3 generated usable ultrasonic signals through the conclusion of the irradiation campaign. It was found that high temperature epoxy tends to ultrasonically couple the sensors to the substrates better than three types of refractory ceramic cements studied, as is supported by post irradiation examination. The results obtained through this experimental study will be utilized in the achievement of the overall goal of development of a sensor system to perform online monitoring of primary loop components.This manuscript has been authored in part by UT-Battelle, LLC, under contract DE AC05-00OR22725 with the US Department of Energy (DOE). The publisher acknowledges the US government license to provide public access under the DOE Public Access Plan (https://energy.gov/downloads/doe-public-access-plan [energy.gov]).

Physics design of 14 MeV neutron generator facility at the Institute for Plasma Research

A high energy and high yield neutron source is a prime requirement for technological studies related to fusion reactor development. It provides a high-energy neutron environment for small-scale fusion reactor components research and testing such as tritium breeding, shielding, plasma-facing materials, reaction cross-section data study for fusion materials, etc. Along with ITER participation, the Institute of Plasma Research, India is developing an accelerator-based 14 MeV neutron source with a yield of 1012 n s−1. The design of the source is based on the deuterium–tritium fusion reaction. The deuterium beam is accelerated and delivered to the tritium target to generate 14 MeV neutrons. The deuterium beam energy and tritium availability in the tritium target are the base parameters of the accelerator-based neutron source design. The paper gives the physics design of the neutron generator facility of the Institute for Plasma Research. It covers the requirements, design basis, and physics parameters of the neutron generator. As per the analytical results generator can produce more than 1 × 1012 n s−1 with a 110 keV D+ ion beam of 10 mA and a minimum 5 Ci tritium target. However, the detailed simulation with the more realistic conditions of deuteron ion interaction with the tritium titanium target shows that the desired results cannot be achieved with 110 keV. The safe limit of the ion energy should be 300 keV as per the simulation. At 300 keV ion energy and 20 mA current, it reaches 1.6 × 1012 n s−1. Moreover, it was found that to ensure sufficiently long operation time a tritium target of more than 20 Ci should be used. The scope of the neutron source is not limited to the fusion reactor research studies, it is extended to other areas such as medical radioisotopes research, semiconductor devices irradiations, and many more.