Reactor Components Research Articles

Enhanced heavy oil fouling resistance of cobalt-based spray-deposited coatings by alloy design and heat treatment

Heavy oil hydrocracking fouls critical reactor components, causing premature turnarounds. Developing materials with superior resistance to coking, sulfidation, and wear proves crucial. High-velocity oxy-fuel (HVOF) coating improves materials' properties as an industrial-ready solution. This study explores the effect of alloying elements on the HVOF-sprayed CoCrMoSi coating model and their role in protecting against heavy oil fouling. Heat treatment reduced the coatings' inhomogeneity and microstructural defects, substantially improving their protection against fouling. A novel optical image treatment has been developed to semi-quantitatively assess the surface fouling intensity. The study provides insights into the complex interplay between alloy composition, microstructure, heat treatment, and surface energy in optimizing coating protection against heavy oil fouling.

Measurement of gamma field inside the biological concrete shielding of VVER-1000 Mock-Up at the LR-0 reactor

The long-term operation of existing nuclear power reactors is a crucial concern due to the complexities and expenses associated with replacing key components, such as the reactor pressure vessel and reactor internals. Gamma radiation, a byproduct of nuclear reactions and radioactive decay, significantly influences the lifetime of these components. This radiation is responsible for various degradation pathways leading to void swelling in steel reactor components and cracking or other radiation damage in concrete structures.A study conducted at a full-scale mock-up of the VVER-1000 reactor at the LR-0 zero-power reactor employed HPGe and stilbene measurements to analyze gamma spectra behind the reactor pressure vessel and within concrete biological shielding. While simulations behind the reactor pressure vessel aligned with measurements, notably, amarked overestimation of stilbene spectrum calculations occurred deep in concrete, suggesting potential inaccuracies in radiation predictions for power plant structures.

Corrosion behaviour of nitroxy-QPQ additively manufactured 17-4 PH steel on marine and nuclear reactor components

The study investigates the corrosion behavior of untreated and nitroxy QPQ treated 17-4 PH stainless steel under NaCl environments. Potentiodynamic polarization tests reveal that the nitroxy layer forms an oxide coating, leading to a passivation phase with gradually increasing current density. Oxidation at different temperatures shows variation in corrosion potential, with the lowest current density observed at 450 °C, indicating enhanced corrosion resistance. SEM images of untreated steel show severe crevice and pitting corrosion, whereas nitroxy QPQ treated surfaces exhibit reduced corrosion, with minimal disturbance at 450 °C. Elemental analysis confirms the formation of oxides and nitrides, with increased oxygen content in oxide layers and decreased nitrogen content due to the formation of chromium nitrides. The oxide layer, enriched with oxygen, nitrogen, and chromium, acts as a passive film, protecting against corrosion even in NaCl environments. Elemental mapping further illustrates the correlation between oxidation temperature and corrosion resistance, with higher temperatures resulting in reduced corrosion rates. The combined nitriding and oxidation process enhances corrosion resistance by forming a dense and adherent oxide layer, serving as a barrier against corrosive agents.

SAM: A Modern System Code for Advanced Non-LWR Safety Analysis

The System Analysis Module (SAM), developed at Argonne National Laboratory and by collaborators at other organizations, is for advanced non–light water reactor safety analysis. SAM aims to provide fast-running, modest-fidelity, whole-plant transient analysis capabilities that are essential for fast-turnaround design scoping and engineering analyses of advanced reactor concepts. To facilitate code development, SAM utilizes the MOOSE object-oriented application framework, its underlying finite element library, and linear and nonlinear solvers to leverage modern advanced software environments and numerical methods. SAM aims to solve tightly coupled physical phenomena, including fission reaction, heat transfer, fluid dynamics, and thermal-mechanical responses in advanced reactor structures, systems, and components with high accuracy and efficiency. This paper gives an overview of the SAM code development, including goals and functional requirements, physical models, current capabilities, verification and validation, software quality assurance, and examples of simulations for advanced nuclear reactor applications.

Using the Radial-Shear Rolling Method for Casted Zirconium Alloy Ingot Structure Improvement

In developing materials for the nuclear industry, it is crucial to enhance both alloy composition and processing methods. This study focuses on investigations of applying radial-shear rolling (RSR) to a Zr-1%Nb alloy ingot, aiming to refine its microstructure and improve its properties for nuclear applications. This method, with complex vortex metal flow inside of a casted workpiece, has not been previously tested for processing zirconium ingots, so experimental verification of its applicability is of scientific interest. The 30 mm diameter ingot, produced by vacuum induction melting, was initially rolled to 20 mm at 800 °C to eliminate defects and refine the cast structure. A second rolling stage reduced the diameter to 13 mm at 530 °C, resulting in an ultrafine-grained structure. The RSR method effectively combines structural refinement and defect healing within fewer cycles, making it suitable for producing components for nuclear reactors. This approach demonstrates a potential reduction in traditional processing steps, providing a more efficient route for preparing high-quality materials for nuclear applications.

Influence of temperature gradients on helium diffusion and clustering in tungsten: A molecular dynamics study

Tungsten and its alloys, used as plasma-facing materials in fusion reactors, endure high-flux, low-energy helium ion irradiation and temperature gradients induced by thermal load and shocks. This study utilizes molecular dynamics (MD) simulations to explore the diffusion and clustering behavior of helium (He) in tungsten (W) under temperature gradients. The migration behavior of isolated helium atoms, small helium clusters, and helium self-interstitial atom (He-SIA) clusters within tungsten is analyzed. It is revealed that both helium and He-SIA clusters tend to migrate towards higher temperature regions, exhibiting negative thermophoresis. The nucleation of helium clusters, the formation of Frenkel pairs, and the interactions between self-interstitial clusters and helium clusters are also investigated. The results indicate that helium concentration significantly impacts He nucleation behavior. Directional diffusion may lead to areas of high concentration over time, potentially promoting the formation of He clusters and bubbles. These insights enhance the understanding of tungsten's performance and durability as a plasma-facing component in fusion reactors.

Fast low-temperature irradiation creep driven by athermal defect dynamics

The occurrence of high stress concentrations in reactor components is a still intractable phenomenon encountered in fusion reactor design. Here, we observe and quantitatively model a non-linear high-dose radiation mediated microstructure evolution effect that facilitates fast stress relaxation in the most challenging low-temperature limit. In situ observations of a tensioned tungsten wire exposed to a high-energy ion beam show that internal stress of up to 2 GPa relaxes within minutes, with the extent and time-scale of relaxation accurately predicted by a parameter-free multiscale model informed by atomistic simulations. As opposed to conventional notions of radiation creep, the effect arises from the self-organisation of nanoscale crystal defects, athermally coalescing into extended polarized dislocation networks that compensate and alleviate the external stress.

Design of experimental facility for helium gas component testing

The current state-of-the-art helium gas test loops for component testing are constrained by the maximum operating temperatures, pressures, and flow rates of such facilities. Idaho National Laboratory is designing a new experimental facility known as HECTOR (HElium Component Testing Out-of-pile Research), which is a helium gas test loop for evaluating the performance of components to helium gas pressures, temperatures, and flow rates up to 8 MPa, 800 °C, and 0.15 kg/s, respectively. The construction of this facility will enable researchers to perform experiments across a wide range of Reynolds and Nusselt numbers, which could result in the maturation of technology for high-temperature gas-cooled reactor (HTGR) components.

Insight into particle size distribution and particle-scale reactions in a supercritical water fluidized bed reactor by CPFD method

Supercritical water gasification (SCWG) provides a promising solution for converting coal into hydrogen-rich gaseous. However, the modelling of the coal SCWG process still poses challenges due to the heterogeneous particle distribution, large quantity of particles, and complex chemical reactions. This study presents a novel three-dimensional numerical model based on computational particle fluid dynamics (CPFD) to study coal particle gasification in the reactor. The model integrates realistic coal particle size distribution, heat and mass transfer phenomena alongside heterogeneous/homogeneous chemical reactions across phases, elucidating multi-scale reacting flow behaviors and illustrating particle volume fraction, size, and velocity distributions. The obtained results reveal the evolution of coal composition, particle scale gasification process, and distribution of gaseous product components in the supercritical fluidized bed reactor. This work aims to serve as a reference for understanding the flow and heat transfer phenomena between supercritical water and coal particles in SCWG reactors.

Helium retention feature in the boron deposited layer on tungsten substrate by laser-induced breakdown spectroscopy and machine learning approach

ITER is designed for a burning plasma operation in which Tungsten (W) tiles are used as the first wall and diverter materials. Studying He dynamics in B-coated W wall is essential to understanding the effect of boronized plasma-facing components in fusion reactors, as these post-conditioning materials significantly influence helium retention and release. Plasma-wall interactions (PWI) are an important issue in ITER fusion reactors. PWI would lead to wall erosion and impurity redeposition. As a product of the D-T burning plasma, helium (He) ash would be retained or co-deposited on plasma-facing components (PFCs), affecting the stable operation of burning plasma. Laser-induced breakdown spectroscopy (LIBS) is proposed as a promising in-situ diagnostic approach for monitoring D/T and He retention and impurity deposition on PFCs of tokamak devices. In this study, the LIBS technique was used to investigate the helium retention feature in the boron (B) layer on tungsten substrate in 10−5 mbar. Five He-retention samples on the boron deposition layer on tungsten substrates were prepared by pulsed laser deposition (PLD) method at different ambient pressures in our lab. The investigations indicate that the atomic spectral line of helium (He-I 587.56 nm) was observed in the spectra of the first three laser shots. The depth profiles of He, B, and W in the boron-deposited layer on tungsten substrate were performed by LIBS to determine the co-deposition layer thickness. The concentration of He in the co-deposition layer samples measured by TDS is 7 × 1020 He/m2. The plasma parameters, such as plasma electron temperature and electron number density, were calculated to validate the local thermodynamic equilibrium. Machine learning (ML) algorithms are used to classify the co-deposits and substrates. The first three principal components (PC1, PC2 & PC3) of the unsupervised ML algorithm (PCA) give a classification accuracy of 91.7 %. The supervised ML algorithm neural network achieved training and testing accuracy of 100 % and 96.7 %, respectively.

Fuzzy Logic-Based Feedwater Controller for the Super Fast Reactor

Abstract Super fast reactor is a Generation IV reactor with water coolant at supercritical pressure. It has advantages of having high efficiency of about 44%, more economical and excellent safety features. Its last core design which adopted one heating in one coolant cycle, however, results a large density change along the fuel channels. It leads to a sensitive steam temperature change against a change of power which might mismatch with the flow change. The sensitivity should be reduced to maintain the structure integrity of the reactor components. A feedwater controller which based on fuzzy logic was used to reduce the sensitivity. Method of fuzzy-based control is applied. Two signals of the temperature and the power changes were taken as the inputs of the fuzzy controller, and change of the feedwater coolant flow was taken as the control output. Fuzzifications used triangular member functions with Tsukamoto fuzzy model for inferencing process. Two control parameters of member function spread and signal values which lead to fuzzy degree of 1 were optimized so that the criteria of outlet coolant temperature and the stability were fulfilled. Test results of the control system showed that the deviation of the water temperature at outlet side could be kept within the criteria when mismatch change of power and water flow happened. The temperature deviation could be supressed by using the fuzzy-based control.

Kinetic Monte Carlo modelling of nano-oxide precipitation and its associated stability under neutron irradiation for the Fe-Ti-Y-O system

While developing nuclear materials, predicting their behavior under long-term irradiation regimes spanning decades poses a significant challenge. We developed a novel Kinetic Monte Carlo (KMC) model to explore the precipitation behavior of Y-Ti-O oxides along grain boundaries within nanostructured ferritic alloys (NFA). This model also assessed the response of the oxides to neutron irradiation, even up simulated radiation damage levels in the desired long dpa range for reactor components. Our simulations investigated how temperature and grain boundary sinks influenced the oxide characteristics of a 12YWT-like alloy during heat treatments at 1023, 1123, and 1223 K. The oxide characteristics observed in our simulations were in good agreement with existing literature. Furthermore, the impact of grain boundaries on precipitation was found to be minimal. The resulting oxide configurations and positions were used in subsequent simulations that exposed them to simulated neutron irradiation to a total accumulated dose of 8 dpa at three temperatures: 673, 773, and 873 K, and at dose rates of 10−3, 10−4, and 10−5 dpa/s. This demonstrated the expected inverse relationship between oxide size and dose rate. In a long-term irradiation simulation at 873 K and 10−3 dpa/s was taken out to 66 dpa and found the oxides in the vicinity of the grain boundary were more susceptible to dissolution. Additionally, we conducted irradiation simulations of a 14YWT-like alloy to reproduce findings from neutron irradiation experiments. The larger oxides in the 14YWT-like alloy did not dissolve and displayed stability similar to the experimental results.

Understanding the role of alloying elements Cr, Ni in erosion–corrosion process of nuclear grade austenitic stainless steel 304 L by total reflection X-ray fluorescence

Stainless steel (SS) has gained an indispensable position as one of the most widely used industrial material. In nuclear industry, it is one of the technologically most important structural materials used in pressure tube, containment vessel and reprocessing-related applications. In fact, SS 304 L is the workhorse material of the reprocessing plant which faces the hostile conditions of acid, temperature as well as high radiation dose. These conditions can affect the erosion-corrosion behavior of the material and can lead to poor mechanical properties. This also affects the life of the reactor components. In the present study, a systematic work has been carried out to develop an in-depth understanding of the Corrosion Rates (CRs) with respect to the loss of major and minor elements from SS into the corrosive nitric acid medium using Total reflection X-Ray Fluorescence (TXRF). The effect of acid molarity as well as temperature, on the erosion-corrosion behavior of SS304L was studied. The TXRF studies have revealed that the CR was maximum in 2 M HNO3 and minimum in 8 M, showing that as the nitric acid molarity has an important role to play in the corrosion and as the molarity increases, the passivation of SS 304 L also increases. This observation was further corroborated with the concentrations of Fe, Cr and Ni, which had leached into the nitric acid medium. The role of alloying elements, Cr and Ni, in the passivation of SS was also studied. The effect of temperature showed that at 45 °C, the rate of corrosion was minimum, as compared to at 25 °C and 90 °C which reveals that the formation of chromium oxide passive layer was thermodynamically most favorable at this temperature.

Shutdown dose rate calculation for fission reactors: An application of the MS-CADIS method to OPAL

There have been considerable advances of shutdown dose rate (SDDR) calculation methods for fusion related problems, however applications of these methods in the fission reactor field have hitherto been sparse. The present study attempts to bridge this gap by investigating the applicability of SDDR calculation methods for fission reactor problems. Specifically, we aim to assess and validate whether recent advances in SDDR methods can be successfully applied in fission research reactors. To this end, we estimate the shutdown dose rate distribution at the Open Pool Australian Light water reactor (OPAL) using the rigorous two step (R2S) computational method, and we compare the calculated results with the experimental data. This method utilizes a 3D reactor model implemented in the Monte Carlo N-Particle (MCNP) transport code, the AutomeD VAriaNce reduction Generator (ADVANTG) code for geometry discretization and variance reduction calculations, and the Oak Ridge Isotope GENeration (ORIGEN) inventory code for activation calculations. To ensure robustness, we employ two variance reduction techniques, Forward Weighted Consistent Adjoint Driven Importance Sampling (FW-CADIS) and Multi-Step Consistent Adjoint Driven Importance Sampling (MS-CADIS). To the best of our knowledge, this is the first MS-CADIS method implementation for fission reactor problems. The SDDR is estimated at ten locations within the experimental hall, all situated more than 4 m away from the reactor core.The paper shows that, the experimental observations are within the lower and upper bounds of the simulation results for 4 out of 10 locations, while the remaining observations are within a factor of 7, with one significant outlier. The calculated average dose rate is within 5% of the nominal values of the experimental observations for 3 locations. The computational results are within statistical uncertainty by using two different variance reduction techniques, with significant computational advantage of MS-CADIS over FW-CADIS for SDDR calculations. The results indicate that the combination of SDDR distribution maps, estimated dose rate energy dependance, and activation information are powerful tools in identifying the radioisotopes and reactor components dominating the SDDR. These results can contribute to better radiation safety practices in contaminated areas, by enabling the minimal dose path planning or by improving radiation shielding.

Microstructural constituents in compositionally graded NiCrFeSiBC hardfacing alloy/316L SS additive manufactured deposits

Ni-Cr-Fe-Si-B-C hardfacing alloys (D50) provide excellent resistance to abrasive and adhesive wear, high temperature oxidation and corrosion. These alloys are candidate materials for hardface coatings on in-reactor components of fast breeder reactors. Conventionally these deposits are made on steels through the welding route. Due to high fluidity of the molten alloy, generation of residual stress and significant dilution of the deposit, achieving crack free deposits is often a very difficult and time consuming task. To address such issues arising out of conventional fabrication routes, compositionally graded coatings have been tried through laser additive manufacturing. A 3 layered compositionally graded deposit (D50/50D50-50SS/SS) was made on 316 L SS build plate and the phase constituents and possible correlation with micro hardness was analyzed. Uppermost D50 layer consisted of γ-Ni, γ-Ni + Ni3B and γ-Ni + Cr5B3 eutectics, CrB, Cr7C3, Cr23C6, Cr3C2 and Ni3Si precipitates. The middle layer consisted of γ-(Ni,Fe), γ-(Ni,Fe)+Cr5B3, Cr5B3, Cr3C2, Cr7C3,Cr23C6 and Ni3Si. The SS layer closest to the build plate had γ-Fe, Cr3C2 and Cr7C3 precipitates. Corresponding to the variation in microstructure a gradual change in hardness also can be seen starting from the top layer to SS build plate: 580 ± 35 HV0.1, 532 ± 17 HV0.1, 348 ± 21 HV0.1 and 230 ± 10 HV0.1, which is promising.

Study on the fracture behavior and toughening mechanisms of continuous fiber reinforced Wf/Y2O3/W composites fabricated via powder metallurgy

Tungsten (W) is a promising candidate material for the plasma facing components in fusion reactors. However, it has issues regarding the intrinsic brittleness. Tungsten fiber reinforced tungsten composites (Wf/W) have been developed based on the concept of extrinsic toughening mechanisms and they show a pseudo-ductile behavior during the fracture process. In the present work, continuous fiber reinforced Wf/Y2O3/W composites were fabricated via a powder metallurgy (PM) process, and the microstructure and mechanical properties were characterized. The fracture behavior and toughening mechanisms were analyzed in detail combining the results of experiments and numerical simulation. The Wf/Y2O3/W composites is toughened by multiple mechanisms such as fiber bridging, crack bending and deflection, interface de-bonding and plastic deformation of fiber. The energy dissipation by interface de-bonding can be neglected. However, it is a necessary factor to ensure any extrinsic toughening mechanisms. The main contribution of the energy dissipation while composite failure is the plastic deformation of fibers.

Effects of irradiation on interfacial strength and microstructure of double-layer mullite and alumina coating on SiC

Silicon carbide (SiC) ceramics hold great potential for use in nuclear-reactor components due to characteristics such as high-temperature strength and low activation. Despite their resistance against corrosion in harsh environments, they suffer elevated corrosion rates under particle irradiation. Thin anti-corrosion coatings, such as mullite bond layers and alumina top layers, are essential to enhance the irradiation stability of SiC. The objective of this study was to evaluate the irradiation stability of a double-layer coating developed to enhance the corrosion resistance of SiC in nuclear applications. The coating, which comprised a mullite bond layer and an alumina top layer, was applied to a SiC substrate using laser chemical vapor deposition. Irradiation experiments were conducted at 300 °C with 5.1-MeV Si ions up to 10 displacements per atom. To assess the interfacial strength, a novel testing method, called the double-notch shear compression testing method, was developed based on ASTM standards and was implemented using a nanoindenter. This approach enabled precise measurements of the mechanical integrity at the interfaces of the coating under irradiation. The results showed an increase in the interfacial strength at the SiC/mullite and alumina/mullite interfaces with irradiation. Microstructural analysis of the fracture surface through scanning electron microscopy–energy-dispersive X-ray spectroscopy revealed that cracks propagated within the mullite layer, indicating the presence of mullite on the fracture surface. Transmission electron microscopy (TEM) images indicated that a 2-µm-thick transition layer existed at the SiC/mullite interface but not at the alumina/mullite interface. The TEM–electron energy loss spectroscopy suggested that the Al-O bonding structures in the transition layer were changed from tetrahedral (AlO4) to octahedral (AlO6) through irradiation, and this structural transition may directly affect the strength.

One-step fabrication of AlPO4 nanosheet reinforced ceramic coatings with improved fracture toughness and hydrogen resistance

Ceramic coatings have great potential in protecting vital metal structural components in hydrogen energy and nuclear fusion reactors. Fracture toughness and hydrogen permeation resistance are two key characteristics that determine life span and performance of these coatings. However, there is still no strategy to simultaneously enhance both of these two characteristics. Herein, we developed a one-step approach to fabricate AlPO4 nanosheet reinforced α-Al2O3 ceramic coating with simultaneously enhanced fracture toughness and hydrogen permeation resistance. Different from the external nanosheet addition method of the conventional techniques for fabricating nanosheet reinforced coatings, AlPO4 nanosheets were in-situ formed within α-Al2O3 ceramic coating though a one-step heat treatment process. Owing to the positive reinforcing effect of AlPO4 nanosheets, the fracture toughness of the coating increases by 3 times, achieving an ultrahigh KIC value of 7.1 MPa/m2. Meanwhile, the hydrogen permeation resistance of the coating increases by 78 %, reaching as high as 3440 times that of the steel substrate. Additionally, the Cr–P bonds formed between the coating and substrate ensure their good bonding, with their bonding strength increasing up to 36 MPa. The combination of high hydrogen permeation resistance, high fracture toughness, and high bonding strength makes the AlPO4 nanosheet reinforced ceramic coating a promising alternative for reducing hydrogen permeation in various hydrogen-related fields. This study also provides critical insights and practical guidelines for constructing high-performance functional coatings via nanosheet reinforcement.

Effects of triaxial rolling on the microstructure and installation characteristics of reactor pressure vessel studs

Reactor pressure vessel (RPV) studs are key components of nuclear reactors, and their connection with flange ensures the sealing of the RPV under high-pressure and high-temperature conditions. In the present work, the external threads of the RPV stud were prepared by triaxial rolling, and the texture evolution of the external thread root material of an RPV stud was predicted by finite element analysis coupled with viscoplastic self-consistent simulations. The microstructure of the external thread root material of RPV stud was characterized by scanning electron microscope and electron back-scattered diffraction. The installation characteristics of the turned and rolled parts of the RPV stud were tested by installation and pretightening tests. It was found that the dynamic recrystallization at the external thread root formed ultrafine tempered sorbite grains, high-angle grain boundaries (47%), and strong {111} <110> and {111} <112> textures. In the installation and pretightening test, the residual elongation of rolled parts was reduced by 6% under the same loading pressure. The triaxial rolling process distributed the microstructure of the external thread root of the RPV stud in a gradient manner, resulting in improved stud installation characteristics.

Long-term thermal aging effects in ferritic-martensitic steel HT9

The ferritic-martensitic steel HT9 is a candidate material for fuel cladding and core components in advanced nuclear reactors, such as sodium-cooled fast reactors, thanks to their high temperature mechanical properties and low susceptibility to irradiation induced swelling phenomena. However, thermal stability and elevated temperature microstructural evolution in these alloys may impact their long-term behavior and reliability. In this work, the effects of thermal aging on the microstructural and mechanical properties of HT9 have been investigated through complementary electron microscopy, synchrotron X-ray diffraction, microhardness, and thermodynamic modeling. Plates of HT9 were aged up to 50 kh at relevant sodium-cooled fast reactor operational temperatures (360 °C - 700 °C). Trends in microstructure as a function of aging time and temperature were apparent from qualitative and quantitative analysis. These observations were further supported by thermodynamic modeling of the bulk and precipitate phases. Specific phases observed include BCC Fe, FCC M23C6, HCP and FCC MX phase and Laves M2X phase. Through the application of our multi-scale and multi-modal approach, clear information on the aging mechanism of HT9 was obtained, allowing for a more informed prediction, and understanding of the long-term behavior, performance and thermal stability of ferritic-martensitic alloys exposed to elevated temperatures.