Reactor Environment Research Articles

Evaluation of the thermal conductivity of chromium coatings for accident tolerant fuels using thermoreflectance techniques

The deployment of Cr coatings as accident tolerant fuel cladding inside nuclear reactors necessitates a thorough understanding of their performance, including mechanical, chemical, irradiation, and thermal properties. While significant research has delved into the former, the thermal transport properties of Cr coatings, which dictates fuel temperature and impacts fuel safety, have been relatively underexplored. This work presents an investigation into the thermal conductivity of 12 µm-thick Cr coatings fabricated by multi-arc ion plating employing high-resolution frequency and spatial domain thermoreflectance techniques. Vacuum annealing was utilized to control the morphology of coatings to explore the impact of grain size. The results indicate that both as-deposited and annealed coatings, despite possessing distinct grain morphologies, exhibit thermal conductivities similar to their bulk counterparts. The influence of boundaries on thermal transport is anticipated to become prominent as the critical dimension diminishes below 0.5 μm. Future investigation is required to address the potential impact of adverse reactions, such as oxygen ingress and irradiation, on the thermal properties of Cr coatings to ensure safe and efficient operation in nuclear reactor environments.

Methane seepage leads to a specific microplastic aging process in the simulated cold seep environment

Marine microplastics pose a significant threat to ecosystems, and deep-sea regions serve as critical sinks for these pollutants. Among these regions, cold seeps harbor relatively high concentrations of microplastics. However, research on the aging of microplastics under low-temperature, dark, methane-abundant, and high-pressure conditions remains limited. Seawater and sediment were collected from various Haima cold seepage sites to simulate seepage environments in 200-mL high-pressure reactors. Four types of microplastics at high concentrations (approximately 10 %) were cultured and monitored over two months to explore how they aged. The key findings are as follows: (1) Compared to areas of weak seepage, methane seepage accelerated microplastic aging, as evidenced by increased surface roughness, enhanced C-O and (CO)-O bond formation, increased microbial colonization, and reduced contact angles. (2) Microplastic aging is more pronounced in sediments than in seawater, with biodegradable polylactic acid (PLA) exhibiting the most significant aging characteristics and carbon contribution. (3) Aged microplastics induce greater disturbances in inorganic nutrient levels than in organic matter, impacting nitrogen cycle processes involving nitrate, nitrite, and ammonium. This study results reveal the fundamental aging characteristics of microplastics in extremely deep seas and highlight their potential ecological effects.

Microstructure, electrical resistivity, and tensile properties of neutron-irradiated Cu–Cr–Nb–Zr

High strength, high conductivity copper alloys that can resist creep at high temperatures are one of the primary candidates for efficient heat exchangers in fusion reactors. Cu–Cr–Nb–Zr (CCNZ) alloys, which were designed to improve the strength and creep life of ITER Cu–Cr–Zr (CCZ) reference alloys, have been found to have comparable electrical conductivity and tensile properties to CCZ alloys. The measured creep rupture times for these improved alloys is about ten times higher than the ITER reference alloys at 90–125 MPa at 500 °C. However, the effects of neutron irradiation on these alloys, and the ensuing material properties, have not been studied; thus, their utility in a fusion reactor environment is not well understood. This study characterizes the room temperature mechanical and electrical properties of a neutron-irradiated CCNZ alloy and compares them to a neutron-irradiated ITER reference heat sink CCZ alloy. Tensile specimens were neutron irradiated in the High Flux Isotope Reactor (HFIR) to 5 dpa between 250 °C and 325 °C. Post-irradiation characterization included electrical resistivity measurements, hardness, and tensile tests. Microstructural evaluation used scanning electron microscopy, energy dispersive x-ray spectroscopy, and atom probe tomography to characterize the irradiation-produced changes in the microstructure and investigate the mechanistic processes leading to post-irradiation properties. Transmutation calculations were validated with composition measurements from atom probe data and used to calculate contributions to the increased electrical resistivity measured after irradiation. Comparisons with CCZ alloys in the same irradiation heat found that the post-irradiated CCNZ and CCZ alloys had comparable electrical resistivity. Although CCNZ alloys suffered more irradiation hardening than CCZ, the overall tensile behavior deviated very little from non-irradiated values in the temperature range studied.

Investigation of neutron irradiated W/CuCrZr joints

This study investigates the effects of neutron irradiation on tungsten (W) and copper-chromium-zirconium (CuCrZr) joints under conditions mimicking the high neutron flux environment of a tokamak fusion reactor. Samples of W/CuCrZr joints were subjected to irradiation in the Belgian Reactor 2 (BR2) nuclear reactor at SCK CEN (Belgian Nuclear Research Centre) to simulate the intense neutron exposure characteristic for International Thermonuclear Experimental Reactor (ITER) and DEMOnstration power plant reactor (DEMO) operations. The primary objective was to evaluate changes in the mechanical properties and microstructure of these materials, which are critical for their potential use in plasma-facing components.It is revealed that a significant reduction in tensile elongation of the joint, indicating some degree of embrittlement, is observed after the irradiation. Importantly, this effect is independent of the irradiation temperature. Possible physical reasons for the observed phenomenon are discussed.

Formation of hydrogen resistant oxides of titanium alloy in water corrosion environment

The structure changes and hydrogen permeation behavior of Titanium alloy TA17 in simulated water environment of small modular reactor (SMR) at 320°C for 8000h was investigated. We found a corrosion layer consist of TiO2 and FeTiO3 with a good hydrogen permeation resistance. The structure and hydrogen permeation resistance change with the increase of corrosion period. At the first corrosion period of 3000h, FeTiO3 grows rapidly with an exponential parameter of 0.663 fitted by parabolic law. As the initial TiO2 layer destroyed, the deuterium permeation flux increases. After the thickness of FeTiO3 reaches a limit at corrosion time of 4000h, there exists a dynamic balance between the growth and the dissolution of FeTiO3, result in a small fluctuant deuterium permeation flux at corrosion periods of 5000－8000h. The deuterium permeation flux of the corrosion layer can be two orders of magnitude lower to that of Titanium alloy TA17 substrate. The formation process of corrosion layer and its effect on hydrogen permeation resistance are discussed in detail.

Microbial approach towards anode biofilm engineering enhances extracellular electron transfer for bioenergy production

Applying microbial electrolysis cells (MEC) is a biological approach to enhance the growth of high amounts of electroactive biofilm for extracellular electron transfer. The electroactive biofilm degrades the organics by oxidizing them at the anode and producing electrical energy. Addition of waste-activated sludge (WAS) with fat grease oil (FOG) produces an optimal reactor environment for microbial growth to enhance the exchange of electrons between cells via microbial electrolysis. The present work aimed to investigate the microbial approach to increase the extracellular electron transfer (EET) in microbial electrolysis cells. Results revealed that metabolites in electroactive microbes (EAM) grow viable cells that initiate high EET at anode sites. At optimum WAS with FOG addition, volatile fatty acid and current generation yield production was 2.94 ± 0.19 g/L and 17.91 ± 7.23 mA, accompanied by COD removal efficiency of 89.5 ± 14.4%, respectively. This study introduces a novel approach to anode biofilm engineering that significantly enhances extracellular electron transfer, offering a fresh perspective on bioenergy production. Our approach, which demonstrates that anodic biofilm enhances intercellular electron transfer, increases NADH-NAD ratio, and increases metabolite yield-fluxes, has the potential to revolutionize bio-electricity production. Results indicated that the electrolysis highlights MEC performance in power generation of 788 mV with 200 mL of anode volume of active viable cells by utilizing WAS with 11% FOG. The achievements of this study provide critical parameters for the anode biofilm engineering, demonstrating how growth cell volume, intercellular electron transfer, and increases in NADH-NAD ratio are evidence of an increase in the EET, compelling evidence for the resilience treatment and efficient current production. These findings are significant in advancing our understanding of bioenergy production.

Advanced corrosion-resistant materials for enhanced nuclear fuel performance: A conceptual review of innovations in fuel cladding against molten salt degradation

The increasing demand for efficient and sustainable nuclear power has led to the development of advanced corrosion-resistant materials that can significantly enhance nuclear fuel performance. This conceptual review focuses on the innovations in fuel cladding materials designed to resist degradation in molten salt environments, which are crucial for advanced nuclear reactors like molten salt reactors (MSRs). Fuel cladding serves as a critical barrier between the nuclear fuel and the reactor environment, and its ability to withstand extreme temperatures and corrosive conditions is essential for safe and efficient reactor operation. Recent advancements in materials science have introduced novel alloys and coatings that demonstrate superior corrosion resistance in molten salts, which are used as both coolants and fuel solvents in MSRs. These innovations include high-temperature nickel-based alloys, refractory metals, and ceramic coatings, which are engineered to reduce the effects of molten salt corrosion, such as oxidation, pitting, and embrittlement. By enhancing the durability of fuel cladding, these materials contribute to improved fuel performance, longer reactor lifespans, and increased safety. This review explores key developments in corrosion-resistant materials, emphasizing the mechanisms by which these materials mitigate molten salt degradation. Additionally, it highlights the challenges associated with material selection, fabrication, and long-term performance in the highly corrosive environments of advanced nuclear reactors. Ongoing research into these materials offers promising avenues for the future of nuclear energy, particularly in addressing the limitations of current fuel cladding technologies. In conclusion, the use of advanced corrosion-resistant materials represents a significant step toward enhancing the performance and safety of nuclear fuel, ensuring the viability of MSRs and other advanced reactor designs. Continued innovation in this field is essential for the future of nuclear power.

Tailoring ZrWNb-Based Refractory High Entropy Alloys: A Multidirectional Characterization Through Elemental, Physical, Thermodynamic, and Nuclear Radiation Attenuation Properties

This study explores the development and characterization of 18 novel ZrWNb[(X)(Y)] (where XY are Hf, Ta, Mo, V, Ti) refractory high entropy alloys (RHEAs) designed to enhance mechanical robustness, thermodynamic stability, and radiation shielding effectiveness. Through a rigorous analysis, we evaluated the elastic modulus, entropy, and enthalpy of mixing, atomic size difference, valence electron concentration, and the omega parameter to understand the alloys’ phase behavior and structural integrity. Our findings revealed a broad range of elastic modulus values, indicating diverse mechanical adaptability suitable for high-stress applications. The thermodynamic assessment highlighted several RHEAs with favorable phase stability, particularly ZrWNbTaHf and ZrWNbTiTaHf, which are poised for high-performance usage due to their balanced mixing entropy and enthalpy. This study also benchmarks the RHEAs against conventional alloys, with sample R1 exhibiting the lowest fast neutron effective removal cross section, underscoring its superior neutron shielding capabilities. Additionally, R1 demonstrated exceptional photon attenuation, as evidenced by its competitive half-value layer performance across an extensive energy spectrum. Collectively, these results position R1 as a standout candidate, offering a significant advancement in materials science for applications demanding stringent radiation shielding, such as in nuclear reactor environments and space exploration.

Open‐ESyn: A 3D‐Printed Toolkit for Expanded ElectraSyn Functionality and Reproducibility for Electrosynthesis

AbstractElectrosynthetic reactions are performed in either custom‐made reactors that are developed and machined in‐house or commercially available systems that offer good reproducibility but come at a high cost. To bridge this divide between customizability and reproducibility, we have developed the Open‐ESyn, which is a suite of 3D‐printed components compatible with the popular ElectraSyn. This collection of parts increases the electrosynthesis that can be performed with the ElectraSyn, expanding, for example, the scale, temperature and the type of electrodes that can be used. The standardized reactor environment can be inexpensively recreated, thereby maintaining the reproducibility of the ElectraSyn ecosystem.

Characteristics of oxide films on TA16 alloy after long-term exposure in simulated secondary circuit water environment of small modular reactor

Currently, little is known about the characteristics of oxide films on Ti alloys exposed to the secondary circuit water environment of small modular reactors (SMRs). In this study, multiple microstructural characterization techniques were employed to meticulously characterize the oxide film characteristics of the TA16 alloy after 6000 h and 8000 h of exposure to a simulated SMR secondary circuit water environment. The results indicate that the oxide films on the 6000 h specimen are composed of a microcrystalline anatase TiO2 interior layer and a coarse-grained anatase TiO2 exterior layer, while the oxide film on the 8000 h specimen typically consists of a microcrystalline anatase TiO2 inner layer, an NiFe2O4 middle layer, and a coarse-grained anatase TiO2 outer layer. As the corrosion time is extended to 8000 h, the presence of NiFe2O4 oxide particles can be attributed to oxide deposition by the 316L stainless steel autoclave used in the corrosion experiments, which results in the migration of Ni and Fe ions onto the surface of the Ti alloy.

A Novel Ultrasonic Leak Detection System in Nuclear Power Plants Using Rigid Guide Tubes with FCOG and SNR.

Leak detection in nuclear reactor coolant systems is crucial for maintaining the safety and operational integrity of nuclear power plants. Traditional leak detection methods, such as acoustic emission sensors and spectroscopy, face challenges in sensitivity, response time, and accurate leak localization, particularly in complex piping systems. In this study, we propose a novel leak detection approach that incorporates a rigid guide tube into the insulation layer surrounding reactor coolant pipes and combines this with an advanced detection criterion based on Frequency Center of Gravity shifts and Signal-to-Noise Ratio analysis. This dual-method strategy significantly improves the sensitivity and accuracy of leak detection by providing a stable transmission path for ultrasonic signals and enabling robust signal analysis. The rigid guide tube-based system, along with the integrated criteria, addresses several limitations of existing technologies, including the detection of minor leaks and the complexity of installation and maintenance. By enhancing the early detection of leaks and enabling precise localization, this approach contributes to increased reactor safety, reduced downtime, and lower operational costs. Experimental evaluations demonstrate the system's effectiveness, focusing on its potential as a valuable addition to the current array of nuclear power plant maintenance technologies. Future research will focus on optimizing key parameters, such as the threshold frequency shift (Δf) and the number of randomly selected frequencies (N), using machine learning techniques to further enhance the system's accuracy and reliability in various reactor environments.

Corrosivity of KOH(g) towards superheater materials in a simulated air reactor environment for chemical looping combustion of biomass

Chemical looping combustion (CLC) of biomass has the potential to improve the electrical efficiency in power plants that utilize carbon capture and storage (CCS). This is attributed to the mild corrosive environment anticipated in the air reactor (AR). However, previous studies have measured alkali emissions in the AR, likely in the form of KOH(g), suggesting that alkali may transfer from the fuel reactor (FR) to the AR. To investigate the corrosive effect of KOH(g) release in the AR, a novel experimental set-up was developed to simulate two scenarios for the AR 1) Clean scenario: no release of KOH(g) in the AR (5 % O2 + 3 % H2O + N2 bal.) and 2) Alkali slip scenario: continuous release of KOH(g) in the AR (5 % O2 + 3 % H2O + N2 bal. + 16 ppm KOH(g)). The exposure was carried out at 700 °C and six alloys, relevant as superheater material, ranging from stainless steels to nickel-base (Ni-base) alloys as well as a FeCrAl alloy were investigated. The samples were exposed for a total of 168 h and the morphology of the corrosion products was investigated using SEM-EDX and XRD. The presented results suggest that KOH(g) significantly accelerates the corrosion of the stainless steels and Ni-base alloys investigated, by rapidly destroying the protective Cr-rich oxide scale, resulting in the formation of a multilayer oxide scale with inferior protective properties. On the contrary, the FeCrAl alloy retained a protective Al-rich oxide scale irrespective of the presence of KOH(g). The findings in this study highlight that the release of KOH(g) in the AR during combustion of biomass in CLC could introduce corrosion challenges for installed superheaters that can be significantly mitigated by utilizing FeCrAl alloys.

Валідація пакета програм MCPV з безперервною за енергією бібліотекою мікроскопічних констант

The results of the validation of the code package MCPV for the calculation of radiation transport within VVER-1000 reactor type near-vessel space using a continuous-energy data library are presented. The validation is carried out based on the experimental data obtained in mock-up and full-scale experiments. It is shown that the code package MCPV can be used to simulate neutrons and gamma quanta transport through the complicated heterogeneous environment of a nuclear reactor and to obtain the valid values of the functionals of the neutron and gamma fluxes within VVER-1000 near-vessel space.

Lifetime of nitric oxide produced by surface dielectric barrier discharge in controlled atmospheres: Role of O2 content

Nitric oxide (NO) is an invaluable multifunctional chemical in agri-food applications; for example, it plays a role in plant growth, development, and stress tolerance. One of the interesting implications of NO is the suppression of fruit ripening and the resulting increase in shelf life. Since a high concentration of NO is producible in atmospheric plasmas, increasing attempts in plasma agriculture have been made to study several types of plasma reactors operated in air. However, as NO is rapidly oxidized by oxygen (O2) and/or ozone (O3), the use of NO produced in atmospheric plasmas is naturally nontrivial, and the lifetime of NO is highly sensitive to the reactor environments. Here, we investigated the time development of O3 and nitrogen oxides (NOx=1–3) in a surface dielectric barrier discharge (sDBD) reactor with respect to the O2 content of the controlled atmosphere (N2+O2). In situ optical absorption spectroscopy enabled the observation of the dynamics of O3 and NOx under gas-tight conditions. In these plasma reactors, NO became a dominant species after the chemical mode transition from O3 to NO occurred. We found that a lower O2 content correlated to a faster appearance of NO in the plasma reactor. The NO lifetime significantly increased as the O2 content decreased from 20 to 5 % in the plasma reactor, while the maximum concentration of NO decreased. Our findings indicated that the appropriate control of the O2 content is essential in atmospheric plasma reactors dependent on O3 and/or NOx applications.

Dynamic Deformation in Nuclear Graphite and Underlying Mechanisms.

Time-dependent deformation in nuclear graphite is influenced by the creation and migration of radiation-induced defects in the reactor environment. This study investigates the role of pre-existing defects such as point defect clusters and Mrozowski cracks in nuclear graphite IG-110. Separate specimens were irradiated with a 2.8 MeV Au2+ beam with a fluence of 4.38 × 1014 cm-2 and an 8 MeV C2+ beam with a fluence of 1.24 × 1016 cm-2. Microscopic specimens were either mechanically loaded inside a transmission electron microscope (TEM) or subjected to ex situ indentation-based creep loading. In situ TEM tests showed significant plasticity in regions highly localized around the Mrozowski cracks, resembling slip or ripplocation bands. Slip bands were also seen near regions without pre-existing defects but at very high stresses. Ex situ self-ion irradiation embrittled the specimens and decreased the creep displacement and rate, while heavy ion irradiation resulted in the opposite behavior. We hypothesize that the large-sized gold ions (compared to the carbon atoms) induced interplanar swelling as well as cross-plane channels for increased defect mobility. These findings illustrate the role of pre-existing defects in the dynamic relaxation of stresses during irradiation and the need for more studies into the radiation environment's impact on the mechanical response of nuclear graphite.

Incorporating Radioactive Decay Batteries into the USA's Energy Grid: Solutions for Winter Power Challenges

This paper explores the integration of radioactive decay batteries into the USA's energy grid as a strategic solution for addressing winter power challenges. Radioactive decay batteries, leveraging isotopes such as tritium or americium-241, offer significant advantages over traditional fission and fusion-based nuclear batteries. Their design inherently ensures enhanced durability, providing long-lasting energy outputs that can span several decades with minimal maintenance. Unlike conventional nuclear reactors, these batteries operate safely outside a reactor environment, significantly reducing the risks associated with nuclear power generation. This operational safety is further augmented by the relatively low radiation levels they emit, making them viable for diverse applications, including residential, commercial, and remote off-grid systems. The implementation of these batteries promises a stable and reliable energy supply during winter months when energy demands peak, mitigating the risk of power outages and enhancing the resilience of the energy grid. This study examines the potential deployment strategies, technical feasibility, and regulatory considerations necessary to incorporate radioactive decay batteries effectively into the USA's energy infrastructure, positioning them as a transformative technology for sustainable and reliable winter energy solutions.

Development and experimental evaluation of an acoustic flow meter prototype for measuring air flow

This study outlines the development and testing of a flow meter prototype specifically designed for use in nuclear reactor environments. The meter uses flow-induced vibration to measure flow rates and is designed for remote monitoring in harsh environments where radiation can damage electronics. The device generates a vibration signal produced by a feedback system when fluid flows out of an orifice and interacts with a downstream wedge. The frequency of the vibration signal corresponds directly to flow rate and velocity. The study investigates geometric parameters intrinsic to the prototype, such as orifice height, width, edge distance, and chamber length. Different configurations were also tested, including a bypass loop with variable cross-sections around the device. The investigation results reveal the impact of these geometric parameters on the vibration signal output, providing valuable insights for future design improvements and enhancing the effectiveness of flow metering systems used in nuclear reactor applications.

Corrosion behavior of additively manufactured FeCrAl in out-of-pile light water reactor environments

Iron-Chromium-Aluminum (FeCrAl) alloys are candidate materials for the cladding of light water reactor (LWR) fuels. The FeCrAl alloys in general range in Cr composition from 12% (C26M) to 21% (APMT). In this work, the general corrosion behavior of Additively Manufactured (AM) C26M coupons was compared to the behavior of traditional Powder Metallurgy (PM) coupons. Immersion testing were conducted for 12 months at 288 °C and 330 °C in pure water containing either oxygen or hydrogen. Results show that the mass change of AM specimens in hydrogenated water was like the mass change of PM specimens. In oxygenated water, the mass change of AM coupons was higher and less reproducible than for the PM coupons. Porosity in the AM specimens makes their behavior less predictable in high-temperature water.

Status and issues of high-temperature and high-pressure water corrosion research of fusion structural materials in Japanese DEMO reactor development

Abstract The activated corrosion product assessments of fusion structural materials are essential to designing components and evaluating workers’ radiation exposure. This paper first gives the R&D status of the high-temperature pressurized water corrosion study of reduced activation ferritic/martensitic steels and chromium–zirconium–copper (CuCrZr) alloys, which are the leading candidate materials of fusion reactor in-vessel components such as breeding blanket and divertor, which are utilized in high-temperature and high-pressure water, and the recent progress of corrosion test apparatus simulating the unique environment of a fusion reactor will also be presented.

Development of a boron-containing reduced activation Ferritic-Martensitic (B-RAFM) steel

Steel will be an essential part of any commercial fusion reactor design. Applications in this area involve extreme conditions, imposing particular performance requirements, such as high operational temperature and creep resistance, and also a limitation on the elements that can be used due to the activation that occurs on interaction with irradiation. This work begins with a steel developed for conventional power plant applications, the IBN1 grade developed by IMPACT (a UK consortium of industrial and academic research organisations). This grade has shown excellent properties at high temperature due to high temperature-stable precipitate phases, but contains several elements that would become radiologically active to a degree that is incompatible with the required disposal routes after exposure to the fusion reactor environment. In this study, modifications of the composition are made to remove these elements, and thermodynamic modelling and experimental assessment of the phases that form are undertaken. In this, we have paid particular attention to the prediction of transformation temperatures (to understand if normalisation and tempering can be applied successfully) and the precipitates, to see if suitable phases that are likely to impart creep strength and other desirable properties would be formed. The modifications made include the removal of Nb, Mo, Ni, Co, Cu and Al from the starting alloy, and the substitution of Ta (intended to form carbides, replacing the effect of Nb). Modifications of the amount of retained elemental components, such as C, Mn and Cr, have been made with Thermo-Calc modelling, to ensure preservation of comparable phase transformation temperatures and microstructures. The predicted changes to the alloy are compared to the observations from experimental investigation, finding that tantalum can substitute for niobium in these systems and form similar carbides with similar distribution in the material, and that reduction of Cr to 8 wt% and increase of C to 0.12 wt% raises the Ae4 temperature to allow a high-temperature heat treatment without δ-ferrite formation. While assessment of the mechanical properties of this alloy would be required, the perspectives for these alloys to perform at high temperature that can be inferred from the microstructure are discussed.