This article presents the development and commissioning of a supercritical carbon dioxide (sCO(2)) autoclave for high-temperature corrosion testing of structural materials under representative operating conditions. The facility is intended to support materials development and corrosion assessment for sCO(2)-based energy conversion systems, with particular relevance to advanced nuclear applications, including Generation IV reactors and small modular reactors (SMRs). The autoclave enables dynamic corrosion testing at elevated temperatures and pressures, with independent control and continuous monitoring of temperature, pressure, and moisture content in the sCO(2) environment. The performance of the system is demonstrated through corrosion testing of high-temperature-resistant Ni-based alloys and the advanced austenitic steel Sanicro 25. The observed corrosion behavior, governed by coupled oxidation and carburization processes, is consistent with previously reported studies, validating the capability of the experimental setup. The developed autoclave allows materials to be evaluated under conditions closely approaching those expected in practical sCO(2) power cycles, providing a reliable platform for materials screening and selection for advanced nuclear energy systems.
In the present study, the activity concentrations of the radionuclides 238U, 232Th, and 40K were determined in four groups of foodstuffs (meat, grains, vegetables, and fruits) commonly consumed by the Iraqi population. The assessment was conducted for three age groups: infants, children, and adults. Food samples were collected from local markets in Iraq and analyzed using gamma-ray spectrometry. The annual effective dose (AED) and excess lifetime cancer risk (ELCR) resulting from the ingestion of 238U, 232Th, and 40K were calculated for each age group. The mean activity concentrations (Bq & centerdot;kg-1) of 238U, 232Th, and 40K in meat samples were 0.49 +/- 0.16, 0.77 +/- 0.43, and 0.43 +/- 0.17, respectively. For grain samples, the corresponding values were 0.44 +/- 0.18, 0.47 +/- 0.13, and 0.42 +/- 0.17, respectively. In vegetable samples, the mean activity concentrations were 0.26 +/- 0.14, 1.41 +/- 0.48, and 0.61 +/- 0.23, respectively, while fruit samples exhibited mean values of 0.10 +/- 0.008, 0.61 +/- 0.06, and 0.63 +/- 0.09, respectively. The average total AED (mSv yr-1) from all analyzed food samples was 0.31 for infants, 0.22 for children, and 0.19 for adults. In addition, the average ELCR values (& times;10-3) for infants, children, and adults were 1.09, 0.69, and 0.68, respectively. Comparison of the estimated AED values with internationally recommended public exposure reference levels indicates that dietary intake of the investigated foodstuffs does not pose a significant radiological health concern for any age group.
The safe operation of medical cyclotrons used for producing PET radiopharmaceuticals like fluorine-18 necessitates a thorough understanding of the surrounding radiation fields. This study presents a comprehensive analysis of the neutron and gamma radiation environment around an IBA Cyclone 18/9. Using the MCNP6.2 code, a detailed model of the cyclotron vault was developed, incorporating radiation sources from both the 18O(p,xn) reaction in the water target and proton beam losses on accelerator components. The simulated dose equivalents were validated against experimental measurements from neutron and gamma probes located outside the vault. Results indicate that while primary gamma rays are effectively shielded, neutrons are the dominant contributor to the dose outside the vault. Furthermore, the gamma dose measured externally is primarily a result of secondary gamma rays produced from neutron interactions with the shielding and structural materials. The agreement between simulation and measurement demonstrates the reliability of the model for evaluating radiation exposure and designing effective shielding around medical cyclotrons.
Broad energy spectrum of neutrons is produced in medical cyclotrons during Fluorine-18 production. In this work, a detailed three-dimensional (3D) spatial neutronic analysis was conducted within a cyclotron facility in the vicinity of a non-self-shielded PETtrace cyclotron target, supported by a geant4-based Monte Carlo simulation. Moreover, the presence of a shielding column, made of lead (Pb), which has a relatively high neutron scattering cross section, was evaluated experimentally. Multiple sets of gold (Au) and cadmium (Cd) foils were used in three arrangements to evaluate the thermal and epithermal neutron fluxes utilizing neutron activation analysis (NAA). The evaluated total neutron fluxes were found to be on the order of 1 & times; 10+6 neutrons/cm2 s in the forward direction at all investigated positions. The findings indicate that the area around the target exhibits a high intensity of both thermal and epithermal neutron flux. Operational factors such as routine cyclotron maintenance and increased target volume also contribute to a rise in thermal neutron flux. In addition, the Pb column causes a significant increase in thermal neutron flux, with enhancements reaching up to 86%. The obtained results will serve as a reference for future safety assessments, including estimates of radioactive waste generation and worker dose exposure over the operational lifetime of the cyclotron facility.
At nuclear power plants (NPPs), critical structures, systems, and components (SSCs) must comply with national codes and standards. To ensure this compliance, an In-Service Inspection (ISI) program is mandatory. The ISI program relies on Non-Destructive Testing (NDT), which uses various methods and techniques to verify the quality and integrity of SSCs. One of these methods is phased array ultrasonic testing (PAUT), which can sometimes produce many indications that require evaluation. The final assessment of these indications is crucial for the continued safe operation of the NPP. This process can be time-consuming and demanding for NDT personnel, especially during plant shutdowns. To support this task, artificial intelligence (AI) can be used to assist NDT personnel in the evaluation of PAUT data. This article presents an approach for applying machine learning (ML) techniques to train a neural network for this purpose.
This study presents a data-driven approach for predicting nuclear reactor operation state parameters using machine learning algorithms. We developed a multi-input single-output (MISO) framework to model key parameters such as temperature, boron concentration, R-rod position, burnup, and power level, monitored by the digital control system (DCS). Using extensive datasets from a certain nuclear reactor's unit 5/6, we evaluated the performance of K-nearest neighbors (KNN), random forest (RF), and extreme gradient boosting (XGBoost) algorithms. The predictive accuracy was significantly improved through meticulous parameter tuning and the introduction of a comprehensive set of evaluation metrics, including mean absolute error (MAE), mean square error (MSE), and mean relative error (MRE). RF and XGBoost outperformed KNN, indicating the superiority of ensemble models. The models also showed strong generalization capabilities when tested on a separate dataset. The resulting regression model offers an effective tool for accurate prediction of reactor state parameters, enhancing the safety and efficiency of nuclear power plant operations.
Accurate quantification of nuclear material holdup is essential for nuclear safeguards, radiation protection, and criticality safety. The conventional generalized geometry holdup (GGH) method, while widely used with gamma spectroscopy, requires manual classification of deposits into point, line, or area sources-a subjective and error-prone step that hinders automation. To overcome this, the fundamental innovation of this work is a novel algorithmic architecture that fuses image recognition with the GGH framework. The process begins by applying an improved Canny edge detector to thermal maps of the holdup distribution to extract source geometries. An intelligent system then automatically classifies these geometries and matches them to the appropriate correction model from a predefined library, fully automating the workflow and removing human subjectivity. Building upon this core architecture, a secondary yet vital innovation is introduced: an adaptive correction logic for sedimentation thickness. This module addresses the significant bias in self-attenuation factors that arises with thicker deposits. By comparing preliminary results against the theoretical model's asymptotic behavior, the algorithm estimates the effective source thickness and dynamically refines the correction factor. Validation demonstrates that this integrated approach effectively handles diverse source morphologies and thickness variations, consistently achieving a calculation error of less than 10%. The proposed method thus offers a robust, automated solution with strong potential for practical deployment in nuclear facilities.
This study presents a unified nonlinear control design for power regulation during load-following operations in large pressurized water reactors (PWRs) while coordinating axial offset (AO) suppression. Building on the authors' earlier two-zone nonlinear dynamic inversion (NDI) studies (Yadav et al., 2020, "Design of NDI-SMC Based Robust Hybrid Nonlinear Controller for Load Following Operation in Pressurized Water Reactor," Nucl. Eng. Des., 363, p. 110604 and Yadav et al., 2018, "Non-Linear Dynamic Inversion Based Controller Design for Load Following Operations in Pressurized Water Reactors With Bounded Xenon Oscillations," Nucl. Eng. Des., 328, pp. 241-254), a single MIMO NDI-based control law is developed that uses plant-level outputs, namely, total core power (P) and AO, and generates coordinated reactivity commands for the two rod banks. A simple algebraic reconstruction relates (P, AO) to zonal neutron densities, enabling implementation without direct zonal-power measurements. Controller gains are tuned offline to meet the tracking requirements while maintaining bounded reactivity demand. Simulation studies of a two-zone nonlinear PWR model with xenon and thermal-hydraulic feedback demonstrate accurate load-following with bounded xenon-induced oscillations and effective AO regulation. The proposed unified formulation reduces sensitivity to mismatch between independent zonal loops and provides a consistent framework for coordinated power and AO control in large PWR cores. Some of the highlights are as follows: (1) unified NDI-based MIMO controller uses (P, AO) feedback to compute coordinated reactivity commands for two rod banks; (2) two-zone nonlinear xenon and thermal-hydraulic model retained; zonal neutron densities are reconstructed from measurable outputs; (3) simulation results demonstrate accurate load-following with bounded xenon oscillations and effective AO suppression with bounded reactivity demand.
The LVR-15 research reactor is operated as a multipurpose reactor focused primarily on material research. Production of radioisotopes is a well-established activity of the LVR-15 facility, as well as the reactor also represents a stable and reliable producer in the supply chain of medical radioisotopes, especially 99Mo. Different types of radioisotopes have been produced over the years. With the increased demand for radioisotopes, especially for nuclear medicine, and the limited availability of the current reactor fleet, a study was performed to analyze a possible increase and optimization of production capabilities of the LVR-15. The present production possibilities are discussed in the following paper, with their possible expansion involving an irradiation channel mounted at another position in the reactor core marked E7. The study is performed using validated calculation models and methods comprising neutron spectra calculation using the MCNP code and further activity determination with the FISPACT-II code, focusing on an irradiation of samples in a drilled cask. Additionally, heavy water and graphite were considered as a moderator inside the cask for some of the samples. Reasonably achievable activities of the products are determined. Self-shielding effect of the target material size was also discussed.
Predictions of corium behavior are inconsistent and often conservative because severe accident tools frequently use simplified corium thermophysical properties and high-temperature data remain scarce. We present high-temperature density measurements for prototypical oxidic corium produced by cold crucible induction melting, combining in situ melt height determination with ex situ skull and ingot measurements. Independently, scanning electron microscopy (SEM)/energy dispersive spectroscopy (EDS) was used to verify composition and microstructure. Our measurements extend the sparse dataset for multicomponent corium density and reveal systematic differences in the modeling of density and its temperature dependence. At the same time, code comparisons show that updating models is essential for credible predictions of both in-vessel and ex-vessel behavior.
The paper presents a validation of the evaluated fluorine total cross section using a broomstick experiment conducted at the VR-1 reactor. Precise knowledge of fluorine cross sections is crucial for nuclear applications. A collimated neutron beam was directed through PTFE (polytetrafluoroethylene) cylinders of various thicknesses, and the transmitted fast neutron spectra were measured using a stilbene scintillation spectrometer. The setup's sensitivity to total cross sections enables a robust comparison with evaluated nuclear data files. Last updates to the fluorine cross section within the inden collaboration framework, to be adopted by endf/b-viii.1 and jeff-4.0, have shown improvements in criticality benchmarks and integral neutron spectrum measurements using PTFE. The measurements revealed that current standard libraries (endf/b-viii.1 and jeff-3.3) underestimate the transmitted neutron flux by approximately 10-20% in the 1.0-2.0 MeV energy range, indicating an overestimation of the total cross section. The new inden f19f4t4_tot6 evaluation, which incorporates a reduction in the total cross section below 2 MeV, corrects this discrepancy, bringing the calculated transmitted spectra within the 1 sigma experimental uncertainty of the measured data.
Accurate prediction of gamma flux in reactor internals is essential for structural assessments studies. This paper presents gamma spectrum measurements in a VVER-1000 full-scale mock-up at the LR-0 reactor, using a stilbene scintillator connected to the NGA-01 spectrometer. Deconvolution of the gamma spectrum was performed using maximum likelihood estimation with a validated response matrix. Measurements were conducted in the reactor baffle and in front of the reactor pressure vessel (RPV) and compared with mcnp6.2 simulations using endf/b-viii.0 and jeff-3.3 libraries. Good agreement was observed in the baffle region below 5 MeV for endf/b-viii.0. However, both libraries underpredicted the measured spectrum in front of the RPV, particularly at higher energies. These results indicate persisting deficiencies in gamma production data for key reactor materials and emphasize the need for continued validation.
Light water ice is an important material for cold neutron moderator applications or future criticality safety evaluations in low-temperature systems. Depending on temperature and pressure conditions, crystalline and amorphous phases may coexist in the same light water ice. The purpose of this study is to elucidate the differences between crystalline and amorphous phases in light water ice in terms of thermal neutron scattering. To achieve this purpose, molecular dynamics simulations are performed to evaluate velocity autocorrelation function (VACF) of light water ice for the crystalline and amorphous models with scigress, and from the simulation results, thermal neutron scattering cross section data are evaluated with an analysis code called Kyoto University Neutron Scattering Cross section Analysis code (kunsca). The total cross section in the crystalline model shows good agreement with experimental values of light water ice at 115 K, and this suggests that the analysis in the present study is generally valid. The thermal neutron scattering cross sections in the amorphous model show different behavior from that in the crystalline model. In the crystalline model, the inelastic scattering is dominant and higher in the cold neutron energy region (<5 meV) than that in the amorphous model. This suggests that it may be preferable to minimize the formation of amorphous ice when using light water ice as a cold neutron source.
Chloride salts are promising fuel carriers for fast-spectrum molten-salt reactors (MSRs), but current chlorine nuclear data introduce significant uncertainties. To support their validation, critical experiments were performed at the thermal spectrum LR-0 reactor using several chlorine-containing compounds inserted into a benchmark core. Accompanying calculations with endf/b-vii.1, endf/b-viii.0, endf/b-viii.1, jeff-4, and jendl-5 generally underestimate reactivity, in some cases by more than 400 pcm. Sensitivity analysis identifies 35Cl capture, scattering, and (n,p) reactions as the dominant contributors to the observed discrepancies. The results highlight the need for improved chlorine evaluations and provide new integral benchmarks for validating nuclear data used in chloride-salt reactor designs.
The molten fluoride salts used in Molten Salt Reactors (MSRs) are known for being a challenging environment for construction materials. Nickel alloys are usually used in this application for their corrosion and temperature resistance. The goal of this work was to test MoNiCr alloy in a molten mixture of lithium fluoride and beryllium fluoride (FLiBe) and to evaluate the effect of laser shock peening (LSP) and hot isostatic pressing (HIP) methods of thermomechanical treatment on its corrosion resistance. The material was first processed by LSP, and then a HIP heat treatment followed. The aim is to create a modified surface layer through a combination of these processes. The corrosion test itself was performed by immersion of specimens in molten FLiBe at 700 degrees C for 1000 h. The results showed that both LSP and HIP have observable influence on corrosion of MoNiCr alloy. Although the depth of corrosion damage did not differ greatly, the mechanism varied considerably. Regardless of the treatment method, all samples showed chromium depletion to a depth of approximately 30 mu m. A zone of finer grains was formed in the same area. Material processed with LSP alone was more susceptible to intergranular corrosion, while the combination of LSP and HIP suppressed it. The number of passes and type of ablation layer in LSP process have a big role. The best results were achieved with LSP processing with a higher number of repeated passes (10) and without ablation polymer tape. In combination with HIP, intergranular corrosion was completely suppressed.
This study examines the effects of gamma and combined neutron-gamma irradiation on cement paste and mortar, focusing on the micromechanical response of the main hydrated phases and associated microstructural changes. Samples were irradiated under both neutron and gamma conditions. Nanoindentation was performed to assess the mechanical behavior of individual phases. Results show that irradiation induces stiffening of the primary hydrates and causes compaction and densification of C-S-H gels, including partial breakdown of fibrillar structures and reorganization of colloids into clustered, rosette-like morphologies. Irradiation also led to the formation of microcracks within the cement paste. While these microcracks did not affect the micromechanical response of individual phases, they are likely responsible for the reductions in compressive strength observed in cement composites under irradiation. The findings demonstrate that irradiation produces irreversible nanoscale and microscale modifications in cementitious materials, and that local phase stiffening combined with microcrack formation provides a mechanistic explanation for the degradation of bulk mechanical properties.
Reliable electrical power is needed for space exploration and development, including supporting human outposts on the surface of the Moon, Mars, asteroid mining, and other applications, and this power could be provided by micronuclear reactors designed to operate for a decade or more in such remote locations and extreme environments. Based on anticipated near-term needs for a Moon outpost, it is expected that the design and performance targets for a space microreactor will be to provide 100-kWe for 10 years, and to have a total mass (core + reflector + shield) that is below 3000 kg, to accommodate the additional mass of secondary systems (power converter, radiators, and other components), and limitations for launch and Moon landing vehicles. Exploratory scoping studies have been carried out using the serpent reactor physics analysis code for a simplified homogeneous model to evaluate the performance characteristics of a potential thermal-spectrum space microreactor concept with a nominal power level of 500 kWth/100 kWe, using the early system nuclear auxiliary power (SNAP)-8 space reactor design as a guide. It is anticipated that such a space reactor would provide power for human outposts on the Moon or Mars. A reference design concept (50-cm diameter, 100-cm height) has been developed, and permutations and modifications to that concept have been tested, using different combinations of fuel, coolant, moderator, and reflector materials. As a simplifying approximation, the core materials are homogenized for the computational reactor physics modeling with serpent to evaluate impacts of design permutations on the core operation life and total system mass (core + reflector + shield). A reactor concept that closely meets the fuel/core lifetime requirement of 10 years has been developed, although its mass exceeds the target 3000 kg by over 400 kg. This concept is based on modifications to the historical SNAP-8 reactor concept with low-enriched uranium (LEU) fuel. The reference design concept uses U-75Zr fuel, ZrH1.8 moderator, Na coolant, and a BeO reflector. The clad for the fuel and other components is INOR-8/Hastelloy-N, while the outer shield is LiH. Alternative materials for the fuel, moderator, coolant, and reflector were also tested, while adjusting the fuel and moderator volume fractions to ensure a constant H/U-235 = 79 atom ratio. It is anticipated that further optimization studies with different combinations of materials could potentially reduce the core, reflector, and shield sizes to achieve performance objectives.
Sodium-cooled fast small modular reactors (SFR-SMRs) are being proposed for potential deployment in Canada. SFR-SMRs have the capability of recycling all actinide nuclides in a closed fuel cycle, and thus SFR-SMR fuel cycle can significantly increase the energy extraction from natural uranium in comparison with thermal-spectrum reactors. The continuous recycling of plutonium and minor actinides (MAs) can make great contributions to the reduction of long-term radioactive waste liabilities. Given the potential importance of the implementation SFR-SMR technology within Canada as part of a long-term sustainable nuclear energy strategy, it is important and necessary to evaluate the fuel supply and fabrication needs, and the associated issues and options for the implementation of SFR-SMRs in Canada. The objective of this paper is to provide an overview of such an assessment that was carried out recently at Canadian Nuclear Laboratories (CNL). As part of this assessment, a number of different SFR fuel types (from current and historical SFR projects) were reviewed along with their associated fabrication methods and fuel supply options. Cylindrical and metallic fuel slugs for an SFR-SMR can be fabricated using a variety of methods. Injection casting is the most established fabrication method. The different kinds of fuel used in various SFR designs need special facilities to fabricate/produce fuels, and none exist in Canada currently, although expertise in fabrication of fuels relevant to metallic fuel and the mixed oxide (MOX) fuel still exist in CNL. Based on the industrial experience already gained on fast reactors within the international community, and the significant synergies with the light water reactor (LWR) and pressure-tube heavy water reactor (PT-HWR) fuel cycles, the SFR reactor fuel cycle can be considered as being relatively mature, or at least highly developed for oxide and metal-alloy fuel options. Based on an assessment of current and near-term future fuel supply and manufacturing capabilities, technical impediments can be overcome in providing the supply of fuel for SFR-SMR design.
Fertilizer production plays a vital role in supporting agriculture, yet its processes may introduce radiological risks that warrant careful monitoring. In this study, an in situ estimation of radionuclide activity concentrations and background gamma radiation levels was conducted at different sections of an inorganic fertilizer factory in Asaba, Nigeria, using a portable scintillation gamma radiometer. Three radionuclides, 238-U, 232-Th, and 40-K, were detected across all sections, with high mean activity concentrations of 238-U and 40-K in raw material and production areas exceeding world mean values (WMV). The absorbed gamma dose rate in the raw material section (56.98 nGy/h) approached the global limit of 59 nGy/h, while other sections remained below. Radiological hazard indices, including annual gonadal dose equivalent (0.36-1.11 mSv/y) and excess lifetime cancer risk (0.528-1.42), were above maximum permissible limits (MPL), indicating potential long-term health risks for workers and surrounding communities. These results highlight the novelty of direct, section-specific radiological assessment within fertilizer production facilities and underscore the need for continuous monitoring, restricted occupational exposure, and evaluation of finished fertilizer products to safeguard public health and the environment.
The molten stable salt small modular reactor (MSR-SSR-SMR) is a molten chloride salt fast-spectrum reactor. Moltex Energy Canada is developing the SSR-W-300, a 300-MWe-class MSR-SSR-SMR. The MSR-SSR-SMR (referred here on as the "SSR-SMR" for brevity) is of potential interest in Canada because of its potential to use recycled uranium, plutonium, and minor actinides (isotopes of Np, Am, and Cm) obtained from stockpiles of slightly utilized nuclear fuel (SUNF) from pressure-tube heavy water reactors (PT-HWRs), such as CANDU (CANada Deuterium Uranium), and also from light water reactors (LWRs), including both pressurized water reactors (PWRs) and boiling water reactors (BWRs). Thus, SSR-SMRs have the potential to enable closure of the nuclear fuel cycle. This approach can maximize the utilization of fissile and fertile nuclear fuel resources and minimize the mass and volume of high-level radioactive waste to be placed into long-term storage. The objective of this study is to assess synthesis/fabrication needs for fuel for a generic 300-MWe-class SSR-SMR (similar to the Moltex SSR-W-300). The current Canadian nuclear industry and facilities could adapt to supply the fuel that is needed for an SSR-SMR. However, a significant technical and financial investment is anticipated to be required to develop the infrastructure necessary for the reprocessing of SUNF from currently operating reactors and for the production of molten chloride salt fuel and fuel assemblies to contain the stable fuel salt.
