Post-irradiation examination of the Molten Salt Reactor Experiment from the 1970s revealed intergranular cracking of the salt-facing material, Hastelloy-N, from the penetration of fission products, specifically tellurium (Te), into the components. Stainless steel 316H is a candidate salt-facing structural material for future molten salt reactors due to its excellent corrosion, oxidation, and neutron irradiation resistance. Thus, studies are needed to verify if Te may lead to material degradation of salt-facing components made from 316H. This work examined the behavior of stainless steel 316H in three conditions: as received, heat treated to 800°C for 100 h without Te, and with a highly concentrated Te environment. After exposure, mechanical testing was performed on all samples to reveal the loss of strength and ductility in the Te-exposed samples. Additional analysis of the Te-exposed 316H samples using scanning electron microscopy displayed intergranular embrittlement and energy-dispersive X-ray spectroscopy maps highlighted the infiltration of Te within grain boundary cracks. These results present the need for additional experiments to understand how Te weakens the structural material, especially in molten salt, and to eventually identify the driving mechanism for this observed behavior.

In high-dose radiological environments, where precision and safety are of utmost importance, the ability to acquire accurate and clear visual information is of paramount importance for ensuring safety and reliability in critical industrial processes. However, these environments inherently introduce significant challenges due to the adverse effects of radiation on imaging equipment. As a consequence, inspection videos captured within such high-radiation environments often contain a significant amount of noise. This noise substantially complicates the task of identifying and assessing potential defects in vital components. It also diverts attention and resources toward investigating false positives created by noise, leading to inefficiencies, and for industrial processes on the critical path, this can further prolong the outage. Addressing this noise is essential not only for precision, but also for ensuring safety, reliability, and efficiency in critical industrial processes. In this paper, we present a custom-designed filter utilizing a priori information about camera position and trajectory to remove the noise from the inspection videos, making the defects easier to manually identify. As the camera movement is in one direction at a constant speed, the proposed approach uses this temporal and spatial information to accurately remove the noise. This approach applies to a subset of visual inspection problems throughout the nuclear industry, as well as many other industries where there is knowledge available about the camera speed and direction of travel. The proposed approach is compared with three accepted/well-known approaches, median filtering, bilateral filtering, and fast nonlocal means denoising, and an additional state-of-the-art deep learning model is also used for comparison. It was found that the proposed approach produces the most accurate video denoising in terms of visual quality and the retainment of the defect features throughout the videos tested.

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.

With an interest in reducing carbon emissions, Canada has committed to deploying a variety of small modular reactor (SMR) designs. This review covers Canada’s research reactor capabilities relevant to SMR development and highlights gaps in these capabilities. Following the commissioning of the subcritical facility at Ontario Tech University, Canada’s capabilities in reactor physics, education, and training will be relatively sufficient. In contrast, materials/fuel irradiation capabilities and neutron scattering techniques are insufficient, with a notable gap in in-reactor test loops needed to develop new fuels. Recommendations are made for a new multipurpose research reactor (MPRR) development and deployment of SMRs in the long term while increasing accessibility to Canada’s existing research reactors in the near term, particularly in Western Canada.

This paper describes the development of a new thermal-hydraulic (TH) module for the Fast REactor NEutronics/Thermal-hydraulICs (FRENETIC) multiphysics code for the full-core simulation of liquid metal–cooled fast reactors, developed at Politecnico di Torino. The code performs steady-state and transient neutronic (NE) and TH coupled calculations while maintaining a relatively low computational cost thanks to the adoption of simplified physical models. The NE module implements the nodal formulation of the multigroup neutron diffusion equations with delayed neutron precursors whereas the TH module treats the reactor hexagonal assemblies as separate channels, which are individually modeled as one-dimensional in the axial direction, accounting for the thermal coupling in the horizontal direction. The new TH module is more robust and portable while providing improved performance with respect to the previous implementation—also thanks to the adopted OpenMP parallelization. Some physical aspects that were previously neglected, such as the thermal inertia of nonfuel rods, have also been included in the model. The development was carried out in accordance with current best practices for code design, implementation, and testing, thus rendering the code easier to be maintained and possibly to be extended in the future. The code usability has also been improved by means of a set of Python classes purposely developed to simplify the input generation and postprocessing phases. This can potentially widen the usage of FRENETIC within the fast reactor community for the simulation of full-core coupled NE-TH transients and/or as a platform to test new solution methods. The paper also includes the application of this new FRENETIC version to a representative configuration of the Advanced Lead-cooled Fast Reactor European Demonstrator (ALFRED) core design. First, the nominal operating conditions of the ALFRED reactor were simulated, proving that the new upgrade of the code is faster and more robust with respect to the previous one. Then, two accidental transients where NE-TH feedback effects are relevant—an unprotected loss of flow and a total instantaneous blockage of a single fuel assembly—were considered.

Until a long-term solution for the disposal of spent nuclear fuel (SNF) is available, interim dry casks will be increasingly used for the storage of SNF discharged from civilian nuclear power reactors. Dry casks containing commercial SNF may hold several significant quantities of plutonium, so appropriate nuclear material safeguards monitoring is needed. An external remote monitoring system (RMS) has been developed to advance dry cask safeguards monitoring beyond the current method of containment and surveillance used to maintain continuity of knowledge. In this study, neutron transport simulations of SNF assemblies in a dry cask were performed for several special nuclear material diversion scenarios. The simulations considered various loading patterns and fuel storage durations as long as 100 years. For each fuel loading pattern and storage time investigated, the simulation results were used to calculate the required measurement time to achieve a nondetection probability β ≤ 10% for the diversion of any single fuel assembly in the cask. The calculations were performed for false alarm probabilities α as low as 0.0001% (or 10−6). A Monte Carlo postprocessing approach was developed to consider the impact on the required measurement time of uncertainty in the burnup of fuel assemblies. The study found that the external RMS is well suited for the surveillance of SNF in dry cask storage for nuclear safeguards or other purposes and is able to detect the diversion of a single SNF assembly even after decades of storage and with a very low false alarm probability.

Several advanced reactor designs are now under active development in the United States and elsewhere, promising sustainable solutions to the growing world energy needs. The designs currently being considered are quite diverse and different from the more established light water reactor technology that has dominated the operating commercial nuclear landscape. While advanced reactor concepts were first explored in the dawn of the nuclear age, they are now being reconsidered under the light of modern needs, and specifically, for their flexible operating conditions and inherent safety characteristics. In response, the U.S. Nuclear Regulatory Commission staff is moving forward with development of 10 CFR Part 53 rulemaking, which is a more risk-informed, technology-agnostic framework for licensing and regulating such new designs. The nuclear industry response to this regulatory initiative resulted in the technical report by the Nuclear Energy Institute, NEI 18-04 Revision 1, which provides an implementation roadmap of the risk-informed approach when defining the safety case for a new plant design. The implementation of this safety case may be a nontrivial exercise for an actual reactor design. This paper provides a demonstration of performing such an analysis for a representative advanced reactor. Public information from the General Atomics high-temperature gas reactor design was considered in this demonstration. The analysis workflow was facilitated with the FPoliSolutions’ proprietary Risk-Informed System Engineering (RISE) digital platform, a product that was presented in previous publications. RISE is one application of FPoli’s enterprise digital platform, which was created to facilitate orchestration of complex workflows leveraging recent technologies developed at national laboratories, such as Idaho National Laboratory’s RAVEN and EMRALD frameworks. The analysis described in the paper includes the selection and classifications of events, the integration of probabilistic risk analysis artifacts, and event modeling simulations for consequence evaluations. The results are then used for system, structures, and components safety classification and a synthesis of the safety case for the design in line with the frequency-consequence targets presented in NEI 18-04. The purpose of the analysis, as framed in RISE, is to readily produce outputs and views that can aid users and regulators in making risk-informed decisions to demonstrate their plant safety case.

Differing from land-based systems, conventional passive residual heat removal systems (PRHRSs) in nuclear-powered ships encounter challenges for use as safety systems due to limited space and difficulties in adopting safety aids. To resolve this problem, the authors have previously suggested a compact and long-term-operable PRHRS that uses air cooling and seawater cooling in tandem, and its feasibility was confirmed through a design study and transient analysis. Nevertheless, the system performance depends on various parameters, and their sensitivities have yet to be analyzed. The purpose of the current study is to conduct a parametric analysis of the system focusing on four parameters: filling ratio, thermal center difference between heat source and sink, heeling angle, and heat source temperature. The Multi-dimensional Analysis of Reactor Safety (MARS-KS) code was used for the parametric analysis. The results show that the parameters are highly interrelated with each other, and consequently, the system exhibits significantly different behavior depending on the parameter conditions. In particular, the filling ratio determines the overall trends of the system. It was also found that the presence of two heat sinks with a single heat source in the natural circulation loop of the system complicate the natural circulation behavior. Based on the results, we conclude that a thorough examination and careful control of the parameters are necessary to implement the proposed system. The highlights are as follows: 1. Parametric analysis of compact and long-term-operable PRHRS is conducted. 2. Analysis reveals that the filling ratio is the most dominant parameter for system performance. 3. The unique behavior of the natural circulation loop with two heat sinks and a single heat source is analyzed.

Research regarding small modular reactors (SMRs) is increasingly receiving attention due to the advanced stage of development of several SMR concepts. Especially, integral pressurized water reactors (iPWRs) are a promising technology for nuclear new builds. Due to this fact, the applicability of system codes to simulate phenomena in SMRs should be investigated in more detail. In order to ensure that AC2-ATHLET can adequately simulate SMRs and their most important phenomena, an input deck to simulate the International Collaborative Standard Problem test SP-3 at the OSU-MASLWR (Oregon State University–Multi Application Small Light Water Reactor) was further developed and analyzed within a national research project. Since numerous iPWRs use helically coiled steam generators (HCSGs) for heat removal from the primary side, the focus is on HCSG behavior. In comparison to straight tubes or U-tubes, centrifugal force affects the fluid in the pipes and forms a secondary flow normal to the main flow direction. This secondary flow influences heat transfer and is intended to increase it due to increased mixing. Therefore, correlations for calculating the heat transfer coefficients in HCSGs were determined, implemented in an in-house version, and tested as part of a national project. Now, these are available as a first option in the current release of version AC2 2023.0. In addition to the new correlations, an option has also been provided to declare Reynolds-dependent pressure losses for specific components that are important for mass flow rate calculations. Comparative analyses are being carried out between AC2 2021.0 and 2023.0 in order to identify deviations and improvements. Both AC2 versions show good agreement with the experimental data, and there are only slight deviations between the versions. In addition to the HCSG phenomena, there are also deviations in the calculation of the mass flow rate on the primary side. In principle, AC2 enables the adequate simulation of phenomena that are relevant for iPWRs.

Nuclear energy and renewables, both being low-carbon energy sources that are likely to play an increasingly larger role in energy systems of the future, are increasingly being considered from an integrated standpoint. However, the deployment of baseload nuclear reactors as part of such integrated systems may present some challenges in long-term planning, such as surplus energy generation, inflexibility, and increased energy storage requirements. On the other hand, flexible advanced nuclear reactors can be utilized to tackle the limitations of baseload reactors, such as inflexibility of load-following and high initial capital cost. This paper aims to investigate from techno-economic aspects whether energy modelers should use a flexible nuclear reactor model, specifically a small modular pressurized water reactor technology, or a baseload reactor model of comparable size in long-term integrated nuclear renewable (NR) integrated energy system planning simulations. We mathematically develop an off-grid NR integrated system in a MATLAB environment. An advanced small modular reactor is incorporated in this study and is operated in a baseload and flexible mode of operation for comparative analysis. Two metaheuristic optimization algorithms, pelican optimization algorithm and particle swarm optimization, are employed to obtain and validate the optimal configurations of two different NR systems (e.g. baseload reactor system and flexible reactor system). A sensitivity analysis is conducted to reinforce the key research findings. The results indicate that a flexible nuclear reactor reduces the total annualized cost by a very small amount (roughly 2%) and the energy storage sizing by around 3% for NR integrated system planning, compared to a baseload reactor. This study provides insights into the operational assumptions (e.g. baseload or flexible operation) to consider during the modeling of a long-term planning problem of for NR integrated systems.