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In nuclear reactors, tracking the loss and production of neutrons is crucial for the safe operation of such devices. In this regard, the microscopic cross section with the Doppler broadening function is a way to represent the thermal agitation movement in a reactor core. This function usually considers the Maxwell–Boltzmann statistics for the velocity distribution. However, this distribution cannot be applied on every occasion, i.e., in conditions outside the thermal equilibrium. In order to overcome this potential limitation, Kaniadakis entropy has been used over the last seven years to generate generalised nuclear data. This short review article summarises what has been conducted so far and what has to be conducted yet.

The paper describes in brief the functional capabilities of a computer code for optimizing the neutronics model parameters (neutron data, technological parameters, and their covariance matrices) based on results of reactor physics experiments using conditional nonlinear multi-parameter optimization algorithms. The code’s application scope includes adjustment of neutron constants, technological parameters and their covariance matrices based on integral measurement results, formulation of requiremen117198ts with respect to the neutron data uncertainties for achieving the target accuracies in calculation of the reactor functionals, and estimation of the reactor performance prediction accuracy, as well as the informativity and similarity metrics of reactor physics experiments with respect to each other and in relation to the target reactor system. The paper also considers some examples of using the code to refine the neutronics models of nuclear reactor and fuel cycle systems based on results of reactor physics experiments.

In the early 1970s Dr. John T. Mihalczo (team leader), J.J. Lynn, and J.R. Taylor performed experiments at the Oak Ridge Critical Experiments Facility (ORCEF) with highly enriched uranium (HEU) metal (called Oak Ridge Alloy or ORALLOY) to recreate GODIVA I results with greater accuracy than those performed at Los Alamos National Laboratory in the 1950s. The purpose of the Oak Ridge ORALLOY Sphere (ORSphere) experiments was to estimate the unreflected and unmoderated critical mass of an idealized sphere of uranium metal corrected to a density, purity, and enrichment such that it could be compared with the GODIVA I experiments. This critical configuration has been evaluated. Preliminary results were presented at ND2013. Since then, the evaluation was finalized and judged to be an acceptable benchmark experiment for the International Criticality Safety Benchmark Experiment Project (ICSBEP). Additionally, reactor physics measurements were performed to determine surface button worths, central void worth, delayed neutron fraction, prompt neutron decay constant, fission density and neutron importance. These measurements have been evaluated and found to be acceptable experiments and are discussed in full detail in the International Handbook of Evaluated Reactor Physics Benchmark Experiments. The purpose of this paper is to summarize all the evaluated critical and reactor physics measurements evaluations.

The space asymptotic theory has constituted a powerful tool for the determination of neutron energy spectra in nuclear reactors, which are the basis of the generation of group constants for the neutronic core design. The method can provide a deep physical insight into the basics of reactor physics and may still give new ideas for modern computational methods. This contribution presents a re-visitation of the method, illustrating its most important general results, some of which may not be well known. In particular, the criticality theory and the space–energy separability theorem are presented. The validity of such theorem is extended also to the net neutron current. The procedure allows to generalize the Fick’s law with a consistent definition of the energy-dependent diffusion coefficient. Some numerical examples are given in simple multigroup models to illustrate the relevant features of the theory.

A new method of Serpent–OpenFOAM coupling is developed as a multi-physics model for Advanced Lead Fast Reactor Demonstrator. The reactor core is simulated in Serpent, a continuous-energy Monte Carlo reactor physics code for neutronic analysis. A three-dimensional geometry is modeled for the calculation of neutronic parameters of the reactor initial core operation. The calculated parameters are evaluated with a good agreement compared to the available reference. The fuel assembly with the maximum power is pinpointed in order to be used for thermal-hydraulic analysis. A thermal-hydraulic model is developed to perform computational fluid dynamics calculation using OpenFOAM software, with an application of a heat conjugate transfer solver written in C++ language. A symmetric one-six of the fuel assembly with the highest power is considered in order to reduce the time of calculation. A multi-physics approach is adopted to map Serpent and OpenFOAM coupling for neutronic and thermal-hydraulic analysis. Moreover, a procedure is implemented in order to evaluate the convergence while coupling Serpent and OpenFOAM. With the implementation of multi-physics model, the maximum temperature of the fuel and coolant is investigated in order to observe any malfunctions. The results show that the coolant and the fuel temperature limits for the ALFRED’s initial core are well preserved for the fuel assembly with the highest power.

The current situation and possible future developments for nuclear power—including fission and fusion processes—is presented. The fission nuclear power continues to be an essential part of the low-carbon electricity generation in the world for decades to come. There are breakthrough possibilities in the development of new generation nuclear reactors where the life-time of the nuclear waste can be reduced to some hundreds of years instead of the present time-scales of hundred thousand of years. Research on the fourth generation reactors is needed for the realisation of this development. For the fast nuclear reactors, a substantial research and development effort is required in many fields—from material sciences to safety demonstration—to attain the envisaged goals. Fusion provides a long-term vision for an efficient energy production. The fusion option for a nuclear reactor for efficient production of electricity has been set out in a focussed European programme including the international project of ITER after which a fusion electricity DEMO reactor is envisaged.

This paper presents a comparative simulation study of xenon and fission product behavior in a CANDU-600 reactor core operating at nominal full power (100%) and an aging CANDU-600 core limited to 90% power due to end-of-life constraints. The analysis includes not only steady-state operating conditions but also transient phases such as shutdown and start-up, where xenon dynamics significantly impact reactivity management. Using theoretical reactor physics simulations, the study investigates xenon reactivity worth in both reactor conditions, with particular attention to the influence of fission product buildup and depletion. The simulations are supported by detailed evaluations of the reactivity compensation provided by the Liquid Zonal Controllers (LZCs), Adjuster Banks and Mechanical Control Absorbers (MCAs). The results highlight that while the 90% power core experiences reduced xenon production, it exhibits higher sensitivity to xenon-induced reactivity swings, particularly during power maneuvers and recovery from shutdown. The total reactivity worth of xenon, in conjunction with the control mechanisms, is quantified throughout the operational cycle. The findings reveal a tighter reactivity balance and reduced control flexibility in the aged core, emphasizing the increasing operational constraints as the reactor approaches end-of-life. These insights are critical for refining control strategies, ensuring safe operation, and optimizing performance across the full power range and operational scenarios of CANDU reactors.

Thermal neutron detection plays a crucial role in numerous scientific and technical applications such as nuclear reactor physics, particle accelerators, radiotherapy, materials analysis and space exploration. There are several challenges associated with the accurate identification and quantification of thermal neutrons. The present work proposes a detailed characterization of a Timepix3 (TPX3) detector equipped with a Lithium Fluoride ( 6 LiF) converter in order to study its response to thermal neutrons that are identified through the 6 Li(n, α ) 3 H reaction. The TPX3-based test system has been installed at the HOTNES facility in ENEA and the analysis highlighted its excellent performance showing high effectiveness in the identification of neutrons through morphological analysis of tracks produced by alpha and triton particles, after accurate discrimination from the gamma background. With the use of Monte Carlo simulations, it has been demonstrated that the main contribution is due to tritons and its signal can be used effectively in the identification of thermal neutrons obtaining an efficiency of 0.9 % for 25 meV neutrons. This allows the TPX3 to have important applications as an environmental monitor for thermal neutrons. This monitoring system can be simply realized and is easy to manage because of its compact size and its digital acquisition that allows a real-time analysis.

A brief comparison study on fuel cycles of four prospective Indonesian power reactors has been completed. The study oversees 160 MWt NuScale Small Modular Reactor (SMR), 10 MWt Reactor Daya Eksperimental (RDE-10), 40 MWt Pembangkit Listrik dan Uap Panas Industri (PeLUIt-40) and Thorcon’s TMSR-500. Both RDE-10 and PeLUIt-40 were fueled with 17wt% HALEU while the other two were fueled by 4.95wt% LEU, without thorium. The study was based on a set of data generated through Monte Carlo reactor physics simulation using OpenMC. The study found that NuScale SMR is the least 235 U consumer by 0.84 g.MW −1 d −1 while the RDE-10 is the most 235 U consumer by 0.97 g.MW −1 d −1 . The simulation results show that TMSR-500 is the most uranium consumer by 1.41 g.MW −1 d −1 while the uprating of RDE-10 to PeLUIt-40 makes PeLUIt-40 as the least uranium consumer by 1.14 g.MW −1 d −1 . As the least uranium consumer, both RDE-10 and PeLUIt-40 provide the highest attainable discharged fuel burnup of extractable energy of about 146 GWd/MTU. Assuming a one batch depletion scheme, NuScale SMR, RDE-10, PeLUIt-40 TMSR-500 has fuel consumption cycles in 2,210 days, 2140 days, 550 days, and 320 days, respectively.

The CROCUS nuclear reactor is primarily motivated by education, from public visits to advanced teaching at the Bachelor’s, Master’s, and doctoral levels, as well as professional training. It is a uranium-fueled, light-water moderated, zero power reactor with a maximum power of 100 W. As simple as an operational reactor can be, CROCUS effectively supports conceptual understanding through its tangible materiality, offering virtually no distance from the system, no significant risk, and negligible dose rates in its vicinity. Once the basic concepts and systems are grasped, however, a deeper effort is required to fully comprehend the machine and the underlying physics. Various tools are employed to accompany this learning process, including Monte Carlo modeling, operator licensing for PhD students, and tangible scale models – the focus of this contribution. Following the limited use of an earlier 3D-printed model, a sturdier and more manipulation-friendly version was built using LEGO ® bricks. This model reproduces all key features with minimal approximations, helping to visualize otherwise hidden elements, such as the startup source and the expansion vases. We present the educational objectives, the learning methods applied, and the benefits of using CROCUS and its models. The impact of the LEGO ® model was evaluated in two teaching contexts: a large group of 2 nd -year Physics Bachelor students during a single 4-hour experiment, and a smaller group of 1 st -year Nuclear Engineering Master students over a semester-long practical course. Statistical analysis indicates improvement in the Master students’ understanding of CROCUS and reactor functioning.