The European Journal of Physics N (EPJ-N)
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Optimization progress of large-scale radiation shielding Monte Carlo simulation software based on AIS variance reduction technique system: MCShield
No full textThe development of novel nuclear facilities has brought safer and more efficient energy options but also poses significant challenges for radiation shielding calculations. MCShield, developed by the Radiation Protection and Environmental Protection Laboratory at Tsinghua University, is a Monte Carlo program designed for coupled neutron/photon/electron transport in radiation shielding calculations. It incorporates a system of variance reduction techniques based on Auto-Importance Sampling (AIS) to address the deep penetration problem commonly encountered in the field of radiation shielding. The accuracy and computational efficiency of MCShield have been validated through benchmark problems and real-world applications. However, the current AIS system faces limitations in complex scenarios, user-friendliness, and reliance on user experience. To address these issues, we optimized the size, shape, and energy parameters for the AIS variance reduction method, expanded the use of regular virtual surfaces, and introduced an irregular AIS virtual surface method. Additionally, we developed an automatic generation method for AIS virtual surfaces and implemented automatic calculation for these surfaces. AIS Energy Bias Method was proposed to improve convergence across different energy intervals. These improvements enhance the applicability and refinement of the AIS virtual surface parameters, significantly boosting the overall performance of MCShield
Capability overview of the DIANE multiparticle transport code
No full textThe DIANE code is a simulation software that solves the transport equation for neutrons, photons, electrons and light ions using the Monte Carlo method. The DIANE code can perform various kinds of calculations, such as criticality or shielding simulations. This paper presents an overview of the DIANE code capabilities, going through the description of input data, the transport simulation and some examples of applications
Exploring the safety and performance of molten salt reactors for their deployment in the European Union: the MIMOSA and ENDURANCE projects
No full textMolten Salt Reactors (MSR) are Generation IV nuclear systems in which the fuel is dissolved in a molten salt circulating through the primary system. There is growing interest in this advanced technology in Europe, but also in the US, China, South Korea, Japan and Russia, due to their inherently high safety level, flexibility, reliability, load-following capabilities, and potential for multi-recycling of materials contained in light-water reactors’ spent nuclear fuels. These advantages could position MSRs as ideal complements to other decarbonized energy sources in a future sustainable energy mix. In this respect, it is probably one of the most promising advanced technologies and, at the same time, the least mature and studied one. Two ongoing EURATOM-funded projects, MIMOSA and ENDURANCE projects are exploring molten salt reactors’ safety and performance features, as well as fuel cycle aspects, in order to assess and demonstrate their potential for future deployment in Europe. The MIMOSA and ENDURANCE projects have the common objective of improving the maturity of MSR technology. The MIMOSA project develops and analyses multi-recycling strategies for the European Union based on the use of MSR and demonstrates several key aspects of their technical feasibility and performance by both calculations and experimental investigations. The ENDURANCE project supports the safe operation and the development of Critical Technology Elements by connecting design developers and industry with universities and research centres while ensuring alignment with regulatory requirements. Whereas ENDURANCE is in its starting phase, MIMOSA has already delivered important results
European collaborations for safe and efficient dismantling: digital twins, ontology and data exchange
No full textDue to economic considerations and political decisions, an increasing number of nuclear facilities is to be dismantled in the coming decades. The large number of nuclear decommissioning projects must comply with the reliability and safety requirements in order to make the dismantling operations more efficient, safer and more cost-effective.
This paper gives an overview of European coordinated efforts to develop and demonstrate the use of digital tools and methods for safe and efficient decommissioning activities through the projects PLEIADES (PLatform based on Emerging and Interoperable Applications for enhanced Decommissioning processES) and DORADO (Digital twins and Ontology for Robot Assisted Decommissioning Operations).
Achieved by the end of 2023, the PLEIADES project defined a common ontology specifically designed for nuclear decommissioning projects. It developed a central server for combining data while ensuring compatibility and it provided the first pilot integration of digital decommissioning and waste management support tools. PLEIADES demonstrated the usefulness and efficiency of this concept using data of three real nuclear sites.
Starting in the second half of 2024, the DORADO project will continue this work by creating a holistic digital data-driven platform as a BIM/DT (Building-Information-Model/Digital-Twin) and by integrating new digital tools into a coherent suite customized for decommissioning applications. Eight digital technologies will be integrated, including point-cloud data, 3D models and change detection, sensors data fusion, ALARA (As Low as Reasonably Achievable) dose estimation, robot mission optimization, and smart voice assistant interface
Neutronic analysis of accident tolerant fuel concepts in spectral shift regulation conditions
No full textThe study of Accident Tolerant Fuels (ATFs) has gained significant attention due to their potential to enhance the safety and reliability of nuclear reactors, especially under severe accident conditions. This work investigates the neutron-physical performance of ATF concepts in Light Water Reactor (LWR) using spectral shift regulation (SSR). The SSR mechanism, which adjusts the moderator-to-fuel ratio, by means of mechanical displacers that run in the guide tubes, enables spectral hardening during the initial cycle and softening toward the end, optimizing fuel utilization and reactor safety. Using collision probability method with the GETERA simulation tool and BNAB library, this study evaluates various ATF designs, including chromium-coated zirconium (CrZry), FeCrAl alloys, and SiC composites, combined with uranium silicide (U3Si2) and uranium nitride (UN) high density fuels. Key metrics such as infinite multiplication factor, nuclide inventory, and plutonium buildup are analyzed. Results demonstrate that SiC cladding paired with UO2 or U3Si2 fuels offers the best neutronic characteristics, requiring slightly lower uranium enrichment. Thin FeCrAl cladding, on the other hand, shows higher enrichment needs but still around enrichment level of the fuel in Light Water Reactors, when paired with combination with U3Si2. The integration of SSR in the LWR reactor improves burn-up efficiency while burning produced plutonium and lower natural uranium consumption, resulting in reduced waste. Among the ATF options studied, SiC cladding shows good promise for guide tube and displacer applications, due to their neutron transparency and thermal stability
Simulating hydrogen diffusion in a zirconium hydride moderator block and its impact on steady state neutronic-thermal behavior
No full textZirconium hydride is a widely used moderator in compact reactor designs due to its high thermal limits and high hydrogen density, both of which being desirable feature. However, a notable characteristic of zirconium hydride is the substantial mobility of hydrogen within the metal lattice, especially at high temperatures and under large thermal gradients. Variations in hydrogen distribution can significantly affect neutron moderation and, consequently, the reactor’s power profile. This study employs a coupled OpenMC-MOOSE simulation framework to model this complex feedback between hydrogen transport, heat transfer, and neutron behavior. A hypothetical epithermal reactor configuration is analyzed, where zirconium hydride serves as a monolithic moderator with embedded fuel pins and heat pipes. The simulation results illustrate the redistribution of hydrogen and its subsequent impact on both the thermal and neutronic behavior of the system. The magnitudes of the neutronic-thermal impacts vary depending on input reactor power and heat pipe boundary condition; this work found reactivity impacts ranging from 100 to 3500 pcm and some minor spatial impacts to power distribution
ICARUS: development, optimization, and harmonization of innovative characterization techniques for large volumes of radioactive waste
No full textIn the framework of the EURAD-2 partnership, the ICARUS (Innovative ChARacterization techniques for large volUmeS) work package aims at further developing, optimizing and harmonizing innovative techniques for the radiological, physical and chemical characterization of large volume of low/intermediate-level mixed waste, as it could be critical for the safe implementation of radioactive waste management programmes of the member states. Destructive Techniques (DT), Non-Destructive Techniques (NDT), and Scaling Factors (SF) will be investigated. Four use cases will be considered according to the end users’ needs: (i) enabling fast and sufficiently accurate characterization of gamma activity distribution by NDT in complex large packages, including mixed wastes as heterogeneous legacy waste; (ii) improving and simplifying NDT for determining physico-chemical properties; (iii) improving sensitivity, accuracy, and uncertainty and replacing the expensive and time-consuming radiochemical analysis of long-lived Difficult To Measure (DTM) radionuclides by DT; (iv) improving accuracy, uncertainty, and reliability of the SF approach to estimate DTM radionuclides in raw mixed waste. Specific education and training materials and opportunities will be provided to foster career advancement of young professionals and researchers. The research activities will be carried out by 29 organizations from 17 countries
Derivation and verification of the direct-sampling method for simulating Monte Carlo flight paths in tetrahedral meshes with linear finite-element cross sections
No full textThis paper provides a derivation of a direct-sampling approach for modeling continuously varying cross sections in tetrahedral-mesh-based Monte Carlo codes. Specifically, cross sections are spatially approximated using linear nodal finite elements. A linearization strategy is provided for non-linearly varying cross sections. The method is verified against seven analytical pure-absorber test problems. These test problems also highlight the benefit of using linear finite elements over element-wise-constant cross sections
Overview of kinetic Monte Carlo methods used to simulate microstructural evolution of materials under irradiation
No full textKinetic Monte Carlo (KMC) methods are commonly used to simulate the microstructure evolution of metals under irradiation due to their ability to generate the random walks underlying defect-mediated diffusion processes at the atomic scale. However, the range of applicability of KMC methods is severely limited by the kinetic trapping of the simulated trajectories within low energy basins presenting small intra-basin barriers. This results in dramatically reducing the efficiency of the classical KMC algorithm. Kinetic trapping can be alleviated by implementing non-local jumps relying on the theory of absorbing Markov chains. A factorization of an auxiliary absorbing transition matrix then allows to generate escaping paths and first-passage times out of trapping basins. Although the speed-up can be of several orders of magnitudes, this is sometimes not enough for very long-term prediction. We must then turn to homogenized rate-equation formulation of the problem. Usually solved deterministically, the corresponding large ordinary differential equation system often suffers from the curse of dimensionality. Dedicated Monte Carlo schemes can simulate the coarse-grained rate equations based on a chemical master equation. Finally, we show the relevance of relaxing the rigid-lattice assumption in the calculation of the free energy barriers and attempt frequencies to capture elastic effects that are important for certain systems, such as high entropy alloys or other concentrated alloys such as austenitic stainless steels. A new activation-relaxation technique combining barriers and prefactors on-the-fly calculations can be used for this purpose in kinetic Monte Carlo studies of slow diffusion processes
Accelerating split-exponential track length estimator on GPU for Monte Carlo simulations
No full textIn the context of computing 3D volumetric tallies for nuclear applications, the combination of Monte Carlo methods and high-performance computing is essential to achieve accurate yet computationally feasible simulations that meet industrial time constraints. The next-event Split Exponential Track-Length Estimator (seTLE) is particularly well suited for estimating tallies on meshes. To alleviate the computational burden associated with seTLE, such as sampling numerous outgoing pseudoparticles at each collision, estimating cross sections, performing ray tracing through complex geometries, and accumulating scores across the geometry, we leverage the parallel computing capabilities of Graphics Processing Units (GPUs). We assess the performance of our implementation using two shielding configurations and one criticality benchmark. Both photon and neutron transport simulations are considered. Scores are evaluated over Cartesian meshes, material volumes, and energy group structures. In all cases, acceleration factors greater than unity are observed in the detectors, reaching several hundred in selected regions of the phase space. In a final experiment, we demonstrate that our GPU-based implementation achieves a net energy gain (in Watt) even when compared to a conventional CPU-based TLE, despite the additional computational cost of GPU use