The European Journal of Physics N (EPJ-N)
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An introduction to Spent Nuclear Fuel decay heat for Light Water Reactors: a review from the NEA WPNCS
No full textThis paper summarized the efforts performed to understand decay heat estimation from existing spent nuclear fuel (SNF), under the auspices of the Working Party on Nuclear Criticality Safety (WPNCS) of the OECD Nuclear Energy Agency. Needs for precise estimations are related to safety, cost, and optimization of SNF handling, storage, and repository. The physical origins of decay heat (a more correct denomination would be decay power) are then introduced, to identify its main contributors (fission products and actinides) and time-dependent evolution. Due to limited absolute prediction capabilities, experimental information is crucial; measurement facilities and methods are then presented, highlighting both their relevance and our need for maintaining the unique current full-scale facility and developing new ones. The third part of this report is dedicated to the computational aspect of the decay heat estimation: calculation methods, codes, and validation. Different approaches and implementations currently exist for these three aspects, directly impacting our capabilities to predict decay heat and to inform decision-makers. Finally, recommendations from the expert community are proposed, potentially guiding future experimental and computational developments. One of the most important outcomes of this work is the consensus among participants on the need to reduce biases and uncertainties for the estimated SNF decay heat. If it is agreed that uncertainties (being one standard deviation) are on average small (less than a few percent), they still substantially impact various applications when one needs to consider up to three standard deviations, thus covering more than 95% of cases. The second main finding is the need of new decay heat measurements and validation for cases corresponding to more modern fuel characteristics: higher initial enrichment, higher average burnup, as well as shorter and longer cooling time. Similar needs exist for fuel types without public experimental data, such as MOX, VVER, or CANDU fuels. A third outcome is related to SNF assemblies for which no direct validation can be performed, representing the vast majority of cases (due to the large number of SNF assemblies currently stored, or too short or too long cooling periods of interest). A few solutions are possible, depending on the application. For the final repository, systematic measurements of quantities related to decay heat can be performed, such as neutron or gamma emission. This would provide indications of the SNF decay heat at the time of encapsulation. For other applications (short- or long-term cooling), the community would benefit from applying consistent and accepted recommendations on calculation methods, for both decay heat and uncertainties. This would improve the understanding of the results and make comparisons easier
Need for precise nuclear structure data for reactor studies
No full textThis paper highlights the strong need for precise nuclear structure and decay data measurements in order to perform high-quality modelling on nuclear reactors and other applications. The context of nuclear data evaluation, as well as the importance of low uncertainty evaluations, will be first presented. The importance of such data for interpreting nuclear data experimental measurements is stressed throughout. To demonstrate this, we will explain how mass and charge-dependent fission yields, decay data (in particular for the purpose of residual heat calculations), and inelastic neutrons scattering cross section rely on nuclear structure and decay information and how new and higher quality in such data can lead to improved accuracy in the precision of evaluated nuclear data
PRATIC: A soluble-boron-free, pressurized water cooled, SMR core benchmark
No full textCurrent nuclear reactor research is actively exploring small modular pressurized water reactors (PWRs), particularly soluble-boron-free (SBF) configurations. SBF designs utilize control rods and gadolinium-poisoned fuel rods to manage reactivity. Additionally, small modular reactors (SMRs) commonly integrate steel reflectors to minimize neutron leakage. However, the compactness of SMRs and the adoption of these technical solutions result in notable fluctuations in neutron flux within the core during normal cycle operation. Hence, comprehensive analysis is essential, especially concerning reactor performance under normal and accident conditions, the reliability of neutronics calculation assumptions, etc. Research into these issues requires reactor core neutronics benchmarks that are consistent with industrial concepts so that the analysis results can be applied to real reactors. In this context, this article introduces PRATIC, a SBF-PWR SMR core neutronics benchmark designed to match the global performances of industrial concepts. The development of PRATIC was conducted using a deterministic calculation scheme coupling the APOLLO2 and CRONOS2 codes. PRATIC features a thermal power of 350 MWth, an equilibrium cycle length of 1.9 years, and an average discharge burnup of about 34 GWd/t, while maintaining controlled power distributions. The article elaborates on the design assumptions for PRATIC, then details the reactor core and its equilibrium cycle. Access to the PRATIC modeling data is available via a GIT repository, accessible upon request via email at [email protected]
Development of a GPU-enhanced time-dependent Monte Carlo neutron transport version of McCARD
No full textTo enhance the performance of the TDMC neutron transport analysis, a new version of McCARD utilizing GPUs, named McCARD/G, is under development. This version introduces an MC neutron tracking module that employs an event-based algorithm and features a generalized lattice geometry treatment module. Additionally, significant adaptations of other major modules have been made to optimize for the GPU architecture. The capabilities of McCARD/G have been verified through the C5G7-TD benchmark, which shows good agreements with the reference results. Furthermore, McCARD/G’s efficacy is also demonstrated via an experimental benchmark conducted at the Kyoto University Critical Assembly
Status of Mercury and Imp: Two Monte Carlo Transport Codes Developed Using Shared Infrastructure at Lawrence Livermore National Laboratory
No full textThe Monte Carlo Transport Project at Lawrence Livermore National Laboratory develops two Monte Carlo transport codes used in production by a sizable internal user community. Mercury is a Monte Carlo particle transport code used to model the interaction of neutrons, gammas, and light ions with a material. Imp is an implicit Monte Carlo thermal x-ray photon transport code used to model the interaction of x-ray photons with a material. This paper describes the two codes and highlights recent developments
European partnership on radioactive waste management
No full textThe European Commission supports a co-funded European Partnership on radioactive waste management within the EURATOM Work Programme for 2023-2025. This initiative, known as EURAD-2, aims to continue and merge the efforts of the EURAD Programme and the PREDIS project, building on their successes and lessons learned. EURAD-2 addresses the European Union’s 3.5 million m3 radioactive waste inventory with a holistic approach, targeting the whole waste management chain from cradle to grave, up until its final disposal in surface, shallow, and deep geological facilities. As a co-funded Partnership, it supports Member States in meeting the Research, Development and Demonstration requirements of the Waste Directive 2011/70. The programme includes diverse participation, with 51 beneficiary organisations and 69 Affiliated Entities from 21 Member States, funded at 60% by the European Commission and 40% by the participating Member States. Furthermore, it welcomes participation from associated countries and interacts proactively with international organisations such as NEA and IAEA.
The EURAD-2 work plan addresses strategic aspects identified in the updated EURAD Strategic Research and Knowledge Management Agenda. To this end, the Partnership employs different instruments (different types of work packages): research and development – performing cutting-edge science and technology research and innovation; Strategic Studies – shorter-term transversal collaborative actions bringing together relevant actors and addressing emerging needs; Knowledge Management – supporting the transfer of knowledge between different programmes and between generations.
EURAD-2 aspires to contribute as one of the leading European platforms for radioactive waste management competence, know-how, and capabilities, supporting scientific excellence and driving innovation in research and technology for its end users. It aims to be a central hub for training new experts and facilitating high-visibility position papers on emerging topics, further establishing the European Union as the forerunner in safe, long-term radioactive waste management
Overview of the new capabilities in the Monte-Carlo particle-transport code NECP-MCX V2.0
No full textNECP-MCX is a Monte-Carlo particle-transport code developed by the Nuclear Engineering Computational Physics (NECP) Lab. of Xi’an Jiaotong University in 2018. The first version of NECP-MCX is focused on addressing the challenge of deep-penetration radiation-shielding problems. In recent years, new capabilities of unstructured-mesh geometries, dose engine for Boron Neutron Capture Therapy (BNCT), neutron noise analysis, sensitivity analysis, depletion and activation calculation, simulation of randomly dispersed media, homogenization and gamma dose calculation based on point-kernel integral method are implemented in NECP-MCX V2.0. A user interface of NECP-MCX V2.0 is also under active development. This paper will introduce these new capabilities
The Reactor Monte Carlo code RMC: The state-of-the-art technologies, advancements, applications, and next
No full textBased on academic research and industrial applications over more than 20 years, the Reactor Monte Carlo code (RMC) developed by the REAL (Reactor Engineering Analysis Laboratory) team at Tsinghua University since 2000 has become a powerful, innovative, and versatile simulation platform for nuclear reactor analysis, shielding simulations, criticality safety calculations, fusion neutronics analysis and beyond. Utilizing collaborative and agile development technology, advanced methods and the most cutting-edge algorithms can be tested and implemented in RMC quickly and efficiently. RMC has been deployed on many world-class supercomputers in China and played an irreplaceable role in the design and analysis of commercial nuclear power plants and newly designed types of advanced nuclear reactors. This paper reviews the state-of-the-art technologies developed in RMC in recent years, such as stochastic and continuous-varying media modeling, advanced transient simulation capability, more accurate energy deposition model, etc. Parallel acceleration on heterogeneous architecture supercomputers and machine learning algorithms would be incorporated in ongoing research and future development plans
Language and design evolution of the OpenMC Monte Carlo particle transport code
No full textThe OpenMC Monte Carlo particle transport code has been continuously developed for 13 years by a large community of contributors. In that time span, the codebase has undergone significant changes that have redefined what OpenMC is and made it an enduring presence in the nuclear science and engineering community. In this paper, we discuss the evolution of programming language use in OpenMC, trends in the overall design of the programming interfaces, and implications for the future of the code