The European Journal of Physics N (EPJ-N)
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Analysis of the preliminary campaign for the PETALE program
No full textThe PETALE program aims to provide new experimental data to constrain the stainless steel nuclear data. In this frame, a preliminary measurement campaign has been performed to characterize the neutron flux in key positions of the CROCUS reactor and to develop analysis tools. During this preliminary campaign detailed in the present paper, an efficiency ratio technique has been developed and tested to speed up HPGe measurements by a factor of 30. A second objective of the campaign concerns the propagation of nuclear data uncertainty from the core neutron cross-sections to the reaction rates in the dosimeters. Uncertainties in the core cross sections, such as the uranium cross section, are nuisance parameters that add uncertainty to the dosimeter reaction rate calculation. This component must be fully characterized with covariances to constrain the metal reflector component for Bayesian assimilation. The experimental results are compared to the calculations with different nuclear databases for the nuclear data uncertainty propagation. A good agreement is obtained with the ENDF/B-VII.1 database and a systematic underestimation of around 5–10% in the fast range is observed with the ENDF/B-VIII.0 and JEFF-3.3 databases
Developments and experiences of the CHANCE, MICADO and PREDIS projects in radioactive waste characterization
No full textCharacterization is a very important step in dealing with materials and waste streams generated during the operational and decommissioning phases of nuclear installations, including nuclear power plants. Characterization allows differentiation between materials that can be released from regulatory control and those that require further treatment and conditioning to become a stable waste form suitable for future storage and final disposal according to its waste classification. Characterization is also needed in the pre-disposal stages of radioactive waste management to demonstrate compliance with the waste acceptance criteria of the facilities that will accept the different waste forms. This work will present the strategies developed and implemented by the three projects for in-depth and accurate waste characterization and investigation of the different radioactive waste packages considered. CHANCE, MICADO, and PREDIS will present their goals, the methods developed, the technologies used and the (preliminary) results contributing to the improvement of the safety and the data and information quality of the waste packages analyzed at the different stages of the waste management process. Special emphasis will also be given to complementary approaches highlighting the usability of the technologies, the accessibility of the data, and the problem-solving of the three projects within the European panorama
On the estimation of nuclide inventory and decay heat: a review from the EURAD European project
No full textIn this work, a study dedicated to the characterization of the neutronics aspect of the Spent Nuclear Fuel (SNF), as part of the European project EURAD (Work Package 8), is presented. Both measured nuclide concentrations from Post Irradiation Examination samples and decay heat from calorimetric measurements are compared to simulations performed by different partners of the project. Based on these detailed studies and data from the published literature, recommendations are proposed with respect to best practices for SNF modelling, as well as biases and uncertainties for a number of important nuclides and the SNF decay heat for a cooling period from 1 to 1000 years. Finally, specific needs are presented for the improvement of current code prediction capabilities
Codes and methods improvements for safety assessment and LTO: varied approaches
No full textNuclear safety has always been at the heart of the concerns of nuclear power plant operators and developers, as well as of various nuclear research organizations and regulatory authorities. Over the last decades, all these nuclear actors have developed and integrated a large number of calculation codes and other tools into their safety work. From the system approach to the local understanding of a phenomenon on a given component, from neutronics to operation optimization for long-term operation, these methods and codes have been constantly evolving since their appearance, in order to be able to integrate new plant designs and components, to improve the results of modeling physical phenomena or quantify and thus reduce the uncertainties on these results. Currently, several H2020 Euratom projects are working on the improvement of these codes and methods. This article will focus on three of these projects: CAMIVVER (Codes And Methods Improvements for VVER comprehensive safety assessment), APAL (Advanced PTS Analysis for LTO), and sCO2-4-NPP (innovative SCO2-based heat removal technology for an increased level of safety of Nuclear Power Plants) in order to illustrate our thinking on the improvement of calculation frameworks. First, we will present the work and the approach adopted with regard to the different calculation codes and methods used in each of these three projects. We will then conclude with an overall analysis of these three approaches, highlighting the difficulties and successes of these three projects, and identifying areas of work for the general improvement of the calculation codes
Multi-group analysis of Minor Actinides transmutation in a Fusion Hybrid Reactor
No full textNew nuclear technologies are currently being study to face High Level Waste treatment and disposal issues. Generally, GEN-IV fission Fast Reactors (FR) are considered the waste-burners of the future. In fact, a fast flux turns out to be the best choice for actinides irradiation in critical reactors because of favorable cross section conditions. Differently, Fusion Fission Hybrid Reactors (FFHR) are futuristic devices based on the combination of fusion and fission systems and could represent an alternative to FRs. In such systems, the choice spectrum of the neutron flux that irradiates HLW may be non-obvious due to some operational constraints which have to be considered. To design and optimize these systems as waste-burners, one should fully understand the transmutation dynamics occurring into the fission region. A multi-energy-group analysis by FISPACT-II code has been set to analyze the conversion processes in scenarios characterized by different neutron energy spectra and fluences. The results of this study show that, despite fast fluxes are characterized by better behaviors in terms of radiotoxicity treatment, the difficulties of reaching high reaction yields may require solutions involving moderators or broadened neutron fluxes to increase the reactions probabilities and, consequently, actinides mass conversion yield
Templates of expected measurement uncertainties for total neutron cross-section observables
No full textThis paper provides a template of expected uncertainties and correlations for measurements of total neutron cross-section observables by transmission. Measurements with time-of-flight and mono-energetic neutron sources are covered. The information required for evaluations in the resonance region and high energy region is detailed, along with the template of uncertainties and correlations that can be used in the absence of other information
SHARE: Stakeholder based analysis of research for decommissioning
No full textThe H2020 EU-funded SHARE project (Stakeholders- based Analysis of REsearch for Decommissioning) is a forerunner to the establishment of a framework for collaboration on research activities related to the decommissioning of nuclear facilities. SHARE aimed to provide an inclusive roadmap for decommissioning research, in both technical and non-technical areas, in the EU and abroad, to enable stakeholders to improve jointly safety, reduce costs and minimise environmental impact. SHARE has been built on a consultation process considering the needs and perspectives of different stakeholders all across the decommissioning value chain. The project also considered existing and emerging innovative technologies solutions as well as the international best practices in the field of decommissioning. After a three-year process, the project provides a Strategic Research Agenda and a Roadmap built on the participation of the international Stakeholder community in a multi-step process including a questionnaire survey, a state-of-the-art review, a gap analysis and multiple workshops. As the final output, the SHARE roadmap effectively set the framework for organizing the priorities identified in the SHARE SRA
Innovation and qualification of LEU research reactor fuels and materials
No full textTwo projects within the Euratom Research and Training Programmes 2014–2018 and 2019–2020 are focused on the innovation and qualification of novel nuclear fuels for conversion from highly-enriched uranium to low-enriched uranium (LEU) and for securing the supply chain of EU research reactors into the future. The LEU-FOREvER project is drawing to a close and has made significant progress in developing and demonstrating the uranium-molybdenum fuel system, demonstrating the viability of a high-density uranium-silicide fuel for EU high-performance research reactors (BR2, RHF, FRM-II, JHR). This project has significantly increased the fabrication know-how and fuel performance understanding of the uranium-molybdenum and high-density uranium-silicide dispersion fuel systems. Further, a new, innovative and increased performance design for the LVR-15 research reactor fuel assembly has been engineered and a demonstration is planned in 2022. In the EU-QUALIFY project, which began in 2020, the planning of four demonstration irradiation tests has been nearly completed and fabrication development of the various fuel systems is ongoing, including the establishment of an EU monolithic uranium-molybdenum fabrication capability. It is expected that the results of this project will begin or complete the data gathering necessary for generic fuel qualification of the LEU uranium-molybdenum dispersion and monolithic fuel systems, and the LEU high-density uranium-silicide fuel system
UMAN – a pluralistic view of uncertainty management
No full textDecisions associated with Radioactive Waste (RW) Management programmes are made in the presence of irreducible and reducible uncertainties. Responsibilities and roles of each actor, the nature of the RW disposal programme and the stage in its implementation influence the preferences of each category of actors in approaching uncertainty management. UMAN (UMAN – Uncertainties Management Multi-Actor Network is a Work Package of the European Radioactive Waste Management Programme – EURAD) carries out a strategic study about the management of uncertainties based on extended exchanges among actors representing Waste Management Organisations, Technical Support Organisations, Research Entities and Civil Society, a review of knowledge generated by past and ongoing R&D projects, and findings of international organisations. UMAN discusses the classification schemes and approaches applied in uncertainty management, and identifies possible actions to be considered in the uncertainty treatment. The relevance for the safety of the uncertainties associated with waste inventory, including spent fuel, near-field, site and geosphere and human aspects, as perceived by each type of actors, and approaches used in their management are explored with the aim to reach either a common understanding on how uncertainties relate to risk and safety and how to deal with them along the programme implementation, or at least arrive at a mutual understanding of each individual view. Finally, uncertainties assessed as highly significant and the associated R&D issues that can be further investigated are being identified
Ambient dose simulation of the ProtherWal proton therapy centre radioactive shielding decay using BDSIM and FISPACT-II
No full textNext-generation proton therapy centres couple treatment and research programs, leading to higher beam currents and longer irradiation times than in clinical conditions. Large fluxes of energetic secondary particles are produced and long- and short-term radioactive nuclides are generated in the concrete shielding of the cyclotron vault. While the overall long-term activation of the centre is well known from the shielding design activation studies, the short-term activation peaks are still of importance when radiation protection studies are involved. The centre shielding design was validated using the BDSIM/FISPACT-II methodology combining particle tracking and Monte Carlo particle-matter interactions simulations using Beam Delivery Simulation (BDSIM) and the computation of the activation using FISPACT-II. We establish, as the next stage of our methodology, the simulation of the decay radiation of the activated concrete shielding and the accurate scoring of the related radiation protection quantities. A single BDSIM simulation per radioactive nuclide is performed based on the nuclide concentration obtained from the prior FISPACT-II activation computations at the start of a given cooling period. The evolution of the radiation protection quantities is obtained by scaling the results with the nuclides activity obtained at later times from fast FISPACT-II computations. We show the evolution of the ambient dose equivalent in the centre vault when considering regular concrete and Low Activation Concrete (LAC) as shielding material to demonstrate the efficiency of LAC mix in mitigating the shielding activation