The European Journal of Physics N (EPJ-N)
Not a member yet
448 research outputs found
Sort by
CONRAD – a code for nuclear data modeling and evaluation
No full textThe CONRAD code is an object-oriented software tool developed at CEA since 2005. It aims at providing nuclear reaction model calculations, data assimilation procedures based on Bayesian inference and a proper framework to treat all uncertainties involved in the nuclear data evaluation process: experimental uncertainties (statistical and systematic) as well as model parameter uncertainties. This paper will present the status of CONRAD-V1 developments concerning the theoretical and evaluation aspects. Each development is illustrated with examples and calculations were validated by comparison with existing codes (SAMMY, REFIT, ECIS, TALYS) or by comparison with experiment. At the end of this paper, a general perspective for CONRAD (concerning the evaluation and theoretical modules) and actual developments will be presented
Muon radiography to visualise individual fuel rods in sealed casks
No full textCosmic-ray muons can be used for the non-destructive imaging of spent nuclear fuel in sealed dry storage casks. The scattering data of the muons after traversing provides information on the thereby penetrated materials. Based on these properties, we investigate and discuss the theoretical feasibility of detecting single missing fuel rods in a sealed cask for the first time. We perform simulations of a vertically standing generic cask model loaded with fuel assemblies from a pressurized water reactor and muon detectors placed above and below the cask. By analysing the scattering angles and applying a significance ratio based on the Kolmogorov-Smirnov test statistic we conclude that missing rods can be reliably identified in a reasonable measuring time period depending on their position in the assembly and cask, and on the angular acceptance criterion of the primary, incoming muons
Enrichment dynamics for advanced reactor HALEU support
No full textTransitioning to High Assay Low Enriched Uranium-fueled reactors will alter the material requirements of the current nuclear fuel cycle, in terms of the mass of enriched uranium and Separative Work Unit capacity. This work simulates multiple fuel cycle scenarios using Cyclus to compare how the type of the advanced reactor deployed and the energy growth demand affect the material requirements of the transition to High Assay Low Enriched Uranium-fueled reactors. Fuel cycle scenarios considered include the current fleet of Light Water Reactors in the U.S. as well as a no-growth and a 1% growth transition to either the Ultra Safe Nuclear Corporation Micro Modular Reactor or the X-energy Xe-100 reactor from the current fleet of U.S. Light Water Reactors. This work explored parameters of interest including the number of advanced reactors deployed, the mass of enriched uranium sent to the reactors, and the Separative Work Unit capacity required to enrich natural uranium for the reactors. Deploying Micro Modular Reactors requires a higher average mass and Separative Work Unit capacity than deploying Xe-100 reactors, and a lower enriched uranium mass and a higher Separative Work Unity capacity than required to fuel Light Water Reactors before the transition. Fueling Xe-100 reactors requires less enriched uranium and Separative Work Unit capacity than fueling Light Water Reactors before the transition
Analysis for the ARIANE GU3 sample: nuclide inventory and decay heat
No full textThis study presents an analysis of the ARIANE GU3 sample, in terms of nuclide inventory, as well as sample rod and assembly decay heat. The validation of a number of CASMO5 and library versions are performed with regards to the measured nuclide inventory, taking into account two dimensional lattice simulations. Uncertainties due to various sources (nuclear data, operating conditions and manufacturing tolerances) are also provided, and are combined with biases into expanded uncertainties. This study is similar to a previous one on the GU1 sample and fit in the framework of code validation, as well as in the estimation of code predictive power for spent fuel characterization
France–Japan synthesis concept on sodium-cooled fast reactor review of a joint collaborative work
No full textIn the frame of the France-Japan agreement on nuclear collaboration, a bilateral collaboration agreement on nuclear energy was signed on March 21st, 2017, including a topic dedicated to Sodium-cooled Fast Reactor (SFR). This agreement has set the framework to start a bilateral discussion on a joint view of an SFR concept. France (CEA and FRAMATOME) and Japan (JAEA, MHI and MFBR) have carried out these studies from 2017 to 2019. Based on the beginning of the basic design phase of ASTRID project − ASTRID − 600 MWe (ASTRID for Advanced Sodium Technological Reactor for Industrial Demonstration), the two countries performed a common work to examine ways to develop a feasible common design concept, which could be realized both in France and in Japan. The subject was then extended and extrapolated with the ASTRID − 150 MWe data (reduced power reactor and enhanced experimental capabilities) in a second phase of this study. France and Japan first focused on design requirements. Common requirements were identified, as well as differences in the safety approach and the structural design requirements, according to national standards and respective site conditions, in particular considering seismic hazards. The teams developed common Top-Level design Requirements (TLRs) to allow common specification data, then joint design. This collaborative work was carried out through the implementation of twelve France-Japan Working Groups, working jointly. This paper is providing a review of this joint synthesis on Sodium Fast Reactor design concept. It is summarizing the context and objectives, then the definition and approaches of the Top Level Requirements. This paper is then dealing with the major design features: the core design and their related safety aspects, and the nuclear island design. Thus, this paper is providing a comprehensive review of this joint work gathering French and Japan nuclear design teams during two full years
Impact of disruption between options of plutonium multi-recycling in PWRs and in SFRs
No full textDuring the recent ten years, the estimation of future uranium demands has changed greatly, and SFR competitiveness is called again into question. In this context, a planning of plutonium multi-recycling in PWRs for the near-term decades has been announced in France, which replaces the objective of future SFR deployment. However, the mid-term policy concerning the future reactor system is always uncertain, and the demand of SFR deployment may re-increase significantly. This study looks into this possibility and analyzes the consequences of such back and forth between different plutonium multi-recycling strategies. The newly developed methodology of robustness assessment is applied to the problem, considering the objective disruptions to take into account the deep uncertainties about nuclear future. Two prior trajectories of plutonium multi-recycling, one involving the use of MIX fuel in PWRs and the other considering the SFR deployment, are analyzed first. The disruption of the strategy using MIX is then supposed under the re-consideration of future SFR deployment. To quantify the impacts of using MIX on deployment timing, we investigate the earliest time for which the fleet substitution with SFRs can be completed. To supplement, the prior strategy of SFR deployment is also disrupted under the context of halting the start of new SFR. The plutonium multi-recycling in PWRs, regarded as adaptive strategy, aims to minimize the idle plutonium. In these robustness assessments, numerous outputs of interests are analyzed, in order to provide a comprehensive evaluation of consequences of prior strategies, regarding the uncertain disruptions and optimal readjustments
Prediction of crack propagation kinetics through multipoint stochastic simulations of microscopic fields
No full textPrediction of crack propagation kinetics in the components of nuclear plant primary circuits undergoing Stress Corrosion Cracking (SCC) can be improved by a refinement of the SCC models. One of the steps in the estimation of the time to rupture is the crack propagation criterion. Current models make use of macroscopic measures (e.g. stress, strain) obtained for instance using the Finite Element Method. To go down to the microscopic scale and use local measures, a two-step approach is proposed. First, synthetic microstructures representing the material under specific loadings are simulated, and their quality is validated using statistical measures. Second, the shortest path to rupture in terms of propagation time is computed, and the distribution of those synthetic times to rupture is compared with the time to rupture estimated only from macroscopic values. The first step is realized with the cross-correlation-based simulation (CCSIM), a multipoint simulation algorithm that produces synthetic stochastic fields from a training field. The Earth Mover’s Distance is the metric which allows to assess the quality of the realizations. The computation of shortest paths is realized using Dijkstra’s algorithm. This approach allows to obtain a refinement in the prediction of the kinetics of crack propagation compared to the macroscopic approach. An influence of the loading conditions on the distribution of the computed synthetic times to rupture was observed, which could be reduced through a more robust use of the CCSIM
A simplified analysis of the Chernobyl accident
No full textWe show with simplified numerical models, that for the kind of RBMK operated in Chernobyl:
The core was unstable due to its large size and to its weak power counter-reaction coefficient, so that the power of the reactor was not easy to control even with an automatic system
A hybrid approach for neutronics calculations in the neutron shielding of sodium fast reactors
No full textNeutron calculations in the neutron shielding of fast neutron reactors are a complex problem as deterministic schemes are usually not suited for such calculations while Monte-Carlo codes have poor computational performance due to the very low flux levels in neutron shields. In this article, both methods are studied, as well as a hybrid scheme on the neutron shielding of the ASTRID fast reactor benchmark. This hybrid scheme uses a fission source calculated by a deterministic code in order to precisely calculate neutron fluxes in the shielding with a Monte-Carlo code using variance reduction techniques. This provides reference results in order to validate deterministic calculations. Comparisons between deterministic codes and this hybrid reference show that large biases are obtained, up to 50%. Further studies are made to reduce the biases, showing that many physical phenomena should be treated, including anisotropy of the scattering law at high energies and spatial self-shielding inside the boron carbide shielding. These improvements reduce the biases to less than 10%. Finally, some applications to designing criteria for the neutron shielding are presented, such as gas production in the neutron shielding and activation of secondary sodium at the intermediate heat exchanger (IHX)
Application of sensitivity analysis in DYMOND/Dakota to fuel cycle transition scenarios
No full textThe ability to perform sensitivity analysis has been enabled for the nuclear fuel cycle simulator DYMOND through its coupling with the design and analysis toolkit Dakota. To test and demonstrate these new capabilities, a transition scenario and multi-parameter study were devised. The transition scenario represents a partial transition from the US nuclear fleet to a closed fuel cycle with small modular LWRs and fast reactors fueled by reprocessed used nuclear fuel. Four uncertain parameters in this transition were studied – start date of reprocessing, total reprocessing capacity, the nuclear energy demand growth, and the rate at which the fast reactors are deployed – with respect to their impact on four response metrics. The responses – total natural uranium consumed, maximum annual enrichment capacity required, total disposed mass, and total cost of the nuclear fuel cycle – were chosen based on measures known to be of interest in transition scenarios [