The European Journal of Physics N (EPJ-N)
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Modelling of fine fragmentation and fission gas release of UO
No full textIn reactor accidents that involve rapid overheating of oxide fuel, overpressurization of gas-filled bubbles and pores may lead to rupture of these cavities, fine fragmentation of the fuel material, and burst-type release of the cavity gas. Analytical rupture criteria for various types of cavities exist, but application of these criteria requires that microstructural characteristics of the fuel, such as cavity size, shape and number density, are known together with the gas content of the cavities. In this paper, we integrate rupture criteria for two kinds of cavities with models that calculate the aforementioned parameters in UO2 LWR fuel for a given operating history. The models are intended for implementation in engineering type computer programs for thermal-mechanical analyses of LWR fuel rods. Here, they have been implemented in the FRAPCON and FRAPTRAN programs and validated against experiments that simulate LOCA and RIA conditions. The capabilities and shortcomings of the proposed models are discussed in light of selected results from this validation. Calculated results suggest that the extent of fuel fragmentation and transient fission gas release depends strongly on the pre-accident fuel microstructure and fission gas distribution, but also on rapid changes in the external pressure exerted on the fuel pellets during the accident
Quick calculation of damage for ion irradiation: implementation in Iradina and comparisons to SRIM
No full textBinary collision approximation (BCA) calculation allows for two types of damage calculation: full cascade and quick calculations. Full cascade mode describes fully the cascades while in quick calculations, only the trajectory of the ion is followed and effective formulas give an estimation of the damage resulting from each collision of the ion. We implement quick calculation of damage in the Iradina code both for elemental and multi-component solids. Good agreement is obtained with SRIM. We show that quick calculations are unphysical in multi-component systems. The choice between full cascade and quick calculations is discussed. We advise to favour full cascade over quick calculation because it is more grounded physically and applicable to all materials. Quick calculations remain a good option for pure solids in the case of actual quantitative comparisons with neutron irradiations simulations in which damage levels are estimated with the NRT (Norgett-Robinson and Torrens) formulas
Microstructure and mechanical properties relationship of additively manufactured 316L stainless steel by selective laser melting
No full textAdditive manufacturing (AM) is rapidly expanding in many industrial applications because of the versatile possibilities of fast and complex fabrication of added value products. This manufacturing process would significantly reduce manufacturing time and development cost for nuclear components. However, the process leads to materials with complex microstructures, and their structural stability for nuclear application is still uncertain. This study focuses on 316L stainless steel fabricated by selective laser melting (SLM) in the context of nuclear application, and compares with a cold-rolled solution annealed 316L sample. The effect of heat treatment (HT) and hot isostatic pressing (HIP) on the microstructure and mechanical properties is discussed. It was found that after HT, the material microstructure remains mostly unchanged, while the HIP treatment removes the materials porosity, and partially re-crystallises the microstructure. Finally, the tensile tests showed excellent results, satisfying RCC-MR code requirements for all AM materials
3D convolutional and recurrent neural networks for reactor perturbation unfolding and anomaly detection
No full textWith Europe's ageing fleet of nuclear reactors operating closer to their safety limits, the monitoring of such reactors through complex models has become of great interest to maintain a high level of availability and safety. Therefore, we propose an extended Deep Learning framework as part of the CORTEX Horizon 2020 EU project for the unfolding of reactor transfer functions from induced neutron noise sources. The unfolding allows for the identification and localisation of reactor core perturbation sources from neutron detector readings in Pressurised Water Reactors. A 3D Convolutional Neural Network (3D-CNN) and Long Short-Term Memory (LSTM) Recurrent Neural Network (RNN) have been presented, each to study the signals presented in frequency and time domain respectively. The proposed approach achieves state-of-the-art results with the classification of perturbation type in the frequency domain reaching 99.89% accuracy and localisation of the classified perturbation source being regressed to 0.2902 Mean Absolute Error (MAE)
Development of a cold plug valve with fluoride salt
No full textExperimental studies have been developed on a new freeze plug concept for safety valves in facilities using molten salt. They are designed to allow the closure of an upstream circuit by solidifying the molten salt in a section of the device and to passively melt in case of a loss of electric power, thus releasing the upper fluid. The working principle of these cold plug designs relies on the control of the heat transfer balance inside the device, which determines whether the salt inside the cold plug solidifies or melts. The device is mainly composed of steel masses that are dimensioned to provide sufficient thermal heat storage to melt the salt and thus open the cold plug after the electric power is stopped. The final goal of the work is to provide useful recommendations and guidelines for the design of a cold plug for the emergency draining system of a molten salt reactor. Some numerical thermal simulations were performed with ANSYS mechanical (Finite Element Method) to be compared with results of the experiments and to make extrapolations for a new component to be used in a reactor
Simplified 0-D semi-analytical model for fuel draining in molten salt reactors
No full textA key feature of molten salt reactors is the possibility to reconfigure the fuel geometry (actively or passively driven by gravitational forces) in case of accidents. In this regard, the design of reference molten salt reactor of Generation IV International Forum, the MSFR, foresees the Emergency core Draining System (EDS). Therefore, the research and development of MSFRs move in the direction to study and investigate the dynamics of the fuel salt when it is drained in case of accidental situations. In case of emergency, the salt could be drained out from the core, actively or passively triggered by melting of salt plugs, and stored into a draining tank underneath the core. During the draining transient, it is relevant from a safety point of view that thermal and mechanical damages to core internal surfaces and to EDS structure – caused by the temperature increase due to the decay heat – are avoided. In addition, the subcriticality of the fuel salt should be granted during all the draining transients. A simplified zero-dimensional semi-analytical model is developed in this paper to capture the multiphysics interactions, to separate and analyse the different physical phenomena involved and to focus on time evolutions of temperature and system reactivity. Results demonstrate that the fuel draining occurs in safe conditions, both from the thermal (temperature-related internal surface damages) and neutronic (sub-critical states dominate the transient) view points and show which are the main characteristics of the fuel salt draining transient
New composite material based on heavy concrete reinforced by basalt-boron fiber for radioactive waste management
No full textA new composite material with neutron radiation shielding properties is presented. This fiber reinforced concrete material incorporates basalt-boron fiber, with different concentrations of boron oxide in fiber, and is applicable to nuclear energy and nuclear waste management. The methodology for production of boron oxide (B2O3) infused basalt fiber has been developed. First experimental samples of basalt boron fiber containing 6% of B2O3 and 12% B2O3 have been produced in laboratory conditions. The concrete samples reinforced by two types of basalt-boron fiber with different dosages have been prepared for neutron experiment. The neutron experimental investigations on radiation shielding properties of concrete reinforced by basalt-boron fiber have been performed by means of Pu-Be neutron source. The prepared samples have been tested in the course of several series of tests. It is shown that basalt-boron fibers in concrete improve neutron radiation shielding properties for neutrons with different energies, but it appears to be most effective when it comes to thermal neutrons
How to produce accurate inelastic cross sections from an indirect measurement method?
No full textInelastic reactions ((n,xn) for x ≥ 1) play a key role in reactor cores as they influence the slowing down of the neutrons. A reactor neutron energy spectrum depends thus on this process which is in strong competition with elastic scattering and fission; a nice example is the case of 238U. Inelastic scattering (x = 1) impacts keff and radial power distribution in the nuclear reactor. For several years, it has been shown that the knowledge of the inelastic cross sections in nuclear databases is not good enough to accurately simulate reactor cores and a strong demand for new measurements has emerged with very tight objectives (only a few percent) for the uncertainties on the cross section. To bypass the well-known experimental difficulty to detect neutrons, the prompt γ-ray spectroscopy method is a powerful but indirect way to obtain inelastic cross sections. Our collaboration has used this method for more than ten years and have produced a lot of (n,n′γ) cross sections for nuclei from 7Li to 238U. In this article, we will first discuss the issues of the prompt γ-ray spectroscopy regarding the control of all the uncertainties involved in the (n,n′γ) cross section estimation. Secondly, we will focus on the role of theoretical modeling which, in certain cases, is crucial to reach the objectives of a few percent uncertainty on the (n,n′) cross sections
From fission yield measurements to evaluation: status on statistical methodology for the covariance question
No full textStudies on fission yields have a major impact on the characterization and the understanding of the fission process and are mandatory for reactor applications. Fission yield evaluation represents the synthesis of experimental and theoretical knowledge to perform the best estimation of mass, isotopic and isomeric yields. Today, the output of fission yield evaluation is available as a function of isotopic yields. Without the explicitness of evaluation covariance data, mass yield uncertainties are greater than those of isotopic yields. This is in contradiction with experimental knowledge where the abundance of mass yield measurements is dominant. These last years, different covariance matrices have been suggested but the experimental part of those are neglected. The collaboration between the LPSC Grenoble and the CEA Cadarache starts a new program in the field of the evaluation of fission products in addition to the current experimental program at Institut Laue-Langevin. The goal is to define a new methodology of evaluation based on statistical tests to define the different experimental sets in agreement, giving different solutions for different analysis choices. This study deals with the thermal neutron induced fission of 235U. The mix of data is non-unique and this topic will be discussed using the Shannon entropy criterion in the framework of the statistical methodology proposed
On the use of the BMC to resolve Bayesian inference with nuisance parameters
No full textNuclear data are widely used in many research fields. In particular, neutron-induced reaction cross sections play a major role in safety and criticality assessment of nuclear technology for existing power reactors and future nuclear systems as in Generation IV. Because both stochastic and deterministic codes are becoming very efficient and accurate with limited bias, nuclear data remain the main uncertainty sources. A worldwide effort is done to make improvement on nuclear data knowledge thanks to new experiments and new adjustment methods in the evaluation processes. This paper gives an overview of the evaluation processes used for nuclear data at CEA. After giving Bayesian inference and associated methods used in the CONRAD code [P. Archier et al., Nucl. Data Sheets 118, 488 (2014)], a focus on systematic uncertainties will be given. This last can be deal by using marginalization methods during the analysis of differential measurements as well as integral experiments. They have to be taken into account properly in order to give well-estimated uncertainties on adjusted model parameters or multigroup cross sections. In order to give a reference method, a new stochastic approach is presented, enabling marginalization of nuisance parameters (background, normalization...). It can be seen as a validation tool, but also as a general framework that can be used with any given distribution. An analytic example based on a fictitious experiment is presented to show the good ad-equations between the stochastic and deterministic methods. Advantages of such stochastic method are meanwhile moderated by the time required, limiting it's application for large evaluation cases. Faster calculation can be foreseen with nuclear model implemented in the CONRAD code or using bias technique. The paper ends with perspectives about new problematic and time optimization