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New composite material based on heavy concrete reinforced by basalt-boron fiber for radioactive waste management
A 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.
6/18/2020 5:48:13 AM +00:00
Innovative technologies in training and education for maintenance team of NPPs
Many industries, such as nuclear power plants, chemical industry, oil and gas industry have dangerous working environments and hazardous conditions for employees. Maintenance, inspection and decommissioning activities in these safety-critical areas mean a serious risk, downtime is a significant financial loss.
6/18/2020 5:48:07 AM +00:00
3D convolutional and recurrent neural networks for reactor perturbation unfolding and anomaly detection
With 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.
6/18/2020 5:48:01 AM +00:00
Stability studies of GANEX system under different irradiation conditions
In this paper, is addressed the stability of one of most promising extractants (TODGA, N,N,N0,N0-tetraoctyldiglycolamide) but also the importance of designing realistic model to simulate and study the degradation of the systems.
6/18/2020 5:47:55 AM +00:00
Modelling of fine fragmentation and fission gas release of UO2 fuel in accident conditions
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.
6/18/2020 5:47:48 AM +00:00
Application of multiphysics model order reduction to doppler/neutronic feedback
In this paper, a proper orthogonal decomposition based reduced-order model is presented for parametrized multiphysics computations. Our application physics is Doppler feedback in a simplified model of the molten salt fast reactor concept.
6/18/2020 5:47:42 AM +00:00
Application of the lines of defence method to the molten salt fast reactor in the framework of the SAMOFAR project
This paper presents the main characteristics of the method, along with some practical guidelines to apply it to the specific case of the MSFR; moreover, some initiating events are analyzed through the implementation of the LoD tool. The outcomes of this analysis drive the design evolution.
6/18/2020 5:47:36 AM +00:00
Simplified 0-D semi-analytical model for fuel draining in molten salt reactors
A 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).
6/18/2020 5:47:30 AM +00:00
Feedback from experimental isotopic compositions of used nuclear fuels on neutron cross sections and cumulative fission yields of the JEFF-3.1.1 library by using integral data assimilation
Comparisons of calculated and experimental isotopic compositions of used nuclear fuels can provide valuable information on the quality of nuclear data involved in neutronic calculations.
6/18/2020 5:47:23 AM +00:00
Development of a control-oriented power plant simulator for the molten salt fast reactor
In this paper, modelling and simulation of a control-oriented plant-dynamics tool for the molten salt fast reactor (MSFR) is presented. The objective was to develop a simulation tool aimed at investigating the plant response to standard control transients, in order to support the system design finalization and the definition of control strategies.
6/18/2020 5:47:16 AM +00:00
Influence of molten salt-(FLiNaK) thermophysical properties on a heated tube using CFD RANS turbulence modeling of an experimental testbed
The goal of this study is to demonstrate and quantify the impact that the uncertainty in thermophysical properties has on key metrics of thermal hydraulic importance for MSRs, in particular on the heat transfer coefficient. In order to achieve this, computational fluid dynamics (CFD) simulations using the RANS k-ω SST model were compared to published experiment data on molten salt.
6/18/2020 5:47:10 AM +00:00
Breed-and-burn fuel cycle in molten salt reactors
In this paper the applicability of breed-and-burn to molten salt reactors is investigated first on a cell level using a modified neutron excess method. Several candidate fuel salts are selected and their performance in a conceptual three-dimensional reactor is investigated.
6/18/2020 5:47:04 AM +00:00
Development of a cold plug valve with fluoride salt
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.
6/18/2020 5:46:57 AM +00:00
Heat exchanger design studies for molten salt fast reactor
In this study, conceptual design for primary heat exchanger of the Molten Salt Fast Reactor is made. The design was carried out to remove the produced heat from the reactor developed under the SAMOFAR project.
6/18/2020 5:46:51 AM +00:00
Evaluation of cobalt free coatings as hardfacing material candidates in sodium-cooled fast reactor and effect of oxygen in sodium on the tribological behaviour
The feedback produced by operating Sodium-cooled Fast Reactors (SFRs) has shown the importance of material tribological properties. Where galling or adhesive wear cannot be allowed, hardfacing alloys, known to be galling-resistant coatings, are usually applied on rubbing surfaces.
6/18/2020 5:46:44 AM +00:00
Quick calculation of damage for ion irradiation: implementation in Iradina and comparisons to SRIM
Binary 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.
6/18/2020 5:46:38 AM +00:00
Microstructure and mechanical properties relationship of additively manufactured 316L stainless steel by selective laser melting
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.
6/18/2020 5:46:31 AM +00:00
Karlsruhe Nuclide Chart – New 10th edition 2018
In this article, we present the scientific highlights of the new 10th Edition of the Karlsruhe Nuclide Chart. The main focus of this Chart is to provide structured, accurate information on the half-lives and decay modes, as well as energies of the emitted radiation for over 4000 experimentally observed ground states and isomer nuclides to an interdisciplinary audience.
6/18/2020 5:46:25 AM +00:00
Neutronic benchmark of the molten salt fast reactor in the frame of the EVOL and MARS collaborative projects
This paper describes the neutronic benchmarks and the results obtained by the various participants of the FP7 project EVOL and the ROSATOM project MARS. The aim of the benchmarks was two-fold: first to verify and validate each of the code packages of the project partners, adapted for liquid-fueled reactors, and second to check the dependence of the core characteristics to nuclear data set for application on a molten salt fast reactor (MSFR).
6/18/2020 5:46:18 AM +00:00
How to obtain an enhanced extended uncertainty associated with decay heat calculations of industrial PWRs using the DARWIN2.3 package
This paper focuses on the strategy that could be used to resolve this issue with the complement and the exploitation of the DARWIN2.3 experimental validation.
6/18/2020 5:46:11 AM +00:00
The Uranie platform: an open-source software for optimisation, meta-modelling and uncertainty analysis
This paper introduces the Uranie platform, an open-source framework developed at the Alternative Energies and Atomic Energy Commission (CEA), in the nuclear energy division, in order to deal with uncertainty propagation, surrogate models, optimisation issues, code calibration, etc.
6/18/2020 5:46:04 AM +00:00
Stimulating niche nurturing process for heat production with nuclear plants in France: A multi-level perspective
This paper examines how intermediaries could interact with other important actors identified by the multi-level perspective (MLP) framework, the niche actors and regime actors, to create niches for nuclear heat production in France.
6/18/2020 5:45:58 AM +00:00
Characterization of fresh EMPIrE and SEMPER FIDELIS U(Mo)/Al fuel plates made with PVD-coated U(Mo) particles
This paper focuses on analyses made at CEA Cadarache on seven fresh plates made of atomized particles, with or without Mo homogenization, and with ZrN coating.
6/18/2020 5:45:52 AM +00:00
Pin to pin neutron flux reconstruction in a PWR reactor using support vector regression (SVR) technique
Coarse mesh nodal methods are widely used in the analysis of nuclear reactors. However, these methods provide only average values of the neutron fluxes. From a safety point of view, it is important to have an accurate analysis of the pin to pin flux distribution that nodal methods are not able to provide.
6/18/2020 5:45:46 AM +00:00
Extension of Bayesian inference for multi-experimental and coupled problem in neutronics a revisit of the theoretical approach
In this document, we rewrite Orlov theory to extend to multiple experimental values and parameters adjustment. We found that the multidimensional system expression looks like can be written as the monodimensional system in a matrix form.
6/18/2020 5:45:40 AM +00:00
Efficient use of Monte Carlo: the fast correlation coefficient
Random sampling methods are used for nuclear data (ND) uncertainty propagation, often in combination with the use of Monte Carlo codes (e.g., MCNP). One example is the Total Monte Carlo (TMC) method.
6/18/2020 5:45:33 AM +00:00
Critical review of CIELO evaluations of n + 235U, 238U using differential experiments
Key reactions have been selected to compare JEFF-3.3 (CIELO 2) and IAEA CIELO (CIELO 1) evaluated nuclear data files for neutron induced reactions on 235U and 238U targets. IAEA CIELO evaluation uses reaction models to construct the evaluation prior, but strongly relied on differential data including all reaction cross sections fitted within the IAEA Neutron Standards project.
6/18/2020 5:45:28 AM +00:00
Evaluating nuclear data and their uncertainties
This paper discusses some uncertainty quantification methodologies in use today, their strengths, their pitfalls, and alternative approaches that have proved to be highly successful in other fields.
6/18/2020 5:45:21 AM +00:00
Use of integral data assimilation and differential measurements as a contribution to improve 235U and 238U cross sections evaluations in the fast and epithermal energy range
This work makes use of the Generalized Least Square method to solve Bayes equation, as implemented in the CONRAD code. Experimental database used includes ICSBEP Uranium based critical experiments and benefits from recent re-analyses of MASURCA and FCA-IX criticality experiments (with Monte-Carlo calculations) and of PROFIL irradiation experiments.
6/18/2020 5:45:15 AM +00:00
Nuclear data uncertainty analysis for the Po-210 production in MYRRHA
MYRRHA is a multi-purpose research reactor able to operate in sub-critical and critical modes and currently in the design phase at SCK•CEN. The choice of LBE was driven by its chemical stability, low melting temperature, high boiling point, low chemical reactivity with water and air and a good neutronic performance.
6/18/2020 5:45:09 AM +00:00