Search Results (7)

The Dual Fluid Reactor (DFR) is a Generation IV concept that relies on a phased development pathway using a low-temperature microdemonstrator ($\mu$DEMO) and a high-temperature minidemonstrator (mDEMO). A rigorous methodology is required to scale experimental data between these facilities to ensure the reliable design of the final reactor. This paper establishes such a methodology grounded in Similarity Theory. The Cathare-2 system code was used to perform a parametric study on a simplified model of the demonstrators, which use lead–bismuth eutectic and pure liquid lead, respectively. This study focused on identifying the specific operating conditions required to match key “defining” dimensionless numbers—the Reynolds number (Re) for dynamic similarity and the Peclet number (Pe_h) for thermal similarity. The analysis successfully identified and presented the distinct operating ranges of fluid velocity and mass flow required to achieve either state. Results show that matching the Reynolds number allows for the dimensionless pressure drop to be scaled with a deviation below 0.2%, while matching the Peclet number allows for the dimensionless temperature profile to be scaled with a deviation under 2.5%. The central finding is that dynamic and thermal similarity cannot be achieved simultaneously due to the different working fluids and temperatures of the demonstrators. This forces a strategic choice in experimental design, where an experiment must be tailored to investigate either fluid dynamics or heat transfer. This work provides the foundational “rulebook” for designing these crucial experiments, ensuring that data from the DFR demonstrator program is both reliable and scalable. Full article

The image of the Virgin Mary appears with increasing frequency in written sources from the 12th and 13th centuries compared to earlier periods. Three major works produced by four eyewitness authors of the Albigensian Crusade (Historia Albigensis, Chronica, and Canso de la Crozada) reflect on and respond to this popular theme. These sources focus on the Albigensian Crusade against heretical groups, particularly the Cathars, and employ the Virgin Mary motif for various purposes. The Virgin Mary is presented as a Catholic model for women drawn to Catharism (a movement in which female spiritual leadership was also present) as a divine protector of the just side in war and as a means of legitimizing the authors’ claims. While Mary appears sporadically in Peter of Vaux-de-Cernay’s Historia Albigensis, she is extensively invoked in the Canso by both William and his anonymous successor. In contrast, the image of the Virgin Mary is scarcely mentioned in Chronica, likely due to the narrative’s intended audience and objectives. This article aims to provide a comparative analysis of how the image of the Virgin Mary is utilized in these primary sources from the Albigensian Crusade and to offer a new perspective on the relationship between historical events and authors’ intentions, laying the groundwork for further research. Full article

To achieve the decarbonisation goal by 2050, nuclear energy can be a useful element for the future energy mix, complementing intermittent renewable sources. Additionally, heat from the core can be used for cogeneration, aiding the decarbonisation of several energy sectors. In this context, Small Modular Reactors (SMRs) are being studied when introduced in Nuclear–Renewable Hybrid Energy Systems for cogeneration applications. However, nuclear cogeneration with SMRs is still an emerging area of study, requiring careful considerations regarding technical, safety, and economic aspects. European research initiatives, such as the TANDEM project, are exploring the integration of light–water SMRs into hybrid systems. This paper investigates the impact of cogeneration transients on the primary system of an SMR using a novel coupling approach. For this scope, the thermal–hydraulic system code CATHARE 3 and the dynamic modelling language MODELICA are adopted. Three transient scenarios were analysed: cogeneration transitions, core power variations, and thermal load rejection. The results achieved provide insights about the robustness of the numerical coupling and the primary system response to cogeneration-induced transients. As a matter of fact, the analysis shows that the reactor system is mildly influenced by cogeneration changes, and the findings suggest future improvements for both the coupling methodology and modelling assumptions. Full article

Best estimate system thermal-hydraulic codes in the nuclear engineering community, e.g., TRACE, RELAP3D, CATHARE-3, etc., were extended with 3D coarse-mesh components to better describe the 3D Thermal-Hydraulic (TH) phenomena taking place within the Reactor Pressure Vessel (RPV) and the core. The RPV is usually shaped like a cylinder while the core is mostly a cube. Hence, the TRACE code is equipped with a Cylindrical VESSEL and a Cartesian VESSEL. The former one is to represent the RPV (including core), pressurizer, and steam generator. The latter one is more appropriate to represent the core. The two components are connected by two Vessel-Junctions (VJ) at the core inlet and outlet. Due to the different nodalization between the two VESSELs, the analyst needs to do repetitive and error-prone work defining the cell-to-cell junctions and their TH parameters. To facilitate this process, the Karlsruhe Institute of Technology (KIT) has developed an automatic approach based on a mesh-constructing and field-mapping library, namely the MEDCoupling. These new capabilities of TRACE are demonstrated by the analysis of the coolant mixing for an academic case and the AP1000 reactor. Full article

Codes and methods are subjected to a continuous update process for answering the regulatory requirements concerning the long-term operation of existing reactors and new concept deployment. In this continuous improvement process, new generation codes are developed for supporting industrial applications and the long-term strategy. In this paper, attention is devoted to selecting codes under development in France for lattice and core steady-state and transient calculations. These codes, APOLLO3^® and CATHARE3, have been selected for carrying out the activities of the H2020 CAMIVVER Project oriented to the 3D-multiphysics couplings improvements. Multiscale and multiphysics solutions are key topics to keep competitiveness and answer to newer industrial needs in plant operations, licensing, and safe operating envelope requirements. The paper presents an overview of the activities performed by Framatome to support the definition of benchmarks exercises proposed in the CAMIVVER Project. Small core configurations subjected to a Reactivity Insertion Accident (RIA) are presented with associated preliminary results. To open the discussions toward the development of Best-Estimate Plus Uncertainties (BEPU) solutions, the URANIE statistical platform was used for sensitivity analysis over different configurations. These preliminary results are also presented. Full article

In line with Polish national activities and research programs investigating non-electrical-reactor use, the national GOSPOSTRATEG-HTR project was launched, aiming at the development of a novel pre-conceptual design of a High-Temperature Gas-cooled Reactor (HTGR). The 40 MW_th research reactor would serve as a technology demonstrator for future industrial purposes. In the paper, the proposal of an established thermal-hydraulic and neutronic core design is presented as a result of the National Centre for Nuclear Research team studies, in the scope of the project, including the areas of fluid mechanics, heat exchange and reactor neutronic core design support analyses. The undertaken analyses were confirmed by the series of code investigations involving integral thermal-hydraulic (MELCOR (Sandia National Laboratories, USA), CATHARE (CEA, France)), neutronic (Serpent (VTT, Finland), MCB (AGH University’s Department of Nuclear Energy, Poland)), Computational Fluid Dynamics (ANSYS Fluent (ANSYS, USA)) and others. The calculations performed within the preliminary safety analysis on the pre-concept showed its compliance with international safety standards for the normal operation and Design Basis Accident sequences. Full article

A predictive transient two-phase flow rotodynamic pump model has been developed in the Code for Analysis of THermalhydraulics during an Accident of Reactor and safety Evaluation (CATHARE-3). Flow inside parts of the pump (suction, impeller, diffuser and volute) is computed according to a one-dimensional discretisation following a mean flow path. Transient governing equations of the model are solved using an implicit resolution method and integrated along the curvilinear abscissa of the element. This model has been previously qualified at the component scale by comparison to an existing experimental database. The present study aims at extending the validation at the system scale: a whole experimental test loop is modelled. The ability of the transient pump model to predict flow rate, head and torque as a function of time during a 1-s pump fast start-up is evaluated. The transient evolution of the pressure upstream and downstream from the centrifugal pump is well predicted by the simulation compared to the measurements. Local quantities such as pressure and velocity inside elements of the circuit are analysed. In the considered case, inertial effects of the global circuit are dominant when compared to pump inertial effects due to the high characteristic lengths of the pipes. The main perspective of this work consists in the simulation of similar pump transients, in cavitating conditions. Full article