Assessing the accuracy of energy turbulent diffusion dispersion correlation in a porous two-fluid model dedicated to PWR core simulations

International audienceCATHARE is a 2-fluid thermal-hydraulic code, capable of simulating thermal and mechanical phenomena occurring in the primary and secondary circuits of Pressurized Water Reactor under a wide variety of accidental situations. One of the medium-term objectives of system code CATHARE-3 is modeling a PWR core at assembly scale to simulate various accidental situations such as the loss of coolant accident (LOCA) and steam line break accident. This requires the monophasic and two-phase models that adapted to the assembly scale. However, there exists 3D models for the whole core and sub-channel scale models, which have a certain degree of validation. For more macroscopic three-dimensional models, we only have global validations without local measurements, which is necessary for the validations of each closure law's separate effects. The objective of my PhD project is improving the sub-channel scale models and developing the assembly scale models in CATHARE-3 system code with the sub-channel scale simulations and experiments results

The european project NURISP for nuclear reactor simulation

The NURISP project aims at developing the European NURESIM reference simulation platform [1] for nuclear reactor. A first version of NURESIM was delivered in 2008. 22 organizations from 14 European countries contribute to the further development of this platform. NURISP also includes a User’s Group (UG) whose members are not NURISP partners and come from the industrial nuclear sector or European and non-European R&D labs. Users can benefit from the use of the NURESIM platform, methods, results and modules and they provide concrete input and feedback on the use of these elements

Prospective For Nuclear Thermal Hydraulic Created By Ongoing And New Networks

International audienceThis paper introduces the FONESYS, SILENCE and CONUSAF projects run by some of the leading organizations working in the nuclear sector.The FONESYS members are developers of some of the major System Thermal-Hydraulic (SYS-TH) codes adopted worldwide, whereas the SILENCE members own and operate important thermal-hydraulic experimental facilities. The two networks work in a cooperative manner and have at least one meeting per year where top-level experts in the areas of thermal-hydraulic code development and experimentation are gathered.The FONESYS members address various topics such as hyperbolicity and numerics in SYS-TH codes, 3-field modeling, transport of interfacial area, 3D modeling, scaling of thermal-hydraulic phenomena, two-phase critical flow (TPCF), critical heat flux (CHF), and others. As part of the working modalities, some numerical benchmarks were proposed and successfully conducted by the network, addressing some of the most relevant topics selected by the FONESYS members.On the other hand, SILENCE addresses topics such as identification of current measurement needs and main gaps for further SYS-TH and CFD codes development and validation, definition of similar tests and counterpart tests in Integral Tests Facilities (including containment thermal-hydraulics) to be possibly conducted on Members' test facilities, scaling issue, and other subjects. Furthermore, SILENCE organized a Specialists Workshop on Advanced Instrumentation and Measurement Techniques for Nuclear Reactor Thermal-Hydraulics (SWINTH) which was held in Italy on June 2016. A second edition of the Workshop, namely SWINTH-2019, will be held in Italy in 2019 under the umbrella of the OECD/NEA/CSNI/WGAMA.Recently a new initiative is being taken by launching an international consortium of nuclear thermal-hydraulics code users, the CONUSAF. The main idea is to enhance the interactions between the users of computational tools in nuclear TH, noticeably including SYS-TH and CFD codes, the code developers and the experimentalists. The proposed initiative is expected to have a positive impact on the entire ecosystem by pursuing the assessment of the current code limitations and capabilities, analyzing and addressing issues raised by the users and promoting common RandD efforts on topics of high relevance

Role of carbon allocation efficiency in the temperature dependence of autotroph growth rate

To predict how plant growth rate will respond to temperature requires understanding how temperature drives the underlying metabolic rates. Although past studies have considered the temperature dependences of photosynthesis and respiration rates underlying growth, they have largely overlooked the temperature dependence of carbon allocation efficiency. By combining a mathematical model that links exponential growth rate of a population of photosynthetic cells to photosynthesis, respiration, and carbon allocation; to an experiment on a freshwater alga; and to a database covering a wide range of taxa, we show that allocation efficiency is crucial for predicting how growth rates will respond to temperature change across aquatic and terrestrial autotrophs, at both short and long (evolutionary) timescales

Phase appearance or disappearance in two-phase flows

This paper is devoted to the treatment of specific numerical problems which appear when phase appearance or disappearance occurs in models of two-phase flows. Such models have crucial importance in many industrial areas such as nuclear power plant safety studies. In this paper, two outstanding problems are identified: first, the loss of hyperbolicity of the system when a phase appears or disappears and second, the lack of positivity of standard shock capturing schemes such as the Roe scheme. After an asymptotic study of the model, this paper proposes accurate and robust numerical methods adapted to the simulation of phase appearance or disappearance. Polynomial solvers are developed to avoid the use of eigenvectors which are needed in usual shock capturing schemes, and a method based on an adaptive numerical diffusion is designed to treat the positivity problems. An alternate method, based on the use of the hyperbolic tangent function instead of a polynomial, is also considered. Numerical results are presented which demonstrate the efficiency of the proposed solutions

Eco-evolutionary dynamics in fragmented landscapes

Peer reviewedPostprin

Critical flow prediction by system codes – Recent analyses made within the FONESYS network

A benchmark activity on Two-Phase Critical Flow (TPCF) prediction was conducted in the framework of the Forum & Network of System Thermal-Hydraulics Nuclear Reactor Thermal-Hydraulics (FONESYS). FONESYS is a network among code developers who share the common objective to strengthen current technology. The aim of the FONESYS Network is to highlight the capabilities and the robustness as well as the limitations of current SYSTH codes to predict the main phenomena during transient scenarios in nuclear reactors for safety issues. Six separate effect test facilities, more than 90 tests, both in steady and transient conditions, were considered for the activity. Moreover, two ideal tests were designed for code to code comparison in clearly defined conditions. Overall eight System Thermal-Hydraulic (SYS-TH) codes were adopted, mostly by the developers themselves, ensuring the minimization of the user effect. Results from selected tests were also compared against Delayed Equilibrium Model, not yet implemented in industrial version of SYS-TH codes. Generally, the results of the benchmark show an improvement of the capability of SYS-TH codes to predict TPCF in the last three decades. However, predicting break flowrate remains a major source of uncertainty in accidental transient simulations of Water-Cooled Nuclear Reactors (WCNR). A set of possible actions is proposed to go beyond the current limitations of choked flow models. More detailed guidelines for using 0-D choked flow models is possible by using the experience gained by the benchmark results as well as all available validation results. Progress in understanding and 1-D modelling of flashing and choked flow might be achieved by a deeper physical analysis leading to more mechanistic models based on specific flow regime maps for high speed flow. Also the use of advanced 3-D numerical tools may help to understand and predict the complex 3-D geometrical effect

Review of Available Data for Validation of Nuresim Two-Phase CFD Software Applied to CHF Investigations

The NURESIM Project of the 6th European Framework Program initiated the development of a new-generation common European Standard Software Platform for nuclear reactor simulation. The thermal-hydraulic subproject aims at improving the understanding and the predictive capabilities of the simulation tools for key two-phase flow thermal-hydraulic processes such as the critical heat flux (CHF). As part of a multi-scale analysis of reactor thermal-hydraulics, a two-phase CFD tool is developed to allow zooming on local processes. Current industrial methods for CHF mainly use the sub-channel analysis and empirical CHF correlations based on large scale experiments having the real geometry of a reactor assembly. Two-phase CFD is used here for understanding some boiling flow processes, for helping new fuel assembly design, and for developing better CHF predictions in both PWR and BWR. This paper presents a review of experimental data which can be used for validation of the two-phase CFD application to CHF investigations. The phenomenology of DNB and Dry-Out are detailed identifying all basic flow processes which require a specific modeling in CFD tool. The resulting modeling program of work is given and the current state-of-the-art of the modeling within the NURESIM project is presented

Accuracy of Eulerian–Eulerian, two-fluid CFD boiling models of subcooled boiling flows

Boiling flows are frequently found in industry and engineering due to the large amount of heat that can be transferred within such flows with minimum temperature differences. In the nuclear industry, boiling affects in different ways the operation of almost all water-cooled nuclear reactors. Recently, the use of computational fluid dynamic (CFD) approaches to predict boiling flows is increasing and, in the nuclear area, CFD is being developed to solve thermal hydraulic safety issues such as establishing the critical heat flux, which is perhaps the major threat to the integrity of nuclear fuel rods. In this paper, the accuracy of an Eulerian–Eulerian, two-fluid CFD model is evaluated over a large database of subcooled boiling flows, avoiding the rather popular case-by-case tuning of descriptive models to a limited number of experiments. The model includes a Reynolds stress turbulence model, the method of moments-based S-gamma population balance approach and a boiling model derived using the heat flux partitioning approach. The database covers a large range of conditions in subcooled boiling flows of water and refrigerants in vertical pipes and annular channels. Overall, a satisfactory predictive accuracy is achieved for some quantities of interest, such as the void fraction and the turbulence and liquid temperature fields, but results are less satisfactory in other areas, more specifically for the average bubble diameter and the mean velocity profiles close to the wall in annular channels. Agreement may be improved with advances in the treatment of large bubbles and bubble break-up and coalescence, as well as in improved modelling of the boiling region close to the wall, and more specifically the bubble departure diameter, the wall treatment and the contribution of bubbles to turbulence