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

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

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

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

OECD/NEA/CSNI state-of-the-art report on scaling in system thermal-hydraulics applications to nuclear reactor safety and design (The S-SOAR)

The present paper deals with scaling in nuclear-system thermal-hydraulics (SYS TH), including the connection with Nuclear Reactor Safety Technology (NST). The paper is entirely derived from the S-SOAR document issued by CSNI of NEA, NEA/CSNI/R(2016)14, 2016 (Bestion et al., 2016). Scaling has constituted ‘an issue’ since the beginning of the exploitation of nuclear energy for civil purposes, with main reference to the generation of electricity. A Nuclear Power Plant (NPP) constitutes a technologically complex industrial system and it is characterized by the impossibility of, or the large difficulty in, characterizing the system's performance under the conditions of the design. So, models were designed, constructed, and operated under downscaled ranges of values for one or more of selected parameters (e.g. power, volume, height, pressure, etc). These features lay at the origin of the scaling issue, i.e. the difficulty in demonstrating that a model behaves like the prototype. Integrated definitions of the widely adopted terms, ‘scaling’, ‘scaling issue’, and ‘addressing the scaling issue’ are part of the present document. The related application domain includes the NST, and the licensing for water-cooled nuclear reactors under operation, under construction, or under an advanced design stage at the time of publication of the (Bestion et al., 2016). Scaling-related analyses are done in different areas of SYS TH and NST. These include the design of test facilities (both integral and separate-effect test facilities, IETF and SETF), the design of experiments (including Counterpart Test, CT), the demonstration of the capability of any computational tool, and the evaluation of uncertainty affecting the prediction of the same computational tools. A variety of approaches have been used to address the scaling issue, including non-dimensional analysis of mass, energy- and momentum-balance equations, derivation and application of scaling factors, including the hierarchy of relative importance, performing experiments at different scales, and running the SYS TH computer codes. This paper discusses the key areas and the key approach for scaling. It was found that the SYS TH computer codes, following their application to differently scaled experiments, demonstrate that the accuracy of their predictions may not depend upon the scale of the considered experiments. The TH codes also may constitute an additional valuable tool for addressing the issue of scaling