Coupled Calculation on Fluid Structure Interaction in Plate-Type Fuel Element

In nuclear reactors that use plate-type fuel, the fuel plates are thermally managed with coolant flowing through channels between the plates. Depending on the flow rates and sizes of the fluid channels, the hydraulic forces exerted on a plate can be quite large. Currently, there is a worldwide effort to convert research reactors that use highly enriched uranium (HEU) fuel, some of which are plate-type, to low-enriched uranium (LEU). Because of the proposed changes to the fuel structure and thickness, a need exists to characterize the potential for flow-induced deflection of the LEU fuel plates. In this study, as an initial step, calculations of Fluid-Structure Interaction (FSI) for a flat aluminum plate separating two parallel rectangular channels are performed using the commercial code STAR-CCM+ and the integrated multi-physics code SHARP, developed under the Nuclear Energy Advanced Modeling and Simulation program. SHARP contains the high-fidelity single physics packages Diablo and Nek5000, both highly scalable and extensively validated. In this work, verification studies are performed to assess the results from both STAR-CCM+ and SHARP. The predicted deflections of the plate agree well with each other as well as exhibiting good agreement with simulations performed by the University of Missouri utilizing STAR-CCM+ coupled with the commercial structural mechanics code ABAQUS. The study provides a solid basis for FSI modeling capability for plate-type fuel element with SHARP.

Numerical Simulation of Corium Jet Breakup During Fuel-Coolant Interaction Based on the ALISA Test Performed at KROTOS Test Facility

Steam explosion is one of the consequences of fuel-coolant interactions in a severe accident. Melt jet fragmentation, which is the key phenomenon during steam explosion, has not been clarified sufficiently which prevents the precise prediction of steam explosion. The focus of this paper is on the numerical simulation of the melt jet behavior falling into a coolant pool in order to get a qualitative and quantitative understanding of initial premixing stage of fuel-coolant interaction. The objective of our first phase is the simulation of the fragmentation process and the estimation of the jet breakup length. A commercial CFD code COMSOL is used for the 2D numerical analysis employing the phase field method. The simulation condition is similar to our steam explosion test supported by the ALISA (Access to Large Infrastructure for Severe Accidents) project between European Union and China, and carried out in the KROTOS test facility at CEA, France. The simulation result is in relatively good agreement with the experimental data. Then the effect of the initial jet velocity, the jet diameter and the instability theory are presented. The preliminary data of melt jet fragmentation is helpful to understand the premixing stage of the fuel-coolant interaction.

Optimization of Moving Particle Semi-Implicit (MPS) Method and Studying of Flow Instability

Multiphase flow widely exists in the nature and engineering. The two-phase flow is the highlight of the studies about the flow in the vessel and steam explosion in nuclear severe accidents. The Moving Particle Semi-implicit (MPS) method is a fully-Lagrangian particle method without grid mesh which focuses on tracking the single particle and concerns with its movement. It has advantages in tracking complex multiphase flows compared with gird methods, and thus shows great potential in predicting multiphase flows. The objective of this thesis is to develop a general multiphase particle method based on the original MPS method and thus this work is of great significance for improving the numerical method for simulating the instability in reactor severe accident and two-phase flows in vessel. This research is intended to provide a study of the instability based on the MPS method. Latest achievements of mesh-free particle methods in instability are researched and a new multiphase MPS method, which is based on the original one, for simulating instability has been developed and validated. Based on referring to other researchers’ papers, the Pressure Poisson Equation (PPE), the viscosity term, the free surface particle determination part and the surface tension model are optimized or added. The numerical simulation on stratification behavior of two immiscible flows is carried out and results are analyzed after data processing. It is proved that the improved MPS method is more accurate than the original method in analysis of multiphase flows. In this paper, the main purposes are simulating and discussing Rayleigh-Taylor (R-T) instability and Kelvin-Helmholtz (K-H) instability. R-T and K-H instability play an important role in the mixing process of many layered flows. R-T instability occurs when a lower density fluid is supported by another density higher fluid or higher density fluid is accelerated by lower density fluid, and the resulting small perturbation increases and eventually forms turbulence. K-H instability is a small disturbance for two different densities, such as waves, at the interface of the two-phase fluid after giving a fixed acceleration in the fluid. Turbulence generated by R-T instability and K-H instability has an important effect in applications such as astrophysics, geophysics, and nuclear science.

Numerical Simulation of the Performance of an Axial Compressor Operating With Supercritical Carbon Dioxide

Using supercritical carbon dioxide (SCO2) as the working fluid of a closed Brayton cycle gas turbine is widely recognized nowadays, because of its compact layout and high efficiency for modest turbine inlet temperature. It is an attractive option for geothermal, nuclear and solar energy conversion. Compressor is one of the key components for the supercritical carbon dioxide Brayton cycle. With established or developing small power supercritical carbon dioxide test loop, centrifugal compressor with small mass flow rate is mainly investigated and manufactured in the literature; however, nuclear energy conversion contains more power, and axial compressor is preferred to provide SCO2 compression with larger mass flow rate which is less studied in the literature. The performance of the axial supercritical carbon dioxide compressor is investigated in the current work. An axial supercritical carbon dioxide compressor with mass flow rate of 1000kg/s is designed. The thermodynamic region of the carbon dioxide is slightly above the vapor-liquid critical point with inlet total temperature 310K and total pressure 9MPa. Numerical simulation is then conducted to assess this axial compressor with look-up table adopted to handle the nonlinear variation property of supercritical carbon dioxide near the critical point. The results show that the performance of the design point of the designed axial compressor matches the primary target. Small corner separation occurs near the hub, and the flow motion of the tip leakage fluid is similar with the well-studied air compressor. Violent property variation near the critical point creates troubles for convergence near the stall condition, and the stall mechanism predictions are more difficult for the axial supercritical carbon dioxide compressor.

Technical Workforce Education and Training Program at Abu Dhabi Polytechnic: Integration of Academia and Industry Requirements

The United Arab Emirates (UAE) is a developing affluent nation. The leaders of the UAE announced the pursuit of peaceful nuclear power in 2008 and by the end of the following year established its Nuclear Energy Program Implementing Organization (the Emirates Nuclear Energy Corporation (ENEC)), Federal Authority for Nuclear Regulation (FANR), and ordered four APR-1400 pressurized water reactors from the Korean Electric Power Company (KEPCO). Nuclear Engineering programs were initiated soon afterwards at Khalifa University for graduate students and the University of Sharjah for undergraduate students. The technical workforce including nuclear power plant local operators and chemistry and radiation protection personnel was established by ENEC and the Institute of Applied Technology as an inaugural program of Abu Dhabi Polytechnic (AD Poly) in 2011. This paper describes the development of the dual education and training program at AD Poly, the experience of the initial cohorts who conducted their training at the APR-1400 units at the Shin Kori Nuclear Power Plant in Korea, and the current program between the AD Poly Abu Dhabi campus and the new Barakah Nuclear Power Plant based on lessons learned from the earlier years.

Post-Test CFD Analysis of Non-Uniformly Heated 19-Pin Fuel Bundle Cooled by HLM

In the context of the studies on GEN. IV/ADS nuclear systems, the correct evaluations of the temperature distribution in the fuel pin bundle is of central interest. In particular, the use of lead or lead-bismuth eutectic (LBE) as coolant for the new generation fast reactors is one of the most promising choices. Due to the high density and high conductivity of lead or LBE, a detailed analysis of the thermo-fluid dynamic behavior of the heavy liquid metal (HLM) inside the sub-channels of a fuel rod bundle is necessary in order to support the Front-End Engineering Design (FEED) of GEN. IV/ADS prototypes and demonstrators. In this frame, the synergy between numerical analysis by CFD and data coming from large experimental facilities seems to be crucial to assess the feasibility of the components. At ENEA-Brasimone R.C., large experimental facilities exist to study HLM free, forced and mixed convection in loops and pools: e.g. NACIE-UP is a large scale LBE loop for mixed convection experiments. The MYRRHA-19 like Fuel Pin Bundle Simulator installed in the NACIE-UP facility allows to make non-uniform and dissymmetric tests with only a few pins heated. This technical feature of the FPS is very interesting for CFD validation and this kind of data tests in HLM fuel bundles are not so common in the literature. In the present paper, a post-test validation is made by a detailed CFD model of the test section. Experimental data, statistically treated by the error propagation theory, are briefly presented and a preliminary comparison with CFD results using different models/turbulent Prandtl numbers are shown. Three monitored section at different levels are compared both for wall and bulk temperatures. This post-test comparison with this experimental configuration is unique and represents a further step towards the validation of the CFD models and methods in fuel bundle geometries cooled by HLM.

Generation IV International Forum Education and Training Webinars: Education Tools for the Next Generation Workforce

The Gen IV International Forum (GIF) Education and Training Task Force was created to respond to the challenge of not only forming, training and/or retaining qualified Gen IV workforce but also educating and informing a more general public, policy makers on topics related to Gen IV reactor systems and cross-cutting subjects. The task force serves as a platform to enhance open education and training as well as communication and networking in support of GIF, and its objectives are to maintain the know-how in this field, to increase the knowledge of new advanced concepts, and to avoid the loss of the knowledge and competences that could seriously and adversely affect the future of nuclear energy. While many countries are either ramping up or developing nuclear power production as an important step towards economic development and environmental protection, a decrease or uncertainty of the fiscal year budgets have left organizations and agencies looking for new avenues for training and educating a qualified workforce. This has led to an increase in those looking for readily available education and training resources. Using modern internet technologies, the GIF Education and Training Task Force has launched a webinar series on Gen IV systems in September 2016, which is accessible to a broad audience and is educating and strengthening the knowledge of participants in applications to advanced reactors. This achievement is the direct result of partnering with university professors and subject matter experts who conduct live webinars on a monthly basis. The live webinars are recorded and archived as an online educational resource to the public from the GIF website (www.gen-4.org). In addition, the webinars offer unprecedented opportunities for interdisciplinary crosslinking and collaboration in education and research. The GIF webinars, with their expansion of topics, targets a large spectrum of those that do not know but are desiring to learn about the many aspects of advanced reactor systems. The details and examples of the GIF webinar modules will be presented in our paper.

CFD Modelling of Loss of Vacuum Accident (LOVA) for CFETR

Chinese Fusion Engineering Testing Reactor (CFETR) is a tokamak-type machine and next device in the roadmap for the realization of fusion energy in China, which aims to bridge the gaps between the International Thermonuclear Experimental Reactor (ITER) and the demonstration reactor (DEMO) [1]. The accident sequence starting from loss of vacuum accident (LOVA) is an important issue concerning the performance of CFETR. During LOVA, air will leak into the vacuum vessel (VV) causing fast pressurization of VV. At the same time, the high speed airflow jet will result in migration and re-suspension of the large quantity of tungsten dust produced and deposited in the lower part of plasma chamber, causing possibilities of radioactive dust leakage into the workshop and environment. In order to conduct a comprehensive analysis of the accident sequence, firstly, the airflow characteristics of LOVA should be studied. In this article, a postulated rupture of different section area is assumed due to a failed component at the equatorial port level. The computational fluid dynamic (CFD) modelling of LOVA was conducted by ANSYS CFX. The results show that the break area has significant influences on the characteristics of the airflow. Two swirling airflows are formed in the upper and lower part of the torus. The airflow characteristics are quite different when the LOVA happens during maintenance or during normal operation. A reverse flow occurs when the LOVA happens during normal operation. Yet can not be observed when LOVA occurs during maintenance. The results are the basis to the further safety study of CFETR such as the re-suspension, migration and explosion of dust.

15 Years of the European Nuclear Education Network (ENEN Association)

The European Nuclear Education Network (ENEN) was established in 2003 through an EU Fifth Framework Programme (FP) project, as a legal nonprofit-making body. Its main objective is the preservation and further development of expertise in the nuclear fields by higher education and training. This objective is realized through the cooperation between EU universities involved in education and research in nuclear disciplines, nuclear research centers and the nuclear industry. As of March 2018, ENEN has 66 members in 18 EU countries and has concluded Memoranda of Understanding (MoU) with partners beyond Europe for further cooperation, including organizations in, Russian Federation, Ukraine, Canada and Japan. ENEN also has good collaboration with national networks and international organizations such as the Belgian Nuclear Education Network (BNEN) and the International Atomic Energy Agency (IAEA). The main activities developed, and results achieved, within the first 15 years of the ENEN Association are presented and discussed. These include, for example, the launch of the European Master of Science in Nuclear Engineering (EMSNE), the annual ENEN Ph.D. competition and the portfolio of more than 10 EURATOM projects dealing with nuclear education, training and knowledge management through development of teaching methods and materials, courses, and exchange of students and teachers within EU and beyond. Those projects were all supported by the European Commission with the ENEN Association acting as the coordinator or as a partner.