Attila Aszódi

Numerical simulation on a HPLWR fuel assembly flow with one revolution of wrapped wire spacers

Abstract Three dimensional computational-fluid-dynamics simulations are performed for the fluid flow within a 40 rod fuel bundle in a square arrangement with a central moderator channel. To ensure spacing between the rods, the design of the bundle uses thin wires wrapped counter-clockwise around each rod. This geometry is presently investigated in the framework of the European High-Performance Light-Water Reactor (HPLWR), which operates at supercritical pressure of 25 MPa. A section with one revolution located in the evaporator region of the HPLWR core is investigated using hydraulic (to ensure fully developed flow inlet boundary conditions and reference for heated cases) and thermal-hydraulic boundary conditions. The geometry of wrapped wires gives rise to additional mixing and a circulating or ‘sweeping’ flow near the outer and inner regions of the fuel element next to the wall of the so called fuel assembly and moderator box. Some interesting flow features associated with the complex three-dimensional flow with significant transverse velocity components are visualized as the first evaluated result of this diversified investigation.

SUMMARY FOR THREE DIFFERENT VALIDATION CASES OF COOLANT FLOW IN SUPERCRITICAL WATER TEST SECTIONS WITH THE CFD CODE ANSYS CFX 11.0

Computational fluid dynamics (CFD) codes have become promising tools for the investigation of thermal hydraulics in revolutionary reactor concepts in the last decade. In Reynolds-averaged Navier-Stokes calculations, the CFD codes (for example, the ANSYS CFX code used here) use turbulence modeling, wall functions, and other approaches. Therefore, the accuracy of CFD codes for water flow under supercritical conditions has to be examined. The first aim of this work is to investigate the effects of different material property definition methods on the numerical results obtained with CFX code. The second aim is to assess the accuracy of the conventional turbulence models (such as k-ε, k-ω, and SST) under supercritical water conditions. The results and comparison of three independent validations for supercritical water flow in vertical smooth-bore tubes with upward flow direction are presented in this paper. It is well known that the material properties strongly depend on the temperature and the pressure near and above the thermodynamic critical point. It is demonstrated that rather than analytical or discrete point methods, the IAPWS-IF97 material table best represents the strongly changing material properties. A nonaxialsymmetric effect on result fields was not found based on the three validations; therefore, a rotational periodic or two-dimensional grid approach is recommended for further validations of homogenously heated, vertically installed, smooth-bore straight tubes cooled by supercritical water. The calculation results have been compared with measurements, and the computational errors for the three validations were found to be in the ranges of 0 to 25%, 0 to 18%, and 2 to 40% for the Swenson, Yamagata, and Herkenrath experiments, respectively. The results of the three validations indicate the need to improve a turbulence model to take into account the buoyancy effect on the turbulence for thermal-hydraulic calculations of the supercritical water.

Detailed CFD Analysis of Coolant Mixing in VVER-440 Fuel Assembly Heads Performed With the Code CFX-5.5

3D modeling of the thermal hydraulical processes in a fuel assembly head means a great challenge for the CFD technique due to the complexity of its structure and the flow domain. On the other hand, this field is of great importance since detailed knowledge on mixing processes in the assembly heads and calculations on the signals of the thermocouples positioned just above the heads would give very significant information for the safety analyses connected to the power upgrading of nuclear power plants. Therefore development of a complex fuel assembly model was started in the Institute of Nuclear Techniques of the Budapest University of Technology and Economics in the near past. In this paper, the fuel assembly head model, the sensitivity study of it, calculations and results are presented. The main goal of our work is investigating the signal of the thermocouples which are placed just above the fuel assemblies. The calculations were performed with consideration of four kinds of different fuel assemblies. The inlet velocity and temperature fields were calculated by the COBRA subchannel code of the Paks Nuclear Power Plant of Hungary. With all kind of fuel assemblies, calculations were performed with assumptions of normal symmetrical and highly asymmetrical heat source profiles of inner assemblies and assemblies positioned beside absorber elements.Copyright

Experimental investigation of the MSFR molten salt reactor concept

Abstract In the paper experimental modelling and investigation of the MSFR concept will be presented. MSFR is a homogeneous, single region liquid fuelled fast reactor concept. In case of molten salt reactors the core neutron flux and fission distribution is determined by the flow field through distribution and transport of fissile material and delayed neutron precursors. Since the MSFR core is a single region homogeneous volume without internal structures, it is a difficult task to ensure stable flow field, which is strongly coupled to the volumetric heat generation. These considerations suggest that experimental modelling would greatly help to understand the flow phenomena in such geometry. A scaled and segmented experimental mock-up of MSFR was designed and built in order to carry out particle image velocimetry measurements. Basic flow behaviour inside the core region can be investigated and the measurement data can also provide resource for the validation of computational fluid dynamics models. Measurement results of steady state conditions will be presented and discussed.

Uncertainty analysis and flow measurements in an experimental mock-up of a molten salt reactor concept

Abstract In the paper measurement results from the experimental modelling of a molten salt reactor concept will be presented along with detailed uncertainty analysis of the experimental system. Non-intrusive flow measurements are carried out on the scaled and segmented mock-up of a homogeneous, single region molten salt fast reactor concept. Uncertainty assessment of the particle image velocimetry (PIV) measurement system applied with the scaled and segmented model is presented in detail. The analysis covers the error sources of the measurement system (laser, recording camera, etc.) and the specific conditions (de-warping of measurement planes) originating in the geometry of the investigated domain. Effect of sample size in the ensemble averaged PIV measurements is discussed as well. An additional two-loop-operation mode is also presented and the analysis of the measurement results confirm that without enhancement nominal and other operation conditions will lead to strong unfavourable separation in the core flow. It implies that use of internal flow distribution structures will be necessary for the optimisation of the core coolant flow. Preliminary CFD calculations are presented to help the design of a perforated plate located above the inlet region. The purpose of the perforated plate is to reduce recirculation near the cylindrical wall and enhance the uniformity of the core flow distribution.

Experimental and numerical thermal-hydraulics investigation of a molten salt reactor concept core

Abstract In the paper measurement results of experimental modelling of a molten salt fast reactor concept will be presented and compared with three-dimensional computational fluid dynamics (CFD) simulation results. Purpose of this article is twofold, on one hand to introduce a geometry modification in order to avoid the disadvantages of the original geometry and discuss new measurement results. On the other hand to present an analysis in order to suggest a method of proper numerical modelling of the problem based on the comparison of calculation results and measurement data for the new, modified geometry. The investigated concept has a homogeneous cylindrical core without any internal structures. Previous measurements on the scaled and segmented plexiglas model of the concept core and simulation results have shown that this core geometry could be optimized for better thermal-hydraulics characteristics. In case of the original geometry strong undesired flow separation could develop, that could negatively affect the characteristics of the core from neutronics point of view as well. An internal flow distributor plate was designed and installed with the purpose of optimizing the flow field in the core by enhancing its uniformity. Particle image velocimetry (PIV) measurement results of the modified experimental model will be presented and compared to numerical simulation results with the purpose of CFD model validation.

Experimental Modelling of a Molten Salt Reactor Concept

Based on the MSFR (Molten Salt Fast Reactor) reactor concept proposed within the framework of the EVOL (Evaluation and Viability of Liquid Fuel Fast Reactor System, EU FP7) international research project a scaled and segmented experimental model of the MSFR and first measurement result will be presented in the paper.MSFR is a single region, homogeneous liquid fuelled fast reactor concept. The reactor uses fluoride-based molten salts as fuel and coolant, with fissile uranium and/or thorium and other heavy nuclei content with the purpose of applying the thorium cycle and the burn-up of transuranic elements. The concept has a single region cylindrical core with sixteen radial inlet and outlet nozzles located at the bottom and top of the core. The external circuit (internal heat exchanger, pump, pipes) is broken up in sixteen identical modules distributed around the core.A scaled and segmented experimental model of the MSFR concept was designed and built in order to carry out Particle Image Velocimetry (PIV) measurements. Purpose of the experimental mock-up is to provide measurement data for validation and benchmarking of CFD simulations, and also to study specific problems or phenomena related to the MSFR, such as design of inlet geometry, effects of internal structures, coolant mixing.The experimental model uses water as working fluid with 50 μm polyamide seeding particles added for PIV measurement. Geometrical scaling was applied in order to reduce size and necessary pumping power and the geometry represents a 90 degree segment of the original cylindrical geometry. It was not possible to maintain the nominal value of the Reynolds-number (∼1E+06 for the core) however a highly turbulent flow (Re>1E+05) can be reproduced in the system.Final design of the scaled and segmented plexiglas model will be presented, capabilities and limitations of the measurement assembly will be discussed together with the presentation of first measurements results.Copyright

Preliminary Thermal-Hydraulic Analyses for Designing an Experimental Model of a Molten Salt Reactor Concept

Based on the MSFR (Molten Salt Fast Reactor) reactor concept presented within the framework of the EVOL (Evaluation and Viability of Liquid Fuel Fast Reactor System, EU FP7) international research project preliminary three-dimensional thermal-hydraulic analyses and the discussion of scaled experimental modelling will be presented.The MSFR concept is a single region, homogeneous liquid fuelled fast reactor. The reactor concept uses fluoride-based molten salts with fissile uranium and/or thorium and other heavy nuclei content with the purpose of applying the thorium cycle and the burn-up of transuranic elements. The concept has a single region cylindrical core with sixteen radial inlet and outlet nozzles located at the bottom and top of the core. The external circuit (internal heat exchanger, pump, pipes) is broken up into sixteen identical modules distributed around the core.Purpose of the three-dimensional computational fluid dynamics (CFD) calculations is to study the possibility of experimental investigation of the fluid flow in the core of the proposed MSFR concept using a scaled model and Particle Image Velocimetry (PIV) flow measurement technique.First the main properties of the proposed MSFR concept are introduced, and the information on other experimental thermal-hydraulic modelling of different reactors, including MSRE (Molten Salt Reactor Experiment) are summarised.With a scaled plexiglas MSFR model it would be possible to carry out flow field measurements under laboratory conditions using PIV method. Possible way of scaling are presented and a series of preliminary CFD calculations are discussed. Possibilities and limitations of such scaling and segmenting of a model and the use of water as substitute fluid for the experimental mock-up will be discussed.Objectives of such a measurement series would be validation, benchmarking of CFD calculations and codes, application of CFD modelling experience in the detailed thermal-hydraulic design of the MSFR concept, possible measurements for the study of specific problems or phenomena, for example refinement of inlet geometry, effects of internal structures, coolant mixing.Copyright

Expedition to the 30-km Chernobyl Exclusion Zone and the Utilization of Its Experience in Education and Communication

Between May 28 – June 4, 2005, under the organization of the Hungarian Nuclear Society (HNS) and the Hungarian Young Generation Network (HYGN) — which operates within the framework of the HNS — a scientific expedition visited the Chernobyl Nuclear Power Plant and the surrounding exclusion zone. The participants were young Hungarian nuclear professionals supervised by more experienced experts. The main scientific goals of the expedition were the followings: • Get personal experiences in a direct way about the current status of the Chernobyl Power Plant and its surroundings, the contamination of the environment and about the doses. • Gather information about the state of the shut down power plant and the shelter built above the damaged 4th unit. • Training of young nuclear experts by performing on site measurements. The Hungarian expedition successfully achieved its objectives by performing wide-range of environmental and dosimetric measurements and collecting numerous biological and soil samples. Within the 30-km exclusion zone the influence of the accident occurred 20 years ago still could be measured clearly; however the level of the radioactivity is manageable in most places. The dosimetrical measurements showed that no considerable exposure occurred among the members of the expedition. The analysis of samples has been started at the International Chernobyl Center in Slavutich. During the expedition not only environmental sampling and in-situ measurements were carried out but it was also well documented with photos and video recordings for educational, training and PR purposes. A documentary TV film was recorded during the expedition. The first-hand knowledge acquired during the expedition helps the authentic communication of the accident and its present-day consequences, which is especially important in 2006, 20 years after the Chernobyl accident. Since Ukraine and Hungary are neighbor countries the media constantly discuss the accident, the consequences and the risks of using nuclear energy. In addition in November 2005 Hungary’s parliament approved plans to extend the lifetime of the country’s four-unit nuclear power plant. In order to have the crucial public support for nuclear energy it is very important to dispel unrealistic dismay and misbelieves regarding these questions. Thus it is extremely beneficial to have a film on this topic created by nuclear professionals especially for the public audience. In 2005 a book on the Chernobyl accident was published in Hungary that covers this expedition in a full chapter [2]. We plan to present the film to the audience of the conference.Copyright

CFD Analyses of Damaged Fuel Inside a Cleaning Vessel

On 10–11th of April, 2003, a serious incident occurred in a special fuel assembly cleaning tank, which was installed into the service shaft of the 2nd unit of the Paks NPP in Hungary. During this incident, most of the 30 fuel assemblies put into the cleaning tank have seriously damaged. In the Institute of Nuclear Techniques of the Budapest University of Technology and Economics several CFD investigations were performed concerning the course of the incident, the post incidental conditions and the recovery work. The main reason of the incident can be originated from the defective design of the cleaning tank which resulted in the insufficient cooling of the system in a special operational mode. Our investigation performed with a complex 3D CFX model clearly showed how could as strong temperature stratification develop inside the cleaning tank that it was able to block the coolant flow through the fuel assemblies. After the blocking of the flow, the coolant turned into boiling and the assemblies became uncovered. The temperature of the surfaces of the fuel assemblies went above 1000 °C. With the aid of the radiative heat transfer model of the CFX-5.6 code, the surface temperatures were analyzed. When the cleaning instrument got opened the fuel assemblies suffered a serious thermal shock and the assemblies highly damaged. The post-incident thermo-hydraulic state inside the cleaning vessel was investigated with a rather complex CFX model. The uncertainties were decreased by a wide parameter study. The recovery work is planned to be started in the close future. The operators of the damaged fuel removing equipments will work standing on a platform which will be placed into the service shaft just above the surface of the coolant of decreased level. Protecting the workers against unnecessary personal doses is a very important task. In this situation, while the coolant is important part of the biological shielding, it is also a source of radiation due to the considerable amount of radioactive contamination dispersed into it. Therefore, the 3D distribution of the contamination in the service shaft was estimated for different operational and incidental scenarios with a wide parameter study made by a 3D CFX model. This comprehensive work performed with several models and calculations is tersely outlined according to the limited extent of the paper.Copyright