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Sample records for thermal hydraulic behaviors

  1. Hydrogen Behaviors Coupled with Thermal Hydraulics in APR1400 Containment during an SBLOCA

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Jongtae; Hong, Seong-Wan [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2016-10-15

    There are two hydrogen mitigation strategies. One is burning hydrogen by igniters as it is released in the containment. The other is using PARs (passive auto-catalytic recombiners) to remove the hydrogen. The strategy based on PAR installation requires well-mixing of the released hydrogen with steam in a containment atmosphere. It means that hydrogen distribution in a containment is very important for the hydrogen mitigation by PAR. Hydrogen behaviors in a NPP containment is strongly coupled with thermal hydraulics such as turbulent mixing, stratification by buoyancy, steam condensation, heat transfer et al. The purpose of this study is to investigate hydrogen behaviors in the APR1400 containment during an SBLOCA with a reactor core damage. The GASFLOW code is used to simulate hydrogen behaviors with thermal hydraulics in the APR1400 containment. In this paper, the hydrogen behaviors in the APR1400 containment during an SBLOCA were investigated by numerical simulation using GASFLOW. It was found that hydrogen mixture cloud may move downward relying on thermal hydraulic effect occurring in the containment. It is thought that condensation of steam included in the hydrogen mixture is very important in the hydrogen behaviors.

  2. Thermally Actuated Hydraulic Pumps

    Science.gov (United States)

    Jones, Jack; Ross, Ronald; Chao, Yi

    2008-01-01

    Thermally actuated hydraulic pumps have been proposed for diverse applications in which direct electrical or mechanical actuation is undesirable and the relative slowness of thermal actuation can be tolerated. The proposed pumps would not contain any sliding (wearing) parts in their compressors and, hence, could have long operational lifetimes. The basic principle of a pump according to the proposal is to utilize the thermal expansion and contraction of a wax or other phase-change material in contact with a hydraulic fluid in a rigid chamber. Heating the chamber and its contents from below to above the melting temperature of the phase-change material would cause the material to expand significantly, thus causing a substantial increase in hydraulic pressure and/or a substantial displacement of hydraulic fluid out of the chamber. Similarly, cooling the chamber and its contents from above to below the melting temperature of the phase-change material would cause the material to contract significantly, thus causing a substantial decrease in hydraulic pressure and/or a substantial displacement of hydraulic fluid into the chamber. The displacement of the hydraulic fluid could be used to drive a piston. The figure illustrates a simple example of a hydraulic jack driven by a thermally actuated hydraulic pump. The pump chamber would be a cylinder containing encapsulated wax pellets and containing radial fins to facilitate transfer of heat to and from the wax. The plastic encapsulation would serve as an oil/wax barrier and the remaining interior space could be filled with hydraulic oil. A filter would retain the encapsulated wax particles in the pump chamber while allowing the hydraulic oil to flow into and out of the chamber. In one important class of potential applications, thermally actuated hydraulic pumps, exploiting vertical ocean temperature gradients for heating and cooling as needed, would be used to vary hydraulic pressures to control buoyancy in undersea research

  3. MyrrhaFoam: A CFD model for the study of the thermal hydraulic behavior of MYRRHA

    Energy Technology Data Exchange (ETDEWEB)

    Koloszar, Lilla; Buckingham, Sophia; Planquart, Philippe [von Karman Institute, Chaussée de Waterloo 72, B-1640 Rhode-St-Genèse (Belgium); Keijers, Steven [SCK-CEN, Boeretang 200, 2400 Mol (Belgium)

    2017-02-15

    Highlights: • Development of a modeling approach for simulating the thermal hydraulics of heavy liquid metal nuclear reactors. • Detailed description of the modeling of each component through the MYRRHA reactor. • Detailed analysis of the flow field of the MYRRHA reactor under operating condition. • Assessment of the thermal load on the structures as well as the thermal stratification in the upper and the lower plenum. - Abstract: Numerical analysis of the thermohydraulic behavior of the innovative flexible fast spectrum research reactor, MYRRHA, under design by the Belgian Nuclear Research Center (SCK• CEN) is a very challenging task. The primary coolant of the reactor is Lead Bismuth Eutectic, LBE, which is an opaque heavy liquid metal with low Prandtl number. The simulation tool needs to involve many complex physical phenomena to be able to predict accurately the flow and thermal field in the pool type reactor. In the past few years, within the frame of a collaboration between SCK• CEN and the von Karman Institute, a new platform, MyrrhaFoam, was developed based on the open source simulation environment, OpenFOAM. The current tool can deal with incompressible buoyancy corrected steady/unsteady single phase flows. It takes into account conjugate heat transfer in the solid parts which is mandatory due to the expected high temperature gradients between the different parts of the reactor. The temperature dependent properties of LBE are also considered. MyrrhaFoam is supplemented with the most relevant thermal turbulence models for low Prandtl number liquids up to date.

  4. Thermal hydraulic behavior and efficiency analysis of an all-vanadium redox flow battery

    Science.gov (United States)

    Xiong, Binyu; Zhao, Jiyun; Tseng, K. J.; Skyllas-Kazacos, Maria; Lim, Tuti Mariana; Zhang, Yu

    2013-11-01

    Vanadium redox flow batteries (VRBs) are very competitive for large-capacity energy storage in power grids and in smart buildings due to low maintenance costs, high design flexibility, and long cycle life. Thermal hydraulic modeling of VRB energy storage systems is an important issue and temperature has remarkable impacts on the battery efficiency, the lifetime of material and the stability of the electrolytes. In this paper, a lumped model including auxiliary pump effect is developed to investigate the VRB temperature responses under different operating and surrounding environmental conditions. The impact of electrolyte flow rate and temperature on the battery electrical characteristics and efficiencies are also investigated. A one kilowatt VRB system is selected to conduct numerical simulations. The thermal hydraulic model is benchmarked with experimental data and good agreement is found. Simulation results show that pump power is sensitive to hydraulic design and flow rates. The temperature in the stack and tanks rises up about 10 °C under normal operating conditions for the stack design and electrolyte volume selected. An optimal flow rate of around 90 cm3 s-1 is obtained for the proposed battery configuration to maximize battery efficiency. The models developed in this paper can also be used for the development of a battery control strategy to achieve satisfactory thermal hydraulic performance and maximize energy efficiency.

  5. Assessment study of RELAP5/SCDAP capability to reproduce TALL facility thermal hydraulic behavior

    Energy Technology Data Exchange (ETDEWEB)

    Fiori, F., E-mail: filippofiori85@gmail.com [INET, Tsinghua University, Beijing 100084 (China); Institute of Nuclear and New Energy Technology, Collaborative Innovation Center of Advanced Nuclear Energy Technology, Key Laboratory of Advanced Reactor Engineering and Safety of Ministry of Education, Tsinghua University, Beijing 100084 (China); Zhou, Z.W. [INET, Tsinghua University, Beijing 100084 (China); Institute of Nuclear and New Energy Technology, Collaborative Innovation Center of Advanced Nuclear Energy Technology, Key Laboratory of Advanced Reactor Engineering and Safety of Ministry of Education, Tsinghua University, Beijing 100084 (China)

    2015-12-15

    Highlights: • We modified the RELAP5/SCDAP code to work with LBE and glycerol. • A RELAP5 model has been set up following preset choices and guidelines. • Seven different transients have been simulated. • Two different HTC correlation have been studied. - Abstract: The paper presents the assessment of RELAP5/SCDAP code capabilities to simulate the thermal–hydraulic behavior of liquid metal. The code has been recently modified to work with liquid metal; new heat transfer correlations have been implemented. As the code is widely used in our institute during the design process of the Chinese ADS reactor the assessment of the newly modified RELAP5/SCDAP is seen as a necessary step to ensure the quality of the code results. The present paper focuses on the simulation of the transients performed on the TALL facility. TALL has been constructed and operated at KTH Royal Institute of Technology of Stockholm. The full height facility was designed and operated to investigate the heat transfer performance of different heat exchangers and the thermal–hydraulic characteristics of natural and forced circulation flow under steady and transient conditions. Two different configurations are available for the TALL facility however only one is simulated for the present study with seven transients analyzed. Different LBE heat transfer correlations are compared for the calculations. A consistent and systematic approach for the nodalization development and assessment procedures that respond to the IAEA guidelines is discussed and thoroughly applied. The procedures and the database developed constitute the base in our institute for further study when more experimental data is made available.

  6. The R&D PERFROI Project on Thermal Mechanical and Thermal Hydraulics Behaviors of a Fuel Rod Assembly during a Loss of Coolant Accident

    Energy Technology Data Exchange (ETDEWEB)

    Repetto, G. [Institut de Radioprotection et de Surete Nucleaire, Cadarache (France); Dominguez, C. [Institut de Radioprotection et de Surete Nucleaire, Cadarache (France); Durville, B. [Institut de Radioprotection et de Surete Nucleaire, Cadarache (France); Carnemolla, S. [Institut de Radioprotection et de Surete Nucleaire, Cadarache (France); Campello, D. [Institut National des Sciences Appliques, Lyon (France); Tardiff, N. [Institut National des Sciences Appliques, Lyon (France); Gradeck, M. [Univ. de Lorraine, Nancy, France. LEMTA

    2015-09-04

    The safety principle in case of a LOCA is to preserve the short and long term coolability of the core. The associated safety requirements are to ensure the resistance of the fuel rods upon quench and post-quench loads and to maintain a coolable geometry in the core. An R&D program has been launched by IRSN with the support of EDF, to perform both experimental and modeling activities in the frame of the LOCA transient, on technical issues such as: - flow blockage within a fuel rods bundle and its potential impact on coolability, - fuel fragment relocation in the ballooned areas: its potential impact on cladding PCT (Peak Cladding Temperature) and on the maximum oxidation rate, - potential loss of cladding integrity upon quench and post-quench loads. The PERFROI project (2014-2019) focusing on the first above issue, is structured in two axes: 1. axis 1: thermal mechanical behavior of deformation and rupture of cladding taking into account the contact between fuel rods; specific research at LaMCoS laboratory focus on the hydrogen behavior in cladding alloys and its impact on the mechanical behavior of the rod; and, 2. axis 2: thermal hydraulics study of a partially blocked region of the core (ballooned area taking into account the fuel relocation with local over power), during cooling phase by water injection; More detailed activities foreseen in collaboration with LEMTA laboratory will focus on the characterization of two phase flows with heat transfer in deformed structures.

  7. Thermal hydraulic test for reactor safety system; a visualization study on flow boiling and bubble behavior

    Energy Technology Data Exchange (ETDEWEB)

    Chang, Soon Heung; Baek, Won Pil; Ban, In Cheol [Korea Advanced Institute of Science and Technology, Taejeon (Korea)

    2002-03-01

    The project contribute to understand and to clarify the physical mechanism of flow nucleate boiling and CHF phenomena through the visualization experiments. the results are useful in the development of the enhancement device of heat transfer and to enhance nuclear fuel safety 1. Visual experimental facility 2. Application method of visualization Technique 3. Visualization results of flow nucleate boiling regime - Overall Bubble Behavior on the Heated Surface - Bubble Behavior near CHF Condition - Identification of Flow Structure - Three-layer flow structure 4. Quantifying of bubble parameter through a digital image processing - Image Processing Techniques - Classification of objects and measurements of the size - Three dimensional surface plot with using the luminance 5. Development and estimation of a correlation between bubble diameter and flow parameter - The effect of system parameter on bubble diameter - The development of a bubble diameter correlation . 49 refs., 42 figs., 7 tabs. (Author)

  8. HANARO thermal hydraulic accident analysis

    Energy Technology Data Exchange (ETDEWEB)

    Park, Chul; Kim, Heon Il; Lee, Bo Yook; Lee, Sang Yong [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1996-06-01

    For the safety assessment of HANARO, accident analyses for the anticipated operational transients, accident scenarios and limiting accident scenarios were conducted. To do this, the commercial nuclear reactor system code. RELAP5/MOD2 was modified to RELAP5/KMRR; the thermal hydraulic correlations and the heat exchanger model was changed to incorporate HANARO characteristics. This report summarizes the RELAP/KMRR calculation results and the subchannel analyses results based on the RELAP/KMRR results. During the calculation, major concern was placed on the integrity of the fuel. For all the scenarios, the important accident analysis parameters, i.e., fuel centerline temperatures and the minimum critical heat flux ratio(MCHFR), satisfied safe design limits. It was verified, therefore, that the HANARO was safely designed. 21 tabs., 89 figs., 39 refs. (Author) .new.

  9. Application of RELAP5/MOD3.3 to Calculate Thermal Hydraulic Behavior of the Pressurizer Safety Valve Performance Test Facility

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Chang Hyun; Kim, Young Ae; Oh, Seung Jong; Park, Jong Woon [Korea Hydro and Nuclear Power Co., Ltd., Daejeon (Korea, Republic of)

    2007-10-15

    The increase of the acceptance tolerance of Pressurizer Safety Valve (PSV) test is vital for the safe operation of nuclear power plants because the frequent tests may make the valves decrepit and become a cause of leak. Recently, Korea Hydro and Nuclear Power Company (KHNP) is building a PSV performance test facility to provide the technical background data for the relaxation of the acceptance tolerance of PSV including the valve pop-up characteristics and the loop seal dynamics (if the plant has the loop seal in the upstream of PSV). The discharge piping and supports must be designed to withstand severe transient hydrodynamic loads when the safety valve actuates. The evaluation of hydrodynamic loads is a two-step process: first the thermal hydraulic behavior in the piping must be defined, and then the hydrodynamic loads are calculated from the thermal hydraulic parameters such as pressure and mass flow. The hydrodynamic loads are used as input to the structural analysis.

  10. Thermal-Hydraulics analysis of pressurized water reactor core by using single heated channel model

    Directory of Open Access Journals (Sweden)

    Reza Akbari

    2017-08-01

    Full Text Available Thermal hydraulics of nuclear reactor as a basis of reactor safety has a very important role in reactor design and control. The thermal-hydraulic analysis provides input data to the reactor-physics analysis, whereas the latter gives information about the distribution of heat sources, which is needed to perform the thermal-hydraulic analysis. In this study single heated channel model as a very fast model for predicting thermal hydraulics behavior of pressurized water reactor core has been developed. For verifying the results of this model, we used RELAP5 code as US nuclear regulatory approved thermal hydraulics code. The results of developed single heated channel model have been checked with RELAP5 results for WWER-1000. This comparison shows the capability of single heated channel model for predicting thermal hydraulics behavior of reactor core.

  11. Thermal-hydraulic analysis of nuclear reactors

    CERN Document Server

    Zohuri, Bahman

    2015-01-01

    This text covers the fundamentals of thermodynamics required to understand electrical power generation systems and the application of these principles to nuclear reactor power plant systems. It is not a traditional general thermodynamics text, per se, but a practical thermodynamics volume intended to explain the fundamentals and apply them to the challenges facing actual nuclear power plants systems, where thermal hydraulics comes to play.  Written in a lucid, straight-forward style while retaining scientific rigor, the content is accessible to upper division undergraduate students and aimed at practicing engineers in nuclear power facilities and engineering scientists and technicians in industry, academic research groups, and national laboratories. The book is also a valuable resource for students and faculty in various engineering programs concerned with nuclear reactors. This book also: Provides extensive coverage of thermal hydraulics with thermodynamics in nuclear reactors, beginning with fundamental ...

  12. Thermal Hydraulic Design of PWT Accelerating Structures

    CERN Document Server

    Yu, David; Chen Ping; Lundquist, Martin; Luo, Yan

    2005-01-01

    Microwave power losses on the surfaces of accelerating structures will transform to heat which will deform the structures if it is not removed in time. Thermal hydraulic design of the disk and cooling rods of a Plane Wave Transformer (PWT) structure is presented. Experiments to measure the hydraulic (pressure vs flow rate) and cooling (heat removed vs flow rate) properties of the PWT disk are performed, and results compared with simulations using Mathcad models and the COSMOSM code. Both experimental and simulation results showed that the heat deposited on the structure could be removed effectively using specially designed water-cooling circuits and the temperature of the structure could be controlled within the range required.

  13. Mercury Thermal Hydraulic Loop (MTHL) Summary Report

    Energy Technology Data Exchange (ETDEWEB)

    Felde, David K. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Crye, Jason Michael [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Wendel, Mark W. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Yoder, Jr, Graydon L. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Farquharson, George [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Jallouk, Philip A. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); McFee, Marshall T. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Pointer, William David [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Ruggles, Art E. [Univ. of Tennessee, Knoxville, TN (United States); Carbajo, Juan J. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2017-03-01

    The Spallation Neutron Source (SNS) is a high-power linear accelerator built at Oak Ridge National Laboratory (ORNL) which incorporates the use of a flowing liquid mercury target. The Mercury Thermal Hydraulic Loop (MTHL) was constructed to investigate and verify the heat transfer characteristics of liquid mercury in a rectangular channel. This report provides a compilation of previously reported results from the water-cooled and electrically heated straight and curved test sections that simulate the geometry of the window cooling channel in the target nose region.

  14. Thermal Hydraulic Tests for Reactor Core Safety

    Energy Technology Data Exchange (ETDEWEB)

    Moon, S. K.; Baek, W. P.; Chun, S. Y. (and others)

    2007-06-15

    The main objectives of the present project are to resolve the current issues of reactor core thermal hydraulics, to develop an advanced measurement and analytical techniques, and to perform reactor core safety verification tests. 6x6 reflood experiments, various heat transfer experiments using Freon, and experiments on the spacer grids effects on the post-dryout are carried out using spacer grids developed in Korea in order to resolve the current issues of the reactor core thermal hydraulics. In order to develop a reflood heat transfer model, the detailed reflood phenomena are visualized and measured using round tube and 2x2 rod bundle. A detailed turbulent mixing phenomenon for subchannels is measured using advanced measurement techniques such as LDV and PIV. MARS and MATRA codes developed in Korea are assessed, verified and improved using the obtained experimental data. Finally, a systematic quality assurance program and experimental data generation system has been constructed in order to increase the reliability of the experimental data.

  15. Multiphase flow dynamics 5 nuclear thermal hydraulics

    CERN Document Server

    Kolev, Nikolay Ivanov

    2015-01-01

    This Volume 5 of the successful book package "Multiphase Flow Dynamics" is devoted to nuclear thermal hydraulics which is a substantial part of nuclear reactor safety. It provides knowledge and mathematical tools for adequate description of the process of transferring the fission heat released in materials due to nuclear reactions into its environment. It step by step introduces into the heat release inside the fuel, temperature fields in the fuels, the "simple" boiling flow in a pipe described using ideas of different complexity like equilibrium, non equilibrium, homogeneity, non homogeneity. Then the "simple" three-fluid boiling flow in a pipe is described by gradually involving the mechanisms like entrainment and deposition, dynamic fragmentation, collisions, coalescence, turbulence. All heat transfer mechanisms are introduced gradually discussing their uncertainty. Different techniques are introduced like boundary layer treatments or integral methods. Comparisons with experimental data at each step demons...

  16. Multiphase Flow Dynamics 5 Nuclear Thermal Hydraulics

    CERN Document Server

    Kolev, Nikolay Ivanov

    2012-01-01

    The present Volume 5 of the successful book package "Multiphase Flow Dynamics" is devoted to nuclear thermal hydraulics which is a substantial part of nuclear reactor safety. It provides knowledge and mathematical tools for adequate description of the process of transferring the fission heat released in materials due to nuclear reactions into its environment. It step by step introduces into the heat release inside the fuel, temperature fields in the fuels, the "simple" boiling flow in a pipe described using ideas of different complexity like equilibrium, non equilibrium, homogeneity, non homogeneity. Then the "simple" three-fluid boiling flow in a pipe is described by gradually involving the mechanisms like entrainment and deposition, dynamic fragmentation, collisions, coalescence, turbulence. All heat transfer mechanisms are introduced gradually discussing their uncertainty. Different techniques are introduced like boundary layer treatments or integral methods. Comparisons with experimental data at each step...

  17. Assessment of the Thermal Hydraulic Models in THALES

    Energy Technology Data Exchange (ETDEWEB)

    Jang, Byeong Il; Kim, Hong Ju; Jang, Beomjun; Woo, Hae-Seuk [KEPCO Nuclear Fuel, Daejeon (Korea, Republic of)

    2016-10-15

    THALES (Thermal Hydraulic AnaLyzer for Enhanced Simulation of core) developed by KEPCO Nuclear Fuel is a subchannel analysis code on the basis of the single-stage core analysis model. THALES calculates the local fluid conditions and DNBR in the PWR (Pressurized Water Reactor) core. Currently, THALES is limited to the licensed type of the nuclear power plant because the thermal hydraulic models and CHF (Critical Heat Flux) correlations for OPR1000 and APR1400 are only licensed. KEPCO NF intends to apply THALES to WH typed nuclear power plants in Korea. To expand the applicable types of the nuclear power plants, the existing thermal hydraulic models were modified and new thermal hydraulic models were added to THALES. In this study, the thermal hydraulic models tested and added in THALES are reviewed and a preliminary calculation is performed. KEPCO NF intends to apply THALES to various typed nuclear power plants in Korea. So, the existing thermal hydraulic models implemented in THALES are modified and the void model which are generally used in the subchannel analysis code is added. Through the preliminary calculation, it is confirmed that the thermal hydraulic models are properly modified and implemented in THALES, which shows the possibility to apply THALES in various typed nuclear power plants in Korea.

  18. Investigation of feedback on neutron kinetics and thermal hydraulics from detailed online fuel behavior modeling during a boron dilution transient in a PWR with the two-way coupled code system DYN3D-TRANSURANUS

    Energy Technology Data Exchange (ETDEWEB)

    Holt, L., E-mail: lars.holt@tuev-sued.de [TÜV SÜD Energietechnik GmbH Baden-Württemberg, Gottlieb-Daimler-Str. 7, 70794 Filderstadt (Germany); Technical University München, Department of Nuclear Engineering, Boltzmannstr. 15, D-85748 Garching bei München (Germany); Rohde, U.; Kliem, S.; Baier, S. [Helmholtz-Zentrum Dresden—Rossendorf, Reactor Safety Division, PO Box 510119, D-01314 Dresden (Germany); Seidl, M. [E.ON Kernkraft GmbH, Tresckowstr. 5, D-30457 Hannover (Germany); Van Uffelen, P. [European Commission, Joint Research Centre, Institute for Transuranium Elements, Hermann-von-Helmholtz-Platz 1, D-76344 Eggenstein-Leopoldshafen (Germany); Macián-Juan, R. [Technical University München, Department of Nuclear Engineering, Boltzmannstr. 15, D-85748 Garching bei München (Germany)

    2016-02-15

    Highlights: • General coupling interface was developed for the fuel performance code TRANSURANUS. • With this new tool simplified fuel behavior models in codes can be replaced. • The reactor dynamics code DYN3D was coupled to TRANSURANUS at assembly level. • The feedback from detailed online fuel behavior modeling is analyzed for reactivity initiated accident (RIA). • The thermal hydraulics can be affected strongly even in fresh fuel assemblies. - Abstract: Recently the reactor dynamics code DYN3D (including an internal fuel behavior model) was coupled to the fuel performance code TRANSURANUS at assembly level. The coupled code system applies the new general TRANSURANUS coupling interface, hence it can be used for one-way or two-way coupling. In the coupling, DYN3D provides process time, time-dependent rod power and thermal hydraulics conditions to TRANSURANUS, which in case of the two-way coupling approach replaces completely the internal DYN3D fuel behavior model and transfers parameters like radial fuel temperature distribution and cladding temperature back to DYN3D. For the first time results of the coupled code system are presented for a post-critical-heat-flux heat transfer. The corresponding heat transfer regime is mostly film boiling, where the cladding temperature can rise several hundreds of degrees. The simulated boron dilution transient assumed an injection of a 36 m{sup 3} slug of under-borated coolant into a German pressurized water reactor (PWR) core initiated from a sub-critical reactor state (extreme reactivity initiated accident (RIA) conditions). The feedback from detailed fuel behavior modeling was found negligible on the neutron kinetics and thermal hydraulics during the first power rise. In a later phase of the transient, the node injected energy can differ 25 J/g, even still around 20 J/g for nodes without film boiling. Furthermore, the thermal hydraulics can be affected strongly even in fresh fuel assemblies, where film boiling

  19. Study on thermal-hydraulics during a PWR reflood phase

    Energy Technology Data Exchange (ETDEWEB)

    Iguchi, Tadashi [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    1998-10-01

    In-core thermal-hydraulics during a PWR reflood phase following a large-break LOCA are quite unique in comparison with two-phase flow which has been studied widely in previous researches, because the geometry of the flow path is complicated (bundle geometry) and water is at extremely low superficial velocity and almost under stagnant condition. Hence, some phenomena realized during a PWR reflood phase are not understood enough and appropriate analytical models have not been developed, although they are important in a viewpoint of reactor safety evaluation. Therefore, author investigated some phenomena specified as important issues for quantitative prediction, i.e. (1) void fraction in a bundle during a PWR reflood phase, (2) effect of radial core power profile on reflood behavior, (3) effect of combined emergency core coolant injection on reflood behavior, and (4) the core separation into two thermal-hydraulically different regions and the in-core flow circulation behavior observed during a combined injection PWR reflood phase. Further, author made analytical models for these specified issues, and succeeded to predict reflood behaviors at representative types of PWRs, i.e.cold leg injection PWRs and Combined injection PWRs, in good accuracy. Above results were incorporated into REFLA code which is developed at JAERI, and they improved accuracy in prediction and enlarged applicability of the code. In the present study, models were intended to be utilized in a practical use, and hence these models are simplified ones. However, physical understanding on the specified issues in the present study is basic and principal for reflood behavior, and then it is considered to be used in a future advanced code development and improvement. (author). 110 refs.

  20. Process management using component thermal-hydraulic function classes

    Science.gov (United States)

    Morman, James A.; Wei, Thomas Y. C.; Reifman, Jaques

    1999-01-01

    A process management expert system where following malfunctioning of a component, such as a pump, for determining system realignment procedures such as for by-passing the malfunctioning component with on-line speeds to maintain operation of the process at full or partial capacity or to provide safe shut down of the system while isolating the malfunctioning component. The expert system uses thermal-hydraulic function classes at the component level for analyzing unanticipated as well as anticipated component malfunctions to provide recommended sequences of operator actions. Each component is classified according to its thermal-hydraulic function, and the generic and component-specific characteristics for that function. Using the diagnosis of the malfunctioning component and its thermal hydraulic class, the expert system analysis is carried out using generic thermal-hydraulic first principles. One aspect of the invention employs a qualitative physics-based forward search directed primarily downstream from the malfunctioning component in combination with a subsequent backward search directed primarily upstream from the serviced component. Generic classes of components are defined in the knowledge base according to the three thermal-hydraulic functions of mass, momentum and energy transfer and are used to determine possible realignment of component configurations in response to thermal-hydraulic function imbalance caused by the malfunctioning component. Each realignment to a new configuration produces the accompanying sequence of recommended operator actions. All possible new configurations are examined and a prioritized list of acceptable solutions is produced.

  1. Thermal Hydraulic Integral Effect Tests for Pressurized Water Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Baek, W. P.; Song, C. H.; Kim, Y. S. and others

    2005-02-15

    The objectives of the project are to construct a thermal-hydraulic integral effect test facility and to perform various integral effect tests for design, operation, and safety regulation of pressurized water reactors. During the first phase of this project (1997.8{approx}2002.3), the basic technology for thermal-hydraulic integral effect tests was established and the basic design of the test facility was accomplished: a full-height, 1/300-volume-scaled full pressure facility for APR1400, an evolutionary pressurized water reactor that was developed by Korean industry. Main objectives of the present phase (2002.4{approx}2005.2), was to optimize the facility design and to construct the experimental facility. We have performed following researches: 1) Optimization of the basic design of the thermal-hydraulic integral effect test facility for PWRs - ATLAS (Advanced Thermal-hydraulic Test Loop for Accident Simulation) - Reduced height design for APR1400 (+ specific design features of KSNP safety injection systems) - Thermal-hydraulic scaling based on three-level scaling methodology by Ishii et al. 2) Construction of the ATLAS facility - Detailed design of the test facility - Manufacturing and procurement of components - Installation of the facility 3) Development of supporting technology for integral effect tests - Development and application of advanced instrumentation technology - Preliminary analysis of test scenarios - Development of experimental procedures - Establishment and implementation of QA system/procedure.

  2. Horizontal steam generator PGV-1000 thermal-hydraulic analysis

    Energy Technology Data Exchange (ETDEWEB)

    Ubra, O. [Skoda Company, Prague (Switzerland); Doubek, M. [Czech Technical Univ., Prague (Switzerland)

    1995-12-31

    A computer program for the steady state thermal-hydraulic analysis of horizontal steam generator PGV-1000 is presented. The program provides the capability to analyze steam generator PGV-1000 primary side flow and temperature distribution, primary side pressure drops, heat transfer between the primary and secondary sides and multidimensional heat flux distribution. A special attention is paid to the thermal-hydraulics of the secondary side. The code predicts 3-D distribution of the void fraction at the secondary side, mass redistribution under the submerged perforated sheet and the steam generator level profile. By means of developed computer program a detailed thermal-hydraulic study of the PGV-1000 has been carried out. A wide range of calculations has been performed and a set of important steam generator characteristics has been obtained. Some of them are presented in the paper. (orig.). 5 refs.

  3. Current and anticipated uses of thermal hydraulic codes at the Japan Atomic Energy Research Institute

    Energy Technology Data Exchange (ETDEWEB)

    Akimoto, Hajime; Kukita; Ohnuki, Akira [Japan Atomic Energy Research Institute, Ibaraki (Japan)

    1997-07-01

    The Japan Atomic Energy Research Institute (JAERI) is conducting several research programs related to thermal-hydraulic and neutronic behavior of light water reactors (LWRs). These include LWR safety research projects, which are conducted in accordance with the Nuclear Safety Commission`s research plan, and reactor engineering projects for the development of innovative reactor designs or core/fuel designs. Thermal-hydraulic and neutronic codes are used for various purposes including experimental analysis, nuclear power plant (NPP) safety analysis, and design assessment.

  4. Current and anticipated uses of thermal hydraulic codes in Korea

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Kyung-Doo; Chang, Won-Pyo [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1997-07-01

    In Korea, the current uses of thermal hydraulic codes are categorized into 3 areas. The first application is in designing both nuclear fuel and NSSS. The codes have usually been introduced based on the technology transfer programs agreed between KAERI and the foreign vendors. Another area is in the supporting of the plant operations and licensing by the utility. The third category is research purposes. In this area assessments and some applications to the safety issue resolutions are major activities using the best estimate thermal hydraulic codes such as RELAP5/MOD3 and CATHARE2. Recently KEPCO plans to couple thermal hydraulic codes with a neutronics code for the design of the evolutionary type reactor by 2004. KAERI also plans to develop its own best estimate thermal hydraulic code, however, application range is different from KEPCO developing code. Considering these activities, it is anticipated that use of the best estimate hydraulic analysis code developed in Korea may be possible in the area of safety evaluation within 10 years.

  5. Development of coupled neutronics/thermal-hydraulics test case for HPLWR

    Science.gov (United States)

    Pham, P.; Gamtsemlidze, I. D.; Bahdanovich, R. B.; Nikonov, S. P.; Smirnov, A. D.

    2017-01-01

    The High-Performance Light Water Reactor (HPLWR) is the European concept of a supercritical water reactor (SCWR) which is one of the most promising and innovative designs of the Generation IV nuclear reactor concepts. The thermal-hydraulics behavior of supercritical water is significantly different from water at sub-critical pressure because of the difference in the specific heat value. Coupled analysis of HPLWR assembly neutronics and thermal-hydraulics has become important because of the strong influence of the water density on the neutron spectrum and power distribution. Programs MCU (Monte-Carlo Universal) and ATHLET (Analysis of Thermal-hydraulics of Leaks and Transients) were used for better estimation of power and temperature distribution in HPLWR assembly.

  6. HELOKA-HP thermal-hydraulic model validation and calibration

    Energy Technology Data Exchange (ETDEWEB)

    Jin, Xue Zhou; Ghidersa, Bradut-Eugen; Badea, Aurelian Florin

    2016-11-01

    Highlights: • The electrical heater in HELOKA-HP has been modeled with RELAP5-3D using experimental data as input. • The model has been validated using novel techniques for assimilating experimental data and the representative model parameters with BEST-EST. • The methodology is successfully used for reducing the model uncertainties and provides a quantitative measure of the consistency between the experimental data and the model. - Abstract: The Helium Loop Karlsruhe High Pressure (HELOKA-HP) is an experimental facility for the testing of various helium-cooled components at high temperature (500 °C) and high pressure (8 MPa) for nuclear fusion applications. For modeling the loop thermal dynamics, a thermal-hydraulic model has been created using the system code RELAP5-3D. Recently, new experimental data covering the behavior of the loop components under relevant operational conditions have been made available giving the possibility of validating and calibrating the existing models in order to reduce the uncertainties of the simulated responses. This paper presents an example where such process has been applied for the HELOKA electrical heater model. Using novel techniques for assimilating experimental data, implemented in the computational module BEST-EST, the representative parameters of the model have been calibrated.

  7. Characterization of thermal, hydraulic, and gas diffusion properties in variably saturated sand grades

    DEFF Research Database (Denmark)

    Deepagoda Thuduwe Kankanamge Kelum, Chamindu; Smits, Kathleen; Ramirez, Jamie

    2016-01-01

    porous media transport properties, key transport parameters such as thermal conductivity and gas diffusivity are particularly important to describe temperature-induced heat transport and diffusion-controlled gas transport processes, respectively. Despite many experimental and numerical studies focusing....../70) in relation to physical properties, water retention, hydraulic conductivity, thermal conductivity, and gas diffusivity. We used measured basic properties and transport data to accurately parameterize the characteristic functions (particle- and pore-size distributions and water retention) and descriptive...... transport models (thermal conductivity, saturated hydraulic conductivity, and gas diffusivity). An existing thermal conductivity model was improved to describe the distinct three-region behavior in observed thermal conductivity–water saturation relations. Applying widely used parametric models for saturated...

  8. Development of fuel performance and thermal hydraulic technology

    Energy Technology Data Exchange (ETDEWEB)

    Jung, Youn Ho; Song, K. N.; Kim, H. K. and others

    2000-03-01

    Space grid in LWR fuel assembly is a key structural component to support fuel rods and to enhance heat transfer from fuel rod to the coolant. Therefore, the original spacer grid has been developed. In addition, new phenomena in fuel behavior occurs at the high burnup, so that models to analyze those new phenomena were developed. Results of this project can be summarized as follows. - Seven different spacer grid candidates have been invented and submitted for domestic and US patents. Spacer grid test specimen(3x3 array and 5x5 array) were fabricated for each candidate and the mechanical tests were performed. - Basic technologies in the mechanical and thermal hydraulic behavior in the spacer grid development are studied and relevant test facilities were established - Fuel performance analysis models and programs were developed for the high burnup pellet and cladding, and fuel performance data base were compiled - Procedures of fuel characterization and in-/out of-pile tests were prepared - Conceptual design of fuel rod for integral PWR was carried out. (author)

  9. Thermal and hydraulic analyses of the System 81 cold traps

    Energy Technology Data Exchange (ETDEWEB)

    Kim, K.

    1977-06-15

    Thermal and hydraulic analyses of the System 81 Type I and II cold traps were completed except for thermal transients analysis. Results are evaluated, discussed, and reported. Analytical models were developed to determine the physical dimensions of the cold traps and to predict the performance. The FFTF cold trap crystallizer performances were simulated using the thermal model. This simulation shows that the analytical model developed predicts reasonably conservative temperatures. Pressure drop and sodium residence time calculations indicate that the present design will meet the requirements specified in the E-Specification. Steady state temperature data for the critical regions were generated to assess the magnitude of the thermal stress.

  10. FONESYS: The FOrum and NEtwork of SYStem Thermal-Hydraulic Codes in Nuclear Reactor Thermal-Hydraulics

    Energy Technology Data Exchange (ETDEWEB)

    Ahn, S.H., E-mail: k175ash@kins.re.kr [Korea Institute of Nuclear Safety (KINS) (Korea, Republic of); Aksan, N., E-mail: nusr.aksan@gmail.com [University of Pisa San Piero a Grado Nuclear Research Group (GRNSPG) (Italy); Austregesilo, H., E-mail: henrique.austregesilo@grs.de [Gesellschaft für Anlagen- und Reaktorsicherheit (GRS) (Germany); Bestion, D., E-mail: dominique.bestion@cea.fr [Commissariat à l’énergie atomique et aux énergies alternatives (CEA) (France); Chung, B.D., E-mail: bdchung@kaeri.re.kr [Korea Atomic Energy Research Institute (KAERI) (Korea, Republic of); D’Auria, F., E-mail: f.dauria@ing.unipi.it [University of Pisa San Piero a Grado Nuclear Research Group (GRNSPG) (Italy); Emonot, P., E-mail: philippe.emonot@cea.fr [Commissariat à l’énergie atomique et aux énergies alternatives (CEA) (France); Gandrille, J.L., E-mail: jeanluc.gandrille@areva.com [AREVA NP (France); Hanninen, M., E-mail: markku.hanninen@vtt.fi [VTT Technical Research Centre of Finland (VTT) (Finland); Horvatović, I., E-mail: i.horvatovic@ing.unipi.it [University of Pisa San Piero a Grado Nuclear Research Group (GRNSPG) (Italy); Kim, K.D., E-mail: kdkim@kaeri.re.kr [Korea Atomic Energy Research Institute (KAERI) (Korea, Republic of); Kovtonyuk, A., E-mail: a.kovtonyuk@ing.unipi.it [University of Pisa San Piero a Grado Nuclear Research Group (GRNSPG) (Italy); Petruzzi, A., E-mail: a.petruzzi@ing.unipi.it [University of Pisa San Piero a Grado Nuclear Research Group (GRNSPG) (Italy)

    2015-01-15

    Highlights: • We briefly presented the project called Forum and Network of System Thermal-Hydraulics Codes in Nuclear Reactor Thermal-Hydraulics (FONESYS). • We presented FONESYS participants and their codes. • We explained FONESYS projects motivation, its main targets and working modalities. • We presented FONESYS position about projects topics and subtopics. - Abstract: The purpose of this article is to present briefly the project called Forum and Network of System Thermal-Hydraulics Codes in Nuclear Reactor Thermal-Hydraulics (FONESYS), its participants, the motivation for the project, its main targets and working modalities. System Thermal-Hydraulics (SYS-TH) codes, also as part of the Best Estimate Plus Uncertainty (BEPU) approaches, are expected to achieve a more-and-more relevant role in nuclear reactor technology, safety and design. Namely, the number of code-users can easily be predicted to increase in the countries where nuclear technology is exploited. Thus, the idea of establishing a forum and a network among the code developers and with possible extension to code users has started to have major importance and value. In this framework the FONESYS initiative has been created. The main targets of FONESYS are: • To promote the use of SYS-TH Codes and the application of the BEPU approaches. • To establish acceptable and recognized procedures and thresholds for Verification and Validation (V and V). • To create a common ground for discussing envisaged improvements in various areas, including user-interface, and the connection with other numerical tools, including Computational Fluid Dynamics (CFD) Codes.

  11. Thermal hydraulic feasibility assessment for the Spent Nuclear Fuel Project

    Energy Technology Data Exchange (ETDEWEB)

    Heard, F.J.; Cramer, E.R.; Beaver, T.R. [Westinghouse Hanford Co., Richland, WA (United States); Thurgood, M.J. [Marvin (John), Inc. (United States)

    1996-01-01

    A series of scoping analyses have been completed investigating the thermal-hydraulic performance and feasibility of the Spent Nuclear Fuel Project (SNFP) Integrated Process Strategy (IPS). The SNFP was established to develop engineered solutions for the expedited removal, stabilization, and storage of spent nuclear fuel from the K Basins at the U.S. Department of Energy`s Hanford Site in Richland, Washington. The subject efforts focused on independently investigating, quantifying, and establishing the governing heat production and removal mechanisms for each of the IPS operations and configurations, obtaining preliminary results for comparison with and verification of other analyses, and providing technology-based recommendations for consideration and incorporation into the design bases for the SNFP. The goal was to develop a series fo thermal-hydraulic models that could respond to all process and safety-related issues that may arise pertaining to the SNFP. A series of sensitivity analyses were also performed to help identify those parameters that have the greatest impact on energy transfer and hence, temperature control. It is anticipated that the subject thermal-hydraulic models will form the basis for a series of advanced and more detailed models that will more accurately reflect the thermal performance of the IPS and alleviate the necessity for some of the more conservative assumptions and oversimplifications, as well as form the basis for the final process and safety analyses.

  12. Thermal hydraulic model descrition of TASS/SMR

    Energy Technology Data Exchange (ETDEWEB)

    Yoon, Han Young; Kim, H. C.; Chung, Y. J.; Lim, H. S.; Yang, S. H

    2001-04-01

    The TASS/SMR code has been developed for the safety analysis of SMART. The governing equations were applied only to the primary coolant system in TASS which had been developed at KAERI. In TASS/SMR, the solution method is improved so that the primary and secondary coolant systems are solved simultaneously. Besides the solution method, thermal-hydraulic models are incorporated, in TASS/SMR, such as non-condensible gas model, helical steam generator heat transfer model, and passive residual heat removal system (PRHRS) heat transfer model for the application to SMART. The governing equtions of TASS/SMR are based on the drift-flux model so that the accidents and transients accompaning with two-phase flow can be analized. This report describes the governing equations and solution methods used in TASS/SMR and also includes the description for the thermal hydraulic models for SMART design.

  13. Thermal-hydraulic modeling of reactivity accidents in MTR reactors

    Directory of Open Access Journals (Sweden)

    Khater Hany

    2006-01-01

    Full Text Available This paper describes the development of a dynamic model for the thermal-hydraulic analysis of MTR research reactors during a reactivity insertion accident. The model is formulated for coupling reactor kinetics with feedback reactivity and reactor core thermal-hydraulics. To represent the reactor core, two types of channels are considered, average and hot channels. The developed computer program is compiled and executed on a personal computer, using the FORTRAN language. The model is validated by safety-related benchmark calculations for MTR-TYPE reactors of IAEA 10 MW generic reactor for both slow and fast reactivity insertion transients. A good agreement is shown between the present model and the benchmark calculations. Then, the model is used for simulating the uncontrolled withdrawal of a control rod of an ETRR-2 reactor in transient with over power scram trip. The model results for ETRR-2 are analyzed and discussed.

  14. Thermal-hydraulic interfacing code modules for CANDU reactors

    Energy Technology Data Exchange (ETDEWEB)

    Liu, W.S.; Gold, M.; Sills, H. [Ontario Hydro Nuclear, Toronto (Canada)] [and others

    1997-07-01

    The approach for CANDU reactor safety analysis in Ontario Hydro Nuclear (OHN) and Atomic Energy of Canada Limited (AECL) is presented. Reflecting the unique characteristics of CANDU reactors, the procedure of coupling the thermal-hydraulics, reactor physics and fuel channel/element codes in the safety analysis is described. The experience generated in the Canadian nuclear industry may be useful to other types of reactors in the areas of reactor safety analysis.

  15. Views on the future of thermal hydraulic modeling

    Energy Technology Data Exchange (ETDEWEB)

    Ishii, M. [Purdue Univ., West Lafayette, IN (United States)

    1997-07-01

    It is essential for the U.S. NRC to sustain the highest level of the thermal-hydraulics and reactor safety research expertise and continuously improve their accident analysis capability. Such expertise should span over four different areas which are strongly related to each other. These are: (1) Reactor Safety Code Development, (2) Two-phase Flow Modeling, (3) Instrumentation and Fundamental Experimental Research, and (4) Separate Effect and Integral Test. The NRC is already considering a new effort in the area of advanced thermal-hydraulics effort. Its success largely depends on the availability of a significantly improved two-phase flow formulation and constitutive relations supported by detailed experimental data. Therefore, it is recommended that the NRC start significant research efforts in the areas of two-phase flow modeling, instrumentation, basic and separate effect experiments which should be pursued systematically and with clearly defined objectives. It is desirable that some international program is developed in this area. This paper is concentrated on those items in the thermal-hydraulic area which eventually determine the quality of future accident analysis codes.

  16. Quantification of LOCA core damage frequency based on thermal-hydraulics analysis

    Energy Technology Data Exchange (ETDEWEB)

    Cho, Jaehyun, E-mail: chojh@kaeri.re.kr; Park, Jin Hee; Kim, Dong-San; Lim, Ho-Gon

    2017-04-15

    Highlights: • We quantified the LOCA core damage frequency based on the best-estimated success criteria analysis. • The thermal-hydraulic analysis using MARS code has been applied to Korea Standard Nuclear Power Plants. • Five new event trees with new break size boundaries and new success criteria were developed. • The core damage frequency is 5.80E−07 (/y), which is 12% less than the conventional PSA event trees. - Abstract: A loss-of-coolant accident (LOCA) has always been significantly considered one of the most important initiating events. However, most probabilistic safety assessment models, up to now, have undoubtedly adopted the three groups of LOCA, and even an exact break size boundary that used in WASH-1400 reports was published in 1975. With an awareness of the importance of a realistic PSA for a risk-informed application, several studies have tried to find the realistic thermal-hydraulic behavior of a LOCA, and improve the PSA model. The purpose of this research is to obtain realistic results of the LOCA core damage frequency based on a success criteria analysis using the best-estimate thermal-hydraulics code. To do so, the Korea Standard Nuclear Power Plant (KSNP) was selected for this study. The MARS code was used for a thermal hydraulics analysis and the AIMS code was used for the core damage quantification. One of the major findings in the thermal hydraulics analysis was that the decay power is well removed by only a normal secondary cooling in LOCAs of below 1.4 in and by only a high pressure safety injection in LOCAs of 0.8–9.4 in. Based on the thermal hydraulics results regarding new break size boundaries and new success criteria, five new event trees (ETs) were developed. The core damage frequency of new LOCA ETs is 5.80E−07 (/y), which is 12% less than the conventional PSA ETs. In this research, we obtained not only thermal-hydraulics characteristics for the entire break size of a LOCA in view of the deterministic safety

  17. A THERMAL-HYDRAULIC SYSTEM FOR THE CONVERSION AND THE STORAGE OF ENERGY

    OpenAIRE

    MITRAN Tudor; CHIOREANU Nicolae; ABAITANCAI Horia; RUS Alexandru

    2016-01-01

    The paper proposes the concept design of a thermal-hydraulic system that converts the thermal energy (from the geothermal water, from the cooling water of power equipment, from exhaust gasses, and so.) in hydrostatic energy, that is stored in a hydraulic accumulator. The hydraulic energy can be converted into electrical energy when needed.

  18. A THERMAL-HYDRAULIC SYSTEM FOR THE CONVERSION AND THE STORAGE OF ENERGY

    Directory of Open Access Journals (Sweden)

    MITRAN Tudor

    2016-05-01

    Full Text Available The paper proposes the concept design of a thermal-hydraulic system that converts the thermal energy (from the geothermal water, from the cooling water of power equipment, from exhaust gasses, and so. in hydrostatic energy, that is stored in a hydraulic accumulator. The hydraulic energy can be converted into electrical energy when needed.

  19. Neutronic and thermal-hydraulic coupling for 3D reactor core modeling combining MCB and fluent

    Directory of Open Access Journals (Sweden)

    Królikowski Igor P.

    2015-09-01

    Full Text Available Three-dimensional simulations of neutronics and thermal hydraulics of nuclear reactors are a tool used to design nuclear reactors. The coupling of MCB and FLUENT is presented, MCB allows to simulate neutronics, whereas FLUENT is computational fluid dynamics (CFD code. The main purpose of the coupling is to exchange data such as temperature and power profile between both codes. Temperature required as an input parameter for neutronics is significant since cross sections of nuclear reactions depend on temperature. Temperature may be calculated in thermal hydraulics, but this analysis needs as an input the power profile, which is a result from neutronic simulations. Exchange of data between both analyses is required to solve this problem. The coupling is a better solution compared to the assumption of estimated values of the temperatures or the power profiles; therefore the coupled analysis was created. This analysis includes single transient neutronic simulation and several steady-state thermal simulations. The power profile is generated in defined points in time during the neutronic simulation for the thermal analysis to calculate temperature. The coupled simulation gives information about thermal behavior of the reactor, nuclear reactions in the core, and the fuel evolution in time. Results show that there is strong influence of neutronics on thermal hydraulics. This impact is stronger than the impact of thermal hydraulics on neutronics. Influence of the coupling on temperature and neutron multiplication factor is presented. The analysis has been performed for the ELECTRA reactor, which is lead-cooled fast reactor concept, where the coolant fl ow is generated only by natural convection

  20. Research on the fundamental process of thermal-hydraulic behaviors in severe accident. Estimation of trigger condition for vapor explosion. JAERI's nuclear research promotion program, H10-027-1. Contract research

    Energy Technology Data Exchange (ETDEWEB)

    Nariai, Hideki [Tsukuba Univ., Institute of Engineering Mechanics and Systems, Tsukuba, Ibaraki (Japan)

    2002-03-01

    The experimental and analytical researches were conducted to study melted core material and coolant interaction including solidification and vapor explosion which is one of the most unidentified thermal hydraulic phenomena during sever accident of nuclear reactor. At first, the effect of the material properties on vapor explosion and solidification was examined to clarify the dominant factors for the spontaneous vapor explosion. Next, the interfacial phenomena of the high temperature melt material and violent boiling behavior of water at the interface was visually observed in the experiment. The interfacial phenomena were physically modeled. Finally, trigger phenomena from liquid-liquid contact to atomization were clarified through the forced collapse experiment of vapor film around a molten droplet by using pressure wave generation device. It is indicated by applying the results obtained in the present study to the actual reactor conditions that the possibility of the vapor explosion is extremely unlikely in the actual reactor accident sequence, since the surface of the molten uranium oxide is solidified in the water and the liquid-liquid contact can not be achieved. It should be noted that the decrease of the solidified temperature by metal compounds and the increase of the molten core temperature. (author)

  1. 78 FR 8202 - Meeting of the Joint ACRS Subcommittees on Thermal Hydraulic Phenomena and Materials, Metallurgy...

    Science.gov (United States)

    2013-02-05

    ... Hydraulic Phenomena and Materials, Metallurgy and Reactor Fuels; Notice of Meeting The Joint ACRS Subcommittees on Thermal Hydraulic Phenomena and Materials, Metallurgy and Reactor Fuels will hold a meeting on...

  2. Review of the nuclear reactor thermal hydraulic research in ocean motions

    Energy Technology Data Exchange (ETDEWEB)

    Yan, B.H., E-mail: yanbh3@mail.sysu.edu.cn

    2017-03-15

    The research and development of small modular reactor in floating platform has been strongly supported by Chinese government and enterprises. Due to the effect of ocean waves, the thermal hydraulic behavior and safety characteristics of floating reactor are different from that of land-based reactor. Many scholars including the author have published their research and results in open literatures. Much of these literatures are valuable but there are also some contradictory conclusions. In this wok, the nuclear reactor thermal hydraulic research in ocean motions was systematically summarized. Valuable results and experimental data were analyzed and classified. Inherent mechanism for controversial issues in different experiments was explained. Necessary work needed in the future was suggested. Through this work, we attempt to find as many valuable results as possible for the designing and subsequent research.

  3. INL Experimental Program Roadmap for Thermal Hydraulic Code Validation

    Energy Technology Data Exchange (ETDEWEB)

    Glenn McCreery; Hugh McIlroy

    2007-09-01

    Advanced computer modeling and simulation tools and protocols will be heavily relied on for a wide variety of system studies, engineering design activities, and other aspects of the Next Generation Nuclear Power (NGNP) Very High Temperature Reactor (VHTR), the DOE Global Nuclear Energy Partnership (GNEP), and light-water reactors. The goal is for all modeling and simulation tools to be demonstrated accurate and reliable through a formal Verification and Validation (V&V) process, especially where such tools are to be used to establish safety margins and support regulatory compliance, or to design a system in a manner that reduces the role of expensive mockups and prototypes. Recent literature identifies specific experimental principles that must be followed in order to insure that experimental data meet the standards required for a “benchmark” database. Even for well conducted experiments, missing experimental details, such as geometrical definition, data reduction procedures, and manufacturing tolerances have led to poor Benchmark calculations. The INL has a long and deep history of research in thermal hydraulics, especially in the 1960s through 1980s when many programs such as LOFT and Semiscle were devoted to light-water reactor safety research, the EBRII fast reactor was in operation, and a strong geothermal energy program was established. The past can serve as a partial guide for reinvigorating thermal hydraulic research at the laboratory. However, new research programs need to fully incorporate modern experimental methods such as measurement techniques using the latest instrumentation, computerized data reduction, and scaling methodology. The path forward for establishing experimental research for code model validation will require benchmark experiments conducted in suitable facilities located at the INL. This document describes thermal hydraulic facility requirements and candidate buildings and presents examples of suitable validation experiments related

  4. Neutronic and thermal-hydraulic analysis of fission molybdenum-99 production at Tehran Research Reactor using LEU plate targets.

    Science.gov (United States)

    Abedi, Ebrahim; Ebrahimkhani, Marzieh; Davari, Amin; Mirvakili, Seyed Mohammad; Tabasi, Mohsen; Maragheh, Mohammad Ghannadi

    2016-12-01

    Efficient and safe production of molybdenum-99 ( 99 Mo) radiopharmaceutical at Tehran Research Reactor (TRR) via fission of LEU targets is studied. Neutronic calculations are performed to evaluate produced 99 Mo activity, core neutronic safety parameters and also the power deposition values in target plates during a 7 days irradiation interval. Thermal-hydraulic analysis has been also carried out to obtain thermal behavior of these plates. Using Thermal-hydraulic analysis, it can be concluded that the safety parameters are satisfied in the current study. Consequently, the present neutronic and thermal-hydraulic calculations show efficient 99 Mo production is accessible at significant activity values in TRR current core configuration. Copyright © 2016 Elsevier Ltd. All rights reserved.

  5. A Study on thermal-hydraulic characteristics of the coolant materials for the transmutation reactor

    Energy Technology Data Exchange (ETDEWEB)

    Chung, Chang Hyun; You, Young Woo; Cho, Jae Seon; Kim, Ju Youl; Kim, Do Hyoung; Kim, Yoon Ik; Yang, Hui Chang [Seoul National University, Taejon (Korea)

    1998-03-01

    The objective of this study is to provide the direction of transmutation reactor design in terms of thermal hydraulics especially through the analysis of thermal hydraulic characteristics of various candidate materials for the transmutation reactor coolant. In this study, the characteristics of coolant materials used in current nuclear power plants and candidate materials for transmutation reactor are analyzed and compared. To evaluate the thermal hydraulic characteristics, the preliminary thermal-hydraulic calculation is performed for the candidate coolant materials of transmutation reactor. An analysis of thermal-hydraulic characteristics of transmutation reactor. An analysis of thermal-hydraulic characteristics of Sodium, Lead, Lead-Bismuth, and Lead-Lithium among the liquid metals considered as the coolant of transmutation reactor is performed by using computational fluid dynamics code FLUENT, and SIMPLER algorithm. (author). 50 refs., 40 figs., 30 tabs.

  6. Thermal-hydraulic modeling needs for passive reactors

    Energy Technology Data Exchange (ETDEWEB)

    Kelly, J.M. [Nuclear Regulatory Commission, Washington, DC (United States)

    1997-07-01

    The U.S. Nuclear Regulatory Commission has received an application for design certification from the Westinghouse Electric Corporation for an Advanced Light Water Reactor design known as the AP600. As part of the design certification process, the USNRC uses its thermal-hydraulic system analysis codes to independently audit the vendor calculations. The focus of this effort has been the small break LOCA transients that rely upon the passive safety features of the design to depressurize the primary system sufficiently so that gravity driven injection can provide a stable source for long term cooling. Of course, large break LOCAs have also been considered, but as the involved phenomena do not appear to be appreciably different from those of current plants, they were not discussed in this paper. Although the SBLOCA scenario does not appear to threaten core coolability - indeed, heatup is not even expected to occur - there have been concerns as to the performance of the passive safety systems. For example, the passive systems drive flows with small heads, consequently requiring more precision in the analysis compared to active systems methods for passive plants as compared to current plants with active systems. For the analysis of SBLOCAs and operating transients, the USNRC uses the RELAP5 thermal-hydraulic system analysis code. To assure the applicability of RELAP5 to the analysis of these transients for the AP600 design, a four year long program of code development and assessment has been undertaken.

  7. Numerical simulations of subcritical reactor kinetics in thermal hydraulic transient phases

    Energy Technology Data Exchange (ETDEWEB)

    Yoo, J.; Park, W. S. [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1998-12-31

    A subcritical reactor driven by a linear proton accelerator has been considered as a nuclear waste incinerator at Korea Atomic Energy Research Institute (KAERI). Since the multiplication factor of a subcritical reactor is less than unity, to compensate exponentially decreasing fission neutrons, external neutrons form spallation reactions are essentially required for operating the reactor in its steady state. Furthermore, the profile of accelerator beam currents is very important in controlling a subcritical reactor, because the reactor power varies in accordance to the profile of external neutrons. We have developed a code system to find numerical solutions of reactor kinetics equations, which are the simplest dynamic model for controlling reactors. In a due course of our previous numerical study of point kinetics equations for critical reactors, however, we learned that the same code system can be used in studying dynamic behavior of the subcritical reactor. Our major motivation of this paper is to investigate responses of subcritical reactors for small changes in thermal hydraulic parameters. Building a thermal hydraulic model for the subcritical reactor dynamics, we performed numerical simulations for dynamic responses of the reactor based on point kinetics equations with a source term. Linearizing a set of coupled differential equations for reactor responses, we focus our research interest on dynamic responses of the reactor to variations of the thermal hydraulic parameters in transient phases. 5 refs., 8 figs. (Author)

  8. Engineered Barrier Systems Thermal-Hydraulic-Chemical Column Test Report

    Energy Technology Data Exchange (ETDEWEB)

    W.E. Lowry

    2001-12-13

    The Engineered Barrier System (EBS) Thermal-Hydraulic-Chemical (THC) Column Tests provide data needed for model validation. The EBS Degradation, Flow, and Transport Process Modeling Report (PMR) will be based on supporting models for in-drift THC coupled processes, and the in-drift physical and chemical environment. These models describe the complex chemical interaction of EBS materials, including granular materials, with the thermal and hydrologic conditions that will be present in the repository emplacement drifts. Of particular interest are the coupled processes that result in mineral and salt dissolution/precipitation in the EBS environment. Test data are needed for thermal, hydrologic, and geochemical model validation and to support selection of introduced materials (CRWMS M&O 1999c). These column tests evaluated granular crushed tuff as potential invert ballast or backfill material, under accelerated thermal and hydrologic environments. The objectives of the THC column testing are to: (1) Characterize THC coupled processes that could affect performance of EBS components, particularly the magnitude of permeability reduction (increases or decreases), the nature of minerals produced, and chemical fractionation (i.e., concentrative separation of salts and minerals due to boiling-point elevation). (2) Generate data for validating THC predictive models that will support the EBS Degradation, Flow, and Transport PMR, Rev. 01.

  9. Strategic need for a multi-purpose thermal hydraulic loop for support of advanced reactor technologies

    Energy Technology Data Exchange (ETDEWEB)

    O' Brien, James E. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Sabharwall, Piyush [Idaho National Lab. (INL), Idaho Falls, ID (United States); Yoon, Su -Jong [Idaho National Lab. (INL), Idaho Falls, ID (United States); Housley, Gregory K. [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2014-09-01

    This report presents a conceptual design for a new high-temperature multi fluid, multi loop test facility for the INL to support thermal hydraulic, materials, and thermal energy storage research for nuclear and nuclear-hybrid applications. In its initial configuration, the facility will include a high-temperature helium loop, a liquid salt loop, and a hot water/steam loop. The three loops will be thermally coupled through an intermediate heat exchanger (IHX) and a secondary heat exchanger (SHX). Research topics to be addressed with this facility include the characterization and performance evaluation of candidate compact heat exchangers such as printed circuit heat exchangers (PCHEs) at prototypical operating conditions, flow and heat transfer issues related to core thermal hydraulics in advanced helium-cooled and salt-cooled reactors, and evaluation of corrosion behavior of new cladding materials and accident-tolerant fuels for LWRs at prototypical conditions. Based on its relevance to advanced reactor systems, the new facility has been named the Advanced Reactor Technology Integral System Test (ARTIST) facility. Research performed in this facility will advance the state of the art and technology readiness level of high temperature intermediate heat exchangers (IHXs) for nuclear applications while establishing the INL as a center of excellence for the development and certification of this technology. The thermal energy storage capability will support research and demonstration activities related to process heat delivery for a variety of hybrid energy systems and grid stabilization strategies. Experimental results obtained from this research will assist in development of reliable predictive models for thermal hydraulic design and safety codes over the range of expected advanced reactor operating conditions. Proposed/existing IHX heat transfer and friction correlations and criteria will be assessed with information on materials compatibility and instrumentation

  10. Thermal-Hydraulic Experiments and Modelling for Advanced Nuclear Reactor Systems

    Energy Technology Data Exchange (ETDEWEB)

    Song, C. H.; Chung, M. K.; Park, C. K. and others

    2005-04-15

    The objectives of the project are to study thermal hydraulic characteristics of reactor primary system for the verification of the reactor safety and to evaluate new safety concepts of new safety design features. To meet the research goal, several thermal hydraulic experiments were performed and related thermal hydraulic models were developed with the experimental data which were produced through the thermal hydraulic experiments. Followings are main research topics; - Multi-dimensional Phenomena in a Reactor Vessel Downcomer - Condensation Load and Thermal Mixing in the IRWST - Development of Thermal-Hydraulic Models for Two-Phase Flow - Development of Measurement Techniques for Two-Phase Flow - Supercritical Reactor T/H Characteristics Analysis From the above experimental and analytical studies, new safety design features of the advanced power reactors were verified and lots of the safety issues were also resolved.

  11. Coupled neutronics - thermal-hydraulics programs for SCWRS

    Energy Technology Data Exchange (ETDEWEB)

    Reiss, T. [Institute of Nuclear Techniques, Budapest University of Technology and Economics, Muegyetem rkp. 9., 1111 Budapest (Hungary)

    2010-07-01

    The Supercritical Water Cooled Reactor (SCWR) was chosen as one of the Generation IV reactors by GIF. At the moment, a number of concepts - thermal as well as fast ones - exist. The reference parameters for a thermal SCWR have been taken from the European High Performance Light Water Reactor (HPLWR). Since the pressure is higher than the critical pressure (22.1 MPa) there is no change in the phase of the water in the core. On the other hand, due to the significant changes in the physical properties of water at supercritical pressure, the system is susceptible to local temperature, density and power oscillations. This inclination is increased by the pseudo-critical transformation of the water used as coolant. Thus, for modelling a system of this type coupled neutronics - thermal-hydraulics programs are required. Such a program system has been developed with the following main features: great modularity which allows for easy modifications, thus several SCWR concepts can be studied; detailed assembly calculations (with MCNP) and full-core analysis (with SCALE) are supported; the differential equations of xenon poisoning are implemented to study xenon oscillations. The program system was used to examine the assembly of the HPLWR, to design the assembly and the core of the Simplified Supercritical Water Cooled Reactor (SSCWR) and to model xenon oscillations in SCWRs. (authors)

  12. submitter Thermal, Hydraulic, and Electromagnetic Modeling of Superconducting Magnet Systems

    CERN Document Server

    Bottura, L

    2016-01-01

    Modeling techniques and tailored computational tools are becoming increasingly relevant to the design and analysis of large-scale superconducting magnet systems. Efficient and reliable tools are useful to provide an optimal forecast of the envelope of operating conditions and margins, which are difficult to test even when a prototype is available. This knowledge can be used to considerably reduce the design margins of the system, and thus the overall cost, or increase reliability during operation. An integrated analysis of a superconducting magnet system is, however, a complex matter, governed by very diverse physics. This paper reviews the wide spectrum of phenomena and provides an estimate of the time scales of thermal, hydraulic, and electromagnetic mechanisms affecting the performance of superconducting magnet systems. The analysis is useful to provide guidelines on how to divide the complex problem into building blocks that can be integrated in a design and analysis framework for a consistent multiphysic...

  13. Investigation and Development of the Thermal Preparation System of the Trailbuilder Machinery Hydraulic Actuator

    Science.gov (United States)

    Konev, V.; Polovnikov, E.; Krut, O.; Merdanov, Sh; Zakirzakov, G.

    2017-07-01

    It’s determined that the main part of trailbuilders operated in the North is the technology equipped by the hydraulic actuator. Further development of the northern territories will demand using of various means and ways machinery thermal preparation, and also the machinery of the northern fulfillment. On this basis problems in equipment operation are defined. One of the main is efficiency supplying of a hydraulic actuator. On the basis of the operating conditions’ analysis of trailbuilder hydraulic actuator operation it is determined, that under low negative temperatures the means of thermal preparation are necessary. The existing systems warm up only a hydraulic tank or warming up of the hydro equipment before the machinery operation is carried out under loading with intensive wears. Thus, with the purpose to raise the efficiency of thermal hydraulic actuator, operated far from stationary bases autonomous, energy saving, not expensive in creation and operation systems are necessary. In accordance with the analysis of means and ways of the thermal preparation of the hydraulic actuator and the thermal balance calculations of the (internal) combustion engine the system of the hydraulic actuator heating is offered and is being investigated. It contains a local hydraulic actuator warming up and the system of internal combustion engine heat utilization. Within research operation conditions of the local hydraulic actuator heating are viewed and determined, taking into account constructive changes to the local hydraulic actuator heating. Mathematical modelling of the heat technical process in the modernized hydraulic actuator is considered. As a result temperature changes of the heat-transfer and the hydraulic cylinder in time are determined. To check the theoretical researches and to define dependences on hydraulic actuator warming up, the experimental installation is made. It contains the measuring equipment, a small tank with the heat exchanger of the burnt gases

  14. THERMIT2. BWR & PWR Thermal-Hydraulic Analysis

    Energy Technology Data Exchange (ETDEWEB)

    Kazimi, M.S.; Kao, S.P.; Kelly, J.E. [Massachusetts Institute of Technology, Cambridge, MA (United States)

    1992-02-27

    THERMIT2, the most recent release of THERMIT, is intended for thermal-hydraulic analysis of both boiling and pressurized water reactor cores. It solves the three-dimensional, two-fluid equations describing the two-phase flow and heat transfer dynamics in rectangular coordinates. The two-fluid model uses separate partial differential equations expressing conservation of mass, momentum, and energy for each fluid. By expressing the exchange of mass, momentum, and energy between the fluids with physically-based mathematical models, the relative motion and thermal non-equilibrium between the fluids can exist. THERMIT2 offers the choice of either pressure or velocity boundary conditions at the top and bottom of the core. THERMIT2 includes a two-phase turbulent mixing model which provides subchannel analysis capability. THERMIT2 also solves the radial heat conduction equations for fuel pin temperatures, and calculates the heat flux from fuel pin to coolant with appropriate heat transfer models described by a boiling curve.

  15. ATHENA (Advanced Thermal Hydraulic Energy Network Analyzer) simulation of a loss of coolant accident in a space reactor

    Energy Technology Data Exchange (ETDEWEB)

    Roth, P.A.; Shumway, R.W.

    1988-01-01

    The Advanced Thermal Hydraulic Energy Network Analyzer (ATHENA) code was used to simulate a loss-of-coolant accident (LOCA) in a conceptual space reactor design. ATHENA provides the capability of simulating the thermal-hydraulic behavior of the wide variety of systems which are being considered for use in space reactors. Flow loops containing any one of several available working fluids may interact through thermal connections with other loops containing the same or a different working fluid. The code can be used to model special systems such as: heat pipes, point reactor kinetics, plant control systems, turbines, valves, and pumps. This work demonstrates the application of the thermal radiation model which has been recently incorporated into ATHENA and verifies the need for supplemental reactor cooling to prevent reactor fuel damage in the event of a LOCA.

  16. Dynamic behavior of upper hydraulic drive control rod

    Energy Technology Data Exchange (ETDEWEB)

    Li Xiaotian [Institute of Nuclear and New Energy Technology, Beijing 100084 (China)]. E-mail: lixiaotian@tsinghua.edu.cn; He Shuyan [Institute of Nuclear and New Energy Technology, Beijing 100084 (China)

    2006-12-15

    A 200 MW nuclear heating reactor (NHR) was developed based on the 5 MW testing heating reactor (THR-5). In order to improve its features, the INET was used to optimize its structure design. One task was to design the upper hydraulic drive control rod, which is located at the upper part of the reactor core, so that the fuel rod in each corner of the bundle will not be removed as they are in the THR-5. A program based on the method of characteristics was run to investigate the dynamic behavior of the upper hydraulic control rod drive system. The steady-state and dynamic characteristics of the control rod were studied. The influence of the reactor temperature and structure parameters on the operation reliability was investigated. Finally the optimal design parameters of the upper hydraulic drive control rod were obtained.

  17. Thermal Hydraulic Analysis Using GIS on Application of HTR to Thermal Recovery of Heavy Oil Reservoirs

    Directory of Open Access Journals (Sweden)

    Yangping Zhou

    2012-01-01

    Full Text Available At present, large water demand and carbon dioxide (CO2 emissions have emerged as challenges of steam injection for oil thermal recovery. This paper proposed a strategy of superheated steam injection by the high-temperature gas-cooled reactor (HTR for thermal recovery of heavy oil, which has less demand of water and emission of CO2. The paper outlines the problems of conventional steam injection and addresses the advantages of superheated steam injection by HTR from the aspects of technology, economy, and environment. A Geographic Information System (GIS embedded with a thermal hydraulic analysis function is designed and developed to analyze the strategy, which can make the analysis work more practical and credible. Thermal hydraulic analysis using this GIS is carried out by applying this strategy to a reference heavy oil field. Two kinds of injection are considered and compared: wet steam injection by conventional boilers and superheated steam injection by HTR. The heat loss, pressure drop, and possible phase transformation are calculated and analyzed when the steam flows through the pipeline and well tube and is finally injected into the oil reservoir. The result shows that the superheated steam injection from HTR is applicable and promising for thermal recovery of heavy oil reservoirs.

  18. Development of a best estimate auditing code for CANDU thermal hydraulic safety analysis

    Energy Technology Data Exchange (ETDEWEB)

    Hwang, M. K.; Lee, W. J. [Korea Atomic Energy Research Institute, Taejeon (Korea)

    2001-03-01

    The main purpose of this study is to develop a thermal hydraulic auditing code for the CANDU reactor, modifying the model of existing PWR auditing tool , i.e. RELAP5/MOD3. This scope of project is a fourth step of the whole project, applying the RELAP5/MOD3/CANDU+ version for the real CANDU plant LOCA Analysis and D2O leakage incident. There are three main models under investigation, i.e. Moody critical flow model, flow regime model of horizontal CANDU bundle, and fuel element heatup model when the stratification occurs, especially when CANDU LOCA is tested. Also, for Wolsung unit 1 D2O leakage incident analysis, the plant behavior is predicted with the newly developed version for the first 1000 seconds after onset of the incident, with the main interest aiming for system pressure, level control system, and thermal hydraulic transient behavior of the secondary system. The model applied for this particular application includes heat transfer model of nuclear fuel assembly, decay heat model, and MOV (Motor Operated Valve) model. Finally, the code maintenance work, mainly correcting the known errors, is presented. 12 refs., 26 figs., 3 tabs. (Author)

  19. Parametric analyses of DEMO Divertor using two dimensional transient thermal hydraulic modelling

    Science.gov (United States)

    Domalapally, Phani; Di Caro, Marco

    2017-11-01

    Among the options considered for cooling of the Plasma facing components of the DEMO reactor, water cooling is a conservative option because of its high heat removal capability. In this work a two-dimensional transient thermal hydraulic code is developed to support the design of the divertor for the projected DEMO reactor with water as a coolant. The mathematical model accounts for transient 2D heat conduction in the divertor section. Temperature-dependent properties are used for more accurate analysis. Correlations for single phase flow forced convection, partially developed subcooled nucleate boiling, fully developed subcooled nucleate boiling and film boiling are used to calculate the heat transfer coefficients on the channel side considering the swirl flow, wherein different correlations found in the literature are compared against each other. Correlation for the Critical Heat Flux is used to estimate its limit for a given flow conditions. This paper then investigates the results of the parametric analysis performed, whereby flow velocity, diameter of the coolant channel, thickness of the coolant pipe, thickness of the armor material, inlet temperature and operating pressure affect the behavior of the divertor under steady or transient heat fluxes. This code will help in understanding the basic parameterś effect on the behavior of the divertor, to achieve a better design from a thermal hydraulic point of view.

  20. APT target/blanket design and thermal hydraulics

    Energy Technology Data Exchange (ETDEWEB)

    Cappiello, M.; Pitcher, E.; Pasamehmetoglu, K.

    1999-04-01

    The Accelerator Production of Tritium (APT) Target/Blanket (T/B) system is comprised of an assembly of tritium producing modules supported by control, heat removal, shielding and retargeting systems. The T/B assembly produces tritium using a high-energy proton beam, a tungsten/lead spallation neutron source and {sup 3}He gas as the tritium producing feedstock. For the nominal production mode, protons are accelerated to an energy of 1030 MeV at a current of 100 mA and are directed onto the T/B assembly. The protons are expanded using a raster/expansion system to illuminate a 0.19m by 1.9m beam spot on the front face of a centrally located tungsten neutron source. A surrounding lead blanket produces additional neutrons from scattered high-energy particles. The tungsten neutron source consists of nested, Inconel-718 clad tungsten cylinders assembled in horizontal Inconel-718 tubes. Each tube contains up to 6 cylinders with annular flow channel gaps of 0.102 cm. These horizontal tubes are manifolded into larger diameter vertical inlet and outlet pipes, which provide coolant. The horizontal and vertical tubes make up a structure similar to that of rungs on a ladder. The entire tungsten neutron source consists of 11 such ladders separated into two modules, one containing five ladders and the other six. Ladders are separated by a 0.3 m void region to increase nucleon leakage. The peak thermal-hydraulic conditions in the tungsten neutron source occur in the second ladder from the front. Because tungsten neutron source design has a significant number of parallel flow channels, the limiting thermal-hydraulic parameter is the onset of significant void (OSV) rather than critical heat flux (CHF). A blanket region surrounds the tungsten neutron source. The lateral blanket region is approximately 120 cm thick and 400 cm high. Blanket material consists of lead, {sup 3}He gas, aluminum, and light-water coolant. The blanket region is subdivided into rows based on the local power

  1. Thermal hydraulic analysis of the JMTR improved LEU-core

    Energy Technology Data Exchange (ETDEWEB)

    Tabata, Toshio; Nagao, Yoshiharu; Komukai, Bunsaku; Naka, Michihiro; Fujiki, Kazuo [Japan Atomic Energy Research Inst., Oarai, Ibaraki (Japan). Oarai Research Establishment; Takeda, Takashi [Radioactive Waste Management and Nuclear Facility Decommissioning Technology Center, Tokai, Ibaraki (Japan)

    2003-01-01

    After the investigation of the new core arrangement for the JMTR reactor in order to enhance the fuel burn-up and consequently extend the operation period, the ''improved LEU core'' that utilized 2 additional fuel elements instead of formerly installed reflector elements, was adopted. This report describes the results of the thermal-hydraulic analysis of the improved LEU core as a part of safety analysis for the licensing. The analysis covers steady state, abnormal operational transients and accidents, which were described in the annexes of the licensing documents as design bases events. Calculation conditions for the computer codes were conservatively determined based on the neutronic analysis results and others. The results of the analysis, that revealed the safety criteria were satisfied on the fuel temperature, DNBR and primary coolant temperature, were used in the licensing. The operation license of the JMTR with the improved LEU core was granted in March 2001, and the reactor operation with new core started in November 2001 as 142nd operation cycle. (author)

  2. Sensitivity theory applied to a transient thermal-hydraulics problem

    Energy Technology Data Exchange (ETDEWEB)

    Weber, C.F.; Oblow, E.M.

    1979-10-01

    A new method for sensitivity analysis of transient nonlinear problems is developed and applied to a reactor thermal-hydraulics problem. The method resembles the differential sensitivity methods currently used in the linear problems of reactor physics, but it is applicable to nonlinear systems as well. The equations governing heat transfer and fluid flow in a fuel pin and surrounding coolant are given and used to derive a second set of equations (commonly known as the adjoint equations) used in the sensitivity analysis. Both systems contain one second-order parabolic and one first-order hyperbolic partial differential equation. Difference equations are derived to approximate both systems and the convergence properties of these discrete systems are evaluated, yielding a useful analysis of the numerical solution. The solution functions are used to derive sensitivity coefficients for any desired integral response. These sensitivity coefficients are used in a first-order perturbation theory to predict changes in a response resulting from changes in parameter values. The results of a test problem are shown, verifying that this procedure is indeed useful for a wide variety of sensitivity calculations.

  3. Thermal hydraulics of accelerator driven system windowless targets

    Directory of Open Access Journals (Sweden)

    Bruno ePanella

    2015-07-01

    Full Text Available The study of the fluid dynamics of the windowless spallation target of an Accelerator Driven System (ADS is presented. Several target mockup configurations have been investigated: the first one was a symmetrical target, that was made by two concentric cylinders, the other configurations are not symmetrical. In the experiments water has been used as hydraulic equivalent to lead-bismuth eutectic fluid. The experiments have been carried out at room temperature and flow rate up to 24 kg/s. The fluid velocity components have been measured by an ultrasound technique. The velocity field of the liquid within the target region either for the approximately axial-symmetrical configuration or for the not symmetrical ones as a function of the flow rate and the initial liquid level is presented. A comparison of experimental data with the prediction of the finite volume FLUENT code is also presented. Moreover the results of a 2D-3D numerical analysis that investigates the effect on the steady state thermal and flow fields due to the insertion of guide vanes in the windowless target unit of the EFIT project ADS nuclear reactor are presented, by analysing both the cold flow case (absence of power generation and the hot flow case (nominal power generation inside the target unit.

  4. Comparative Analysis of CTF and Trace Thermal-Hydraulic Codes Using OECD/NRC PSBT Benchmark Void Distribution Database

    Directory of Open Access Journals (Sweden)

    M. Avramova

    2013-01-01

    Full Text Available The international OECD/NRC PSBT benchmark has been established to provide a test bed for assessing the capabilities of thermal-hydraulic codes and to encourage advancement in the analysis of fluid flow in rod bundles. The benchmark was based on one of the most valuable databases identified for the thermal-hydraulics modeling developed by NUPEC, Japan. The database includes void fraction and departure from nucleate boiling measurements in a representative PWR fuel assembly. On behalf of the benchmark team, PSU in collaboration with US NRC has performed supporting calculations using the PSU in-house advanced thermal-hydraulic subchannel code CTF and the US NRC system code TRACE. CTF is a version of COBRA-TF whose models have been continuously improved and validated by the RDFMG group at PSU. TRACE is a reactor systems code developed by US NRC to analyze transient and steady-state thermal-hydraulic behavior in LWRs and it has been designed to perform best-estimate analyses of LOCA, operational transients, and other accident scenarios in PWRs and BWRs. The paper presents CTF and TRACE models for the PSBT void distribution exercises. Code-to-code and code-to-data comparisons are provided along with a discussion of the void generation and void distribution models available in the two codes.

  5. Thermal-hydraulics/thermal-mechanics temporal coupling for unprotected loss of flow accidents simulations on a SFR

    Directory of Open Access Journals (Sweden)

    Patricot Cyril

    2016-01-01

    Full Text Available In the frame of ASTRID designing, unprotected loss of flow (ULOF accidents are considered. As the reactor is not scrammed, power evolution is driven by neutronic feedbacks, among which Doppler effect, linked to fuel temperature, is prominent. Fuel temperature is calculated using thermal properties of fuel pins (we will focus on heat transfer coefficient between fuel pellet and cladding, Hgap, and on fuel thermal conductivity, λfuel which vary with irradiation conditions (neutronic flux, mass flow and history for instance and during transient (mainly because of dilatation of materials with temperature. In this paper, we propose an analysis of the impact of spatial variation and temporal evolution of thermal properties of fuel pins on a CFV-like core [M.S. Chenaud et al., Status of the ASTRID core at the end of the pre-conceptual design phase 1, in Proceedings of ICAPP 2013, Jeju Island, Korea (2013] behavior during an ULOF accident. These effects are usually neglected under some a priori conservative assumptions. The vocation of our work is not to provide a best-estimate calculation of ULOF transient, but to discuss some of its physical aspects. To achieve this goal, we used TETAR, a thermal-hydraulics system code developed by our team to calculate ULOF transients, GERMINAL V1.5, a CEA code dedicated to SFR pin thermal-mechanics calculations and APOLLO3®, a neutronic code in development at CEA.

  6. Thermal-hydraulics of lead bismuth for accelerator driven systems

    Energy Technology Data Exchange (ETDEWEB)

    Thomas Schulenberg; Xu Cheng; Robert Stieglitz [Institute for Nuclear and Energy Technologies, Forschungszentrum Karlsruhe, P.O. Box 3640, 76021 Karlsruhe (Germany)

    2005-07-01

    Full text of publication follows: Lead bismuth has been selected as one of the most suitable coolants to be used in accelerator driven systems (ADS) for transmutation of minor actinides. It serves both, as a target material of the spallation source to balance the neutron economy, and as a coolant with high thermal inertia to provide a safe and reliable heat transfer to the secondary power cycle. With the aim to develop the required technologies to enable the later design of such ADS systems, the Karlsruhe Lead bismuth LAboratory KALLA, consisting of three test loops, has been built and set into operation at the Forschungszentrum Karlsruhe since 2000, keeping more than 45 t of PbBi in operation at temperatures up to 550 deg. C. The test program includes oxygen control systems, heat flux simulation tools, electro-magnetic and mechanical pump technologies, heat transfer and flow measurements, reliability and corrosion tests. In a first test campaign, a technology loop called THESYS was built to develop measurement technologies for the acquisition of scalar quantities, like pressures, temperatures, concentrations, and flow rates, as well as velocity fields, which are required for both operational and scientific purposes. THESYS also allowed to perform generic turbulent heat transfer experiments necessary to provide liquid metal adapted turbulent heat transfer models for ADS design analyses. The second loop, the thermalhydraulic loop THEADES with an installed power of 2.5 MW, has been built to conduct prototypical component experiments for beam windows (e.g. MEGAPIE or MYHRRA) or fuel rod configurations. First test results will be reported. The experimental team is supported by a numerical team who studied the thermal hydraulics of the tested components in order to enable a later transfer of the results to industrial systems. Three different types of codes are being improved: lumped parameter codes (e.g. ATHLET) to perform system analyses for lead bismuth in loops

  7. Issues and future direction of thermal-hydraulics research and development in nuclear power reactors

    Energy Technology Data Exchange (ETDEWEB)

    Saha, P., E-mail: pradip.saha@ge.com [GE Hitachi Nuclear Energy, Wilmington, NC (United States); Aksan, N. [GRNSPG Group, University of Pisa (Italy); Andersen, J. [GE Hitachi Nuclear Energy, Wilmington, NC (United States); Yan, J. [Westinghouse Electric Co., Columbia, SC (United States); Simoneau, J.P. [AREVA, Lyon (France); Leung, L. [Atomic Energy of Canada Ltd., Chalk River, Ontario (Canada); Bertrand, F. [CEA, DEN, DER, F-13108 Saint-Paul-Lez-Durance (France); Aoto, K.; Kamide, H. [Japan Atomic Energy Agency, Chiyoda-ku, Tokyo (Japan)

    2013-11-15

    The paper archives the proceedings of an expert panel discussion on the issues and future direction of thermal-hydraulic research and development in nuclear power reactors held at the NURETH-14 conference in Toronto, Canada, in September 2011. Thermal-hydraulic issues related to both operating and advanced reactors are presented. Advances in thermal-hydraulics have significantly improved the performance of operating reactors. Further thermal-hydraulics research and development is continuing in both experimental and computational areas for operating reactors, reactors under construction or ready for near-term deployment, and advanced Generation-IV reactors. As the computing power increases, the fine-scale multi-physics computational models, coupled with the systems analysis code, are expected to provide answers to many challenging problems in both operating and advanced reactor designs.

  8. Development of thermal hydraulic models for the reliable regulatory auditing code

    Energy Technology Data Exchange (ETDEWEB)

    Chung, B. D.; Song, C. H.; Lee, Y. J.; Kwon, T. S. [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    2003-04-15

    The objective of this project is to develop thermal hydraulic models for use in improving the reliability of the regulatory auditing codes. The current year fall under the first step of the 3 year project, and the main researches were focused on identifying the candidate thermal hydraulic models for improvement and to develop prototypical model development. During the current year, the verification calculations submitted for the APR 1400 design certification have been reviewed, the experimental data from the MIDAS DVI experiment facility in KAERI have been analyzed and evaluated, candidate thermal hydraulic models for improvement have been identified, prototypical models for the improved thermal hydraulic models have been developed, items for experiment in connection with the model development have been identified, and preliminary design of the experiment has been carried out.

  9. IMPROVEMENT OF THE LOCA PSA MODEL USING A BEST-ESTIMATE THERMAL-HYDRAULIC ANALYSIS

    Directory of Open Access Journals (Sweden)

    DONG HYUN LEE

    2014-08-01

    Full Text Available Probabilistic Safety Assessment (PSA has been widely used to estimate the overall safety of nuclear power plants (NPP and it provides base information for risk informed application (RIA and risk informed regulation (RIR. For the effective and correct use of PSA in RIA/RIR related decision making, the risk estimated by a PSA model should be as realistic as possible. In this work, a best-estimate thermal-hydraulic analysis of loss-of-coolant accidents (LOCAs for the Hanul Nuclear Units 3&4 is first carried out in a systematic way. That is, the behaviors of peak cladding temperature (PCT were analyzed with various combinations of break sizes, the operating conditions of safety systems, and the operator's action time for aggressive secondary cooling. Thereafter, the results of the thermal-hydraulic analysis have been reflected in the improvement of the PSA model by changing both accident sequences and success criteria of the event trees for the LOCA scenarios.

  10. Thermal-hydraulic code selection for modular high temperature gas-cooled reactors

    Energy Technology Data Exchange (ETDEWEB)

    Komen, E.M.J.; Bogaard, J.P.A. van den

    1995-06-01

    In order to study the transient thermal-hydraulic system behaviour of modular high temperature gas-cooled reactors, the thermal-hydraulic computer codes RELAP5, MELCOR, THATCH, MORECA, and VSOP are considered at the Netherlands Energy Research Foundation ECN. This report presents the selection of the most appropriate codes. To cover the range of relevant accidents, a suite of three codes is recommended for analyses of HTR-M and MHTGR reactors. (orig.).

  11. Research on the improvement of nuclear safety -Thermal hydraulic tests for reactor safety system-

    Energy Technology Data Exchange (ETDEWEB)

    Jung, Moon Kee; Park, Choon Kyung; Yang, Sun Kyoo; Chun, Se Yung; Song, Chul Hwa; Jun, Hyung Kil; Jung, Heung Joon; Won, Soon Yun; Cho, Yung Roh; Min, Kyung Hoh; Jung, Jang Hwan; Jang, Suk Kyoo; Kim, Bok Deuk; Kim, Wooi Kyung; Huh, Jin; Kim, Sook Kwan; Moon, Sang Kee; Lee, Sang Il [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1995-06-01

    The present research aims at the development of the thermal hydraulic verification test technology for the safety system of the conventional and advanced nuclear power plant and the development of the advanced thermal hydraulic measuring techniques. In this research, test facilities simulating the primary coolant system and safety system are being constructed for the design verification tests of the existing and advanced nuclear power plant. 97 figs, 14 tabs, 65 refs. (Author).

  12. Current and anticipated uses of thermal-hydraulic codes in NFI

    Energy Technology Data Exchange (ETDEWEB)

    Tsuda, K. [Nuclear Fuel Industries, Ltd., Tokyo (Japan); Takayasu, M. [Nuclear Fuel Industries, Ltd., Sennann-gun (Japan)

    1997-07-01

    This paper presents the thermal-hydraulic codes currently used in NFI for the LWR fuel development and licensing application including transient and design basis accident analyses of LWR plants. The current status of the codes are described in the context of code capability, modeling feature, and experience of code application related to the fuel development and licensing. Finally, the anticipated use of the future thermal-hydraulic code in NFI is briefly given.

  13. Thermal hydraulic core simulation of the MYRRHA Reactor in steady state operation

    Energy Technology Data Exchange (ETDEWEB)

    Ferandes, Gustavo H.N.; Ramos, Mário C.; Carvalho, Athos M.S.S.; Cabrera, Carlos E.V.; Costa, Antonella L.; Pereira, Claubia, E-mail: ghnfernandes@gmail.com, E-mail: marc5663@gmail.com, E-mail: athos1495@yahoo.com.br, E-mail: carlosvelcab@hotmail.com, E-mail: antonella@nuclear.ufmg.br, E-mail: claubia@nuclear.ufmg.br [Universidade Federal de Minas Gerais (UFMG), Belo Horizonte, MG (Brazil). Departamento de Engenharia Nuclear; Instituto Nacional de Ciência e Tecnologia de Reatores Nucleares Inovadores/CNPq (Brazil)

    2017-07-01

    MYRRHA (Multi-purpose Hybrid Research Reactor for High-tech Applications) is a prototype nuclear subcritical reactor driven by a particle accelerator. As a special property, the reactor maintains the nuclear fission chain reaction by means of an external neutron source provided by a particle accelerator. The main aim of this work is to study two types of coolants, LBE (Lead-Bismuth Eutectic) and Na (Sodium) that are two strong candidates to be used in ADS systems as well as in Generation IV (GEN-IV) reactors. Firstly, it was developed a thermal hydraulic model of the MYRRHA core using the RELAP5-3D, considering LBE as coolant (original project). After this, the LBE was substituted by Na coolant to investigate the reactor behavior in such case. Results have demonstrated the high heat transfer capacity of the LBE coolant in this type of system. (author)

  14. Thermal hydraulics modeling of the US Geological Survey TRIGA reactor

    Science.gov (United States)

    Alkaabi, Ahmed K.

    The Geological Survey TRIGA reactor (GSTR) is a 1 MW Mark I TRIGA reactor located in Lakewood, Colorado. Single channel GSTR thermal hydraulics models built using RELAP5/MOD3.3, RELAP5-3D, TRACE, and COMSOL Multiphysics predict the fuel, outer clad, and coolant temperatures as a function of position in the core. The results from the RELAP5/MOD3.3, RELAP5-3D, and COMSOL models are similar. The TRACE model predicts significantly higher temperatures, potentially resulting from inappropriate convection correlations. To more accurately study the complex fluid flow patterns within the core, this research develops detailed RELAP5/MOD3.3 and COMSOL multichannel models of the GSTR core. The multichannel models predict lower fuel, outer clad, and coolant temperatures compared to the single channel models by up to 16.7°C, 4.8°C, and 9.6°C, respectively, as a result of the higher mass flow rates predicted by these models. The single channel models and the RELAP5/MOD3.3 multichannel model predict that the coolant temperatures in all fuel rings rise axially with core height, as the coolant in these models flows predominantly in the axial direction. The coolant temperatures predicted by the COMSOL multichannel model rise with core height in the B-, C-, and D-rings and peak and then decrease in the E-, F-, and G-rings, as the coolant tends to flow from the bottom sides of the core to the center of the core in this model. Experiments at the GSTR measured coolant temperatures in the GSTR core to validate the developed models. The axial temperature profiles measured in the GSTR show that the flow patterns predicted by the COMSOL multichannel model are consistent with the actual conditions in the core. Adjusting the RELAP5/MOD3.3 single and multichannel models by modifying the axial and cross-flow areas allow them to better predict the GSTR coolant temperatures; however, the adjusted models still fail to predict accurate axial temperature profiles in the E-, F-, and G-rings.

  15. Application of flow network models of SINDA/FLUINT{sup TM} to a nuclear power plant system thermal hydraulic code

    Energy Technology Data Exchange (ETDEWEB)

    Chung, Ji Bum [Institute for Advanced Engineering, Yongin (Korea, Republic of); Park, Jong Woon [Korea Electric Power Research Institute, Taejon (Korea, Republic of)

    1998-12-31

    In order to enhance the dynamic and interactive simulation capability of a system thermal hydraulic code for nuclear power plant, applicability of flow network models in SINDA/FLUINT{sup TM} has been tested by modeling feedwater system and coupling to DSNP which is one of a system thermal hydraulic simulation code for a pressurized heavy water reactor. The feedwater system is selected since it is one of the most important balance of plant systems with a potential to greatly affect the behavior of nuclear steam supply system. The flow network model of this feedwater system consists of condenser, condensate pumps, low and high pressure heaters, deaerator, feedwater pumps, and control valves. This complicated flow network is modeled and coupled to DSNP and it is tested for several normal and abnormal transient conditions such turbine load maneuvering, turbine trip, and loss of class IV power. The results show reasonable behavior of the coupled code and also gives a good dynamic and interactive simulation capabilities for the several mild transient conditions. It has been found that coupling system thermal hydraulic code with a flow network code is a proper way of upgrading simulation capability of DSNP to mature nuclear plant analyzer (NPA). 5 refs., 10 figs. (Author)

  16. Thermal hydraulic analysis of China fusion engineering test reactor during thermal quenching by comparative approach of Relap5 and THEATRe codes

    Science.gov (United States)

    Khan, Salah Ud-Din; Song, Yuntao; Khan, Shahab Ud-Din

    2016-10-01

    Thermal quenching in Tokamak reactor is the most obvious phenomenon happens during plasma disruption conditions. The current research is focused on the thermal behavior of different parameters of China fusion engineering test reactor (CFETR) including reactor power, pressure and mass flow rate conditions. The analysis was performed by two thermal hydraulic codes, i.e. THEATRe and Relap5. During the first phase of research, thermal quenching behavior and trends that can be possible during the reactor operation was performed. In the next phase, nodalization diagram of THEATRe and Relap5 codes were developed. The listed parameters were calculated and analyzed for the safety aspects of the reactor. The main objective of the research was to analyze the blanket system of CFETR (Tokamak) for safety concerns during disruption condition. The research will be extended to other components for safe operation of reactor as well.

  17. Development of thermal hydraulic models for main circulation circuit of RBMK-1500 reactor using Apros and Cathare 2 codes

    Energy Technology Data Exchange (ETDEWEB)

    Zemulis, G.; Jasiulevicius, A. [Kaunas University of Technology, Dept. of Thermal and Nuclear Energy, Kaunas, (Lithuania)

    2001-07-01

    Reactor safety is the most important issue in nuclear engineering. It concerns the capability of the nuclear object to withhold the main safety and reliability criterion within specified range during both normal operation and transient conditions. Three types of assessment are to be performed in order to establish the nuclear power plant safety level: neutronic calculations; thermal hydraulic calculations; mechanical design calculations. Calculations of the thermal hydraulic parameters of the RBMK-1500 reactor main circulation circuit (MCC) are presented in this paper. The aim of this work was to test the capability of the APROS code to simulate the behavior of the RBMK-1500 type reactor main circulation circuit during normal operation and transients. (author)

  18. Modeling thermal stress propagation during hydraulic stimulation of geothermal wells

    Science.gov (United States)

    Jansen, Gunnar; Miller, Stephen A.

    2017-04-01

    A large fraction of the world's water and energy resources are located in naturally fractured reservoirs within the earth's crust. Depending on the lithology and tectonic history of a formation, fracture networks can range from dense and homogeneous highly fractured networks to single large scale fractures dominating the flow behavior. Understanding the dynamics of such reservoirs in terms of flow and transport is crucial to successful application of engineered geothermal systems (also known as enhanced geothermal systems or EGS) for geothermal energy production in the future. Fractured reservoirs are considered to consist of two distinct separate media, namely the fracture and matrix space respectively. Fractures are generally thin, highly conductive containing only small amounts of fluid, whereas the matrix rock provides high fluid storage but typically has much smaller permeability. Simulation of flow and transport through fractured porous media is challenging due to the high permeability contrast between the fractures and the surrounding rock matrix. However, accurate and efficient simulation of flow through a fracture network is crucial in order to understand, optimize and engineer reservoirs. It has been a research topic for several decades and is still under active research. Accurate fluid flow simulations through field-scale fractured reservoirs are still limited by the power of current computer processing units (CPU). We present an efficient implementation of the embedded discrete fracture model, which is a promising new technique in modeling the behavior of enhanced geothermal systems. An efficient coupling strategy is determined for numerical performance of the model. We provide new insight into the coupled modeling of fluid flow, heat transport of engineered geothermal reservoirs with focus on the thermal stress changes during the stimulation process. We further investigate the interplay of thermal and poro-elastic stress changes in the reservoir

  19. Thermal-hydraulic analysis of PWR core including intermediate flow mixers with the THYC code

    Energy Technology Data Exchange (ETDEWEB)

    Mur, J. [Electricite de France (EDF), 78 - Chatou (France); Meignin, J.C. [Electricite de France (EDF), 69 - Villeurbanne (France)

    1997-07-01

    Departure from nucleate boiling (DNB) is one of the major limiting factors of pressurized water reactors (PWRs). Safety requires that occurrence of DNB should be precluded under normal or incidental operating conditions. The thermal-hydraulic THYC code developed by EDF is described. The code is devoted to heat and mass transfer in nuclear components. Critical Heat Flux (CHF) is predicted from local thermal-hydraulic parameters such as pressure, mass flow rate, and quality. A three stage methodology to evaluate thermal margins in order to perform standard core design is described. (K.A.) 8 refs.

  20. Transient thermal hydraulic modeling and analysis of ITER divertor plate system

    Energy Technology Data Exchange (ETDEWEB)

    El-Morshedy, Salah El-Din [Argonne National Laboratory, Argonne, IL (United States); Atomic Energy Authority, Cairo (Egypt)], E-mail: selmorshedy@etrr2-aea.org.eg; Hassanein, Ahmed [Purdue University, West Lafayette, IN (United States)], E-mail: hassanein@purdue.edu

    2009-12-15

    A mathematical model has been developed/updated to simulate the steady state and transient thermal-hydraulics of the International Thermonuclear Experimental Reactor (ITER) divertor module. The model predicts the thermal response of the armour coating, divertor plate structural materials and coolant channels. The selected heat transfer correlations cover all operating conditions of ITER under both normal and off-normal situations. The model also accounts for the melting, vaporization, and solidification of the armour material. The developed model is to provide a quick benchmark of the HEIGHTS multidimensional comprehensive simulation package. The present model divides the coolant channels into a specified axial regions and the divertor plate into a specified radial zones, then a two-dimensional heat conduction calculation is created to predict the temperature distribution for both steady and transient states. The model is benchmarked against experimental data performed at Sandia National Laboratory for both bare and swirl tape coolant channel mockups. The results show very good agreements with the data for steady and transient states. The model is then used to predict the thermal behavior of the ITER plasma facing and structural materials due to plasma instability event where 60 MJ/m{sup 2} plasma energy is deposited over 500 ms. The results for ITER divertor response is analyzed and compared with HEIGHTS results.

  1. Application of computational fluid dynamics methods to improve thermal hydraulic code analysis

    Science.gov (United States)

    Sentell, Dennis Shannon, Jr.

    A computational fluid dynamics code is used to model the primary natural circulation loop of a proposed small modular reactor for comparison to experimental data and best-estimate thermal-hydraulic code results. Recent advances in computational fluid dynamics code modeling capabilities make them attractive alternatives to the current conservative approach of coupled best-estimate thermal hydraulic codes and uncertainty evaluations. The results from a computational fluid dynamics analysis are benchmarked against the experimental test results of a 1:3 length, 1:254 volume, full pressure and full temperature scale small modular reactor during steady-state power operations and during a depressurization transient. A comparative evaluation of the experimental data, the thermal hydraulic code results and the computational fluid dynamics code results provides an opportunity to validate the best-estimate thermal hydraulic code's treatment of a natural circulation loop and provide insights into expanded use of the computational fluid dynamics code in future designs and operations. Additionally, a sensitivity analysis is conducted to determine those physical phenomena most impactful on operations of the proposed reactor's natural circulation loop. The combination of the comparative evaluation and sensitivity analysis provides the resources for increased confidence in model developments for natural circulation loops and provides for reliability improvements of the thermal hydraulic code.

  2. Status and subjects of thermal-hydraulic analysis for next-generation LWRs

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2000-03-01

    The status and subjects on thermal-hydraulic analysis for next-generation light water reactors (LWRs) with passive safety systems were surveyed through about 5 years until March 1999 by subcommittee on improvement of reactor thermal-hydraulic analysis codes under the nuclear code committee in Japan Atomic Energy Research Institute. Based on the survey results and discussion, the status and subjects on system analysis for various types of proposed reactor were summarized in 1998 and those on multidimensional two-phase flow analysis were also reviewed, since the multidimensional analysis was recognized as one of the most important subjects through the investigation on system analysis. In this report, the status and subjects for the following were summarized from the survey results and discussion in 1998 and 1999; (1) BWR neutronic/thermal-hydraulic coupled analysis, (2) Evaluation of passive safety system performance and (3) Gas-liquid two-phase flow analysis. The contents in this report are the forefront of thermal-hydraulic analysis for LWRs including test results from several large-scale facilities. We expect that the contents can offer a guideline to improve reactor thermal-hydraulic analysis codes in future. (author)

  3. Thermal-Hydraulic Analysis Tasks for ANAV NPPs in Support of Plant Operation and Control

    Directory of Open Access Journals (Sweden)

    L. Batet

    2007-11-01

    Full Text Available Thermal-hydraulic analysis tasks aimed at supporting plant operation and control of nuclear power plants are an important issue for the Asociación Nuclear Ascó-Vandellòs (ANAV. ANAV is the consortium that runs the Ascó power plants (2 units and the Vandellòs-II power plant. The reactors are Westinghouse-design, 3-loop PWRs with an approximate electrical power of 1000 MW. The Technical University of Catalonia (UPC thermal-hydraulic analysis team has jointly worked together with ANAV engineers at different levels in the analysis and improvement of these reactors. This article is an illustration of the usefulness of computational analysis for operational support. The contents presented were operational between 1985 and 2001 and subsequently changed slightly following various organizational adjustments. The paper has two different parts. In the first part, it describes the specific aspects of thermal-hydraulic analysis tasks related to operation and control and, in the second part, it briefly presents the results of three examples of analyses that were performed. All the presented examples are related to actual situations in which the scenarios were studied by analysts using thermal-hydraulic codes and prepared nodalizations. The paper also includes a qualitative evaluation of the benefits obtained by ANAV through thermal-hydraulic analyses aimed at supporting operation and plant control.

  4. Development of system analysis code for thermal-hydraulic simulation of integral reactor, Rex-10

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2010-10-15

    Rex-10 is an environment-friendly and economical small-scale nuclear reactor to provide the energy for district heating as well as the electric power in micro-grid. This integral reactor comprises several innovative concepts supported by advanced primary circuit components, low coolant parameters and natural circulation cooling. To evaluate the system performance and thermal-hydraulic behavior of the reactor, a system analysis code is being developed so that the new designs and technologies adopted in Rex-10 can be reflected. The research efforts are absorbed in programming the simple and fast-running thermal-hydraulic analysis software. The details of hydrodynamic governing equations component models and numerical solution scheme used in this code are presented in this paper. On the basis of one-dimensional momentum integral model, the models of point reactor neutron kinetics for thorium-fueled core, physical processes in the steam-gas pressurizer, and heat transfers in helically coiled steam generator are implemented to the system code. Implicit numerical scheme is employed to momentum and energy equations to assure the numerical stability. The accuracy of simulation is validated by applying the solution method to the Rex-10 test facility. Calculated natural circulation flow rate and coolant temperature at steady-state are compared to the experimental data. The validation is also carried out for the transients in which the sudden reduction in the core power or the feedwater flow takes place. The code's capability to predict the steady-state flow by natural convection and the qualitative behaviour of the primary system in the transients is confirmed. (Author)

  5. Current and anticipated uses of the thermal hydraulics codes at the NRC

    Energy Technology Data Exchange (ETDEWEB)

    Caruso, R.

    1997-07-01

    The focus of Thermal-Hydraulic computer code usage in nuclear regulatory organizations has undergone a considerable shift since the codes were originally conceived. Less work is being done in the area of {open_quotes}Design Basis Accidents,{close_quotes}, and much more emphasis is being placed on analysis of operational events, probabalistic risk/safety assessment, and maintenance practices. All of these areas need support from Thermal-Hydraulic computer codes to model the behavior of plant fluid systems, and they all need the ability to perform large numbers of analyses quickly. It is therefore important for the T/H codes of the future to be able to support these needs, by providing robust, easy-to-use, tools that produce easy-to understand results for a wider community of nuclear professionals. These tools need to take advantage of the great advances that have occurred recently in computer software, by providing users with graphical user interfaces for both input and output. In addition, reduced costs of computer memory and other hardware have removed the need for excessively complex data structures and numerical schemes, which make the codes more difficult and expensive to modify, maintain, and debug, and which increase problem run-times. Future versions of the T/H codes should also be structured in a modular fashion, to allow for the easy incorporation of new correlations, models, or features, and to simplify maintenance and testing. Finally, it is important that future T/H code developers work closely with the code user community, to ensure that the code meet the needs of those users.

  6. Validation of CESAR Thermal-hydraulic Module of ASTEC V1.2 Code on BETHSY Experiments

    Science.gov (United States)

    Tregoures, Nicolas; Bandini, Giacomino; Foucher, Laurent; Fleurot, Joëlle; Meloni, Paride

    The ASTEC V1 system code is being jointly developed by the French Institut de Radioprotection et Sûreté Nucléaire (IRSN) and the German Gesellschaft für Anlagen und ReaktorSicherheit (GRS) to address severe accident sequences in a nuclear power plant. Thermal-hydraulics in primary and secondary system is addressed by the CESAR module. The aim of this paper is to present the validation of the CESAR module, from the ASTEC V1.2 version, on the basis of well instrumented and qualified integral experiments carried out in the BETHSY facility (CEA, France), which simulates a French 900 MWe PWR reactor. Three tests have been thoroughly investigated with CESAR: the loss of coolant 9.1b test (OECD ISP N° 27), the loss of feedwater 5.2e test, and the multiple steam generator tube rupture 4.3b test. In the present paper, the results of the code for the three analyzed tests are presented in comparison with the experimental data. The thermal-hydraulic behavior of the BETHSY facility during the transient phase is well reproduced by CESAR: the occurrence of major events and the time evolution of main thermal-hydraulic parameters of both primary and secondary circuits are well predicted.

  7. Improving Neutron Kinetics and Thermal Hydraulics coupled tools for BEPU calculations

    Energy Technology Data Exchange (ETDEWEB)

    Pericas, R.; Reventós, F.; Batet, Il.

    2015-07-01

    The BEPU methodology is capable of providing a solution in terms of increasing the nuclear power production without compromising the safety margins. This study presents different improvements performed using tools available at UPC in the field of Neutron Kinetics and Thermal Hydraulics coupled systems. The paper describes a comparison between the BEPU methodology and the Conservative Bounding methodology within the framework of the Neutron Kinetics and Thermal Hydraulics coupled systems. To perform such comparison the following tools have been selected: TRACE for thermal-hydraulic system calculations, PARCS for reactor kinetics core simulator code. A Main Steam Line Break (MSLB) in a Pressurized Water Reactor (PWR) is the selected simulated transient to show the improvements performed. (Author)

  8. A comparative assessment of independent thermal-hydraulic models for research reactors: The RSG-GAS case

    Energy Technology Data Exchange (ETDEWEB)

    Chatzidakis, S., E-mail: schatzid@purdue.edu [Purdue University, School of Nuclear Engineering, West Lafayette, IN 47907 (United States); Hainoun, A. [Atomic Energy Commission of Syria (AECS), Nuclear Engineering Department, P.O. Box 6091, Damascus (Syrian Arab Republic); Doval, A. [Nuclear Engineering Department, Av. Cmdt. Luis Piedrabuena 4950, C.P. 8400, San Carlos de Bariloche, Rio Negro (Argentina); Alhabet, F. [Atomic Energy Commission of Syria (AECS), Nuclear Engineering Department, P.O. Box 6091, Damascus (Syrian Arab Republic); Francioni, F. [Nuclear Engineering Department, Av. Cmdt. Luis Piedrabuena 4950, C.P. 8400, San Carlos de Bariloche, Rio Negro (Argentina); Ikonomopoulos, A. [Institute of Nuclear and Radiological Sciences, Energy, Technology and Safety, National Center for Scientific Research ‘Demokritos’, 15130, Aghia Paraskevi, Athens (Greece); Ridikas, D. [Department of Nuclear Sciences and Applications, International Atomic Energy Agency, Vienna International Centre, A-1400 Vienna (Austria)

    2014-03-15

    Highlights: • Increased use of thermal-hydraulic codes requires assessment of important phenomena in RRs. • Three independent modeling teams performed analysis of loss of flow transient. • Purpose of this work is to examine the thermal-hydraulic codes response. • To perform benchmark analysis comparing the different codes with experimental measurements. • To identify the impact of the user effect on the computed results, performed with the same codes. - Abstract: This study presents the comparative assessment of three thermal-hydraulic codes employed to model the Indonesian research reactor (RSG-GAS) and simulate the reactor behavior under steady state and loss of flow transient (LOFT). The RELAP5/MOD3, MERSAT and PARET-ANL thermal-hydraulic codes are used by independent research groups to perform benchmark analysis against measurements of coolant and clad temperatures, conducted on an instrumented fuel element inside RSG-GAS core. The results obtained confirm the applicability of RELAP5/MOD3, MERSAT and PARET-ANL on the modeling of loss of flow transient in research reactors. In particular, the three codes are able to simulate flow reversal from downward forced to upward natural convection after pump trip and successful reactor scram. The benchmark results show that the codes predict maximum clad temperature of hot channel conservatively with a maximum overestimation of 27% for RELAP5/MOD3, 17% for MERSAT and 8% for PARET-ANL. As an additional effort, the impact of user effect on the simulation results has been assessed for the code RELAP5/MOD3, where the main differences among the models are presented and discussed.

  9. SUPERENERGY-2: a multiassembly, steady-state computer code for LMFBR core thermal-hydraulic analysis

    Energy Technology Data Exchange (ETDEWEB)

    Basehore, K.L.; Todreas, N.E.

    1980-08-01

    Core thermal-hydraulic design and performance analyses for Liquid Metal Fast Breeder Reactors (LMFBRs) require repeated detailed multiassembly calculations to determine radial temperature profiles and subchannel outlet temperatures for various core configurations and subassembly structural analyses. At steady-state, detailed core-wide temperature profiles are required for core restraint calculations and subassembly structural analysis. In addition, sodium outlet temperatures are routinely needed for each reactor operating cycle. The SUPERENERGY-2 thermal-hydraulic code was designed specifically to meet these designer needs. It is applicable only to steady-state, forced-convection flow in LMFBR core geometries.

  10. Thermal-hydraulic analysis of a cylindrical blanket module using ATHENA code

    Energy Technology Data Exchange (ETDEWEB)

    Hwang, J.G.; Herring, J.S.; Carlson, K.E.; Ransom, V.H.

    1981-01-01

    ATHENA (Advanced Thermal-Hydraulic Energy Network Analyzer) is a new computer code for thermal-hydraulic analyses of many energy systems. Multiple-loop and multiple-fluid capabilities have been emphasized during the code development. A pilot version of ATHENA has incorporated a fusion kinetic package to model the effect of first wall temperature variation on the reactor conditions. The capability has been demonstrated by analyzing the performance under various conditions of a cylindrical fusion blanket module. The results have shown the viability of using ATHENA for fusion reactor design and safety analyses.

  11. Neutronic and thermal hydraulic analysis for production of fission molybdenum-99 at Pakistan Research Reactor-1

    Energy Technology Data Exchange (ETDEWEB)

    Mushtaq, A. [Isotope Production Division, Pakistan Institute of Nuclear Science and Technology, P.O. Nilore, Islamabad (Pakistan)], E-mail: mushtaqa@pinstech.org.pk; Iqbal, Massod; Bokhari, Ishtiaq Hussain; Mahmood, Tariq; Mahmood, Tayyab; Ahmad, Zahoor; Zaman, Qamar [Nuclear Engineering Division, Pakistan Institute of Nuclear Science and Technology, P.O. Nilore, Islamabad (Pakistan)

    2008-02-15

    Neutronic and thermal hydraulic analysis for the fission molybdenum-99 production at PARR-1 has been performed. Low enriched uranium foil (<20% {sup 235}U) will be used as target material. Annular target designed by ANL (USA) will be irradiated in PARR-1 for the production of 100 Ci of molybdenum-99 at the end of irradiation, which will be sufficient to prepare required {sup 99}Mo/{sup 99m}Tc generators at PINSTECH and its supply in the country. Neutronic and thermal hydraulic analysis were performed using various codes. Data shows that annular targets can be safely irradiated in PARR-1 for production of required amount of fission molybdenum-99.

  12. Thermal-hydraulic simulation of mercury target concepts for a pulsed spallation neutron source

    Energy Technology Data Exchange (ETDEWEB)

    Siman-Tov, M.; Wendel, M.; Haines, J. [Oak Ridge National Lab., TN (United States)

    1996-06-01

    The Oak Ridge Spallation Neutron Source (ORSNS) is a high-power, accelerator-based pulsed spallation neutron source being designed by a multi-laboratory team led by Oak Ridge National Laboratory to achieve very high fluxes of neutrons for scientific experiments. The ORSNS is projected to have a 1 MW proton beam upgradable to 5 MW. About 60% of the beam power (1-5 MW, 17-83 kJ/pulse in 0.5 microsec at 60 cps) is deposited in the liquid metal (mercury) target having the dimensions of 65x30x10 cm (about 19.5 liter). Peak steady state power density is about 150 and 785 MW/m{sup 3} for 1 MW and 5 MW beam respectively, whereas peak pulsed power density is as high as 5.2 and 26.1 GW/m{sup 3}, respectively. The peak pulse temperature rise rate is 14 million C/s (for 5 MW beam) whereas the total pulse temperature rise is only 7 C. In addition to thermal shock and materials compatibility, key feasibility issues for the target are related to its thermal-hydraulic performance. This includes proper flow distribution, flow reversals, possible {open_quotes}hot spots{close_quotes} and the challenge of mitigating the effects of thermal shock through possible injection of helium bubbles throughout the mercury volume or other concepts. The general computational fluid dynamics (CFD) code CFDS-FLOW3D was used to simulate the thermal and flow distribution in three preliminary concepts of the mercury target. Very initial CFD simulation of He bubbles injection demonstrates some potential for simulating behavior of He bubbles in flowing mercury. Much study and development will be required to be able to `predict`, even in a crude way, such a complex phenomena. Future direction in both design and R&D is outlined.

  13. Thermal hydraulic-Mechanic Integrated Simulation for Advanced Cladding Thermal Shock Fracture Analysis during Reflood Phase in LBLOCA

    Energy Technology Data Exchange (ETDEWEB)

    Son, Seong Min; Lee, You Ho; Cho, Jae Wan; Lee, Jeong Ik [KAERI, Daejeon (Korea, Republic of)

    2016-05-15

    This study suggested thermal hydraulic-mechanical integrated stress based methodology for analyzing the behavior of ATF type claddings by SiC-Duplex cladding LBLOCA simulation. Also, this paper showed that this methodology could predict real experimental result well. That concept for enhanced safety of LWR called Advanced Accident-Tolerance Fuel Cladding (ATF cladding, ATF) is researched actively. However, current nuclear fuel cladding design criteria for zircaloy cannot be apply to ATF directly because those criteria are mainly based on limiting their oxidation. So, the new methodology for ATF design criteria is necessary. In this study, stress based analysis methodology for ATF cladding design criteria is suggested. By simulating LBLOCA scenario of SiC cladding which is the one of the most promising candidate of ATF. Also we'll confirm our result briefly through comparing some facts from other experiments. This result is validating now. Some of results show good performance with 1-D failure analysis code for SiC fuel cladding that already developed and validated by Lee et al,. It will present in meeting. Furthermore, this simulation presented the possibility of understanding the behavior of cladding deeper. If designer can predict the dangerous region and the time precisely, it may be helpful for designing nuclear fuel cladding geometry and set safety criteria.

  14. Deterministic and Monte Carlo transport models with thermal-hydraulic feedback

    Energy Technology Data Exchange (ETDEWEB)

    Seubert, A.; Langenbuch, S.; Velkov, K.; Zwermann, W. [Gesellschaft fuer Anlagen- und Reaktorsicherheit (GRS) mbH, Garching (Germany)

    2008-07-01

    This paper gives an overview of recent developments concerning deterministic transport and Monte Carlo methods with thermal-hydraulic feedback. The timedependent 3D discrete ordinates transport code TORT-TD allows pin-by-pin analyses of transients using few energy groups and anisotropic scattering by solving the timedependent transport equation using the unconditionally stable implicit method. To account for thermal-hydraulic feedback, TORT-TD has been coupled with the system code ATHLET. Applications to, e.g., a control rod ejection in a 2 x 2 PWR fuel assembly arrangement demonstrate the applicability of the coupled code TORT-TD/ATHLET for test cases. For Monte Carlo steady-state calculations with nuclear point data and thermalhydraulic feedback, MCNP has been prepared to incorporate thermal-hydraulic parameters. As test case has been chosen the uncontrolled steady state of the 2 x 2 PWR fuel assembly arrangement for which the thermal-hydraulic parameter distribution has been obtained from a preceding coupled TORT-TD/ATHLET analysis. The result demonstrates the applicability of MCNP to problems with spatial distributions of thermal-fluiddynamic parameters. The comparison with MCNP results confirms that the accuracy of deterministic transport calculations with pin-wise homogenised few-group cross sections is comparable to Monte Carlo simulations. The presented cases are considered as a pre-stage of performing calculations of larger configurations like a quarter core which is in preparation. (orig.)

  15. Hydraulic Behavior in The Downflow Hanging Sponge Bioreactor

    Directory of Open Access Journals (Sweden)

    Izarul Machdar

    2016-12-01

    Full Text Available Performance efficiency in a Downflow Hanging Sponge (DHS bioreactor is associated with the amount of time that a wastewater remains in the bioreactor. The bioreactor is considered as a plug flow reactor and its hydraulic residence time (HRT depends on the void volume of packing material and the flow rate. In this study, hydraulic behavior of DHS bioreactor was investigated by using tracer method. Two types of sponge module covers, cylindrical plastic frame (module-1 and plastic hair roller (module-2, were investigated and compared. A concentrated NaCl solution used as an inert tracer and input as a pulse at the inlet of DHS bioreactor. Analysis of the residence time distribution (RTD curves provided interpretation of the index distribution or holdup water (active volume, the degree of short-circuiting, number of tanks in series (the plug flow characteristic, and the dispersion number. It was found that the actual HRT was primarily shorter than theoretical HRT of each test. Holdup water of the DHS bioreactor ranged from 60% to 97% and 36% to 60% of module-1 and module-2, respectively. Eventhough module-1 has higher effective volume than module-2, result showed that the dispersion numbers of the two modules were not significant difference. Furthermore, N-values were found larger at a higher flow rate. It was concluded that a DHS bioreactor design should incorporated a combination of water distributor system, higher loading rate at startup process to generate a hydraulic behavior closer to an ideal plug flow.ABSTRAKEfisiensi unjuk kerja bioreactor Downflow Hanging Sponge (DHS berkaitan dengan lamanya waktu tinggal limbah berada di dalam bioreaktor tersebut. Bioreaktor DHS dianggap sebagai seuatu reaktor aliran sumbat (plug flow dimana waktu tinggal hidraulik (HRT tergantung pada volume pori material isian dan laju alir. Dua jenis modul digunakan dalam penelitian ini, yang diberi nama dengan module-1 dan module-2 untuk melihat pengaruh jenis modul

  16. Hydraulic assessment of the Buda Thermal Karst area and its vulnerability (Budapest, Hungary)

    Science.gov (United States)

    Czauner, Brigitta; Erőss, Anita; Erhardt, Ildikó; Ötvös, Viktória; Simon, Szilvia; Mádl-Szőnyi, Judit

    2017-04-01

    Thermal and medicinal water resources of Budapest (Hungary), the "City of Spas", are provided by the Buda Thermal Karst area. Assessment of its vulnerability requires the understanding of the discharge phenomena and thus the groundwater flow conditions in the area. Accordingly, BTK has already been the objective of several hydrogeological investigations, including numerical simulations as well, which led to conceptual models. The aim of the present study was the hydraulic evaluation of the flow systems based on the complex analysis of real, i.e. measured, archival hydraulic data of wells in order to i) get acquainted with the real flow systems, and ii) hydraulically confirm or disprove the previous conceptual models, in particular the applicability of gravity-driven regional groundwater flow concept and hydraulic continuity, separation of the natural discharge zones, and hypogenic karstification. Considering the data distribution, pressure vs. elevation profiles, tomographic fluid-potential maps, and hydraulic cross-sections were constructed for the first time in this area. As a result, gravitational flow systems and the modifying effects of aquitard units and faults were identified. Consequently, the differences in temperature, hydrochemistry, discharge distribution (one and two-components), and related cave forming processes between the Central (Rózsadomb) and Southern (Gellért Hill) natural discharge areas could be explained, as well as the hydraulic behaviour of the Northeastern Margin-fault of the Buda Hills could be determined. Regarding the on-going hypogenic karstification processes, regional upward flow conditions were confirmed along the main discharge zone of the Danube. Identification of gravity as the main fluid flow driving force, as well as the hydraulic effects of heterogeneities can significantly contribute to the recognition of the risk factors regarding the vulnerability of the Buda Thermal Karst area. The research was supported by the

  17. Thermal-hydraulic analysis of LTS cables for the DEMO TF coil using simplified models

    Directory of Open Access Journals (Sweden)

    Lewandowska Monika

    2017-03-01

    Full Text Available The conceptual design activities for the DEMOnstration reactor (DEMO – the prototype fusion power plant – are conducted in Europe by the EUROfusion Consortium. In 2015, three design concepts of the DEMO toroidal field (TF coil were proposed by Swiss Plasma Center (EPFL-SPC, PSI Villigen, Italian National Agency for New Technologies (ENEA Frascati, and Atomic Energy and Alternative Energies Commission (CEA Cadarache. The proposed conductor designs were subjected to complete mechanical, electromagnetic, and thermal-hydraulic analyses. The present study is focused on the thermal-hydraulic analysis of the candidate conductor designs using simplified models. It includes (a hydraulic analysis, (b heat removal analysis, and (c assessment of the maximum temperature and the maximum pressure in each conductor during quench. The performed analysis, aimed at verification whether the proposed design concepts fulfil the established acceptance criteria, provides the information for further improvements of the coil and conductors design.

  18. Thermal Hydraulic Characteristics of Fuel Defects in Plate Type Nuclear Research Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Bodey, Isaac T [ORNL

    2014-05-01

    Turbulent flow coupled with heat transfer is investigated for a High Flux Isotope Reactor (HFIR) fuel plate. The Reynolds Averaged Navier-Stokes Models are used for fluid dynamics and the transfer of heat from a thermal nuclear fuel plate using the Multi-physics code COMSOL. Simulation outcomes are compared with experimental data from the Advanced Neutron Source Reactor Thermal Hydraulic Test Loop. The computational results for the High Flux Isotope Reactor core system provide a more physically accurate simulation of this system by modeling the turbulent flow field in conjunction with the diffusion of thermal energy within the solid and fluid phases of the model domain. Recommendations are made regarding Nusselt number correlations and material properties for future thermal hydraulic modeling efforts

  19. Analytical Thermal Field Theory Applicable to Oil Hydraulic Fluid Film Lubrication

    DEFF Research Database (Denmark)

    Johansen, Per; Roemer, Daniel Beck; Pedersen, Henrik C.

    2014-01-01

    An analytical thermal field theory is derived by a perturbation series expansion solution to the energy conservation equation. The theory is valid for small values of the Brinkman number and the modified Peclet number. This condition is sufficiently satisfied for hydraulic oils, whereby...

  20. Thermal hydraulic studies of spallation target for one-way coupled ...

    Indian Academy of Sciences (India)

    pp. 355–363. Thermal hydraulic studies of spallation target for one-way coupled Indian accelerator driven systems with low energy proton beam. V MANTHA1, A K MOHANTY2 and P SATYAMURTHY1. 1Laser and Plasma Technology Division; 2Nuclear Physics Division, Bhabha Atomic. Research Centre, Mumbai 400 085, ...

  1. Proceedings of the OECD/CSNI workshop on transient thermal-hydraulic and neutronic codes requirements

    Energy Technology Data Exchange (ETDEWEB)

    Ebert, D.

    1997-07-01

    This is a report on the CSNI Workshop on Transient Thermal-Hydraulic and Neutronic Codes Requirements held at Annapolis, Maryland, USA November 5-8, 1996. This experts` meeting consisted of 140 participants from 21 countries; 65 invited papers were presented. The meeting was divided into five areas: (1) current and prospective plans of thermal hydraulic codes development; (2) current and anticipated uses of thermal-hydraulic codes; (3) advances in modeling of thermal-hydraulic phenomena and associated additional experimental needs; (4) numerical methods in multi-phase flows; and (5) programming language, code architectures and user interfaces. The workshop consensus identified the following important action items to be addressed by the international community in order to maintain and improve the calculational capability: (a) preserve current code expertise and institutional memory, (b) preserve the ability to use the existing investment in plant transient analysis codes, (c) maintain essential experimental capabilities, (d) develop advanced measurement capabilities to support future code validation work, (e) integrate existing analytical capabilities so as to improve performance and reduce operating costs, (f) exploit the proven advances in code architecture, numerics, graphical user interfaces, and modularization in order to improve code performance and scrutibility, and (g) more effectively utilize user experience in modifying and improving the codes.

  2. A review on the thermal hydraulic characteristics of the air-cooled ...

    Indian Academy of Sciences (India)

    Home; Journals; Sadhana; Volume 40; Issue 3. A review on the thermal hydraulic characteristics of the air-cooled heat exchangers in forced convection. Ankur Kumar Jyeshtharaj B Joshi Arun K Nayak Pallippattu K Vijayan. Section I – Fluid Mechanics and Fluid Power (FMFP) Volume 40 Issue 3 May 2015 pp 673-755 ...

  3. Advances in thermal hydraulic and neutronic simulation for reactor analysis and safety

    Energy Technology Data Exchange (ETDEWEB)

    Tentner, A.M.; Blomquist, R.N.; Canfield, T.R.; Ewing, T.F.; Garner, P.L.; Gelbard, E.M.; Gross, K.C.; Minkoff, M.; Valentin, R.A.

    1993-03-01

    This paper describes several large-scale computational models developed at Argonne National Laboratory for the simulation and analysis of thermal-hydraulic and neutronic events in nuclear reactors and nuclear power plants. The impact of advanced parallel computing technologies on these computational models is emphasized.

  4. Thermal-hydraulically corrected neutron cross-sections for PWR reactors

    Energy Technology Data Exchange (ETDEWEB)

    Santiago, Daniela M.N.; Alvim, Antonio C.M.; Silva, Fernando C., E-mail: dsantiago@con.ufrj.b, E-mail: alvim@con.ufrj.b, E-mail: fernando@con.ufrj.b [Universidade Federal do Rio de Janeiro (PEN/COPPE/UFRJ), RJ (Brazil). Coordenacao dos Programas de Pos-Graduacao de Engenharia. Programa de Engenharia Nuclear

    2011-07-01

    Reactor core simulation codes ought to have a thermal-hydraulics feedback module. This module calculates, among other effects, the fuel temperature thermal-hydraulics feedback, that corrects neutron cross sections. In the nodal code developed at PEN/COPPE/UFRJ, the fuel temperature is obtained through an empirical model. This work presents a physical model to calculate this temperature. A finite volume technique was used to discretize the equation for temperature distribution, while the moderator coefficient of heat transfer was calculated using ASME routines, appended to the developed code. This model allows calculation of an average radial temperature per node, since the thermal-hydraulic feedback must follow the conditions imposed by the nodal code. The results obtained were compared with the ones obtained by the empirical model. The results show that, for fuel elements near core periphery, the empirical model overestimates the temperature in the fuel, as compared to our model, which may indicate that the physical model is more appropriate to calculate the thermal-hydraulic feedback temperatures. (author)

  5. Thermal-hydraulics analysis of a PWR reactor using zircaloy and carbide silicon reinforced with type S fibers as fuel claddings: Simulation of a channel blockage transient

    Energy Technology Data Exchange (ETDEWEB)

    Matuck, Vinicius; Ramos, Mario C.; Faria, Rochkhudson B.; Reis, Patricia A.L.; Costa, Antonella L.; Pereira, Claubia, E-mail: rochkdefaria@yahoo.com.br, E-mail: matuck747@gmail.com, E-mail: patricialire@yahoo.com.br, E-mail: marc5663@gmail.com, E-mail: antonella@nuclear.ufmg.br, E-mail: claubia@nuclear.ufmg.br [Universidade Federal de Minas Gerais (UFMG), Belo Horizonte (Brazil). Departamento de Engenharia Nuclear

    2017-11-01

    A detailed thermal-hydraulic reactor model using as reference data from the Angra 2 Final Safety Analysis Report (FSAR) has been developed and SiC reinforced with Hi-Nicalon type S fibers (SiC HNS) was used as fuel cladding. The goal is to compare its behavior from the thermal viewpoint with the Zircaloy, at the steady- state and transient conditions. The RELAP-3D was used to perform the thermal-hydraulic analysis and a blockage transient has been investigated at full power operation. The transient considered is related to total obstruction of a core cooling channel of one fuel assembly. The calculations were performed using a point kinetic model. The reactor behavior after this transient was analyzed and the time evolution of cladding and coolant temperatures mass flow and void fraction are presented. (author)

  6. Design and Construction of Experiment for Direct Electron Irradiation of Uranyl Sulfate Solution: Bubble Formation and Thermal Hydraulics Studies

    Energy Technology Data Exchange (ETDEWEB)

    Chemerisov, Sergey [Argonne National Lab. (ANL), Argonne, IL (United States); Gromov, Roman [Argonne National Lab. (ANL), Argonne, IL (United States); Makarashvili, Vakho [Argonne National Lab. (ANL), Argonne, IL (United States); Heltemes, Thad [Argonne National Lab. (ANL), Argonne, IL (United States); Sun, Zaijing [Argonne National Lab. (ANL), Argonne, IL (United States); Wardle, Kent E. [Argonne National Lab. (ANL), Argonne, IL (United States); Bailey, James [Argonne National Lab. (ANL), Argonne, IL (United States); Quigley, Kevin [Argonne National Lab. (ANL), Argonne, IL (United States); Stepinski, Dominique [Argonne National Lab. (ANL), Argonne, IL (United States); Vandegrift, George [Argonne National Lab. (ANL), Argonne, IL (United States)

    2014-10-01

    Argonne is assisting SHINE Medical Technologies in developing SHINE, a system for producing fission-product 99Mo using a D/T-accelerator to produce fission in a non-critical target solution of aqueous uranyl sulfate. We have developed an experimental setup for studying thermal-hydraulics and bubble formation in the uranyl sulfate solution to simulate conditions expected in the SHINE target solution during irradiation. A direct electron beam from the linac accelerator will be used to irradiate a 20 L solution (sector of the solution vessel). Because the solution will undergo radiolytic decomposition, we will be able to study bubble formation and dynamics and effects of convection and temperature on bubble behavior. These experiments will serve as a verification/ validation tool for the thermal-hydraulic model. Utilization of the direct electron beam for irradiation allows homogeneous heating of a large solution volume and simplifies observation of the bubble dynamics simultaneously with thermal-hydraulic data collection, which will complement data collected during operation of the miniSHINE experiment. Irradiation will be conducted using a 30-40 MeV electron beam from the high-power linac accelerator. The total electron-beam power will be 20 kW, which will yield a power density on the order of 1 kW/L. The solution volume will be cooled on the front and back surfaces and central tube to mimic the geometry of the proposed SHINE solution vessel. Also, multiple thermocouples will be inserted into the solution vessel to map thermal profiles. The experimental design is now complete, and installation and testing are in progress.

  7. Geomechanical, Hydraulic and Thermal Characteristics of Deep Oceanic Sandy Sediments Recovered during the Second Ulleung Basin Gas Hydrate Expedition

    Directory of Open Access Journals (Sweden)

    Yohan Cha

    2016-09-01

    Full Text Available This study investigates the geomechanical, hydraulic and thermal characteristics of natural sandy sediments collected during the Ulleung Basin gas hydrate expedition 2, East Sea, offshore Korea. The studied sediment formation is considered as a potential target reservoir for natural gas production. The sediments contained silt, clay and sand fractions of 21%, 1.3% and 77.7%, respectively, as well as diatomaceous minerals with internal pores. The peak friction angle and critical state (or residual state friction angle under drained conditions were ~26° and ~22°, respectively. There was minimal or no apparent cohesion intercept. Stress- and strain-dependent elastic moduli, such as tangential modulus and secant modulus, were identified. The sediment stiffness increased with increasing confining stress, but degraded with increasing strain regime. Variations in water permeability with water saturation were obtained by fitting experimental matric suction-water saturation data to the Maulem-van Genuchen model. A significant reduction in thermal conductivity (from ~1.4–1.6 to ~0.5–0.7 W·m−1·K−1 was observed when water saturation decreased from 100% to ~10%–20%. In addition, the electrical resistance increased quasi-linearly with decreasing water saturation. The geomechanical, hydraulic and thermal properties of the hydrate-free sediments reported herein can be used as the baseline when predicting properties and behavior of the sediments containing hydrates, and when the hydrates dissociate during gas production. The variations in thermal and hydraulic properties with changing water and gas saturation can be used to assess gas production rates from hydrate-bearing deposits. In addition, while depressurization of hydrate-bearing sediments inevitably causes deformation of sediments under drained conditions, the obtained strength and stiffness properties and stress-strain responses of the sedimentary formation under drained loading conditions

  8. Design of a Resistively Heated Thermal Hydraulic Simulator for Nuclear Rocket Reactor Cores

    Science.gov (United States)

    Litchford, Ron J.; Foote, John P.; Ramachandran, Narayanan; Wang, Ten-See; Anghaie, Samim

    2007-01-01

    A preliminary design study is presented for a non-nuclear test facility which uses ohmic heating to replicate the thermal hydraulic characteristics of solid core nuclear reactor fuel element passages. The basis for this testing capability is a recently commissioned nuclear thermal rocket environments simulator, which uses a high-power, multi-gas, wall-stabilized constricted arc-heater to produce high-temperature pressurized hydrogen flows representative of reactor core environments, excepting radiation effects. Initially, the baseline test fixture for this non-nuclear environments simulator was configured for long duration hot hydrogen exposure of small cylindrical material specimens as a low cost means of evaluating material compatibility. It became evident, however, that additional functionality enhancements were needed to permit a critical examination of thermal hydraulic effects in fuel element passages. Thus, a design configuration was conceived whereby a short tubular material specimen, representing a fuel element passage segment, is surrounded by a backside resistive tungsten heater element and mounted within a self-contained module that inserts directly into the baseline test fixture assembly. With this configuration, it becomes possible to create an inward directed radial thermal gradient within the tubular material specimen such that the wall-to-gas heat flux characteristics of a typical fuel element passage are effectively simulated. The results of a preliminary engineering study for this innovative concept are fully summarized, including high-fidelity multi-physics thermal hydraulic simulations and detailed design features.

  9. THR-TH: a high-temperature gas-cooled nuclear reactor core thermal hydraulics code

    Energy Technology Data Exchange (ETDEWEB)

    Vondy, D.R.

    1984-07-01

    The ORNL version of PEBBLE, the (RZ) pebble bed thermal hydraulics code, has been extended for application to a prismatic gas cooled reactor core. The supplemental treatment is of one-dimensional coolant flow in up to a three-dimensional core description. Power density data from a neutronics and exposure calculation are used as the basic information for the thermal hydraulics calculation of heat removal. Two-dimensional neutronics results may be expanded for a three-dimensional hydraulics calculation. The geometric description for the hydraulics problem is the same as used by the neutronics code. A two-dimensional thermal cell model is used to predict temperatures in the fuel channel. The capability is available in the local BOLD VENTURE computation system for reactor core analysis with capability to account for the effect of temperature feedback by nuclear cross section correlation. Some enhancements have also been added to the original code to add pebble bed modeling flexibility and to generate useful auxiliary results. For example, an estimate is made of the distribution of fuel temperatures based on average and extreme conditions regularly calculated at a number of locations.

  10. Development of MCATHAS system of coupled neutronics/thermal-hydraulics in supercritical water reactor

    Energy Technology Data Exchange (ETDEWEB)

    An, P.; Yao, D. [Science and Tech. on Reactor System Design Tech. Laboratory, Chengdu (China)

    2011-07-01

    The MCATHAS system of coupled neutronics/Thermal-hydraulics in supercritical water reactor is described, which considers the mutual influence between the obvious axial and radial evolution of material temperature, water density and the relative power distribution. This system can obtain the main neutronics and thermal parameters along with burn-up. MCATHAS system is parallel processing coupling. The MCNP code is used for neutronics analysis with the continuous cross section library at any temperature calculated by interpolation algorithm; The sub-channel code ATHAS is for thermal-hydraulics analysis and the ORIGEN Code for burn-up calculation. We validate the code with the assembly of HPLWR and analyze the assembly SCLWR- H. (author)

  11. Whole core pin-by-pin coupled neutronic-thermal-hydraulic steady state and transient calculations using COBAYA3 code

    Energy Technology Data Exchange (ETDEWEB)

    Jimenez, J.; Herrero, J. J.; Cuervo, D.; Aragones, J. M., E-mail: jimenez@din.upm.e [Universidad Politecnica de Madrid, Departamento de Ingenieria Nuclear, Jose Gutierrez Abascal No. 2, 28006 Madrid (Spain)

    2010-10-15

    Nowadays coupled 3-dimensional neutron kinetics and thermal-hydraulic core calculations are performed by applying a radial average channel approach using a meshing of one quarter of assembly in the best case. This approach does not take into account the subchannels effects due to the averaging of the physical fields and the loose of heterogeneity in the thermal-hydraulic model. Therefore the models do not have enough resolution to predict those subchannels effects which are important for the fuel design safety margins, because it is in the local scale, where we can search the hottest pellet or the maximum heat flux. The Polytechnic University of Madrid advanced multi-scale neutron-kinetics and thermal-hydraulics methodologies being implemented in COBAYA3 include domain decomposition by alternate core dissections for the local 3-dimensional fine-mesh scale problems (pin cells/subchannels) and an analytical nodal diffusion solver for the coarse mesh scale coupled with the thermal-hydraulic using a model of one channel per assembly or per quarter of assembly. In this work, we address the domain decomposition by the alternate core dissections methodology applied to solve coupled 3-dimensional neutronic-thermal-hydraulic problems at the fine-mesh scale. The neutronic-thermal-hydraulic coupling at the cell-subchannel scale allows the treatment of the effects of the detailed thermal-hydraulic feedbacks on cross-sections, thus resulting in better estimates of the local safety margins at the pin level. (Author)

  12. The 25 kWe solar thermal Stirling hydraulic engine system: Conceptual design

    Science.gov (United States)

    White, Maurice; Emigh, Grant; Noble, Jack; Riggle, Peter; Sorenson, Torvald

    1988-01-01

    The conceptual design and analysis of a solar thermal free-piston Stirling hydraulic engine system designed to deliver 25 kWe when coupled to a 11 meter test bed concentrator is documented. A manufacturing cost assessment for 10,000 units per year was made. The design meets all program objectives including a 60,000 hr design life, dynamic balancing, fully automated control, more than 33.3 percent overall system efficiency, properly conditioned power, maximum utilization of annualized insolation, and projected production costs. The system incorporates a simple, rugged, reliable pool boiler reflux heat pipe to transfer heat from the solar receiver to the Stirling engine. The free-piston engine produces high pressure hydraulic flow which powers a commercial hydraulic motor that, in turn, drives a commercial rotary induction generator. The Stirling hydraulic engine uses hermetic bellows seals to separate helium working gas from hydraulic fluid which provides hydrodynamic lubrication to all moving parts. Maximum utilization of highly refined, field proven commercial components for electric power generation minimizes development cost and risk.

  13. Thermal-hydraulic design concept of the solid-target system of spallation neutron source

    Energy Technology Data Exchange (ETDEWEB)

    Tanaka, F.; Hibiki, T.; Saito, Y.; Takeda, T.; Mishima, K. [Kyoto Univ., Research Reactor Institute (Japan)

    2001-07-01

    In relation to thermal-hydraulic design of the N-Arena solid-target system of the JHF project, heat transfer experiments were performed to obtain experimental data systematically on heat transfer coefficient and CHF for vertical upward and horizontal flows in a thin rectangular channel simulating a coolant channel of the proposed spallation neutron source. Thermal-hydraulic correlations which can be used for design calculations were proposed based on the obtained data. Finally tentative results of feasibility study on maximum beam power which could be attained with a solid target were presented. The result indicated that the condition for the onset of nucleate boiling is the most significant limiting factor to the maximum beam power. (author)

  14. Thermal hydraulic simulations, error estimation and parameter sensitivity studies in Drekar::CFD

    Energy Technology Data Exchange (ETDEWEB)

    Smith, Thomas Michael; Shadid, John N; Pawlowski, Roger P; Cyr, Eric C; Wildey, Timothy Michael

    2014-01-01

    This report describes work directed towards completion of the Thermal Hydraulics Methods (THM) CFD Level 3 Milestone THM.CFD.P7.05 for the Consortium for Advanced Simulation of Light Water Reactors (CASL) Nuclear Hub effort. The focus of this milestone was to demonstrate the thermal hydraulics and adjoint based error estimation and parameter sensitivity capabilities in the CFD code called Drekar::CFD. This milestone builds upon the capabilities demonstrated in three earlier milestones; THM.CFD.P4.02 [12], completed March, 31, 2012, THM.CFD.P5.01 [15] completed June 30, 2012 and THM.CFD.P5.01 [11] completed on October 31, 2012.

  15. Current and anticipated uses of thermal-hydraulic codes in Germany

    Energy Technology Data Exchange (ETDEWEB)

    Teschendorff, V.; Sommer, F.; Depisch, F.

    1997-07-01

    In Germany, one third of the electrical power is generated by nuclear plants. ATHLET and S-RELAP5 are successfully applied for safety analyses of the existing PWR and BWR reactors and possible future reactors, e.g. EPR. Continuous development and assessment of thermal-hydraulic codes are necessary in order to meet present and future needs of licensing organizations, utilities, and vendors. Desired improvements include thermal-hydraulic models, multi-dimensional simulation, computational speed, interfaces to coupled codes, and code architecture. Real-time capability will be essential for application in full-scope simulators. Comprehensive code validation and quantification of uncertainties are prerequisites for future best-estimate analyses.

  16. Development of best estimate auditing code for CANDU thermal hydraulic safety analysis

    Energy Technology Data Exchange (ETDEWEB)

    Chung, B. D.; Lee, W. J.; Lim, H. S. [Korea Atomic Energy Research Institute, Taejeon (Korea)

    1998-04-01

    The main purpose of this study is to develop a thermal hydraulic auditing code for the CANDU reactor, modifying the model of existing PWR auditing tool, i.e. RELAP5/MOD3. This scope of project is first step of the whole project, thus focus to the establishment of improvement area. The study was performed by reconsideration of the previous code assessment works and investigation of AECL design analysis tools. In order to identify the thermal hydraulic phenomena for events, the whole system of CANDU plant was divided into main functional systems and subcomponents. Each phenomena was addressed to the each subcomponent. Finally improvement areas of model development for auditing tool were established based on the identified phenomena. 8 refs., 21 figs., 19 tabs. (Author)

  17. The SESAME project. State of the art liquid metal thermal hydraulics and beyond

    Energy Technology Data Exchange (ETDEWEB)

    Roelofs, F.; Shams, A.; Batta, A.; Moreau, V.; Di Piazza, I.; Gerschenfeld, A.; Planquart, P.; Tarantino, M. [Nuclear Research and Consultancy Group (NRG), Petten (Netherlands)

    2017-08-15

    The European Sustainable Nuclear Industry Initiative (ESNII) aims at industrial application of fast reactor technology for a sustainable nuclear energy production. Currently four demonstration projects have a promising outlook in Europe, i.e. the ASTRID project in France, the MYRRHA project in Belgium, the ALFRED pan-European project to be realized in Romania, and SEALER in Sweden. Sodium and lead(-alloys) are envisaged as coolants for these reactors. Obviously, in the development of these reactors, thermal-hydraulics is recognized as a key challenge with emphasis on safety issues. This paper discusses the state-of-the-art knowledge with respect to experiments and simulation techniques as pursued in the Horizon 2020 SESAME (thermal hydraulics Simulations and Experiments for the Safety Assessment of MEtal cooled reactors) project.

  18. ATHENA (Advanced Thermal Hydraulic Energy Network Analyzer) solutions to developmental assessment problems

    Energy Technology Data Exchange (ETDEWEB)

    Carlson, K.E.; Ransom, V.H.; Roth, P.A.

    1987-03-01

    The ATHENA (Advanced Thermal Hydraulic Energy Network Analyzer) code has been developed to perform transient simulation of the thermal hydraulic systems that may be found in fusion reactors, space reactors, and other advanced systems. As an assessment of current capability the code was applied to a number of physical problems, both conceptual and actual experiments. Results indicate that the numerical solution to the basic conservation equations is technically sound, and that generally good agreement can be obtained when modeling relevant hydrodynamic experiments. The assessment also demonstrates basic fusion system modeling capability and verifies compatibility of the code with both CDC and CRAY mainframes. Areas where improvements could be made include constitutive modeling, which describes the interfacial exchange term. 13 refs., 84 figs.

  19. Development of best estimate auditing code for CANDU thermal hydraulic safety analysis

    Energy Technology Data Exchange (ETDEWEB)

    Chung, B. D.; Lee, W. J.; Lim, H. S. [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1998-04-15

    The main purpose of this study is to develop a thermal hydraulic auditing code for the CANDU reactor, modifying the model of existing PWR auditing tool, i.e. RELAP5/MOD3. This scope of project is first step of the whole project, thus focus to the establishment of improvement area. The study was performed by reconsideration of the previous code assessment works and investigation of AECL design analysis tools. In order to identify the thermal hydraulic phenomena for events, the whole system of CANDU plant was divided into main functional systems and subcomponents. Each phenomena was addressed to the each subcomponent. FinaIly improvement areas of model development for auditing tool were established based on the identified phenomena.

  20. Reactor thermal-hydraulic FY 1986 status report for the multimegawatt Space Nuclear Power Program

    Energy Technology Data Exchange (ETDEWEB)

    Krotiuk, W.J.; Antoniak, Z.I.

    1986-10-01

    PNL's 1986 activities can be divided into three basic areas: code assessment, correlation assessment and experimental activities. The ultimate goal of all these activities is developing computer codes and verifying their use to perform the thermal-hydraulic analysis and design of the reactor core and plenum of the various proposed concepts. To perform this task as assessment is made of existing computer codes, models, correlations, and microgravity experimental data.

  1. Coupled neutronic-thermal-hydraulics analysis in a coolant subchannel of a PWR using CFD techniques

    Energy Technology Data Exchange (ETDEWEB)

    Ribeiro, Felipe P.; Su, Jian, E-mail: sujian@nuclear.ufrj.br [Coordenacao de Pos-Graduacao e Pesquisa de Engenharia (COPPE/UFRJ), Rio de Janeiro, RJ (Brazil). Programa de Engenharia Nuclear

    2017-07-01

    The high capacity of Computational Fluid Dynamics code to predict multi-dimensional thermal-hydraulics behaviour and the increased availability of capable computer systems are making that method a good tool to simulate phenomena of thermal-hydraulics nature in nuclear reactors. However, since there are no neutron kinetics models available in commercial CFD codes to the present day, the application of CFD in the nuclear reactor safety analyses is still limited. The present work proposes the implementation of the point kinetics model (PKM) in ANSYS - Fluent to predict the neutronic behaviour in a Westinghouse Sequoyah nuclear reactor, coupling with the phenomena of heat conduction in the rod and thermal-hydraulics in the cooling fluid, via the reactivity feedback. Firstly, a mesh convergence and turbulence model study was performed, using the Reynolds-Average Navier-Stokes method, with square arrayed rod bundle featuring pitch to diameter ratio of 1:32. Secondly, simulations using the k-! SST turbulence model were performed with an axial distribution of the power generation in the fuel to analyse the heat transfer through the gap and cladding, and its in fluence on the thermal-hydraulics behaviour of the cooling fluid. The wall shear stress distribution for the centre-line rods and the dimensionless velocity were evaluated to validate the model, as well as the in fluence of the mass flow rate variation on the friction factor. The coupled model enabled to perform a dynamic analysis of the nuclear reactor during events of insertion of reactivity and shutdown of primary coolant pumps. (author)

  2. ITHNA.SYS: An Integrated Thermal Hydraulic and Neutronic Analyzer SYStem for NUR research reactor

    Energy Technology Data Exchange (ETDEWEB)

    Mazidi, S., E-mail: samirmazidi@gmail.com [Division Physique et Applications Nucléaires, Centre de Recherche Nucléaire de Draria (CRND), BP 43 Sebala, Draria, Alger (Algeria); Meftah, B., E-mail: b_meftah@yahoo.com [Division Physique et Applications Nucléaires, Centre de Recherche Nucléaire de Draria (CRND), BP 43 Sebala, Draria, Alger (Algeria); Belgaid, M., E-mail: belgaidm@yahoo.com [Faculté de Physique, Université Houari Boumediene, USTHB, BP 31, Bab Ezzouar, Alger (Algeria); Letaim, F., E-mail: fletaim@yahoo.fr [Faculté des Sciences et Technologies, Université d’El-oued, PO Box 789, El-oued (Algeria); Halilou, A., E-mail: hal_rane@yahoo.fr [Division Réacteur NUR, Centre de Recherche Nucléaire de Draria, BP 43 Sebala, Draria, Alger (Algeria)

    2015-08-15

    Highlights: • We develop a neutronic and thermal hydraulic MTR reactor analyzer. • The analyzer allows a rapid determination of the reactor core parameters. • Some NUR reactor parameters have been analyzed. - Abstract: This paper introduces the Integrated Thermal Hydraulic and Neutronic Analyzer SYStem (ITHNA.SYS) that has been developed for the Algerian research reactor NUR. It is used both as an operating aid tool and as a core physics engineering analysis tool. The system embeds three modules of the MTR-PC software package developed by INVAP SE: the cell calculation code WIMSD, the core calculation code CITVAP and the program TERMIC for thermal hydraulic analysis of a material testing reactor (MTR) core in forced convection. ITHNA.SYS operates both in on-line and off-line modes. In the on-line mode, the system is linked, via the computer parallel port, to the data acquisition console of the reactor control room and allows a real time monitoring of major physical and safety parameters of the NUR core. PC-based ITHNA.SYS provides a viable and convenient way of using an accumulated and often complex reactor physics stock of knowledge and frees the user from the intricacy of adequate reactor core modeling. This guaranties an accurate, though rapid, determination of a variety of neutronic and thermal hydraulic parameters of importance for the operation and safety analysis of the NUR research reactor. Instead of the several hours usually required, the processing time for the determination of such parameters is now reduced to few seconds. Validation of the system was performed with respect to experimental measurements and to calculations using reference codes. ITHNA.SYS can be easily adapted to accommodate other kinds of MTR reactors.

  3. Neutronic and thermal-hydraulic studies of aqueous homogeneous reactor for medical isotopes production

    Energy Technology Data Exchange (ETDEWEB)

    Perez, Daniel Milian; Lorenzo, Daniel E. Milian; Lira, Carlos A. Brayner de Oliveira; Garcia, Lorena P. Rodríguez, E-mail: milianperez89@gmail.com, E-mail: dmilian@instec.cu, E-mail: lorenapilar1109@gmail.com, E-mail: cabol@ufpe.br [Higher Institute of Technologies and Applied Sciences (InSTEC), Havana (Cuba); Universidade Federal de Pernambuco (UFPE), Recife, PE (Brazil). Departamento de Energia Nuclear

    2017-11-01

    The use of Aqueous Homogenous Reactors (AHR) is one of the most promissory alternatives to produce medical isotopes, mainly {sup 99}Mo. Compare to multipurpose research reactors, an AHR dedicated for {sup 99}Mo production has advantages because of their low cost, small critical mass, inherent passive safety, and simplified fuel handling, processing, and purification characteristics. This article presents the current state of research in our working group on this topic. Are presented and discussed the group validation efforts with benchmarking exercises that include neutronic and thermal-hydraulic results of two solution reactors, the SUPO and ARGUS reactors. Neutronic and thermal-hydraulic results of 75 kWth AHR based on the ARGUS reactor LEU configuration are presented. The neutronic studies included the determination of parameters such as reflector thickness, critical height, medical isotopes production and others. Thermal-hydraulics studies were focused on demonstrating that sufficient cooling capacity exists to prevent fuel overheating. In addition, the effects of some calculation parameters on the computational modeling of temperature, velocity and gas volume fraction during steady-state operation of an AHR are discussed. The neutronic and thermal-hydraulics studies have been performed with the MCNPX version 2.6e computational code and the version 14 of ANSYS CFX respectively. Our group studies and the results obtained contribute to demonstrate the feasibility of using AHR for the production of medical isotopes, however additional studies are still necessary to confirm these results and contribute to development and demonstration of their technical, safety, and economic viability. (author)

  4. Mechanical and hydraulic behavior of a rock fracture under shear deformation

    National Research Council Canada - National Science Library

    Nishiyama, Satoshi; Ohnishi, Yuzo; Ito, Hisao; Yano, Takao

    2014-01-01

    ...-scale hydraulic behavior. To this end, a simultaneous permeability and shear test device is developed, and shear-flow coupling tests are conducted on specimens having fractures with varied levels of surface roughness...

  5. Predicting Formation Damage in Aquifer Thermal Energy Storage Systems Utilizing a Coupled Hydraulic-Thermal-Chemical Reservoir Model

    Science.gov (United States)

    Müller, Daniel; Regenspurg, Simona; Milsch, Harald; Blöcher, Guido; Kranz, Stefan; Saadat, Ali

    2014-05-01

    In aquifer thermal energy storage (ATES) systems, large amounts of energy can be stored by injecting hot water into deep or intermediate aquifers. In a seasonal production-injection cycle, water is circulated through a system comprising the porous aquifer, a production well, a heat exchanger and an injection well. This process involves large temperature and pressure differences, which shift chemical equilibria and introduce or amplify mechanical processes. Rock-fluid interaction such as dissolution and precipitation or migration and deposition of fine particles will affect the hydraulic properties of the porous medium and may lead to irreversible formation damage. In consequence, these processes determine the long-term performance of the ATES system and need to be predicted to ensure the reliability of the system. However, high temperature and pressure gradients and dynamic feedback cycles pose challenges on predicting the influence of the relevant processes. Within this study, a reservoir model comprising a coupled hydraulic-thermal-chemical simulation was developed based on an ATES demonstration project located in the city of Berlin, Germany. The structural model was created with Petrel, based on data available from seismic cross-sections and wellbores. The reservoir simulation was realized by combining the capabilities of multiple simulation tools. For the reactive transport model, COMSOL Multiphysics (hydraulic-thermal) and PHREEQC (chemical) were combined using the novel interface COMSOL_PHREEQC, developed by Wissmeier & Barry (2011). It provides a MATLAB-based coupling interface between both programs. Compared to using COMSOL's built-in reactive transport simulator, PHREEQC additionally calculates adsorption and reaction kinetics and allows the selection of different activity coefficient models in the database. The presented simulation tool will be able to predict the most important aspects of hydraulic, thermal and chemical transport processes relevant to

  6. An assessment of the CORCON-MOD3 code. Part 1: Thermal-hydraulic calculations

    Energy Technology Data Exchange (ETDEWEB)

    Strizhov, V.; Kanukova, V.; Vinogradova, T.; Askenov, E. [Russian Academy of Sciences, Moscow (Russian Federation). Inst. of Nuclear Safety; Nikulshin, V. [Russian Research Center, Moscow (Russian Federation). Kurchatov Inst.

    1996-09-01

    This report deals with the subject of CORCON-Mod3 code validation (thermal-hydraulic modeling capability only) based on MCCI (molten core concrete interaction) experiments conducted under different programs in the past decade. Thermal-hydraulic calculations (i.e., concrete ablation, melt temperature, melt energy, concrete temperature, and condensible and non-condensible gas generation) were performed with the code, and compared with the data from 15 experiments, conducted at different scales using both simulant (metallic and oxidic) and prototypic melt materials, using different concrete types, and with and without an overlying water pool. Sensitivity studies were performed in a few cases involving, for example, heat transfer from melt to concrete, condensed phase chemistry, etc. Further, special analysis was performed using the ACE L8 experimental data to illustrate the differences between the experimental and the reactor conditions, and to demonstrate that with proper corrections made to the code, the calculated results were in better agreement with the experimental data. Generally, in the case of dry cavity and metallic melts, CORCON-Mod3 thermal-hydraulic calculations were in good agreement with the test data. For oxidic melts in a dry cavity, uncertainties in heat transfer models played an important role for two melt configurations--a stratified geometry with segregated metal and oxide layers, and a heterogeneous mixture. Some discrepancies in the gas release data were noted in a few cases.

  7. Changes of soil thermal and hydraulic regimes in the Heihe River Basin.

    Science.gov (United States)

    Peng, Xiaoqing; Mu, Cuicui

    2017-09-02

    Soil thermal and hydraulic regimes are critical factors influencing terrestrial processes in cold regions. Collection of field data from frozen ground has occurred at point scales, but limited data exist that characterize changes of soil thermal and hydraulic regimes at the scale of the whole Heihe River Basin. This study uses a long-term regional climate model coupled with land surface model to investigate the soil thermal and hydraulic regime changes at a large spatial scale. It also explores potential factors, including the climate and non-climate factors. Results show that there is significant variability in mean annual air temperature (MAAT) of about 0.47 °C/decade during 1980-2013. A time series of area-averaged mean annual soil temperature (MAST) over the whole Heihe River Basin shows a significant increase between 0.25 and 0.36 °C/decade during 1984-2013, with a net change of 0.9 °C. A trend of increasing wetness is found in soil moisture. Frozen days (FD) decreased significantly both in seasonally frozen ground (SFG) regions and permafrost regions, with a net change between 7 and 13 days during 1984-2013. Freezing index (FI) had a positive effect on FD, while thawing index (TI), MAAT, precipitation, and normalized difference vegetation index (NDVI) had a negative effect. These results are important to understand dynamic mechanisms of soil freeze/thaw cycles.

  8. Thermal-Hydraulic Integral Effect Test with the ATLS for Investigation on CEDM Penetration Nozzle Integrity

    Energy Technology Data Exchange (ETDEWEB)

    Kang, Kyoungho; Seokcho; Park, Hyunsik; Choi, Namhyun; Park, Yusun; Kim, Jongrok; Bae, Byounguhn; Kim, Yeonsik; Choi, Kiyong; Song, Chulhwa [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2013-05-15

    In this study, thermal-hydraulic integral effect test with the ATLAS (Advanced Thermal-Hydraulic Test Loop for Accident Simulation) was performed for simulating a failure of CEDM penetration nozzle. The main objectives of the present test were not only to provide physical insight into the system response during a failure of CEDM penetration nozzle but also to establish an integral effect test database for the validation of the safety analysis codes. Furthermore, present experimental data were utilized to resolve the safety issue raised by the PWSCC at the CEDM penetration nozzle of the YGN-3. Thermal-hydraulic integral effect test with the ATLAS was performed for simulating a failure of CEDM penetration nozzle. Failure of two penetration nozzles of the CEDM in the APR1400 was simulated. Initial and boundary conditions were determined with respect to the reference conditions of the APR1400. However, with an aim of corresponding to the YGN-3 situation, the safety injection water was supplied via CLI mode. Compared to the cold leg break SBLOCA, the consequences of the event were milder in terms of a loop seal clearance, break flow rate, collapsed water level, and PCT. This could be mainly attributed to the small break flow rate in case of the failure in the RPV upper head. Present experimental data were utilized to resolve the safety issue raised by the PWSCC at the CEDM penetration nozzle of the YGN-3.

  9. Thermal-Hydraulic System Codes in Nulcear Reactor Safety and Qualification Procedures

    Directory of Open Access Journals (Sweden)

    Alessandro Petruzzi

    2008-01-01

    Full Text Available In the last four decades, large efforts have been undertaken to provide reliable thermal-hydraulic system codes for the analyses of transients and accidents in nuclear power plants. Whereas the first system codes, developed at the beginning of the 1970s, utilized the homogenous equilibrium model with three balance equations to describe the two-phase flow, nowadays the more advanced system codes are based on the so-called “two-fluid model” with separation of the water and vapor phases, resulting in systems with at least six balance equations. The wide experimental campaign, constituted by the integral and separate effect tests, conducted under the umbrella of the OECD/CSNI was at the basis of the development and validation of the thermal-hydraulic system codes by which they have reached the present high degree of maturity. However, notwithstanding the huge amounts of financial and human resources invested, the results predicted by the code are still affected by errors whose origins can be attributed to several reasons as model deficiencies, approximations in the numerical solution, nodalization effects, and imperfect knowledge of boundary and initial conditions. In this context, the existence of qualified procedures for a consistent application of qualified thermal-hydraulic system code is necessary and implies the drawing up of specific criteria through which the code-user, the nodalization, and finally the transient results are qualified.

  10. Thermal Hydraulic Analysis of 3 MW TRIGA Research Reactor of Bangladesh Considering Different Cycles of Burnup

    Directory of Open Access Journals (Sweden)

    M.H. Altaf

    2014-12-01

    Full Text Available Burnup dependent steady state thermal hydraulic analysis of TRIGA Mark-II research reactor has been carried out utilizing coupled point kinetics, neutronics and thermal hydraulics code EUREKA-2/RR. From the previous calculations of neutronics parameters including percentage burnup of individual fuel elements performed so far for 700 MWD burnt core of TRIGA reactor showed that the fuel rod predicted as hottest at the beginning of cycle (fresh core was found to remain as the hottest until 200 MWD of burn, but, with the progress of core burn, the hottest rod was found to be shifted and another rod in the core became the hottest. The present study intends to evaluate the thermal hydraulic parameters of these hottest fuel rods at different cycles of burnup, from beginning to 700 MWD core burnt considering reactor operates under steady state condition. Peak fuel centerline temperature, maximum cladding and coolant temperatures of the hottest channels were calculated. It revealed that maximum temperature reported for fuel clad and fuel centerline found to lie below their melting points which indicate that there is no chance of burnout on the fuel cladding surface and no blister in the fuel meat throughout the considered cycles of core burnt.

  11. Stomatal control and leaf thermal and hydraulic capacitances under rapid environmental fluctuations.

    Directory of Open Access Journals (Sweden)

    Stanislaus J Schymanski

    Full Text Available Leaves within a canopy may experience rapid and extreme fluctuations in ambient conditions. A shaded leaf, for example, may become exposed to an order of magnitude increase in solar radiation within a few seconds, due to sunflecks or canopy motions. Considering typical time scales for stomatal adjustments, (2 to 60 minutes, the gap between these two time scales raised the question whether leaves rely on their hydraulic and thermal capacitances for passive protection from hydraulic failure or over-heating until stomata have adjusted. We employed a physically based model to systematically study effects of short-term fluctuations in irradiance on leaf temperatures and transpiration rates. Considering typical amplitudes and time scales of such fluctuations, the importance of leaf heat and water capacities for avoiding damaging leaf temperatures and hydraulic failure were investigated. The results suggest that common leaf heat capacities are not sufficient to protect a non-transpiring leaf from over-heating during sunflecks of several minutes duration whereas transpirative cooling provides effective protection. A comparison of the simulated time scales for heat damage in the absence of evaporative cooling with observed stomatal response times suggested that stomata must be already open before arrival of a sunfleck to avoid over-heating to critical leaf temperatures. This is consistent with measured stomatal conductances in shaded leaves and has implications for water use efficiency of deep canopy leaves and vulnerability to heat damage during drought. Our results also suggest that typical leaf water contents could sustain several minutes of evaporative cooling during a sunfleck without increasing the xylem water supply and thus risking embolism. We thus submit that shaded leaves rely on hydraulic capacitance and evaporative cooling to avoid over-heating and hydraulic failure during exposure to typical sunflecks, whereas thermal capacitance provides

  12. TAPINS: A THERMAL-HYDRAULIC SYSTEM CODE FOR TRANSIENT ANALYSIS OF A FULLY-PASSIVE INTEGRAL PWR

    Directory of Open Access Journals (Sweden)

    YEON-GUN LEE

    2013-08-01

    Full Text Available REX-10 is a fully-passive small modular reactor in which the coolant flow is driven by natural circulation, the RCS is pressurized by a steam-gas pressurizer, and the decay heat is removed by the PRHRS. To confirm design decisions and analyze the transient responses of an integral PWR such as REX-10, a thermal-hydraulic system code named TAPINS (Thermal-hydraulic Analysis Program for INtegral reactor System is developed in this study. Based on a one-dimensional four-equation drift-flux model, TAPINS incorporates mathematical models for the core, the helical-coil steam generator, and the steam-gas pressurizer. The system of difference equations derived from the semi-implicit finite-difference scheme is numerically solved by the Newton Block Gauss Seidel (NBGS method. TAPINS is characterized by applicability to transients with non-equilibrium effects, better prediction of the transient behavior of a pressurizer containing non-condensable gas, and code assessment by using the experimental data from the autonomous integral effect tests in the RTF (REX-10 Test Facility. Details on the hydrodynamic models as well as a part of validation results that reveal the features of TAPINS are presented in this paper.

  13. Thermal hydraulic and neutron kinetic simulation of the Angra 2 reactor using a RELAP5/PARCS coupled model

    Energy Technology Data Exchange (ETDEWEB)

    Reis, Patricia A.L.; Costa, Antonella L.; Hamers, Adolfo R.; Pereira, Claubia; Rodrigues, Thiago D.A.; Mantecon, Javier G.; Veloso, Maria A.F., E-mail: patricialire@yahoo.com.br, E-mail: antonella@nuclear.ufmg.br, E-mail: adolforomerohamers@hotmail.com, E-mail: claubia@nuclear.ufmg.br, E-mail: thiagodanielbh@gmail.com, E-mail: mantecon1987@gmail.com, E-mail: dora@nuclear.ufmg.br [Universidade Federal de Minas Gerais (UFMG), Belo Horizonte, MG (Brazil). Departamento de Engenharia Nuclear; Instituto Nacional de Ciencias e Tecnologia de Reatores Nucleares Inovadores (INCT/CNPq), Belo Horizonte (Brazil); Miro, Rafael; Verdu, Gumersindo, E-mail: rmiro@iqn.upv.es, E-mail: gverdu@iqn.upv.es [Universidad Politecnica de Valencia (Spain). Departamento de Ingenieria Quimica y Nuclear

    2015-07-01

    The computational advances observed in the last two decades have been provided direct impact on the researches related to nuclear simulations, which use several types of computer codes, including coupled between them, allowing representing with very accuracy the behavior of nuclear plants. Studies of complex scenarios in nuclear reactors have been improved by the use of thermal-hydraulic (TH) and neutron kinetics (NK) coupled codes. This technique consists in incorporating three-dimensional (3D) neutron modeling of the reactor core into codes, mainly to simulate transients that involve asymmetric core spatial power distributions and strong feedback effects between neutronics and reactor thermal-hydraulics. Therefore, this work presents preliminary results of TH RELAP5 and the NK PARCS calculations applied to model of the Angra 2 reactor. The WIMSD-5B code has been used to generate the macroscopic cross sections used in the NK code. The results obtained are satisfactory and represent important part of the development of this methodology. The next step is to couple the codes. (author)

  14. Development of numerical simulation system for thermal-hydraulic analysis in fuel assembly of sodium-cooled fast reactor

    Energy Technology Data Exchange (ETDEWEB)

    Ohshima, Hiroyuki; Uwaba, Tomoyuki [Japan Atomic Energy Agency (4002 Narita, O-arai, Ibaraki 311-1393, Japan) (Japan); Hashimoto, Akihiko; Imai, Yasutomo [NDD Corporation (1-1-6 Jounan, Mito, Ibaraki 310-0803, Japan) (Japan); Ito, Masahiro [NESI Inc. (4002 Narita, O-arai, Ibaraki 311-1393, Japan) (Japan)

    2015-12-31

    A numerical simulation system, which consists of a deformation analysis program and three kinds of thermal-hydraulics analysis programs, is being developed in Japan Atomic Energy Agency in order to offer methodologies to clarify thermal-hydraulic phenomena in fuel assemblies of sodium-cooled fast reactors under various operating conditions. This paper gives the outline of the system and its applications to fuel assembly analyses as a validation study.

  15. Thermal-hydraulic analysis on the whole module of water cooled ceramic breeder blanket for CFETR

    Energy Technology Data Exchange (ETDEWEB)

    Jiang, Kecheng; Ma, Xuebin [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei, Anhui, 230031 (China); University of Science and Technology of China, Hefei, Anhui, 230027 (China); Cheng, Xiaoman [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei, Anhui, 230031 (China); Lin, Shuang [University of Science and Technology of China, Hefei, Anhui, 230027 (China); Huang, Kai [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei, Anhui, 230031 (China); Liu, Songlin, E-mail: slliu@ipp.ac.cn [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei, Anhui, 230031 (China); University of Science and Technology of China, Hefei, Anhui, 230027 (China)

    2016-11-15

    Highlights: • The 3D thermal hydraulic analysis on the whole module of WCCB is performed by CFD method. • Temperature field and mass flow distribution have been obtained. • The design of WCCB is reasonable from the perspective of thermal-hydraulics. • The scheme for further optimization has been proposed. - Abstract: The Water Cooled Ceramic Breeder blanket (WCCB) is being researched for Chinese Fusion Engineering Test Reactor (CFETR). The thermal-hydraulic analysis is essential because the blanket should remove the high heat flux from the plasma and the volumetric heat generated by neutrons. In this paper, the detailed three dimensional (3D) thermal hydraulic analysis on the whole module of WCCB blanket has been performed by Computational Fluid Dynamics (CFD) method, which is capable of solving conjugate heat transfer between solid structure and fluid. The main results, including temperature field, distribution of mass flow rate and coolant pressure drop, have been calculated simultaneously. These provides beneficial guidance data for the further structural optimization and for the design arrangement of primary and secondary circuit. Under the total heat source of 1.23 MW, the coolant mass flow rate of 5.457 kg/s is required to make coolant water corresponding to the Pressurized Water Reactor (PWR) condition (15.5 MPa, 285 °C–325 °C), generating the total coolant pressure drop (△P) of 0.467 MPa. The results show that the present structural design can make all the materials effectively cooled to the allowable temperature range, except for a few small modifications on the both sides of FW. The main components, including the first wall (FW), cooling plates (CPs), side wall (SWs)&stiffening plates (SPs) and the manifold(1–4), dominate 4.7%/41.7%/13%/40.6% of the total pressure drop, respectively. Additionally, the mass flow rate of each channel has been obtained, showing the peak relative deviation of 3.4% and 2% from the average for the paratactic

  16. Effects of temperature and thermally-induced microstructure change on hydraulic conductivity of Boom Clay

    Directory of Open Access Journals (Sweden)

    W.Z. Chen

    2017-06-01

    Full Text Available Boom Clay is one of the potential host rocks for deep geological disposal of high-level radioactive nuclear waste in Belgium. In order to investigate the mechanism of hydraulic conductivity variation under complex thermo-mechanical coupling conditions and to better understand the thermo-hydro-mechanical (THM coupling behaviour of Boom Clay, a series of permeability tests using temperature-controlled triaxial cell has been carried out on the Boom Clay samples taken from Belgian underground research laboratory (URL HADES. Due to its sedimentary nature, Boom Clay presents across-anisotropy with respect to its sub-horizontal bedding plane. Direct measurements of the vertical (Kv and horizontal (Kh hydraulic conductivities show that the hydraulic conductivity at 80 °C is about 2.4 times larger than that at room temperature (23 °C, and the hydraulic conductivity variation with temperature is basically reversible during heating–cooling cycle. The anisotropic property of Boom Clay is studied by scanning electron microscope (SEM tests, which highlight the transversely isotropic characteristics of intact Boom Clay. It is shown that the sub-horizontal bedding feature accounts for the horizontal permeability higher than the vertical one. The measured increment in hydraulic conductivity with temperature is lower than the calculated one when merely considering the changes in water kinematic viscosity and density with temperature. The nuclear magnetic resonance (NMR tests have also been carried out to investigate the impact of microstructure variation on the THM properties of clay. The results show that heating under unconstrained boundary condition will produce larger size of pores and weaken the microstructure. The discrepancy between the hydraulic conductivity experimentally measured and predicted (considering water viscosity and density changes with temperature can be attributed to the microstructural weakening effect on the thermal volume change

  17. Use of a hydraulic brake as a source of thermal energy for the railway rolling stock

    Directory of Open Access Journals (Sweden)

    V.A.Gabrinets

    2012-12-01

    Full Text Available Introduction: In this paper the braking issues of passenger trains which have a great speed and frequent stops are examines. Problem statement: These processes are ехpensive and have big energy losses. The proposed solution to the problem: The kinetic energy of braking prosses propose to turn into thermal energy of heating fluid. For this purpose special hydraulic brake is proposed. The brake is connected with the wheel carriage pairs. The process is based on the energy dissipation in liqid when the disks with spikes rotate in it. Because the real liquid has friction and viscosity, it will be heat up, when the mechanical parts of the hydraulic brake are moved in it. The design, operating principle and characteristics of the hydraulic brake are proposed. Transmission of kinetic energy of carriage motion to brake system executed by mechanical clutches. It connected with the wheel pair and transmitting the energy the wheels rotation to hydraulic brake discs. The cylindrical rods are installed on the discs. Rods location fits the profile of the curved centrifugal pump vanes. As result, the fluid heatind prosess by rotatinge discs with rods take place also at the same time with the liquid pumping through the inner volume of brake system.Conclusions: Affordable passenger carriage braking dynamic is achieved by varying the size and number of rods. The heated liquid may be subsequently used for household needs and for heating the passenger carriage.

  18. Simulations of thermal-hydraulic processes in heat exchangers- station of the cogeneration power plant

    Energy Technology Data Exchange (ETDEWEB)

    Studovic, M.; Stevanovic, V.; Ilic, M.; Nedeljkovic, S. [Faculty of Mechanical Engineering of Belgrade (Croatia)

    1995-12-31

    Design of the long district heating system to Belgrade (base load 580 MJ/s) from Thermal Power Station `Nikola Tesla A`, 30 km southwest from the present gas/oil burning boilers in New Belgrade, is being conducted. The mathematical model and computer code named TRP are developed for the prediction of the design basis parameters of heat exchangers station, as well as for selection of protection devices and formulation of operating procedures. Numerical simulations of heat exchangers station are performed for various transient conditions: up-set and abnormal. Physical model of multi-pass, shell and tube heat exchanger in the station represented is by unique steam volume, and with space discretised nodes both for water volume and tube walls. Heat transfer regimes on steam and water side, as well as hydraulic calculation were performed in accordance with TEMA standards for transient conditions on both sides, and for each node on water side. Mathematical model is based on balance equations: mass and energy for lumped parameters on steam side, and energy balances for tube walls and water in each node. Water mass balance is taken as boundary/initial condition or as specified control function. The physical model is proposed for (s) heat exchangers in the station and (n) water and wall volumes. Therefore, the mathematical model consists of 2ns+2, non-linear differential equations, including equations of state for water, steam and tube material, and constitutive equations for heat transfer on steam and water side, solved by the Runge-Kutt method. Five scenarios of heat exchangers station behavior have been simulated with the TRP code and obtained results are presented. (author)

  19. Sensitivity analysis of hydraulic and thermal parameters inducing anomalous heat flow in the Lower Yarmouk Gorge

    Science.gov (United States)

    Goretzki, Nora; Inbar, Nimrod; Kühn, Michael; Möller, Peter; Rosenthal, Eliyahu; Schneider, Michael; Siebert, Christian; Magri, Fabien

    2016-04-01

    The Lower Yarmouk Gorge, at the border between Israel and Jordan, is characterized by an anomalous temperature gradient of 46 °C/km. Numerical simulations of thermally-driven flow show that ascending thermal waters are the result of mixed convection, i.e. the interaction between the regional flow from the surrounding heights and buoyant flow within permeable faults [1]. Those models were calibrated against available temperature logs by running several forward problems (FP), with a classic "trial and error" method. In the present study, inverse problems (IP) are applied to find alternative parameter distributions that also lead to the observed thermal anomalies. The investigated physical parameters are hydraulic conductivity and thermal conductivity. To solve the IP, the PEST® code [2] is applied via the graphical interface FEPEST® in FEFLOW® [3]. The results show that both hydraulic and thermal conductivity are consistent with the values determined with the trial and error calibrations, which precede this study. However, the IP indicates that the hydraulic conductivity of the Senonian Paleocene aquitard can be 8.54*10-3 m/d, which is three times lower than the originally estimated value in [1]. Moreover, the IP suggests that the hydraulic conductivity in the faults can increase locally up to 0.17 m/d. These highly permeable areas can be interpreted as local damage zones at the faults/units intersections. They can act as lateral pathways in the deep aquifers that allow deep outflow of thermal water. This presentation provides an example about the application of FP and IP to infer a wide range of parameter values that reproduce observed environmental issues. [1] Magri F, Inbar N, Siebert C, Rosenthal E, Guttman J, Möller P (2015) Transient simulations of large-scale hydrogeological processes causing temperature and salinity anomalies in the Tiberias Basin. Journal of Hydrology, 520, 342-355 [2] Doherty J (2010) PEST: Model-Independent Parameter Estimation. user

  20. Response of Compacted Bentonites to Thermal and Thermo-Hydraulic Loadings at High Temperatures

    Directory of Open Access Journals (Sweden)

    Snehasis Tripathy

    2017-07-01

    Full Text Available The final disposal of high-level nuclear waste in many countries is preferred to be in deep geological repositories. Compacted bentonites are proposed for use as the buffer surrounding the waste canisters which may be subjected to both thermal and hydraulic loadings. A significant increase in the temperature is anticipated within the buffer, particularly during the early phase of the repository lifetime. In this study, several non-isothermal and non-isothermal hydraulic tests were carried on compacted MX80 bentonite. Compacted bentonite specimens (water content = 15.2%, dry density = 1.65 Mg/m3 were subjected to a temperature of either 85 or 150 °C at one end, whereas the temperature at the opposite end was maintained at 25 °C. During the non-isothermal hydraulic tests, water was supplied from the opposite end of the heat source. The temperature and relative humidity were monitored along predetermined depths of the specimens. The profiles of water content, dry density, and degree of saturation were established after termination of the tests. The test results showed that thermal gradients caused redistribution of the water content, whereas thermo-hydraulic gradients caused both redistribution and an increase in the water content within compacted bentonites, both leading to development of axial stress of various magnitudes. The applied water injection pressures (5 and 600 kPa and temperature gradients appeared to have very minimal impact on the magnitude of axial stress developed. The thickness of thermal insulation layer surrounding the testing devices was found to influence the temperature and relative humidity profiles thereby impacting the redistribution of water content within compacted bentonites. Under the influence of both the applied thermal and thermo-hydraulic gradients, the dry density of the bentonite specimens increased near the heat source, whereas it decreased at the opposite end. The test results emphasized the influence of

  1. Transmutation technology development; thermal hydraulic power analysis and structure analysis of the HYPER target beam window

    Energy Technology Data Exchange (ETDEWEB)

    Choi, J. H.; Ju, E. S.; Song, M. K.; Jeon, Y. Z. [Gyeongsang National University, Jinju (Korea)

    2002-03-01

    A thermal hydraulic power analysis, a structure analysis and optimization computation for some design factor for the design of spallation target suitable for HYPER with 1000 MW thermal power in this study was performed. Heat generation formula was used which was evaluated recently based on the LAHET code, mainly to find the maximum beam current under given computation conditions. Thermal hydraulic power of HYPER target system was calculated using FLUENT code, structure conducted by inputting the data into ANSYS. On the temp of beam windows and the pressure distribution calculated using FLUENT. Data transformation program was composed apply the data calculated using FLUENT being commercial CFD code and ANSYS being FEM code for CFX structure analysis. A basic study was conducted on various singular target to obtain fundamental data on the shape for optimum target design. A thermal hydraulic power analysis and structure analysis were conducted on the shapes of parabolic, uniform, scanning beams to choose the optimum shape of beam current analysis was done according to some turbulent model to simulate the real flow. To evaluate the reliability of numerical analysis result, benchmarking of FLUENT code reformed at SNU and Korea Advanced Institute of Science and Technology and it was compared to CFX in the possession of Korea Atomic Energy Research Institute and evaluated. Reliable deviation was observed in the results calculated using FLUENT code, but temperature deviation of about 200 .deg. C was observed in the result from CFX analysis at optimum design condition. Several benchmarking were performed on the basis of numerical analysis concerning conventional HYPER. It was possible to allow a beam arrests of 17.3 mA in the case of the {phi} 350 mm parabolic beam suggested to the optimum in nuclear transmutation when stress equivalent to VON-MISES was calculated to be 140 MPa. 29 refs., 109 figs. (Author)

  2. Transient cases analyses of the TRIGA IPR-R1 using thermal hydraulic and neutron kinetic coupled codes

    Energy Technology Data Exchange (ETDEWEB)

    Reis, Patricia A.L.; Costa, Antonella L.; Pereira, Claubia; Veloso, Maria A.F.; Scari, Maria E., E-mail: patricialire@yahoo.com.br, E-mail: antonella@nuclear.ufmg.br, E-mail: claubia@nuclear.ufmg.br, E-mail: dora@nuclear.ufmg.br, E-mail: melizabethscari@yahoo.com [Universidade Federal de Minas Gerais (UFMG), Belo Horizonte, MG (Brazil). Departamento de Engenharia Nuclear; Instituto Nacional de Ciencias e Tecnologia de Reatores Nucleares Inovadores (INCT/CNPq), Belo Horizonte (Brazil); Miro, Rafael; Verdu, Gumersindo, E-mail: rmiro@iqn.upv.es, E-mail: gverdu@iqn.upv.es [Universidad Politecnica de Valencia (Spain). Departamento de Ingenieria Quimica y Nuclear

    2015-07-01

    Simulations and analyses of nuclear reactors have been improved by utilization of coupled thermal-hydraulic (TH) and neutron kinetics (NK) system codes especially to simulate transients that involve strong feedback effects between NK and TH. The TH-NK coupling technique was initially developed and used to simulate the behavior of power reactors; however, several coupling methodologies are now being applied for research reactors. This work presents the coupling methodology application between RELAP5 and PARCS codes using as a model the TRIGA IPR-R1 research reactor. Analyses of steady state and transient conditions and comparisons with results from simulations using only the RELAP5 code are being presented in this paper. (author)

  3. Thermal-hydraulic analysis under partial loss of flow accident hypothesis of a plate-type fuel surrounded by two water channels using RELAP5 code

    Directory of Open Access Journals (Sweden)

    Itamar Iliuk

    2016-01-01

    Full Text Available Thermal-hydraulic analysis of plate-type fuel has great importance to the establishment of safety criteria, also to the licensing of the future nuclear reactor with the objective of propelling the Brazilian nuclear submarine. In this work, an analysis of a single plate-type fuel surrounding by two water channels was performed using the RELAP5 thermal-hydraulic code. To realize the simulations, a plate-type fuel with the meat of uranium dioxide sandwiched between two Zircaloy-4 plates was proposed. A partial loss of flow accident was simulated to show the behavior of the model under this type of accident. The results show that the critical heat flux was detected in the central region along the axial direction of the plate when the right water channel was blocked.

  4. Calibrating thermal behavior of electronics

    Science.gov (United States)

    Chainer, Timothy J.; Parida, Pritish R.; Schultz, Mark D.

    2016-05-31

    A method includes determining a relationship between indirect thermal data for a processor and a measured temperature associated with the processor, during a calibration process, obtaining the indirect thermal data for the processor during actual operation of the processor, and determining an actual significant temperature associated with the processor during the actual operation using the indirect thermal data for the processor during actual operation of the processor and the relationship.

  5. Verification of combined thermal-hydraulic and heat conduction analysis code FLOWNET/TRUMP

    Science.gov (United States)

    Maruyama, Soh; Fujimoto, Nozomu; Kiso, Yoshihiro; Murakami, Tomoyuki; Sudo, Yukio

    1988-09-01

    This report presents the verification results of the combined thermal-hydraulic and heat conduction analysis code, FLOWNET/TRUMP which has been utilized for the core thermal hydraulic design, especially for the analysis of flow distribution among fuel block coolant channels, the determination of thermal boundary conditions for fuel block stress analysis and the estimation of fuel temperature in the case of fuel block coolant channel blockage accident in the design of the High Temperature Engineering Test Reactor(HTTR), which the Japan Atomic Energy Research Institute has been planning to construct in order to establish basic technologies for future advanced very high temperature gas-cooled reactors and to be served as an irradiation test reactor for promotion of innovative high temperature new frontier technologies. The verification of the code was done through the comparison between the analytical results and experimental results of the Helium Engineering Demonstration Loop Multi-channel Test Section(HENDEL T(sub 1-M)) with simulated fuel rods and fuel blocks.

  6. IAEA coordinated research project on thermal-hydraulics of Supercritical Water-Cooled Reactors (SCWRs)

    Energy Technology Data Exchange (ETDEWEB)

    Yamada, K. [Vienna International Centre, P.O. Box 100, 1400 Vienna (Austria); Aksan, S. N. [International Atomic Energy Agency, 1400 Vienna (Austria)

    2012-07-01

    The Supercritical Water-Cooled Reactor (SCWR) is an innovative water-cooled reactor concept, which uses supercritical pressure water as reactor coolant. It has been attracting interest of many researchers in various countries mainly due to its benefits of high thermal efficiency and simple primary systems, resulting in low capital cost. The IAEA started in 2008 a Coordinated Research Project (CRP) on Thermal-Hydraulics of SCWRs as a forum to foster the exchange of technical information and international collaboration in research and development. This paper summarizes the activities and current status of the CRP, as well as major progress achieved to date. At present, 15 institutions closely collaborate in several tasks. Some organizations have been conducting thermal-hydraulics experiments and analysing the data, and others have been participating in code-to-test and/or code-to-code benchmark exercises. The expected outputs of the CRP are also discussed. Finally, the paper introduces several IAEA activities relating to or arising from the CRP. (authors)

  7. Coupled neutronic and thermal-hydraulic code benchmark activities at the International Nuclear Safety Center.

    Energy Technology Data Exchange (ETDEWEB)

    Podlazov, L. N.

    1998-07-29

    Two realistic benchmark problems are defined and used to assess the performance of coupled thermal-hydraulic and neutronic codes used in simulating dynamic processes in VVER-1000 and RBMK reactor systems. One of the problems simulates a design basis accident involving the ejection of three control and protection system rods from a VVER-1000 reactor. The other is based on a postulated rod withdrawal from an operating RBMK reactor. Preliminary results calculated by various codes are compared. While these results show significant differences, the intercomparisons performed so far provide a basis for further evaluation of code limitations and modeling assumptions.

  8. Coupled neutronics and thermal hydraulics of high density cores for FRM II

    Energy Technology Data Exchange (ETDEWEB)

    Breitkreutz, Harald

    2011-03-04

    According to the 'Verwaltungsvereinbarung zwischen Bund und Land vom 30.5.2003' and its updating on 13.11.2010, the Forschungs-Neutronenquelle Heinz Maier-Leibnitz, Frm II, has to convert its fuel element to an uranium enrichment which is significantly lower than the current 93%, in case this is economically reasonable and doesn't impact the reactor performance immoderate. In the framework of this conversion, new calculations regarding neutronics and thermal hydraulics for the anticipated core configurations have to be made. The computational power available nowadays allows for detailed 3D calculations, on the neutronic as well as on the thermal hydraulic side. In this context, a new program system, 'X{sup 2}', was developed. It couples the Monte Carlo code McnpX, the computational fluid dynamics code Cfx and the burn-up code sequence MonteBurns. The codes were modified and extended to meet the requirements of the coupled calculation concept. To verify the new program system, highly detailed calculations for the current fuel element were made and compared to simulations and measurements that were performed in the past. The results strengthen the works performed so far and show that the original, conservative approach overestimates all critical thermal hydraulic values. Using the CFD software, effects like the impact of the combs that fix the fuel plates and the pressure drop at the edges of the fuel plates were studied in great detail for the first time. Afterwards, a number of possible new fuel elements with lower enrichment, based on disperse and monolithic UMo (uranium with 8 wt.-% Mo) were analysed. A number of straight-forward conversion scenarios was discussed, showing that a further compaction of the fuel element, an extended cycle length or an increased reactor power is needed to compensate the flux loss, which is caused by the lower enrichment. This flux loss is in excess of 7%. The discussed new fuel elements include a 50

  9. A General Model for Thermal, Hydraulic and Electric Analysis of Superconducting Cables

    CERN Document Server

    Bottura, L; Rosso, C

    2000-01-01

    In this paper we describe a generic, multi-component and multi-channel model for the analysis of superconducting cables. The aim of the model is to treat in a general and consistent manner simultaneous thermal, electric and hydraulic transients in cables. The model is devised for most general situations, but reduces in limiting cases to most common approximations without loss of efficiency. We discuss here the governing equations, and we write them in a matrix form that is well adapted to numerical treatment. We finally demonstrate the model capability by comparison with published experimental data on current distribution in a two-strand cable.

  10. Thermal Hydraulics Design and Analysis Methodology for a Solid-Core Nuclear Thermal Rocket Engine Thrust Chamber

    Science.gov (United States)

    Wang, Ten-See; Canabal, Francisco; Chen, Yen-Sen; Cheng, Gary; Ito, Yasushi

    2013-01-01

    Nuclear thermal propulsion is a leading candidate for in-space propulsion for human Mars missions. This chapter describes a thermal hydraulics design and analysis methodology developed at the NASA Marshall Space Flight Center, in support of the nuclear thermal propulsion development effort. The objective of this campaign is to bridge the design methods in the Rover/NERVA era, with a modern computational fluid dynamics and heat transfer methodology, to predict thermal, fluid, and hydrogen environments of a hypothetical solid-core, nuclear thermal engine the Small Engine, designed in the 1960s. The computational methodology is based on an unstructured-grid, pressure-based, all speeds, chemically reacting, computational fluid dynamics and heat transfer platform, while formulations of flow and heat transfer through porous and solid media were implemented to describe those of hydrogen flow channels inside the solid24 core. Design analyses of a single flow element and the entire solid-core thrust chamber of the Small Engine were performed and the results are presented herein

  11. Status and subjects of thermal-hydraulic analysis for next-generation LWRs with passive safety systems

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1998-03-01

    The present status and subjects on thermal-hydraulic analysis for next-generation light water reactors (LWRs) with passive safety systems were summarized based on survey results and discussion by subcommittee on improvement of reactor thermal-hydraulic analysis codes under nuclear code committee in Japan Atomic Energy Research Institute. This survey was performed to promote the research of improvement of reactor thermal-hydraulic analysis codes in future. In the first part of this report, the status and subjects on system analysis and those on evaluation of passive safety system performance are summarized for various types of reactor proposed before. In the second part, the status and subjects on multidimensional two-phase flow analysis are reviewed, since the multidimensional analysis was recognized as one of most important subjects through the investigation in the first part. Besides, databases for bubbly flow and annular dispersed flow were explored, those are needed to assess and verify each multidimensional analytical method. The contents in this report are the forefront of thermal-hydraulic analysis for LWRs and those include current findings for the development of multidimensional two-phase flow analytical method. Thus, we expect that the contents can offer various useful information against the improvement of reactor thermal-hydraulic analysis codes in future. (author)

  12. Thermal-hydraulic analysis techniques for axisymmetric pebble bed nuclear reactor cores. [PEBBLE code

    Energy Technology Data Exchange (ETDEWEB)

    Stroh, K.R.

    1979-03-01

    The pebble bed reactor's cylindrical core volume contains a random bed of small, spherical fuel-moderator elements. These graphite spheres, containing a central region of dispersed coated-particle fissile and fertile material, are cooled by high pressure helium flowing through the connected interstitial voids. A mathematical model and numerical solution technique have been developed which allow calculation of macroscopic values of thermal-hydraulic variables in an axisymmetric pebble bed nuclear reactor core. The computer program PEBBLE is based on a mathematical model which treats the bed macroscopically as a generating, conducting porous medium. The steady-state model uses a nonlinear Forchheimer-type relation between the coolant pressure gradient and mass flux, with newly derived coefficients for the linear and quadratic resistance terms. The remaining equations in the model make use of mass continuity, and thermal energy balances for the solid and fluid phases.

  13. Thermal-hydraulic analysis of an irregular sector of the ITER vacuum vessel by means of CFD tools

    Energy Technology Data Exchange (ETDEWEB)

    Fradera, J., E-mail: jorge.fradera@idom.com [Idom Nuclear Services, Gran Vía Carlos III, 97 Bajos, 08028 Barcelona (Spain); Colomer, C.; Fabbri, M.; Martín, M.; Martínez-Sabán, E.; Zamora, I.; Alemán, A. [Idom Nuclear Services, Gran Vía Carlos III, 97 Bajos, 08028 Barcelona (Spain); Izquierdo, J. [F4E, Barcelona (Spain); Le Barbier, R.; Utin, Y. [ITER IO, Cadarache (France)

    2015-03-15

    Highlights: • 3D Geometry healing and simplification for CFD simulations. • Meshing of large domains for CFD simulations. • Meshing procedure for fluid–solid interface coupling. • Hydraulics of the ITER VV Irregular Sector number 2 (IrS#2). • Thermal-hydraulics of the ITER VV IrS#2. - Abstract: The present work exposes the 3D thermal-hydraulic analysis of the Irregular Sector number 2 (IrS#2) of the ITER Vacuum Vessel (VV) by means of CFD (computational fluid dynamics). IrS#2 geometry has been simplified and healed in order to be suitable for CFD analysis. A polyhedral cell based mesh has been generated so as to enhance accuracy and calculation stability. Nuclear heat deposition has been implemented through several subroutines and an in-house MCNP data converter. Water coolant and stainless steel shell are solved coupled as a steady-state conjugate heat transfer problem in order to assess the impact of the nuclear heat deposition on the IrS#2 cooling scheme. Hence, the IrS#2 is simulated as a whole without splitting the domain. Results show the total IrS#2 pressure drop as well as the flow and temperature distribution all over the IrS#2. Moreover, heat transfer coefficient has been calculated at the water–shell interface in order to assess the behavior of shell cooling scheme. Velocity magnitude in the water coolant has an average value of 2 cm/s and the inboard to outboard mass flow rate distribution is 10.2% and 89.8% respectively. Pressure drop, mainly at inlet and outlet ducts, is of 60.21 kPa. Temperature at the liquid–solid interface has an average value of 106.4 °C and the heat transfer coefficient (HTC) stays always above 638 W/(m{sup 2} °C), way above the limit of 500 W/(m{sup 2} °C). Shell temperature stays at an average value of 130.0 °C. Exposed results, with a significant importance regarding design and safety, give a valuable insight on current cooling scheme and system behavior for the IrS#2 of the ITER VV.

  14. 2-D CFD time-dependent thermal-hydraulic simulations of CANDU-6 moderator flows

    Energy Technology Data Exchange (ETDEWEB)

    Mehdi Zadeh, Foad [Department of Engineering Physics/Polytechnique Montréal, Montréal, QC (Canada); Étienne, Stéphane [Department of Mechanical Engineering/Polytechnique Montréal, Montréal, QC (Canada); Teyssedou, Alberto, E-mail: alberto.teyssedou@polymtl.ca [Department of Engineering Physics/Polytechnique Montréal, Montréal, QC (Canada)

    2016-12-01

    Highlights: • 2-D time-dependent CFD simulations of CANDU-6 moderator flows are presented. • A thermal-hydraulic code using thermal physical fluid properties is used. • The numerical approach and convergence is validated against available data. • Flow configurations are correlated using Richardson’s number. • Frequency components indicate moderator flow oscillations vs. Richardson numbers. - Abstract: The distribution of the fluid temperature and mass density of the moderator flow in CANDU-6 nuclear power reactors may affect the reactivity coefficient. For this reason, any possible moderator flow configuration and consequently the corresponding temperature distributions must be studied. In particular, the variations of the reactivity may result in major safety issues. For instance, excessive temperature excursions in the vicinity of the calandria tubes nearby local flow stagnation zones, may bring about partial boiling. Moreover, steady-state simulations have shown that for operating condition, intense buoyancy forces may be dominant, which can trigger a thermal stratification. Therefore, the numerical study of the time-dependent flow transition to such a condition, is of fundamental safety concern. Within this framework, this paper presents detailed time-dependent numerical simulations of CANDU-6 moderator flow for a wide range of flow conditions. To get a better insight of the thermal-hydraulic phenomena, the simulations were performed by covering long physical-time periods using an open-source code (Code-Saturne V3) developed by Électricité de France. The results show not only a region where the flow is characterized by coherent structures of flow fluctuations but also the existence of two limit cases where fluid oscillations disappear almost completely.

  15. Thermal-Hydraulic Sensitivity Study of Intermediate Loop Parameters for Nuclear Hydrogen Production System

    Energy Technology Data Exchange (ETDEWEB)

    Jeong, Jong Hwa; Lee, Heung Nae; Park, Jea Ho [KONES Corp., Seoul (Korea, Republic of); Lee, Won Jae [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Lee, Sang Il; Yoo, Yeon Jae [Hyundai Engineering Co., Seoul (Korea, Republic of)

    2016-10-15

    The heat generated from the VHTR is transferred to the intermediate loop through Intermediate Heat Exchanger (IHX). It is further passed on to the Sulfur-Iodine (SI) hydrogen production system (HPS) through Process Heat Exchanger (PHX). The IL provides the safety distance between the VHTR and HPS. Since the IL performance affects the overall nuclear HPS efficiency, it is required to optimize its design and operation parameters. In this study, the thermal-hydraulic sensitivity of IL parameters with various coolant options has been examined by using MARS-GCR code, which was already applied for the case of steam generator. Sensitivity study of the IL and PHX parameters has been carried out based on their thermal-hydraulic performance. Several parameters for design and operation, such as the pipe diameter, safety distance and surface area, are considered for different coolant options, He, CO{sub 2} and He-CO{sub 2} (2:8). It was found that the circulator work is the major factor affecting on the overall nuclear hydrogen production system efficiency. Circulator work increases with the safety distance, and decreases with the operation pressure and loop pipe diameter. Sensitivity results obtained from this study will contribute to the optimization of the IL design and operation parameters and the optimal coolant selection.

  16. Neutronic and Thermal-hydraulic Modelling of High Performance Light Water Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Seppaelae, Malla [VTT Technical Research Centre of Finland, P.O.Box 1000, FI02044 VTT (Finland)

    2008-07-01

    High Performance Light Water Reactor (HPLWR), which is studied in EU project 'HPLWR2', uses water at supercritical pressures as coolant and moderator to achieve higher core outlet temperature and thus higher efficiency compared to present reactors. At VTT Technical Research Centre of Finland, functionality of the thermal-hydraulics in the coupled reactor dynamics code TRAB3D/ SMABRE was extended to supercritical pressures for the analyses of HPLWR. Input models for neutronics and thermal-hydraulics were made for TRAB3D/ SMABRE according to the latest HPLWR design. A preliminary analysis was performed in which the capability of SMABRE in the transition from supercritical pressures to subcritical pressures was demonstrated. Parameterized two-group cross sections for TRAB3D neutronics were received from Hungarian Academy of Sciences KFKI Atomic Energy Research Institute together with a subroutine for handling them. PSG, a new Monte Carlo transport code developed at VTT, was also used to generate two-group constants for HPLWR and comparisons were made with the KFKI cross sections and MCNP calculations. (author)

  17. Development of thermal-hydraulic models for the safety evaluation of CANDU reactors

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Jae Young; Jung, Yun Sik; Hwang, Gi Suk; Kim, Nam Seok [Handong Univ., Pohang (Korea, Republic of); No, Hee Cheon [Korea Advanced Institute of Science and Technology, Taejon (Korea, Republic of)

    2004-02-15

    The objective of the present research is to evaluate the safety analysis for CANDU and to improve the Horizontal Stratification Entrainment Model (HSEM) of RELAP5/MOD3.3. This report includes two items the one is the experimental study of entrainment at horizontal pipe with {+-} 36 .deg. C , {+-} 72 .deg. C branch pies, the other is the model improvement of the moderator heat sink in the Calandria. The off-take experiments on onset of entrainment and branch quality were investigated by using water and air as working fluid, and the experimental data were compared by the previous correlations. The previous correlations could not expect experimental results, thus the weak points of the previous correlations were investigated. The improvement of the previous model continues as the next year research. The thermal hydraulic scaling analysis of SPEL, STERN and ideal linear scaling analysis have been studied. As a result, a new scaling method were needed to design a new experimental facility (HGU). A new scaling method with 1/8 length scale was applied. From these results, the thermal hydraulic model for CFD code simulation was designed and test apparatus has been made. The moderator temperature distribution experiments and CFD code simulation will be continued in next year.

  18. Thermal Hydraulic design parameters study for severe accidents using neural networks

    Energy Technology Data Exchange (ETDEWEB)

    Roh, Chang Hyun; Chang, Soon Heung [Korea Advanced Institute of Science and Technology, Taejon (Korea, Republic of); Chang, Keun Sun [Sunmoon University, Asan (Korea, Republic of)

    1997-12-31

    To provide the information on severe accident progression is very important for advanced or new type of nuclear power plant (NPP) design. A parametric study, therefore, was performed to investigate the effect of thermal hydraulic design parameters on severe accident progression of pressurized water reactors (PWRs). Nine parameters, which are considered important in NPP design or severe accident progression, were selected among the various thermal hydraulic design parameters. The backpropagation neural network (BPN) was used to determine parameters, which might more strongly affect the severe accident progression, among nine parameters. For training, different input patterns were generated by the latin hypercube sampling (LHS) technique and then different target patterns that contain core uncovery time and vessel failure time were obtained for Young Gwang Nuclear (YGN) Units 3 and 4 using modular accident analysis program (MAAP) 3.0B code. Three different severe accident scenarios, such as two loss of coolant accidents (LOCAs) and station blackout (SBO), were considered in this analysis. Results indicated that design parameters related to refueling water storage tank (RWST), accumulator and steam generator (S/G) have more dominant effects on the progression of severe accidents investigated, compared to the other six parameters. 9 refs., 5 tabs. (Author)

  19. Investigation of film boiling thermal hydraulics under FCI conditions. Results of a numerical study

    Energy Technology Data Exchange (ETDEWEB)

    Dinh, T.N.; Dinh, A.T.; Nourgaliev, R.R.; Sehgal, B.R. [Div. of Nuclear Power Safety Royal Inst. of Tech. (RIT), Brinellvaegen 60, 10044 Stockholm (Sweden)

    1998-01-01

    Film boiling on the surface of a high-temperature melt jet or of a melt particle is one of key phenomena governing the physics of fuel-coolant interactions (FCIs) which may occur during the course of a severe accident in a light water reactor (LWR). A number of experimental and analytical studies have been performed, in the past, to address film boiling heat transfer and the accompanying hydrodynamic aspects. Most of the experiments have, however, been performed for temperature and heat flux conditions, which are significantly lower than the prototypic conditions. For ex-vessel FCIs, high liquid subcooling can significantly affect the FCI thermal hydraulics. Presently, there are large uncertainties in predicting natural-convection film boiling of subcooled liquids on high-temperature surfaces. In this paper, research conducted at the Division of Nuclear Power Safety, Royal Institute of Technology (RIT/NPS), Stockholm, concerning film-boiling thermal hydraulics under FCI condition is presented. Notably, the focus is placed on the effects of (1) water subcooling, (2) high-temperature steam properties, (3) the radiation heat transfer and (4) mixing zone boiling dynamics, on the vapor film characteristics. Numerical investigations are performed using a novel CFD modeling concept named as the local-homogeneous-slip model (LHSM). Results of the analytical and numerical studies are discussed with respect to boiling dynamics under FCI conditions. (author)

  20. Porosity Effect in the Core Thermal Hydraulics for Ultra High Temperature Gas-cooled Reactor

    Directory of Open Access Journals (Sweden)

    Motoo Fumizawa

    2008-12-01

    Full Text Available This study presents an experimental method of porosity evaluation and a predictive thermal-hydraulic analysis with packed spheres in a nuclear reactor core. The porosity experiments were carried out in both a fully shaken state with the closest possible packing and in a state of non-vibration. The predictive analysis considering the fixed porosity value was applied as a design condition for an Ultra High Temperature Reactor Experiment (UHTREX. The thermal-hydraulic computer code was developed and identified as PEBTEMP. The highest outlet coolant temperature of 1316 oC was achieved in the case of an UHTREX at Los Alamos Scientific Laboratory, which was a small scale UHTR. In the present study, the fuel was changed to a pebble type, a porous media. In order to compare the present pebble bed reactor and UHTREX, a calculation based on HTGR-GT300 was carried out in similar conditions with UHTREX; in other words, with an inlet coolant temperature of 871oC, system pressure of 3.45 MPa and power density of 1.3 w/cm3. As a result, the fuel temperature in the present pebble bed reactor showed an extremely lower value compared to that of UHTREX.

  1. Comparative study of boron transport models in NRC Thermal-Hydraulic Code Trace

    Energy Technology Data Exchange (ETDEWEB)

    Olmo-Juan, Nicolás; Barrachina, Teresa; Miró, Rafael; Verdú, Gumersindo; Pereira, Claubia, E-mail: nioljua@iqn.upv.es, E-mail: tbarrachina@iqn.upv.es, E-mail: rmiro@iqn.upv.es, E-mail: gverdu@iqn.upv.es, E-mail: claubia@nuclear.ufmg.br [Institute for Industrial, Radiophysical and Environmental Safety (ISIRYM). Universitat Politècnica de València (Spain); Universidade Federal de Minas Gerais (UFMG), Belo Horizonte, MG (Brazil). Departamento de Engenharia Nuclear

    2017-07-01

    Recently, the interest in the study of various types of transients involving changes in the boron concentration inside the reactor, has led to an increase in the interest of developing and studying new models and tools that allow a correct study of boron transport. Therefore, a significant variety of different boron transport models and spatial difference schemes are available in the thermal-hydraulic codes, as TRACE. According to this interest, in this work it will be compared the results obtained using the different boron transport models implemented in the NRC thermal-hydraulic code TRACE. To do this, a set of models have been created using the different options and configurations that could have influence in boron transport. These models allow to reproduce a simple event of filling or emptying the boron concentration in a long pipe. Moreover, with the aim to compare the differences obtained when one-dimensional or three-dimensional components are chosen, it has modeled many different cases using only pipe components or a mix of pipe and vessel components. In addition, the influence of the void fraction in the boron transport has been studied and compared under close conditions to BWR commercial model. A final collection of the different cases and boron transport models are compared between them and those corresponding to the analytical solution provided by the Burgers equation. From this comparison, important conclusions are drawn that will be the basis of modeling the boron transport in TRACE adequately. (author)

  2. Thermal-hydraulic analysis of research reactor core with different LEU fuel types using RELAP5

    Energy Technology Data Exchange (ETDEWEB)

    El-Sahlamy, Neama M. [Nuclear and Radiological Regulatory Authority, Cairo (Egypt)

    2017-11-15

    In the current work, comparisons between the core performances when using different LEU fuels are done. The fuels tested are UA1{sub X}-A1, U{sub 3}O{sub 8}-Al, and U{sub 3}Si{sub 2}-Al fuels with 19.7 % enrichment. Calculations are done using RELAP5 code to evaluate the thermal-hydraulic performance of the IAEA benchmark 10 MW reactor. First, a reassessment of the slow reactivity insertion transient with UA1{sub X}-A1 LEU fuel to compare the results with those reported in the IAEA TECDOC [1]. Then, comparisons between the thermal-hydraulic core performances when using the three LEU fuels are done. The assessment is performed at initial power of 1.0 W. The reactor power is calculated using the RELAP5 point kinetic model. The reactivity feedback, from changes in water density and fuel temperature, is considered for all cases. From the results it is noticed that U{sub 3}Si{sub 2}-Al fuel gives the best fuel performance since it has the minimum value of peak fuel temperature and the minimum peak clad surface temperature, as operating parameters. Also, it gives the maximum value of the Critical Heat Flux Ratio and the lowest tendency to flow instability occurrence.

  3. Development of the NSSS thermal-hydraulic program for YGN unit 1 simulator

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Kyung Doo; Jeong, Jae Jun; Lee, Won Jae; Chung, Bub Dong; Ha, Kwi Seok; Kang, Kyung Ho

    2000-09-01

    The NSSS thermal-hydraulic programs installed in the domestic full-scope power plant simulators were provided in early 1980s by foreign vendors. Because of limited computational capability at that time, they usually adopt very simplified physical models for a real-time simulation of NSSS thermal-hydraulic phenomena, which entails inaccurate results and the possibility of so-called 'negative training', especially for complicated two-phase flows in the reactor coolant system. To resolve the problem, we developed a realistic NSSS T/H program (named 'ARTS' code) for use in YongGwang Nuclear Unit 1 full-scope simulator. The best-estimate code RETRAN03, developed by EPRI and approved by USNRC, was selected as a reference code of ARTS. For the development of ARTS, the followings have been performed: -Improvement of the robustness of RETRAN - Improvement of the real-time simulation capability of RETRAN - Optimum input data generation for the NSSS simulation - New model development that cannot be efficiently modeled by RETRAN - Assessment of the ARTS code. The systematic assessment of ARTS has been conducted in both personal computers (Windows 98, Visual fortran) and the simulator development environment (Windows NT, GSE simulator development tool). The results were resonable in terms of accuracy, real-time simulation and robustness.

  4. Thermal-hydraulic criteria for the APT tungsten neutron source design

    Energy Technology Data Exchange (ETDEWEB)

    Pasamehmetoglu, K.

    1998-03-01

    This report presents the thermal-hydraulic design criteria (THDC) developed for the tungsten neutron source (TNS). The THDC are developed for the normal operations, operational transients, and design-basis accidents. The requirements of the safety analyses are incorporated into the design criteria, consistent with the integrated safety management and the safety-by-design philosophy implemented throughout the APT design process. The phenomenology limiting the thermal-hydraulic design and the confidence level requirements for each limit are discussed. The overall philosophy of the uncertainty analyses and the confidence level requirements also are presented. Different sets of criteria are developed for normal operations, operational transients, anticipated accidents, unlikely accidents, extremely unlikely accidents, and accidents during TNS replacement. In general, the philosophy is to use the strictest criteria for the high-frequency events. The criteria is relaxed as the event frequencies become smaller. The THDC must be considered as a guide for the design philosophy and not as a hard limit. When achievable, design margins greater than those required by the THDC must be used. However, if a specific event sequence cannot meet the THDC, expensive design changes are not necessary if the single event sequence results in sufficient margin to safety criteria and does not challenge the plant availability or investment protection considerations.

  5. Development of an integrated thermal-hydraulics capability incorporating RELAP5 and PANTHER neutronics code

    Energy Technology Data Exchange (ETDEWEB)

    Page, R.; Jones, J.R.

    1997-07-01

    Ensuring that safety analysis needs are met in the future is likely to lead to the development of new codes and the further development of existing codes. It is therefore advantageous to define standards for data interfaces and to develop software interfacing techniques which can readily accommodate changes when they are made. Defining interface standards is beneficial but is necessarily restricted in application if future requirements are not known in detail. Code interfacing methods are of particular relevance with the move towards automatic grid frequency response operation where the integration of plant dynamic, core follow and fault study calculation tools is considered advantageous. This paper describes the background and features of a new code TALINK (Transient Analysis code LINKage program) used to provide a flexible interface to link the RELAP5 thermal hydraulics code with the PANTHER neutron kinetics and the SIBDYM whole plant dynamic modelling codes used by Nuclear Electric. The complete package enables the codes to be executed in parallel and provides an integrated whole plant thermal-hydraulics and neutron kinetics model. In addition the paper discusses the capabilities and pedigree of the component codes used to form the integrated transient analysis package and the details of the calculation of a postulated Sizewell `B` Loss of offsite power fault transient.

  6. Multi channel thermal hydraulic analysis of gas cooled fast reactor using genetic algorithm

    Energy Technology Data Exchange (ETDEWEB)

    Drajat, R. Z.; Su' ud, Z.; Soewono, E.; Gunawan, A. Y. [Department of Mathematics, Institut Teknologi Bandung, Bandung 40132 (Indonesia); Department of Physics, Institut Teknologi Bandung, Bandung 40132 (Indonesia); Department of Mathematics, Institut Teknologi Bandung, Bandung 40132 (Indonesia)

    2012-05-22

    There are three analyzes to be done in the design process of nuclear reactor i.e. neutronic analysis, thermal hydraulic analysis and thermodynamic analysis. The focus in this article is the thermal hydraulic analysis, which has a very important role in terms of system efficiency and the selection of the optimal design. This analysis is performed in a type of Gas Cooled Fast Reactor (GFR) using cooling Helium (He). The heat from nuclear fission reactions in nuclear reactors will be distributed through the process of conduction in fuel elements. Furthermore, the heat is delivered through a process of heat convection in the fluid flow in cooling channel. Temperature changes that occur in the coolant channels cause a decrease in pressure at the top of the reactor core. The governing equations in each channel consist of mass balance, momentum balance, energy balance, mass conservation and ideal gas equation. The problem is reduced to finding flow rates in each channel such that the pressure drops at the top of the reactor core are all equal. The problem is solved numerically with the genetic algorithm method. Flow rates and temperature distribution in each channel are obtained here.

  7. Improvement of auditing technology of safety analysis through thermal-hydraulic separate effect tests

    Energy Technology Data Exchange (ETDEWEB)

    No, Hee Cheon; Moon, Young Min; Lee, Dong Won; Lee, Sang Ik; Kim, Eung Soo; Yeom, Keum Soo [Korea Advanced Institute of Science and Technology, Taejon (Korea, Republic of)

    2002-03-15

    The objective of the present research is to perform the separate effect tests and to assess the RELAP5/MOD3.2 code for the analysis of thermal-hydraulic behavior in the reactor coolant system and the improvement of the auditing technology of safety analysis. Three Separate Effect Tests (SETs) are the reflux condensation in the U-tube, the direct contact condensation in the hot-leg and the mixture level buildup in the pressurizer. The experimental data and the empirical correlations are obtained through SETs. On the ases of the three SET works, models in RELAP5 are modified and improved, which are compared with the data. The Korea Standard Nuclear Power Plant (KSNP) are assessed using the modified RELAP5. In the reflux condensation test, the data of heat transfer coefficients and flooding are obtained and the condensation models are modified using the non-iterative model, as results, modified code better predicts the data. In the direct contact condensation test, the data of heat transfer coefficients are obtained for the cocurrent and countercurrent flow between the mixture gas and the water in condition of horizontal stratified flow. Several condensation and friction models are modified, which well predict the present data. In the mixture level test, the data for the mixture level and the onset of water draining into the surge line are obtained. The standard RELAP5 over-predicts the mixture level and the void fraction in the pressurizer. Simple modification of model related to the pool void fraction is suggested. The KSNP is assessed using the standard and the modified RELAP5 resulting from the experimental and code works for the SETs. In case of the pressurizer manway opening with available secondary side of the steam generators, the modified code predicts that the collapsed level in the pressurizer is little accumulated. The presence and location of the opening and the secondary condition of the steam generators have an effect on the coolant inventory. The

  8. Creation of an Enhanced Geothermal System through Hydraulic and Thermal Stimulation

    Energy Technology Data Exchange (ETDEWEB)

    Rose, Peter Eugene [Energy and Geoscience Institute at the University of Utah

    2013-04-15

    This report describes a 10-year DOE-funded project to design, characterize and create an Engineered Geothermal System (EGS) through a combination of hydraulic, thermal and chemical stimulation techniques. Volume 1 describes a four-year Phase 1 campaign, which focused on the east compartment of the Coso geothermal field. It includes a description of the geomechanical, geophysical, hydraulic, and geochemical studies that were conducted to characterize the reservoir in anticipation of the hydraulic stimulation experiment. Phase 1 ended prematurely when the drill bit intersected a very permeable fault zone during the redrilling of target stimulation well 34-9RD2. A hydraulic stimulation was inadvertently achieved, however, since the flow of drill mud from the well into the formation created an earthquake swarm near the wellbore that was recorded, located, analyzed and interpreted by project seismologists. Upon completion of Phase 1, the project shifted focus to a new target well, which was located within the southwest compartment of the Coso geothermal field. Volume 2 describes the Phase 2 studies on the geomechanical, geophysical, hydraulic, and geochemical aspects of the reservoir in and around target-stimulation well 46A-19RD, which is the deepest and hottest well ever drilled at Coso. Its total measured depth exceeding 12,000 ft. It spite of its great depth, this well is largely impermeable below a depth of about 9,000 ft, thus providing an excellent target for stimulation. In order to prepare 46A-19RD for stimulation, however, it was necessary to pull the slotted liner. This proved to be unachievable under the budget allocated by the Coso Operating Company partners, and this aspect of the project was abandoned, ending the program at Coso. The program then shifted to the EGS project at Desert Peak, which had a goal similar to the one at Coso of creating an EGS on the periphery of an existing geothermal reservoir. Volume 3 describes the activities that the Coso team

  9. Development of the thermal hydraulic analysis code for a copper bonded steam generator

    Energy Technology Data Exchange (ETDEWEB)

    Kim, W. K.; Wei, M. H.; Yeo, J. H.; Kim, S. O. [KAERI, Taejon (Korea, Republic of); Back, B. J. [Chonbuk National Univ., Jeonju (Korea, Republic of)

    2002-10-01

    An one-dimensional thermal-hydraulic analysis computer code was developed for the thermal sizing of copper bonded steam generator. It was assumed that the conduction heat transfer of copper region between hot side and cold side tube is one-dimensional and its thermal resistance of the function of a tube pitch was derived. The flow regions of water/steam side were devided into four regions, which are sub-cooled, saturated, film boiling, and super-heated regions. The numbers of tube were selected from 250 to 3500 for the parameter study calculation. The pitch over tube diameter ratios were 1.4, 1.6 and 1.8. The calculation results showed that when the number of tube was 2500, the length of heating tube was about 10 m and the diameter was about 3 m. If P/D ratio increases, the thermal resistance of copper component also increases, however the length of heating tube is not increasing so much.

  10. Thermal-Hydraulic Performance of the TREAT Multi-SERTTA for Reactivity Initiated Accident Experiments

    Energy Technology Data Exchange (ETDEWEB)

    Jensen, Colby B.; Folsom, Charles P.; Davis, Cliff B.; Woolstenhulme, Nicolas E.; Bess, John D.; O' Brien, Robert C.; Ban, Heng; Wachs, Daniel M.

    2016-08-01

    Experimental testing in the Multi-Static Environment Rodlet Transient Test Apparatus (SERTTA) will lead the rebirth of transient fuel testing in the United States as part of the Accident Tolerant Fuels (ATF) progam. The Multi-SERTTA is comprised of four isolated pressurized environments capable of a wide variety of working fluids and thermal conditions. Ultimately, the TREAT reactor as well as the Multi-SERTTA test vehicle serve the purpose of providing desired thermal-hydraulic boundary conditions to the test specimen. The initial ATF testing in TREAT will focus on reactivity insertion accident (RIA) events using both gas and water environments including typical PWR operating pressures and temperatures. For the water test environment, a test configuration is envisioned using the expansion tank as part of the gas-filled expansion volume seen by the test to provide additional pressure relief. The heat transfer conditions during the high energy power pulses of RIA events remains a subject of large uncertainty and great importance for fuel performance predictions. To support transient experiments, the Multi-SERTTA vehicle has been modeled using RELAP5 with a baseline test specimen composed of UO2 fuel in zircaloy cladding. The modeling results show the influence of the designs of the specimen, vehicle, and transient power pulses. The primary purpose of this work is to provide input and boundary conditions to fuel performance code BISON. Therefore, studies of parameters having influence on specimen performance during RIA transients are presented including cladding oxidation, power pulse magnitude and width, cladding-to-coolant heat fluxes, fuel-to-cladding gap, transient boiling effects (modified CHF values), etc. The results show the great flexibility and capacity of the TREAT Multi-SERTTA test vehicle to provide testing under a wide range of prototypic thermal-hydraulic conditions as never done before.

  11. Thermal hydraulic-severe accident code interfaces for SCDAP/RELAP5/MOD3.2

    Energy Technology Data Exchange (ETDEWEB)

    Coryell, E.W.; Siefken, L.J.; Harvego, E.A. [Idaho National Engineering Lab., Idaho Falls, ID (United States)] [and others

    1997-07-01

    The SCDAP/RELAP5 computer code is designed to describe the overall reactor coolant system thermal-hydraulic response, core damage progression, and fission product release during severe accidents. The code is being developed at the Idaho National Engineering Laboratory under the primary sponsorship of the Office of Nuclear Regulatory Research of the U.S. Nuclear Regulatory Commission. The code is the result of merging the RELAP5, SCDAP, and COUPLE codes. The RELAP5 portion of the code calculates the overall reactor coolant system, thermal-hydraulics, and associated reactor system responses. The SCDAP portion of the code describes the response of the core and associated vessel structures. The COUPLE portion of the code describes response of lower plenum structures and debris and the failure of the lower head. The code uses a modular approach with the overall structure, input/output processing, and data structures following the pattern established for RELAP5. The code uses a building block approach to allow the code user to easily represent a wide variety of systems and conditions through a powerful input processor. The user can represent a wide variety of experiments or reactor designs by selecting fuel rods and other assembly structures from a range of representative core component models, and arrange them in a variety of patterns within the thermalhydraulic network. The COUPLE portion of the code uses two-dimensional representations of the lower plenum structures and debris beds. The flow of information between the different portions of the code occurs at each system level time step advancement. The RELAP5 portion of the code describes the fluid transport around the system. These fluid conditions are used as thermal and mass transport boundary conditions for the SCDAP and COUPLE structures and debris beds.

  12. An analytical study on excitation of nuclear-coupled thermal-hydraulic instability due to seismically induced resonance in BWR

    Energy Technology Data Exchange (ETDEWEB)

    Hirano, Masashi [Japan Atomic Energy Research Institute, Ibaraki-ken (Japan)

    1997-07-01

    This paper describes the results of a scoping study on seismically induced resonance of nuclear-coupled thermal-hydraulic instability in BWRs, which was conducted by using TRAC-BF1 within a framework of a point kinetics model. As a result of the analysis, it is shown that a reactivity insertion could occur accompanied by in-surge of coolant into the core resulted from the excitation of the nuclear-coupled instability by the external acceleration. In order to analyze this phenomenon more in detail, it is necessary to couple a thermal-hydraulic code with a three-dimensional nuclear kinetics code.

  13. Thermal Properties for the Thermal-Hydraulics Analyses of the BR2 Maximum Nominal Heat Flux

    Energy Technology Data Exchange (ETDEWEB)

    Dionne, B. [Argonne National Lab. (ANL), Argonne, IL (United States). Nuclear Engineering Division; Bergeron, A. [Argonne National Lab. (ANL), Argonne, IL (United States). Nuclear Engineering Division; Licht, J. R. [Argonne National Lab. (ANL), Argonne, IL (United States). Nuclear Engineering Division; Kim, Y. S. [Argonne National Lab. (ANL), Argonne, IL (United States). Nuclear Engineering Division; Hofman, G. L. [Argonne National Lab. (ANL), Argonne, IL (United States). Nuclear Engineering Division

    2015-02-01

    This memo describes the assumptions and references used in determining the thermal properties for the various materials used in the BR2 HEU (93% enriched in 235U) to LEU (19.75% enriched in 235U) conversion feasibility analysis. More specifically, this memo focuses on the materials contained within the pressure vessel (PV), i.e., the materials that are most relevant to the study of impact of the change of fuel from HEU to LEU. Section 2 provides a summary of the thermal properties in the form of tables while the following sections and appendices present the justification of these values. Section 3 presents a brief background on the approach used to evaluate the thermal properties of the dispersion fuel meat and specific heat capacity. Sections 4 to 7 discuss the material properties for the following materials: i) aluminum, ii) dispersion fuel meat (UAlx-Al and U-7Mo-Al), iii) beryllium, and iv) stainless steel. Section 8 discusses the impact of irradiation on material properties. Section 9 summarizes the material properties for typical operating temperatures. Appendix A elaborates on how to calculate dispersed phase’s volume fraction. Appendix B provides a revised methodology for determining the thermal conductivity as a function of burnup for HEU and LEU.

  14. Integrated assessment of thermal hydraulic processes in W7-X fusion experimental facility

    Energy Technology Data Exchange (ETDEWEB)

    Kaliatka, T., E-mail: tadas.kaliatka@lei.lt; Uspuras, E.; Kaliatka, A.

    2017-02-15

    Highlights: • The model of Ingress of Coolant Event experiment facility was developed using the RELAP5 code. • Calculation results were compared with Ingress of Coolant Event experiment data. • Using gained experience, the numerical model of Wendelstein 7-X facility was developed. • Performed analysis approved pressure increase protection system for LOCA event. - Abstract: Energy received from the nuclear fusion reaction is one of the most promising options for generating large amounts of carbon-free energy in the future. However, physical and technical problems existing in this technology are complicated. Several experimental nuclear fusion devices around the world have already been constructed, and several are under construction. However, the processes in the cooling system of the in-vessel components, vacuum vessel and pressure increase protection system of nuclear fusion devices are not widely studied. The largest amount of radioactive materials is concentrated in the vacuum vessel of the fusion device. Vacuum vessel is designed for the vacuum conditions inside the vessel. Rupture of the in-vessel components of the cooling system pipe may lead to a sharp pressure increase and possible damage of the vacuum vessel. To prevent the overpressure, the pressure increase protection system should be designed and implemented. Therefore, systematic and detailed experimental and numerical studies, regarding the thermal-hydraulic processes in cooling system, vacuum vessel and pressure increase protection system, are important and relevant. In this article, the numerical investigation of thermal-hydraulic processes in cooling systems of in-vessel components, vacuum vessels and pressure increase protection system of fusion devices is presented. Using the experience gained from the modelling of “Ingress of Coolant Event” experimental facilities, the numerical model of Wendelstein 7-X (W7-X) experimental fusion device was developed. The integrated analysis of the

  15. Links between matrix bulk density, macropore characteristics and hydraulic behavior of soils

    DEFF Research Database (Denmark)

    Katuwal, Sheela; Møldrup, Per; Lamandé, Mathieu

    2013-01-01

    The relationship of soil bulk density with the hydraulic behavior of soil and the role of macropores in preferential flow and transport has been extensively studied in literatures. Yet, the influence of soil structural heterogeneity as simultaneous variation of bulk density and macropore characte......The relationship of soil bulk density with the hydraulic behavior of soil and the role of macropores in preferential flow and transport has been extensively studied in literatures. Yet, the influence of soil structural heterogeneity as simultaneous variation of bulk density and macropore...... characteristics on soil hydraulic functions has rarely been studied. With the objective of studying the links between these parameters we quantified macropore characteristics of intact soil columns (19 cm diameter x 20 cm high) from two agricultural field sites (Silstrup and Faardrup) in Denmark using coarse....... However, in Faardrup soils macroporosity and matrix bulk density were weakly correlated. Though macroporosity and interconnectivity of macropores could explain some variation in the hydraulic behavior of a few samples from Faardrup, the preferential flow behavior was observed to be primarily influenced...

  16. Thermal and Hydraulic Performances of Nanofluids Flow in Microchannel Heat Sink with Multiple Zigzag Flow Channels

    Directory of Open Access Journals (Sweden)

    Duangthongsuk Weerapun

    2017-01-01

    Full Text Available This article presents an experimental investigation on the heat transfer performance and pressure drop characteristic of two types of nanofluids flowing through microchannel heat sink with multiple zigzag flow channel structures (MZMCHS. SiO2 nanoparticles dispersed in DI water with concentrations of 0.3 and 0.6 vol.% were used as working fluid. MZMCHS made from copper material with dimension of 28 × 33 mm. Hydraulic diameter of MZMCHs is designed at 1 mm, 7 number of flow channels and heat transfer area is about 1,238 mm2. Effects of particle concentration and flow rate on the thermal and hydraulic performances are determined and then compare with the common base fluid. The results indicated that the heat transfer coefficient of nanofluids was higher than that of the water and increased with increasing particle concentration as well as Reynolds number. For pressure drop, the particle concentrations have no significant effect on the pressure drop across the test section.

  17. Development of a thermal-hydraulic code for reflood analysis in a PWR experimental loop

    Energy Technology Data Exchange (ETDEWEB)

    Alves, Sabrina P.; Mesquita, Amir Z.; Rezende, Hugo C., E-mail: sabrinapral@gmail.com, E-mail: amir@cdtn.brm, E-mail: hcr@cdtn.br, E-mail: hcr@cdtn.br [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN/CNEN-MG), Belo Horizonte, MG (Brazil); Palma, Daniel A.P., E-mail: dapalma@cnen.gov.br [Comissão Nacional de Energia Nuclear (CNEN), Rio de Janeiro, RJ (Brazil)

    2017-07-01

    A process of fundamental importance in the event of Loss of Coolant Accident (LOCA) in Pressurized Water nuclear Reactors (PWR) is the reflood of the core or rewetting of nuclear fuels. The Nuclear Technology Development Center (CDTN) has been developing since the 70’s programs to allow Brazil to become independent in the field of reactor safety analysis. To that end, in the 80’s was designed, assembled and commissioned one Rewetting Test Facility (ITR in Portuguese). This facility aims to investigate the phenomena involved in the thermal hydraulic reflood phase of a Loss of Coolant Accident in a PWR nuclear reactor. This work aim is the analysis of physical and mathematical models governing the rewetting phenomenon, and the development a thermo-hydraulic simulation code of a representative experimental circuit of the PWR reactors core cooling channels. It was possible to elaborate and develop a code called REWET. The results obtained with REWET were compared with the experimental results of the ITR, and with the results of the Hydroflut code, that was the old program previously used. An analysis was made of the evolution of the wall temperature of the test section as well as the evolution of the front for two typical tests using the two codes calculation, and experimental results. The result simulated by REWET code for the rewetting time also came closer to the experimental results more than those calculated by Hydroflut code. (author)

  18. Uncertainty propagation applied to multi-scale thermal-hydraulics coupled codes. A step towards validation

    Energy Technology Data Exchange (ETDEWEB)

    Geffray, Clotaire Clement

    2017-03-20

    The work presented here constitutes an important step towards the validation of the use of coupled system thermal-hydraulics and computational fluid dynamics codes for the simulation of complex flows in liquid metal cooled pool-type facilities. First, a set of methods suited for uncertainty and sensitivity analysis and validation activities with regards to the specific constraints of the work with coupled and expensive-to-run codes is proposed. Then, these methods are applied to the ATHLET - ANSYS CFX model of the TALL-3D facility. Several transients performed at this latter facility are investigated. The results are presented, discussed and compared to the experimental data. Finally, assessments of the validity of the selected methods and of the quality of the model are offered.

  19. Development of a thermal-hydraulic analysis code for annular fuel assemblies

    Energy Technology Data Exchange (ETDEWEB)

    Vishnoi, A.K.; Chandraker, D.K.; Vijayan, P.K. [Bhabha Atomic Reseach Centre (BARC), Mumbai, Maharashtra (India)

    2012-03-15

    In this work a detailed study of the annular fuel has been carried out. A thermal hydraulics code, ANUFAN (Annular Fuel Analysis), based on the bundle average method, capable of modeling both internally and externally cooled annular fuel pins is developed. Code predictions have been compared with calculations from Korea Atomic Energy Research Institute (KAERI) and MIT. Heat transfer fraction difference between ANUFAN and RELAP was found about 1.7%. Analysis of a 54 - fuel rod assembly is carried out with 36 and 45 numbers of annular fuel pins keeping the same channel size and bundle power as of the solid fuel assembly. Fuel pin maximum temperature of the annular fuel is found much less than the solid fuel. MCHFR value for annular fuel is found much higher compared to that of the solid fuel of 54 - fuel rod assembly. The full paper covers the details of the computer code, the analysis carried out and the results obtained. (orig.)

  20. Development of a best estimate auditing code for CANDU thermal hydraulic safety analysis

    Energy Technology Data Exchange (ETDEWEB)

    Chung, B.D.; Lee, W.J.; Lim, H.S. [Korea Atomic Energy Research Institute, Taejeon (Korea)

    2000-03-01

    The main purpose of this study is to develop a thermal hydraulic auditing code for the CANDU reactor, modifying the model of existing PWR auditing tool , i.e. RELAP5/MOD3. This scope of project is a third step of the whole project, and expand the RELAP5/MOD3/CANDU version for implementation of LOCA Analysis. There are three main area of model development, i.e. Moody critical flow model, flow regime model of horizontal CANDU bundle, and fuel element heatup model when the stratification occurs. Newly developed version, namely RELAP5/MOD3/CANDU+ is applicable to CANDU plant analysis with keeping the function of light water reactor analysis. The limited validations of model installation were performed. Assessment of CHF model using AECL separated effect test and calculation for Wolsong 2 plant were performed also for the applicability test of the developed version. 15 refs., 37 figs., 8 tabs. (Author)

  1. Advances in thermal-hydraulic studies of a transmutation advanced device for sustainable energy applications

    Energy Technology Data Exchange (ETDEWEB)

    Fajardo, Laura Garcia, E-mail: laura.gf@cern.ch [European Organization for Nuclear Research (CERN), Geneva (Switzerland). Technology Department; Hernandez, Carlos Garcia; Mazaira, Leorlen Rojas, E-mail: cgh@instec.cu, E-mail: irojas@instec.cu [Higher Institute of Technologies and Applied Sciences (INSTEC), Habana (Cuba); Castells, Facundo Alberto Escriva, E-mail: aescriva@iqn.upv.es [University of Valencia (UV), Valencia (Spain). Energetic Engineering Institute; Lira, Carlos Brayner de Olivera, E-mail: cabol@ufpe.br [Universidade Federal de Pernambuco (UFPE), Recife, PE (BRazil). Dept. de Engenharia Nuclear

    2013-07-01

    The Transmutation Advanced Device for Sustainable Energy Applications (TADSEA) is a pebble-bed Accelerator Driven System (ADS) with a graphite-gas configuration, designed for nuclear waste trans- mutation and for obtaining heat at very high temperatures to produce hydrogen. In previous work, the TADSEA's nuclear core was considered as a porous medium performed with a CFD code and thermal-hydraulic studies of the nuclear core were presented. In this paper, the heat transfer from the fuel to the coolant was analyzed for three core states during normal operation. The heat transfer inside the spherical fuel elements was also studied. Three critical fuel elements groups were defined regarding their position inside the core. Results were compared with a realistic CFD model of the critical fuel elements groups. During the steady state, no critical elements reached the limit temperature of this type of fuel. (author)

  2. Burn-up dependent steady-state thermal hydraulic analysis of Pakistan research reactor-1

    Directory of Open Access Journals (Sweden)

    Muhammad Atta

    2011-01-01

    Full Text Available The burn-up dependent steady-state thermal hydraulic analysis of Pakistan research reactor-1, reference operating core, has been carried out utilizing standard computer codes WIMS/D4, CITATION, and RELAP5/MOD3.4. Reactor codes WIMS/D4 and CITATION have been used for the calculations of neutronic parameters including peaking factors and power profiles at different burn-up considering a xenon free core and also the equilibrium xenon values. RELAP5/MOD3.4 code was utilized for the determination of peak fuel centerline, clad and coolant temperatures to ensure the safety of the reactor throughout the cycle. The calculations reveal that the reactor is safe and no nucleate boiling will commence at any part of the core throughout the cycle and that the safety margin increases with burnup as peaking factors decrease.

  3. Development of a best estimate auditing code for CANDU thermal hydraulic safety analysis

    Energy Technology Data Exchange (ETDEWEB)

    Chung, B.D.; Lee, W.J.; Lim, H.S. [Korea Atomic Energy Research Institute, Taejon (Korea)

    1999-04-01

    The main purpose of this study is to develop a thermal hydraulic auditing code for the CANDU reactor, modifying the model of existing PWR auditing tool, i.e. RELAP5/MOD3. This scope of project is a second step of the whole project, and focus to the implementation of CANDU models based on the previous study. FORTRAN 90 language have been used for the development of RELAP5.MOD3/CANDU PC version. For the convenience of the previous Workstation users, the FOTRAN 77 version has been coded also and implanted into the original RELAP5 source file. The verification of model implementation has been performed through the simple verification calculations using the CANDU version. 6 refs., 15 figs., 7 tabs. (Author)

  4. Thermal-hydraulic processes involved in loss of residual heat removal during reduced inventory operation

    Energy Technology Data Exchange (ETDEWEB)

    Fletcher, C.D.; McHugh, P.R.; Naff, S.A.; Johnsen, G.W.

    1991-02-01

    This paper identifies the topics needed to understand pressurized water reactor response to an extended loss of residual heat removal event during refueling and maintenance outages. By identifying the possible plant conditions and cooling methods that would be used for each cooling mode, the controlling thermal-hydraulic processes and phenomena were identified. Controlling processes and phenomena include: gravity drain, core water boil-off, and reflux cooling processes. Important subcategories of the reflux cooling processes include: the initiation of reflux cooling from various plant conditions, the effects of air on reflux cooling, core level depression effects, issues regarding the steam generator secondaries, and the special case of boiler-condenser cooling with once-through steam generators. 25 refs., 6 figs., 1 tab.

  5. Thermal hydraulic codes for LWR safety analysis - present status and future perspective

    Energy Technology Data Exchange (ETDEWEB)

    Staedtke, H. [Commission of the European Union, Ispra (Italy)

    1997-07-01

    The aim of the present paper is to give a review on the current status and future perspective of present best-estimate Thermal Hydraulic codes. Reference is made to internationally well-established codes which have reached a certain state of maturity. The first part of the paper deals with the common basic code features with respect to the physical modelling and their numerical methods used to describe complex two-phase flow and heat transfer processes. The general predictive capabilities are summarized identifying some remaining code deficiencies and their underlying limitations. The second part discusses various areas including physical modelling, numerical techniques and informatic structure where the codes could be substantially improved.

  6. Thermal-hydraulic and structural safety analysis of SLSF P3 experiment

    Energy Technology Data Exchange (ETDEWEB)

    Ragland, W A; Ariman, T; Tessier, J H

    1979-01-01

    The Sodium Loop Safety Facility (SLSF) P3 experiment was the fourth in a series of in-reactor tests in the Engineering Test Reactor and simulated an unprotected flow coastdown in a 37 pin bundle. A comprehensive thermal-hydraulic analysis of the SLSF loop was coupled with a structural analysis of the test section hexcan outer duct in order to provide assurance, before the transient, that the loop would safely contain the P3 experiment. This analysis was performed for both the expected transient and for an off-normal condition of failure of redundant logic circuits to provide the proper pump power demand signal. Results of the analysis showed the hexcan outer duct to safely contain the P3 experiment for both conditions. The analysis for the expected transient has been verified by the successful completion of the P3 experiment.

  7. 2D Thermal Hydraulic Analysis and Benchmark in Support of HFIR LEU Conversion using COMSOL

    Energy Technology Data Exchange (ETDEWEB)

    Freels, James D [ORNL; Bodey, Isaac T [ORNL; Lowe, Kirk T [ORNL; Arimilli, Rao V [ORNL

    2010-09-01

    The research documented herein was funded by a research contract between the Research Reactors Division (RRD) of Oak Ridge National Laboratory (ORNL) and the University of Tennessee, Knoxville (UTK) Mechanical, Aerospace and Biomedical Engineering Department (MABE). The research was governed by a statement of work (SOW) which clearly defines nine specific tasks. This report is outlined to follow and document the results of each of these nine specific tasks. The primary goal of this phase of the research is to demonstrate, through verification and validation methods, that COMSOL is a viable simulation tool for thermal-hydraulic modeling of the High Flux Isotope Reactor (HFIR) core. A secondary goal of this two-dimensional phase of the research is to establish methodology and data base libraries that are also needed in the full three-dimensional COMSOL simulation to follow. COMSOL version 3.5a was used for all of the models presented throughout this report.

  8. ATHENA (Advanced Thermal-Hydraulic Energy Network Analyzer) transient analysis of a fusion engineering test reactor

    Energy Technology Data Exchange (ETDEWEB)

    Wareing, T.A.

    1988-02-01

    Two potential undercooling transients are of concern in the design of TIBER-II (Tokamak Ignition/Burn Experimental Reactor), namely loss of coolant and loss of flow accidents. The major area of concern for TIBER-II is the inboard shield, where, due to tungsten material, the decay heat is extremely high. The purpose of this study was to analyze these transients using the thermal-hydraulic code ATHENA (Advanced Thermal-Hydraulic Energy Network Analyzer). The most comprehensive portion of this project involved creating a simple, yet complete, ATHENA model representative of TIBER-II. The completed model represents the case when the plasma is off and contains the inboard shield, the outboard shield, the divertor shields, and the primary loop. The primary loop contains the piping, pump, pressurizer, and heat exchanger. The heat exchanger is at the same elevation as the reactor, the least favorable to establishing natural circulation. The only transient analyzed so far, however, is a loss of flow accident. Results from the loss of flow analysis show that there is sufficient natural circulation in the inboard, outboard, and lower divertor shield to remove the decay heat, assuming that the secondary side flow is at full capacity. Although the upper divertor shield does not have sufficient natural circulation, cooling is provided due to vaporization and re-flood oscillations. However, one must recognize that there may be some local hot sport where the flow geometry inhibits cooling in a LOFA; the ATHENA model would not detect any localized problem. 9 refs., 18 figs., 4 tabs.

  9. Interface requirements to couple thermal-hydraulic codes to severe accident codes: ATHLET-CD

    Energy Technology Data Exchange (ETDEWEB)

    Trambauer, K. [GRS, Garching (Germany)

    1997-07-01

    The system code ATHLET-CD is being developed by GRS in cooperation with IKE and IPSN. Its field of application comprises the whole spectrum of leaks and large breaks, as well as operational and abnormal transients for LWRs and VVERs. At present the analyses cover the in-vessel thermal-hydraulics, the early phases of core degradation, as well as fission products and aerosol release from the core and their transport in the Reactor Coolant System. The aim of the code development is to extend the simulation of core degradation up to failure of the reactor pressure vessel and to cover all physically reasonable accident sequences for western and eastern LWRs including RMBKs. The ATHLET-CD structure is highly modular in order to include a manifold spectrum of models and to offer an optimum basis for further development. The code consists of four general modules to describe the reactor coolant system thermal-hydraulics, the core degradation, the fission product core release, and fission product and aerosol transport. Each general module consists of some basic modules which correspond to the process to be simulated or to its specific purpose. Besides the code structure based on the physical modelling, the code follows four strictly separated steps during the course of a calculation: (1) input of structure, geometrical data, initial and boundary condition, (2) initialization of derived quantities, (3) steady state calculation or input of restart data, and (4) transient calculation. In this paper, the transient solution method is briefly presented and the coupling methods are discussed. Three aspects have to be considered for the coupling of different modules in one code system. First is the conservation of masses and energy in the different subsystems as there are fluid, structures, and fission products and aerosols. Second is the convergence of the numerical solution and stability of the calculation. The third aspect is related to the code performance, and running time.

  10. Experiments and analytical studies related to blowdown and containment thermal hydraulics on CSF

    Energy Technology Data Exchange (ETDEWEB)

    Dutta, Anu, E-mail: adutta@barc.gov.in; Thangamani, I.; Shanware, V.M.; Rao, K.S.; Gera, B.; Ravi Kiran, A.; Goyal, P.; Verma, Vishnu; Sharma, P.K.; Agrawal, M.K.; Ganju, S.; Singh, R.K.

    2015-12-01

    Highlights: • Blowdown and containment thermal hydraulics experiments conducted in CSF. • RELAP5, ASTEC and CONTRAN codes used for analysis. • Containment peak pressure and temp predicted close to experimental values. • CONTRAN and ASTEC codes predict early containment depressurization. • Numerical procedure, benchmarked for loss of coolant accident in nuclear reactors. - Abstract: Containment Studies Facility (CSF) is volumetrically scaled down model of Indian Pressurized Heavy Water Reactor (IPHWR) containment for simulating LOCA/MSLB conditions which consists of concrete containment model (CM) and Primary Heat Transport Model (PHTM) vessel. Blowdown experiments at different initial vessel pressure conditions were recently conducted at CSF and the vessel and containment parameters such as pressure, temperature and level transients have been recorded during the experiments. The experimental results have been used for benchmarking of numerical procedure adopted for evaluating LOCA/MSLB conditions in nuclear containment. The numerical procedure involves simulation of blowdown phenomena using RELAP5 code for evaluating mass and energy discharge rates, which are then used for calculating containment pressure–temperature transients using ASTEC and in-house CONTRAN codes. Predictions of major parameters of vessel and containment model were found to be in good agreement with that of experimental data. In containment thermal hydraulic calculations, condensation heat transfer coefficient affects the containment pressure–temperature transients. Various empirical condensation models like Tagami, Uchida and Diffusion models have been incorporated in CONTRAN code and suitable condensation model has been identified for which predicted pressure values are close to the experimental one. The details of the experimental and analytical studies conducted are presented in this paper.

  11. Current and anticipated use of thermal-hydraulic codes for BWR transient and accident analyses in Japan

    Energy Technology Data Exchange (ETDEWEB)

    Arai, Kenji; Ebata, Shigeo [Toshiba Corp., Yokohama (Japan)

    1997-07-01

    This paper summarizes the current and anticipated use of the thermal-hydraulic and neutronic codes for the BWR transient and accident analyses in Japan. The codes may be categorized into the licensing codes and the best estimate codes for the BWR transient and accident analyses. Most of the licensing codes have been originally developed by General Electric. Some codes have been updated based on the technical knowledge obtained in the thermal hydraulic study in Japan, and according to the BWR design changes. The best estimates codes have been used to support the licensing calculations and to obtain the phenomenological understanding of the thermal hydraulic phenomena during a BWR transient or accident. The best estimate codes can be also applied to a design study for a next generation BWR to which the current licensing model may not be directly applied. In order to rationalize the margin included in the current BWR design and develop a next generation reactor with appropriate design margin, it will be required to improve the accuracy of the thermal-hydraulic and neutronic model. In addition, regarding the current best estimate codes, the improvement in the user interface and the numerics will be needed.

  12. Mechanical and hydraulic behavior of a rock fracture under shear deformation

    Science.gov (United States)

    Nishiyama, Satoshi; Ohnishi, Yuzo; Ito, Hisao; Yano, Takao

    2014-12-01

    With regard to crystalline rock that constitutes deep geology, attempts have been made to explore its hydraulic characteristics by focusing on the network of numerous fractures within. As the hydraulic characteristics of a rock are the accumulation of hydraulic characteristics of each fracture, it is necessary to develop the hydraulic model of a single fracture to predict the large-scale hydraulic behavior. To this end, a simultaneous permeability and shear test device is developed, and shear-flow coupling tests are conducted on specimens having fractures with varied levels of surface roughness in the constant normal stiffness conditions. The results show that the permeability characteristics in the relation between shear displacement and transmissivity change greatly at the point where the stress path reaches the Mohr-Coulomb failure curve. It is also found that there exists a range in which transmissivity is not proportional to the cube of mechanical aperture width, which seems to be because of the occurrence of channeling phenomenon at small mechanical aperture widths. This channeling flow disappears with increasing shear and is transformed into a uniform flow. We develop a simulation technique to evaluate the macroscopic permeability characteristics by the lattice gas cellular automaton method, considering the microstructure of fracture, namely the fracture surface roughness. With this technique, it is shown that the formation of the Hagen-Poiseuille flow is affected by the fracture microstructure under shear, which as a result determines the relationship between the mechanical aperture width and transmissivity.

  13. Thermal-Hydraulic Results for the Boiling Water Reactor Dry Cask Simulator.

    Energy Technology Data Exchange (ETDEWEB)

    Durbin, Samuel [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Lindgren, Eric R. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)

    2017-09-01

    The thermal performance of commercial nuclear spent fuel dry storage casks is evaluated through detailed numerical analysis. These modeling efforts are completed by the vendor to demonstrate performance and regulatory compliance. The calculations are then independently verified by the Nuclear Regulatory Commission (NRC). Carefully measured data sets generated from testing of full sized casks or smaller cask analogs are widely recognized as vital for validating these models. Recent advances in dry storage cask designs have significantly increased the maximum thermal load allowed in a cask in part by increasing the efficiency of internal conduction pathways and by increasing the internal convection through greater canister helium pressure. These same canistered cask systems rely on ventilation between the canister and the overpack to convect heat away from the canister to the environment for both aboveground and belowground configurations. While several testing programs have been previously conducted, these earlier validation attempts did not capture the effects of elevated helium pressures or accurately portray the external convection of aboveground and belowground canistered dry cask systems. The purpose of this investigation was to produce validation-quality data that can be used to test the validity of the modeling presently used to determine cladding temperatures in modern vertical dry casks. These cladding temperatures are critical to evaluate cladding integrity throughout the storage cycle. To produce these data sets under well-controlled boundary conditions, the dry cask simulator (DCS) was built to study the thermal-hydraulic response of fuel under a variety of heat loads, internal vessel pressures, and external configurations. An existing electrically heated but otherwise prototypic BWR Incoloy-clad test assembly was deployed inside of a representative storage basket and cylindrical pressure vessel that represents a vertical canister system. The symmetric

  14. The relevance of thermal hydraulics pipeline simulation as a regulatory support tool

    Energy Technology Data Exchange (ETDEWEB)

    Silva, Patricia Mannarino; Santos, Almir Beserra dos [Agencia Nacional do Petroleo, Gas Natural e Biocombustiveis (ANP), Rio de Janeiro, RJ (Brazil)

    2009-07-01

    The capacity definition of a pipeline, along with its allocation, is very relevant to assure market transparency, nondiscriminatory access, security of supply, and also to give consistent signs for expansion needs. Nevertheless, the capacity definition is a controversial issue, and may widely vary depending on the technical and commercial assumptions made. To calculate a pipeline's nominal capacity, there are a variety of simulation tools, which include steady state, transient and on-line computer programs. It is desirable that the simulation tool is robust enough to predict the pipeline's capacity under different conditions. There are many variables that impact the flow through a pipeline, like gas characteristics, pipe and environmental variables. Designing a thermal model is a time-consuming task that requests understanding the level of detail need, in order to achieve success in its application. This article discusses the capacity definition, its role and calculation guidelines, describes ANP's experience with capacity calculation and further challenges according to the new regulation, and debates the role of thermal hydraulic simulation as a regulatory tool. (author)

  15. Thermal Hydraulic Performance in a Solar Air Heater Channel with Multi V-Type Perforated Baffles

    Directory of Open Access Journals (Sweden)

    Anil Kumar

    2016-07-01

    Full Text Available This article presents heat transfer and fluid flow characteristics in a solar air heater (SAH channel with multi V-type perforated baffles. The flow passage has an aspect ratio of 10. The relative baffle height, relative pitch, relative baffle hole position, flow attack angle, and baffle open area ratio are 0.6, 8.0, 0.42, 60°, and 12%, respectively. The Reynolds numbers considered in the study was in the range of 3000–10,000. The re-normalization group (RNG k-ε turbulence model has been used for numerical analysis, and the optimum relative baffle width has been investigated considering relative baffle widths of 1.0–7.0.The numerical results are in good agreement with the experimental data for the range considered in the study. Multi V-type perforated baffles are shown to have better thermal performance as compared to other baffle shapes in a rectangular passage. The overall thermal hydraulic performance shows the maximum value at the relative baffle width of 5.0.

  16. A review on improving thermal-hydraulic performance of fin-and-tube heat exchangers

    Science.gov (United States)

    Nickolas, N.; Moorthy, P.; Oumer, A. N.; Ishak, M.

    2017-10-01

    Fin-and-tube heat exchangers are one of the most common type of heat exchangers that are normally used in sectors that require small size and light weight but high heat transfer capabilities. Compact fin-and-tube heat exchangers experiences high convective thermal resistance at the air-side due to the thermo-physical properties of air. Thus, the purpose of this paper is to provide an overview of research works that are relevant to improving thermalhydraulic performance at the air-side of fin-and-tube heat exchangers. This paper covers a variety of parameters such as tube parameters like tube arrangement, tube shapes and tube inclination angles; extended surfaces such as different shapes of fins and different parameters of vortex generators like attack angles, shapes and locations. Overall, for most modifications there was increment in heat transfer but accompanied with a pressure drop penalty. However, this varies for different combinations of parameters thus this review is to help understand how every mentioned parameters influences the thermal-hydraulic performance.

  17. Data report of BWR post-CHF tests. Transient core thermal-hydraulic test program. Contract research

    Energy Technology Data Exchange (ETDEWEB)

    Iguchi, Tadashi; Itoh, Hideo; Kiuchi, Toshio; Watanabe, Hironori; Kimura, Mamoru; Anoda, Yoshinari [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    2001-03-01

    JAERI has been performing transient core thermal-hydraulic test program. In the program, authors performed BWR/ABWR DBE simulation tests with a test facility, which can simulate BWR/ABWR transients. The test facility has a 4 x 4 bundle core simulator with 15-rod heaters and one non-heated rod. Through the tests, authors quantified the thermal safety margin for core cooling. In order to quantify the thermal safety margin, authors collected experimental data on post-CHF. The data are essential for the evaluation of clad temperature transient when core heat-up occurs during DBEs. In comparison with previous post-CHF tests, present experiments were performed in much wider experimental condition, covering high clad temperature, low to high pressure and low to high mass flux. Further, data at wider elevation (lower to higher elevation of core) were obtained in the present experiments, which make possible to discuss the effect of axial position on thermal-hydraulics, while previous works usually discuss the thermal-hydraulics at the position where the first heat-up occurs. This data report describes test procedure, test condition and major experimental data of post-CHF tests. (author)

  18. Analysis of the Thermal and Hydraulic Stimulation Program at Raft River, Idaho

    Science.gov (United States)

    Bradford, Jacob; McLennan, John; Moore, Joseph; Podgorney, Robert; Plummer, Mitchell; Nash, Greg

    2017-05-01

    The Raft River geothermal field, located in southern Idaho, roughly 100 miles northwest of Salt Lake City, is the site of a Department of Energy Enhanced Geothermal System project designed to develop new techniques for enhancing the permeability of geothermal wells. RRG-9 ST1, the target stimulation well, was drilled to a measured depth of 5962 ft. and cased to 5551 ft. The open-hole section of the well penetrates Precambrian quartzite and quartz monzonite. The well encountered a temperature of 282 °F at its base. Thermal and hydraulic stimulation was initiated in June 2013. Several injection strategies have been employed. These strategies have included the continuous injection of water at temperatures ranging from 53 to 115 °F at wellhead pressures of approximately 275 psi and three short-term hydraulic stimulations at pressures up to approximately 1150 psi. Flow rates, wellhead and line pressures and fluid temperatures are measured continuously. These data are being utilized to assess the effectiveness of the stimulation program. As of August 2014, nearly 90 million gallons have been injected. A modified Hall plot has been used to characterize the relationships between the bottom-hole flowing pressure and the cumulative injection fluid volume. The data indicate that the skin factor is decreased, and/or the permeability around the wellbore has increased since the stimulation program was initiated. The injectivity index also indicates a positive improvement with values ranging from 0.15 gal/min psi in July 2013 to 1.73 gal/min psi in February 2015. Absolute flow rates have increased from approximately 20 to 475 gpm by February 2 2015. Geologic, downhole temperature and seismic data suggest the injected fluid enters a fracture zone at 5650 ft and then travels upward to a permeable horizon at the contact between the Precambrian rocks and the overlying Tertiary sedimentary and volcanic deposits. The reservoir simulation program FALCON developed at the Idaho National

  19. Preliminary fluid channel design and thermal-hydraulic analysis of glow discharge cleaning permanent electrode

    Energy Technology Data Exchange (ETDEWEB)

    Cai, Lijun, E-mail: cailj@swip.ac.cn [Southwestern Institute of Physics, Chengdu (China); Lin, Tao; Wang, Yingqiao; Wang, Mingxu [Southwestern Institute of Physics, Chengdu (China); Maruyama, So; Yang, Yu; Kiss, Gabor [ITER Organization, Route de Vinon-sur-Verdon, CS 90 046, 13067 St. Paul Lez Durance Cedex (France)

    2016-11-01

    Highlights: • The plasma facing closure cap has to survive after 30,000 thermal heat load cycles. • 0.35 MW/m2 radiation heat load plus nuclear heat load are very challenging for stainless steel. • Multilayer structure has been designed by using advanced welding and drilling technology to solve the neutron heating problem. • Accurate volumetric load application in analysis model by CFX has been mastered. - Abstract: Glow discharge cleaning (GDC) shall be used on ITER device to reduce and control impurity and hydrogenic fuel out-gassing from in-vessel plasma facing components. After first plasma, permanent electrode (PE) will be used to replace Temporary Electrode (TE) for subsequent operation. Two fundamental scenarios i.e., GDC and Plasma Operation State (POS) should be considered for electrode design, which requires the heat load caused by plasma radiation and neutron heating must be taken away by cooling water flowing inside the electrode. In this paper, multilayer cooling channels inside PE are preliminarily designed, and snakelike route in each layer is adopted to improve the heat exchange. Detailed thermal-hydraulic analyses have been done to validate the design feasibility or rationality. The analysis results show that during GDC the cooling water inlet and outlet temperature difference is far less than the allowable temperature rise under water flow rate 0.15 kg/s compromised by many factors. For POS, the temperature rise and pressure drop are within the design goals, but high thermal stress occurs on the front surface of closure cap of electrode. After several iterations of optimization of the closure cap, the equivalent strain range after 30,000 loading cycles for POS is well below 0.3% design goals.

  20. Methods and Models for the Coupled Neutronics and Thermal-Hydraulics Analysis of the CROCUS Reactor at EFPL

    Directory of Open Access Journals (Sweden)

    A. Rais

    2015-01-01

    Full Text Available In order to analyze the steady state and transient behavior of the CROCUS reactor, several methods and models need to be developed in the areas of reactor physics, thermal-hydraulics, and multiphysics coupling. The long-term objectives of this project are to work towards the development of a modern method for the safety analysis of research reactors and to update the Final Safety Analysis Report of the CROCUS reactor. A first part of the paper deals with generation of a core simulator nuclear data library for the CROCUS reactor using the Serpent 2 Monte Carlo code and also with reactor core modeling using the PARCS code. PARCS eigenvalue, radial power distribution, and control rod reactivity worth results were benchmarked against Serpent 2 full-core model results. Using the Serpent 2 model as reference, PARCS eigenvalue predictions were within 240 pcm, radial power was within 3% in the central region of the core, and control rod reactivity worth was within 2%. A second part reviews the current methodology used for the safety analysis of the CROCUS reactor and presents the envisioned approach for the multiphysics modeling of the reactor.

  1. Integral and Separate Effects Tests for Thermal Hydraulics Code Validation for Liquid-Salt Cooled Nuclear Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Peterson, Per

    2012-10-30

    The objective of the 3-year project was to collect integral effects test (IET) data to validate the RELAP5-3D code and other thermal hydraulics codes for use in predicting the transient thermal hydraulics response of liquid salt cooled reactor systems, including integral transient response for forced and natural circulation operation. The reference system for the project is a modular, 900-MWth Pebble Bed Advanced High Temperature Reactor (PB-AHTR), a specific type of Fluoride salt-cooled High temperature Reactor (FHR). Two experimental facilities were developed for thermal-hydraulic integral effects tests (IETs) and separate effects tests (SETs). The facilities use simulant fluids for the liquid fluoride salts, with very little distortion to the heat transfer and fluid dynamics behavior. The CIET Test Bay facility was designed, built, and operated. IET data for steady state and transient natural circulation was collected. SET data for convective heat transfer in pebble beds and straight channel geometries was collected. The facility continues to be operational and will be used for future experiments, and for component development. The CIET 2 facility is larger in scope, and its construction and operation has a longer timeline than the duration of this grant. The design for the CIET 2 facility has drawn heavily on the experience and data collected on the CIET Test Bay, and it was completed in parallel with operation of the CIET Test Bay. CIET 2 will demonstrate start-up and shut-down transients and control logic, in addition to LOFC and LOHS transients, and buoyant shut down rod operation during transients. Design of the CIET 2 Facility is complete, and engineering drawings have been submitted to an external vendor for outsourced quality controlled construction. CIET 2 construction and operation continue under another NEUP grant. IET data from both CIET facilities is to be used for validation of system codes used for FHR modeling, such as RELAP5-3D. A set of

  2. Validation and Calibration of Nuclear Thermal Hydraulics Multiscale Multiphysics Models - Subcooled Flow Boiling Study

    Energy Technology Data Exchange (ETDEWEB)

    Anh Bui; Nam Dinh; Brian Williams

    2013-09-01

    In addition to validation data plan, development of advanced techniques for calibration and validation of complex multiscale, multiphysics nuclear reactor simulation codes are a main objective of the CASL VUQ plan. Advanced modeling of LWR systems normally involves a range of physico-chemical models describing multiple interacting phenomena, such as thermal hydraulics, reactor physics, coolant chemistry, etc., which occur over a wide range of spatial and temporal scales. To a large extent, the accuracy of (and uncertainty in) overall model predictions is determined by the correctness of various sub-models, which are not conservation-laws based, but empirically derived from measurement data. Such sub-models normally require extensive calibration before the models can be applied to analysis of real reactor problems. This work demonstrates a case study of calibration of a common model of subcooled flow boiling, which is an important multiscale, multiphysics phenomenon in LWR thermal hydraulics. The calibration process is based on a new strategy of model-data integration, in which, all sub-models are simultaneously analyzed and calibrated using multiple sets of data of different types. Specifically, both data on large-scale distributions of void fraction and fluid temperature and data on small-scale physics of wall evaporation were simultaneously used in this work’s calibration. In a departure from traditional (or common-sense) practice of tuning/calibrating complex models, a modern calibration technique based on statistical modeling and Bayesian inference was employed, which allowed simultaneous calibration of multiple sub-models (and related parameters) using different datasets. Quality of data (relevancy, scalability, and uncertainty) could be taken into consideration in the calibration process. This work presents a step forward in the development and realization of the “CIPS Validation Data Plan” at the Consortium for Advanced Simulation of LWRs to enable

  3. Validation of coupled neutronic / thermal-hydraulic codes for VVER reactors. Final report

    Energy Technology Data Exchange (ETDEWEB)

    Mittag, S.; Grundmann, U.; Kliem, S.; Kozmenkov, Y.; Rindelhardt, U.; Rohde, U.; Weiss, F.-P.; Langenbuch, S.; Krzykacz-Hausmann, B.; Schmidt, K.-D.; Vanttola, T.; Haemaelaeinen, A.; Kaloinen, E.; Kereszturi, A.; Hegyi, G.; Panka, I.; Hadek, J.; Strmensky, C.; Darilek, P.; Petkov, P.; Stefanova, S.; Kuchin, A.; Khalimonchuk, V.; Hlbocky, P.; Sico, D.; Danilin, S.; Ionov, V.; Nikonov, S.; Powney, D.

    2004-08-01

    In recent years, the simulation methods for the safety analysis of nuclear power plants have been continuously improved to perform realistic calculations. Therefore in VALCO work package 2 (WP 2), the usual application of coupled neutron-kinetic / thermal-hydraulic codes to VVER has been supplemented by systematic uncertainty and sensitivity analyses. A comprehensive uncertainty analysis has been carried out. The GRS uncertainty and sensitivity method based on the statistical code package SUSA was applied to the two transients studied earlier in SRR-1/95: A load drop of one turbo-generator in Loviisa-1 (VVER-440), and a switch-off of one feed water pump in Balakovo-4 (VVER-1000). The main steps of these analyses and the results obtained by applying different coupled code systems (SMABRE - HEXTRAN, ATHLET - DYN3D, ATHLET - KIKO3D, ATHLET - BIPR-8) are described in this report. The application of this method is only based on variations of input parameter values. No internal code adjustments are needed. An essential result of the analysis using the GRS SUSA methodology is the identification of the input parameters, such as the secondary-circuit pressure, the control-assembly position (as a function of time), and the control-assembly efficiency, that most sensitively affect safety-relevant output parameters, like reactor power, coolant heat-up, and primary pressure. Uncertainty bands for these output parameters have been derived. The variation of potentially uncertain input parameter values as a consequence of uncertain knowledge can activate system actions causing quite different transient evolutions. This gives indications about possible plant conditions that might be reached from the initiating event assuming only small disturbances. In this way, the uncertainty and sensitivity analysis reveals the spectrum of possible transient evolutions. Deviations of SRR-1/95 coupled code calculations from measurements also led to the objective to separate neutron kinetics from

  4. Development of nuclear transmutation technology - A study on the thermal-hydraulic characteristics of Pb-Bi coolant material

    Energy Technology Data Exchange (ETDEWEB)

    Chung, Chang Hyun; You, Young Woo; Cho, Jae Seon; Kim, Ju Youl; Yang, Hui Chang; Huh, Byung Gil [Seoul National University, Seoul (Korea)

    2000-03-01

    The objective of this study is to provide the direction of HYPER design in terms of thermal hydraulics especially through the analysis of thermal hydraulic characteristics of lead-bismuth material as a HYPER coolant and of proton accelerator target system. In this study, in order to evaluate the thermal-hydraulic characteristics of HYPER system, the FLUENT calculation is performed with liquid metal lead-bismuth(43%) and the turbulent Prandtl number model is developed. Also, the heat transfer analyses including temperature rising are performed for accelerator beam window, solid tungsten target and liquid target which is composed of liquid lead and lead-bismuth, respectively and the thermal stress analyses are performed for accelerator beam window. Through this study, the BASECASE whose parameter is HYPER system design specification is calculated by FLUENT. It is shown that the coolant velocity must exceeds 1.6 m/s for supporting the core coolant temperature in operating temperature range. The suggested turbulent Prandtl number model is applicable to liquid metal. And in order to maintain the integrity of proton beam target system, it is necessary to investigate the target structure associated with smoothing the flow path and beam window cooling. 43 refs., 67 figs., 27 tabs. (Author)

  5. Neutronics and thermal hydraulic analysis of TRIGA Mark II reactor using MCNPX and COOLOD-N2 computer code

    Science.gov (United States)

    Tiyapun, K.; Wetchagarun, S.

    2017-06-01

    The neutronic analysis of TRIGA Mark II reactor has been performed. A detailed model of the reactor core was conducted including standard fuel elements, fuel follower control rods, and irradiation devices. As the approach to safety nuclear design are based on determining the criticality (keff), reactivity worth, reactivity excess, hot rod power factor and power peaking of the reactor, the MCNPX code had been used to calculate the nuclear parameters for different core configuration designs. The thermal-hydraulic model has been developed using COOLOD-N2 for steady state, using the nuclear parameters and power distribution results from MCNPX calculation. The objective of the thermal-hydraulic model is to determine the thermal safety margin and to ensure that the fuel integrity is maintained during steady state as well as during abnormal condition at full power. The hot channel fuel centerline temperature, fuel surface temperature, cladding surface temperature, the departure from nucleate boiling (DNB) and DNB ratio were determined. The good agreement between experimental data and simulation concerning reactor criticality proves the reliability of the methodology of analysis from neutronic and thermal hydraulic perspective.

  6. Experimental research in neutron physic and thermal-hydraulic at the CDTN Triga reactor

    Energy Technology Data Exchange (ETDEWEB)

    Mesquita, Amir Z.; Souza, Rose Mary G.P.; Ferreira, Andrea V.; Pinto, Antonio J.; Costa, Antonio C.L.; Rezende, Hugo C., E-mail: amir@cdtn.b, E-mail: souzarm@cdtn.b, E-mail: avf@cdtn.b, E-mail: ajp@cdtn.b, E-mail: aclc@cdtn.b, E-mail: hcr@cdtn.b [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN/CNEN-MG), Belo Horizonte, MG (Brazil)

    2011-07-01

    The IPR-R1 TRIGA (Training, Research, Isotopes production, General Atomics) at Nuclear Technology Development Center (CDTN) is a pool type reactor cooled by natural circulation of light water and an open surface. TRIGA reactors, developed by General Atomics (GA), are the most widely used research reactor in the world and characterized by inherent safety. The IPR-R1 is the only Brazilian nuclear research reactor available and able to perform experiments in which interaction between neutronic and thermal-hydraulic areas occurs. The IPR-R1 has started up on November 11th, 1960. At that time the maximum thermal power was 30 kW. The present forced cooling system was built in the 70th and the power was upgraded to 100 kW. Recently the core configuration and instrumentation was upgraded again to 250 kW at steady state, and is awaiting the license of CNEN to operate definitely at this new power. This paper describes the experimental research project carried out in the IPR-R1 reactor that has as objective evaluate the behaviour of the reactor operational parameters, and mainly to investigate the influence of temperature on the neutronic variables. The research was supported by Research Support Foundation of the State of Minas Gerais (FAPEMIG) and Brazilian Council for Scientific and Technological Development (CNPq). The research project meets the recommendations of the IAEA, for safety, modernization and development of strategic plan for research reactors utilization. This work is in line with the strategic objectives of Brazil, which aims to design and construct the Brazilian Multipurpose research Reactor (RMB). (author)

  7. Study of thermal behavior of phytic acid

    Directory of Open Access Journals (Sweden)

    André Luis Máximo Daneluti

    2013-06-01

    Full Text Available Phytic acid is a natural compound widely used as depigmenting agent in galenic cosmetic emulsions. However, we have observed experimentally that phytic acid, when heated to 150 ºC for around one hour, shows evidence of thermal decomposition. Few studies investigating this substance alone with regard to its stability are available in the literature. This fact prompted the present study to characterize this species and its thermal behavior using thermal analysis (TG/DTG and DSC and to associate the results of these techniques with those obtained by elemental analysis (EA and absorption spectroscopy in the infrared region. The TG/DTG and DSC curves allowed evaluation of the thermal behavior of the sample of phytic acid and enabled use of the non-isothermal thermogravimetric method to study the kinetics of the three main mass-loss events: dehydration I, dehydration II and thermal decomposition. The combination of infrared absorption spectroscopy and elemental analysis techniques allowed evaluation of the intermediate products of the thermal decomposition of phytic acid. The infrared spectra of samples taken during the heating process revealed a reduction in the intensity of the absorption band related to O-H stretching as a result of the dehydration process. Furthermore, elemental analysis results showed an increase in the carbon content and a decrease in the hydrogen content at temperatures of 95, 150, 263 and 380 °C. Visually, darkening of the material was observed at 150 °C, indicating that the thermal decomposition of the material started at this temperature. At a temperature of 380 °C, thermal decomposition progressed, leading to a decrease in carbon and hydrogen. The results of thermogravimetry coupled with those of elemental analysis allow us to conclude that there was agreement between the percentages of phytic acid found in aqueous solution. The kinetic study by the non-isothermal thermogravimetric method showed that the dehydration

  8. Thermal Hydraulic Computational Fluid Dynamics Simulations and Experimental Investigation of Deformed Fuel Assemblies

    Energy Technology Data Exchange (ETDEWEB)

    Mays, Brian [AREVA Federal Services, Lynchburg, VA (United States); Jackson, R. Brian [TerraPower, Bellevue, WA (United States)

    2017-03-08

    The project, Toward a Longer Life Core: Thermal Hydraulic CFD Simulations and Experimental Investigation of Deformed Fuel Assemblies, DOE Project code DE-NE0008321, was a verification and validation project for flow and heat transfer through wire wrapped simulated liquid metal fuel assemblies that included both experiments and computational fluid dynamics simulations of those experiments. This project was a two year collaboration between AREVA, TerraPower, Argonne National Laboratory and Texas A&M University. Experiments were performed by AREVA and Texas A&M University. Numerical simulations of these experiments were performed by TerraPower and Argonne National Lab. Project management was performed by AREVA Federal Services. The first of a kind project resulted in the production of both local point temperature measurements and local flow mixing experiment data paired with numerical simulation benchmarking of the experiments. The project experiments included the largest wire-wrapped pin assembly Mass Index of Refraction (MIR) experiment in the world, the first known wire-wrapped assembly experiment with deformed duct geometries and the largest numerical simulations ever produced for wire-wrapped bundles.

  9. Study of thermal and hydraulic efficiency of supersonic tube of temperature stratification

    Science.gov (United States)

    Tsynaeva, Anna A.; Nikitin, Maxim N.; Tsynaeva, Ekaterina A.

    2017-10-01

    Efficiency of supersonic pipe for temperature stratification with finned subsonic surface of heat transfer is the major of this paper. Thermal and hydraulic analyses of this pipe were conducted to asses effects from installation of longitudinal rectangular and parabolic fins as well as studs of cylindrical, rectangular and parabolic profiles. The analysis was performed based on refined empirical equations of similarity, dedicated to heat transfer of high-speed gas flow with plain wall, and Kármán equation with Nikuradze constants. Results revealed cylindrical studs (with height-to-diameter ratio of 5:1) to be 1.5 times more efficient than rectangular fins of the same height. At the same time rectangular fins (with height-to-thickness ratio of 5:1) were tend to enhance heat transfer rate up to 2.67 times compared to bare walls from subsonic side of the pipe. Longitudinal parabolic fins have minuscule effect on combined efficiency of considered pipe since extra head losses void any gain of heat transfer. Obtained results provide perspective of increasing efficiency of supersonic tube for temperature stratification. This significantly broadens device applicability in thermostatting systems for equipment, cooling systems for energy converting machinery, turbine blades and aerotechnics.

  10. Theoretical investigation of the thermal hydraulic behaviour of a slab-type liquid metal target

    Energy Technology Data Exchange (ETDEWEB)

    Dury, T.V.; Smith, B.L. [Paul Scherrer Institut, Villigen (Switzerland)

    1996-06-01

    The thermal hydraulics codes CFDS-FLOW3D and ASTEC have been used to simulate a slabtype design of ESS spallation target. This design is single-skinned, and of tapering form (in the beam direction), with rounded sides in a cross-section through a plane normal to the beam. The coolant fluid used is mercury, under forced circulation, with an inlet temperature of 180{degrees}C. The goal of these computer studies was to understand the behaviour of the coolant flow, and hence to arrive at a design which optimises the heat extraction for a given beam power - in the sense of: (1) minimising the peak local fluid temperature within the target, (2) maintaining an acceptable temperature level and distribution over and through the target outer wall, (3) keeping the overall fluid pressure loss through the complete target to a minimum, (4) staying within the physical limits of overall size required, particularly in the region of primary spallation. Two- and three-dimensional models have been used, with different arrangements and design of internal baffles, and different coolant flow distributions at the target inlet. Nominal total inlet mass flow was 245 kg/s, and a heat deposition profile used which was based on the proton beam energy distribution. This gave a nominal total heat load of 3.23 MW - of which 8.2kW were deposited in the window steel.

  11. Thermal hydraulic investigations and optimization on the EVC system of a PWR by CFD simulation

    Energy Technology Data Exchange (ETDEWEB)

    Xi, Mengmeng [Department of Nuclear Science and Technology, State Key Laboratory of Multiphase Flow in Power Engineering, Xi’an Jiaotong University, 710049 Xi’an (China); Zhang, Dalin, E-mail: dlzhang@mail.xjtu.edu.cn [Department of Nuclear Science and Technology, State Key Laboratory of Multiphase Flow in Power Engineering, Xi’an Jiaotong University, 710049 Xi’an (China); Tang, Mao [China Nuclear Power Design Engineering Co., Ltd., 518124 Shenzhen (China); Wang, Chenglong; Zheng, Meiyin; Qiu, Suizheng [Department of Nuclear Science and Technology, State Key Laboratory of Multiphase Flow in Power Engineering, Xi’an Jiaotong University, 710049 Xi’an (China)

    2015-08-15

    Highlights: • This study constructs a full CFD model for the EVC system of a PWR. • The complex fluid and solid coupling is treated in the computation. • Primary characteristics of the velocity, pressure and temperature distributions in the EVC system are investigated. • The optimization of the EVC system with different inlet boundaries are performed. - Abstract: In order to optimize the design of Reactor Pit Ventilation (EVC) system in a Pressurized Water Reactor (PWR), it is necessary to study the characteristics of the velocity, pressure and temperature fields in the EVC system. A full computational fluid dynamics (CFD) model for the EVC system is constructed by a commercial CFD code, where the complex fluid and solid coupling is treated. The Shear Stress Transport (SST) model is adopted to perform the turbulence calculation. This paper numerically investigates the characteristics of the velocity, pressure and temperature distributions in the EVC system. In particular, the effects of inlet air parameters on the thermal hydraulic characteristics and the reactor pit structure are also discussed for the EVC system optimization. Simulations are carried out with different mesh sizes and boundary conditions for sensitivity analysis. The computational results are important references to optimize the design and verify the rationality of the EVC system.

  12. Stabilized FE simulation of prototype thermal-hydraulics problems with integrated adjoint-based capabilities

    Science.gov (United States)

    Shadid, J. N.; Smith, T. M.; Cyr, E. C.; Wildey, T. M.; Pawlowski, R. P.

    2016-09-01

    A critical aspect of applying modern computational solution methods to complex multiphysics systems of relevance to nuclear reactor modeling, is the assessment of the predictive capability of specific proposed mathematical models. In this respect the understanding of numerical error, the sensitivity of the solution to parameters associated with input data, boundary condition uncertainty, and mathematical models is critical. Additionally, the ability to evaluate and or approximate the model efficiently, to allow development of a reasonable level of statistical diagnostics of the mathematical model and the physical system, is of central importance. In this study we report on initial efforts to apply integrated adjoint-based computational analysis and automatic differentiation tools to begin to address these issues. The study is carried out in the context of a Reynolds averaged Navier-Stokes approximation to turbulent fluid flow and heat transfer using a particular spatial discretization based on implicit fully-coupled stabilized FE methods. Initial results are presented that show the promise of these computational techniques in the context of nuclear reactor relevant prototype thermal-hydraulics problems.

  13. Normal zone propagation and Thermal Hydraulic Quenchback in a cable-in-conduit superconductor

    Energy Technology Data Exchange (ETDEWEB)

    Lue, J.W.; Dresner, L.

    1993-11-01

    When a local normal zone appears in a cable-in-conduit superconductor, a slug of hot helium is produced. The pressure rises and the hot helium expands. Thus the normal zone propagation in such a conductor can be governed by the hot helium expansion, rather than the heat conduction along the conductor. The expansion of the hot helium compresses the cold helium outside of the normal zone. This raises th@ temperature of the cold helium. When the temperature rise reaches the current sharing limit, the superconductor in contact goes normal. Thus a rapid increase in normal zone propagation occur. This phenomenon is termed Thermal Hydraulic Quenchback (THQ). An experiment was performed to investigate this process. The existence of THQ was verified. Thresholds of THQ were also observed by varying the conductor current, the magnetic field, the temperature, and the initial normal zone length. When THQ occurred, normal zone propagation approaching the velocity of sound was observed. A better picture of THQ is obtained by a careful comparison of the data with analytical studies.

  14. Thermal-hydraulic design of tungsten rod bundles for the APT 3He neutron spallation target

    Science.gov (United States)

    Willcutt, Gordon J. E.

    1995-01-01

    A preconceptual design has been developed for the 3He Target/Blanket System for the Accelerator Production of Tritium Project. The design use tungsten wire-wrapped rods to produce neutrons when the rods are struck by a proton beam. The rods are contained in bundles inside hexagonal Inconel ducts and cooled by D2O. Rod bundles are grouped in patterns in the proton beam inside a chamber filled with 3He that is transmuted to tritium by the neutrons coming from the tungsten rods. Additional 3He is transmuted in a blanket region surrounding the helium chamber. This paper describes the initial thermal-hydraulic design and testing that has been completed to confirm the designed calculations for pressure drop through the bundle and heat transfer in the bundle. Heat transfer tests were run to verify steady-state operation. These tests were followed by increasing power until nucleate boiling occurs to determine operating margins. Changes that improve the initial design are described.

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

    Energy Technology Data Exchange (ETDEWEB)

    Yamaji, Bogdan; Aszodi, Attila [Budapest Univ. of Technology and Economics (Hungary). Inst. of Nuclear Techniques

    2017-09-15

    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.

  16. Thermal-Hydraulic Research Review and Cooperation Outcome for Light Water Reactor Fuel

    Energy Technology Data Exchange (ETDEWEB)

    In, Wang Kee; Shin, Chang Hwan; Lee, Chan; Chun, Tae Hyun; Oh, Dong Seok [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Lee, Chi Young [Pukyong Nat’l Univ., Busan (Korea, Republic of)

    2016-12-15

    The fuel assembly for pressurized water reactor (PWR) consists of fuel rod bundle, spacer grid and bottom/top end fittings. The cooling water in high pressure and temperature is introduced in lower plenum of reactor core and directed to upper plenum through the subchannel which is formed between the fuel rods. The main thermalhydraulic performance parameters for the PWR fuel are pressure drop and critical heat flux in normal operating condition, and quenching time in accident condition. The Korea Atomic Energy Research Institute (KAERI) has been developing an advanced PWR fuel, dual-cooled annular fuel and accident tolerant fuel for the enhancement of fuel performance and the localization. For the key thermal-hydraulic technology development of PWR fuel, the KAERI LWR fuel team has conducted the experiments for pressure drop, turbulent flow mixing and heat transfer, critical heat flux(CHF) and quenching. The computational fluid dynamics (CFD) analysis was also performed to predict flow and heat transfer in fuel assembly including the spent fuel assembly in dry cask for interim repository. In addition, the research cooperation with university and nuclear fuel company was also carried out to develop a basic thermalhydraulic technology and the commercialization.

  17. THERMAL HYDRAULIC ISSUES OF CONTAINMENT FILTERED VENTING SYSTEM FOR A LONG OPERATING TIME

    Directory of Open Access Journals (Sweden)

    YOUNG SU NA

    2014-12-01

    Full Text Available This study investigated the thermal hydraulic issues in the Containment Filtered Venting System (CFVS for a long operating time using the MELCOR computer code. The modeling of the CFVS, including the models for pool scrubbing and the filter, was added to the input file for the OPR-1000, and a Station Blackout (SBO was chosen as an accident scenario. Although depressurization in the containment building as a primary objective of the CFVS was successful, the decontamination feature by scrubbing and filtering in the CFVS for a long operating time could fail by the continuous evaporation of the scrubbing solution. After the operation of the CFVS, the atmosphere temperature in the CFVS became slightly above the water saturation temperature owing to the release of an amount of steam with high temperature from the containment building to the scrubbing solution. Reduced pipe diameters at the inlet and outlet of the CFVS vessel mitigated the evaporation of scrubbing water by controlling the amount of high-temperature steam and the water saturation temperature.

  18. Summary of papers on current and anticipated uses of thermal-hydraulic codes

    Energy Technology Data Exchange (ETDEWEB)

    Caruso, R.

    1997-07-01

    The author reviews a range of recent papers which discuss possible uses and future development needs for thermal/hydraulic codes in the nuclear industry. From this review, eight common recommendations are extracted. They are: improve the user interface so that more people can use the code, so that models are easier and less expensive to prepare and maintain, and so that the results are scrutable; design the code so that it can easily be coupled to other codes, such as core physics, containment, fission product behaviour during severe accidents; improve the numerical methods to make the code more robust and especially faster running, particularly for low pressure transients; ensure that future code development includes assessment of code uncertainties as integral part of code verification and validation; provide extensive user guidelines or structure the code so that the `user effect` is minimized; include the capability to model multiple fluids (gas and liquid phase); design the code in a modular fashion so that new models can be added easily; provide the ability to include detailed or simplified component models; build on work previously done with other codes (RETRAN, RELAP, TRAC, CATHARE) and other code validation efforts (CSAU, CSNI SET and IET matrices).

  19. Current and anticipated uses of thermal-hydraulic codes in Spain

    Energy Technology Data Exchange (ETDEWEB)

    Pelayo, F.; Reventos, F. [Consejo de Seguridad Nuclear, Barcelona (Spain)

    1997-07-01

    Spanish activities in the field of Applied Thermal-Hydraulics are steadily increasing as the codes are becoming practicable enough to efficiently sustain engineering decision in the Nuclear Power industry. Before reaching this point, a lot of effort has been devoted to achieve this goal. This paper briefly describes this process, points at the current applications and draws conclusions on the limitations. Finally it establishes the applications where the use of T-H codes would be worth in the future, this in turn implies further development of the codes to widen the scope of application and improve the general performance. Due to the different uses of the codes, the applications mainly come from the authority, industry, universities and research institutions. The main conclusion derived from this paper establishes that further code development is justified if the following requisites are considered: (1) Safety relevance of scenarios not presently covered is established. (2) A substantial gain in margins or the capability to use realistic assumptions is obtained. (3) A general consensus on the licensability and methodology for application is reached. The role of Regulatory Body is stressed, as the most relevant outcome of the project may be related to the evolution of the licensing frame.

  20. Simplified thermal-hydraulic analysis of single phase natural circulation circuit with two heat exchangers

    Energy Technology Data Exchange (ETDEWEB)

    Pinheiro, Larissa Cunha; Su, Jian, E-mail: larissa@lasme.coppe.ufrj.br, E-mail: sujian@lasme.coppe.ufrj.br [Coordenacao dos Programas de Pos-Graduacao em Engenharia (COPPE/UFRJ), Rio de Janeiro, RJ (Brazil). Programa de Engenhraria Nuclear; Cotta, Renato Machado, E-mail: cotta@mecanica.coppe.ufrj.br [Coordenacao dos Programas de Pos-Graduacao em Engenharia (POLI/COPPE/UFRJ), Rio de Janeiro, RJ (Brazil). Dept. de Engenharia Mecanica

    2015-07-01

    Single phase natural circulation circuits composed of two convective heat exchangers and connecting tubes are important for the passive heat removal from spent fuel pools (SFP). To keep the structural integrity of the stored spent fuel assemblies, continuously cooling has to be provided in order to avoid increase at the pool temperature and subsequent uncovering of the fuel and enhanced reaction between water and metal releasing hydrogen. Decay heat can achieve considerably high amounts of energy e.g. in the AP1000, considering the emergency fuel assemblies, the maximum heat decay will reach 13 MW in the 15th day (Westinghouse Electric Company, 2010). A highly efficient alternative to do so is by means of natural circulation, which is cost-effective compared to active cooling systems and is inherently safer since presents less associated devices and no external work is required. Many researchers have investigated safety and stability aspects of natural circulation loops (NCL). However, there is a lack of literature concerning the improvement of NCL through a standard unified methodology, especially for natural circulation circuits with two heat exchangers. In the present study, a simplified thermal-hydraulic analysis of single phase natural circulation circuit with two heat exchanges is presented. Relevant dimensionless key groups were proposed to for the design and safety analysis of a scaled NCL for the cooling of spent fuel storage pool with convective cooling and heating. (author)

  1. Investigation of Correlations for the Thermal-hydraulic Analysis of Liquid Metal Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Chang, Won Pyo; Jeong, Hae Yong; Lee, Yong Bum

    2007-08-15

    The present investigation is aimed at reducing favorable constitutive correlations from those developed for the thermal-hydraulic analysis of Liquid Metal Reactors (LMR), for reliable safety analyses of KALIMER. It is achieved by analyzing them in a point of their accuracies. The study is particularly important because its outcomes can provide an essential knowledge on their relative errors including their conservatisms to be analyzed in the future KALIMER licensing stage. The predictions of the correlations have been compared with available experimental data on both friction factors for the wired-wrapped rod bundles in the core and the heat transfer coefficients in the system. As a result, the heat transfer coefficient inside pipe currently featured in SSC-K has been found acceptable. It, however, has shown a discrepancy of about 60 % and thus an alternative one has been proposed for improvement. Meanwhile, the friction factor model in the current SSC-K has not shown a prominent discrepancy in prediction trend but it has not backed an enough theoretical basis so that another model has been proposed. A systematic assessment for effects of those factors to the conservatism must be fully understood for the future licensing stage, and systematic calculations must be followed by designing an assessment matrix. Besides, it is essential to conduct experiments under similar conditions for constitutive parts of geometries which represent the KALIMER design.

  2. Numerical simulation of thermal-hydraulic processes in the riser chamber of installation for clinker production

    Directory of Open Access Journals (Sweden)

    Borsuk Grzegorz

    2016-03-01

    Full Text Available Clinker burning process has a decisive influence on energy consumption and the cost of cement production. A new problem is to use the process of decarbonization of alternative fuels from waste. These issues are particularly important in the introduction of a two-stage combustion of fuel in a rotary kiln without the typical reactor-decarbonizator. This work presents results of numerical studies on thermal-hydraulic phenomena in the riser chamber, which will be designed to burn fuel in the system where combustion air is supplied separately from the clinker cooler. The mathematical model is based on a combination of two methods of motion description: Euler description for the gas phase and Lagrange description for particles. Heat transfer between particles of raw material and gas was added to the numerical calculations. The main aim of the research was finding the correct fractional distribution of particles. For assumed particle distribution on the first stage of work, authors noted that all particles were carried away by the upper outlet to the preheater tower, what is not corresponding to the results of experimental studies. The obtained results of calculations can be the basis for further optimization of the design and operating conditions in the riser chamber with the implementation of the system.

  3. Development of best estimate auditing code for CANDU thermal-hydraulic safety analysis

    Energy Technology Data Exchange (ETDEWEB)

    Chung, Bub Dong; Lee, Won Jae; Hwang, Moon Kyu; Lim, Hong Sik [Korea Atomic Energy Research Institute, Taejeon (Korea)

    2002-04-01

    The main purpose of this study is to develop a thermal hydraulic auditing code for the CANDU reactor, modifying the model of existing PWR auditing tool, i.e. RELAP5/MOD3.The study was performed by reconsideration of the previous code assessment works and phenomena identification for essential accident scenario. Improvement areas of model development for auditing tool were identified based on the code comparison and PIRT results. Nine models have been improved significantly for the analysis of LOCA and Mon LOCA event. Conceptual problem or separate effect assessment have been performed to verify the model improvement. The linking calculation with CONTAIN 2.0 has been also enabled to establish the unified auditing code system. Analysis for the CANDU plant real transient and hypothetical LOCA bas been performed using the improved version. It has been concluded that the developed version can be utilized for the auditing analysis of LOCA and non-LOCA event for the CANDU reactor. 25 refs., 84 figs., 36 tabs. (Author)

  4. Roles of morphology, anatomy, and aquaporins in determining contrasting hydraulic behavior of roots.

    Science.gov (United States)

    Bramley, Helen; Turner, Neil C; Turner, David W; Tyerman, Stephen D

    2009-05-01

    The contrasting hydraulic properties of wheat (Triticum aestivum), narrow-leafed lupin (Lupinus angustifolius), and yellow lupin (Lupinus luteus) roots were identified by integrating measurements of water flow across different structural levels of organization with anatomy and modeling. Anatomy played a major role in root hydraulics, influencing axial conductance (L(ax)) and the distribution of water uptake along the root, with a more localized role for aquaporins (AQPs). Lupin roots had greater L(ax) than wheat roots, due to greater xylem development. L(ax) and root hydraulic conductance (L(r)) were related to each other, such that both variables increased with distance from the root tip in lupin roots. L(ax) and L(r) were constant with distance from the tip in wheat roots. Despite these contrasting behaviors, the hydraulic conductivity of root cells (Lp(c)) was similar for all species and increased from the root surface toward the endodermis. Lp(c) was largely controlled by AQPs, as demonstrated by dramatic reductions in Lp(c) by the AQP blocker mercury. Modeling the root as a series of concentric, cylindrical membranes, and the inhibition of AQP activity at the root level, indicated that water flow in lupin roots occurred primarily through the apoplast, without crossing membranes and without the involvement of AQPs. In contrast, water flow across wheat roots crossed mercury-sensitive AQPs in the endodermis, which significantly influenced L(r). This study demonstrates the importance of examining root morphology and anatomy in assessing the role of AQPs in root hydraulics.

  5. Thermal-Hydraulics and Electrochemistry of a Boiling Solution in a Porous Sludge Pile A Test Methodology

    Energy Technology Data Exchange (ETDEWEB)

    R.F. Voelker

    2001-05-03

    When boiling occurs in a pile of porous corrosion products (sludge), chemical species can concentrate. These species can react with the corrosion products and transform the sludge into a rock hard mass and/or create a corrosive environment. In-situ measurements are required to improve the understanding of this process, and the thermal-hydraulic and electrochemical environment in the pile. A test method is described that utilizes a water heated instrumented tube array in an autoclave to perform the in-situ measurements. As a proof of method feasibility, tests were performed in an alkaline phosphate solution. The test data is discussed. Temperature changes and electrochemical potential shifts were used to indicate when chemicals concentrate and if/when the pile hardens. Post-test examinations confirmed hardening occurred. Experiments were performed to reverse the hardening process. A one-dimensional model, utilizing capillary forces, was developed to understand the thermal-hydraulic measurements.

  6. Development of a preliminary PIRT (Phenomena Identification and Ranking Table) of thermal-hydraulic phenomena for SMART

    Energy Technology Data Exchange (ETDEWEB)

    Chung, Bub Dong; Lee, Won Jae; Kim, Hee Cheol; Song, Jin Ho; Sim, Suk Ku [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1997-12-31

    The work reported in this paper identifies the thermal-hydraulic phenomena that are expected to occur during a number of key transients in SMART (System-integrated Modular Advanced ReacTor) which is under development at KAERI. The result of this effort is based on the current design concept of SMART integral reactor. Although the design is still evolving, the preliminary Phenomena Identification and Ranking Table (PIRT) has been developed based on the experts` knowledge and experience. The preliminary PIRT has been developed by consensus of KAERI expert panelists and AHP (Analytical Hierarchy Process). Preliminary PIRT developed in this paper is intended to be used to identify and integrate development areas of further experimental tests needed, thermal hydraulic models and correlations and code improvements for the safety analysis of the SMART. 8 refs., 4 tabs (Author)

  7. Overview of the Use of ATHENA for Thermal-Hydraulic Analysis of Systems with Lead-Bismuth Coolant

    Energy Technology Data Exchange (ETDEWEB)

    Davis, Cliff Bybee; Shieh, Arthur Shan Luk

    2000-04-01

    The INEEL and MIT are investigating the suitability of lead-bismuth cooled fast reactor for producing low-cost electricity as well as for actinide burning. This paper is concerned with the general area of thermal-hydraulics of lead-bismuth cooled reactors. The ATHENA code is being used in the thermal-hydraulic design and analysis of lead-bismuth cooled reactors. The ATHENA code was reviewed to determine its applicability for simulating lead-bismuth cooled reactors. Two modifications were made to the code as a result of this review. Specifically, a correlation to represent heat transfer from rod bundles to a liquid metal and a void correlation based on data taken in a mixture of lead-bismuth and steam were added the code. The paper also summarizes the analytical work that is being performed with the code and plans for future analytical work.

  8. Overview of the use of ATHENA for thermal-hydraulic analysis of systems with lead-bismuth coolant

    Energy Technology Data Exchange (ETDEWEB)

    C. B. Davis; A. S. Shieh

    2000-04-02

    The INEEL and MIT are investigating the suitability of lead-bismuth cooled fast reactor for producing low-cost electricity as well as for actinide burning. This paper is concerned with the general area of thermal-hydraulics of lead-bismuth cooled reactors. The ATHENA code is being used in the thermal-hydraulic design and analysis of lead-bismuth cooled reactors. The ATHENA code was reviewed to determine its applicability for simulating lead-bismuth cooled reactors. Two modifications were made to the code as a result of this review. Specifically, a correlation to represent heat transfer from rod bundles to a liquid metal and a void correlation based on data taken in a mixture of lead-bismuth and steam were added the code. The paper also summarizes the analytical work that is being performed with the code and plans for future analytical work.

  9. Multi-scale, coupled Reactor Physics / Thermal-Hydraulics system and applications to the HPLWR 3 Pass Core

    OpenAIRE

    Monti, Lanfranco

    2009-01-01

    The HPLWR is an innovative reactor concept cooled with water at supercritical pressure. The pronounced changes of water properties with the heat-up demands advanced analyses tools which have been developed and successfully applied. Coupled neutronic/thermal-hydraulic analyses have been performed for the whole core and the coupled solution has been successively investigated at sub-channel resolution evaluating local quantities. The obtained results represent a new quality in core analyses.

  10. History of the 185-/189-D thermal hydraulics laboratory and its effects on reactor operations at the Hanford Site

    Energy Technology Data Exchange (ETDEWEB)

    Gerber, M.S.

    1994-09-01

    The 185-D deaeration building and the 189-D refrigeration building were constructed at Hanford during 1943 and 1944. Both buildings were constructed as part of the influent water cooling system for D reactor. The CMS studies eliminated the need for 185-D function. Early gains in knowledge ended the original function of the 189-D building mission. In 1951, 185-D and 189-D were converted to a thermal-hydraulic laboratory. The experiments held in the thermal-hydraulic lab lead to historic changes in Hanford reactor operations. In late 1951, the exponential physics experiments were moved to the 189-D building. In 1958, new production reactor experiments were begun in 185/189-D. In 1959, Plutonium Recycle Test Reactor experiments were added to the 185/189-D facility. By 1960, the 185/189-D thermal hydraulics laboratory was one of the few full service facilities of its type in the nation. During the years 1961--1963 tests continued in the facility in support of existing reactors, new production reactors, and the Plutonium Recycle Test Reactor. In 1969, Fast Flux Test Facility developmental testings began in the facility. Simulations in 185/189-D building aided in the N Reactor repairs in the 1980`s. In 1994 the facility was nominated to the National Register of Historic Places, because of its pioneering role over many years in thermal hydraulics, flow studies, heat transfer, and other reactor coolant support work. During 1994 and 1995 it was demolished in the largest decontamination and decommissioning project thus far in Hanford Site history.

  11. Experimental studies into the thermal-hydraulic performance of the VK-300 reactor based on a draft tube model

    Directory of Open Access Journals (Sweden)

    N.P. Serdun

    2015-12-01

    Full Text Available The paper presents an experimental study into the thermal-hydraulic performance of the VK-300 reactor based on a model of a single draft tube at a pressure of 3.4MPa, various flow rates and the model inlet relative enthalpies of –0.05 to 0.2. The experimental procedures include generation of a steam-water mixture circulation with a preset flow rate and a relative enthalpy through the test section at a pressure of 3.3 to 3.4MPa, and measurement of thermal-hydraulic parameters within the circuit's representative upflow and downflow lengths of practical interest. There have been confirmed the designs used to support the reactor facility serviceability and the assumptions concerning the thermal-hydraulic performance of a natural circulation circuit used in the analysis thereof. It has been shown that, across the analyzed range of the relative enthalpy values, the draft tube has an annular-dispersed or an annular flow of the steam-water mixture, both providing for the significant separation of the steam-water mixture (Ksep=0.4 at the draft tube edges and in the mixing chamber. The perforation in the upper part of the draft tubes allows the separation coefficient to be increased at the first stage and creates more favorable conditions for the second-stage separation. The measured values of the void fraction in the mixing chamber and in the draft tube are in a satisfactory agreement with calculations based on Z.L. Miropolskiy's method and the RELAP code and may be used to verify the VK-300 thermal-hydraulic codes. It has been shown that steam may enter the ring slit that simulates the annular space and reach the reactor core inlet. Further investigations need to be conducted to study this effect for its guaranteed exclusion and for the development of emergency response procedures.

  12. Development of Effective Algorithm for Coupled Thermal-Hydraulics – Neutron-Kinetics Analysis of Reactivity Transient

    OpenAIRE

    Peltonen, Joanna

    2009-01-01

    Analyses of nuclear reactor safety have increasingly required coupling of full three dimensional neutron kinetics (NK) core models with system transient thermal-hydraulics (TH) codes. To produce results within a reasonable computing time, the coupled codes use different spatial description of the reactor core. The TH code uses few, typically 5 to 20 TH channels, which represent the core. The NK code uses explicit node for each fuel assembly. Therefore, a spatial mapping of coarse grid TH and ...

  13. Parametric studies by means of uncertainty and sensitivity methods for coupled thermal-hydraulic/neutron-physics application

    Energy Technology Data Exchange (ETDEWEB)

    Jaeger, W.; Sanchez, V.; Cheng, X. [Karlsruhe Institute of Technology (KIT), Eggenstein-Leopoldshafen (Germany). Inst. for Neutron Physics and Reactor Technology; Monti, L. [Karlsruhe Institute of Technology (KIT), Eggenstein-Leopoldshafen (Germany). Inst. for Nuclear and Energy Technologies; Hurtado, A. [Technical Univ. of Dresden (Germany). Inst. of Power Engineering

    2011-07-01

    At the Institute for Neutron Physics and Reactor Technology (INR) at the Karlsruhe Institute of Technology (KIT), the development and validation of coupled codes systems is one major activity. In this paper, a 2-step method is proposed to perform uncertainty and sensitivity analysis of a nuclear fuel bundle. At first, the SUSA package (Software system for Uncertainty and Sensitivity Analysis), 2 is applied to the thermal hydraulic results of the TRACE (TRACE/RELAP Advanced Computational Engine) code to identify crucial thermal hydraulic parameter combinations which are successively used in the TH/NP coupled system TRACEERANOS to account for the neutronic feedbacks. This 2-step method was applied since the TRACE-ERANOS system runs 1 input in approximately 1 day (depending on the computer configurations). Since the uncertainty and sensitivity analysis requires about 100 runs of the thermal hydraulic input (with altered parameters, running within minutes) an integral TRACE-SUSA-ERANOS analysis would need around 100 days. For this analysis a fuel assembly model of the HPLWR (High Performance Light Water Reactor) was selected. Due to the general structure of the coupling and code communication scripts, the system can be used for any kind of reactor/system which can be described with TRACE and ERANOS (e.g., fast systems) while SUSA can be applied to anything. (orig.)

  14. The time-dependent 3D discrete ordinates code TORT-TD with thermal-hydraulic feedback by ATHLET models

    Energy Technology Data Exchange (ETDEWEB)

    Seubert, A.; Velkov, K.; Langenbuch, S. [Gesellschaft fuer Anlagen- und Reaktorsicherheit (GRS) mbH, Forschungsinstitute, D-85748 Garching (Germany)

    2008-07-01

    This paper describes the time-dependent 3D discrete ordinates transport code TORT-TD. Thermal-hydraulic feedback is considered by coupling TORT-TD with the thermal-hydraulics system code ATHLET. The coupled code TORT-TD/ATHLET allows 3D pin-by-pin analyses of transients in few energy groups and anisotropic scattering by solving the time-dependent transport equation using the unconditionally stable implicit method. The nuclear cross sections are interpolated between pre-calculated table values of fuel temperature, moderator density and boron concentration. For verification of the implementation, selected test cases have been calculated by TORT-TD/ATHLET. They include a control rod ejection transient in a small PWR fuel assembly arrangement and a local boron concentration change in a single PWR fuel assembly. In the latter, special attention has been paid to study the influence of the thermal-hydraulic feedback modelling in ATHLET. The results obtained for a control rod ejection accident in a PWR quarter core demonstrate the applicability of TORT-TD/ATHLET. (authors)

  15. Development of CFD thermal hydraulics and neutron kinetics coupling methodologies for the prediction of local safety parameters for light water reactors

    Energy Technology Data Exchange (ETDEWEB)

    Perez Manes, Jorge

    2013-02-26

    This dissertation contributes to the development of high-fidelity coupled neutron kinetic and thermal hydraulic simulation tools with high resolution of the spatial discretization of the involved domains for the analysis of Light Water Reactors transient scenarios.

  16. Coupled neutronics/thermal-hydraulics analysis of a high-performance light-water reactor fuel assembly

    Energy Technology Data Exchange (ETDEWEB)

    Waata, C.L.

    2006-07-15

    The use of water at supercritical pressure as coolant and moderator introduces a challenge in the design of a High-Performance Light-Water Reactor (HPLWR) fuel assembly. At supercritical pressure condition (P=25 MPa), the thermal-hydraulics behaviour of water differs strongly from that at sub-critical pressure due to a rapid variation of the thermal-physical properties across the pseudo-critical line. Due of the strong link between the water (moderation) and the neutron spectrum and subsequently the power distribution, a coupling of neutronics and thermal-hydraulics has become a necessity for reactor concepts operating at supercritical pressure condition. The effect of neutron moderation on the local parameters of thermal-hydraulics and vice-verse in a fuel assembly has to be considered for an accurate design analysis. In this study, the Monte Carlo N-Particle code (MCNP) and the sub-channel code STAFAS (Sub-channel Thermal-hydraulics Analysis of a Fuel Assembly under Supercritical conditions) have been coupled for the design analysis of a fuel assembly with supercritical water as coolant and moderator. Both codes are well known for complex geometry modelling. The MCNP code is used for neutronics analyses and for the prediction of power profiles of individual fuel rods. The sub-channel code STAFAS for the thermal-hydraulics analyses takes into account the coolant properties beyond the critical point as well as separate moderator channels. The coupling procedure is realized automatically. MCNP calculates the power distribution in each fuel rod, which is then transferred into STAFAS to obtain the corresponding thermal-hydraulic conditions in each sub-channel. The new thermal-hydraulic conditions are used to generate a new input deck for the next MCNP calculation. This procedure is repeated until a converged state is achieved. The coupled code system was tested on a proposed fuel assembly design of a HPLWR. An under-relaxation was introduced to achieve convergence

  17. Argonne Liquid-Metal Advanced Burner Reactor : components and in-vessel system thermal-hydraulic research and testing experience - pathway forward.

    Energy Technology Data Exchange (ETDEWEB)

    Kasza, K.; Grandy, C.; Chang, Y.; Khalil, H.; Nuclear Engineering Division

    2007-06-30

    This white paper provides an overview and status report of the thermal-hydraulic nuclear research and development, both experimental and computational, conducted predominantly at Argonne National Laboratory. Argonne from the early 1970s through the early 1990s was the Department of Energy's (DOE's) lead lab for thermal-hydraulic development of Liquid Metal Reactors (LMRs). During the 1970s and into the mid-1980s, Argonne conducted thermal-hydraulic studies and experiments on individual reactor components supporting the Experimental Breeder Reactor-II (EBR-II), Fast Flux Test Facility (FFTF), and the Clinch River Breeder Reactor (CRBR). From the mid-1980s and into the early 1990s, Argonne conducted studies on phenomena related to forced- and natural-convection thermal buoyancy in complete in-vessel models of the General Electric (GE) Prototype Reactor Inherently Safe Module (PRISM) and Rockwell International (RI) Sodium Advanced Fast Reactor (SAFR). These two reactor initiatives involved Argonne working closely with U.S. industry and DOE. This paper describes the very important impact of thermal hydraulics dominated by thermal buoyancy forces on reactor global operation and on the behavior/performance of individual components during postulated off-normal accident events with low flow. Utilizing Argonne's LMR expertise and design knowledge is vital to the further development of safe, reliable, and high-performance LMRs. Argonne believes there remains an important need for continued research and development on thermal-hydraulic design in support of DOE's and the international community's renewed thrust for developing and demonstrating the Global Nuclear Energy Partnership (GNEP) reactor(s) and the associated Argonne Liquid Metal-Advanced Burner Reactor (LM-ABR). This white paper highlights that further understanding is needed regarding reactor design under coolant low-flow events. These safety-related events are associated with the transition

  18. First vapor explosion calculations performed with MC3D thermal-hydraulic code

    Energy Technology Data Exchange (ETDEWEB)

    Brayer, C.; Berthoud, G. [CEA Centre d`Etudes de Grenoble, 38 (France). Direction des Reacteurs Nucleaires

    1998-01-01

    This paper presents the first calculations performed with the `explosion` module of the multiphase computer code MC3D, which is devoted to the fine fragmentation and explosion phase of a fuel coolant interaction. A complete description of the physical laws included in this module is given. The fragmentation models, taking into account two fragmentation mechanisms, a thermal one and an hydrodynamic one, are also developed here. Results to some calculations to test the numerical behavior of MC3D and to test the explosion models in 1D or 2D are also presented. (author)

  19. Simulating High Flux Isotope Reactor Core Thermal-Hydraulics via Interdimensional Model Coupling

    Energy Technology Data Exchange (ETDEWEB)

    Travis, Adam R [ORNL

    2014-05-01

    A coupled interdimensional model is presented for the simulation of the thermal-hydraulic characteristics of the High Flux Isotope Reactor core at Oak Ridge National Laboratory. The model consists of two domains a solid involute fuel plate and the surrounding liquid coolant channel. The fuel plate is modeled explicitly in three-dimensions. The coolant channel is approximated as a twodimensional slice oriented perpendicular to the fuel plate s surface. The two dimensionally-inconsistent domains are linked to one another via interdimensional model coupling mechanisms. The coupled model is presented as a simplified alternative to a fully explicit, fully three-dimensional model. Involute geometries were constructed in SolidWorks. Derivations of the involute construction equations are presented. Geometries were then imported into COMSOL Multiphysics for simulation and modeling. Both models are described in detail so as to highlight their respective attributes in the 3D model, the pursuit of an accurate, reliable, and complete solution; in the coupled model, the intent to simplify the modeling domain as much as possible without affecting significant alterations to the solution. The coupled model was created with the goal of permitting larger portions of the reactor core to be modeled at once without a significant sacrifice to solution integrity. As such, particular care is given to validating incorporated model simplifications. To the greatest extent possible, the decrease in solution time as well as computational cost are quantified versus the effects such gains have on the solution quality. A variant of the coupled model which sufficiently balances these three solution characteristics is presented alongside the more comprehensive 3D model for comparison and validation.

  20. Thermal-hydraulic analysis for changing feedwater check valve leakage rate testing methodology

    Energy Technology Data Exchange (ETDEWEB)

    Fuller, R.; Harrell, J.

    1996-12-01

    The current design and testing requirements for the feedwater check valves (FWCVs) at the Grand Gulf Nuclear Station are established from original licensing requirements that necessitate extremely restrictive air testing with tight allowable leakage limits. As a direct result of these requirements, the original high endurance hard seats in the FWCVs were modified with elastomeric seals to provide a sealing surface capable of meeting the stringent air leakage limits. However, due to the relatively short functional life of the elastomeric seals compared to the hard seats, the overall reliability of the sealing function actually decreased. This degraded performance was exhibited by frequent seal failures and subsequent valve repairs. The original requirements were based on limited analysis and the belief that all of the high energy feedwater vaporized during the LOCA blowdown. These phenomena would have resulted in completely voided feedwater lines and thus a steam environment within the feedwater leak pathway. To challenge these criteria, a comprehensive design basis accident analysis was developed using the RELAP5/MOD3.1 thermal-hydraulic code. Realistic assumptions were used to more accurately model the post-accident fluid conditions within the feedwater system. The results of this analysis demonstrated that no leak path exists through the feedwater lines during the reactor blowdown phase and that sufficient subcooled water remains in various portions of the feedwater piping to form liquid water loop seals that effectively isolate this leak path. These results provided the bases for changing the leak testing requirements of the FWCVs from air to water. The analysis results also established more accurate allowable leakage limits, determined the real effective margins associated with the FWCV safety functions, and led to design changes that improved the overall functional performance of the valves.

  1. Parametric studies on containment thermal hydraulic loads during high pressure melt ejection in a BWR

    Energy Technology Data Exchange (ETDEWEB)

    Silde, A.; Lindholm, I. [VTT Energy, Espoo (Finland)

    1997-12-01

    The containment thermal hydraulic loads during high pressure melt ejection in a Nordic BWR are studied parametrically with the CONTAIN and the MELCOR codes. The work is part of the Nordic RAK-2 project. The containment analyses were divided into two categories according to composition of the discharged debris: metallic and oxidic debris cases. In the base case with highly metallic debris, all sources from the reactor coolant system to the containment were based on the MELCOR/BH calculation. In the base case with the oxidic debris, the source data was specified assuming that {approx} 15% of the whole core material inventory and 34,000 kg of saturated water was discharged from the reactor pressure vessel (RPV) during 30 seconds. In this case, the debris consisted mostly of oxides. The highest predicted containment pressure peaks were about 8.5 bar. In the scenarios with highly metallic debris source, very high gas temperature of about 1900 K was predicted in the pedestal, and about 1400 K in the upper drywell. The calculations with metallic debris were sensititive to model parameters, like the particle size and the parameters, which control the chemical reaction kinetics. In the scenarios with oxidic debris source, the predicted pressure peaks were comparable to the cases with the metallic debris source. The maximum gas temperatures (about 450-500 K) in the containment were, however, significantly lower than in the respective metallic debris case. The temperatures were also insensitive to parametric variations. In addition, one analysis was performed with the MELCOR code for benchmarking of the MELCOR capabilities against the more detailed CONTAIN code. The calculations showed that leak tightness of the containment penetrations could be jeopardized due to high temperature loads, if a high pressure melt ejection occurred during a severe accident. Another consequence would be an early containment venting. (au). 28 refs.

  2. ANTEO+: A subchannel code for thermal-hydraulic analysis of liquid metal cooled systems

    Energy Technology Data Exchange (ETDEWEB)

    Lodi, F., E-mail: francesco.lodi5@unibo.it [DIN – Laboratory of Montecuccolino, University of Bologna, Via dei Colli 16, 40136 Bologna (Italy); Grasso, G., E-mail: giacomo.grasso@enea.it [ENEA (Italian National Agency for New Technologies, Energy and Sustainable Economic Development), via Martiri di Monte Sole, 4, 40129 Bologna (Italy); Mattioli, D., E-mail: davide.mattioli@enea.it [ENEA (Italian National Agency for New Technologies, Energy and Sustainable Economic Development), via Martiri di Monte Sole, 4, 40129 Bologna (Italy); Sumini, M., E-mail: marco.sumini@unibo.it [DIN – Laboratory of Montecuccolino, University of Bologna, Via dei Colli 16, 40136 Bologna (Italy)

    2016-05-15

    Highlights: • The code structure is presented in detail. • The performed validation is outlined. • Results are critically discussed assessing code accuracy. • Conclusions are drawn and ground for future work identified. - Abstract: Liquid metal cooled fast reactors are promising options for achieving the high degrees of safety and sustainability demanded by the Generation IV paradigm. Among the critical aspects to be addressed in the design process, thermal-hydraulics is one of the most challenging; in order to embed safety in the core conceptualization, these aspects are to be considered at the very beginning of the design process, and translated in a design perspective. For achieving these objectives the subchannel code ANTEO+ has been conceived, able to simulate pin bundle arrangements cooled by liquid metals. The main purposes of ANTEO+ are simplifying the problem description maintaining the required accuracy, enabling a more transparent interface with the user, and having a clear and identifiable application domain, in order to help the user interpreting the results and, mostly, defining their confidence. Since ANTEO+ relies on empirical correlations, the validation phase is of paramount importance along with a clear discussion on the simplifications adopted in modeling the conservation equations. In the present work a detailed description of ANTEO+ structure is given along with a thorough validation of the main models implemented for flow split, pressure drops and subchannel temperatures. The analysis confirmed the ability of ANTEO+ in reproducing experimental data in its anticipated validity domain, with a relatively high degree of accuracy when compared to other classical subchannel tools like ENERGY-II, COBRA-IV-I-MIT and BRS-TVS.

  3. Development of a multi-dimensional realistic thermal-hydraulic system analysis code, MARS 1.3 and its verification

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Won Jae; Chung, Bub Dong; Jeong, Jae Jun; Ha, Kwi Seok [Korea Atomic Energy Research Institute, Taejon (Korea)

    1998-06-01

    A multi-dimensional realistic thermal-hydraulic system analysis code, MARS version 1.3 has been developed. Main purpose of MARS 1.3 development is to have the realistic analysis capability of transient two-phase thermal-hydraulics of Pressurized Water Reactors (PWRs) especially during Large Break Loss of Coolant Accidents (LBLOCAs) where the multi-dimensional phenomena domain the transients. MARS code is a unified version of USNRC developed COBRA-TF, domain the transients. MARS code is a unified version of USNRC developed COBRA-TF, three-dimensional (3D) reactor vessel analysis code, and RELAP5/MOD3.2.1.2, one-dimensional (1D) reactor system analysis code., Developmental requirements for MARS are chosen not only to best utilize the existing capability of the codes but also to have the enhanced capability in code maintenance, user accessibility, user friendliness, code portability, code readability, and code flexibility. For the maintenance of existing codes capability and the enhancement of code maintenance capability, user accessibility and user friendliness, MARS has been unified to be a single code consisting of 1D module (RELAP5) and 3D module (COBRA-TF). This is realized by implicitly integrating the system pressure matrix equations of hydrodynamic models and solving them simultaneously, by modifying the 1D/3D calculation sequence operable under a single Central Processor Unit (CPU) and by unifying the input structure and the light water property routines of both modules. In addition, the code structure of 1D module is completely restructured using the modular data structure of standard FORTRAN 90, which greatly improves the code maintenance capability, readability and portability. For the code flexibility, a dynamic memory management scheme is applied in both modules. MARS 1.3 now runs on PC/Windows and HP/UNIX platforms having a single CPU, and users have the options to select the 3D module to model the 3D thermal-hydraulics in the reactor vessel or other

  4. Development of a kinetics analysis code for fuel solution combined with thermal-hydraulics analysis code PHOENICS and analysis of natural-cooling characteristic test of TRACY. Contract research

    Energy Technology Data Exchange (ETDEWEB)

    Watanabe, Shouichi; Yamane, Yuichi; Miyoshi, Yoshinori [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    2003-03-01

    Since exact information is not always acquired in the criticality accident of fuel solution, parametric survey calculations are required for grasping behaviors of the thermal-hydraulics. On the other hand, the practical methods of the calculation with can reduce the computation time with allowable accuracy will be also required, since the conventional method takes a long calculation time. In order to fulfill the requirement, a two-dimensional (R-Z) nuclear-kinetics analysis code considering thermal-hydraulic based on the multi-region kinetic equations with one-group neutron energy was created by incorporating with the thermal-hydraulics analysis code PHOENICS for all-purpose use the computation time of the code was shortened by separating time mesh intervals of the nuclear- and heat-calculations from that of the hydraulics calculation, and by regulating automatically the time mesh intervals in proportion to power change rate. A series of analysis were performed for the natural-cooling characteristic test using TRACY in which the power changed slowly for 5 hours after the transient power resulting from the reactivity insertion of a 0.5 dollar. It was found that the code system was able to calculate within the limit of practical time, and acquired the prospect of reproducing the experimental values considerably for the power and temperature change. (author)

  5. Fracture opening/propagation behavior and their significance on pressure-time records during hydraulic fracturing

    Energy Technology Data Exchange (ETDEWEB)

    Takashi Kojima; Yasuhiko Nakagawa; Koji Matsuki; Toshiyuki Hashida

    1992-01-01

    Hydraulic fracturing with constant fluid injection rate was numerically modeled for a pair of rectangular longitudinal fractures intersecting a wellbore in an impermeable rock mass, and numerical calculations have been performed to investigate the relations among the form of pressure-time curves, fracture opening/propagation behavior and permeability of the mechanically closed fractures. The results have shown that both permeability of the fractures and fluid injection rate significantly influence the form of the pressure-time relations on the early stage of fracture opening. Furthermore it has been shown that wellbore pressure during fracture propagation is affected by the pre-existing fracture length.

  6. Thermal behavior of ground talc mineral

    Directory of Open Access Journals (Sweden)

    Balek V.

    2008-01-01

    Full Text Available Microstructure development and thermal behavior of ground talc mineral samples (from locality Puebla de Lillo, Spain was characterized by X-ray diffraction, scanning electron microscopy(SEM,surface area measurements differential thermal analysis, thermogravimetry and emanation thermal analysis (ETA.A vibratory mill and grinding time 5 min. was used to prepare the ground talc sample. It was found that grinding caused an increase in the surface area of the natural talc from 3 m2g-1 to 110 m2g-1. A decrease of particle size after sample grinding was observed by SEM. The increase of structure disorder of the ground sample and the crystallite size reduction from 40 to 10 nm were determined from the XRD results. ETA results revealed a closing up of surface micro-cracks and healing of microstructure irregularities on heating in the range 200-500ºC of both un-ground and ground talc samples. For the ground talc sample a crystallization of non-crystalline phase into orthorhombic enstatite was characterized as by a decrease of radon mobility in the range 785-825ºC and by a DTA exothermal effect with the maximum at 830ºC. ETA results were used for the assessment of transport properties of the talc samples on heating.

  7. IAEA Coordinated Research Project on HTGR Reactor Physics, Thermal-hydraulics and Depletion Uncertainty Analysis

    Energy Technology Data Exchange (ETDEWEB)

    Strydom, Gerhard [Idaho National Lab. (INL), Idaho Falls, ID (United States); Bostelmann, F. [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2015-09-01

    The continued development of High Temperature Gas Cooled Reactors (HTGRs) requires verification of HTGR design and safety features with reliable high fidelity physics models and robust, efficient, and accurate codes. The predictive capability of coupled neutronics/thermal-hydraulics and depletion simulations for reactor design and safety analysis can be assessed with sensitivity analysis (SA) and uncertainty analysis (UA) methods. Uncertainty originates from errors in physical data, manufacturing uncertainties, modelling and computational algorithms. (The interested reader is referred to the large body of published SA and UA literature for a more complete overview of the various types of uncertainties, methodologies and results obtained). SA is helpful for ranking the various sources of uncertainty and error in the results of core analyses. SA and UA are required to address cost, safety, and licensing needs and should be applied to all aspects of reactor multi-physics simulation. SA and UA can guide experimental, modelling, and algorithm research and development. Current SA and UA rely either on derivative-based methods such as stochastic sampling methods or on generalized perturbation theory to obtain sensitivity coefficients. Neither approach addresses all needs. In order to benefit from recent advances in modelling and simulation and the availability of new covariance data (nuclear data uncertainties) extensive sensitivity and uncertainty studies are needed for quantification of the impact of different sources of uncertainties on the design and safety parameters of HTGRs. Only a parallel effort in advanced simulation and in nuclear data improvement will be able to provide designers with more robust and well validated calculation tools to meet design target accuracies. In February 2009, the Technical Working Group on Gas-Cooled Reactors (TWG-GCR) of the International Atomic Energy Agency (IAEA) recommended that the proposed Coordinated Research Program (CRP) on

  8. Improvement of auditing technology of safety analysis through thermal-hydraulic separate effect tests

    Energy Technology Data Exchange (ETDEWEB)

    No, Hee Cheon; Park, Hyun Sik; Kim, Hyoung Tae; Moon, Young Min; Choi, Sung Won; Hwang, Do Hyun [Korea Advanced Institute of Science and Technology, Taejon (Korea, Republic of)

    2000-03-15

    The direct-contact condensation hear transfer coefficients are experimentally obtained in the following conditions : pure steam/steam in the presence of noncondensible gas, horizontal/slightly inclined pipe, cocurrent/countercurrent stratified flow with water. The empirical correlation for liquid Nusselt number is developed in conditions of the slightly inclined pipe and the cocurrent stratified flow. The several models - the wall friction coefficient, the interfacial friction coefficient, the correlation of direct-contact condensation with noncondensible gases, and the correlation of wall film condensation - in the RELAP5/MOD3.2 code are modified, As results, RELAP5/MOD3.2 is improved. The present experimental data is used for evaluating the improved code. The standard RELAP5/MOD3.2 code is modified using the non-iterative modeling, which is a mechanistic model and does not require any interfacial information such as the interfacial temperature, The modified RELAP5/MOD3.2 code os used to simulate the horizontally stratified in-tube condensation experiment which represents the direct-contact condensation phenomena in a hot leg of a nuclear reactor. The modeling capabilities of the modified code as well as the standard code are assessed using several hot-leg condensation experiments. The modified code gives better prediction over local experimental data of liquid void fraction and interfacial heat transfer coefficient than the standard code. For the separate effect test of the thermal-hydraulic phenomena in the pressurizer, the scaling analysis is performed to obtain a similarity of the phenomena between the Korea Standard Nuclear Power Plant(KSNPP) and the present experimental facility. The diameters and lengths of the hot-leg, the surge line and the pressurizer are scaled down with the similitude of CCFL and velocity. The ratio of gas flow rate is 1/25. The experimental facility is composed of the air-water supply tank, the horizontal pipe, the surge line and the

  9. Screening of Gas-Cooled Reactor Thermal-Hydraulic and Safety Analysis Tools and Experimental Database

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Won Jae; Kim, Min Hwan; Lee, Seung Wook (and others)

    2007-08-15

    This report is a final report of I-NERI Project, 'Screening of Gas-cooled Reactor Thermal Hydraulic and Safety Analysis Tools and Experimental Database 'jointly carried out by KAERI, ANL and INL. In this study, we developed the basic technologies required to develop and validate the VHTR TH/safety analysis tools and evaluated the TH/safety database information. The research tasks consist of; 1) code qualification methodology (INL), 2) high-level PIRTs for major nucleus set of events (KAERI, ANL, INL), 3) initial scaling and scoping analysis (ANL, KAERI, INL), 4) filtering of TH/safety tools (KAERI, INL), 5) evaluation of TH/safety database information (KAERI, INL, ANL) and 6) key scoping analysis (KAERI). The code qualification methodology identifies the role of PIRTs in the R and D process and the bottom-up and top-down code validation methods. Since the design of VHTR is still evolving, we generated the high-level PIRTs referencing 600MWth block-type GT-MHR and 400MWth pebble-type PBMR. Nucleus set of events that represents the VHTR safety and operational transients consists of the enveloping scenarios of HPCC (high pressure conduction cooling: loss of primary flow), LPCC/Air-Ingress (low pressure conduction cooling: loss of coolant), LC (load changes: power maneuvering), ATWS (anticipated transients without scram: reactivity insertion), WS (water ingress: water-interfacing system break) and HU (hydrogen-side upset: loss of heat sink). The initial scaling analysis defines dimensionless parameters that need to be reflected in mixed convection modeling and the initial scoping analysis provided the reference system transients used in the PIRTs generation. For the PIRTs phenomena, we evaluated the modeling capability of the candidate TH/safety tools and derived a model improvement need. By surveying and evaluating the TH/safety database information, a tools V and V matrix has been developed. Through the key scoping analysis using available database, the

  10. Optimization study and neutronic and thermal-hydraulic design calculations of a 75 KWTH aqueous homogeneous reactor for medical isotopes production

    Energy Technology Data Exchange (ETDEWEB)

    Perez, Daniel Milian; Lorenzo, Daniel E. Milian; Garcia, Lorena P. Rodriguez; Llanes, Jesus Salomon; Hernandez, Carlos R. Garcia, E-mail: dperez@instec.cu, E-mail: dmilian@instec.cu, E-mail: lorenapilar@instec.cu, E-mail: cgh@instec.cu [Instituto Superior de Tecnologias y Ciencias Aplicadas (InSTEC), La Habana (Cuba); Lira, Carlos A. Brayner de Oliveira, E-mail: cabol@ufpe.br [Universidade Federal de Pernambuco (UFPE), Recife (Brazil); Rodriguez, Manuel Cadavid, E-mail: mcadavid2001@yahoo.com [Tecnologia Nuclear Medica Spa, TNM (Chile)

    2015-07-01

    {sup 99m}Tc is the most common radioisotope used in nuclear medicine. It is a very useful radioisotope, which is used in about 30-40 million procedures worldwide every year. Medical diagnostic imaging techniques using {sup 99m}Tc represent approximately 80% of all nuclear medicine procedures. Although {sup 99m}Tc can be produced directly on a cyclotron or other type of particle accelerator, currently is almost exclusively produced from the beta-decay of its 66-h parent {sup 99}Mo. {sup 99}Mo production system in an Aqueous Homogeneous Reactor (AHR) is potentially advantageous because of its low cost, small critical mass, inherent passive safety, and simplified fuel handling, processing and purification characteristics. In this paper, an AHR conceptual design using Low Enriched Uranium (LEU) is studied and optimized for the production of {sup 99}Mo. Aspects related with the neutronic behavior such as optimal reflector thickness, critical height, medical isotopes production and the reactivity feedback introduced in the solution by the volumetric expansion of the fuel solution due to thermal expansion of the fuel solution and the void volume generated by radiolytic gas bubbles were evaluated. Thermal-hydraulics studies were carried out in order to show that sufficient cooling capacity exists to prevent fuel overheating. The neutronic and thermal-hydraulics calculations have been performed with the MCNPX computational code and the version 14 of ANSYS CFX respectively. The neutronic calculations demonstrated that the reactor is able to produce 370 six-day curies of {sup 99}Mo in 5 days operation cycles and the CFD simulation demonstrated that the heat removal systems provide sufficient cooling capacity to prevent fuel overheating, the maximum temperature reached by the fuel (89.29 deg C) was smaller to the allowable temperature limit (90 deg C). (author)

  11. Neutronic and thermal-hydraulic calculations for the AP-1000 NPP with the MCNP6 and SERPENT codes

    Energy Technology Data Exchange (ETDEWEB)

    Stefani, Giovanni Laranjo; Maiorino, Jose R.; Santos, Thiago A., E-mail: giovanni.laranjo@ufabc.edu.br, E-mail: joserubens.maiorino@ufabc.edu.br, E-mail: thiago.santos@ufabc.edu.br [Universidade Federal do ABC (CECS/UFABC), Santo Andre, SP (Brazil). Centro de Engenharia, Modelagem e Ciencias Sociais; Rossi, Pedro R., E-mail: pedro.russorossi@gmail.com [FERMIUM - Tecnologia Nuclear, Sao Paulo, SP (Brazil)

    2015-07-01

    The AP-1000 is an evolutionary PWR reactor designed as an evolution of the AP-600 project. The reactor is already pre-licensed by NRC, and is considered to have achieved high standards of safety, possible short construction time and good economic competitiveness. The core is a 17x17 typical assembly using Zirlo as cladding, 3 different enrichment regions, and is controlled by boron, control banks, and burnable poison. The expected fuel final burnup is 62 MWD/ton U and a cycle of 18 months. In this paper we present results for neutronic and thermal-hydraulic calculations for the AP-1000. We use the MCNP6 and SERPENT codes to calculate the first cycle of operation. The calculated parameters are K{sub eff} at BOL and EOL and its variation with burnup and neutron flux, and reactivity coefficients. The production of transuranic elements such as Pu-239 and Pu-241, and burning fuel are calculated over time. In the work a complete reactor was burned for 450 days with no control elements, boron or burnable poison were considered, these results were compared with data provided by the Westinghouse. The results are compared with those reported in the literature. A simple thermal hydraulic analysis allows verification of thermal limits such as fuel and cladding temperatures, and MDNB. (author)

  12. Development of a 1D thermal-hydraulic analysis code for once-through steam generator in SMRs using straight tubes

    Energy Technology Data Exchange (ETDEWEB)

    Park, Youngjae; Kim, Iljin; Kim, Hyungdae [Kyung Hee University, Yongin (Korea, Republic of)

    2015-10-15

    Diverse integral/small-modular reactors (SMRs) have been developed. Once-through steam generator (OTSG) which generates superheated steam without steam separator and dryer was used in the SMRs to reduce volume of steam generator. It would be possible to design a new steam generator with best estimate thermal-hydraulic codes such as RELAP and MARS. However, it is not convenience to use the general purpose thermal-hydraulic analysis code to design a specific component of nuclear power plants. A widely used simulation tool for thermal-hydraulic analysis of drum-type steam generators is ATHOS, which allows 3D analysis. On the other hand, a simple 1D thermal-hydraulic analysis code might be accurate enough for the conceptual design of OTSG. In this study, thermal-hydraulic analysis code for conceptual design of OTSG was developed using 1D homogeneous equilibrium model (HEM). A benchmark calculation was also conducted to verify and validate the prediction accuracy of the developed code by comparing with the analysis results with MARS. Finally, conceptual design of OTSG was conducted by the developed code. A simple 1D thermal-hydraulic analysis code was developed for the purpose of conceptual design OTSG for SMRs. A set of benchmark calculations was conducted to verify and validate the analysis accuracy of the developed code by comparing results obtained with a best-estimated thermal-hydraulic analysis code, MARS. Finally, analysis of two different OTSG design concepts with superheating and recirculation was demonstrated using the developed code.

  13. Dynamic Response in Transient Stress-Field Behavior Induced by Hydraulic Fracturing

    Science.gov (United States)

    Jenkins, Andrew

    magnitude. These types of shifts are of great concern because they can impact subsequent fracture development causing non-uniform fracture propagation and the potential overlapping of fracture paths as they extend from the wellbore at the point of injection. The dynamics of stress variation that occur with respect to hydraulic fracturing is a somewhat new area of study. In order to accomplish the goals of this thesis and continue future research in this area a new transient model has been developed in order to asses these dynamic systems and determine their influence on fracture behavior. This applies the use of a fully coupled finite element method in 2-D using linear elastic fracture mechanics which is then expanded using displacement discontinuity to a cohesive zone model in 3-D. A static boundary element model was also used to determine stress fields surrounding static, predetermined fracture geometries. These models have been verified against analytical solutions for simple cases and are now being applied to more detailed case studies and analysis. These models have been briefly discussed throughout this thesis in order to give insight on their current capabilities and application as well as their future potential within this area of research. The majority of this work introduces transient stress field prediction to cases of single and multiple hydraulic fractures. The static assessment of these stresses is determined for verification of results to those found in publication which leads into these transient stress field variations. A new method has been developed and applied to the stress state prediction for the first time in a transient fracture model which is partly based upon a critical distance theory. These dynamic interactions can provide useful insight to pertinent issues within the petroleum and natural gas industry such as those to hydraulic fracturing fluid loss and induced seismic events, as well as to applications of efficiency and optimization of the

  14. Parameter estimation of soil hydraulic and thermal property functions for unsaturated porous media using the HYDRUS-2D code

    Directory of Open Access Journals (Sweden)

    Nakhaei Mohammad

    2014-03-01

    Full Text Available Knowledge of soil hydraulic and thermal properties is essential for studies involving the combined effects of soil temperature and water input on water flow and redistribution processes under field conditions. The objective of this study was to estimate the parameters characterizing these properties from a transient water flow and heat transport field experiment. Real-time sensors built by the authors were used to monitor soil temperatures at depths of 40, 80, 120, and 160 cm during a 10-hour long ring infiltration experiment. Water temperatures and cumulative infiltration from a single infiltration ring were monitored simultaneously. The soil hydraulic parameters (the saturated water content θ s, empirical shape parameters α and n, and the saturated hydraulic conductivity Ks and soil thermal conductivity parameters (coefficients b1, b2, and b3 in the thermal conductivity function were estimated from cumulative infiltration and temperature measurements by inversely solving a two-dimensional water flow and heat transport using HYDRUS-2D. Three scenarios with a different, sequentially decreasing number of optimized parameters were considered. In scenario 1, seven parameters (θ s, Ks, α, n, b1, b2, and b3 were included in the inverse problem. The results indicated that this scenario does not provide a unique solution. In scenario 2, six parameters (Ks, α, n, b1, b2, and b3 were included in the inverse problem. The results showed that this scenario also results in a non-unique solution. Only scenario 3, in which five parameters (α, n, b1, b2, and b3 were included in the inverse problem, provided a unique solution. The simulated soil temperatures and cumulative infiltration during the ring infiltration experiment compared reasonably well with their corresponding observed values.

  15. Core Thermal-Hydraulic Design of a Sodium Cooled Fast Reactor for the U/TRU Fuel Modification

    Energy Technology Data Exchange (ETDEWEB)

    Choi, Sun Rock; Cho, Chung Ho; Kim, Young Gyun; Song, Hoon; Park, Won Seok; Kim, Sang Ji [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2011-05-15

    The Korea Atomic energy Research Institute (KAERI) has been developing an advanced SFR design technology with the final goal of constructing a demonstration plant by 2028. The main objective of the SFR demonstration plant is to verify TRU metal fuel performance, large-scale reactor operation, and transmutation ability of high-level wastes. However, in the early stage, the SFR will run on low enriched uranium fuel due to a lack of TRU fuel qualification. After sequential evaluations of the fuel performance, the fissile fuel material will transform from uranium to LTRU (LWR-TRU), and then finally to MTRU (Mixed TRU of LTRU and recycled TRU). At the same time, the core configurations will be modified to meet the nuclear design requirements. Therefore, there is also a strong need to ensure a proper cooling capability during modifications of the entire core. In this work, the core thermal-hydraulic design for U/TRU fuel modification is performed using the SLTHEN (Steady-State LMR Thermal-Hydraulic Analysis Code Based on ENERGY Model) code. As the power distribution in a reactor core is not uniform, it requires a suitable flow allocation to each assembly. There are two ways of allocating the flow rates depending on the orifice positions. The inner officering scheme locates orifice plates in the lower part of the fuel assembly. Therefore, it is possible that the flow distribution is redesigned according to the core configurations. On the other hand, the outer officering scheme fixes orifice plates within the receptacle body throughout the entire plant lifetime. This has the advantage lower of fabrication costs and operating errors but included insufficient design flexibility. This paper provides comparative studies of orifice position for the core thermal-hydraulic design

  16. Thermal-hydraulic analysis of an innovative decay heat removal system for lead-cooled fast reactors

    Energy Technology Data Exchange (ETDEWEB)

    Giannetti, Fabio; Vitale Di Maio, Damiano; Naviglio, Antonio; Caruso, Gianfranco, E-mail: gianfranco.caruso@uniroma1.it

    2016-08-15

    Highlights: • LOOP thermal-hydraulic transient analysis for lead-cooled fast reactors. • Passive decay heat removal system concept to avoid lead freezing. • Solution developed for the diversification of the decay heat removal functions. • RELAP5 vs. RELAP5-3D comparison for lead applications. - Abstract: Improvement of safety requirements in GEN IV reactors needs more reliable safety systems, among which the decay heat removal system (DHR) is one of the most important. Complying with the diversification criteria and based on pure passive and very reliable components, an additional DHR for the ALFRED reactor (Advanced Lead Fast Reactor European Demonstrator) has been proposed and its thermal-hydraulic performances are analyzed. It consists in a coupling of two innovative subsystems: the radiative-based direct heat exchanger (DHX), and the pool heat exchanger (PHX). Preliminary thermal-hydraulic analyses, by using RELAP5 and RELAP5-3D© computer programs, have been carried out showing that the whole system can safely operate, in natural circulation, for a long term. Sensitivity analyses for: the emissivity of the DHX surfaces, the PHX water heat transfer coefficient (HTC) and the lead HTC have been carried out. In addition, the effects of the density variation uncertainty on the results has been analyzed and compared. It allowed to assess the feasibility of the system and to evaluate the acceptable range of the studied parameters. A comparison of the results obtained with RELAP5 and RELAP5-3D© has been carried out and the analysis of the differences of the two codes for lead is presented. The features of the innovative DHR allow to match the decay heat removal performance with the trend of the reactor decay heat power after shutdown, minimizing at the same time the risk of lead freezing. This system, proposed for the diversification of the DHR in the LFRs, could be applicable in the other pool-type liquid metal fast reactors.

  17. 3D neutronic codes coupled with thermal-hydraulic system codes for PWR, and BWR and VVER reactors

    Energy Technology Data Exchange (ETDEWEB)

    Langenbuch, S.; Velkov, K. [GRS, Garching (Germany); Lizorkin, M. [Kurchatov-Institute, Moscow (Russian Federation)] [and others

    1997-07-01

    This paper describes the objectives of code development for coupling 3D neutronics codes with thermal-hydraulic system codes. The present status of coupling ATHLET with three 3D neutronics codes for VVER- and LWR-reactors is presented. After describing the basic features of the 3D neutronic codes BIPR-8 from Kurchatov-Institute, DYN3D from Research Center Rossendorf and QUABOX/CUBBOX from GRS, first applications of coupled codes for different transient and accident scenarios are presented. The need of further investigations is discussed.

  18. Calibration of hydrodynamic behavior and biokinetics for TOC removal modeling in biofilm reactors under different hydraulic conditions.

    Science.gov (United States)

    Zeng, Ming; Soric, Audrey; Roche, Nicolas

    2013-09-01

    In this study, total organic carbon (TOC) biodegradation was simulated by GPS-X software in biofilm reactors with carriers of plastic rings and glass beads under different hydraulic conditions. Hydrodynamic model by retention time distribution and biokinetic measurement by in-situ batch test served as two significant parts of model calibration. Experimental results showed that TOC removal efficiency was stable in both media due to the enough height of column, although the actual hydraulic volume changed during the variation of hydraulic condition. Simulated TOC removal efficiencies were close to experimental ones with low theil inequality coefficient values (below 0.15). Compared with glass beads, more TOC was removed in the filter with plastic rings due to the larger actual hydraulic volume and lower half saturation coefficient in spite of its lower maximum specific growth rate of biofilm, which highlighted the importance of calibrating hydrodynamic behavior and biokinetics. Copyright © 2013 Elsevier Ltd. All rights reserved.

  19. Thermal-hydraulic analysis of heat transfer in subchannels of the European high performance supercritical Water-Cooled Reactor for different CFD turbulence models

    Energy Technology Data Exchange (ETDEWEB)

    Castro, Landy Y.; Rojas, Leorlen Y.; Gamez, Abel; Rosales, Jesus; Gonzalez, Daniel; Garcia, Carlos, E-mail: lcastro@instec.cu, E-mail: leored1984@gmail.com, E-mail: agamezgmf@gmail.com, E-mail: jrosales@instec.cu, E-mail: danielgonro@gmail.com, E-mail: cgh@instec.cu [Instituto Superior de Tecnologias y Ciencias Aplicadas (InSTEC), La Habana (Cuba); Oliveira, Carlos Brayner de, E-mail: cabol@ufpe.br [Universidade Federal de Pernambuco (UFPE), Recife, PE (Brazil); Dominguez, Dany S., E-mail: dsdominguez@gmail.com [Universidade Estadual de Santa Cruz (UESC), Ilheus, BA (Brazil). Pos-Graduacao em Modelagem Computacional

    2015-07-01

    Chosen as one of six Generation‒IV nuclear-reactor concepts, Supercritical Water-cooled Reactors (SCWRs) are expected to have high thermal efficiencies within the range of 45 - 50% owing to the reactor's high pressures and outlet temperatures. In this reactor, the primary water enters the core under supercritical-pressure condition (25 MPa) at a temperature of 280 deg C and leaves it at a temperature of up to 510 deg C. Due to the significant changes in the physical properties of water at supercritical-pressure, the system is susceptible to local temperature, density and power oscillations. The behavior of supercritical water into the core of the SCWR, need to be sufficiently studied. Most of the methods available to predict the effects of the heat transfer phenomena within the pseudocritical region are based on empirical one-directional correlations, which do not capture the multidimensional effects and do not provide accurate results in regions such as the deteriorated heat transfer regime. In this paper, computational fluid dynamics (CFD) analysis was carried out to study the thermal-hydraulic behavior of supercritical water flows in sub-channels of a typical European High Performance Light Water Reactor (HPLWR) fuel assembly using commercial CFD code CFX-14. It was determined the steady-state equilibrium parameters and calculated the temperature and density distributions. A comparative study for different turbulence models were carried out and the obtained results are discussed. (author)

  20. Hydraulic and mechanical behavior of landfill clay liner containing SSA in contact with leachate.

    Science.gov (United States)

    Zhang, Qian; Lu, Haijun; Liu, Junzhu; Wang, Weiwei; Zhang, Xiong

    2017-05-24

    Sewage sludge ash (SSA) produced by municipal sludge can be used as a modified additive for clay liner, and improves the working performance of landfill clay liner in contact with leachate. Under the action of landfill leachate, the permeability, shear strength, phase composition, and pore structure of the modified clay are investigated through the flexible wall permeability test, triaxial shear test, thermal gravimetric and differential thermal analysis, and low-temperature nitrogen adsorption test, respectively. The hydraulic conductivity of the modified clay containing 0-5% SSA is in the range of 3.94 × 10 -8 -1.16 × 10 -7  cm/s, and the pollutant concentration of the sample without SSA was higher than others. The shear strength of the modified clay is more than that of the traditional clay liner, the cohesion rate of modified clay increases from 32.5 to 199.91 kPa, and the internal friction angle decreases from 32.5° to 15.6°. Furthermore, the weight loss rates of the samples are 15.69%, 17.92%, 18.06%, and 20.68%, respectively, when the SSA content increases from 0% to 5%. The total pore volume and average pore diameter of the modified clay decrease with the increase in the SSA content, respectively. However, the specific area of the modified clay increases with the increase in the SSA content.

  1. Thermal and hydraulic characteristics of the JEN-1 Reactor; Caracteristicas hidraulicas y termicas del Reactor JEN-1

    Energy Technology Data Exchange (ETDEWEB)

    Otra Otra, F.; Leira Rey, G.

    1971-07-01

    In this report an analysis is made of the thermal and hydraulic performances of the JEN-1 reactor operating steadily at 3 Mw of thermal power. The analysis is made separately for the core, main heat exchanger and cooling tower. A portion of the report is devoted to predict the performances of these three main components when and if the reactor was going to operate at a power higher than the maximum 3 Mw attainable today. Finally an study is made of the unsteady operation of the reactor, focusing the attention towards the pumping characteristics and the temperatures obtained in the fuel elements. Reference is made to several digital calculation programmes that nave been developed for such purpose. (Author) 21 refs.

  2. Transient analysis in the 3D nodal kinetics and thermal-hydraulics ANDES/COBRA coupled system

    Energy Technology Data Exchange (ETDEWEB)

    Lozano, Juan Andres; Aragones, Jose Maria; Garcia-Herranz, Nuria [Universidad Politecnica de Madrid, Madrid (Spain)

    2008-07-01

    Neutron kinetics has been implemented in the 3D nodal solver ANDES, which has been coupled to the core thermal-hydraulics (TH) code COBRA-III for core transient analysis. The purpose of this work is, first, to discuss and test the ability of the kinetics solver ANDES to model transients; and second, by means of a systematic analysis, including alternate kinetics schemes, time step size, nodal size, neutron energy groups and spectrum, to serve as a basis for the development of more accurate and efficient neutronics/thermal-hydraulics tools for general transient simulations. The PWR MOX/UO{sub 2} transient benchmark provided by the OECD/NEA and US NRC was selected for these goals. The obtained ANDES/COBRA-III results were consistent with other solutions to the benchmark; the differences in the TH feedback led to slight differences in the core power evolution, whereas very good agreements were found in the other requested parameters. The performed systematic analysis highlighted the optimum kinetics iterative scheme, and showed that neutronics spatial discretization effects have stronger influence than time discretization effects, in the semi-implicit scheme adopted, on the numerical solution. On the other hand, the number of energy groups has an important influence on the transient evolution, whereas the assumption of using the prompt neutron spectrum for delayed neutrons is acceptable as it leads to small relative errors. (authors)

  3. Neutronics - thermal-hydraulics coupling: application to the helium-cooled fast reactor; Couplage neutronique - thermohydraulique: application au reacteur a neutrons rapides refroidi a l'helium

    Energy Technology Data Exchange (ETDEWEB)

    Vaiana, F.

    2009-11-15

    This thesis focuses on the study of interactions between neutron-kinetics and thermal-hydraulics. Neutron-kinetics allow to calculate the power in a nuclear reactor and the temperature evolution of materials where this power is deposited is known thanks to thermal-hydraulics. Moreover, when the temperatures evolve, the densities and cross sections change. These two disciplines are thus coupled. The first part of this work corresponds to the study and development of a method which allows to simulate transients in nuclear reactors and especially with a Monte-Carlo code for neutron-kinetics. An algorithm for the resolution of the neutron transport equation has been established and validated with a benchmark. In thermal-hydraulics, a porous media approach, based on another thesis, is considered. This gives the opportunity to solve the equations on the whole core without unconscionable computation time. Finally, a theoretical study has been performed on the statistical uncertainties which result from the use of a Monte-Carlo code and which spread from the reactivity to the power and from the power to the temperatures. The second part deals with the study of a misplaced control rod withdrawing in a GFR (helium-cooled fast reactor), a fourth generation reactor. Some models allowing to calculate neutron-kinetics and thermal-hydraulics in the core (which contains assemblies built up with fuel plates) were defined. In thermal-hydraulics, a model for the core based on the porous media approach and a fuel plate homogenization model have been set up. A similar homogenization model has been studied for neutron-kinetics. Finally, the control rod withdrawing transient where we can observe the power raising and the stabilisation by thermal feedback has been performed with the Monte-Carlo code Tripoli for neutron-kinetics and the code Trio-U for thermal-hydraulics. (author)

  4. Oxidation and thermal behavior of Jatropha curcas biodiesel ...

    African Journals Online (AJOL)

    Oxidation and thermal behavior of Jatropha curcas biodiesel influenced by antioxidants and metal contaminants. ... International Journal of Engineering, Science and Technology. Journal Home ... The present paper is dealing with the study of oxidation and thermal behavior of JCB with respect to different metal contents.

  5. Development of a coupled neutronic/thermal-hydraulic tool with multi-scale capabilities and applications to HPLWR core analysis

    Energy Technology Data Exchange (ETDEWEB)

    Monti, Lanfranco, E-mail: lanfranco.monti@gmail.co [Karlsruhe Institute of Technology (KIT), Institute for Nuclear and Energy Technologies (IKET), Hermann-von-Helmholtzplatz 1, 76344 Eggenstein-Leopoldshafen (Germany); Starflinger, Joerg, E-mail: joerg.starflinger@ike.uni-stuttgart.d [Universitaet Stuttgart, Institut fuer Kernenergetik und Energiesysteme, Pfaffenwaldring 31, 70569 Stuttgart (Germany); Schulenberg, Thomas, E-mail: schulenberg@kit.ed [Karlsruhe Institute of Technology (KIT), Institute for Nuclear and Energy Technologies (IKET), Hermann-von-Helmholtzplatz 1, 76344 Eggenstein-Leopoldshafen (Germany)

    2011-05-15

    Highlights: Advanced analysis and design techniques for innovative reactors are addressed. Detailed investigation of a 3 pass core design with a multi-physics-scales tool. Coupled 40-group neutron transport/equivalent channels TH core analyses methods. Multi-scale capabilities: from equivalent channels to sub-channel pin-by-pin study. High fidelity approach: reduction of conservatism involved in core simulations. - Abstract: The High Performance Light Water Reactor (HPLWR) is a thermal spectrum nuclear reactor cooled and moderated with light water operated at supercritical pressure. It is an innovative reactor concept, which requires developing and applying advanced analysis tools as described in the paper. The relevant water density reduction associated with the heat-up, together with the multi-pass core design, results in a pronounced coupling between neutronic and thermal-hydraulic analyses, which takes into account the strong natural influence of the in-core distribution of power generation and water properties. The neutron flux gradients within the multi-pass core, together with the pronounced dependence of water properties on the temperature, require to consider a fine spatial resolution in which the individual fuel pins are resolved to provide precise evaluation of the clad temperature, currently considered as one of the crucial design criteria. These goals have been achieved considering an advanced analysis method based on the usage of existing codes which have been coupled with developed interfaces. Initially neutronic and thermal-hydraulic full core calculations have been iterated until a consistent solution is found to determine the steady state full power condition of the HPLWR core. Results of few group neutronic analyses might be less reliable in case of HPLWR 3-pass core than for conventional LWRs because of considerable changes of the neutron spectrum within the core, hence 40 groups transport theory has been preferred to the usual 2 groups

  6. Numerical simulation of the LAGEOS thermal behavior and thermal accelerations

    NARCIS (Netherlands)

    Andrés, J.I.; Noomen, R.; Vecellio None, S.

    2006-01-01

    The temperature distribution throughout the LAGEOS satellites is simulated numerically with the objective to determine the resulting thermal force. The different elements and materials comprising the spacecraft, with their energy transfer, have been modeled with unprecedented detail. The radiation

  7. Thermal-hydraulic performance of a water-cooled tungsten-rod target for a spallation neutron source

    Energy Technology Data Exchange (ETDEWEB)

    Poston, D.I.

    1997-08-01

    A thermal-hydraulic (T-H) analysis is conducted to determine the feasibility and limitations of a water-cooled tungsten-rod target at powers of 1 MW and above. The target evaluated has a 10-cm x 10-cm cross section perpendicular to the beam axis, which is typical of an experimental spallation neutron source - both for a short-pulse spallation source and long-pulse spallation source. This report describes the T-H model and assumptions that are used to evaluate the target. A 1-MW baseline target is examined, and the results indicate that this target should easily handle the T-H requirements. The possibility of operating at powers >1 MW is also examined. The T-H design is limited by the condition that the coolant does not boil (actual limits are on surface subcooling and wall heat flux); material temperature limits are not approached. Three possible methods of enhancing the target power capability are presented: reducing peak power density, altering pin dimensions, and improving coolant conditions (pressure and temperature). Based on simple calculations, it appears that this target concept should have little trouble reaching the 2-MW range (from a purely T-H standpoint), and possibly much higher powers. However, one must keep in mind that these conclusions are based solely on thermal-hydraulics. It is possible, and perhaps likely, that target performance could be limited by structural issues at higher powers, particularly for a short-pulse spallation source because of thermal shock issues.

  8. Steady-state thermal hydraulic analysis and flow channel blockage accident analysis of JRR-4 silicide LEU core

    Energy Technology Data Exchange (ETDEWEB)

    Kaminaga, Masanori; Yamamoto, Kazuyoshi; Watanabe, Shukichi [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    1996-09-01

    JRR-4 is a light water moderated and cooled, graphite reflected pool type research reactor using high enriched uranium (HEU) plate-type fuels. Its thermal power is 3.5 MW. The core conversion program from HEU fuel to uranium-silicon-aluminum (U{sub 3}Si{sub 2}-Al) dispersion type fuel (Silicide fuel) with low enriched uranium (LEU) is currently conducted at the JRR-4. This report describes about the steady-state thermal hydraulic analysis results and the flow channel blockage accident analysis result. In JRR-4, there are two operation mode. One is high power operation mode up to 3.5 MW, under forced convection cooling using the primary and the secondary cooling systems. The other is low power operation mode up to 200 kW, under natural circulation cooling between the reactor core and the reactor pool without the primary and the secondary cooling systems. For the analysis of the high power operation mode under forced convection cooling and the flow channel blockage accident, COOLOD code was used. On the other hand, for the analysis of low power operation under natural convection cooling, COOLOD-N2 code was used. From steady-state thermal hydraulic analysis results of both forced and natural convection cooling, fuel temperature, minimum DNBR etc. meet the design criteria and JRR-4 LEU silicide core has enough safety margin under normal operation conditions. Furthermore, flow channel blockage accident analysis results show that one channel flow blockage accident meet the safety criteria for accident conditions which have been established for JRR-4 LEU silicide core. (author)

  9. Thermal fluid mixing behavior during medium break LOCA in evaluation of pressurized thermal shock

    Energy Technology Data Exchange (ETDEWEB)

    Jung, Jae Won; Bang, Young Seok; Seul, Kwang Won; Kim, Hho Jung [Korea Institute of Nuclear Safety, Taejon (Korea, Republic of)

    1998-12-31

    Thermal fluid mixing behavior during a postulated medium-size hot leg break loss of coolant accident is analyzed for the international comparative assessment study on pressurized thermal shock (PTS-ICAS) proposed by OECD-NEA. The applicability of RELAP5 code to analyze the thermal fluid mixing behavior is evaluated through a simple modeling relevant to the problem constraints. Based on the calculation result, the onset of thermal stratification is investigated using Theofanous`s empirical correlation. Sensitivity calculations using a fine node model and crossflow model are also performed to evaluate the modeling capability on multi-dimensional characteristics related to thermal fluid mixing. 6 refs., 8 figs. (Author)

  10. Impact of mineral fouling on hydraulic behavior of permeable reactive barriers.

    Science.gov (United States)

    Lin, Li; Benson, Craig H; Lawson, Elizabeth M

    2005-01-01

    This paper describes reactive transport simulations conducted to assess the impact of mineral fouling on the hydraulic behavior of continuous-wall permeable reactive barriers (PRBs) employing granular zero-valent iron (ZVI) in carbonate-rich alluvial aquifers. The reactive transport model included a geochemical algorithm for simulating corrosion and mineral precipitation reactions that have been observed in ZVI PRBs. Results of simulations show that porosity and hydraulic conductivity of the ZVI decrease over time and that flows are redistributed throughout the PRB in response to fouling of the pore space. Under typical conditions, only subtle changes occur within the first 10 years (i.e., duration of the current field experience record with PRBs), and the most significant changes do not occur until the PRB has operated for at least 30 years. However, changes can occur sooner (or later) if the rate at which mineral-forming ions are delivered to the PRB is higher (or lower) than that expected under typical conditions (i.e., due to higher/lower flow rate or inflowing ground water that has higher/lower ionic strength). When the PRB is more permeable than the aquifer, the median Darcy flux in the PRB does not change appreciably over time because the aquifer controls the rate of flow through the PRB. However, seepage velocities in the PRB increase, and residence times decrease, due to porosity reductions caused by accumulation of minerals in the pore space. When fouling becomes extensive, bypassing and reductions in flow rate in the PRB occur.

  11. Conceptual Thermal Hydraulic Design of a 20MW Multipurpose Research Reactor (KAERI/VAEC joint study on a new research reactor for Vietnam)

    Energy Technology Data Exchange (ETDEWEB)

    Chae, Hee Taek; Seo, Chul Gyo; Park, Jong Hark; Park, Cheol [Kaeri, Daejeon (Korea, Republic of); Vinh, Le Vinh; Nghiem, Huynh Ton; Dang, Vo Doan Hai [Dalat Nuclear Research Reactor, Hanoi (Viet Nam)

    2007-08-15

    The conceptual thermal hydraulics design analyses for the 20 MW reference AHR core have been jointly performed by the KAERI and DNRI(VAEC). The preliminary core thermal hydraulic characteristics and safety margins for the AHR core were studied for various core flow rates, fuel assembly powers and core inlet temperatures. Statistical method was applied to the thermal hydraulic design of the reactor core. The MATRA{sub h} subchannel code has been applied to evaluate the thermal hydraulic performances of the AHR and the resulting thermal margins of the core under the forced convection cooling mode during a nominal power operation and the natural circulation mode during a reactor shutdown condition. In addition, typical accident analyses were carried out for a loss of flow accident by a primary pump seizure and a reactivity induced accident by a CAR rod withdrawal during a normal full power operation. The normal full power operation of the AHR was ensured with a sufficient safety margin for the onset of nucleate boiling phenomena. The AHR also had a sufficient natural circulation cooling capability to cool the core without the onset of nucleate boiling in the channel after a normal reactor shutdown and the anticipated transients. It was confirmed by the typical accident analyses that the AHR core was sufficiently protected from the loss of flow by the primary cooling pump seizure and the overpower transients by the CAR withdrawal from the MCHFR and fuel temperature points of view.

  12. Progress of thermal hydraulic evaluation methods and experimental studies on a sodium-cooled fast reactor and its safety in Japan

    Energy Technology Data Exchange (ETDEWEB)

    Kamide, Hideki, E-mail: kamide.hideki@jaea.go.jp; Ohshima, Hiroyuki, E-mail: ohshima.hiroyuki@jaea.go.jp; Sakai, Takaaki, E-mail: sakai.takaaki@jaea.go.jp; Tanaka, Masaaki, E-mail: tanaka.masaaki@jaea.go.jp

    2017-02-15

    Highlights: • Thermal hydraulic issues for safety design criteria of sodium cooled fast reactors. • Measurement of velocity data in a subchannel surrounded by wire wrapped fuel-pins. • Statistical evaluation of core hot spot temperature during natural circulation. • Simulation of dynamics of molten fuel pool in a core disruptive accident. • V&V procedure of a multi-dimensional thermal hydraulic code on thermal striping. - Abstract: In the framework of the Generation-IV International Forum, the safety design criteria (SDC) incorporating safety-related R&D results on innovative technologies and lessons learned from Fukushima Dai-ichi nuclear power plants accident has been established to provide the set of general criteria for the safety designs of structures, systems and components of Generation-IV Sodium-cooled Fast Reactors (Gen-IV SFRs). A number of thermal-hydraulic evaluations are necessary to meet the concept of the criteria in the design studies of Gen-IV SFRs. This paper focuses on four kinds of thermal-hydraulic issues associated with the SDC, i.e., fuel subassembly thermal-hydraulics, natural circulation decay heat removal, core disruptive accidents, and thermal striping. Progress of evaluation methods on these issues is shown with activities on verification and validation (V&V) and experimental studies towards commercialization of SFR in Japan. These evaluation methods are planned to be eventually integrated into a comprehensive numerical simulation system that can be applied to all possible phenomena in SFR systems and that can be expected to become an effective tool for the development of human resource and the handing our knowledge and technologies down.

  13. Effects of Geometrical Clearances, Supports Friction, and Wear Rings on Hydraulic Actuators Bending Behavior

    Directory of Open Access Journals (Sweden)

    Sergio Baragetti

    2016-01-01

    Full Text Available Hydraulic actuators are commonly adopted in machines and structures to provide translating forces with significant magnitudes. Although their application dates back to the industrial revolution, their bending behavior under compression is typically addressed by simple Euler’s instability analysis on the rod, neglecting effects such as the cylinder inertia and stiffness, the presence of contact elements in the cylinder-rod junction and on the piston, geometrical misalignments and imperfections, and friction moments at the support. Such simplifications lead to unjustified reduced critical load calculations on the component. In the present paper, a complete mathematical formulation, which accounts for such effects, is presented and validated against experimental data. A numerical sensitivity analysis is conducted, to assess the contributions of initial rectilinear imperfections, wear rings stiffness and dimension, and supports friction on the actuator’s limit buckling load and bending behavior under compression. Results are presented, including the effect of the cited parameters on the buckling load, providing a reliable tool for the mechanical designer. In particular, an optimum position for the wear ring distance is found. Moreover, increased wear ring stiffness and reduced imperfections increase the buckling load and reduce the bending stresses before the critical load.

  14. International benchmark study of advanced thermal hydraulic safety analysis codes against measurements on IEA-R1 research reactor

    Energy Technology Data Exchange (ETDEWEB)

    Hainoun, A., E-mail: pscientific2@aec.org.sy [Atomic Energy Commission of Syria (AECS), Nuclear Engineering Department, P.O. Box 6091, Damascus (Syrian Arab Republic); Doval, A. [Nuclear Engineering Department, Av. Cmdt. Luis Piedrabuena 4950, C.P. 8400 S.C de Bariloche, Rio Negro (Argentina); Umbehaun, P. [Centro de Engenharia Nuclear – CEN, IPEN-CNEN/SP, Av. Lineu Prestes 2242-Cidade Universitaria, CEP-05508-000 São Paulo, SP (Brazil); Chatzidakis, S. [School of Nuclear Engineering, Purdue University, West Lafayette, IN 47907 (United States); Ghazi, N. [Atomic Energy Commission of Syria (AECS), Nuclear Engineering Department, P.O. Box 6091, Damascus (Syrian Arab Republic); Park, S. [Research Reactor Design and Engineering Division, Basic Science Project Operation Dept., Korea Atomic Energy Research Institute (Korea, Republic of); Mladin, M. [Institute for Nuclear Research, Campului Street No. 1, P.O. Box 78, 115400 Mioveni, Arges (Romania); Shokr, A. [Division of Nuclear Installation Safety, Research Reactor Safety Section, International Atomic Energy Agency, A-1400 Vienna (Austria)

    2014-12-15

    Highlights: • A set of advanced system thermal hydraulic codes are benchmarked against IFA of IEA-R1. • Comparative safety analysis of IEA-R1 reactor during LOFA by 7 working teams. • This work covers both experimental and calculation effort and presents new out findings on TH of RR that have not been reported before. • LOFA results discrepancies from 7% to 20% for coolant and peak clad temperatures are predicted conservatively. - Abstract: In the framework of the IAEA Coordination Research Project on “Innovative methods in research reactor analysis: Benchmark against experimental data on neutronics and thermal hydraulic computational methods and tools for operation and safety analysis of research reactors” the Brazilian research reactor IEA-R1 has been selected as reference facility to perform benchmark calculations for a set of thermal hydraulic codes being widely used by international teams in the field of research reactor (RR) deterministic safety analysis. The goal of the conducted benchmark is to demonstrate the application of innovative reactor analysis tools in the research reactor community, validation of the applied codes and application of the validated codes to perform comprehensive safety analysis of RR. The IEA-R1 is equipped with an Instrumented Fuel Assembly (IFA) which provided measurements for normal operation and loss of flow transient. The measurements comprised coolant and cladding temperatures, reactor power and flow rate. Temperatures are measured at three different radial and axial positions of IFA summing up to 12 measuring points in addition to the coolant inlet and outlet temperatures. The considered benchmark deals with the loss of reactor flow and the subsequent flow reversal from downward forced to upward natural circulation and presents therefore relevant phenomena for the RR safety analysis. The benchmark calculations were performed independently by the participating teams using different thermal hydraulic and safety

  15. LMFBR in-core thermal-hydraulics: the state of the art and US research and development needs

    Energy Technology Data Exchange (ETDEWEB)

    Khan, E.U.

    1980-04-01

    A detailed critical review is presented of the literature relevant to predicting coolant flow and temperature fields in LMFBR core assemblies for nominal and non-nominal rod bundle geometries and reactor operating conditions. The review covers existing thermal-hydraulic models, computational methods, and experimental data useful for the design of an LMFBR core. The literature search made for this review included publications listed by Nuclear Science Abstracts and Energy Data Base as well as papers presented at key nuclear conferences. Based on this extensive review, the report discusses the accuracy with which the models predict flow and temperature fields in rod assemblies, identifying areas where analytical, experimental, and model development needs exist.

  16. Thermal-hydraulic processes involved in loss of residual heat removal during reduced inventory operation. Revision 1

    Energy Technology Data Exchange (ETDEWEB)

    Fletcher, C.D.; McHugh, P.R.; Naff, S.A.; Johnsen, G.W.

    1991-02-01

    This paper identifies the topics needed to understand pressurized water reactor response to an extended loss of residual heat removal event during refueling and maintenance outages. By identifying the possible plant conditions and cooling methods that would be used for each cooling mode, the controlling thermal-hydraulic processes and phenomena were identified. Controlling processes and phenomena include: gravity drain, core water boil-off, and reflux cooling processes. Important subcategories of the reflux cooling processes include: the initiation of reflux cooling from various plant conditions, the effects of air on reflux cooling, core level depression effects, issues regarding the steam generator secondaries, and the special case of boiler-condenser cooling with once-through steam generators. 25 refs., 6 figs., 1 tab.

  17. RELAP5-3D thermal hydraulic analysis of the target cooling system in the SPES experimental facility

    Science.gov (United States)

    Giardina, M.; Castiglia, F.; Buffa, P.; Palermo, G.; Prete, G.

    2014-11-01

    The SPES (Selective Production of Exotic Species) experimental facility, under construction at the Italian National Institute of Nuclear Physics (INFN) Laboratories of Legnaro, Italy, is a second generation Isotope Separation On Line (ISOL) plant for advanced nuclear physic studies. The UCx target-ion source system works at temperature of about 2273 K, producing a high level of radiation (105 Sv/h), for this reason a careful risk analysis for the target chamber is among the major safety issues. In this paper, the obtained results of thermofluid-dynamics simulations of accidental transients in the SPES target cooling system are reported. The analysis, performed by using the RELAP5-3D 2.4.2 qualified thermal-hydraulic system code, proves good safety performance of this system during different accidental conditions.

  18. Molybdenum-99-producing 37-element fuel bundle neutronically and thermal-hydraulically equivalent to a standard CANDU fuel bundle

    Energy Technology Data Exchange (ETDEWEB)

    Nichita, E., E-mail: Eleodor.Nichita@uoit.ca; Haroon, J., E-mail: Jawad.Haroon@uoit.ca

    2016-10-15

    Highlights: • A 37-element fuel bundle modified for {sup 99}Mo production in CANDU reactors is presented. • The modified bundle is neutronically and thermal-hydraulically equivalent to the standard bundle. • The modified bundle satisfies all safety criteria satisfied by the standard bundle. - Abstract: {sup 99m}Tc, the most commonly used radioisotope in diagnostic nuclear medicine, results from the radioactive decay of {sup 99}Mo which is currently being produced at various research reactors around the globe. In this study, the potential use of CANDU power reactors for the production of {sup 99}Mo is investigated. A modified 37-element fuel bundle, suitable for the production of {sup 99}Mo in existing CANDU-type reactors is proposed. The new bundle is specifically designed to be neutronically and thermal-hydraulically equivalent to the standard 37-element CANDU fuel bundle in normal, steady-state operation and, at the same time, be able to produce significant quantities of {sup 99}Mo when irradiated in a CANDU reactor. The proposed bundle design uses fuel pins consisting of a depleted-uranium centre surrounded by a thin layer of low-enriched uranium. The new molybdenum-producing bundle is analyzed using the lattice transport code DRAGON and the diffusion code DONJON. The proposed design is shown to produce 4081 six-day Curies of {sup 99}Mo activity per bundle when irradiated in the peak-power channel of a CANDU core, while maintaining the necessary reactivity and power rating limits. The calculated {sup 99}Mo yield corresponds to approximately one third of the world weekly demand. A production rate of ∼3 bundles per week can meet the global demand of {sup 99}Mo.

  19. Transient thermal, hydraulic, and mechanical analysis of a counter flow offset strip fin intermediate heat exchanger using an effective porous media approach

    Science.gov (United States)

    Urquiza, Eugenio

    This work presents a comprehensive thermal hydraulic analysis of a compact heat exchanger using offset strip fins. The thermal hydraulics analysis in this work is followed by a finite element analysis (FEA) to predict the mechanical stresses experienced by an intermediate heat exchanger (IHX) during steady-state operation and selected flow transients. In particular, the scenario analyzed involves a gas-to-liquid IHX operating between high pressure helium and liquid or molten salt. In order to estimate the stresses in compact heat exchangers a comprehensive thermal and hydraulic analysis is needed. Compact heat exchangers require very small flow channels and fins to achieve high heat transfer rates and thermal effectiveness. However, studying such small features computationally contributes little to the understanding of component level phenomena and requires prohibitive computational effort using computational fluid dynamics (CFD). To address this issue, the analysis developed here uses an effective porous media (EPM) approach; this greatly reduces the computation time and produces results with the appropriate resolution [1]. This EPM fluid dynamics and heat transfer computational code has been named the Compact Heat Exchanger Explicit Thermal and Hydraulics (CHEETAH) code. CHEETAH solves for the two-dimensional steady-state and transient temperature and flow distributions in the IHX including the complicating effects of temperature-dependent fluid thermo-physical properties. Temperature- and pressure-dependent fluid properties are evaluated by CHEETAH and the thermal effectiveness of the IHX is also calculated. Furthermore, the temperature distribution can then be imported into a finite element analysis (FEA) code for mechanical stress analysis using the EPM methods developed earlier by the University of California, Berkeley, for global and local stress analysis [2]. These simulation tools will also allow the heat exchanger design to be improved through an

  20. Hydraulic Behavior and Chemical Characterization of Lapilli as Material for Natural Filtering of Slurry

    Directory of Open Access Journals (Sweden)

    Nereida Falcón-Cardona

    2015-06-01

    Full Text Available Livestock effluents are a beneficial nutrient supply for crops, whereby their use is critical to ensure the sustainability of the farms global management. However, they can cause serious ecological problems if misused, polluting soils and groundwater. Combining “soft technology” and local materials is a low cost solution in terms of finance and energy. The REAGUA project (REuso AGUA, Water reuse in Spanish analyzes the possibility of using “picon” (lapilli as a material for the treatment of liquid manure from ruminants, for later use in subsurface drip irrigation system to produce forage and biofuels, in which the soil acts as a subsequent advanced treatment. A three-phase system, in which the effluent was poured with a vertical subsurface flow in an unsaturated medium, is designed. In order to determine the management conditions that optimize the filter, it was necessary to characterize the hydraulic behavior of lapilli and its ability to remove substances. Using three lapilli-filled columns, unsaturated flux, and a ruminant effluent, the reduction of chemical oxygen demand (COD, biochemical oxygen demand after 5 days (BOD5 and ammonia, phosphorus and suspension solids (SS obtained was over 80%, 90%, and 95% respectively, assumable values for irrigation.

  1. Neutronic and thermal-hydraulic analysis of new irradiation channels inside the Moroccan TRIGA Mark II research reactor core.

    Science.gov (United States)

    Chham, E; El Bardouni, T; Benaalilou, K; Boukhal, H; El Bakkari, B; Boulaich, Y; El Younoussi, C; Nacir, B

    2016-10-01

    This study was conducted to improve the capacity of radioisotope production in the Moroccan TRIGA Mark II research reactor, which is considered as one of the most important applications of research reactors. The aim of this study is to enhance the utilization of TRIGA core in the field of neutron activation and ensure an economic use of the fuel. The main idea was to create an additional irradiation channel (IC) inside the core. For this purpose, three new core configurations are proposed, which differ according to the IC position in the core. Thermal neutron flux distribution and other neutronic safety parameters such as power peaking factors, excess reactivity, and control rods worth reactivity were calculated using the Monte Carlo N-Particle Transport (MCNP) code and neutron cross-section library based on ENDF/B-VII evaluation. The calculated thermal flux in the central thimble (CT) and in the added IC for the reconfigured core is compared with the thermal flux in the CT of the existing core, which is taken as a reference. The results show that all the obtained fluxes in CTs are very close to the reference value, while a remarkable difference is observed between the fluxes in the new ICs and reference. This difference depends on the position of IC in the reactor core. To demonstrate that the Moroccan TRIGA reactor could safely operate at 2MW, with new configurations based on new ICs, different safety-related thermal-hydraulic parameters were investigated. The PARET model was used in this study to verify whether the safety margins are met despite the new modifications of the core. The results show that it is possible to introduce new ICs safely in the reactor core, because the obtained values of the parameters are largely far from compromising the safety of the reactor. Copyright © 2016 Elsevier Ltd. All rights reserved.

  2. Coupling agents in therapeutic ultrasound: Acoustic and thermal behavior

    OpenAIRE

    CASAROTTO, Raquel Aparecida; Adamowski, Julio Cesar; Fallopa, Flávio [UNIFESP; Bacanelli, Fernando

    2004-01-01

    Casarotto RA, Adamowski JC, Fallopa F, Bacanelli F. Coupling agents in therapeutic ultrasound: acoustic and thermal behavior. Arch Phys Med Rehabil 2004;85:162-5.Objective: To compare transmissivity data and thermal behavior of 4 coupling media.Design: Experimental.Setting: Postgraduate rehabilitation program in Brazil.Specimens: Four coupling media: gel, mineral oil, white petrolatum, and degassed water.Interventions: Not applicable.Main Outcome Measures: the transmission, attenuation, refle...

  3. Comparative of the Tribological Performance of Hydraulic Cylinders Coated by the Process of Thermal Spray HVOF and Hard Chrome Plating

    Directory of Open Access Journals (Sweden)

    R.M. Castro

    2014-03-01

    Full Text Available Due to the necessity of obtaining a surface that is resistant to wear and oxidation, hydraulic cylinders are typically coated with hard chrome through the process of electroplating process. However, this type of coating shows an increase of the area to support sealing elements, which interferes directly in the lubrication of the rod, causing damage to the seal components and bringing oil leakage. Another disadvantage in using the electroplated hard chromium process is the presence of high level hexavalent chromium Cr+6 which is not only carcinogenic, but also extremely contaminating to the environment. Currently, the alternative process of high-speed thermal spraying (HVOF - High Velocity Oxy-Fuel, uses composite materials (metal-ceramic possessing low wear rates. Research has shown that some mechanical properties are changed positively with the thermal spray process in industrial applications. It is evident that a coating based on WC has upper characteristics as: wear resistance, low friction coefficient, with respect to hard chrome coatings. These characteristics were analyzed by optical microscopy, roughness measurements and wear test.

  4. Preliminary Thermo-hydraulic Core Design Analysis of Korea Advanced Nuclear Thermal Engine Rocket for Space Application

    Energy Technology Data Exchange (ETDEWEB)

    Nam, Seung Hyun; Lee, Jeong Ik; Chang, Soon Heung [Korea Advanced Institute of Science and Technology, Daejeon (Korea, Republic of)

    2014-05-15

    {sub th} power and electricity generation with 100 kW{sub th} idle power. Consequently, KANUTER has the characteristics of a compact and lightweight system, excellent propellant efficiency, bimodal capability, and mission versatility as indicated in the reference design parameters. This thermo-hydraulic design analysis was carried out to estimate the optimum FWT of the unique SLHC fuel design in the core and thereby the maximum rocket performance. The FWT affects the mechanical strength of the SLHC fuel assembly as well as the thermo-hydraulic capability mainly depending on the heat transfer area of fuel. The thicker fuel wafer is mechanically strong with low pressure drop, while the thinner fuel wafer is thermally robust with less mechanical strength and higher shear stress in the core.

  5. Flexible parallel implicit modelling of coupled thermal-hydraulic-mechanical processes in fractured rocks

    Science.gov (United States)

    Cacace, Mauro; Jacquey, Antoine B.

    2017-09-01

    Theory and numerical implementation describing groundwater flow and the transport of heat and solute mass in fully saturated fractured rocks with elasto-plastic mechanical feedbacks are developed. In our formulation, fractures are considered as being of lower dimension than the hosting deformable porous rock and we consider their hydraulic and mechanical apertures as scaling parameters to ensure continuous exchange of fluid mass and energy within the fracture-solid matrix system. The coupled system of equations is implemented in a new simulator code that makes use of a Galerkin finite-element technique. The code builds on a flexible, object-oriented numerical framework (MOOSE, Multiphysics Object Oriented Simulation Environment) which provides an extensive scalable parallel and implicit coupling to solve for the multiphysics problem. The governing equations of groundwater flow, heat and mass transport, and rock deformation are solved in a weak sense (either by classical Newton-Raphson or by free Jacobian inexact Newton-Krylow schemes) on an underlying unstructured mesh. Nonlinear feedbacks among the active processes are enforced by considering evolving fluid and rock properties depending on the thermo-hydro-mechanical state of the system and the local structure, i.e. degree of connectivity, of the fracture system. A suite of applications is presented to illustrate the flexibility and capability of the new simulator to address problems of increasing complexity and occurring at different spatial (from centimetres to tens of kilometres) and temporal scales (from minutes to hundreds of years).

  6. Thermal-Hydraulic Effect of Pattern of Wire-wrap Spacer in 19-pin Rod Bundle for SFR Fuel Assembly

    Energy Technology Data Exchange (ETDEWEB)

    Jeong, Yeong Shin; Bang, In Cheol [UNIST, Ulsan (Korea, Republic of); Park, Seong Dae [KAERI, Daejeon (Korea, Republic of)

    2016-05-15

    As sodium-cooled fast reactor (SFR) has been considered the most promising reactor type for future and prototype gen-IV SFR has been developed actively in Korea, thermal-hydraulic aspects of the SFR fuel assembly have the important role for the reactor safety analysis. In PGSFR fuel assembly, 271 pins of fuel rods are tightly packed in triangular array inside hexagonal duct, and wire is wound helically per each fuel rod with regular pattern to assure the gap between rods and prevent the collision, which is called wire-wrapped spacer. Due to helical shape of the wire-wrapped spacer, flow inside duct can have stronger turbulent characteristics and thermal mixing effect. However, many studies showed the possible wake from swirl flow inside subchannel, which cause local hot spot. To prevent the wake flow and improve thermal mixing, new pattern of wire wrap spacer was suggested. To evaluate the effect of wire wrap spacer pattern, CFD analysis was performed for 19-pin rod bundle with comparison of conventional and U-pattern wire wrap spacer. To prevent the wake due to same direction of swirl flow, 7-rod unit pattern of wire spacer, which are arranged to have different rotational direction of wire with adjacent rods and center rod without wire wrap was proposed. From simulation results, swirl flow across gap conflicts its rotation direction causing wake flow from the regular pattern of the conventional one, which generates local hot spot near cladding. With U-pattern of wire wrap spacer, heat transfer in subchannel can be enhanced with evenly distributed cross flow without compensating pressure loss. From the results, the pattern of wire wrap spacer can influence the both heat transfer characteristics and pressure drop, with flow structures generated by wire wrap spacer.

  7. Uncertainty and sensitivity analysis for the modeling of transients with interaction of thermal hydraulics and neutron kinetics

    Energy Technology Data Exchange (ETDEWEB)

    Soeren Kliem; Siegfried Mittag [Forschungszentrum Rossendorf (FZR), Institute of Safety Research, P.O.B. 510119, D-01314 Dresden (Germany); Siegfried Langenbuch [Gesellschaft fuer Anlagen- und Reaktorsicherheit mbH, P.O.B. 13 28, D-85748 Garching (Germany)

    2005-07-01

    Full text of publication follows: The transition from the application of conservative models to the use of best-estimate models raises the question about the uncertainty of the obtained results. This question becomes especially important, if the best-estimate models should be used for safety analyses in the field of nuclear engineering. Different methodologies were developed to assess the uncertainty of the calculation results of computer simulation codes. One of them is the methodology developed by Gesellschaft fuer Anlagenund Reaktorsicherheit (GRS) which uses the statistical code package SUSA. In the past, this methodology was applied to the calculation results of the advanced thermal hydraulic system code ATHLET. In the frame of the recently finished EU FP5 funded research project VALCO, that methodology was extended and successfully applied to different coupled code systems, including the uncertainty analysis for neutronics. These code systems consist of a thermal hydraulic system code and a 3D neutron kinetic core model. One of the code systems applied was ATHLET coupled with the Rossendorf kinetics code DYN3D. Two real transients at NPPs with VVER-type reactors documented within the VALCO project were selected for analyses. One was the load drop of one of two turbines to house load level at the Loviisa-1 NPP (VVER-440), the second was a test with the switching-off of one of two main feed water pumps at the VVER-1000 Balakovo-4 NPP. The current paper is dedicated to the different steps of the use and implementation of the GRS methodology to coupled code systems and to the assessment of the results obtained by the DYN3D/ATHLET code. Based on the relevant physical processes in both transients, lists of possible sources of uncertainties were compiled. They are specific for the two transients. Besides control parameters like control rod movement and thermal hydraulic parameters like secondary side pressure, mass flow rates, pressurizer sprayer and heater

  8. Neutronic / thermal-hydraulic coupling with the code system Trace / Parcs; Acoplamiento neutronico / termohidraulico con el sistema de codigos TRACE / PARCS

    Energy Technology Data Exchange (ETDEWEB)

    Mejia S, D. M. [Comision Nacional de Seguridad Nuclear y Salvaguardias, Dr. Barragan 779, Col. Narvarte, 03020 Ciudad de Mexico (Mexico); Del Valle G, E., E-mail: dulcemaria.mejia@cnsns.gob.mx [IPN, Escuela Superior de Fisica y Matematicas, Av. IPN s/n, Col. Lindavista, 07738 Ciudad de Mexico (Mexico)

    2015-09-15

    The developed models for Parcs and Trace codes corresponding for the cycle 15 of the Unit 1 of the Laguna Verde nuclear power plant are described. The first focused to the neutronic simulation and the second to thermal hydraulics. The model developed for Parcs consists of a core of 444 fuel assemblies wrapped in a radial reflective layer and two layers, a superior and another inferior, of axial reflector. The core consists of 27 total axial planes. The model for Trace includes the vessel and its internal components as well as various safety systems. The coupling between the two codes is through two maps that allow its intercommunication. Both codes are used in coupled form performing a dynamic simulation that allows obtaining acceptably a stable state from which is carried out the closure of all the main steam isolation valves (MSIVs) followed by the performance of safety relief valves (SRVs) and ECCS. The results for the power and reactivities introduced by the moderator density, the fuel temperature and total temperature are shown. Data are also provided like: the behavior of the pressure in the steam dome, the water level in the downcomer, the flow through the MSIVs and SRVs. The results are explained for the power, the pressure in the steam dome and the water level in the downcomer which show agreement with the actions of the MSIVs, SRVs and ECCS. (Author)

  9. Thermal-Hydraulic Analysis of an Experimental Reactor Cavity Cooling System with Air. Part I: Experiments; Part II: Separate Effects Tests and Modeling

    Energy Technology Data Exchange (ETDEWEB)

    Corradin, Michael [Univ. of Wisconsin, Madison, WI (United States). Dept. of Engineering Physics; Anderson, M. [Univ. of Wisconsin, Madison, WI (United States). Dept. of Engineering Physics; Muci, M. [Univ. of Wisconsin, Madison, WI (United States). Dept. of Engineering Physics; Hassan, Yassin [Texas A & M Univ., College Station, TX (United States); Dominguez, A. [Texas A & M Univ., College Station, TX (United States); Tokuhiro, Akira [Univ. of Idaho, Moscow, ID (United States); Hamman, K. [Univ. of Idaho, Moscow, ID (United States)

    2014-10-15

    This experimental study investigates the thermal hydraulic behavior and the heat removal performance for a scaled Reactor Cavity Cooling System (RCCS) with air. A quarter-scale RCCS facility was designed and built based on a full-scale General Atomics (GA) RCCS design concept for the Modular High Temperature Gas Reactor (MHTGR). The GA RCCS is a passive cooling system that draws in air to use as the cooling fluid to remove heat radiated from the reactor pressure vessel to the air-cooled riser tubes and discharged the heated air into the atmosphere. Scaling laws were used to preserve key aspects and to maintain similarity. The scaled air RCCS facility at UW-Madison is a quarter-scale reduced length experiment housing six riser ducts that represent a 9.5° sector slice of the full-scale GA air RCCS concept. Radiant heaters were used to simulate the heat radiation from the reactor pressure vessel. The maximum power that can be achieved with the radiant heaters is 40 kW with a peak heat flux of 25 kW per meter squared. The quarter-scale RCCS was run under different heat loading cases and operated successfully. Instabilities were observed in some experiments in which one of the two exhaust ducts experienced a flow reversal for a period of time. The data and analysis presented show that the RCCS has promising potential to be a decay heat removal system during an accident scenario.

  10. Meta-analysis: abundance, behavior, and hydraulic energy shape biotic effects on sediment transport in streams.

    Science.gov (United States)

    Albertson, L K; Allen, D C

    2015-05-01

    An increasing number of studies have emphasized the need to bridge the disciplines of ecology and geomorphology. A large number of case studies show that organisms can affect erosion, but a comprehensive understanding of biological impacts on sediment transport conditions is still lacking. We use meta-analysis to synthesize published data to quantify the effects of the abundance, body size, and behavior of organisms on erosion in streams. We also explore the influence of current velocity, discharge, and sediment grain size on the strength of biotic effects on erosion. We found that species that both increase erosion (destabilizers) and decrease erosion (stabilizers) can alter incipient sediment motion, sediment suspension, and sediment deposition above control conditions in which the organisms were not present. When abundance was directly manipulated, these biotic effects were consistently stronger in the higher abundance treatment, increasing effect sizes by 66%. Per capita effect size and per capita biomass were also consistently positively correlated. Fish and crustaceans were the most studied organisms, but aquatic insects increased the effect size by 550 x compared to other types of organisms after accounting for biomass. In streams with lower discharge and smaller grain sizes, we consistently found stronger biotic effects. Taken collectively, these findings provide synthetic evidence that biology can affect physical processes in streams, and these effects can be mediated by hydraulic energy. We suggest that future studies focus on understudied organisms, such as biofilms, conducting experiments under realistic field conditions, and developing hypotheses for the effect of biology on erosion and velocity currents in the context of restoration to better understand the forces that mediate physical disturbances in stream ecosystems.

  11. Thermal-Hydrologic-Mechanical Behavior of Single Fractures in EGS Reservoirs

    Science.gov (United States)

    Zyvoloski, G.; Kelkar, S.; Yoshioka, K.; Rapaka, S.

    2010-12-01

    Enhanced Geothermal Systems (EGS) rely on the creation a connected fracture system or the enhancement of existing (natural) fractures by hydraulic and chemical treatments. EGS studies at Fenton Hill (New Mexico, USA) and Hijiori (Japan) have revealed that only a limited number of fractures contribute to the effective heat transfer surface area. Thus, the economic viability of EGS depends strongly on the creation and spacing of single fractures in order to efficiently mine heat from given volume of rock. Though there are many similarities between EGS and natural geothermal reservoirs, a major difference between the reservoir types is the (typically) high pumping pressures and induced thermal stresses at the injection wells of an EGS reservoir. These factors can be responsible for fracture dilation/extension and thermal short circuiting and depend strongly on the surrounding state of stress in the reservoir and mechanical properties. We will present results from our study of the thermal-hydrologic-mechanical (THM) behavior of a single fracture in a realistic subsurface stress field. We will show that fracture orientation, the stress environment, fracture permeability structure, and the relationship between permeability changes in a fracture resulting from mechanical displacement are all important when designing and managing an EGS reservoir. Lastly, we present a sensitivity analysis of the important parameters that govern fracture behavior with respect to field measurements. Temperature in high permeability fracture in an EGS reservoir

  12. Development of a thermal hydraulic modelling of ground water of the Malm in the Munich metropolitan area; Entwicklung einer thermisch-hydraulischen Grundwassermodellierung des Malm im Grossraum Muenchen

    Energy Technology Data Exchange (ETDEWEB)

    Dussel, M.; Lueschen, E.; Thomas, R. [Leibniz-Institut fuer Angewandte Geophysik, Hannover (DE)] (and others)

    2011-10-24

    The mutual potential influence of geothermal duplicates and the scientific investigation of the relationship between seismic and hydraulic parameters are investigated in the joint research project 'Geothermal characterization of fractured karst limestone aquifers in the Munich metropolitan area'. Thirteen doublets and triplets being in production or sunk illustrate the great geothermal potential and provide important data on the development of a thermal-hydraulic modeling of the reservoir. 3D seismic Unterhaching, 3D structural model, hydrogeological model and a high-resolution 3D temperature model form the basis of the numerical modeling. Different seismic signatures, seismic attributes and variations in the interval velocities characterize the ground geophysically, and were interpreted under consideration of geological and hydrogeological background information as well as borehole measurements in terms of hydraulically conductive homogeneous areas. For the Munich metropolitan area, the numerical modeling is a decision aid for the future optimized and sustainable hydrothermal utilization of the Malm aquifer.

  13. Supercritical Water Nuclear Steam Supply System: Innovations In Materials, Neutronics & Thermal-Hydraulics

    Energy Technology Data Exchange (ETDEWEB)

    Mark Anderson; M.L. Corradini; K. Sridharan; P. WIlson; D. Cho; T.K. Kim; S. Lomperski

    2004-09-02

    In the 1990's supercritical light-water reactors were considered in conceptual designs. A nuclear reactor cooled by supercritical waster would have a much higher thermal efficiency with a once-through direct power cycle, and could be based on standardized water reactor components (light water or heavy water). The theoretical efficiency could be improved by more than 33% over that of other water reactors and could be simplified with higher reliability; e.g., a boiling water reactor without steam separators or dryers.

  14. Numerical research on the neutronic/thermal-hydraulic/mechanical coupling characteristics of the optimized helium cooled solid breeder blanket for CFETR

    Energy Technology Data Exchange (ETDEWEB)

    Cui, Shijie; Zhang, Dalin, E-mail: dlzhang@mail.xjtu.edu.cn; Cheng, Jie; Tian, Wenxi; Su, G.H.

    2017-01-15

    As one of the candidate tritium breeding blankets for Chinese Fusion Engineering Test Reactor (CFETR), a conceptual structure of the helium cooled solid breeder blanket has recently been proposed. The neutronic, thermal-hydraulic and mechanical characteristics of the blanket directly affect its tritium breeding and safety performance. Therefore, neutronic/thermal-hydraulic/mechanical coupling analyses are of vital importance for a reliable blanket design. In this work, first, three-dimensional neutronics analysis and optimization of the typical outboard equatorial blanket module (No. 12) were performed for the comprehensive optimal scheme. Then, thermal and fluid dynamic analyses of the scheme under both normal and critical conditions were performed and coupled with the previous neutronic calculation results. With thermal-hydraulic boundaries, thermo-mechanical analyses of the structure materials under normal, critical and blanket over-pressurization conditions were carried out. In addition, several parametric sensitivity studies were also conducted to investigate the influences of the main parameters on the blanket temperature distributions. In this paper, the coupled analyses verify the reasonability of the optimized conceptual design preliminarily and can provide an important reference for the further analysis and optimization design of the CFETR helium cooled solid breeder blanket.

  15. Thermal and hydraulic considerations regarding the fate of water discharged by the outflow channels to the Martian northern plains

    Science.gov (United States)

    Clifford, S. M.

    1993-01-01

    The identification of possible shorelines in the Martian northern plains suggests that the water discharged by the circum-Chryse outflow channels may have led to the formation of transient seas, or possibly even an ocean, covering as much as one-third of the planet. Speculations regarding the possible fate of this water have included local ponding and reinfiltration into the crust; freezing, sublimation, and eventual cold-trapping at higher latitudes; or the in situ survival of this now frozen water to the present day -- perhaps aided by burial beneath a protective cover of eolian sediment or lavas. Although neither cold-trapping at higher latitudes nor the subsequent freezing and burial of flood waters can be ruled out, thermal and hydraulic considerations effectively eliminate the possibility that any significant reassimilation of this water by local infiltration has occurred given climatic conditions resembling those of today. The arguments against the local infiltration of flood water into the northern plains are two-fold. First, given the climatic and geothermal conditions that are thought to have prevailed on Mars during the Late Hesperian (the period of peak outflow channel activity in the northern plains), the thickness of the cryosphere in Chryse Planitia is likely to have exceeded 1 km. A necessary precondition for the widespread occurrence of groundwater is that the thermodynamic sink represented by the cryosphere must already be saturated with ice. For this reason, the ice-saturated cryosphere acts as an impermeable barrier that effectively precludes the local resupply of subpermafrost groundwater by the infiltration of water discharged to the surface by catastraphic floods. Note that the problem of local infiltration is not significantly improved even if the cryosphere were initially dry, for as water attempts to infiltrate the cold, dry crust, it will quickly freeze, creating a seal that prevents any further infiltration from the ponded water above

  16. FEM (finite element method) thermal modeling and thermal hydraulic performance of an enhanced thermal conductivity UO2/BeO composite fuel

    Energy Technology Data Exchange (ETDEWEB)

    Zhou, Wenzhong [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Purdue Univ., West Lafayette, IN (United States)

    2011-03-24

    An enhanced thermal conductivity UO2-BeO composite nuclear fuel was studied. A methodology to generate ANSYS (an engineering simulation software) FEM (Finite Element Method) thermal models of enhanced thermal conductivity oxide nuclear fuels was developed. The results showed significant increase in the fuel thermal conductivities and have good agreement with the measured ones. The reactor performance analysis showed that the decrease in centerline temperature was 250-350K for the UO2-BeO composite fuel, and thus we can improve nuclear reactors' performance and safety, and high-level radioactive waste generation.

  17. Thermal-hydraulic analysis of the HL-2M divertor using an homogeneous equilibrium model

    Science.gov (United States)

    Lu, Yong; Cai, Lijun; Liu, Yuxiang; Liu, Jian; Yuan, Yinglong; Zheng, Guoyao; Liu, Dequan

    2017-09-01

    The heat flux of the HL-2M divertor would reach 10 MW m-2 or more at the local area when the device operates at high parameters. Subcooled boiling could occur at high thermal load, which would be simulated based on the homogeneous equilibrium model. The results show that the current design of the HL-2M divertor could withstand the local heat flux 10 MW m-2 at a plasma pulse duration of 5 s, inlet coolant pressure of 1.5 MPa and flow velocity of 4 m s-1. The pulse duration that the HL-2M divertor could withstand is closely related to the coolant velocity. In addition, at the time of 2 min after plasma discharge, the flow velocity decreased from 4 m s-1 to 1 m s-1, and the divertor could also be cooled to the initial temperature before the next plasma discharge commences.

  18. Influences of hydraulic gradient, surface roughness, intersecting angle, and scale effect on nonlinear flow behavior at single fracture intersections

    Science.gov (United States)

    Li, Bo; Liu, Richeng; Jiang, Yujing

    2016-07-01

    Fluid flow tests were conducted on two crossed fracture models for which the geometries of fracture segments and intersections were measured by utilizing a visualization technique using a CCD (charged coupled device) camera. Numerical simulations by solving the Navier-Stokes equations were performed to characterize the fluid flow at fracture intersections. The roles of hydraulic gradient, surface roughness, intersecting angle, and scale effect in the nonlinear fluid flow behavior through single fracture intersections were investigated. The simulation results of flow rate agreed well with the experimental results for both models. The experimental and simulation results showed that with the increment of the hydraulic gradient, the ratio of the flow rate to the hydraulic gradient, Q/J, decreases and the relative difference of Q/J between the calculation results employing the Navier-Stokes equations and the cubic law, δ, increases. When taking into account the fracture surface roughness quantified by Z2 ranging 0-0.42 for J = 1, the value of δ would increase by 0-10.3%. The influences of the intersecting angle on the normalized flow rate that represents the ratio of the flow rate in a segment to the total flow rate, Ra, and the ratio of the hydraulic aperture to the mechanical aperture, e/E, are negligible when J 10-2. Based on the regression analysis on simulation results, a mathematical expression was proposed to quantify e/E, involving variables of J and Rr, where Rr is the radius of truncating circles centered at an intersection. For E/Rr > 10-2, e/E varies significantly and the scale of model has large impacts on the nonlinear flow behavior through intersections, while for E/Rr fracture intersections is suggested as J < 10-3, E/Rr < 10-3, and Z2 = 0.

  19. Making Heat Visible: Promoting Energy Conservation Behaviors Through Thermal Imaging.

    Science.gov (United States)

    Goodhew, Julie; Pahl, Sabine; Auburn, Tim; Goodhew, Steve

    2015-12-01

    Householders play a role in energy conservation through the decisions they make about purchases and installations such as insulation, and through their habitual behavior. The present U.K. study investigated the effect of thermal imaging technology on energy conservation, by measuring the behavioral effect after householders viewed images of heat escaping from or cold air entering their homes. In Study 1 (n = 43), householders who received a thermal image reduced their energy use at a 1-year follow-up, whereas householders who received a carbon footprint audit and a non-intervention control demonstrated no change. In Study 2 (n = 87), householders were nearly 5 times more likely to install draught proofing measures after seeing a thermal image. The effect was especially pronounced for actions that addressed an issue visible in the images. Findings indicate that using thermal imaging to make heat loss visible can promote energy conservation.

  20. ELM - A SIMPLE TOOL FOR THERMAL-HYDRAULIC ANALYSIS OF SOLID-CORE NUCLEAR ROCKET FUEL ELEMENTS

    Science.gov (United States)

    Walton, J. T.

    1994-01-01

    ELM is a simple computational tool for modeling the steady-state thermal-hydraulics of propellant flow through fuel element coolant channels in nuclear thermal rockets. Written for the nuclear propulsion project of the Space Exploration Initiative, ELM evaluates the various heat transfer coefficient and friction factor correlations available for turbulent pipe flow with heat addition. In the past, these correlations were found in different reactor analysis codes, but now comparisons are possible within one program. The logic of ELM is based on the one-dimensional conservation of energy in combination with Newton's Law of Cooling to determine the bulk flow temperature and the wall temperature across a control volume. Since the control volume is an incremental length of tube, the corresponding pressure drop is determined by application of the Law of Conservation of Momentum. The size, speed, and accuracy of ELM make it a simple tool for use in fuel element parametric studies. ELM is a machine independent program written in FORTRAN 77. It has been successfully compiled on an IBM PC compatible running MS-DOS using Lahey FORTRAN 77, a DEC VAX series computer running VMS, and a Sun4 series computer running SunOS UNIX. ELM requires 565K of RAM under SunOS 4.1, 360K of RAM under VMS 5.4, and 406K of RAM under MS-DOS. Because this program is machine independent, no executable is provided on the distribution media. The standard distribution medium for ELM is one 5.25 inch 360K MS-DOS format diskette. ELM was developed in 1991. DEC, VAX, and VMS are trademarks of Digital Equipment Corporation. Sun4 and SunOS are trademarks of Sun Microsystems, Inc. IBM PC is a registered trademark of International Business Machines. MS-DOS is a registered trademark of Microsoft Corporation.

  1. The Numerical Nuclear Reactor for High-Fidelity Integrated Simulation of Neutronic, Thermal-Hydraulic, and Thermo-Mechanical Phenomena

    Energy Technology Data Exchange (ETDEWEB)

    Kim, K. S.; Ju, H. G.; Jeon, T. H. and others

    2005-03-15

    A comprehensive high fidelity reactor core modeling capability has been developed for detailed analysis of current and advanced reactor designs as part of a US-ROK collaborative I-NERI project. High fidelity was accomplished by integrating highly refined solution modules for the coupled neutronic, thermal-hydraulic, and thermo-mechanical phenomena. Each solution module employs methods and models that are formulated faithfully to the first-principles governing the physics, real geometry, and constituents. Specifically, the critical analysis elements that are incorporated in the coupled code capability are whole-core neutron transport solution, ultra-fine-mesh computational fluid dynamics/heat transfer solution, and finite-element-based thermo-mechanics solution, all obtained with explicit (fuel pin cell level) heterogeneous representations of the components of the core. The vast computational problem resulting from such highly refined modeling is solved on massively parallel computers, and serves as the 'numerical nuclear reactor'. Relaxation of modeling parameters were also pursued to make problems run on clusters of workstations and PCs for smaller scale applications as well.

  2. Thermal-hydraulic study of the LBE-cooled fuel assembly in the MYRRHA reactor: Experiments and simulations

    Energy Technology Data Exchange (ETDEWEB)

    Pacio, J., E-mail: Julio.pacio@kit.edu [Karlsruhe Institute of Technology (KIT), Institute for Nuclear and Energy Technologies (IKET), Hermann-von-Helmholtz Platz 1, 76344 Eggenstein-Leopoldshafen (Germany); Wetzel, T. [Karlsruhe Institute of Technology (KIT), Institute for Nuclear and Energy Technologies (IKET), Hermann-von-Helmholtz Platz 1, 76344 Eggenstein-Leopoldshafen (Germany); Doolaard, H.; Roelofs, F. [Nuclear Research and Consultancy Group (NRG), Westerduinweg 3, 1755 ZG Petten (Netherlands); Van Tichelen, K. [Belgian Nuclear Reseach Center (SCK-CEN), Boeretang 200, Mol (Belgium)

    2017-02-15

    Heavy liquid metals (HLMs), such as lead-bismuth eutectic (LBE) and pure lead are prominent candidate coolants for many advanced systems based on fast neutrons. In particular, LBE is used in the first-of-its-kind MYRRHA fast reactor, to be built in Mol (Belgium), which can be operated either in critical mode or as a sub-critical accelerator-driven system. With a strong focus on safety, key thermal-hydraulic aspects of these systems, such as the proper cooling of fuel assemblies, must be assessed. Considering the complex geometry and low Prandtl number of LBE (Pr ∼ 0.025), this flow scenario is challenging for the models used in Computational Fluid Dynamics (CFD), e.g. for relating the turbulent transport of momentum and heat. Thus, reliable experimental data for the relevant scenario are needed for validation. In this general context, this topic is studied both experimentally and numerically in the framework of the European FP7 project SEARCH (2011–2015). An experimental campaign, including a 19-rod bundle with wire spacers, cooled by LBE is undertaken at KIT. With prototypical geometry and operating conditions, it is intended to evaluate the validity of current empirical correlations for the MYRRHA conditions and, at the same time, to provide validation data for the CFD simulations performed at NRG. The results of one benchmarking case are presented in this work. Moreover, this validated approach is then used for simulating a complete MYRRHA fuel assembly (127 rods).

  3. A two-fluid two-phase model for thermal-hydraulic analysis of a U-tube steam generator

    Energy Technology Data Exchange (ETDEWEB)

    Hung, Huanjen; Chieng, Chingchang; Pei, Baushei (National Tsinghua Univ, Hsinchu (Taiwan, Province of China)); Wang, Songfeng (Inst. of Nuclear Energy Research, Lungtan (Taiwan, Province of China))

    1993-02-01

    The Advanced Thermal-Hydraulic Analysis Code for Nuclear Steam Generators (ATHANS) was developed on the basis of the THERMIT-UTSG computer code for U-tube steam generators. The main features of the ATHANS model are as follows: (a) the equations are solved in cylindrical coordinates, (b) the number and the arrangement of the control volumes inside the steam generator can be chosen by the user, (c) the virtual mass effect is incorporated, and (d) the conjugate gradient squared method is employed to accelerate and improve the numerical convergence. The performance of the model is successfully validated by comparison with the test data from a Westinghouse model F steam generator at the Maanshan nuclear power plant. Better agreement with the test data can be obtained by a finer grid system using a cylindrical coordinate system and the virtual mass effect. With these advanced features, ATHANS provides the basic framework for further studies on the problems of steam generators, such as analyses of secondary-side corrosion and tube ruptures.

  4. The Verification of Coupled Neutronics Thermal-Hydraulics Code NODAL3 in the PWR Rod Ejection Benchmark

    Directory of Open Access Journals (Sweden)

    Surian Pinem

    2014-01-01

    Full Text Available A coupled neutronics thermal-hydraulics code NODAL3 has been developed based on the few-group neutron diffusion equation in 3-dimensional geometry for typical PWR static and transient analyses. The spatial variables are treated by using a polynomial nodal method while for the neutron dynamic solver the adiabatic and improved quasistatic methods are adopted. In this paper we report the benchmark calculation results of the code against the OECD/NEA CRP PWR rod ejection cases. The objective of this work is to determine the accuracy of NODAL3 code in analysing the reactivity initiated accident due to the control rod ejection. The NEACRP PWR rod ejection cases are chosen since many organizations participated in the NEA project using various methods as well as approximations, so that, in addition to the reference solutions, the calculation results of NODAL3 code can also be compared to other codes’ results. The transient parameters to be verified are time of power peak, power peak, final power, final average Doppler temperature, maximum fuel temperature, and final coolant temperature. The results of NODAL3 code agree well with the PHANTHER reference solutions in 1993 and 1997 (revised. Comparison with other validated codes, DYN3D/R and ANCK, shows also a satisfactory agreement.

  5. Final report of the 'Nordic thermal-hydraulic and safety network (NOTNET)' - Project

    Energy Technology Data Exchange (ETDEWEB)

    Tuunanen, J.; Tuomainen, M. [VTT Processes (Finland)

    2005-04-01

    A Nordic network for thermal-hydraulics and nuclear safety research was started. The idea of the network is to combine the resources of different research teams in order to carry out more ambitious and extensive research programs than would be possible for the individual teams. From the very beginning, the end users of the research results have been integrated to the network. Aim of the network is to benefit the partners involved in nuclear energy in the Nordic Countries (power companies, reactor vendors, safety regulators, research units). First task within the project was to describe the resources (personnel, know-how, simulation tools, test facilities) of the various teams. Next step was to discuss with the end users about their research needs. Based on these steps, few most important research topics with defined goals were selected, and coarse road maps were prepared for reaching the targets. These road maps will be used as a starting point for planning the actual research projects in the future. The organisation and work plan for the network were established. National coordinators were appointed, as well as contact persons in each participating organisation, whether research unit or end user. This organisation scheme is valid for the short-term operation of NOTNET when only Nordic organisations take part in the work. Later on, it is possible to enlarge the network e.g. within EC framework programme. The network can now start preparing project proposals and searching funding for the first common research projects. (au)

  6. Atucha II NPP full scope simulator modelling with the thermal hydraulic code TRAC{sub R}T

    Energy Technology Data Exchange (ETDEWEB)

    Alonso, Pablo Rey; Ruiz, Jose Antonio; Rivero, Norberto, E-mail: prey@tecnatom.e, E-mail: jaruiz@tecnatom.e, E-mail: nrivero@tecnatom.e [Tecnatom S.A., Madrid (Spain)

    2011-07-01

    In February 2010 NA-SA (Nucleoelectrica Argentina S.A.) awarded Tecnatom the Atucha II full scope simulator project. NA-SA is a public company owner of the Argentinean nuclear power plants. Atucha II is due to enter in operation shortly. Atucha II NPP is a PHWR type plant cooled by the water of the Parana River and has the same design as the Atucha I unit, doubling its power capacity. Atucha II will produce 745 MWe utilizing heavy water as coolant and moderator, and natural uranium as fuel. A plant singular feature is the permanent core refueling. TRAC{sub R}T is the first real time thermal hydraulic six-equations code used in the training simulation industry for NSSS modeling. It is the result from adapting to real time the best estimate code TRACG. TRAC{sub R}T is based on first principle conservation equations for mass, energy and momentum for liquid and steam phases, with two phase flows under non homogeneous and non equilibrium conditions. At present, it has been successfully implemented in twelve full scope replica simulators in different training centers throughout the world. To ease the modeling task, TRAC{sub R}T includes a graphical pre-processing tool designed to optimize this process and alleviate the burden of entering alpha numerical data in an input file. (author)

  7. ANTHEM2000{sup TM}: Integration of the ANTHEM Thermal Hydraulic Model in the ROSE{sup TM} Environment

    Energy Technology Data Exchange (ETDEWEB)

    Boire, R.; Nguyen, M; Salim, G. [CAE Electronics Ltd., Quebec (Canada)

    1999-07-01

    ROSEN{sup TM} is an object oriented, visual programming environment used for many applications, including the development of power plant simulators. ROSE provides an integrated suite of tools for the creation, calibration, test, integration, configuration management and documentation of process, electrical and I and C models. CAE recently undertook an ambitious project to integrate its two phase thermal hydraulic model ANTHEM{sup TM} into the ROSE environment. ANTHEM is a non equilibrium, non-homogenous model based on the drift flux formalism. CAE has used the model in numerous two phase applications for nuclear and fossil power plant simulators. The integration of ANTHEM into ROSE brings the full power of visual based programming to two phase modeling applications. Features include graphical model building, calibration tools, a superior test environment and process visualisation. In addition the integration of ANTHEM into ROSE makes it possible to easily apply the fidelity of ANTHEM to BOP applications. This paper describes the implementation of the ANTHEM model within the ROSE environment and gives examples of its use. (author)

  8. Implicit Treatment of Technical Specification and Thermal Hydraulic Parameter Uncertainties in Gaussian Process Model to Estimate Safety Margin

    Directory of Open Access Journals (Sweden)

    Douglas A. Fynan

    2016-06-01

    Full Text Available The Gaussian process model (GPM is a flexible surrogate model that can be used for nonparametric regression for multivariate problems. A unique feature of the GPM is that a prediction variance is automatically provided with the regression function. In this paper, we estimate the safety margin of a nuclear power plant by performing regression on the output of best-estimate simulations of a large-break loss-of-coolant accident with sampling of safety system configuration, sequence timing, technical specifications, and thermal hydraulic parameter uncertainties. The key aspect of our approach is that the GPM regression is only performed on the dominant input variables, the safety injection flow rate and the delay time for AC powered pumps to start representing sequence timing uncertainty, providing a predictive model for the peak clad temperature during a reflood phase. Other uncertainties are interpreted as contributors to the measurement noise of the code output and are implicitly treated in the GPM in the noise variance term, providing local uncertainty bounds for the peak clad temperature. We discuss the applicability of the foregoing method to reduce the use of conservative assumptions in best estimate plus uncertainty (BEPU and Level 1 probabilistic safety assessment (PSA success criteria definitions while dealing with a large number of uncertainties.

  9. Thermal Behavior of Green Roofs Applied to Tropical Climate

    Directory of Open Access Journals (Sweden)

    Grace Tibério Cardoso

    2013-01-01

    Full Text Available The main goal of this paper is to present results on an experimental field about the green roofs thermal behavior, compared to other traditional roof covering systems. On the one hand, it intends to describe shortly the constructive system of a green roof with a lightweight building system, which has a sustainable building materials character and, on the other, it worries with the water reuse and with the run-off delay. The main methodological procedure adopted to study the thermal behavior of green roof was installing thermocouples to collect surface temperatures and indoor air, later comparing them with existing prototypes in an experimental plot. The thermal behavior analysis of cover systems was assessed by a representative episode of the climate fact, based on the dynamic climate approach. The experimental results from internal air temperature measurements show that the green roofs applied to warm and dry climates also provide an interesting time lag with surface and internal air temperature reduction.

  10. Thermal hydraulics characterization of the core and the reactor vessel type BWR; Caracterizacion termohidraulica del nucleo y de la vasija de un reactor tipo BWR

    Energy Technology Data Exchange (ETDEWEB)

    Zapata Y, M.; Lopez H, L.E. [CFE, Carretera Cardel-Nautla Km. 42.5, Municipio Alto Lucero, Veracruz (Mexico)]. e-mail: marxlenin.zapata@cfe.gob.mx

    2008-07-01

    The thermal hydraulics design of a reactor type BWR 5 as the employees in the nuclear power plant of Laguna Verde involves the coupling of at least six control volumes: Pumps jet region, Stratification region, Core region, Vapor dryer region, Humidity separator region and Reactor region. Except by the regions of the core and reactor, these control volumes only are used for design considerations and their importance as operative data source is limited. It is for that is fundamental to complement the thermal hydraulics relations to obtain major data that allow to determine the efficiency of internal components, such as pumps jet, humidity separator and vapor dryer. Like example of the previous thing, calculations are realized on the humidity of the principal vapor during starting, comparing it with the values at the moment incorporated in the data banks of the computers of process of both units. (Author)

  11. Degradation Behavior of Thermal Stabilized Polyacrylonitrile Fibers

    Directory of Open Access Journals (Sweden)

    LEI Shuai

    2017-05-01

    Full Text Available In the temperature range of 300-800℃, 40%-50% of the mass lost during the processing of polyacrylonitrile based carbon fiber (PANCF. Understanding the degradation behavior will be valuable in understanding the formation mechanism of pseudo-graphite structure, and providing theoretic basis for producing high performance carbon fiber and increasing the carbonization yield. The simulation of the degradation progress was carried out on the thermogravimetric analyzer (TGA, the results show that there are two degradation steps for PAN fiber stabilized in air, and controlled by cyclization coefficient and oxygen content. The cyclization coefficient and oxygen content are effective to the density of carbon fiber by influencing the degradation behavior, which cause defects in the fiber. The higher cyclization coefficient leads to form less structural defects and higher density of the fiber; on the contrary, the higher oxygen content leads to form more structural defects and lower density of the fiber.

  12. Experimental Development and Demonstration of Ultrasonic Measurement Diagnostics for Sodium Fast Reactor Thermal-hydraulics

    Energy Technology Data Exchange (ETDEWEB)

    Tokuhiro, Akira; Jones, Byron

    2013-09-13

    This research project will address some of the principal technology issues related to sodium-cooled fast reactors (SFR), primarily the development and demonstration of ultrasonic measurement diagnostics linked to effective thermal convective sensing under normatl and off-normal conditions. Sodium is well-suited as a heat transfer medium for the SFR. However, because it is chemically reactive and optically opaque, it presents engineering accessibility constraints relative to operations and maintenance (O&M) and in-service inspection (ISI) technologies that are currently used for light water reactors. Thus, there are limited sensing options for conducting thermohydraulic measurements under normal conditions and off-normal events (maintenance, unanticipated events). Acoustic methods, primarily ultrasonics, are a key measurement technology with applications in non-destructive testing, component imaging, thermometry, and velocimetry. THis project would have yielded a better quantitative and qualitative understanding of the thermohydraulic condition of solium under varied flow conditions. THe scope of work will evaluate and demonstrate ultrasonic technologies and define instrumentation options for the SFR.

  13. Model studies of the vertical steam generator thermal-hydraulic characteristics

    Energy Technology Data Exchange (ETDEWEB)

    Desyatun, V.F.; Moskvichev, V.F.; Ulasov, V.M.; Morozov, V.G.; Burkov, V.K.; Grebennikov, V.N.

    1984-03-01

    Results of investigations conducted to clarify the calculation technique and to test the workability of the main elements and units of the PGV-250 vertical steam generator of saturated steam are considered. The steam generating capacity of the plant is 1486 t/h, thermal power is 792 MW. Steam generation follows a multiple circulation scheme. The heat surface comprises 330-shields. The investigations are carried out with a model which reproduces all the main elements of the steam generator excluding the economizer section. The flow rates of feed water, generated steam and coolant of the first circuit as well as temperature, pressure and humidity of the generated steam past the separator are determined. The average heat transfer factors of the heat surface are calculated on the base of the data obtained and a conclusion is drawn on the correctness of the thermohydraulic calculation technique used in development of the PGV-250 steam generator design. Temperature pulsations and heat surface steaming are not observed. The steam humidity at the outlet and steam capture into sinking tubes are within permissible values.

  14. An approach to modeling coupled thermal-hydraulic-chemical processes in geothermal systems

    Science.gov (United States)

    Palguta, Jennifer; Williams, Colin F.; Ingebritsen, Steven E.; Hickman, Stephen H.; Sonnenthal, Eric

    2011-01-01

    Interactions between hydrothermal fluids and rock alter mineralogy, leading to the formation of secondary minerals and potentially significant physical and chemical property changes. Reactive transport simulations are essential for evaluating the coupled processes controlling the geochemical, thermal and hydrological evolution of geothermal systems. The objective of this preliminary investigation is to successfully replicate observations from a series of hydrothermal laboratory experiments [Morrow et al., 2001] using the code TOUGHREACT. The laboratory experiments carried out by Morrow et al. [2001] measure permeability reduction in fractured and intact Westerly granite due to high-temperature fluid flow through core samples. Initial permeability and temperature values used in our simulations reflect these experimental conditions and range from 6.13 × 10−20 to 1.5 × 10−17 m2 and 150 to 300 °C, respectively. The primary mineralogy of the model rock is plagioclase (40 vol.%), K-feldspar (20 vol.%), quartz (30 vol.%), and biotite (10 vol.%). The simulations are constrained by the requirement that permeability, relative mineral abundances, and fluid chemistry agree with experimental observations. In the models, the granite core samples are represented as one-dimensional reaction domains. We find that the mineral abundances, solute concentrations, and permeability evolutions predicted by the models are consistent with those observed in the experiments carried out by Morrow et al. [2001] only if the mineral reactive surface areas decrease with increasing clay mineral abundance. This modeling approach suggests the importance of explicitly incorporating changing mineral surface areas into reactive transport models.

  15. Advanced Multiphysics Thermal-Hydraulics Models for the High Flux Isotope Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Jain, Prashant K [ORNL; Freels, James D [ORNL

    2015-01-01

    Engineering design studies to determine the feasibility of converting the High Flux Isotope Reactor (HFIR) from using highly enriched uranium (HEU) to low-enriched uranium (LEU) fuel are ongoing at Oak Ridge National Laboratory (ORNL). This work is part of an effort sponsored by the US Department of Energy (DOE) Reactor Conversion Program. HFIR is a very high flux pressurized light-water-cooled and moderated flux-trap type research reactor. HFIR s current missions are to support neutron scattering experiments, isotope production, and materials irradiation, including neutron activation analysis. Advanced three-dimensional multiphysics models of HFIR fuel were developed in COMSOL software for safety basis (worst case) operating conditions. Several types of physics including multilayer heat conduction, conjugate heat transfer, turbulent flows (RANS model) and structural mechanics were combined and solved for HFIR s inner and outer fuel elements. Alternate design features of the new LEU fuel were evaluated using these multiphysics models. This work led to a new, preliminary reference LEU design that combines a permanent absorber in the lower unfueled region of all of the fuel plates, a burnable absorber in the inner element side plates, and a relocated and reshaped (but still radially contoured) fuel zone. Preliminary results of estimated thermal safety margins are presented. Fuel design studies and model enhancement continue.

  16. Mold Flux Crystallization and Mold Thermal Behavior

    Science.gov (United States)

    Peterson, Elizabeth Irene

    Mold flux plays a small but critical role in the continuous casting of steel. The carbon-coated powder is added at the top of the water-cooled copper mold, over time it melts and infiltrates the gap between the copper mold and the solidifying steel strand. Mold powders serve five primary functions: (1) chemical insulation, (2) thermal insulation, (3) lubrication between the steel strand and mold, (4) absorption of inclusions, and (5) promotion of even heat flux. All five functions are critical to slab casting, but surface defect prevention is primarily controlled through even heat flux. Glassy fluxes have high heat transfer and result in a thicker steel shell. Steels with large volumetric shrinkage on cooling must have a crystalline flux to reduce the radiative heat transfer and avoid the formation of cracks in the shell. Crystallinity plays a critical role in steel shell formation, therefore it is important to study the thermal conditions that promote each phase and its morphology. Laboratory tests were performed to generate continuous cooling transformation (CCT) and time-temperature-transformation (TTT) diagrams. Continuous cooling transformation tests were performed in an instrumented eight cell step chill mold. Results showed that cuspidine was the only phase formed in conventional fluxes and all observed structures were dendritic. An isothermal tin bath quench method was also developed to isothermally age glassy samples. Isothermal tests yielded different microstructures and different phases than those observed by continuous cooling. Comparison of aged tests with industrial flux films indicates similar faceted structures along the mold wall, suggesting that mold flux first solidifies as a glass along the mold wall, but the elevated temperature devitrifies the glassy structure forming crystals that cannot form by continuous cooling.

  17. Comparison of best-estimate plus uncertainty and conservative methodologies for a PWR MSLB analysis using a coupled 3-D neutron-kinetics/thermal-hydraulic code

    OpenAIRE

    Pericas Casals, Raimon; Ivanov, K.; Reventós Puigjaner, Francesc; Batet Miracle, Lluís

    2017-01-01

    This paper compares the Best-Estimate Plus Uncertainty (BEPU) methodology with the Conservative Bounding methodology for design-basis-accident analysis. Calculations have been performed with TRACE [for thermal-hydraulic (TH) system calculations] and PARCS [for neutron-kinetics (NK) modeling] under the SNAP platform. DAKOTA is used under the SNAP interface for uncertainty and sensitivity analysis. A simplified three-dimensional (3-D) neutronics model of the Ascó II nuclear power plant is used ...

  18. Development and application of the coupled thermal-hydraulics and neutron-kinetics code ATHLET/BIPR-VVER for safety analysis

    Energy Technology Data Exchange (ETDEWEB)

    Lizorkin, M.; Nikonov, S. [Kurchatov Institute for Atomic Energy, Moscow (Russian Federation); Langenbuch, S.; Velkov, K. [Gesellschaft fur Anlagen- und Reaktorsicherheit (GRS) mbH, Garching (Germany)

    2006-07-01

    The coupled thermal-hydraulics and neutron-kinetics code ATHLET/BIPR-VVER was developed within a co-operation between the RRC Kurchatov Institute (KI) and GRS. The modeling capability of this coupled code as well as the status of validation by benchmark activities and comparison with plant measurements are described. The paper is focused on the modeling of flow mixing in the reactor pressure vessel including its validation and the application for the safety justification of VVER plants. (authors)

  19. The thermal-hydraulic for the new technologies: the micro-fluidics; La thermohydraulique au service des nouvelles technologies: la microfluidique

    Energy Technology Data Exchange (ETDEWEB)

    Crecy, F. de; Gruss, A.; Bricard, A.; Excoffon, J

    2000-07-01

    The micro-fluidics can be defined as the fluid flow in little canals. This scale offers a great interest for the biotechnology type. In this paper, the authors present this fluids form and detail the researches performed at the Department of Physics and Thermal-hydraulics of the CEA, in the domain of the physical properties characterization and of the numerical two-phase direct simulation. (A.L.B.)

  20. Using finite mixture models in thermal-hydraulics system code uncertainty analysis

    Energy Technology Data Exchange (ETDEWEB)

    Carlos, S., E-mail: scarlos@iqn.upv.es [Department d’Enginyeria Química i Nuclear, Universitat Politècnica de València, Camí de Vera s.n, 46022 València (Spain); Sánchez, A. [Department d’Estadística Aplicada i Qualitat, Universitat Politècnica de València, Camí de Vera s.n, 46022 València (Spain); Ginestar, D. [Department de Matemàtica Aplicada, Universitat Politècnica de València, Camí de Vera s.n, 46022 València (Spain); Martorell, S. [Department d’Enginyeria Química i Nuclear, Universitat Politècnica de València, Camí de Vera s.n, 46022 València (Spain)

    2013-09-15

    Highlights: • Best estimate codes simulation needs uncertainty quantification. • The output variables can present multimodal probability distributions. • The analysis of multimodal distribution is performed using finite mixture models. • Two methods to reconstruct output variable probability distribution are used. -- Abstract: Nuclear Power Plant safety analysis is mainly based on the use of best estimate (BE) codes that predict the plant behavior under normal or accidental conditions. As the BE codes introduce uncertainties due to uncertainty in input parameters and modeling, it is necessary to perform uncertainty assessment (UA), and eventually sensitivity analysis (SA), of the results obtained. These analyses are part of the appropriate treatment of uncertainties imposed by current regulation based on the adoption of the best estimate plus uncertainty (BEPU) approach. The most popular approach for uncertainty assessment, based on Wilks’ method, obtains a tolerance/confidence interval, but it does not completely characterize the output variable behavior, which is required for an extended UA and SA. However, the development of standard UA and SA impose high computational cost due to the large number of simulations needed. In order to obtain more information about the output variable and, at the same time, to keep computational cost as low as possible, there has been a recent shift toward developing metamodels (model of model), or surrogate models, that approximate or emulate complex computer codes. In this way, there exist different techniques to reconstruct the probability distribution using the information provided by a sample of values as, for example, the finite mixture models. In this paper, the Expectation Maximization and the k-means algorithms are used to obtain a finite mixture model that reconstructs the output variable probability distribution from data obtained with RELAP-5 simulations. Both methodologies have been applied to a separated

  1. Deformation Behavior between Hydraulic and Natural Fractures Using Fully Coupled Hydromechanical Model with XFEM

    Directory of Open Access Journals (Sweden)

    Fei Liu

    2017-01-01

    Full Text Available There has been a growing consensus that preexisting natural fractures play an important role during stimulation. A novel fully coupled hydromechanical model using extended finite element method is proposed. This directly coupled scheme avoids the cumbersome process during calculating the fluid pressure in complicated fracture networks and translating into an equivalent nodal force. Numerical examples are presented to simulate the hydraulic fracture propagation paths for simultaneous multifracture treatments with properly using the stress shadow effects for horizontal wells and to reveal the deformation response and interaction mechanism between hydraulic induced fracture and nonintersected natural fractures at orthotropic and nonorthotropic angles. With the stress shadow effects, the induced hydraulic flexural fracture deflecting to wellbore rather than transverse fracture would be formed during the progress of simultaneous fracturing for a horizontal well. The coupled hydromechanical simulation reveals that the adjacent section to the intersection is opened and the others are closed for orthogonal natural fracture, while the nonorthogonal natural fracture is activated near the intersection firstly and along the whole section with increasing perturbed stresses. The results imply that the induced hydraulic fracture tends to cross orthotropic natural fracture, while it is prior to being arrested by the nonorthotropic natural fracture.

  2. Anisotropic Thermal Behavior of Silicone Polymer, DC 745

    Energy Technology Data Exchange (ETDEWEB)

    Adams, Jillian Cathleen [Univ. of Oregon, Eugene, OR (United States). Dept. of Chemistry; Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Torres, Joseph Angelo [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Volz, Heather Michelle [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Gallegos, Jennifer Marie [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Yang, Dali [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2016-09-02

    In material applications, it is important to understand how polymeric materials behave in the various environments they may encounter. One factor governing polymer behavior is processing history. Differences in fabrication will result in parts with varied or even unintended properties. In this work, the thermal expansion behavior of silicone DC 745 is studied. Thermomechanical analysis (TMA) is used to determine changes in sample dimension resulting from changes in temperature. This technique can measure thermal events such as the linear coefficient of thermal expansion (CTE), melting, glass transitions, cure shrinkage, and internal relaxations. Using a thermomechanical analyzer (Q400 TMA), it is determined that DC 745 expands anisotropically when heated. This means that the material has a different CTE depending upon which direction is being measured. In this study, TMA experiments were designed in order to confirm anisotropic thermal behavior in multiple DC 745 samples of various ages and lots. TMA parameters such as temperature ramp rate, preload force, and temperature range were optimized in order to ensure the most accurate and useful data. A better understanding of the thermal expansion of DC 745 will allow for more accurate modeling of systems using this material.

  3. MODELING OF THE THERMAL BEHAVIOR OF METALS DURING ...

    African Journals Online (AJOL)

    30 juin 2012 ... ABSTRACT. The laser welding provides a very high energy density, this allows the use of welds narrow, deep and at high speeds. Because of its advantages over other joining processes, laser welding is a process of manufacturing the most important. In order to model the thermal behavior, we must study ...

  4. Thermal behavior analysis of PWR fuel during RIA at various fuel burnups using modified theatre code

    Directory of Open Access Journals (Sweden)

    Nawaz Amjad

    2016-01-01

    Full Text Available The fuel irradiation and burnup causes geometrical and dimensional changes in the fuel rod which affects its thermal resistance and ultimately affects the fuel rod behavior during steady-state and transient conditions. The consistent analysis of fuel rod thermal performance is essential for precise evaluation of reactor safety in operational transients and accidents. In this work, analysis of PWR fuel rod thermal performance is carried out under steady-state and transient conditions at different fuel burnups. The analysis is performed by using thermal hydraulic code, THEATRe. The code is modified by adding burnup dependent fuel rod behavior models. The original code uses as-fabricated fuel rod dimensions during steady-state and transient conditions which can be modified to perform more consistent reactor safety analysis. AP1000 reactor is considered as a reference reactor for this analysis. The effect of burnup on steady-state fuel rod parameters has been investigated. For transient analysis, hypothetical reactivity initiated accident was simulated by considering a triangular power pulse of variable pulse height (relative to the full power reactor operating conditions and pulse width at different fuel burnups which corresponds to fresh fuel, low and medium burnup fuels. The effect of power pulse height, pulse width and fuel burnup on fuel rod temperatures has been investigated. The results of reactivity initiated accident analysis show that the fuel failure mechanisms are different for fresh fuel and fuel at different burnup levels. The fuel failure in fresh fuel is expected due to fuel melting as fuel temperature increases with increase in pulse energy (pulse height. However, at relatively higher burnups, the fuel failure is expected due to cladding failure caused by strong pellet clad mechanical interaction, where, the contact pressure increases beyond the cladding yield strength.

  5. Thermal-hydraulic modelling of the SAFARI-1 research reactor using RELAP/SCDAPSIM/MOD3.4

    Energy Technology Data Exchange (ETDEWEB)

    Sekhri, Abdelkrim; Graham, Andy [RRT Radiation and Reactor Theory, South African Nuclear Energy Corporation - NECSA, PO Box 582 Pretoria 0001 (South Africa); D' Arcy, Alan; Oliver, Melissa [SAFARI-1 Research Reactor, South African Nuclear Energy Corporation - NECSA, PO Box 582 Pretoria 0001 (South Africa)

    2008-07-01

    The SAFARI-1 reactor is a tank-in-pool MTR type research reactor operated at a nominal core power of 20 MW. It operates exclusively in the single phase liquid water regime with nominal water and fuel temperatures not exceeding 100 deg. C. RELAP/SCDAPSIM/MOD3.4 is a Best Estimate Code for light water reactors as well as for low pressure transients, as part of the code validation was done against low pressure facilities and research reactor experimental data. The code was used to simulate SAFARI-1 in normal and abnormal operation and validated against the experimental data in the plant and was used extensively in the upgrading of the Safety Analysis Report (SAR) of the reactor. The focus of the following study is the safety analysis of the SAFARI-1 research reactor and describes the thermal hydraulic modelling and analysis approach. Particular emphasis is placed on the modelling detail, the application of the no-boiling rule and predicting the Onset of Nucleate Boiling and Departure from Nucleate Boiling under Loss of Flow conditions. Such an event leads the reactor to switch to a natural convection regime which is an adequate mode to maintain the clad and fuel temperature within the safety margin. It is shown that the RELAP/SCDAPSIM/MOD3.4 model can provide accurate predictions as long as the clad temperature remains below the onset of nucleate boiling temperature and the DNB ratio is greater than 2. The results are very encouraging and the model is shown to be appropriate for the analysis of SAFARI-1 research reactor. (authors)

  6. The axial power distribution validation of the SCWR fuel assembly with coupled neutronics-thermal hydraulics method

    Energy Technology Data Exchange (ETDEWEB)

    Xi, Xi [CNNC Key Laboratory on Nuclear Reactor Thermal Hydraulics Technology, Nuclear Power Institute of China, Chengdu 610041 (China); Xiao, Zejun, E-mail: fabulous_2012@sina.com [CNNC Key Laboratory on Nuclear Reactor Thermal Hydraulics Technology, Nuclear Power Institute of China, Chengdu 610041 (China); Yan, Xiao; Li, Yongliang; Huang, Yanping [CNNC Key Laboratory on Nuclear Reactor Thermal Hydraulics Technology, Nuclear Power Institute of China, Chengdu 610041 (China)

    2013-05-15

    Highlights: ► CFX and MCNP codes are suitable to calculate the axial power profile of the FA. ► The partition method in the calculation will affect the final result. ► The density feedback has little effect on the axial power profile of CSR1000 FA. -- Abstract: SCWR (super critical water reactor) is one of the IV generation nuclear reactors in the world. In a typical SCWR the water enters the reactor from the cold leg with a temperature of 280 °C and then leaves the core with a temperature of 500 °C. Due to the sharp change in temperature, there is a huge density change of the water along the axial direction of the fuel assembly (FA), which will affect the moderating power of the water. So the axial power distribution of the SCWR FA could be different from the traditional PWR FA.In this paper, it is the first time that the thermal hydraulics code CFX and neutronics code MCNP are used to analyze the axial power distribution of the SCWR FA. First, the factors in the coupled method which could affect the result are analyzed such as the initialization value or the partition method especially in the MCNP code. Then the axial power distribution of the Europe HPLWR FA is obtained by the coupled method with the two codes and the result is compared with that obtained by Waata and Reiss. There is a good agreement among the three kinds of results. At last, this method is used to calculate the axial power distribution of the Chinese SCWR (CSR1000) FA. It is found the axial power profile of the CSR1000 FA is not so sensitive to the change of the moderator density.

  7. Constraining groundwater discharge in a large watershed: Integrated isotopic, hydraulic, and thermal data from the Canadian shield

    Science.gov (United States)

    Gleeson, Tom; Novakowski, Kent; Cook, Peter G.; Kyser, T. Kurt

    2009-08-01

    The objective of this study is to evaluate the pattern and rate of groundwater discharge in a large, regulated fractured rock watershed using novel and standard methods that are independent of base flow recession. Understanding the rate and pattern of groundwater discharge to surface water bodies is critical for watershed budgets, as a proxy for recharge rates, and for protecting the ecological integrity of lake and river ecosystems. The Tay River is a low-gradient, warm-water river that flows over exposed and fractured bedrock or a thin veneer of coarse-grained sediments. Natural conservative (δ2H, δ18O, Cl, and specific conductance), radioactive (222Rn), and thermal tracers are integrated with streamflow measurements and a steady state advective model to delimit the discharge locations and quantify the discharge fluxes to lakes, wetlands, creeks, and the Tay River. The groundwater discharge rates to most surface water body types are low, indicating that the groundwater and surface water system may be largely decoupled in this watershed compared to watersheds underlain by porous media. Groundwater discharge is distributed across the watershed rather than localized around lineaments or high-density zones of exposed brittle fractures. The results improve our understanding of the rate, localization, and conceptualization of discharge in a large, fractured rock watershed. Applying hydraulic, isotopic, or chemical hydrograph separation techniques would be difficult because the groundwater discharge "signal" is small compared to the "background" surface water inflows or volumes of the surface water bodies. Although this study focuses on a large watershed underlain by fractured bedrock, the methodology developed is transferable to any large regulated or unregulated watershed. The low groundwater discharge rates have significant implications for the ecology, sustainability, and management of large, crystalline watersheds.

  8. RDS - A systematic approach towards system thermal hydraulics input code development for a comprehensive deterministic safety analysis

    Science.gov (United States)

    Salim, Mohd Faiz; Roslan, Ridha; Ibrahim, Mohd Rizal Mamat @

    2014-02-01

    Deterministic Safety Analysis (DSA) is one of the mandatory requirements conducted for Nuclear Power Plant licensing process, with the aim of ensuring safety compliance with relevant regulatory acceptance criteria. DSA is a technique whereby a set of conservative deterministic rules and requirements are applied for the design and operation of facilities or activities. Computer codes are normally used to assist in performing all required analysis under DSA. To ensure a comprehensive analysis, the conduct of DSA should follow a systematic approach. One of the methodologies proposed is the Standardized and Consolidated Reference Experimental (and Calculated) Database (SCRED) developed by University of Pisa. Based on this methodology, the use of Reference Data Set (RDS) as a pre-requisite reference document for developing input nodalization was proposed. This paper shall describe the application of RDS with the purpose of assessing its effectiveness. Two RDS documents were developed for an Integral Test Facility of LOBI-MOD2 and associated Test A1-83. Data and information from various reports and drawings were referred in preparing the RDS. The results showed that by developing RDS, it has made possible to consolidate all relevant information in one single document. This is beneficial as it enables preservation of information, promotes quality assurance, allows traceability, facilitates continuous improvement, promotes solving of contradictions and finally assisting in developing thermal hydraulic input regardless of whichever code selected. However, some disadvantages were also recognized such as the need for experience in making engineering judgments, language barrier in accessing foreign information and limitation of resources. Some possible improvements are suggested to overcome these challenges.

  9. Design and optimization of the WEST ICRH antenna front face components based on thermal and hydraulic analysis

    Energy Technology Data Exchange (ETDEWEB)

    Chen, Zhaoxi, E-mail: chenzx@ipp.ac.cn [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei 230031 (China); University of Science and Technology of China, Hefei 230026 (China); Vulliez, Karl [Laboratoire d’étanchéité, DEN/DTEC/SDTC, Commissariat à l’énergie atomique et aux énergies alternatives, 2 rue James Watt, 26700 Pierrelatte (France); Ferlay, Fabien; Martinez, André; Mollard, Patrick; Hillairet, Julien; Doceul, Louis; Bernard, Jean-Michel; Larroque, Sébastien; Helou, Walid [CEA, IRFM, F-13108, Saint-Paul-Lez-Durance (France); Song, Yuntao; Yang, Qingxi; Wang, Yongsheng [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei 230031 (China)

    2015-05-15

    Highlights: • Three ICRH antennas are designed to realize continuous-wave operation. • Fully active cooling structure is designed which takes the balance of structure safety and cooling performance. • High cooling efficiency is achieved for the current cooling circuit design based on the thermal-hydraulic simulation. - Abstract: The WEST (Tungsten (W) Environment in Steady-state Tokamak) is an upgrade of Tore-Supra (TS) which aims it into an X-point magnetic configuration tokamak equipped with an actively cooled tungsten divertor. To be a platform of ITER technologies of high heat flux components testing, three sets of Ion Cyclotron Resonant Heating (ICRH) antennas have been designed to inject 9 MW during 30 s or 3 MW during 1000 s. The antenna design is based on a load resilient prototype successfully tested in Tore Supra in 2007. In order to allow continuous-wave (CW) operations, the mechanical design of the WEST ICRH antenna is emphasized on its cooling performances by designing fully active cooling structure. Two kinds of cooling water loops are used, with temperature and pressure of 70 °C/30 bar and 25 °C/5.2 bar, respectively. The hot water loop is used for the Faraday screen (FS) and the housing box (HB), while the cold water loop is used for the straps, the matching capacitors and the impedance transformer. To enhance the heat removal ability and control the pressure drop, the cooling channels in the FS and HB are drilled directly and parallel connected as much as possible. By performing the hydraulic–thermal analysis, the lack of cooling efficiency was found in the front face of lateral collector where 1 MW/m{sup 2} is imposed and fluid dead zones were found in some of the bars. After optimization, the cooling performance of the cooling circuit increased significantly. With a mass flow rate of 2.5 kg/s, the total pressure drop is 3.1 bar, and the peak temperatures on the FS and HB are 500 °C and 261 °C, respectively. Besides, no cavitation is

  10. Coupling of unidimensional neutron kinetics to thermal hydraulics in parallel channels; Acoplamiento de cinetica neutronica unidimensional a canales termohidraulicos en paralelo

    Energy Technology Data Exchange (ETDEWEB)

    Cecenas F, M.; Campos G, R.M. [IIE, Av. Reforma 113, Col. Palmira, Cuernavaca, Morelos (Mexico)]. e-mail: mcf@iie.org.mx

    2003-07-01

    In this work the dynamic behavior of a consistent system in fifteen channels in parallel that represent the reactor core of a BWR type, coupled of a kinetic neutronic model in one dimension is studied by means of time series. The arrangement of channels is obtained collapsing the assemblies that it consists the core to an arrangement of channels prepared in straight lines, and it is coupled to the unidimensional solution of the neutron diffusion equation. This solution represents the radial power distribution, and initially the static solution is obtained to verify that the one modeling core is critic. The coupled set nuclear-thermal hydraulics it is solved numerically by means of a net of CPUs working in the outline teacher-slave by means of Parallel Virtual Machine (PVM), subject to the restriction that the pressure drop is equal for each channel, which is executed iterating on the refrigerant distribution. The channels are dimensioned according to the one Stability Benchmark of the Ringhals swedish plant, organized by the Nuclear Energy Agency in 1994. From the information of this benchmark it is obtained the axial power profile for each channel, which is assumed as invariant in the time. To obtain the time series, the system gets excited with white noise (sequence that statistically obeys to a normal distribution with zero media), so that the power generated in each channel it possesses the same ones characteristics of a typical signal obtained by means of the acquisition of those signals of neutron flux in a BWR reactor. (Author)

  11. Dispersal, behavioral responses and thermal adaptation in Musca domestica

    DEFF Research Database (Denmark)

    Kjaersgaard, Anders; Blackenhorn, Wolf U.; Pertoldi, Cino

    were obtained with flies held for several generations in a laboratory common garden setting, therefore we suggest that exposure to and avoidance of high temperatures under natural conditions has been an important selective agent causing the suggested adaptive differentiation between the populations.......Behavioral traits can have great impact on an organism’s ability to cope with or avoidance of thermal stress, and are therefore of evolutionary importance for thermal adaptation. We compared the morphology, heat resistance, locomotor (walking and flying) activity and flight performance of three...

  12. Contribution to the study of thermal-hydraulic problems in nuclear reactors; Contribution a l`etude de problemes de thermohydraulique dans les reacteurs nucleaires

    Energy Technology Data Exchange (ETDEWEB)

    Cognet, G

    1998-07-07

    In nuclear reactors, whatever the type considered, Pressurized Water Water Reactors (PWRs), Fast Breeder reactors (FBRs)..., thermal-hydraulics, the science of fluid mechanics and thermal behaviour, plays an essential role, both in nominal operating and accidental conditions. Fluid can either be the primary fluid (liquid or gas) or a very specific fluid called corium, which, in case of severe accident, could result from core and environning structure melting. The work reported here represents a 20-year contribution to thermal-hydraulic issues which could occur in FBRs and PWRs. Working on these two types of reactors, both in nominal and severe accident situations, has allowed me to compare the problems and to realize the importance of communication between research teams. The evolution in the complexity of studied problems, unavoidable in order to reduce costs and significantly improve safety, has led me from numerical modelling of single-phase flow turbulence to high temperature real melt experiments. The difficulties encountered in understanding the observed phenomena and in increasing experimental databases for computer code qualification have often entailed my participation in specific measurement device developments or adaptations, in particular non-intrusive devices generally based on optical techniques. Being concerned about the end-use of this research work, I actively participated in `in-situ` thermalhydraulic experiments in the FBRs: Phenix and Super-Phenix, of which I appreciated their undeniable scientific contribution. In my opinion, the thermal-hydraulic questions related to severe accidents are the most complex as they are at the cross-roads of several scientific specialities. Consequently, they require a multi-disciplinary approach and a continuous see-saw motion between experimentalists and modelling teams. After a brief description of the various problems encountered, the main ones are reported. Finally, the importance for research teams to

  13. A new coupling of the 3D thermal-hydraulic code THYC and the thermo-mechanical code CYRANO3 for PWR calculations

    Energy Technology Data Exchange (ETDEWEB)

    Marguet, S.D. [Electricite de France (EDF), 92 - Clamart (France)

    1997-12-31

    Among all parameters, the fuel temperature has a significant influence on the reactivity of the core, because of the Doppler effect on cross-sections. Most neutronic codes use a straightforward method to calculate an average fuel temperature used in their specific feed-back models. For instance, EDF`s neutronic code COCCINELLE uses the Rowland`s formula using the temperatures of the center and the surface of the pellet. COCCINELLE is coupled to the 3D thermal-hydraulic code THYC with calculates TDoppler with is standard thermal model. In order to improve the accuracy of such calculations, we have developed the coupling of our two latest codes in thermal-hydraulics (THYC) and thermo-mechanics (CYRANO3). THYC calculates two-phase flows in pipes or rod bundles and is used for transient calculations such as steam-line break, boron dilution accidents, DNB predictions, steam generator and condenser studies. CYRANO3 calculates most of the phenomena that take place in the fuel such as: 1) heat transfer induced by nuclear power; 2) thermal expansion of the fuel and the cladding; 3) release of gaseous fission`s products; 4) mechanical interaction between the pellet and the cladding. These two codes are now qualified in their own field and the coupling, using Parallel Virtual Machine (PVM) libraries customized in an home-made-easy-to-use package called CALCIUM, has been validated on `low` configurations (no thermal expansion, constant thermal characteristics) and used on accidental transients such as rod ejection and loss of coolant accident. (K.A.) 7 refs.

  14. Thermal shock behavior of rare earth modified alumina ceramic composites

    Energy Technology Data Exchange (ETDEWEB)

    Sun, Junlong; Liu, Changxia [Ludong Univ., Yantai (China). School of Transportation

    2017-05-15

    Alumina matrix ceramic composites toughened by AlTiC master alloys, diopside and rare earths were fabricated by hot-pressing and their thermal shock behavior was investigated and compared with that of monolithic alumina. Results showed that the critical thermal shock temperature (ΔT) of monolithic alumina was 400 C. However, it decreased to 300 C for alumina incorporating only AlTiC master alloys, and increased with further addition of diopside and rare earths. Improvement of thermal shock resistance was obtained for alumina ceramic composites containing 9.5 wt.% AlTiC master alloys and 0.5 wt.% rare earth additions, which was mainly attributed to the formation of elongated grains in the composites.

  15. Thermal and electrical behavior of nano-modified cement mortar

    Science.gov (United States)

    Exarchos, D. A.; Dalla, P. T.; Tragazikis, I. K.; Alafogianni, P.; Barkoula, N.-M.; Paipetis, A. S.; Dassios, K. G.; Matikas, T. E.

    2014-04-01

    This research aims in characterizing modified cement mortar with carbon nanotubes (CNTs) that act as nanoreinforcements leading to the development of innovative materials possessing multi-functionality and smartness. Such multifunctional properties include enhanced mechanical behavior, electrical and thermal conductivity, and piezo-electric characteristics. The effective thermal properties of the modified nano-composites were evaluated using IR Thermography. The electrical resistivity was measured with a contact test method using a custom made apparatus and applying a known D.C. voltage. To eliminate any polarization effects the specimens were dried in an oven before testing. In this work, the thermal and electrical properties of the nano-modified materials were studied by nondestructively monitoring their structural integrity in real time using the intrinsic multi-functional properties of the material as damage sensors.

  16. Cultivar specific metabolic changes in grapevines berry skins in relation to deficit irrigation and hydraulic behavior.

    Science.gov (United States)

    Hochberg, Uri; Degu, Asfaw; Cramer, Grant R; Rachmilevitch, Shimon; Fait, Aaron

    2015-03-01

    Deficit irrigation techniques are widely used in commercial vineyards. Nevertheless, varieties respond differently to water availability, prompting the need to elucidate the physiological and molecular mechanisms involved in the interactions between genotypes and their environment. In the present study, the variability in berry metabolism under deficit irrigation was investigated in the field on Shiraz and Cabernet Sauvignon (CS), known for their hydraulic variability. Berry skin metabolite profiling of the two cultivars was performed by parallel GC-MS and LC-MS at four development stages. Under similar irrigation, the cultivars differed in stomata regulation. In response to water deficit, CS exhibited lessened loss in berry weight and milder metabolic alteration of berry-skin primary metabolites, as compared with Shiraz. The metabolic stress responses were shown to depend on berry phenology. Characteristic metabolic changes included a decrease in amino acids and TCA cycle intermediates from veraison onward. In contrast, water deficit induced the accumulation of stress-related metabolites such as: proline, beta-alanine, raffinose, nicotinate and ascorbate, to a greater extent in Shiraz. Polyphenol metabolism in response to water stress also underwent significant changes, unique to each cultivar. Results suggest a link between the vine hydraulics and water-deficit driven changes in the berry skin metabolism, with significant consequences on the metabolic composition of the fruit. Copyright © 2015 Elsevier Masson SAS. All rights reserved.

  17. Assessing swine thermal comfort by image analysis of postural behaviors.

    Science.gov (United States)

    Xin, H

    1999-01-01

    Postural behavior is an integral response of animals to complex environmental factors. Huddling, nearly contacting one another on the side, and spreading are common postural behaviors of group-housed animals undergoing cold, comfortable, and warm/hot sensations, respectively. These postural patterns have been routinely used by animal caretakers to assess thermal comfort of the animals and to make according adjustment on the environmental settings or management schemes. This manual adjustment approach, however, has the inherent limitations of daily discontinuity and inconsistency between caretakers in interpretation of the animal comfort behavior. The goal of this project was to explore a novel, automated image analysis system that would assess the thermal comfort of swine and make proper environmental adjustments to enhance animal wellbeing and production efficiency. This paper describes the progress and on-going work toward the achievement of our proposed goal. The feasibility of classifying the thermal comfort state of young pigs by neural network (NN) analysis of their postural images was first examined. It included exploration of using certain feature selections of the postural behavioral images as the input to a three-layer NN that was trained to classify the corresponding thermal comfort state as being cold, comfortable, or warm. The image feature selections, a critical step for the classification, examined in this study included Fourier coefficient (FC), moment (M), perimeter and area (P&A), and combination of M and P&A of the processed binary postural images. The result was positive, with the combination of M and P&A as the input feature to the NN yielding the highest correct classification rate. Subsequent work included the development of hardware and computational algorithms that enable automatic image segmentation, motion detection, and the selection of the behavioral images suitable for use in the classification. Work is in progress to quantify the

  18. Thermal Fatigue Behavior of Air-Plasma Sprayed Thermal Barrier Coating with Bond Coat Species in Cyclic Thermal Exposure

    Directory of Open Access Journals (Sweden)

    Ungyu Paik

    2013-08-01

    Full Text Available The effects of the bond coat species on the delamination or fracture behavior in thermal barrier coatings (TBCs was investigated using the yclic thermal fatigue and thermal-shock tests. The interface microstructures of each TBC showed a good condition without cracking or delamination after flame thermal fatigue (FTF for 1429 cycles. The TBC with the bond coat prepared by the air-plasma spray (APS method showed a good condition at the interface between the top and bond coats after cyclic furnace thermal fatigue (CFTF for 1429 cycles, whereas the TBCs with the bond coats prepared by the high-velocity oxygen fuel (HVOF and low-pressure plasma spray (LPPS methods showed a partial cracking (and/or delamination and a delamination after 780 cycles, respectively. The TBCs with the bond coats prepared by the APS, HVOF and LPPS methods were fully delaminated (>50% after 159, 36, and 46 cycles, respectively, during the thermal-shock tests. The TGO thickness in the TBCs was strongly dependent on the both exposure time and temperature difference tested. The hardness values were found to be increased only after the CFTF, and the TBC with the bond coat prepared by the APS showed the highest adhesive strength before and after the FTF.

  19. A simple measuring technique of surface flow velocity to analyze the behavior of velocity fields in hydraulic engineering applications.

    Science.gov (United States)

    Tellez, Jackson; Gomez, Manuel; Russo, Beniamino; Redondo, Jose M.

    2015-04-01

    An important achievement in hydraulic engineering is the proposal and development of new techniques for the measurement of field velocities in hydraulic problems. The technological advances in digital cameras with high resolution and high speed found in the market, and the advances in digital image processing techniques now provides a tremendous potential to measure and study the behavior of the water surface flows. This technique was applied at the Laboratory of Hydraulics at the Technical University of Catalonia - Barcelona Tech to study the 2D velocity fields in the vicinity of a grate inlet. We used a platform to test grate inlets capacity with dimensions of 5.5 m long and 4 m wide allowing a zone of useful study of 5.5m x 3m, where the width is similar of the urban road lane. The platform allows you to modify the longitudinal slopes from 0% to 10% and transversal slope from 0% to 4%. Flow rates can arrive to 200 l/s. In addition a high resolution camera with 1280 x 1024 pixels resolution with maximum speed of 488 frames per second was used. A novel technique using particle image velocimetry to measure surface flow velocities has been developed and validated with the experimental data from the grate inlets capacity. In this case, the proposed methodology can become a useful tools to understand the velocity fields of the flow approaching the inlet where the traditional measuring equipment have serious problems and limitations. References DigiFlow User Guide. (2012), (June). Russo, B., Gómez, M., & Tellez, J. (2013). Methodology to Estimate the Hydraulic Efficiency of Nontested Continuous Transverse Grates. Journal of Irrigation and Drainage Engineering, 139(10), 864-871. doi:10.1061/(ASCE)IR.1943-4774.0000625 Teresa Vila (1), Jackson Tellez (1), Jesus Maria Sanchez (2), Laura Sotillos (1), Margarita Diez (3, 1), and J., & (1), M. R. (2014). Diffusion in fractal wakes and convective thermoelectric flows. Geophysical Research Abstracts - EGU General Assembly 2014

  20. Feasibility Test of Commercial CFD Code for Thermal-Hydraulic Analysis of Wire-wrapped Fuel Pin Bundle in SFR

    Energy Technology Data Exchange (ETDEWEB)

    Choi, Chiwoong; Ha, Kwiseok [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Haeyong, Jeong [Sejong Univ., Daejeon (Korea, Republic of)

    2014-05-15

    One of the major difficulties in the numerical approach is a grid generation of the interface between a wire and rod, because the point connectivity and higher scale deviation of a wire and rod can easily increase the number of meshes. The general way to handle this problem is a simplification of the wire-rod interface. In this study, the wire configuration effect is also studied with a no-wire case and simplified wire geometries. The materials of coolant and cladding are sodium and HT9. Fuel part is not modeled. The detailed design parameters are described in Table 1. In addition, Fig. 2 shows a detailed schematic of the fuel rod and wire. The wire contact distance, Sw, is the length penetrating the clad. Basically, the wire will be firmly contacted with the clad surface. As shown in Table 1, the wire diameter is the same as the gap between two fuel rods. However, it is unavoidable in the modeling of the wire to ignore a singular contact point between the wire and cladding. Therefore, an intersected wire and rod are arbitrarily generated. The boundary conditions for the conjugated heat transfer analysis are described in Table 2. The inlet region is defined with a uniform velocity and constant temperature of 650 K. The outlet is defined with a constant ambient pressure, i. e., zero static pressure. Since only the cladding part was modeled, a uniform heat flux condition is applied on the inner cladding wall surface. All solid surfaces are considered as no slip adiabatic boundaries. The numerical analyses of wire-wrapped 7 pins are conducted using ANSYS-CFX to check the feasibility of the CFD tool in thermal-hydraulic phenomena in a wire-wrapped fuel pin bundle. Typical turbulent models are applied to check the dependency of model selection on the pressure drop and heat transfer on the wire-wrapped fuel rod. In short, the omega-based model shows the highest pressure drop and heat transfer. Comparing the existing correlations, the pressure drop results represent

  1. Safety analysis of RBMK design basis accidents by coupled neutronics thermal hydraulics codes QUABOX/CUBBOX-ATHLET and SADCO-ATHLET

    Energy Technology Data Exchange (ETDEWEB)

    Clemente, M.; Langenbuch, S.; Velkov, K. [Gesellschaft fuer Anlagen- und Reaktorsicherheit mbH (GRS), Garching (Germany); Kuznetcov, P.; Panin, V.; Roghdestvensky, M.; Sakharova, T.; Stenbock, I. [Research and Development Inst. of Power Engineering (RDIPE), Moscow (Russian Federation)

    2006-07-01

    One of the most important issues of safety analysis is the validation and verification (V and V) of coupled neutronics/thermal-hydraulics codes used to study NPP transients and accidental processes. Risks and difficulties of a straight experimental investigation of such regimes of operating NPPs for V and V purposes give rise to the application of coupled 3D code packages cross-verification. - Codes for V and V: Thermal hydraulics and fluid mechanics calculations were performed by the ATHLET code (Analysis of Thermal-hydraulics of Leaks and Transient). Right and left loops of RBMK Main Coolant Circuit (MCC) connected by a steam line system were simulated. The following elements of each loop are reproduced as one group each: 2 Steam Drums, Downcomers and 3 Main Coolant Pumps. 22 Distribution Group Headers are modeled as one group for RIAs; however for LOCAs accidental and neighbouring channels are described separately. Fuel Channels (and the Lower Water Lines) are grouped according to the reactor core power distribution. In a first approach the core power in the ATHLET2.0A input model was simulated by a point kinetics model, which was replaced later by the 3D-core models SADCO from NIKIET and QUABOX/CUBBOX (Q/C) from GRS. - Results of Calculations: A comparison of Q/C and SADCO-ATHLET calculations for the NPP Kursk-1 are presented as example of 3D coupled codes cross-verification, showing a rod withdrawal accident. In the ATHLET model the RBMK core is represented by 6 regions, however the fuel channels around the accidental rod were analyzed separately, resulting in a total number of 79 thermal hydraulic channels in ATHLET to describe the core. The detailed description of modernized control and protection system (CPS) was included in the model. - Conclusions: 1. The results show that coupled codes cross-verification provides a reliable methodology of testing such codes for transient and accident analyses. 2. ATHLET is an effective calculational tool for analyses of

  2. Groundwater Flow and Thermal Modeling to Support a Preferred Conceptual Model for the Large Hydraulic Gradient North of Yucca Mountain

    Energy Technology Data Exchange (ETDEWEB)

    McGraw, D.; Oberlander, P.

    2007-12-18

    The purpose of this study is to report on the results of a preliminary modeling framework to investigate the causes of the large hydraulic gradient north of Yucca Mountain. This study builds on the Saturated Zone Site-Scale Flow and Transport Model (referenced herein as the Site-scale model (Zyvoloski, 2004a), which is a three-dimensional saturated zone model of the Yucca Mountain area. Groundwater flow was simulated under natural conditions. The model framework and grid design describe the geologic layering and the calibration parameters describe the hydrogeology. The Site-scale model is calibrated to hydraulic heads, fluid temperature, and groundwater flowpaths. One area of interest in the Site-scale model represents the large hydraulic gradient north of Yucca Mountain. Nearby water levels suggest over 200 meters of hydraulic head difference in less than 1,000 meters horizontal distance. Given the geologic conceptual models defined by various hydrogeologic reports (Faunt, 2000, 2001; Zyvoloski, 2004b), no definitive explanation has been found for the cause of the large hydraulic gradient. Luckey et al. (1996) presents several possible explanations for the large hydraulic gradient as provided below: The gradient is simply the result of flow through the upper volcanic confining unit, which is nearly 300 meters thick near the large gradient. The gradient represents a semi-perched system in which flow in the upper and lower aquifers is predominantly horizontal, whereas flow in the upper confining unit would be predominantly vertical. The gradient represents a drain down a buried fault from the volcanic aquifers to the lower Carbonate Aquifer. The gradient represents a spillway in which a fault marks the effective northern limit of the lower volcanic aquifer. The large gradient results from the presence at depth of the Eleana Formation, a part of the Paleozoic upper confining unit, which overlies the lower Carbonate Aquifer in much of the Death Valley region. The

  3. INVESTIGATION OF HYDRAULIC OPERATIONAL REGIMES OF CIRCULATING SYSTEM AT THERMAL POWER PLANT OF VOLZHSKY AUTOMOTIVE WORKS USING COMPUTER MODEL

    Directory of Open Access Journals (Sweden)

    V. V. Dikop

    2005-01-01

    Full Text Available Investigation results of hydraulic operational regimes of a circulating system with the help of a computer model are presented in the paper. The models simulates hydraulic processes by means of an iterated solution of a set of algebraic non-linear equations that is formed while using a graph theory The circulating system is considered as a unit in the model. The model program makes it possible to calculate consumption and pressure at any point of the circulating system with indication of flow motion directions along its separate branches and also to execute some economical calculations.

  4. Linking nonstructural carbohydrate dynamics to gas exchange and leaf hydraulic behavior in Pinus edulis and Juniperus monosperma

    Science.gov (United States)

    David R. Woodruff; Frederick C. Meinzer; Danielle E. Marias; Sanna Sevanto; Michael W. Jenkins; Nate G. McDowell

    2014-01-01

    Leaf hydraulics, gas exchange and carbon storage in Pinus edulis and Juniperus monosperma, two tree species on opposite ends of the isohydry–anisohydry spectrum, were analyzed to examine relationships between hydraulic function and carbohydrate dynamics.Leaf hydraulic vulnerability,...

  5. Uncertainty analysis for results of thermal hydraulic codes of best-estimate-type; Analisis de incertidumbre para resultados de codigos termohidraulicos de mejor estimacion

    Energy Technology Data Exchange (ETDEWEB)

    Alva N, J.

    2010-07-01

    In this thesis, some fundamental knowledge is presented about uncertainty analysis and about diverse methodologies applied in the study of nuclear power plant transient event analysis, particularly related to thermal hydraulics phenomena. These concepts and methodologies mentioned in this work come from a wide bibliographical research in the nuclear power subject. Methodologies for uncertainty analysis have been developed by quite diverse institutions, and they have been widely used worldwide for application to results from best-estimate-type computer codes in nuclear reactor thermal hydraulics and safety analysis. Also, the main uncertainty sources, types of uncertainties, and aspects related to best estimate modeling and methods are introduced. Once the main bases of uncertainty analysis have been set, and some of the known methodologies have been introduced, it is presented in detail the CSAU methodology, which will be applied in the analyses. The main objective of this thesis is to compare the results of an uncertainty and sensibility analysis by using the Response Surface Technique to the application of W ilks formula, apply through a loss coolant experiment and an event of rise in a BWR. Both techniques are options in the part of uncertainty and sensibility analysis of the CSAU methodology, which was developed for the analysis of transients and accidents at nuclear power plants, and it is the base of most of the methodologies used in licensing of nuclear power plants practically everywhere. Finally, the results of applying both techniques are compared and discussed. (Author)

  6. Innovative and Advanced Coupled Neutron Transport and Thermal Hydraulic Method (Tool) for the Design, Analysis and Optimization of VHTR/NGNP Prismatic Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Rahnema, Farzad; Garimeela, Srinivas; Ougouag, Abderrafi; Zhang, Dingkang

    2013-11-29

    This project will develop a 3D, advanced coarse mesh transport method (COMET-Hex) for steady- state and transient analyses in advanced very high-temperature reactors (VHTRs). The project will lead to a coupled neutronics and thermal hydraulic (T/H) core simulation tool with fuel depletion capability. The computational tool will be developed in hexagonal geometry, based solely on transport theory without (spatial) homogenization in complicated 3D geometries. In addition to the hexagonal geometry extension, collaborators will concurrently develop three additional capabilities to increase the code’s versatility as an advanced and robust core simulator for VHTRs. First, the project team will develop and implement a depletion method within the core simulator. Second, the team will develop an elementary (proof-of-concept) 1D time-dependent transport method for efficient transient analyses. The third capability will be a thermal hydraulic method coupled to the neutronics transport module for VHTRs. Current advancements in reactor core design are pushing VHTRs toward greater core and fuel heterogeneity to pursue higher burn-ups, efficiently transmute used fuel, maximize energy production, and improve plant economics and safety. As a result, an accurate and efficient neutron transport, with capabilities to treat heterogeneous burnable poison effects, is highly desirable for predicting VHTR neutronics performance. This research project’s primary objective is to advance the state of the art for reactor analysis.

  7. The behavior of multilayer ceramic protections at quick thermal shock

    Directory of Open Access Journals (Sweden)

    Alexandru MIHAILESCU

    2013-06-01

    Full Text Available Protective layers of “hot parts” of the turbo engines as well as co-generative systems of energy industry are exposed to a combination of wear factors which may act together at high values.The main goal of the paper is the behavior of some advanced layers, duplex and triplex, multifunctional, ceramics in relation to the most complex wear factor and disturbing as well, the quick thermal shock.The quick thermal shock test installation designed and constructed by the INCAS covers the domain of some high gradients of heating/cooling and is currently integrated in a network of European infrastructure that evaluates the properties of functional layers for turbo engines.Micro-structure inter- and intra- facial changes gradually induced in ceramic structures are highlighted and on this basis their ranking and selection for application on physical parts are established.

  8. Lipophilic phytosterol derivatives: synthesis, thermal property and nanoemulsion behavior

    DEFF Research Database (Denmark)

    Panpipat, Worawan; Xu, Xuebing; Guo, Zheng

    the productive yield of a series of -sitosteryl fatty acid esters (C2-C18) and to investigate the thermal property and nano-emulsion behaviors of those compounds. This work reported a novel approach to synthesize phytosterol (-sitosterol as a model) fatty acid ester by employing Candida antarctica lipase...... of -sitosterol fatty acid esters was governed by the carbon chain length of fatty acid incorporated. The nano-emulsions of a series of -sitosteryl fatty acid esters were prepared by probe-sonication method. The particle size distributions, zeta potentials and TEM images of those emulsions were different...

  9. Thermal Reactor Safety

    Energy Technology Data Exchange (ETDEWEB)

    1980-06-01

    Information is presented concerning fire risk and protection; transient thermal-hydraulic analysis and experiments; class 9 accidents and containment; diagnostics and in-service inspection; risk and cost comparison of alternative electric energy sources; fuel behavior and experiments on core cooling in LOCAs; reactor event reporting analysis; equipment qualification; post facts analysis of the TMI-2 accident; and computational methods.

  10. Coupling agents in therapeutic ultrasound: acoustic and thermal behavior.

    Science.gov (United States)

    Casarotto, Raquel Aparecida; Adamowski, Julio Cesar; Fallopa, Flávio; Bacanelli, Fernando

    2004-01-01

    To compare transmissivity data and thermal behavior of 4 coupling media. Experimental. Postgraduate rehabilitation program in Brazil. Four coupling media: gel, mineral oil, white petrolatum, and degassed water. Not applicable. The transmission, attenuation, reflection coefficient, and acoustic impedance of gel, mineral oil, white petrolatum, and degassed water were measured with a density measurement cell. The temperature variation in the therapeutic ultrasound transducer was measured with a thermocouple. The transmissivity data showed that the water and gel presented the highest transmission coefficient, the lowest reflection, and an attenuation coefficient and acoustic impedance close to that of the skin. The thermal data revealed the highest heating in the transducer during the insonation with white petrolatum and mineral oil, resulting from the thermal conductivity features of each medium. Transmissivity data obtained showed that water and gel present the best acoustic features. In ultrasound therapy, with the direct contact technique using thin layers of coupling agents, any product may be used, because the effect of the attenuation coefficient does not play a significant role when layers are as thin as those used in this experiment.

  11. ON THE THERMAL BEHAVIOR OF DIFFERENT TANNED BOVINE LEATHERS

    Directory of Open Access Journals (Sweden)

    VARGANICI Cristian-Dragoş

    2017-05-01

    Full Text Available Leather is one of the most globally spread biomaterial which is obtained by the processing of different animal skins. It encompasses a wide palette of applications, from footwear and clothing to upholsteries and different types of furniture [1], [2], [3]. The main constituent of animal skins is collagen, a supramolecular fibrillar protein in the form of a triple helix. This form endows leather with elasticity, good mechanical properties and softness. A major disadvantage resides in the inapplicability of raw animal hides, due to their microbiological instability and decay through rotting. Microbiological stability is obtained through the tanning process, characterized by protein crosslinking and drying afterwards. After tanning the leather exhibits the required properties for the desired specific applications in terms of aspect, availability and sustainability [4], [5]. The study aims to elucidate the thermal decomposition process of chrome-free tanned bovine hide (wet-white using a new product based on titanium and aluminium salts compared with the same hide tanned by chromium salts (wet-blue. The thermal behavior was studied by dynamic thermogravimetry in nitrogen atmosphere, up to 700 oC. A comparative thermal decomposition study between the different tanned bovine leathers was undertaken.

  12. Chlorine characterization and thermal behavior in MSW and RDF.

    Science.gov (United States)

    Ma, Wenchao; Hoffmann, Gaston; Schirmer, Mattias; Chen, Guanyi; Rotter, Vera Susanne

    2010-06-15

    Chlorine, as a key element causing high temperature corrosion and low efficiency in waste-to-energy plants, and its thermal behavior has widely drawn attention. In this study, the chlorine content in eight fractions of municipal solid waste (MSW) was quantified and characterized using five analytical methods. The influence of the operating temperature, and fuel additives like sulfur and silica on the volatilization of chlorine in combustion process was also investigated. The results showed: these fractions cover a wide range of chlorine content from 0.1 wt.% in wood to >6 wt.% in non-packaging plastics (dry basis). Polyvinylchloride (PVC) from packaging, electrical wire insulation etc. in plastics and chloride salts (mainly NaCl) in kitchen waste are the main sources of organic and inorganic chlorine. The increase of the operating temperature from 700 degrees C to 1000 degrees C has more influence on the HCl formation for kitchen waste than that for PVC. Sulfur addition leads to 20-40% higher HCl formation rate in most fractions. Silica supports the chlorine release at relatively low temperatures between 700 degrees C and 850 degrees C. These findings enhance to understand the thermal behavior of chlorine in MSW and RDF (refuse derived fuel) in waste-to-energy plants and lead to the suggestions for a fuel management for waste derived fuels in order to avoid chlorine induced corrosion. Copyright 2010 Elsevier B.V. All rights reserved.

  13. Synthesis and thermal behavior of polyacrylonitrile/vinylidene chloride copolymer

    Directory of Open Access Journals (Sweden)

    Robson Fleming

    2014-06-01

    Full Text Available Polyacrylonitrile fiber encompasses a broad range of products based on acrylonitrile (AN which is readily copolymerized with a wide range of ethylenic unsaturated monomers giving rise to polymers with different characteristics and applications. Such products can be designed for cost-effective, flame and heat resistant solutions for the textile industry, aircraft and automotive markets. In the present work acrylonitrile was copolymerized with vinylidene chloride (VDC by conventional suspension polymerization process via redox system, with an initial content of 10%/mass of the VDC monomer. The copolymer average molecular weight was obtained by Gel Permeation Chromatography (GPC and by intrinsic viscosity analysis. To control the polymerization process continuously, qualitative and quantitative analysis of the chloride content in the PAN AN/VDC copolymer structure was accomplished by using X-ray fluorescence and potentiometric titration techniques. A good correlation was found between these two techniques, leading to a straightforward verification of VDC in the polymer structure. The thermal behavior of PAN AN/VDC copolymer was performed by Differential Scanning Calorimetry (DSC and Thermogravimetric Analysis (TGA. The results showed that VDC monomers exhibited a nearly stoichiometric reaction with acrylonitrile, copolymerizing about 90% of its initial mass. VDC changed significantly the polyacrylonitrile thermal behavior, decreasing the polymer degradation temperature by about 40-50°C.

  14. Kinetic Study and Thermal Decomposition Behavior of Lignite Coal

    Directory of Open Access Journals (Sweden)

    Mehran Heydari

    2015-01-01

    Full Text Available A thermogravimetric analyzer was employed to investigate the thermal behavior and extract the kinetic parameters of Canadian lignite coal. The pyrolysis experiments were conducted in temperatures ranging from 298 K to 1173 K under inert atmosphere utilizing six different heating rates of 1, 6, 9, 12, 15, and 18 K min−1, respectively. There are different techniques for analyzing the kinetics of solid-state reactions that can generally be classified into two categories: model-fitting and model-free methods. Historically, model-fitting methods are broadly used in solid-state kinetics and show an excellent fit to the experimental data but produce uncertain kinetic parameters especially for nonisothermal conditions. In this work, different model-free techniques such as the Kissinger method and the isoconversional methods of Ozawa, Kissinger-Akahira-Sunose, and Friedman are employed and compared in order to analyze nonisothermal kinetic data and investigate thermal behavior of a lignite coal. Experimental results showed that the activation energy values obtained by the isoconversional methods were in good agreement, but Friedman method was considered to be the best among the model-free methods to evaluate kinetic parameters for solid-state reactions. These results can provide useful information to predict kinetic model of coal pyrolysis and optimization of the process conditions.

  15. Thermal-mechanical behavior of fuel element in SCWR design

    Energy Technology Data Exchange (ETDEWEB)

    Xu, R.; Yetisir, M.; Hamilton, H. [Atomic Energy of Canada Limited, Chalk River, ON (Canada)

    2014-07-01

    This paper presents a study on thermal-mechanical behavior of a fuel element proposed for the Canadian Supercritical Water Cooled Reactor (SCWR). In the Canadian SCWR, the coolant pressure is 25 MPa, and the temperature is 350{sup o}C at the inlet and 625{sup o}C at the outlet of the reactor core. Critical design decisions for fuel design will be the selection of the fuel sheath material and details of the fuel element design options (sheath thickness, pellet-clad gap, internal pressure, etc.). The analysis presented in this paper predicted temperature, stress and strain in the fuel element of the Canadian SCWR with a collapsible sheath using ANSYS. Typical conditions for the evaluation of the fuel behavior, such as linear heat generation rate, coolant temperature and sheath surface heat transfer coefficient, were extracted from core and fuel channel designs. The temperature distribution in the fuel element is predicted by a thermal model and then the thermal model is coupled sequentially with a structural model to predict fuel sheath deformation under the predicted temperature distribution and external (coolant) pressure. Nonlinear thermo-mechanical simulations include nonlinear buckling with elastic-plastic deformation. Three sheath collapse phenomena are considered: (1) elastic collapse by buckling, (2) longitudinal ridging and (3) plastic collapse by yielding. The numerical models are validated against analytical and experimental data. The presented results show the temperature distribution, deformed shape, stress and strain of the fuel element, allowing the designers to select appropriate sheath material and element design options for the SCWR fuel element design. (author)

  16. Computer code analysis of steam generator in thermal-hydraulic test facility simulating nuclear power plant; Ydinvoimalaitosta kuvaavan koelaitteiston hoeyrystimien analysointi tietokoneohjelmilla

    Energy Technology Data Exchange (ETDEWEB)

    Virtanen, E.

    1995-12-31

    In the study three loss-of-feedwater type experiments which were preformed with the PACTEL facility has been calculated with two computer codes. The purpose of the experiments was to gain information about the behaviour of horizontal steam generator in a situation where the water level on the secondary side of the steam generator is decreasing. At the same time data that can be used in the assessment of thermal-hydraulic computer codes was assembled. The purpose of the work was to study the capabilities of two computer codes, APROS version 2.11 and RELAP5/MOD3.1, to calculate the phenomena in horizontal steam generator. In order to make the comparison of the calculation results easier the same kind of model of the steam generator was made for both codes. Only the steam generator was modelled, the rest of the facility was given for the codes as a boundary condition. (23 refs.).

  17. Applicability of RELAP5-3D for Thermal-Hydraulic Analyses of a Sodium-Cooled Actinide Burner Test Reactor

    Energy Technology Data Exchange (ETDEWEB)

    C. B. Davis

    2006-07-01

    The Actinide Burner Test Reactor (ABTR) is envisioned as a sodium-cooled, fast reactor that will burn the actinides generated in light water reactors to reduce nuclear waste and ease proliferation concerns. The RELAP5-3D computer code is being considered as the thermal-hydraulic system code to support the development of the ABTR. An evaluation was performed to determine the applicability of RELAP5-3D for the analysis of a sodium-cooled fast reactor. The applicability evaluation consisted of several steps, including identifying the important transients and phenomena expected in the ABTR, identifying the models and correlations that affect the code’s calculation of the important phenomena, and evaluating the applicability of the important models and correlations for calculating the important phenomena expected in the ABTR. The applicability evaluation identified code improvements and additional models needed to simulate the ABTR. The accuracy of the calculated thermodynamic and transport properties for sodium was also evaluated.

  18. INVESTIGATION OF THERMAL BEHAVIOR OF MULTILAYERED FIRE RESISTANT STRUCTURE

    Directory of Open Access Journals (Sweden)

    R. GUOBYS

    2016-09-01

    Full Text Available This paper presents experimental and numerical investigations of thermal behavior under real fire conditions of new generation multilayered fire resistant structure (fire door, dimensions H × W × D: 2090 × 980 × 52 mm combining high strength and fire safety. This fire door consists of two steel sheets (thickness 1.5 and 0.7 mm with stone wool ( = 33 kg/m3, k = 0.037 W/mK, E = 5000 N/m2,  = 0.2 insulating layer in between. One surface of the structure was heated in fire furnace for specified period of time of 60 min. Temperature and deformation of opposite surface were measured from outside at selected measuring points during fire resistance test. Results are presented as temperature-time and thermal deformation-time graphs. Experimental results were compared with numerical temperature field simulation results obtained from SolidWorks®Simulation software. Numerical results were found to be in good agreement with experimental data. The percent differences between door temperatures from simulation and fire resistance test don’t exceed 8%. This shows that thermal behaviour of such multilayered structures can be investigated numerically, thus avoiding costly and time-consuming fire resistance tests. It is established that investigated structure should be installed in a way that places thicker steel sheet closer to the potential heat source than thinner one. It is also obtained that stone wool layer of higher density should be used to improve fire resistance of the structure.

  19. Synthesis and thermal behavior of double copper and potassium pyrophosphate

    Energy Technology Data Exchange (ETDEWEB)

    Ciopec, Mihaela [Faculty of Industrial Chemistry and Environmental Engineering, University ' Politehnica' Timisoara, Piata Victoriei no. 2, 300006 Timisoara (Romania); Muntean, Cornelia, E-mail: cornelia.muntean@chim.upt.ro [Faculty of Industrial Chemistry and Environmental Engineering, University ' Politehnica' Timisoara, Piata Victoriei no. 2, 300006 Timisoara (Romania); Negrea, Adina; Lupa, Lavinia; Negrea, Petru [Faculty of Industrial Chemistry and Environmental Engineering, University ' Politehnica' Timisoara, Piata Victoriei no. 2, 300006 Timisoara (Romania); Barvinschi, Paul [Faculty of Physics, West University of Timisoara, Bv. Vasile Parvan no. 4, 300223 Timisoara (Romania)

    2009-05-05

    This paper presents the synthesis and thermal behavior of double copper and potassium pyrophosphate, which can be used as a PK fertilizer containing copper as micronutrient. In order to find the conditions for the synthesis of this compound from copper sulphate and potassium pyrophosphate, various Cu{sup 2+}:P{sub 2}O{sub 7}{sup 4-} molar ratios (0:1-2:1), various molar concentrations of the solutions (0.075; 0.1; 0.15 and 0.2 mol L{sup -1}) and various temperatures (25, 50, 75 and 100 {sup o}C) have been used. The solid product synthesized in optimum conditions for the separation of micronutrient copper from the reaction mass (Cu{sup 2+}:P{sub 2}O{sub 7}{sup 4-} molar ratio 1:1, concentration 0.1 mol L{sup -1}) was subjected to a complex study: chemical analysis, thermal analysis, energy dispersive X-ray spectroscopy, scanning electron microscopy and X-ray diffractometry. During heating up to 1000 {sup o}C, K{sub 2}Cu{sub 3}(P{sub 2}O{sub 7}){sub 2}.3H{sub 2}O loses the crystallization water; several transformations of the phosphates also take place: the decomposition of pyrophosphates to ortho-phosphates; the transformation of ortho-phosphates; the polymerization of a fraction of ortho-phosphates to amorphous phosphates with longer chains; the reorganization of ortho-phosphates and poly-phosphates to pyrophosphates and their crystallization. The decomposition mechanism was confirmed when using the X-ray diffraction patterns of the compound, thermally treated at several temperatures.

  20. Hydraulic structures

    CERN Document Server

    Chen, Sheng-Hong

    2015-01-01

    This book discusses in detail the planning, design, construction and management of hydraulic structures, covering dams, spillways, tunnels, cut slopes, sluices, water intake and measuring works, ship locks and lifts, as well as fish ways. Particular attention is paid to considerations concerning the environment, hydrology, geology and materials etc. in the planning and design of hydraulic projects. It also considers the type selection, profile configuration, stress/stability calibration and engineering countermeasures, flood releasing arrangements and scouring protection, operation and maintenance etc. for a variety of specific hydraulic structures. The book is primarily intended for engineers, undergraduate and graduate students in the field of civil and hydraulic engineering who are faced with the challenges of extending our understanding of hydraulic structures ranging from traditional to groundbreaking, as well as designing, constructing and managing safe, durable hydraulic structures that are economical ...

  1. Bistability behavior of thermally actuated micro-bridge

    Science.gov (United States)

    Michael, Aron; Yu, Kevin; Kwok, Chee Yee

    2005-02-01

    A multi-layered micro-bridge buckles due to residual stresses in the layers of the beam when it is released during fabrication, and the axial load due to this stress exceeds the Euler load. This residual stress renders intrinsic bi-stability behavior to the bridge. In this paper, the effect of axial, rotational stiffness, and residual moment on the buckled shape, snapping, and bi-stability of multi-layered bridge when it is thermally actuated is studied both theoretically, and experimentally. Theoretical analysis, and ANSYS finite element simulation have been carried out to investigate these effects. Deflection versus temperature plots for different axial, rotational stiffness, and residual moment are obtained. The theoretical investigations are applied to bi-morph micro-bridges of 1000um length, and 40um wide made of PECVD silicon dioxide, and epi-taxial silicon, and a tri-layer structure of Poly/SiO2/epi-silicon. The bi-layer structure is fabricated, and its buckled shape is obtained from SEM. Results show that axial, rotational stiffness, and residual moment strongly affect the buckled shape, and bi-stability of the micro-bridge. It is also shown that for thermally actuated micro-bridge, better bi-stability, and snapping characteristics can be obtained when both rotational and axial stiffnesses are reduced, and the residual moment must not exceed a certain threshold value if bi-stability is to be preserved.

  2. Investigation of the thermal behavior of Philippine natural zeolites

    Science.gov (United States)

    Olegario-Sanchez, Eleanor; Felizco, Jenichi Clairvaux; Mulimbayan, Francis

    2017-12-01

    This study aims to investigate the thermal behavior of natural zeolites from Mangatarem, Pangasinan, Philippines. Results obtained from the XRD pattern revealed that the sample is largely composed of mordenite, a Type I zeolite. In the chemical analysis, Si and Al are proven to be the primary components of the sample, which is typical in an aluminosilicate material. It also has a high Si/Al ratio, which suggests that the material is hydrophilic and thermally stable. The TGA analysis showed that dehydration processes dominate the sample's reaction upon heating up to 1000 °C. The initial dehydration process led to a weight loss that is almost equivalent to the moisture content, as measured by chemical analysis. The succeeding weight loss is most probably due to the dehydration of tightly-bound water and cation dehydration. No structural collapse was observed even up to 1000 °C, which suggests that the material is ideal for industrial applications. Lastly, FTIR results revealed that heating the zeolitic tuff led to possible de-alumination.

  3. Studies on Thermal Degradation Behavior of Siliceous Agriculture Waste (Rice Husk, Wheat Husk and Bagasse)

    National Research Council Canada - National Science Library

    Syed H. Javed; Umair Aslam; Mohsin Kazmi; Masooma Rustam; Sheema Riaz; Zahid Munir

    2015-01-01

    Various siliceous agriculture waste (SAW) such as rice husk, wheat husk and bagasse have been investigated to study their thermal degradation behavior using Thermogravimetric Analyzer (TGA) technique...

  4. Thermal expansion behavior of fluor-chlorapatite crystalline solutions

    Science.gov (United States)

    Hovis, G.; Harlov, D.; Gottschalk, M.; Hudacek, W.; Wildermuth, S.

    2009-04-01

    Apatite Ca5(PO4)3(F,Cl,OH,CO3) occurs widely as an accessory mineral in many igneous and metamorphic rocks and in nature displays a wide range of F-Cl-OH-CO3 mixtures (e.g., O'Reilly and Griffin, 2000) that have been used to interpret the role of fluids, e.g. Cl, F, and OH activities, during metamorphic and igneous processes (e.g., Harlov and Förster, 2002). It is important, therefore, to understand the thermodynamic behavior of these solid solutions, including their thermal expansion properties. Fluorapatite - chlorapatite samples were synthesized at the GFZ-Potsdam (Hovis, Harlov, Hahn and Steigert, 2007) using an adaptation of the molten flux method of Cherniak (2000). Dry CaF2 and CaCl2 (0.1 mole total) were mixed with Ca3(PO4)2 (0.03 moles), placed in a Pt crucible, equilibrated for 15 hours at 1375 °C, cooled to 1220 °C at 3 °C/hour, removed from the oven and cooled in air. Crystals were separated from the flux by boiling the quenched product in water. F:Cl fractions for each sample were determined via Rietveld refinement of X-ray powder diffraction data. Chemical homogeneity was confirmed by Rietveld refinement and high-contrast back-scattered electron imaging. Room-temperature unit-cell volumes were determined at the GFZ-Potsdam through Rietveld analysis of X-ray powder diffraction data and also at Lafayette College by standard unit-cell refinement techniques (Holland and Redfern, 1997) using NBS/NIST 640a Si as an internal standard. High-temperature unit-cell dimensions were calculated from X-ray powder diffraction data collected at Cambridge University from room temperature to 1000 °C on a Bruker D8 X-ray diffractometer. NBS Si again was utilized as an internal standard; high-temperature Si peak positions were taken from Parrish (1953). Results indicate that despite the considerable size difference between fluorine and chlorine ions, reflected by substantially different unit-cell sizes at room temperature, the coefficient of thermal expansion across

  5. Thermal-hydraulics verification of a coarse-mesh OpenFOAM-based solver for a Sodium Fast Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Bonet López, M.

    2015-07-01

    Recently, in the Institute Swiss Paul Scherrer Institut, is has developed a platform Multiphysics, based in OpenFOAM, that is capable of performing an analysis multidimensional of a reactor nuclear. One of the main objectives of this project is to verify the part of the code responsible for the Thermo-hydraulic analysis of the reactor. To carry out simulations this part of the code uses the approximation of thick mesh based on the equations of a porous medium. Therefore, the other objective is demonstrate that this method is applicable to the analysis of a reactor nuclear fast of sodium, focusing is in his capacity of predict the transfer of heat between a subset and the space vacuum between subsets of the core of the reactor. (Author)

  6. Thermal Behavior of Cylindrical Buckling Restrained Braces at Elevated Temperatures

    Directory of Open Access Journals (Sweden)

    Elnaz Talebi

    2014-01-01

    Full Text Available The primary focus of this investigation was to analyze sequentially coupled nonlinear thermal stress, using a three-dimensional model. It was meant to shed light on the behavior of Buckling Restraint Brace (BRB elements with circular cross section, at elevated temperature. Such bracing systems were comprised of a cylindrical steel core encased in a strong concrete-filled steel hollow casing. A debonding agent was rubbed on the core’s surface to avoid shear stress transition to the restraining system. The numerical model was verified by the analytical solutions developed by the other researchers. Performance of BRB system under seismic loading at ambient temperature has been well documented. However, its performance in case of fire has yet to be explored. This study showed that the failure of brace may be attributed to material strength reduction and high compressive forces, both due to temperature rise. Furthermore, limiting temperatures in the linear behavior of steel casing and concrete in BRB element for both numerical and analytical simulations were about 196°C and 225°C, respectively. Finally it is concluded that the performance of BRB at elevated temperatures was the same as that seen at room temperature; that is, the steel core yields prior to the restraining system.

  7. Characterization of thermal-hydraulic and ignition phenomena in prototypic, full-length boiling water reactor spent fuel pool assemblies after a complete loss-of-coolant accident.

    Energy Technology Data Exchange (ETDEWEB)

    Lindgren, Eric Richard; Durbin, Samuel G

    2007-04-01

    The objective of this project was to provide basic thermal-hydraulic data associated with a SFP complete loss-of-coolant accident. The accident conditions of interest for the SFP were simulated in a full-scale prototypic fashion (electrically-heated, prototypic assemblies in a prototypic SFP rack) so that the experimental results closely represent actual fuel assembly responses. A major impetus for this work was to facilitate code validation (primarily MELCOR) and reduce questions associated with interpretation of the experimental results. It was necessary to simulate a cluster of assemblies to represent a higher decay (younger) assembly surrounded by older, lower-power assemblies. Specifically, this program provided data and analysis confirming: (1) MELCOR modeling of inter-assembly radiant heat transfer, (2) flow resistance modeling and the natural convective flow induced in a fuel assembly as it heats up in air, (3) the potential for and nature of thermal transient (i.e., Zircaloy fire) propagation, and (4) mitigation strategies concerning fuel assembly management.

  8. Thermal-hydraulics of helium cooled First Wall channels and scoping investigations on performance improvement by application of ribs and mixing devices

    Energy Technology Data Exchange (ETDEWEB)

    Arbeiter, Frederik, E-mail: frederik.arbeiter@kit.edu [Karlsruhe Institute of Technology, Institute of Neutron Physics and Reactor Technology, Hermann-von-Helmholtz-Platz 1, 76344 Eggenstein-Leopoldshafen (Germany); Bachmann, Christian [EUROfusion – Programme Management Unit, Garching (Germany); Chen, Yuming; Ilić, Milica; Schwab, Florian [Karlsruhe Institute of Technology, Institute of Neutron Physics and Reactor Technology, Hermann-von-Helmholtz-Platz 1, 76344 Eggenstein-Leopoldshafen (Germany); Sieglin, Bernhard [Max-Planck-Institut für Plasmaphysik, Garching (Germany); Wenninger, Ronald [EUROfusion – Programme Management Unit, Garching (Germany)

    2016-11-01

    Highlights: • Existing first wall designs and expected plasma heat loads are reviewed. • Heat transfer enhancement methods are investigated by CFD. • The results for heat transfer and friction are given, compared and explained. • Relations for needed pumping power and gained thermal heat are shown. • A range for the maximum permissible heat loads from the plasma is estimated. - Abstract: The first wall (FW) of DEMO is a component with high thermal loads. The cooling of the FW has to comply with the material's upper and lower temperature limits and requirements from stress assessment, like low temperature gradients. Also, the cooling has to be integrated into the balance-of-plant, in a sense to deliver exergy to the power cycle and require a limited pumping power for coolant circulation. This paper deals with the basics of FW cooling and proposes optimization approaches. The effectiveness of several heat transfer enhancement techniques is investigated for the use in helium cooled FW designs for DEMO. Among these are wall-mounted ribs, large scale mixing devices and modified hydraulic diameter. Their performance is assessed by computational fluid dynamics (CFD), and heat transfer coefficients and pressure drop are compared. Based on the results, an extrapolation to high heat fluxes is tried to estimate the higher limits of cooling capabilities.

  9. Proceedings of the Twenty-First Water Reactor Safety Information Meeting: Volume 1, Plenary session; Advanced reactor research; advanced control system technology; advanced instrumentation and control hardware; human factors research; probabilistic risk assessment topics; thermal hydraulics; thermal hydraulic research for advanced passive LWRs

    Energy Technology Data Exchange (ETDEWEB)

    Monteleone, S. [Brookhaven National Lab., Upton, NY (United States)] [comp.

    1994-04-01

    This three-volume report contains 90 papers out of the 102 that were presented at the Twenty-First Water Reactor Safety Information Meeting held at the Bethesda Marriott Hotel, Bethesda, Maryland, during the week of October 25--27, 1993. The papers are printed in the order of their presentation in each session and describe progress and results of programs in nuclear safety research conducted in this country and abroad. Foreign participation in the meeting included papers presented by researchers from France, Germany, Japan, Russia, Switzerland, Taiwan, and United Kingdom. The titles of the papers and the names of the authors have been updated and may differ from those that appeared in the final program of the meeting. Individual papers have been cataloged separately. This document, Volume 1 covers the following topics: Advanced Reactor Research; Advanced Instrumentation and Control Hardware; Advanced Control System Technology; Human Factors Research; Probabilistic Risk Assessment Topics; Thermal Hydraulics; and Thermal Hydraulic Research for Advanced Passive Light Water Reactors.

  10. A thermal-hydraulic drift-flux based mixture-fluid model for the description of single- and two-phase flow along a general coolant channel

    Energy Technology Data Exchange (ETDEWEB)

    Alois Hoeld [Bernaysstr. 16A, D-80937 Munich (Germany)

    2005-07-01

    Full text of publication follows: Different to the very simple class of homogeneous non-equilibrium models (HEM) an one dimensional thermal-hydraulic theoretical drift-flux based and thus non-homogeneous coolant channel model and, as a result, an in itself complete thermal-hydraulic coolant channel module CCM have been established allowing to simulate in a very general way the steady state and transient behaviour of the most important parameters of a single- or two-phase fluid flowing within any type of heated or non-heated coolant channel (with an eventually varying cross flow area). To avoid mathematical discontinuities at the transition from single- to two-phase flow the coolant channel will, in its general form, be split into different regions, i.e. be looked as a basic channel (BC) which can consist of a number of different flow regimes and can, accordingly, be subdivided into a number of sub-channels (SC-s). All of them belong, obviously, to only two types of SC-s, a SC with an only single-phase or two-phase flow regime separated by corresponding time-dependent phase boundaries. After a nodalization of the BC (and thus the corresponding SC-s) and applying a 'modified finite element method' for the spatial discretization of the partial differential eqs. (PDE-s) representing the conservation equations of thermal-hydraulics and after taking into account the initial and boundary conditions together with the additional constitutive equations a set of non-linear ordinary differential equations (ODE-s) of 1-st order can be derived for each SC type (and thus also the entire BC). Since during a transient a SC boundary can cross the BC node boundaries (so that a SC can eventually shrink to an only single node or even disappear or be created anew) special attention had to be given to the possibility of variable entrance or outlet positions (representing boiling boundaries or mixture levels). A special quadratic polygon approximation procedure (PAX) had to be

  11. The development of a thermal hydraulic feedback mechanism with a quasi-fixed point iteration scheme for control rod position modeling for the TRIGSIMS-TH application

    Science.gov (United States)

    Karriem, Veronica V.

    Nuclear reactor design incorporates the study and application of nuclear physics, nuclear thermal hydraulic and nuclear safety. Theoretical models and numerical methods implemented in computer programs are utilized to analyze and design nuclear reactors. The focus of this PhD study's is the development of an advanced high-fidelity multi-physics code system to perform reactor core analysis for design and safety evaluations of research TRIGA-type reactors. The fuel management and design code system TRIGSIMS was further developed to fulfill the function of a reactor design and analysis code system for the Pennsylvania State Breazeale Reactor (PSBR). TRIGSIMS, which is currently in use at the PSBR, is a fuel management tool, which incorporates the depletion code ORIGEN-S (part of SCALE system) and the Monte Carlo neutronics solver MCNP. The diffusion theory code ADMARC-H is used within TRIGSIMS to accelerate the MCNP calculations. It manages the data and fuel isotopic content and stores it for future burnup calculations. The contribution of this work is the development of an improved version of TRIGSIMS, named TRIGSIMS-TH. TRIGSIMS-TH incorporates a thermal hydraulic module based on the advanced sub-channel code COBRA-TF (CTF). CTF provides the temperature feedback needed in the multi-physics calculations as well as the thermal hydraulics modeling capability of the reactor core. The temperature feedback model is using the CTF-provided local moderator and fuel temperatures for the cross-section modeling for ADMARC-H and MCNP calculations. To perform efficient critical control rod calculations, a methodology for applying a control rod position was implemented in TRIGSIMS-TH, making this code system a modeling and design tool for future core loadings. The new TRIGSIMS-TH is a computer program that interlinks various other functional reactor analysis tools. It consists of the MCNP5, ADMARC-H, ORIGEN-S, and CTF. CTF was coupled with both MCNP and ADMARC-H to provide the

  12. United Arab Emirates limestones: impact of petrography on thermal behavior

    Science.gov (United States)

    Alaabed, Sulaiman; Soltan, Abdel Monem; Abdelghany, Osman; Amin, Bahaa Eldin Mahmoud; El Tokhi, Mohamed; Khaleel, Abbas; Musalim, Abdullah

    2014-12-01

    The thermal behavior of selected limestones from representative localities of the United Arab Emirates is investigated for their suitability for soft-burnt lime production. The limestone samples were collected from the Ghalilah, Musandam, Shauiba, Muthaymimah, Dammam and Asmari formations. The samples were characterized for petrography, mineral and chemical composition, together with physico-mechanical characteristics. Investigative methods included transmitted light microscopy (TLM), cathodoluminescence (CLM) and scanning electron microscopy (SEM), as well as X-ray micro-tomography (μ-CT), XRD, XRF and Archimedes method. The limestone samples were fired in an electrical muffle furnace for 0.25, 0.5, 1 and 2 hours at 800, 900, 1,000 and 1,100 °C. After firing the lime grains were tested to determine their hydration rate and microfabric. The Ghalilah and Musandam limes show the lowest and highest maximum hydration rates, respectively, due mainly to the impure nature of the former, and the smaller lime crystallites and dominance of post-calcination micro-cracks of the latter. The Dammam and Asmari limes preserve a "ghost" microfabric of the original limestone. Higher allochem contents impose lower activation energy requirements for calcination, which implies earlier calcination of the allochems. The Musandam, Shauiba and Muthaymimah limestones may be useful for the production of reactive soft-burnt lime under the applied firing conditions, however, the Dammam and Asmari limestones need more advanced calcination conditions than the applied ones. The Ghalilah limestone was found to be unsuitable for the production of lime.

  13. Anti-icing Behavior of Thermally Sprayed Polymer Coatings

    Science.gov (United States)

    Koivuluoto, Heli; Stenroos, Christian; Kylmälahti, Mikko; Apostol, Marian; Kiilakoski, Jarkko; Vuoristo, Petri

    2017-01-01

    Surface engineering shows an increasing potential to provide a sustainable approach to icing problems. Currently, several passive anti-ice properties adoptable to coatings are known, but further research is required to proceed for practical applications. This is due to the fact that icing reduces safety, operational tempo, productivity and reliability of logistics, industry and infrastructure. An icing wind tunnel and a centrifugal ice adhesion test equipment can be used to evaluate and develop anti-icing and icephobic coatings for a potential use in various arctic environments, e.g., in wind power generation, oil drilling, mining and logistic industries. The present study deals with evaluation of icing properties of flame-sprayed polyethylene (PE)-based polymer coatings. In the laboratory-scale icing tests, thermally sprayed polymer coatings showed low ice adhesion compared with metals such as aluminum and stainless steel. The ice adhesion strength of the flame-sprayed PE coating was found to have approximately seven times lower ice adhesion values compared with metallic aluminum, indicating a very promising anti-icing behavior.

  14. Comparative analysis of thermal behavior in hollow nuclear fuel pellets

    Energy Technology Data Exchange (ETDEWEB)

    Santos, Beatriz M. dos; Alvim, Antonio C.M., E-mail: bmachado@nuclear.ufrj.br, E-mail: aalvim@gmail.com [Coordenacao de Pos-Graduacao e Pesquisa de Engenharia (PEN/COPPE/UFRJ), Rio de Janeiro, RJ (Brazil). Programa de Engenharia Nuclear

    2017-11-01

    The increase in energy demand in Brazil and in the world is a real problem and several solutions are being considered to mitigate it. Maximization of energy generation, within the safety standards of fuel resources already known, is one of them. In this respect, nuclear energy is a crucial technology to sustain energy demand on several countries. Performances of a solid cylindrical and an annular rod have been verified and compared; where it has been proven that the annular rod can reach a higher nominal power in relation to the solid one. In this paper, the temperature profiles of two distinct nuclear fuel pellets, one of them annular and the other in the shape of a hollow biconcave disc (like the cross section of a red blood cell), were compared to analyze the efficiency and safety of both. The finite differences method allowed the evaluation of the thermal behavior of these pellets, where one specific physical condition was analyzed, regarding convection and conduction at the lateral edges. The results show that the temperature profile of the hollow biconcave disc pellet is lower, about 70 deg C below, when compared to the temperature profile of the annular pellet, considering the same simulation parameters for both pellets. (author)

  15. Study of Lactic Acid Thermal Behavior Using Thermoanalytical Techniques

    Directory of Open Access Journals (Sweden)

    Andrea Komesu

    2017-01-01

    Full Text Available Actually, there is a growing interest in the biotechnological production of lactic acid by fermentation aiming to substitute fossil fuel routes. The development of an efficient method for its separation and purification from fermentation broth is very important to assure the economic viability of production. Due to its high reactivity and tendency to decompose at high temperatures, the study of lactic acid thermal behavior is essential for its separation processes and potential application. In the present study, differential scanning calorimetry (DSC analyses showed endothermic peaks related to the process of evaporation. Data of thermogravimetry (TG/DTG were correlated to Arrhenius and Kissinger equations to provide the evaporation kinetic parameters and used to determine the vaporization enthalpy. Activation energies were 51.08 and 48.37 kJ·mol−1 and frequency values were 859.97 and 968.81 s−1 obtained by Arrhenius and Kissinger equations, respectively. Thermogravimetry, coupled with mass spectroscopy (TG-MS, provided useful information about decomposition products when lactic acid was heated at 573 K for approximately 30 min.

  16. Thermal-Hydraulic Integral Effect Test with ATLAS for an Intermediate Break Loss of Coolant Accident at a Pressurizer Surge Line

    Energy Technology Data Exchange (ETDEWEB)

    Kang, Kyoung Ho; Seok Cho; Park, Hyun Sik; Choi, Nam Hyun; Park, Yu Sun; Kim, Jong Rok; Bae, Byoung Uhn; Kim, Yeon Sik; Kim, Kyung Doo; Choi, Ki Yong; Song, Chul Hwa [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2013-10-15

    The main objectives of this test were not only to provide physical insight into the system response of the APR1400 during the pressurizer surge line break accident but also to produce an integral effect test data to validate the SPACE code. In order to simulate a double-ended guillotine break of a pressurizer surge line in the APR1400, the IB-SUR-01R test was performed with ATLAS. The major thermal-hydraulic phenomena such as the system pressures, the collapsed water levels, and the break flow rate were presented and discussed. Despite the core was uncovered, no excursion in the cladding temperature was observed. The pressurizer surge line break can be classified as a hot leg break from a break location point of view. Compared with a cold leg break, coolability in the core may be better in case of a hot leg break due to the enhanced flow in the core region. This integral effect test data will be used to evaluate the prediction capability of existing safety analysis codes of the MARS and the RELAP5 as well as the SPACE code. Furthermore, this data can be utilized to identify any code deficiency for an IBLOCA simulation, especially for DVI-adapted plants. Redefinition of break size for design basis accident (DBA) based on risk information is being extensively investigated due to the potential for safety benefits and unnecessary burden reduction from current LBLOCA (large break loss of coolant accident)-based ECC (Emergency Core Cooling) Acceptance Criteria. As a transition break size (TBS), the rupture of medium-size pipe is considered to be more important than ever in risk-informed regulation (RIR)-relevant safety analysis. As plants age, are up-rated, and continue to seek improved operating efficiencies, the small break and intermediate break LOCA (IBLOCA) can become a concern. In particular, IBLOCA with DVI (Direct Vessel Injection) features will be addressed to support redefinition of a design-basis LOCA. With an aim of expanding code validation to address small

  17. Basic hydraulics

    CERN Document Server

    Smith, P D

    1982-01-01

    BASIC Hydraulics aims to help students both to become proficient in the BASIC programming language by actually using the language in an important field of engineering and to use computing as a means of mastering the subject of hydraulics. The book begins with a summary of the technique of computing in BASIC together with comments and listing of the main commands and statements. Subsequent chapters introduce the fundamental concepts and appropriate governing equations. Topics covered include principles of fluid mechanics; flow in pipes, pipe networks and open channels; hydraulic machinery;

  18. Numerical simulation of the direct contact condensation phenomena for PTS-related in single and combined-effect thermal hydraulic test facilities using TransAT CMFD code

    Energy Technology Data Exchange (ETDEWEB)

    Kadi, Rabah, E-mail: kadi.rkhaled@hotmail.com [Laboratory for Thermal-Hydraulics, Nuclear Research Center of Birine (Algeria); Aissani, Slimane [Hydrocarbons and Chemistry Faculty, University of Boumerdes (Algeria); Bouam, Abdellah [Laboratory for Thermal-Hydraulics, Nuclear Research Center of Birine (Algeria)

    2015-11-15

    Highlights: • TransAT CMFD code application to DCC phenomenon. • LEIS methodology to predict the condensing steam flow rate. • Validation of interfacial phase-change heat transfer and turbulence models. • Correction of damping function at the free surface region. • Numerical validation of previous models using LIM and KAERI & KAIST test facilities. - Abstract: The use of CFD for the industrial studies related to PTS, including DCC is already possible; improvements of the two-phase modeling capabilities have to be undertaken to qualify the codes for the simulation of such flows. The DCC in horizontally stratified flow regime constitutes very considerable challenge exercises for a computational fluid dynamics (CFD) simulation of the thermal hydraulics PTS phenomenon because the interplay between turbulence and interfacial heat and mass transfer problem. The main purpose of our study is to investigate numerically the DCC in horizontally stratified steam water flow in a 2D and 3D channel using TransAT CMFD code. The new methodology known as Large-Eddy & Interface (LEIS) have been implemented for treatment of turbulence combined with interface tracking ITM (level set approach). Among of the so-called ‘coarse-grained’ ITM's models, the modified original surface divergence has been chosen as well as the treatment of the turbulence by URANS and VLES. This contribution addressed on the validation of interfacial phase-change heat transfer and turbulence models with special correction of the damping function at the free surface for single and combined-effect thermal hydraulic studies for LIM and KAERI & KAIST test facilities. The LIES methodology was found to apply successfully to predict the condensing steam flow rate in the all cases of the LIM test case involving a Smooth to Wavy turbulent, concurrent stratified steam-water flow in a 2D channel. The CMFD TransAT code predicting capability is analyzed, comparing the liquid temperature and to much the

  19. Pore Scale Thermal Hydraulics Investigations of Molten Salt Cooled Pebble Bed High Temperature Reactor with BCC and FCC Configurations

    Directory of Open Access Journals (Sweden)

    Shixiong Song

    2014-01-01

    CFD results and empirical correlations’ predictions of pressure drop and local Nusselt numbers. Local pebble surface temperature distributions in several default conditions are investigated. Thermal removal capacities of molten salt are confirmed in the case of nominal condition; the pebble surface temperature under the condition of local power distortion shows the tolerance of pebble in extreme neutron dose exposure. The numerical experiments of local pebble insufficient cooling indicate that in the molten salt cooled pebble bed reactor, the pebble surface temperature is not very sensitive to loss of partial coolant. The methods and results of this paper would be useful for optimum designs and safety analysis of molten salt cooled pebble bed reactors.

  20. Hydraulic Structures

    Data.gov (United States)

    Department of Homeland Security — This table is required whenever hydraulic structures are shown in the flood profile. It is also required if levees are shown on the FIRM, channels containing the...

  1. Hydraulic evaluation of the hypogenic karst area in Budapest (Hungary)

    Science.gov (United States)

    Erhardt, Ildikó; Ötvös, Viktória; Erőss, Anita; Czauner, Brigitta; Simon, Szilvia; Mádl-Szőnyi, Judit

    2017-09-01

    The Buda Thermal Karst area, in central Hungary, is in the focus of research interest because of its thermal water resources and the on-going hypogenic karstification processes at the boundary of unconfined and confined carbonates. Understanding of the discharge phenomena and the karstification processes requires clarification of the groundwater flow conditions in the area. Accordingly, the aim of the present study was to present a hydraulic evaluation of the flow systems based on analyses of the archival measured hydraulic data of wells. Pressure vs. elevation profiles, tomographic fluid-potential maps and hydraulic cross sections were constructed, based on the data distribution. As a result, gravitational flow systems, hydraulic continuity, and the modifying effects of aquitard units and faults were identified in the karst area. The location of natural discharge areas could be explained and the hydraulic behavior of the Northeastern Margin Fault of the Buda Hills could be determined. The flow pattern determines the differences in the discharge distribution (one- and two-component) and related cave-forming processes between the Central System (Rózsadomb area) and Southern System (Gellért Hill area) natural discharge areas. Among the premises of hypogenic karstification, regional upward flow conditions were confirmed along the main discharge zone of the River Danube.

  2. Development of thermal-hydraulic system analysis code SSC-K for pool-type liquid metal reactor

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Kyung Doo; Kwon, Y.M.; Suk, S.D.; Chang, W.P.; Hahn, D.H

    1999-03-01

    The Supper System Code of KAERI (SSC-K) is a best-estimate system code for analyzing an variety of off-normal or accident of a pool type design. It is developed at KAERI on the basis of SSC-L developed at BNL to analyze pool-type LMR transients. Because of inherent difference between th pool and loop design, the major modefications of SSC-L is required for the safety analysis of KALIMER. The major difference between KALIMER and general loop type LMRs exists in the primary heat transport system. In KALIMER, all of the essential components consisted of the primary heat transport system are located within the reactor vessel. This is contrast to the loop type LMRs, in which all the primary components are connected via piping to form loops attached externally to the reactor vessel. KALIMER has only one cover gas space. This eliminates the need for separate cover gas systems over liquid level in pump tanks and upper plenum. Since the sodium in hot pool is separated from cold pool by insulated barrier in KALIMER, The liquid level in hot pool is different from that in the cold pool mainly due to hydraulic losses and pump suction heads occuring during flow through the circulation pathes. In some accident conditions the liquid in the hot pool is flooded into cold pool and forms the natural circulation flow path. During the loss of heat sink transients, this will provided as a major heat rejection mechanism with the passive decay heat removal system. Since the pipes in the primary system exist only between pump discharge and core inlet plenum and are submerged in cold pool, a pipe rupture accident becomes less severe due to a constant back pressure exerted against the coolant flow from break. The intermediate and steam generator systems of both are generally identical. To adapt SSC-K to KALIMER design, the major modification of SSC-L has been made for the safety analysis of KALIMER. Test runs have been performed for the qualitative verification of the developed models. The

  3. Thermal Comfort for Urban Parks in Subtropics: Understanding Visitor’s Perceptions, Behavior and Attendance

    Directory of Open Access Journals (Sweden)

    Chuang-Hung Lin

    2013-01-01

    Full Text Available The paper is an effort toward thermal comfort assessment for urban parks under the climatic conditions of Taiwan to help architects achieve better climatic design. Field interviews, observations, and micrometeorological measurements were conducted in this study. The WBGT was used as the thermophysiological index to investigate the effects of thermal conditions on visitor’s thermal perception and adaptive behavior in outdoor urban spaces. In this study, behavioral adaptations used by visitors as a means of achieving comfort were evaluated. Observational results showed that the overall attendance was influenced by sun and thermal conditions. There was a robust relationship between thermal sensation votes, as well as thermal acceptability, and thermal environment, in terms of WBGT. The upper and lower limits of 80% acceptability are 26°C WBGT and 20°C WBGT, respectively.

  4. Analyses of Instability Events in the Peach Bottom-2 BWR Using Thermal-Hydraulic and 3D Neutron Kinetic Coupled Codes Technique

    Directory of Open Access Journals (Sweden)

    Antonella Lombardi Costa

    2008-01-01

    Full Text Available Boiling water reactor (BWR instabilities may occur when, starting from a stable operating condition, changes in system parameters bring the reactor towards an unstable region. In order to design more stable and safer core configurations, experimental and theoretical studies about BWR stability have been performed to characterise the phenomenon and to predict the conditions for its occurrence. In this work, contributions to the study of BWR instability phenomena are presented. The RELAP5/MOD3.3 thermal-hydraulic (TH system code and the PARCS-2.4 3D neutron kinetic (NK code were coupled to simulate BWR transients. Different algorithms were used to calculate the decay ratio (DR and the natural frequency (NF from the power oscillation predicted by the transient calculations as two typical parameters used to provide a quantitative description of instabilities. The validation of the code model set up for the Peach Bottom Unit 2 BWR plant is performed against low-flow stability tests (LFSTs. The four series of LFST have been performed during the first quarter of 1977 at the end of cycle 2 in Pennsylvania. The tests were intended to measure the reactor core stability margins at the limiting conditions used in design and safety analyses.

  5. The STAT7 Code for Statistical Propagation of Uncertainties In Steady-State Thermal Hydraulics Analysis of Plate-Fueled Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Dunn, Floyd E. [Argonne National Lab. (ANL), Argonne, IL (United States); Hu, Lin-wen [Massachusetts Inst. of Technology (MIT), Cambridge, MA (United States). Nuclear Reactor Lab.; Wilson, Erik [Argonne National Lab. (ANL), Argonne, IL (United States)

    2016-12-01

    The STAT code was written to automate many of the steady-state thermal hydraulic safety calculations for the MIT research reactor, both for conversion of the reactor from high enrichment uranium fuel to low enrichment uranium fuel and for future fuel re-loads after the conversion. A Monte-Carlo statistical propagation approach is used to treat uncertainties in important parameters in the analysis. These safety calculations are ultimately intended to protect against high fuel plate temperatures due to critical heat flux or departure from nucleate boiling or onset of flow instability; but additional margin is obtained by basing the limiting safety settings on avoiding onset of nucleate boiling. STAT7 can simultaneously analyze all of the axial nodes of all of the fuel plates and all of the coolant channels for one stripe of a fuel element. The stripes run the length of the fuel, from the bottom to the top. Power splits are calculated for each axial node of each plate to determine how much of the power goes out each face of the plate. By running STAT7 multiple times, full core analysis has been performed by analyzing the margin to ONB for each axial node of each stripe of each plate of each element in the core.

  6. GEYSER/TONUS: a coupled multi-D lumped parameter code for reactor thermal hydraulics analysis in case of severe accidents

    Energy Technology Data Exchange (ETDEWEB)

    Petit, M.; Durin, M.; Gauvain, J.

    1995-12-31

    The safety requirements for future light water reactors include accounting for severe accidents in the design process. The design must now include mitigation features to limit pressure and temperature inside the building. Hydrogen concentration is also a major issue for severe accidents. Modelling the thermal hydraulics inside the containment requires the description of complex phenomena such as condensation, stratification, transport of gases and aerosols, heat transfers. The effect of mitigation systems will increase the heterogeneities in the building, and most of those phenomena can be coupled. The GEYSER/TONUS multi-dimensional computer code is under development at CEA Saclay to model this complex situation. It allow the coupling of parts of the containment described in a lumped parameter manner, together with meshed parts. Emphasis is put on the numerical methods used to solve the transient problem, and physical models of classical lumped parameters codes will be adapted for the spatially described zones. The code is developed in the environment of the CASTEM 2000/TRIO EF system which allows to construct sophisticated applications based upon it. (J.S.). 22 refs., 1 fig.

  7. Application of ASTEC, MELCOR, and MAAP Computer Codes for Thermal Hydraulic Analysis of a PWR Containment Equipped with the PCFV and PAR Systems

    Directory of Open Access Journals (Sweden)

    Siniša Šadek

    2017-01-01

    Full Text Available The integrity of the containment will be challenged during a severe accident due to pressurization caused by the accumulation of steam and other gases and possible ignition of hydrogen and carbon monoxide. Installation of a passive filtered venting system and passive autocatalytic recombiners allows control of the pressure, radioactive releases, and concentration of flammable gases. Thermal hydraulic analysis of the containment equipped with dedicated passive safety systems after a hypothetical station blackout event is performed for a two-loop pressurized water reactor NPP with three integral severe accident codes: ASTEC, MELCOR, and MAAP. MELCOR and MAAP are two major US codes for severe accident analyses, and the ASTEC code is the European code, joint property of Institut de Radioprotection et de Sûreté Nucléaire (IRSN, France and Gesellschaft für Anlagen und Reaktorsicherheit (GRS, Germany. Codes’ overall characteristics, physics models, and the analysis results are compared herein. Despite considerable differences between the codes’ modelling features, the general trends of the NPP behaviour are found to be similar, although discrepancies related to simulation of the processes in the containment cavity are also observed and discussed in the paper.

  8. TRAC-PF1/MOD1: an advanced best-estimate computer program for pressurized water reactor thermal-hydraulic analysis

    Energy Technology Data Exchange (ETDEWEB)

    Liles, D.R.; Mahaffy, J.H.

    1986-07-01

    The Los Alamos National Laboratory is developing the Transient Reactor Analysis Code (TRAC) to provide advanced best-estimate predictions of postulated accidents in light-water reactors. The TRAC-PF1/MOD1 program provides this capability for pressurized water reactors and for many thermal-hydraulic test facilities. The code features either a one- or a three-dimensional treatment of the pressure vessel and its associated internals, a two-fluid nonequilibrium hydrodynamics model with a noncondensable gas field and solute tracking, flow-regime-dependent constitutive equation treatment, optional reflood tracking capability for bottom-flood and falling-film quench fronts, and consistent treatment of entire accident sequences including the generation of consistent initial conditions. The stability-enhancing two-step (SETS) numerical algorithm is used in the one-dimensional hydrodynamics and permits this portion of the fluid dynamics to violate the material Courant condition. This technique permits large time steps and, hence, reduced running time for slow transients.

  9. PMK-2, the First Integral Thermal-Hydraulics Tests for the Safety Evaluation of VVER-440/213 Nuclear Power Plants

    Directory of Open Access Journals (Sweden)

    Gy. Ézsöl

    2012-01-01

    Full Text Available The PMK-2 facility is a full-pressure thermal-hydraulic model of the primary and partly the secondary circuit of the VVER-type units of Paks NPP. The facility was the first integral-type facility for VVERs. The PMK-2 was followed later by the PACTEL (for VVER-440, the ISB, and PSB for VVER-1000. Since the startup of the facility in 1985, 55 experiments have been performed primarily in international frameworks with the participation of experts from 29 European and overseas countries forming a scientific school to better understand VVER system behaviour and reach a high level of modelling of accident sequences. The ATHLET, CATHARE, and RELAP5 codes have been validated including both qualitative and quantitative assessments. The former was almost exclusively applied to the early phase of validation by integral experiments, while the quantitative assessments have been performed by the Fast Fourier Transform Based Method. Paper gives comprehensive information on the design features of PMK-2 facility with a special respect to the representativeness of phenomena, the experiments performed, and the results of the validation of ATHLET, CATHARE, and RELAP5 codes. Safety significance of the PMK-2 projects is also discussed.

  10. Conversion of Molybdenum-99 production process to low enriched uranium: Neutronic and thermal hydraulic analyses of HEU and LEU target plates for irradiation in Pakistan Research Reactor-1

    Energy Technology Data Exchange (ETDEWEB)

    Mushtaq, Ahmad, E-mail: amushtaq1@hotmail.com [Isotope Production Division, Pakistan Institute of Nuclear Science and Technology, P.O. Nilore, Islamabad (Pakistan); Iqbal, Masood; Bokhari, Ishtiaq Hussain; Mahmood, Tayyab; Muhammad, Atta [Nuclear Engineering Division, Pakistan Institute of Nuclear Science and Technology, P.O. Nilore, Islamabad (Pakistan)

    2012-09-15

    Technetium-99m, the daughter product of Molybdenum-99 is the most widely needed radionuclide for diagnostic studies in Pakistan. Molybdenum-99 Production Facility has been established at PINSTECH. Highly enriched uranium (93% {sup 235}U) U/Al alloy targets have been irradiated in Pakistan Research Reactor-1 (PARR-1) for the generation of fission Mo-99, while basic dissolution technique is used for separation of Mo-99 from target matrix activity. In line with the international objective of minimizing and eventually eliminating the use of HEU in civil commerce, national and international efforts have been underway to shift the production of medical isotopes from HEU to LEU (LEU; <20% {sup 235}U enrichment) targets. To achieve the equivalent amount of {sup 99}Mo with LEU targets, approximately 5 times uranium is needed. LEU aluminum uranium dispersion target has been developed, which may replace existing HEU aluminum/uranium alloy targets for production of {sup 99}Mo using basic dissolution technique. Neutronic and thermal hydraulic calculations were performed for safe irradiation of targets in the core of PARR-1.

  11. Outcomes from the EURATOM–ROSATOM ERCOSAM SAMARA projects on containment thermal-hydraulics for severe accident management

    Energy Technology Data Exchange (ETDEWEB)

    Paladino, Domenico, E-mail: domenico.paladino@psi.ch [Paul Scherrer Institut (Switzerland); Andreani, Michele [Paul Scherrer Institut (Switzerland); Guentay, Salih [Innovative, Technology Development GmbH (Switzerland); Mignot, Guillaume; Kapulla, Ralf; Paranjape, Sidharth; Sharabi, Medhat [Paul Scherrer Institut (Switzerland); Kisselev, Arkadi; Yudina, Tatiana; Filippov, Aleksandr [Nuclear Safety Institute of the Russian Academy of Sciences, Moscow 115191 (Russian Federation); Kamnev, Mikhail; Khizbullin, Akhmir; Tyurikov, Oleg [JSC “Afrikantov OKB Mechanical Engineering”, Nizhny Novgorod 603074 (Russian Federation); Liang, Zhe [CNL-2251 Speakman Drive, Mississauga, ON L5K 1B2 (Canada); Abdo, Daniele; Brinster, Jérôme; Dabbene, Frédéric [CEA, DEN, DM2S, STMF, F-91191 Gif-sur-Yvette Cedex (France); Kelm, Stephan [Forschungszentrum Juelich, 52425 Jülich (Germany); Klauck, Michael; Götz, Lasse [RWTH Aachen University (Germany); and others

    2016-11-15

    Highlights: • Hydrogen distribution in the containment of PWR was investigated for scenario leading to stratification. • The scenario was scaled from a generic PWR containment to four facilities. • Effect of spray, cooler and heat sources was investigated experimentally and with LP and CFD. • Code-to-code benchmarks aiming a scaling up the facilities to a large containment. - Abstract: ERCOSAM and SAMARA are the acronyms for two parallel projects co-financed respectively by EURATOM and ROSATOM during the period 2010–2014 with the general aim to advance the knowledge on the phenomenology associated with the hydrogen and steam spreading and stratification in the LWR containment during a postulated severe accident. The important peculiarity of the projects was in experimental and analytical investigating the impact of systems such as spray, cooler and heat sources (simulating thermal effect of PARs) on the distribution of gas mixture (e.g. hydrogen, steam, air). This paper presents the main outcomes of the ERCOSAM–SAMARA projects.

  12. Development of a model for the thermal-hydraulic characterization of the He-FUS3 loop

    Energy Technology Data Exchange (ETDEWEB)

    Barone, G., E-mail: gianluca.barone@for.unipi.it [University of Pisa, Department of Civil and Industrial Engineering (DICI), Pisa (Italy); Coscarelli, E.; Forgione, N.; Martelli, D. [University of Pisa, Department of Civil and Industrial Engineering (DICI), Pisa (Italy); Del Nevo, A.; Tarantino, M.; Utili, M. [ENEA UTIS-TCI, CR Brasimone, Camugnano (Italy); Ricapito, I.; Calderoni, P. [TBM& MD Project, Fusion for Energy, EU Commission, Carrer J. Pla, 2, Building B3, 08019 Barcelona (Spain)

    2015-10-15

    Highlights: • RELAP5-3D model of He-FUS3 facility and turbo circulator (TC). • Cold and hot facility T/H numerical analysis for the TC operanting range. • Effect of the cold by-pass opening on the facility performances. - Abstract: He-FUS3 is a helium facility designed and realized by ENEA in order to test the thermal-mechanical properties of prototypical breeding blanket module assemblies of a DEMO reactor. The actual facility has been upgraded with a high performance turbo circulator and a water heat exchanger integrating the pre-existent air cooler. In addition, a new test section located in the loop hot zone has been settled down with the objective of investigating safety relevant transient conditions of “In-TBM” LOCA scenarios. A RELAP5-3D{sup ©} model has been developed to perform a set of preliminary simulations on the new He-FUS3 layout. Both cold and hot stationary conditions have been analyzed evaluating the turbo circulator performances for a wide range of helium flow rate. Outcomes have shown that RELAP5-3D{sup ©} is an effective tool in reproducing the most significant phenomena of He-FUS3 system, providing relevant insight supporting future experimental campaigns. The post-test analysis phase will be, of course, fundamental for the qualification of a consistent numerical model.

  13. Study of the stagnation in a thermal solar installation. Influence of the design of the hydraulic circuit; Estudio del estancamiento en una instalacion solar termica. Influencia del diseno del circuito hidraulico

    Energy Technology Data Exchange (ETDEWEB)

    Vicente, P. G.

    2008-07-01

    An experimental study of stagnation in a solar thermal facility has been carried-out. The experimental facility consists of 4 flat solar collectors of 2 m{sup 2} each one and a tank of 450 litres. It has been determined the evolution over time in temperature in 20 points in the hydraulic circuit. These data, together with pressure and flow measurements, permit to show the actual behaviour of the solar system at stagnation conditions. (Author)

  14. Computer simulation of thermal-hydraulic transient events in multi-circuits with multipumps; Simulacao computacional de eventos termo-hidraulicos transitorios em multicircuitos com multibombas

    Energy Technology Data Exchange (ETDEWEB)

    Veloso, Marcelo Antonio

    2003-07-01

    PANTERA-2 (from Programa para Analise Termo-hidraulica de Reatores a Agua - Program for Thermal-hydraulic Analysis of Water Reactors, Version 2), whose fundamentals are described in this work, is intended to carry out rod bundle subchannel analysis in conjunction with multiloop simulation. It solves simultaneously the conservation equations of mass, axial and lateral momentum, and energy for subchannel geometry coupled with the balance equations that describe the fluid flows in any number of coolant loops connected to a pressure vessel containing the rod bundle. As far as subchannel analysis is concerned, the basic computational strategy of PANTERA-2 comes from COBRA codes, but an alternative implicit solution method oriented to the pressure field has been used to solve the finite difference approximations for the balance laws. The results provided by the subchannel model comprise the fluid density, enthalpy, flow rate, and pressure fields in the subchannels. The loop model predicts the individual loop flows, total flow through the pressure vessel, and pump rotational speeds as a function of time subsequent to the failure of any number of the coolant pumps. The flow transients in the loops may initiated by partial, total or sequential loss of electric power to the operating pumps. Transient events caused by either shaft break or rotor locking may also be simulated. The changes in rotational speed of the pumps as a function of rime are determined from a torque balance. Pump dynamic head and hydraulic torque are calculated as a function of rotational speed and volumetric flow from two polar homologous curves supplied to the code in the tabular form. In order to illustrate the analytical capability of PANTERA-2, three sample problems are presented and discussed. Comparisons between calculated and measured results indicate that the program reproduces with a good accuracy experimental data for subchannel exit temperatures and critical heat fluxes in 5x5 rod bundles. It

  15. Oxidation and thermal shock behavior of thermal barrier coated 18/10CrNi alloy with coating modifications

    Energy Technology Data Exchange (ETDEWEB)

    Guergen, Selim [Vocational School of Transportation, Anadolu University, Eskisehir (Turkmenistan); Diltemiz, Seyid Fehmi [Turkish Air Force1st Air Supply and Maintenance Center Command, Eskisehir (Turkmenistan); Kushan, Melih Cemal [Dept. of Mechanical Engineering, Eskisehir Osmangazi University, Eskisehir (Turkmenistan)

    2017-01-15

    In this study, substrates of 18/10CrNi alloy plates were initially sprayed with a Ni-21Cr-10Al-1Y bond coat and then with an yttria stabilized zirconia top coat by plasma spraying. Subsequently, plasma-sprayed Thermal barrier coatings (TBCs) were treated with two different modification methods, namely, vacuum heat treatment and laser glazing. The effects of modifications on the oxidation and thermal shock behavior of the coatings were evaluated. The effect of coat thickness on the bond strength of the coats was also investigated. Results showed enhancement of the oxidation resistance and thermal shock resistance of TBCs following modifications. Although vacuum heat treatment and laser glazing exhibited comparable results as per oxidation resistance, the former generated the best improvement in the thermal shock resistance of the TBCs. Bond strength also decreased as coat thickness increased.

  16. THERMAL-HYDRAULIC ANALYSIS OF SMR WITH NATURALLY CIRCULATING PRIMARY SYSTEM DURING LOSS OF FEED WATER ACCIDENT

    Directory of Open Access Journals (Sweden)

    Susyadi Susyadi

    2016-09-01

    ABSTRAK Reaktor daya kecil modular (SMR memiliki beberapa keunggulan dibanding reaktor daya besar konvensional. Dengan disain yang lebih sederhana dan terintegrasi, penerapan hukum alamiah untuk sistem keselamatannya dan biaya modal yang rendah, reaktor ini sangat cocok untuk dibangun di Indonesia. Salah satunya disain SMR yang sedang dikembangkan menerapkan gaya penggerak alami untuk sistim pendingin primernya. Dengan disain seperti itu, adalah sangat penting untuk memahami implikasinya terhadap aspek keselamatan pada seluruh kondisi operasi. Salah satu yang perlu diinvestigasi adalah kecelakaan kehilangan air umpan (LoFW. Pada studi ini, dilakukan analisis kinerja thermal hidrolik SMR yang menggunakan sistim pendinginan primer sirkulasi alam saat kecelakaan LoFW. Tujuannya adalah untuk menginvestigasi karakteristik aliran sistem primer saat kecelakaan LoFW dan untuk memastikan apakah aliran sirkulasi alam cukup untuk memindahkan panas dari teras guna menjaga kondisi tetap aman selama kecelakaan tersebut. Metoda yang digunakan adalah dengan merepresentasikan sistem reaktor ke dalam model-model generik program RELAP5 dan melakukan simulasi numerik. Hasil perhitungan menunjukkan bahwa setelah kejadian pemicu dan trip reaktor, pada sisi primer laju alirnya berfluktuasi secara signifikan dan temperatur pendinginnya menurun secara bertahap sedangkan  pada sisi sekunder kondisi uap berubah menjadi uap jenuh. Laju alir turun dari ~711 kg/detik menjadi ~263 kg/detik sebelum kembali naik lagi pada t=~46 detik. Saat laju alir di titik terendah, temperatur pusat bahan bakar dan fluida pendingin adalah sekitar  ~565 K dan  ~554 K, yang menujukkan bahwa temperatur bahan bakar masih jauh di bawah batas disain dan temperatur fluidanya juga berada di bawah titik saturasi. Keadaan ini menunjukkan bahwa saat transien kedua parameter utama termohidrolik reaktor tetap dalam kondisi yang dapat diterima sehingga dapat disimpulkan  bahwa saat  kecelakaan kehilangan air umpan, SMR

  17. Thermal Expansion Behavior of Hot-Pressed Engineered Matrices

    Science.gov (United States)

    Raj, S. V.

    2016-01-01

    Advanced engineered matrix composites (EMCs) require that the coefficient of thermal expansion (CTE) of the engineered matrix (EM) matches those of the fiber reinforcements as closely as possible in order to reduce thermal compatibility strains during heating and cooling of the composites. The present paper proposes a general concept for designing suitable matrices for long fiber reinforced composites using a rule of mixtures (ROM) approach to minimize the global differences in the thermal expansion mismatches between the fibers and the engineered matrix. Proof-of-concept studies were conducted to demonstrate the validity of the concept.

  18. Cellulose nanowhiskers from coconut husk fibers: effect of preparation conditions on their thermal and morphological behavior

    Science.gov (United States)

    Cellulose nanowhiskers were prepared by sulfuric acid hydrolysis from coconut husk fibers which had previously been submitted to a delignification process. The effects of preparation conditions on the thermal and morphological behavior of the nanocrystals were investigated. Cellulose nanowhisker sus...

  19. Cracking behavior of tungsten armor under ELM-like thermal shock loads: A computational study

    National Research Council Canada - National Science Library

    Li, Muyuan; Werner, Ewald; You, Jeong-Ha

    2015-01-01

    In this work, the cracking behavior of tungsten under edge-localized mode (ELM)-like thermal shock loads was investigated on the basis of a rigorous computational fracture mechanical analysis combined with the finite element method...

  20. Thermal hydraulic parametric investigation of decay heat removal from degraded core of a sodium cooled fast Breeder reactor

    Energy Technology Data Exchange (ETDEWEB)

    Verma, Lokesh [Department of Physics and Astrophysics, University of Delhi, Delhi 110007 (India); Kumar Sharma, Anil, E-mail: aksharma@igcar.gov.in [Reactor Design Group, Indira Gandhi Centre for Atomic Research, HBNI, Kalpakkam (India); Velusamy, K. [Reactor Design Group, Indira Gandhi Centre for Atomic Research, HBNI, Kalpakkam (India)

    2017-03-15

    established that a single plate core catcher can safely accommodate decay heat arising due to ∼70% of the core debris by establishing natural circulation in the lower sodium pool. The influence of heat removal rate by natural circulation on availability of the number of decay heat exchangers (DHX) dipped in the upper pool is also analyzed. It is seen that the temperatures in the inner vessel, source plate and the maximum debris temperature do not increase significantly even when the DHXs are deployed 5 h after the accident, demonstrating the benefit of large thermal inertia of the pool.

  1. Thermographic measurement of thermal bridges in buildings under dynamic behavior

    Science.gov (United States)

    Ferrarini, G.; Bison, P.; Bortolin, A.; Cadelano, G.; De Carli, M.

    2016-05-01

    The accurate knowledge of the thermal performance could reduce significantly the impact of buildings on global energy consumption. Infrared thermography is widely recognized as one of the key technologies for building surveys, thanks to its ability to acquire at a glance thermal images of the building envelope. However, a spot measurement could be misleading when the building is under dynamic thermal conditions. In this case data should be acquired for hours or days, depending on the thermal properties of the walls. Long term thermographic monitoring are possible but imply strong challenges from a practical standpoint. This work investigates the possibilities and limitations of spot thermographic surveys coupled with contact probes, that are able to acquire continuously the thermal signal for days, to investigate the thermal bridges of a building. The goal is the estimation of the reliability and accuracy of the measurement under realistic environmental conditions. Firstly, numerical simulations are performed to determine the reference value of an experimental case. Then a long term thermographic survey is performed and integrated with the contact probe measurement, assessing the feasibility of the method.

  2. Correlation between dynamic wetting behavior and chemical components of thermally modified wood

    Energy Technology Data Exchange (ETDEWEB)

    Wang, Wang; Zhu, Yuan; Cao, Jinzhen, E-mail: caoj@bjfu.edu.cn; Sun, Wenjing

    2015-01-01

    Highlights: • We studied the dynamic wetting behavior of thermally modified wood by wetting models. • We found lower wetting speed of water droplets on thermally modified wood surface. • Dynamic wetting behavior and surface chemical components show a strong correlation. - Abstract: In order to investigate the dynamic wetting behavior of thermally modified wood, Cathay poplar (Populus cathayana Rehd.) and Scots pine (Pinus sylvestris L.) samples were thermally modified in an oven at 160, 180, 200, 220 or 240 °C for 4 h in this study. The dynamic contact angles and droplet volumes of water droplets on modified and unmodified wood surfaces were measured by sessile drop method, and their changing rates (expression index: K value and wetting slope) calculated by wetting models were illustrated for mapping the dynamic wetting process. The surface chemical components were also measured by X-ray photoelectron spectroscopy analysis (XPS), thus the relationship between dynamic wetting behavior and chemical components of thermally modified wood were determined. The results indicated that thermal modification was capable of decreasing the dynamic wettability of wood, expressed in lowing spread and penetration speed of water droplets on wood surfaces. This change was more obvious with the increased heating temperature. The K values varied linearly with the chemical components parameter (mass loss, O/C ratio, and C{sub 1}/C{sub 2} ratio), indicating a strong correlation between dynamic wetting behavior and chemical components of thermally modified wood.

  3. Experimental Results for Direct Electron Irradiation of a Uranyl Sulfate Solution: Bubble Formation and Thermal Hydraulics Studies

    Energy Technology Data Exchange (ETDEWEB)

    Chemerisov, Sergey [Argonne National Lab. (ANL), Argonne, IL (United States); Gromov, R. [Argonne National Lab. (ANL), Argonne, IL (United States); Makarashvili, Vakhtang [Argonne National Lab. (ANL), Argonne, IL (United States); Heltemes, Thad [Argonne National Lab. (ANL), Argonne, IL (United States); Sun, Zaijing [Argonne National Lab. (ANL), Argonne, IL (United States); Wardle, Kent E. [Argonne National Lab. (ANL), Argonne, IL (United States); Bailey, James [Argonne National Lab. (ANL), Argonne, IL (United States); Stepinski, Dominique [Argonne National Lab. (ANL), Argonne, IL (United States); Jerden, James [Argonne National Lab. (ANL), Argonne, IL (United States); Vandegrift, George F. [Argonne National Lab. (ANL), Argonne, IL (United States)

    2015-01-30

    In support of the development of accelerator-driven production of fission product Mo-99 as proposed by SHINE Medical Technologies, a 35 MeV electron linac was used to irradiate depleted-uranium (DU) uranyl sulfate dissolved in pH 1 sulfuric acid at average power densities of 6 kW, 12 kW, and 15 kW. During these irradiations, gas bubbles were generated in the solution due to the radiolytic decomposition of water molecules in the solution. Multiple video cameras were used to record the behavior of bubble generation and transport in the solution. Seven six-channel thermocouples were used to record temperature gradients in the solution from self-heating. Measurements of hydrogen and oxygen concentrations in a helium sweep gas were recorded by a gas chromatograph to estimate production rates during irradiation. These data are being used to validate a computational fluid dynamics (CFD) model of the experiment that includes multiphase flow and a custom bubble injection model for the solution region.

  4. Thermal-Mechanical and Thermal Behavior of High-Temperature Structural Materials.

    Science.gov (United States)

    1979-12-31

    Physical Constants of Porcelain ," Nagoya Kogyo Gijutsu Shikensko Hokoku, 8 [5] 37-43 (1959); Ceram. Abstracts, 1959, Nov. p. 287a. 6. F. P. Knudsen...engineering materials appropriate for conditions which require high thermal shock resistance in combination with good thermal insulating ability"C. Finally

  5. Thermal shock resistance behavior of a functionally graded ceramic: Effects of finite cooling rate

    Directory of Open Access Journals (Sweden)

    Zhihe Jin

    2014-01-01

    Full Text Available This work presents a semi-analytical model to explore the effects of cooling rate on the thermal shock resistance behavior of a functionally graded ceramic (FGC plate with a periodic array of edge cracks. The FGC is assumed to be a thermally heterogeneous material with constant elastic modulus and Poisson's ratio. The cooling rate applied at the FGC surface is modeled using a linear ramp function. An integral equation method and a closed form asymptotic temperature solution are employed to compute the thermal stress intensity factor (TSIF. The thermal shock residual strength and critical thermal shock of the FGC plate are obtained using the SIF criterion. Thermal shock simulations for an Al2O3/Si3N4 FGC indicate that a finite cooling rate leads to a significantly higher critical thermal shock than that under the sudden cooling condition. The residual strength, however, is relatively insensitive to the cooling rate.

  6. Dynamic Thermal Features of Insulated Blocks: Actual Behavior and Myths

    Directory of Open Access Journals (Sweden)

    Marta Cianfrini

    2017-11-01

    Full Text Available The latest updates in the European directive on energy performance of buildings have introduced the fundamental “nearly zero-energy building (NZEB” concept. Thus, a special focus needs to be addressed to the thermal performance of building envelopes, especially concerning the role played by thermal inertia in the energy requirements for cooling applications. In fact, a high thermal inertia of the outer walls results in a mitigation of the daily heat wave, which reduces the cooling peak load and the related energy demand. The common assumption that high mass means high thermal inertia typically leads to the use of high-mass blocks. Numerical and experimental studies on thermal inertia of hollow envelope components have not confirmed this general assumption, even though no systematic analysis is readily available in the open literature. Yet, the usually employed methods for the calculation of unsteady heat transfer through walls are based on the hypothesis that such walls are composed of homogeneous layers. In this framework, a study of the dynamic thermal performance of insulated blocks is brought forth in the present paper. A finite-volume method is used to solve the two-dimensional equation of conduction heat transfer, using a triangular-pulse temperature excitation to analyze the heat flux response. The effects of both the type of clay and the insulating filler are investigated and discussed at length. The results obtained show that the wall front mass is not the basic independent variable, since clay and insulating filler thermal diffusivities are more important controlling parameters.

  7. Modelling of Thermal Behavior of Borehole Heat Exchangers of Geothermal Heat Pump Heating Systems

    Directory of Open Access Journals (Sweden)

    Gornov V.F.

    2016-01-01

    Full Text Available This article reports results of comparing the accuracy of the software package “INSOLAR.GSHP.12”, modeling non-steady thermal behavior of geothermal heat pump heating systems (GHCS and of the similar model “conventional” using finite difference methods for solving spatial non-steady problems of heat conductivity. The software package is based on the method of formulating mathematical models of thermal behavior of ground low-grade heat collection systems developed by INSOLAR group of companies. Equations of mathematical model of spatial non-steady thermal behavior of ground mass of low-grade heat collection system obtained by the developed method have been solved analytically that significantly reduced computing time spent by the software complex “INSOLAR.GSHP.12” for calculations. The method allows to turn aside difficulties associated with information uncertainty of mathematical models of the ground thermal behavior and approximation of external factors affecting the ground. Use of experimentally obtained information about the ground natural thermal behavior in the software package allows to partially take into account the whole complex of factors (such as availability of groundwater, their velocity and thermal behavior, structure and arrangement of ground layers, the Earth’s thermal background, precipitation, phase transformations of moisture in the pore space, and more, significantly influencing the formation of thermal behavior of the ground mass of a low-grade geothermal heat collection system. Numerical experiments presented in the article confirmed the high convergence of the results obtained through the software package “INSOLAR.GSHP.12” with solutions obtained by conventional finite-difference methods.

  8. Thermal reactor safety

    Energy Technology Data Exchange (ETDEWEB)

    1980-06-01

    Information is presented concerning new trends in licensing; seismic considerations and system structural behavior; TMI-2 risk assessment and thermal hydraulics; statistical assessment of potential accidents and verification of computational methods; issues with respect to improved safety; human factors in nuclear power plant operation; diagnostics and activities in support of recovery; LOCA transient analysis; unresolved safety issues and other safety considerations; and fission product transport.

  9. Field Synergy Analysis and Optimization of the Thermal Behavior of Lithium Ion Battery Packs

    Directory of Open Access Journals (Sweden)

    Hongwen He

    2017-01-01

    Full Text Available In this study, a three dimensional (3D modeling has been built for a lithium ion battery pack using the field synergy principle to obtain a better thermal distribution. In the model, the thermal behavior of the battery pack was studied by reducing the maximum temperature, improving the temperature uniformity and considering the difference between the maximum and maximum temperature of the battery pack. The method is further verified by simulation results based on different environmental temperatures and discharge rates. The thermal behavior model demonstrates that the design and cooling policy of the battery pack is crucial for optimizing the air-outlet patterns of electric vehicle power cabins.

  10. Thermal behavior variations in coating thickness using pulse phase thermography

    Energy Technology Data Exchange (ETDEWEB)

    Ranjit, Shrestha; Chung, Yoonjae; Kim, Won Tae [Kongju National University, Cheonan (Korea, Republic of)

    2016-08-15

    This paper presents a study on the use of pulsed phase thermography in the measurement of thermal barrier coating thickness with a numerical simulation. A multilayer heat transfer model was used to analyze the surface temperature response acquired from one-sided pulsed thermal imaging. The test sample comprised four layers: the metal substrate, bond coat, thermally grown oxide and the top coat. The finite element software, ANSYS, was used to model and predict the temperature distribution in the test sample under an imposed heat flux on the exterior of the TBC. The phase image was computed with the use of the software MATLAB and Thermofit Pro using a Fourier transform. The relationship between the coating thickness and the corresponding phase angle was then established with the coating thickness being expressed as a function of the phase angle. The method is successfully applied to measure the coating thickness that varied from 0.25 mm to 1.5 mm.

  11. Thermal behavior of Liquidambar orientalis mill wood before and after extraction processes

    Directory of Open Access Journals (Sweden)

    Evren Terzi

    2017-11-01

    Full Text Available The effect of extractives on the thermal behavior of Liquidambar orientalis Mill. (storax wood is studied using thermogravimetric analysis (TGA. To evaluate the effects of polar and apolar extractives on the thermal behavior of wood, sawdust samples from the heartwood of L. orientalis are extracted with either cold water (48 h, hot water (48 h, or ethanol/toluene (1:2 v/v (6 h prior to thermal analysis. Thermogravimetry (TG curves show that polar and apolar extractives promote char formation, increase the amount of residue, and improve the thermal behavior of L. orientalis wood. In addition, derivative thermogravimetry (DTG curves demonstrate that thermal degradation of un-extracted and cold water-extracted wood samples occurs in a single step, while a two-step degradation pattern is seen for hot water- and ethanol/toluene-extracted wood samples. It is also observed that first degradation reactions in hot water and ethanol/toluene-extracted wood samples occur faster than those in unextracted and cold water-extracted wood samples. Although there are approximately half the number of extracted apolar compounds compared to polar compounds, the removal of both types of compounds affect the thermal properties of L. orientalis wood to the same degree. It is thus deduced that apolar extractives significantly affect the thermal behavior of L. orientalis wood.

  12. The effect of Y2O3 addition on thermal shock behavior of magnesium aluminate spinel

    Directory of Open Access Journals (Sweden)

    Pošarac Milica

    2009-01-01

    Full Text Available The effect of yttria additive on the thermal shock behavior of magnesium aluminate spinel has been investigated. As a starting material we used spinel (MgAl2O4 obtained by the modified glycine nitrate procedure (MGNP. Sintered products were characterized in terms of phase analysis, densities, thermal shock, monitoring the damaged surface area in the refractory specimen during thermal shock and ultrasonic determination of the Dynamic Young modulus of elasticity. It was found that a new phase between yttria and alumina is formed, which improved thermal shock properties of the spinel refractories. Also densification of samples is enhanced by yttria addition.

  13. Thermal stability, swelling behavior and CO 2 absorption properties of Nanoscale Ionic Materials (NIMs)

    KAUST Repository

    Andrew Lin, Kun-Yi

    2014-11-11

    © The Royal Society of Chemistry 2015. Nanoscale Ionic Materials (NIMs) consist of a nanoscale core, a corona of charged brushes tethered on the surface of the core, and a canopy of the oppositely charged species linked to the corona. Unlike conventional polymeric nanocomposites, NIMs can display liquid-like behavior in the absence of solvents, have a negligible vapor pressure and exhibit unique solvation properties. These features enable NIMs to be a promising CO2 capture material. To optimize NIMs for CO2 capture, their structure-property relationships were examined by investigating the roles of the canopy and the core in their thermal stability, and thermally- and CO2-induced swelling behaviors. NIMs with different canopy sizes and core fractions were synthesized and their thermal stability as well as thermally- and CO2-induced swelling behaviors were determined using thermogravimetry, and ATR FT-IR and Raman spectroscopies. It was found that the ionic bonds between the canopy and the corona, as well as covalent bonds between the corona and the core significantly improved the thermal stability compared to pure polymer and polymer/nanofiller mixtures. A smaller canopy size and a larger core fraction led to a greater enhancement in thermal stability. This thermal stability enhancement was responsible for the long-term thermal stability of NIMs over 100 temperature swing cycles. Owing to their ordered structure, NIMs swelled less when heated or when they adsorbed CO2 compared to their corresponding polymers. This journal is

  14. Effect of Thermal Cycling on the Tensile Behavior of CF/AL Fiber Metal Laminates

    Directory of Open Access Journals (Sweden)

    Muhammad Farhan Noor

    2017-09-01

    Full Text Available The objective of this research work was to estimate the effect of thermal cycling on the tensile behavior of CARALL composites. Fiber metal laminates (FMLs, based on 2D woven carbon fabric and 2024-T3 Alclad aluminum alloy sheet, was manufactured by pressure molding technique followed by hand layup method. Before fabrication, aluminum sheets were anodized with phosphoric acid to produce micro porous alumina layer on surface. This micro-porous layer is beneficial to produce strong bonding between metal and fiber surfaces in FMLs. The effect of thermal cycling (-65 to +70ºC on the tensile behavior of Cf/Al based FML was studied. Tensile strength was increased after 10 thermal cycles, but it was slightly decreased to some extent after 30, and 50 thermal cycles. Tensile modulus also shown the similar behavior as that of tensile strength.

  15. Mapping Thermal Habitat of Ectotherms Based on Behavioral Thermoregulation in a Controlled Thermal Environment

    Science.gov (United States)

    Fei, T.; Skidmore, A.; Liu, Y.

    2012-07-01

    Thermal environment is especially important to ectotherm because a lot of physiological functions rely on the body temperature such as thermoregulation. The so-called behavioural thermoregulation function made use of the heterogeneity of the thermal properties within an individual's habitat to sustain the animal's physiological processes. This function links the spatial utilization and distribution of individual ectotherm with the thermal properties of habitat (thermal habitat). In this study we modelled the relationship between the two by a spatial explicit model that simulates the movements of a lizard in a controlled environment. The model incorporates a lizard's transient body temperatures with a cellular automaton algorithm as a way to link the physiology knowledge of the animal with the spatial utilization of its microhabitat. On a larger spatial scale, 'thermal roughness' of the habitat was defined and used to predict the habitat occupancy of the target species. The results showed the habitat occupancy can be modelled by the cellular automaton based algorithm at a smaller scale, and can be modelled by the thermal roughness index at a larger scale.

  16. Thermal behavior of the Medicina 32-meter radio telescope

    Science.gov (United States)

    Pisanu, Tonino; Buffa, Franco; Morsiani, Marco; Pernechele, Claudio; Poppi, Sergio

    2010-07-01

    We studied the thermal effects on the 32 m diameter radio-telescope managed by the Institute of Radio Astronomy (IRA), Medicina, Bologna, Italy. The preliminary results show that thermal gradients deteriorate the pointing performance of the antenna. Data has been collected by using: a) two inclinometers mounted near the elevation bearing and on the central part of the alidade structure; b) a non contact laser alignment optical system capable of measuring the secondary mirror position; c) twenty thermal sensors mounted on the alidade trusses. Two series of measurements were made, the first series was performed by placing the antenna in stow position, the second series was performed while tracking a circumpolar astronomical source. When the antenna was in stow position we observed a strong correlation between the inclinometer measurements and the differential temperature. The latter was measured with the sensors located on the South and North sides of the alidade, thus indicating that the inclinometers track well the thermal deformation of the alidade. When the antenna pointed at the source we measured: pointing errors, the inclination of the alidade, the temperature of the alidade components and the subreflector position. The pointing errors measured on-source were 15-20 arcsec greater than those measured with the inclinometer.

  17. Dual percolation behaviors of electrical and thermal conductivity in metal-ceramic composites

    Science.gov (United States)

    Sun, K.; Zhang, Z. D.; Qian, L.; Dang, F.; Zhang, X. H.; Fan, R. H.

    2016-02-01

    The thermal and electrical properties including the permittivity spectra in radio frequency region were investigated for copper/yttrium iron garnet (Cu/YIG) composites. Interestingly, the percolation behaviors in electrical and thermal conductivity were obtained due to the formation of copper particles' networks. Beyond the electrical percolation threshold, negative permittivity was observed and plasmon frequency was reduced by several orders of magnitude. With the increase in copper content, the thermal conductivity was gradually increased; meanwhile, the phonon scattering effect and thermal resistance get enhanced, so the rate of increase in thermal conductivity gradually slows down. Hopefully, Cu/YIG composites with tunable electrical and thermal properties have great potentials for electromagnetic interference shielding and electromagnetic wave attenuation.

  18. Buckling Behavior of Cold-Formed Studs with Thermal Perforations

    Directory of Open Access Journals (Sweden)

    Garifullin Marsel

    2016-01-01

    Full Text Available Studies have shown that the optimal structural scheme for low-rise buildings that meets all regulatory requirements is a frame system. In this connection, thin-walled cold-formed steel (CFS profiles seem to be the best material for constructing light steel framed (LSF walls. The framework of LSF walls is usually constructed from CFS C-shaped profiles. To increase the thermal effectiveness of a wall, CFS profiles usually have thermal perforations and thus are called thermoprofiles. However, these openings have a negative impact on bearing capacity of profiles and require accurate evaluation. In this article a relatively new reticular-stretched thermoprofile with diamond-shaped openings is considered. The article deals with the buckling analysis of perforated CFS C-sections subjected to compression.

  19. A-thermal elastic behavior of silicate glasses

    Science.gov (United States)

    Rabia, Mohammed Kamel; Degioanni, Simon; Martinet, Christine; Le Brusq, Jacques; Champagnon, Bernard; Vouagner, Dominique

    2016-02-01

    Depending on the composition of silicate glasses, their elastic moduli can increase or decrease as function of the temperature. Studying the Brillouin frequency shift of these glasses versus temperature allows the a-thermal composition corresponding to an intermediate glass to be determined. In an intermediate glass, the elastic moduli are independent of the temperature over a large temperature range. For sodium alumino-silicate glasses, the a-thermal composition is close to the albite glass (NaAlSi3O8). The structural origin of this property is studied by in situ high temperature Raman scattering. The structure of the intermediate albite glass and of silica are compared at different temperatures between room temperature and 600 °C. When the temperature increases, it is shown that the high frequency shift of the main band at 440 cm-1 in silica is a consequence of the cristobalite-like alpha-beta transformation of 6-membered rings. This effect is stronger in silica than bond elongation (anharmonic effects). As a consequence, the elastic moduli of silica increase as the temperature increases. In the albite glass, the substitution of 25% of Si4+ ions by Al3+ and Na+ ions decreases the proportion of SiO2 6-membered rings responsible for the silica anomaly. The effects of the silica anomaly balance the anharmonicity in albite glass and give rise to an intermediate a-thermal glass. Different networks, formers or modifiers, can be added to produce different a-thermal glasses with useful mechanical or chemical properties.

  20. An integrated systems calculation of a steam generator tube rupture in a modular prismatic HTGR (high-temperature gas-cooled reactor) conceptual design using ATHENA (Advanced Thermal-Hydraulic Energy Network Analyzer)

    Energy Technology Data Exchange (ETDEWEB)

    Beelman, R.J. (Idaho National Engineering Laboratory, Idaho Falls (USA))

    1989-11-01

    The capability to perform integrated systems calculations of modular high-temperature gas-cooled reactor (MHTGR) transients has been developed at the Idaho National Engineering Laboratory (INEL) using the Advanced Thermal-Hydraulic Energy Network Analyzer (ATHENA) computer code. A scoping calculation of a steam generator tube rupture (SGTR) water ingress event in a prismatic 2 {times} 350-MW(thermal) MHTGR conceptual design has been completed at INEL using ATHENA. The proposed MHTGR design incorporates dual, graphite-moderated, helium-cooled, 350-MW(thermal), annular prismatic core concept reactor plants, each configured with an individual helical once-through steam generator steaming a common 280-MW(electric) turbine generator set.

  1. Thermal behavior and ice-table depth within the north polar erg of Mars

    Science.gov (United States)

    Putzig, Nathaniel E.; Mellon, Michael T.; Herkenhoff, Kenneth E.; Phillips, Roger J.; Davis, Brian J.; Ewer, Kenneth J.; Bowers, Lauren M.

    2014-01-01

    We fully resolve a long-standing thermal discrepancy concerning the north polar erg of Mars. Several recent studies have shown that the erg’s thermal properties are consistent with normal basaltic sand overlying shallow ground ice or ice-cemented sand. Our findings bolster that conclusion by thoroughly characterizing the thermal behavior of the erg, demonstrating that other likely forms of physical heterogeneity play only a minor role, and obviating the need to invoke exotic materials. Thermal inertia as calculated from orbital temperature observations of the dunes has previously been found to be more consistent with dust-sized materials than with sand. Since theory and laboratory data show that dunes will only form out of sand-sized particles, exotic sand-sized agglomerations of dust have been invoked to explain the low values of thermal inertia. However, the polar dunes exhibit the same darker appearance and color as that of dunes found elsewhere on the planet that have thermal inertia consistent with normal sand-sized basaltic grains, whereas Martian dust deposits are generally lighter and redder. The alternative explanation for the discrepancy as a thermal effect of a shallow ice table is supported by our analysis of observations from the Mars Global Surveyor Thermal Emission Spectrometer and the Mars Odyssey Thermal Emission Imaging System and by forward modeling of physical heterogeneity. In addition, our results exclude a uniform composition of dark dust-sized materials, and they show that the thermal effects of the dune slopes and bright interdune materials evident in high-resolution images cannot account for the erg’s thermal behavior.

  2. Enhanced RF Behavior Multi-Layer Thermal Insulation.

    Science.gov (United States)

    Mazzinghi, A; Sabbadini, M; Freni, A

    2018-01-08

    This paper shows that it is possible to exploit the modulated metasurface concept to control the unwanted coupling between antennas that are installed on the same satellite. The metasurface is combined with a Multi-Layer thermal Insulation blanket to reduce its specular reflection by spreading the energy incoherently in the surrounding space. In the design, sub-wavelength radiating elements printed on thin substrate have been used to make the metasurface response azimuthally independent, and to keep the weight of blanket down. The comparison between simulations and measurements confirms the validity of the idea.

  3. Preliminary analysis of modeling of Pars and steam injectors to support long-term operation of LWR passive ECCS using a best estimate thermal-hydraulics code

    Energy Technology Data Exchange (ETDEWEB)

    Morales S, J. B.; Sanchez J, J. [UNAM, Facultad de Ingenieria, Circuito Interior s/n, Ciudad Universitaria, 04510 Mexico D. F. (Mexico); Espinosa P, G., E-mail: jaimebmoraless@gmail.co [Universidad Autonoma Metropolitana, Unidad Iztapalapa, Av. San Rafael Atlixco No. 186, Col. Vicentina, 09340 Mexico D. F. (Mexico)

    2010-10-15

    pools. This work presents Pars and steam injectors models, simulations and results of implementations in a best estimated thermal-hydraulics code, which are to be used for evaluations of these systems as long-term supports to ECCS. Lumped parameter models based on heat and mass balances have also been developed to test required additional best estimated code routines needed for the representation of Pars. (Author)

  4. Manometer Behavior Analysis using CATHENA, RELAP and GOTHIC Codes

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Yang Hoon; Han, Kee Soo; Moon, Bok Ja; Jang, Misuk [Nuclear Engineering Service and Solution Co. Ltd., Daejeon (Korea, Republic of)

    2014-05-15

    In this presentation, simple thermal hydraulic behavior is analyzed using three codes to show the possibility of using alternative codes. We established three models of simple u-tube manometer using three different codes. CATHENA (Canadian Algorithm for Thermal hydraulic Network Analysis), RELAP (Reactor Excursion and Leak Analysis Program), GOTHIC (Generation of Thermal Hydraulic Information for Containments) are used for this analysis. CATHENA and RELAP are widely used codes for the analysis of system behavior of CANDU and PWR. And GOTHIC code also has been widely used for the analysis of thermal hydraulic behavior in the containment system. In this paper, the internal behavior of u-tube manometer was analyzed using 3 codes, CATHENA, RELAP and GOTHIC. The general transient behavior is similar among 3 codes. However, the behavior simulated using GOTHIC shows some different trend compared with the results from the other 2 codes at the end of the transient. It would be resulted from the use of different physical model in GOTHIC, which is specialized for the multi-phase thermal hydraulic behavior analysis of containment system unlike the other two codes.

  5. THERMAL-HYDRAULICS PARAMETER ANALYSIS OF THE BANDUNG TRIGA 2000 REACTOR BASED ON CFD AND RELAP5/MOD3.2

    Directory of Open Access Journals (Sweden)

    Reinaldy Nazar

    2015-04-01

    . The result of this study can be used as a valuable information in operating Bandung TRIGA 2000 reactor at the limited power of 1000 kW and revising safety analysis report (SAR of Bandung TRIGA 2000 reactor. Keywords: Bandung TRIGA 2000 reactor, 1000 kW limited power, thermal-hydraulic aspect, computer code of CFD, computer code of RELAP5/Mod3.2.   Reaktor TRIGA 2000 Bandung merupakan hasil upgrading dari reaktor TRIGA Mark II berdaya nominal 1 MW menjadi 2 MW dan telah diresmikan pengoperasiannya pada tahun 2000. Dalam periode tersebut telah terjadi perubahan parameter operasi, terutama parameter yang berkaitan dengan aspek termohidrolik, seperti suhu teras reaktor yang tinggi dan menyebabkan terjadi pembentukan gelembung uap di dalam teras. Hal ini bertentangan dengan aspek keselamatan. Mengingat masalah keselamatan merupakan hal yang utama, maka perlu dilakukan penurunan suhu teras reaktor dan pengurangan pembentukkan gelembung uap di dalam teras, diantaranya dengan mengoperasikan reaktor TRIGA 2000 Bandung pada daya terbatas 1000 kW. Untuk mengetahui tingkat keselamatan pengoperasian reaktor TRIGA 2000 Bandung pada daya 1000 kW, dilakukan analisis karakteristik termohidrolik melalui kajian teoritik menggunakan program computer Computational Fluid Dynamics (CFD dan RELAP5/Mod3.2 (Reactor Excursion and Leak Analysis Program. Hasil kajian menunjukkan bahwa reaktor mencapai kondisi tunak pada daya 1000 kW setelah 1500 detik reaktor kritis, suhu maksimum bahan bakar di dalam teras reaktor berada di posisi C4 dimana suhu maksimum pusat bahan bakar 529,35 °C, suhu maksimum kelongsong bahan bakar 103,12 °C, dan suhu maksimum pendingin pada posisi bahan bakar terkait 90,67 °C. Suhu maksimum kelongsong bahan bakar dan suhu maksimum pendingin yang diperoleh berharga jauh di bawah suhu saturasi 112,4 °C, sehingga pendidihan prajenuh (sub-cooled boiling atau pendidihan saturasi dan pembentukan gelembung uap di dalam teras diprediksi tidak terjadi. Selain itu ketika reaktor

  6. Thermal behavior of a metal embedded fiber Bragg grating sensor

    Science.gov (United States)

    Li, Xiao Chun; Prinz, Fritz; Seim, John

    2001-08-01

    With embedded sensors it is possible to monitor structural parameters at critical locations which are not accessible to ordinary sensors. Recently, the fiber optic sensor has emerged as a promising technology to be integrated with structures. The embedding of fiber optic sensors into composites and some metals, especially those with low melting points, have been reported. However, all reported embedding techniques so far are either complicated or it is difficult to achieve coherent bonding with low residue stresses. Thus, it is of interest to pursue some economical ways to embed fiber optic sensors into metallic structures with low residue stresses. In this work, a new technique is proposed for embedding a fiber optic sensor into metallic structures, such as nickel, with minimized residue stress. Fiber Bragg grating (FBG) sensors have been embedded into nickel structures. The thermal performance of such an embedded FBG sensor is studied. Higher temperature sensitivity is demonstrated for the embedded FBG sensors. For temperature measurements, the embedded FBG sensor yields an accuracy of about 2 °C. Under rapid temperature changes, it is found that thermal stresses due to the temperature gradient in the metallic structures would be the main cause for errors.

  7. Optical, mechanical and thermal behaviors of Nitrilotriacetic acid single crystal

    Science.gov (United States)

    Deepa, B.; Philominathan, P.

    2017-11-01

    An organic nonlinear single crystal of Nitrilotriacetic acid (NTAA) was grown for the first time by employing a simple slow evaporation technique. Single crystal X-ray diffraction (XRD) analysis reveals that the grown crystal belongs to the monoclinic system with noncentrosymmetric space group CC. Fourier transform infrared (FTIR) spectral study ascertains the presence of functional groups in NTAA. The molecular structure of the grown crystal was confirmed by Nuclear Magnetic Resonance (NMR) spectral analysis. The optical parameters such as transmittance, absorption coefficient and band gap were calculated from UV-Visible and fluorescence studies. Dielectric measurements were carried out for different frequency and temperature. The mechanical strength of the grown crystal was measured using Vickers microhardness test. The high thermal stability and the melting point of the grown crystal were also estimated using thermogravimetric (TGA) and differential thermal analyses (DTA). The confirmation of the grown crystals belonging to nonlinear optical crystals was performed by Kurtz-Perry technique and found as suitable candidate for optoelectronics applications.

  8. Numerical model for the thermal behavior of thermocline storage tanks

    Science.gov (United States)

    Ehtiwesh, Ismael A. S.; Sousa, Antonio C. M.

    2017-10-01

    Energy storage is a critical factor in the advancement of solar thermal power systems for the sustained delivery of electricity. In addition, the incorporation of thermal energy storage into the operation of concentrated solar power systems (CSPs) offers the potential of delivering electricity without fossil-fuel backup even during peak demand, independent of weather conditions and daylight. Despite this potential, some areas of the design and performance of thermocline systems still require further attention for future incorporation in commercial CSPs, particularly, their operation and control. Therefore, the present study aims to develop a simple but efficient numerical model to allow the comprehensive analysis of thermocline storage systems aiming better understanding of their dynamic temperature response. The validation results, despite the simplifying assumptions of the numerical model, agree well with the experiments for the time evolution of the thermocline region. Three different cases are considered to test the versatility of the numerical model; for the particular type of a storage tank with top round impingement inlet, a simple analytical model was developed to take into consideration the increased turbulence level in the mixing region. The numerical predictions for the three cases are in general good agreement against the experimental results.

  9. Thermal Behavior Optimization in Multi-MW Wind Power Converter by Reactive Power Circulation

    DEFF Research Database (Denmark)

    Zhou, Dao; Blaabjerg, Frede; Lau, Mogens

    2014-01-01

    for the conditions of constant wind speed and during wind gust. It is concluded that the thermal performance will be improved by injecting proper reactive power circulation within the wind turbine system, thereby being able to reduce the thermal cycling and enhance the reliability of the power converter.......The influence of actively controlled reactive power on the thermal behavior of multi-MW wind power converter with a Doubly-Fed Induction Generator (DFIG) is investigated. First, the allowable range of internal reactive power circulation is mapped depending on the DC-link voltage as well...... as the induction generator and power device capacity. Then, the effects of reactive power circulation on current characteristic and thermal distribution of the two-level back-to-back power converter are analyzed and compared. Finally, the thermal-oriented reactive power control method is introduced to the system...

  10. Thermal Cycling Behavior of Quasi-Columnar YSZ Coatings Deposited by PS-PVD

    Science.gov (United States)

    Yang, Jiasheng; Zhao, Huayu; Zhong, Xinghua; Shao, Fang; Liu, Chenguang; Zhuang, Yin; Ni, Jinxing; Tao, Shunyan

    2017-01-01

    Columnar-structured thermal barrier coatings, owing to their high strain tolerance, are expected for their potential possibilities to substantially extend turbine lives and improve engine efficiencies. In this paper, plasma spray-physical vapor deposition (PS-PVD) process was used to deposit yttria partially stabilized zirconia (YSZ) coatings with quasi-columnar structures. Thermal cyclic tests on burner rigs and thermal shock tests by heating and water-quenching method were involved to evaluate the thermal cycling and thermal shock behaviors of such kind of structured thermal barrier coatings (TBCs). Evolution of the microstructures, phase composition, residual stresses and failure behaviors of quasi-columnar YSZ coatings before and after the thermal tests was investigated. The quasi-columnar coating obtained had an average life of around 623 cycles when the spallation area reached about 10% of the total coating surface during burner rig tests with the coating surface temperature of 1250 °C. Failure of the coating is mainly due to the break and pull-out of center columnar segments.

  11. Comparative evaluation of thermal decomposition behavior and thermal stability of powdered ammonium nitrate under different atmosphere conditions.

    Science.gov (United States)

    Yang, Man; Chen, Xianfeng; Wang, Yujie; Yuan, Bihe; Niu, Yi; Zhang, Ying; Liao, Ruoyu; Zhang, Zumin

    2017-09-05

    In order to analyze the thermal decomposition characteristics of ammonium nitrate (AN), its thermal behavior and stability under different conditions are studied, including different atmospheres, heating rates and gas flow rates. The evolved decomposition gases of AN in air and nitrogen are analyzed with a quadrupole mass spectrometer. Thermal stability of AN at different heating rates and gas flow rates are studied by differential scanning calorimetry, thermogravimetric analysis, paired comparison method and safety parameter evaluation. Experimental results show that the major evolved decomposition gases in air are H2O, NH3, N2O, NO, NO2 and HNO3, while in nitrogen, H2O, NH3, NO and HNO3 are major components. Compared with nitrogen atmosphere, lower initial and end temperatures, higher heat flux and broader reaction temperature range are obtained in air. Meanwhile, higher air gas flow rate tends to achieve lower reaction temperature and to reduce thermal stability of AN. Self-accelerating decomposition temperature of AN in air is much lower than that in nitrogen. It is considered that thermostability of AN is influenced by atmosphere, heating rate and gas flow rate, thus changes of boundary conditions will influence its thermostability, which is helpful to its safe production, storage, transportation and utilization. Copyright © 2017 Elsevier B.V. All rights reserved.

  12. Novel ionic polymeric hydraulic actuators

    Science.gov (United States)

    Shahinpoor, Mohsen; Kim, Kwang J.

    2001-07-01

    It is now well recognized that a strip of ionic polymer- metal composite (IPMC) exhibits a spontaneous bending capability under the influence of an electric potential. A key observation is the appearance and disappearance of water on the expansion and contraction surfaces of the strip, respectively. Such water appearing/disappearing activities occur near the permeable metal electrodes. The imposition of en elctric field causes the mobile cations that are conjugated to the polymeric anions to undergo electrophoretic dynamic migration that can result in local deformation of the material. Such an electrophoretic behavior of the IPMC causes the water to leak out of the permeable electroded boundary so as to lower the actuation performance. This situation is similar to a leaking hydraulic actuator (hydraulic jack), which has the highest force density notwithstanding the compressor unit weight. Herein, a new category of actuators as ionic polymeric hydraulic actuators (IPHA's) is defined. The IPMC is a good example of such ionic polymeric hydraulic actuators. The advantage of ionic polymeric hydraulic actuators is their potential to generate substantially high force densities, theoretically better than current hydraulic actuators. Based upon this ionic polymer hydraulic actuator concept, a certain manufacturing technique was developed to increase the force density of the conventional IPMC's by a factor of two (100% improvement in force). This technology and associated experimental results are presented in this paper.

  13. Thermal Degradation Behavior of Siloxane Elastomer Impregnated Carbon Nanotube Areogel Networks

    Energy Technology Data Exchange (ETDEWEB)

    Lewicki, J P; Worsley, M A

    2010-12-13

    A novel class of nanoporous graphitic carbon foams has been synthesized. Unprecedented properties - electrically conductive, thermally stable (> 1000 C), and mechanically robust. Improved transport properties (DWNT-CA, SWNT-CA) - greater than 100% enhancement in thermal conductivity, 100-400% improvement in electrical conductivity. Rich mechanical deformation behavior (SWNT-CA) - stiff ({approx}100% improvement of elastic modulus), energy dissipation, fracture toughness, and fatigue behavior. Implications for energy-related technologies - hydrogen storage, fusion and fission energy, catalysis, electrochemical energy storage, and composites with foam scaffolds.

  14. The diversity in thermal behavior of novel catanionic cholates: The dominant effect of quaternary ammonium centers

    Energy Technology Data Exchange (ETDEWEB)

    Mihelj, Tea, E-mail: tmihelj@irb.hr [Department of Physical Chemistry, Ruđer Bošković Institute, POB 180, HR-10002 Zagreb (Croatia); Vojta, Danijela [Department of Organic Chemistry and Biochemistry, Ruđer Bošković Institute, POB 180, HR-10002 Zagreb (Croatia); Tomašić, Vlasta [Department of Physical Chemistry, Ruđer Bošković Institute, POB 180, HR-10002 Zagreb (Croatia)

    2014-05-01

    Graphical abstract: - Highlights: • Quaternary ammonium cholates characterized as crystal smectics at room temperature. • The domination of quaternary ammonium center seen through diverse thermal behavior. • Thermal changes of cholates keto–enol balance results with prevalence of one form. • Polynomial regression of isotropisation temperatures vs. dodecyl chain number. • Linear descending trend of isotropisation temperatures with higher headgroup number. - Abstract: The thermal behavior of novel catanionic compounds based on cholate anion was examined. The study explains the effect of the raising dodecyl chain number, as well as of the raising headgroup number in quaternary ammonium salts on their physico-chemical properties. The examined samples are crystal smectic phases at room temperature. Their rich and diverse thermal behavior is seen through polymorphic phase transitions, thermotropic mesomorphism of smectic nature and in some cases, kinetically managed crystallization that lasts in days. The changes of the cholates keto–enol isomer balance during thermal treatment are obtained. For both groups of samples temperatures of isotropisation and the thermodynamic parameters follow polynomial regression. Introduction of the new dodecyl chain leads to more ordered structure, while implementation the new headgroup leads to advanced polymer-like structure. The obtained results demonstrate an effort for getting and controlling the regularity of the physico-chemical and thermotropic properties for new compounds by systematic changing of cationic part of the molecule. These informations could provide in the future the easiest way for selection of potentially new and targeted applicable materials.

  15. Thermal behavior optimization in multi-MW wind power converter by reactive power circulation

    DEFF Research Database (Denmark)

    Zhou, Dao; Blaabjerg, Frede; Lau, Mogens

    2013-01-01

    In the paper, an actively controlled reactive power influence to the thermal behavior of multi-MW wind power converter with Doubly-Fed Induction Generator (DFIG) is investigated. The allowable range of internal reactive power circulation is firstly mapped depending on the DC-link voltage as well ...... in the conditions of constant wind speed and during wind gust. It is concluded that the thermal performance will be improved by injecting proper reactive power circulation in the wind turbine system and thereby be able to reduce the thermal cycling and enhance the reliability.......In the paper, an actively controlled reactive power influence to the thermal behavior of multi-MW wind power converter with Doubly-Fed Induction Generator (DFIG) is investigated. The allowable range of internal reactive power circulation is firstly mapped depending on the DC-link voltage as well...... as the induction generator and power device capacity. Then the effects of reactive power circulation towards current characteristic and thermal distribution of the two-level back-to-back power converter is analyzed and compared. Finally the thermal-oriented reactive power is introduced to the system...

  16. Analysing the Hydraulic Actuator-based Knee Unit Kinematics and Correlating the Numerical Results and Walking Human Knee Joint Behavior

    Directory of Open Access Journals (Sweden)

    K. A. Trukhanov

    2014-01-01

    Full Text Available State-of-the-art machinery development enables people with lost lower limb to continue their previous life despite a loss. International companies dealing with this area pursue a minimization of human behaviour problems because of amputation. Researches to create an optimal design of the artificial knee joint are under way.The work task was to define analytical relationships of changing kinematic parameters of the human walking on the flat surface such as an angle of the knee joint, knee point (moment, definition of reduced knee actuator (A load, as well as to compare obtained results with experimental data.As an A in created design, the article proposes to use a controlled shock absorber based on the hydraulic cylinder.A knee unit is a kinematic two-tier mechanism. One of the mechanism links performs rotational motion, and the other is rotation-translational to provide a rotation of the first one.When studying the hydraulic actuator device dynamics, as a generalized coordinate a coordinate of the piston x (or ρ position is chosen while in the study of link movements an angle β is preferable.Experimental data are obtained for a human with the body weight of 57.6 kg walking on the flat surface to estimate a value of the knee joint angle, speed, acceleration, torque, and capacity in the knee joint and are taken from the published works of foreign authors.A trigonometric approximation was used for fitting the experimental data. The resulting dependence of the reduced load on the stock of A is necessary to perform the synthesis of A. The criterion for linear mechanisms mentioned in the D.N. Popov’s work is advisable to use as a possible criterion for optimization of A.The results obtained are as follows:1. Kinematics linkage mechanism is described using relationships for dependencies of its geometrical parameters, namely a cylinder piston stroke x (or ρ and a links angle β.2. Obtained polynomials of kinematic relationships allow a synthesis of

  17. Study on Subgrain Rotation Behavior at Different Interfaces of a Solder Joint During Thermal Shock

    Science.gov (United States)

    Han, Jing; Tan, Shihai; Guo, Fu

    2016-12-01

    In order to investigate subgrain rotation behavior in the recrystallized region of lead-free solder joints, a ball grid array (BGA) specimen with a cross-sectioned edge row was thermally shocked. Electron backscattered diffraction (EBSD) was used to obtain the microstructure and orientations of Sn grains or subgrains in as-reflowed and thermally shocked conditions. Orientation imaging microscopy (OIM) showed that several subgrains were formed at the tilted twin grain boundaries, near the chip side and near the printed circuit board (PCB) side after 200 thermal shocks due to a highly mismatched coefficient of thermal expansion (CTE) of twin grains. Also, subgrains formed at the chip side and PCB side in the solder joint were selected to research the grain rotation behavior in lead-free solder joints. The analysis of subgrain rotation also indicated that the rotation behavior of subgrains was different between the chip side and PCB side. It was closely related with the large different crystal orientations between the chip side and PCB side. Furthermore, electron backscattered patterns (EBSPs) at several parts of the joint were not obtained after 300 thermal shocks due to the serious deformation caused by mismatched CTE during thermal shock. But 4 subgrains were selected and compared with that of the initial state and 200-thermal shock conditions. The results showed that the subgrains at the chip side were also rotated around the Sn [101] and [001] axes and the subgrains at the PCB side were also rotated around the Sn [100] axis, which indicated a continuous process of subgrain rotation.

  18. Experimental data showing the thermal behavior of a flat roof with phase change material.

    Science.gov (United States)

    Tokuç, Ayça; Başaran, Tahsin; Yesügey, S Cengiz

    2015-12-01

    The selection and configuration of building materials for optimal energy efficiency in a building require some assumptions and models for the thermal behavior of the utilized materials. Although the models for many materials can be considered acceptable for simulation and calculation purposes, the work for modeling the real time behavior of phase change materials is still under development. The data given in this article shows the thermal behavior of a flat roof element with a phase change material (PCM) layer. The temperature and energy given to and taken from the building element are reported. In addition the solid-liquid behavior of the PCM is tracked through images. The resulting thermal behavior of the phase change material is discussed and simulated in [1] A. Tokuç, T. Başaran, S.C. Yesügey, An experimental and numerical investigation on the use of phase change materials in building elements: the case of a flat roof in Istanbul, Build. Energy, vol. 102, 2015, pp. 91-104.

  19. Experimental data showing the thermal behavior of a flat roof with phase change material

    Directory of Open Access Journals (Sweden)

    Ayça Tokuç

    2015-12-01

    Full Text Available The selection and configuration of building materials for optimal energy efficiency in a building require some assumptions and models for the thermal behavior of the utilized materials. Although the models for many materials can be considered acceptable for simulation and calculation purposes, the work for modeling the real time behavior of phase change materials is still under development. The data given in this article shows the thermal behavior of a flat roof element with a phase change material (PCM layer. The temperature and energy given to and taken from the building element are reported. In addition the solid–liquid behavior of the PCM is tracked through images. The resulting thermal behavior of the phase change material is discussed and simulated in [1] A. Tokuç, T. Başaran, S.C. Yesügey, An experimental and numerical investigation on the use of phase change materials in building elements: the case of a flat roof in Istanbul, Build. Energy, vol. 102, 2015, pp. 91–104.

  20. Using thermal and spectroscopic data to investigate the thermal behavior of epinephrine

    Energy Technology Data Exchange (ETDEWEB)

    Bannach, Gilbert; Cervini, Priscila [Instituto de Quimica de Sao Carlos - IQSC/USP, C.P. 780, CEP 13560-970, Sao Carlos, SP (Brazil); Cavalheiro, Eder Tadeu Gomes, E-mail: cavalheiro@iqsc.usp.br [Instituto de Quimica de Sao Carlos - IQSC/USP, C.P. 780, CEP 13560-970, Sao Carlos, SP (Brazil); Ionashiro, Massao [Instituto de Quimica, UNESP, C.P. 355, CEP 14801-970, Araraquara, SP (Brazil)

    2010-02-20

    Epinephrine is a neurotransmitter of the catecholamine class that acts in the mammalian central nervous system. The TG-DTA curves of epinephrine showed that the anhydrous compound starts decomposition at 165 {sup o}C, under the conditions used in this work. The reflectance FTIR spectra and X-ray powder diffraction patterns of epinephrine before and after heating up to 210 {sup o}C, as well as the TG-FTIR spectra of sample heated between 30 and 600 {sup o}C, were obtained and reveled that after heating, structural changes occurred in the sample. At temperatures higher than 205 {sup o}C the thermal decomposition took place with elimination of methylamine in agreement with the first mass loss observed in the TG curve in both air and N{sub 2} atmospheres (TG = 17.0%, calcd. = 17.0%). The melting was observed at 205 {sup o}C (DTA) or 203 {sup o}C (DSC) but this process occurred overlapped with decomposition characteristic of an incongruent melting process.

  1. Overview of pool hydraulic design of Indian prototype fast breeder ...

    Indian Academy of Sciences (India)

    Computational Fluid Dynamics (CFD) plays a critical role in the design of pool type reactors and becomes an increasingly popular tool, thanks to the advancements in computing technology. In this paper, thermal hydraulic characteristics of a fast breeder reactor, design limits and challenging thermal hydraulic investigations ...

  2. Thermal Conductivity and Wear Behavior of HVOF-Sprayed Fe-Based Amorphous Coatings

    Directory of Open Access Journals (Sweden)

    Haihua Yao

    2017-10-01

    Full Text Available To protect aluminum parts in vehicle engines, metal-based thermal barrier coatings in the form of Fe59Cr12Nb5B20Si4 amorphous coatings were prepared by high velocity oxygen fuel (HVOF spraying under two different conditions. The microstructure, thermal transport behavior, and wear behavior of the coatings were characterized simultaneously. As a result, this alloy shows high process robustness during spraying. Both Fe-based coatings present dense, layered structure with porosities below 0.9%. Due to higher amorphous phase content, the coating H-1 exhibits a relatively low thermal conductivity, reaching 2.66 W/(m·K, two times lower than the reference stainless steel coating (5.85 W/(m·K, indicating a good thermal barrier property. Meanwhile, the thermal diffusivity of amorphous coatings display a limited increase with temperature up to 500 °C, which guarantees a steady and wide usage on aluminum alloy. Furthermore, the amorphous coating shows better wear resistance compared to high carbon martensitic GCr15 steel at different temperatures. The increased temperature accelerating the tribological reaction, leads to the friction coefficient and wear rate of coating increasing at 200 °C and decreasing at 400 °C.

  3. Geometrical effects on the concentrated behavior of heat flux in metamaterials thermal harvesting devices

    Science.gov (United States)

    Xu, Guoqiang; Zhang, Haochun; Xie, Ming; Jin, Yan

    2017-10-01

    Thermal harvesting devices based on transformation optics, which can manipulate the heat flux concentration significantly through rational arrangements of the conductivities, have attracted considerable interest owing to several great potential applications of the technique for high-efficiency thermal conversion and collection. However, quantitative studies on the geometrical effects, particularly wedge angles, on the harvesting behaviors are rare. In this paper, we adopt wedge structure-based thermal harvesting schemes, and focus on the effects of the geometrical parameters including the radii ratios and wedge angles on the harvesting performance. The temperature deformations at the boundaries of the compressional region and temperature gradients for the different schemes with varying design parameters are investigated. Moreover, a concept for temperature stabilization was derived to evaluate the fluctuation in the energy distributions. In addition, the effects of interface thermal resistances have been investigated. Considering the changes in the radii ratios and wedge angles, we proposed a modification of the harvesting efficiency to quantitatively assess the concentration performance, which was verified through random tests and previously fabricated devices. In general, this study indicates that a smaller radii ratio contributes to a better harvesting behavior, but causes larger perturbations in the thermal profiles owing to a larger heat loss. We also find that a smaller wedge angle is beneficial to ensuring a higher concentration efficiency with less energy perturbations. These findings can be used to guide the improvement of a thermal concentrator with a high efficiency in reference to its potential applications as novel heat storage, thermal sensors, solar cells, and thermoelectric devices.

  4. Behavior of deep flaws in a thick-wall cylinder under thermal shock loading

    Energy Technology Data Exchange (ETDEWEB)

    Cheverton, R.D.

    1979-01-01

    Behavior of inner-surface flaws in thick-walled vessels was studied in a 991-mm OD x 152 mm wall x 1220 mm length cylinder with toughness properties similar to those for HSST Plate. The initial temperature of 93/sup 0/C and a thermal shock medium of liquid nitrogen (-197/sup 0/C) were employed. The initial flaw selected was a sharp, 16 mm deep, long (1220 mm) axial crack. Crack arrest methodology was shown to be valid for deep flaws under severe thermal shock. (FS)

  5. Phenomenology involved in self-pressurized, natural circulation, low thermo-dynamic quality, nuclear reactors: The thermal-hydraulics of the CAREM-25 reactor

    Energy Technology Data Exchange (ETDEWEB)

    Marcel, C.P., E-mail: christian.marcel@cab.cnea.gov.ar [Instituto Balseiro, 8400 S. C de Bariloche (Argentina); Centro Atomico Bariloche, CNEA, Bustillo 9500, 8400 S. C. de Bariloche (Argentina); Consejo Nacional de Investigaciones Cientificas y Tecnicas (CONICET) (Argentina); Furci, H.F.; Delmastro, D.F. [Instituto Balseiro, 8400 S. C de Bariloche (Argentina); Centro Atomico Bariloche, CNEA, Bustillo 9500, 8400 S. C. de Bariloche (Argentina); Masson, V.P. [Centro Atomico Bariloche, CNEA, Bustillo 9500, 8400 S. C. de Bariloche (Argentina); Consejo Nacional de Investigaciones Cientificas y Tecnicas (CONICET) (Argentina)

    2013-01-15

    Highlights: Black-Right-Pointing-Pointer Low quality, natural circulation, and self-pressurized nuclear reactors are modeled analytically. Black-Right-Pointing-Pointer The feedbacks resulting from the interplay of the acting phenomena are analyzed. Black-Right-Pointing-Pointer By decreasing the nuclear power the core inlet enthalpy increases. Black-Right-Pointing-Pointer The mass flow has to be regulated to set it within a certain range in order. - Abstract: The interwoven phenomena involved in a prototypical self-pressurized natural circulation, low thermo-dynamical quality nuclear reactor such as CAREM-25 are analytically presented. These phenomena present many differences with traditional light water nuclear power plants. The dependence between mass flow and core inlet enthalpy on generated power is found. The need of tuning the mass flow rate in accordance to the design value is found to be important in order to keep the thermal margin and the heat transfer coefficients in the steam generators. The influence of condensation in structures or walls in the upper dome on the two-phase boundary is also studied. The dynamic consequences of all these results are therefore discussed. A numerical code is then used to verify the aforementioned findings and to test the validity of the modeling approximations. From the results it is clear that the way the phenomena interact causes the resulting dynamics in CAREM-25 to be substantially different from that existing in reactors such as PWRs, BWRs and also natural circulation BWRs. It is thus clear that the combination of different effects makes CAREM-25 behavior impossible to be extrapolated from existing knowledge and accumulated experience.

  6. Thermal behavior of latent thermal energy storage unit using two phase change materials: Effects of HTF inlet temperature

    Directory of Open Access Journals (Sweden)

    Fouzi Benmoussa

    2017-09-01

    Full Text Available This work presents a numerical study of the thermal behavior of shell-and-tube latent thermal energy storage (LTES unit using two phase change materials (PCMs. The heat transfer fluid (HTF flow through the inner tube and transfer the heat to PCMs. First, a mathematical model is developed based on the enthalpy formulation and solved through the governing equations. Second, the effects of HTF inlet temperature on the unsteady temperature evolution of PCMs, the total energy stored evolution as well as the total melting time is studied. Numerical results show that for all HTF inlet temperature, melting rate of PCM1 is the fastest and that of PCM2 is the slowest; increasing the HTF inlet temperature considerably increases the temperature evolution of PCMs. The maximum energy stored is observed in PCM2 with high melting temperature and high specific heat; heat storage capacity is large for high HTF inlet temperature. When the HTF inlet temperature increases from 338 K to 353 K, decreasing degree of melting time of PCM2 is the biggest from 1870 s to 490 s, which reduces about 73.8%; decreasing degree of melting time of PCM1 is the smallest from 530 s to 270 s, which reduces about 49.1%.

  7. Fundamentals, current state of the development of, and prospects for further improvement of the new-generation thermal-hydraulic computational HYDRA-IBRAE/LM code for simulation of fast reactor systems

    Science.gov (United States)

    Alipchenkov, V. M.; Anfimov, A. M.; Afremov, D. A.; Gorbunov, V. S.; Zeigarnik, Yu. A.; Kudryavtsev, A. V.; Osipov, S. L.; Mosunova, N. A.; Strizhov, V. F.; Usov, E. V.

    2016-02-01

    The conceptual fundamentals of the development of the new-generation system thermal-hydraulic computational HYDRA-IBRAE/LM code are presented. The code is intended to simulate the thermalhydraulic processes that take place in the loops and the heat-exchange equipment of liquid-metal cooled fast reactor systems under normal operation and anticipated operational occurrences and during accidents. The paper provides a brief overview of Russian and foreign system thermal-hydraulic codes for modeling liquid-metal coolants and gives grounds for the necessity of development of a new-generation HYDRA-IBRAE/LM code. Considering the specific engineering features of the nuclear power plants (NPPs) equipped with the BN-1200 and the BREST-OD-300 reactors, the processes and the phenomena are singled out that require a detailed analysis and development of the models to be correctly described by the system thermal-hydraulic code in question. Information on the functionality of the computational code is provided, viz., the thermalhydraulic two-phase model, the properties of the sodium and the lead coolants, the closing equations for simulation of the heat-mass exchange processes, the models to describe the processes that take place during the steam-generator tube rupture, etc. The article gives a brief overview of the usability of the computational code, including a description of the support documentation and the supply package, as well as possibilities of taking advantages of the modern computer technologies, such as parallel computations. The paper shows the current state of verification and validation of the computational code; it also presents information on the principles of constructing of and populating the verification matrices for the BREST-OD-300 and the BN-1200 reactor systems. The prospects are outlined for further development of the HYDRA-IBRAE/LM code, introduction of new models into it, and enhancement of its usability. It is shown that the program of development and

  8. Analysis of thermal demagnetization behavior of Nd–Fe–B sintered magnets using magnetic domain observation

    Directory of Open Access Journals (Sweden)

    Masaaki Takezawa

    2016-05-01

    Full Text Available We used magnetic domain observation to statistically observe the thermal demagnetization behavior of Nd–Fe–B sintered magnets at elevated temperatures up to 150 °C. Simultaneous magnetization reversal in a hundred adjacent grains occurred at 90 °C because of the magnetic interaction among the grains beyond grain boundaries in the Dysprosium (Dy-free low-coercivity magnet. Conversely, simultaneous magnetization reversal in a hundred grains did not occur in the Dy-added high-coercivity magnets, and the demagnetizing ratio steadily increased with temperature. Furthermore, the addition of Dy induced high thermal stability by eliminating the simultaneous thermal demagnetization, which was caused by the magnetic interaction among the grains.

  9. Improved thermal stability and wettability behavior of thermoplastic polyurethane / barium metaborate composites

    Energy Technology Data Exchange (ETDEWEB)

    Baştürka, Emre; Madakbaş, Seyfullah; Kahraman, Memet Vezir, E-mail: smadakbas@marmara.edu.tr [Department of Chemistry, Marmara University, Istanbul (Turkey)

    2016-03-15

    In this paper, it was targeted to the enhance thermal stability and wettability behavior of thermoplastic polyurethane (TPU) by adding barium metaborate. TPU-Barium metaborate composites were prepared by adding various proportions of barium metaborate to TPU. The chemical structures of the composites were characterised by fourier transform infrared spectroscopy (FTIR) and X-ray diffraction (XRD) analysis. All prepared composites have extremely high Tg and thermal stability as determined from DSC and TGA analysis. All composite materials have the Tg ranging from 15 to 35 °C. The surface morphologies of the composites were investigated by a scanning electron microscopy. Mechanical properties of the samples were characterized with stress-strain test. Hydrophobicity of the samples was determined by the contact angle measurements. The obtained results proved that thermal, hydrophobic and mechanical properties were improved. (author)

  10. Abnormal thermal shock behavior in electrical conductivity of Ti2SnC

    Directory of Open Access Journals (Sweden)

    Linquan Zhang

    2017-08-01

    Full Text Available Some ternary carbide and nitride ceramics have been demonstrated to exhibit abnormal thermal shock behavior in mechanical properties. However, the influence of thermal shock on other properties is not clear. This work reports on the influence of thermal shock on electrical conductivity of Ti2SnC as a representative member of ternary carbides. Abnormal change in electrical conductivity was first demonstrated during quenching Ti2SnC in water at 500–800 °C. The residual electrical conductivity of the quenched Ti2SnC gradually decreased with increasing temperature, but abnormally increased after quenching at 600 °C. The microstructure of surface cracks was characterized. The main mechanism for the abnormal electrical conductivity recovery is that some narrow branching cracks are filled by metallic Sn precipitating from Ti2SnC.

  11. The Correlation between Thermal and Noxious Gas Environments, Pig Productivity and Behavioral Responses of Growing Pigs

    Directory of Open Access Journals (Sweden)

    Won Kyung Chang

    2011-08-01

    Full Text Available Correlations between environmental parameters (thermal range and noxious gas levels and the status (productivity, physiological, and behavioral of growing pigs were examined for the benefit of pig welfare and precision farming. The livestock experiment was conducted at a Seoul National University station in South Korea. Many variations were applied and the physiological and behavioral responses of the growing pigs were closely observed. Thermal and gas environment parameters were different during the summer and winter seasons, and the environments in the treatments were controlled in different manners. In the end, this study finds that factors such as Average Daily Gain (ADG, Adrenocorticotropic Hormone (ACTH, stress, posture, and eating habits were all affected by the controlled environmental parameters and that appropriate control of the foregoing could contribute to the improvement of precision farming and pig welfare.

  12. Characterization of Thermal Behavior of Epoxy Composites Reinforced with Curaua Fibers by Differential Scanning Calorimetry

    Science.gov (United States)

    Barcelos, Mariana A.; Ribeiro, Carolina Gomes D.; Ferreira, Jordana; Vieira, Janaina da S.; Margem, Frederico M.; Monteiro, Sergio N.

    Epoxy composites reinforced with natural lignocellulosic fibers have, in recent times, been gaining attention in engineering areas as lighter and cheaper alternatives for traditional composites such as the "fiberglass". The curaua fiber is the one strongest today being considered as reinforcement of composites for automobile interior parts. In fact, several studies are currently being dedicated to curaua fiber composites since physical and mechanical properties are required for practical uses. In this work, the thermal behavior of epoxy composites reinforced with up to 30 % in volume of curaua fibers was investigated by differential scanning calorimetry, DSC. The results showed endothermic and exothermic events associated with water release and possible molecular chain amorphous transformation. Comparison with similar composites permitted to propose mechanism that explains this DSC thermal behavior.

  13. Hydraulic Power Plant Machine Dynamic Diagnosis

    Directory of Open Access Journals (Sweden)

    Hans Günther Poll

    2006-01-01

    Full Text Available A method how to perform an entire structural and hydraulic diagnosis of prototype Francis power machines is presented and discussed in this report. Machine diagnosis of Francis units consists on a proper evaluation of acquired mechanical, thermal and hydraulic data obtained in different operating conditions of several rotary and non rotary machine components. Many different physical quantities of a Francis machine such as pressure, strains, vibration related data, water flow, air flow, position of regulating devices and displacements are measured in a synchronized way so that a relation of cause an effect can be developed for each operating condition and help one to understand all phenomena that are involved with such kind of machine. This amount of data needs to be adequately post processed in order to allow correct interpretation of the machine dynamics and finally these data must be compared with the expected calculated data not only to fine tuning the calculation methods but also to accomplish fully understanding of the influence of the water passages on such machines. The way how the power plant owner has to operate its Francis machines, many times also determined by a central dispatcher, has a high influence on the fatigue life time of the machine components. The diagnostic method presented in this report helps one to understand the importance of adequate operation to allow a low maintenance cost for the entire power plant. The method how to acquire these quantities is discussed in details together with the importance of correct sensor balancing, calibration and adequate correlation with the physical quantities. Typical results of the dynamic machine behavior, with adequate interpretation, obtained in recent measurement campaigns of some important hydraulic turbines were presented. The paper highlights the investigation focus of the hydraulic machine behavior and how to tailor the measurement strategy to accomplish all goals. Finally some

  14. Macroscopic behavior of fast reactor fuel subjected to simulated thermal transients

    Energy Technology Data Exchange (ETDEWEB)

    Fenske, G.R.; Emerson, J.E.; Savoie, F.E.

    1983-06-01

    High-speed cinematography has been used to characterize the macroscopic behavior of irradiated and unirradiated fuel subjected to thermal transients prototypical of fast reactor transients. The results demonstrate that as the cladding melts, the fuel can disperse via spallation if the fuel contains in excess of approx. 16 ..mu..moles/gm of fission gas. Once the cladding has melted, the macroscopic behavior (time to failure and dispersive nature) was strongly influenced by the presence of volatile fission products and the heating rate.

  15. Thermal-shock behavior of advanced ceramic/composite hot-gas filters

    Energy Technology Data Exchange (ETDEWEB)

    Singh, J.P.; Sutaria, M.; Bielke, W.

    1997-02-01

    The thermal shock/fatigue behavior of monolithic and composite hot-gas candle filters obtained from various manufacturers was evaluated. The composite filters were made of both oxide and nonoxide materials; the monolithic filters were made only of nonoxide materials. During single-cycle thermal shock tests, composite filters show little or no strength degradation when quenched from temperatures between 900 and 1000{degrees}C. At higher quenching temperatures, slow strength degradation was observed. Regular monolithic SiC filters showed no strength degradation when quenched from temperatures up to {approx}700-900{degrees}C, whereas at higher quenching temperatures, the strength decreased at a relatively sharper rate. On the other hand, recrystallized monolithic SiC filters showed higher initial strength and retained this strength to higher quenching temperatures when compared with regular SiC filters. This finding may be related to the difference in the strength of grain boundary phases in the two filters. For thermal cycles between room temperature and 800-1000{degrees}C, composite filters show little (18-24%) strength degradation up to three cycles, beyond which the strength remains unchanged. Similar behavior, with an initial strength drop of 15-28%, was observed for monolithic filter specimens that were thermally cycled between room temperature and 800{degrees}C.

  16. Thermal behavior of underground power transmission lines with SF/sub 6/ insulation

    Energy Technology Data Exchange (ETDEWEB)

    Brueggemann, G.

    1977-12-01

    For the calculation of the stationary load of underground SF/sub 6/ cables, a method is proposed which also takes into account the natural temperature gradient in the thermally not influenced ground as well as the uneven temperature distribution on the outer tube at any given geometric formation. A long-time test confirms the calculations. The transient behavior is described by means of time constants.

  17. Minimization of lumen depreciation in LED lamps using thermal transient behavior analysis and design optimizations.

    Science.gov (United States)

    Khan, M Nisa

    2016-02-10

    We expansively investigate thermal behaviors of various general-purpose light-emitting diode (LED) lamps and apply our measured results, validated by simulation, to establish lamp design rules for optimizing their optical and thermal properties. These design rules provide the means to minimize lumen depreciation over time by minimizing the periods for lamps to reach thermal steady-state while maintaining their high luminous efficacy and omnidirectional light distribution capability. While it is well known that minimizing the junction temperature of an LED leads to a longer lifetime and an increased lumen output, our study demonstrates, for the first time, to the best of our knowledge, that it is also important to minimize the time it takes to reach thermal equilibrium because doing so minimizes lumen depreciation and enhances light output and color stability during operation. Specifically, we have found that, in addition to inadequate heat-sink fin areas for a lamp configuration, LEDs mounted on multiple boards, as opposed to a single board, lead to longer periods for reaching thermal equilibrium contributing to larger lumen depreciation.

  18. From kinetic to collective behavior in thermal transport on semiconductors and semiconductor nanostructures

    Energy Technology Data Exchange (ETDEWEB)

    Tomas, C. de; Lopeandia, A. F.; Alvarez, F. X., E-mail: xavier.alvarez@uab.cat [Department of Physics, Universitat Autònoma de Barcelona, 08193 Bellaterra, Catalonia (Spain); Cantarero, A. [Materials Science Institute, University of Valencia, P. O. Box 22085, 46071 Valencia (Spain)

    2014-04-28

    We present a model which deepens into the role that normal scattering has on the thermal conductivity in semiconductor bulk, micro, and nanoscale samples. Thermal conductivity as a function of the temperature undergoes a smooth transition from a kinetic to a collective regime that depends on the importance of normal scattering events. We demonstrate that in this transition, the key point to fit experimental data is changing the way to perform the average on the scattering rates. We apply the model to bulk Si with different isotopic compositions obtaining an accurate fit. Then we calculate the thermal conductivity of Si thin films and nanowires by only introducing the effective size as additional parameter. The model provides a better prediction of the thermal conductivity behavior valid for all temperatures and sizes above 30 nm with a single expression. Avoiding the introduction of confinement or quantum effects, the model permits to establish the limit of classical theories in the study of the thermal conductivity in nanoscopic systems.

  19. Thermal stability and hydration behavior of ritonavir sulfate: A vibrational spectroscopic approach

    Directory of Open Access Journals (Sweden)

    Kaweri Gambhir

    2015-12-01

    Full Text Available Ritonavir sulfate is a protease inhibitor widely used in the treatment of acquired immunodeficiency syndrome. In order to elucidate the inherent stability and sensitivity characteristics of ritonavir sulfate, it was investigated under forced thermal and hydration stress conditions as recommended by the International Conference on Harmonization guidelines. In addition, competency of vibrational (infrared and Raman spectroscopy was assessed to identify structural changes of the drug symbolizing its stress degradation. High performance liquid chromatography was used as a confirmatory technique for both thermal and hydration stress study, while thermogravimetric analysis/differential thermal analysis and atomic force microscopy substantiated the implementation of vibrational spectroscopy in this framework. The results exhibited high thermal stability of the drug as significant variations were observed in the diffuse reflectance infrared Fourier transform spectra only after the drug exposure to thermal radiations at 100 °C. Hydration behavior of ritonavir sulfate was evaluated using Raman spectroscopy and the value of critical relative humidity was found to be >67%. An important aspect of this study was to utilize vibrational spectroscopic technique to address stability issues of pharmacological molecules, not only for their processing in pharmaceutical industry, but also for predicting their shelf lives and suitable storage conditions.

  20. The Wear behavior of UHMWPE against Surface Modified CP-Titanium by Thermal Oxidation

    Directory of Open Access Journals (Sweden)

    B.T. Prayoga

    2016-12-01

    Full Text Available The effects of thermal oxidation duration on hardness, roughness, and wettability of the CP-titanium surfaces were investigated in this paper. The thermal oxidation treatment was done at 700 oC for 12-36 hours in an air atmosphere. The wear behavior of the UHMWPE sliding against treated thermal oxidation of the CP-titanium was tested by a pin-on-plate tribometer under lubrication of the solution of 75 % distilled water and 25 % bovine serum. The results showed that the layer of the oxide titanium was formed on the surface after being treated by the thermal oxidation for 12-36 hours. The oxide titanium layer was dominated by rutile form of TiO2, that offers an improvement of hardness and wettability of the CP-titanium surfaces. The average wear factor of the UHMWPE reduced significantly when the sliding against of the CP-titanium was modified by the thermal oxidation, and the lowest average wear factor was reached when the sliding against the 12 hour oxidized CP-titanium counterfaces.

  1. Young’s modulus evaluation and thermal shock behavior of a porous SiC/cordierite composite material

    Directory of Open Access Journals (Sweden)

    Pošarac-Marković M.

    2015-01-01

    Full Text Available Porous SiC/Cordierite Composite Material with graphite content (10% was synthesized. Evaluation of Young modulus of elasticity and thermal shock behavior of these samples was presented. Thermal shock behavior was monitored using water quench test, and non destructive methods such are UPVT and image analysis were also used for accompaniment the level of destruction of the samples during water quench test. Based on the level of destruction graphical modeling of critical number of cycles was given. This approach was implemented on discussion of the influence of the graphite content on thermal stability behavior of the samples. [Projekat Ministarstva nauke Republike Srbije, br. III 45012

  2. Quantification and analysis of color stability based on thermal transient behavior in white LED lamps.

    Science.gov (United States)

    Nisa Khan, M

    2017-09-20

    We present measurement and analysis of color stability over time for two categories of white LED lamps based on their thermal management scheme, which also affects their transient lumen depreciation. We previously reported that lumen depreciation in LED lamps can be minimized by properly designing the heat sink configuration that allows lamps to reach a thermal equilibrium condition quickly. Although it is well known that lumen depreciation degrades color stability of white light since color coordinates vary with total lumen power by definition, quantification and characterization of color shifts based on thermal transient behavior have not been previously reported in literature for LED lamps. Here we provide experimental data and analysis of transient color shifts for two categories of household LED lamps (from a total of six lamps in two categories) and demonstrate that reaching thermal equilibrium more quickly provides better stability for color rendering, color tempera