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Sample records for bundle thermal-hydraulic calculations

  1. Thermal Hydraulic Performance of Tight Lattice Bundle

    Science.gov (United States)

    Yamamoto, Yasushi; Akiba, Miyuki; Morooka, Shinichi; Shirakawa, Kenetsu; Abe, Nobuaki

    Recently, the reduced moderation spectrum BWR has been studied. The fast neutron spectrum is obtained through triangular tight lattice fuel. However, there are few thermal hydraulic test data and thermal hydraulic correlation applicable to critical power prediction in such a tight lattice bundle. This study aims to enhance the database of the thermal hydraulic performance of the tight lattice bundle whose rod gap is about 1mm. Therefore, thermal hydraulic performance measurement tests of tight lattice bundles for the critical power, the pressure drop and the counter current flow limiting were performed. Moreover, the correlations to evaluate the thermal-hydraulic performance of the tight lattice bundle were developed.

  2. Post-test calculation of the QUENCH-17 bundle experiment with debris formation and bottom water reflood using thermal hydraulic and severe fuel damage code SOCRAT/V3

    Energy Technology Data Exchange (ETDEWEB)

    Vasiliev, A., E-mail: vasil@ibrae.ac.ru [Nuclear Safety Institute (IBRAE), B. Tulskaya 52, 115191 Moscow (Russian Federation); Stuckert, J., E-mail: juri.stuckert@kit.edu [Karlsruhe Institute of Technology (KIT), Hermann-von-Helmholtz-Platz 1, 76344 Eggenstein-Leopoldshafen (Germany)

    2015-03-15

    Highlights: • Modeling of processes in porous debris regions. • Analysis of coolability of massive debris bed. • Complexity of simulation of flow regime near boiling curve. - Abstract: The thermal hydraulic and SFD (Severe Fuel Damage) best estimate computer modeling code SOCRAT/V3 was used for the post-test analysis of the QUENCH-17 experiment performed at KIT on January 2013. The objective of this test was to examine the formation of a debris bed inside the completely oxidized region of the bundle without melt formation and to investigate the coolability behavior during the reflood. The test bundle for QUENCH-17 test was intentionally changed in comparison to basic QUENCH bundles (usually 21 heated rod simulators) with the emphasis to investigate debris behavior phenomena. Only 12 periphery fuel rod simulators were heated by centerline tungsten heaters. 9 unheated fuel rod simulators were located in the inner part of the test bundle. This is why the massive porous debris formation in the inner part of the bundle was not influenced by the presence of tungsten heaters. The QUENCH-17 test conditions simulated a hypothetical scenario of nuclear power plant severe accident sequence with debris bed formation in which the overheated up to 1800 K core would be flooded from the bottom by ECCS (Emergency Core Cooling System). The QUENCH-17 test included the following phases: (1) heat-up phase (heat-up rate up to 0.25 K/s); (2) oxidation phase (the cladding temperature about 1800 K in hottest region, steam mass flow rate 2 g/s); (3) bottom flood phase (characteristic cooling time about 600 s, water mass flow rate 10 g/s). SOCRAT/V3 computer modeling code was used for calculation of basic thermal hydraulic, oxidation and thermal mechanical behavior during all phases of the experiment. The calculated results are in a good agreement with experimental data which justifies the adequacy of modeling capabilities of SOCRAT code system.

  3. 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.

  4. Optimised Iteration in Coupled Monte Carlo - Thermal-Hydraulics Calculations

    Science.gov (United States)

    Hoogenboom, J. Eduard; Dufek, Jan

    2014-06-01

    This paper describes an optimised iteration scheme for the number of neutron histories and the relaxation factor in successive iterations of coupled Monte Carlo and thermal-hydraulic reactor calculations based on the stochastic iteration method. The scheme results in an increasing number of neutron histories for the Monte Carlo calculation in successive iteration steps and a decreasing relaxation factor for the spatial power distribution to be used as input to the thermal-hydraulics calculation. The theoretical basis is discussed in detail and practical consequences of the scheme are shown, among which a nearly linear increase per iteration of the number of cycles in the Monte Carlo calculation. The scheme is demonstrated for a full PWR type fuel assembly. Results are shown for the axial power distribution during several iteration steps. A few alternative iteration method are also tested and it is concluded that the presented iteration method is near optimal.

  5. Thermal hydraulics of rod bundles: The effect of eccentricity

    Energy Technology Data Exchange (ETDEWEB)

    Chauhan, Amit K., E-mail: amit_fmlab@yahoo.co.in [Fluid Mechanics Laboratory, Department of Applied Mechanics, Indian Institute of Technology Madras, Chennai 600036 (India); Prasad, B.V.S.S.S., E-mail: prasad@iitm.ac.in [Thermal Turbomachines Laboratory, Department of Mechanical Engineering, Indian Institute of Technology Madras, Chennai 600036 (India); Patnaik, B.S.V., E-mail: bsvp@iitm.ac.in [Fluid Mechanics Laboratory, Department of Applied Mechanics, Indian Institute of Technology Madras, Chennai 600036 (India)

    2013-10-15

    Highlights: • Present CFD investigation explores, whole bundle eccentricity for the first time. • Fluid flow and thermal characteristics in various subchannels are analyzed. • Mass flux distribution is particularly analyzed to study eccentricity effect. • Higher eccentricity resulted in a shoot up in rod surface temperature distribution. • Both tangential and radial flow in rod bundles has resulted due to eccentricity. -- Abstract: The effect of eccentricity on the fluid flow and heat transfer through a 19-rod bundle is numerically carried out. When the whole bundle shifts downwards with respect to the outer (pressure) tube, flow redistribution happens. This in turn is responsible for changes in mass flux, pressure and differential flow development in various subchannels. The heat flux imposed on the surface of the fuel rods and the mass flux through the subchannels determines the coolant outlet temperatures. The simulations are performed for a coolant flow Reynolds number of 4 × 10{sup 5}. For an eccentricity value of 0.7, the mass flux in the bottom most subchannel (l) was found to decrease by 10%, while the surface temperature of the fuel rod in the vicinity of this subchannel increased by 250% at the outlet section. Parameters of engineering interest including skin friction coefficient, Nusselt number, etc., have been systematically explored to study the effect of eccentricity on the rod bundle.

  6. Subchannel thermal-hydraulic modeling of an APT tungsten target rod bundle

    Energy Technology Data Exchange (ETDEWEB)

    Hamm, L.L.; Shadday, M.A. Jr.

    1997-09-01

    The planned target for the Accelerator Production of Tritium (APT) neutron source consists of an array of tungsten rod bundles through which D{sub 2}O coolant flows axially. Here, a scoping analysis of flow through an APT target rod bundle was conducted to demonstrate that lateral cross-flows are important, and therefore subchannel modeling is necessary to accurately predict thermal-hydraulic behavior under boiling conditions. A local reactor assembly code, FLOWTRAN, was modified to model axial flow along the rod bundle as flow through three concentric heated annular passages.

  7. Isotope Production Facility Conceptual Thermal-Hydraulic Design Review and Scoping Calculations

    Energy Technology Data Exchange (ETDEWEB)

    Pasamehmetoglu, K.O.; Shelton, J.D.

    1998-08-01

    The thermal-hydraulic design of the target for the Isotope Production Facility (IPF) is reviewed. In support of the technical review, scoping calculations are performed. The results of the review and scoping calculations are presented in this report.

  8. Thermal hydraulic investigations on porous blockage in a prototype sodium cooled fast reactor fuel pin bundle

    Energy Technology Data Exchange (ETDEWEB)

    Raj, M.Naveen; Velusamy, K., E-mail: kvelu@igcar.gov.in; Maity, Ram Kumar

    2016-07-15

    Highlights: • We simulate flow and temperature fields in FBR fuel bundle with porous blockage. • We perform RANS-based CFD simulation for 217 pin bundle of 7 axial pitch lengths. • Flow reduction in fuel bundle due to porous internal blockage is estimated. • Monitoring bulk sodium outlet temperature does not guarantee blockage detection. • Admissible blockage length to avoid sodium boiling is determined. - Abstract: Thermal hydraulic characteristics of sodium flow in a prototype fuel subassembly with porous internal blockage have been investigated by computational fluid dynamics (CFD) simulations. CFD solutions for a subassembly having 217 pin bundle with seven helical pitch length were obtained by parallel processing. The CFD model has been validated against benchmark blockage experiment reported in literature. Wide parametric ranges for blockage radius, porosity, mean particle diameter and location of blockage have been considered. Critical length of blockage that would result in local sodium boiling as a function of aforementioned blockage parameters has been estimated and the parametric zone posing risk of sodium boiling has been identified. Attention has been paid to coolant mixing and flow and temperature fields downstream of the blockage zone. It is seen that for a prototype subassembly with various sections contributing to pressure loss, the total flow reduction is <2.5% for all blockages that can lead to local sodium boiling. This suggests, that global bulk sodium temperature monitoring at subassembly outlet is unlikely to detect slowly growing blockages. Comparing the sodium flow and temperature fields in unblocked and blocked bundles, it is found that the wake-induced temperature non-uniformity persist even upto 3 helical pitch length, highlighting that the sodium temperature non-uniformity at the bundle exit can serve as an efficient blockage indicator, provided that the cross-section temperature is mapped by a proper instrumentation. The peak

  9. ORNL rod-bundle heat-transfer test data. Volume 6. Thermal-hydraulic test facility experimental data report for test 3. 05. 5B - double-ended cold-leg break simulation

    Energy Technology Data Exchange (ETDEWEB)

    Mullins, C.B.; Felde, D.K.; Sutton, A.G.; Gould, S.S.; Morris, D.G.; Robinson, J.J.; Schwinkendorf, K.N.

    1982-05-18

    Thermal-Hydraulic Test Facility (THTF) Test 3.05.5B was conducted by members of the ORNL PWR Blowdown Heat Transfer Separate-Effects Program on July 3, 1980. The objective of the program is to investigate heat transfer phenomena believed to occur in PWRs during accidents, including small and large break loss-of-coolant accidents. Test 3.05.5B was designed to provide transient thermal-hydraulics data in rod bundle geometry under reactor accident-type conditions. Reduced instrument responses are presented. Also included are uncertainties in the instrument responses, calculated mass flows, and calculated rod powers.

  10. Benchmark of SIMULATE5 thermal hydraulics against the Frigg and NUPEC full bundle test experiments

    Energy Technology Data Exchange (ETDEWEB)

    Grandi, G. [Studsvik Scandpower, Inc., Idaho Falls, Idaho (United States); Lindahl, S.-O. [Studsvik Scandpower AB, Vasteras (Sweden)

    2011-07-01

    SIMULATE5 is Studsvik Scandpower's next generation nodal code. The core portion of the thermal hydraulic models of PWR and BWRs are treated as essentially identical, with each assembly having an active channel and a number of parallel water channels. In addition, the BWR assembly may be divided into four radial sub-assemblies. For natural circulation reactors, the BWR thermal hydraulic model is capable of modeling an entire vessel loop: core, chimney, upper plenum, standpipes, steam separators, downcomer, recirculation pumps, and lower plenum. This paper presents results of the validation of the BWR thermal hydraulic model against: (1) pressure drop data measured in the Frigg and NUPEC test facilities; (2) void fraction distribution measured in the Frigg and NUPEC loops; (3) quarter-assembly void fraction measured in the NUPEC experiments and (4) natural and forced circulation flow measurements in the Frigg loop. (author)

  11. A thermal-hydraulic code for transient analysis in a channel with a rod bundle

    Energy Technology Data Exchange (ETDEWEB)

    Khodjaev, I.D. [Research & Engineering Centre of Nuclear Plants Safety, Electrogorsk (Russian Federation)

    1995-09-01

    The paper contains the model of transient vapor-liquid flow in a channel with a rod bundle of core of a nuclear power plant. The computer code has been developed to predict dryout and post-dryout heat transfer in rod bundles of nuclear reactor core under loss-of-coolant accidents. Economizer, bubble, dispersed-annular and dispersed regimes are taken into account. The computer code provides a three-field representation of two-phase flow in the dispersed-annular regime. Continuous vapor, continuous liquid film and entrained liquid drops are three fields. For the description of dispersed flow regime two-temperatures and single-velocity model is used. Relative droplet motion is taken into account for the droplet-to-vapor heat transfer. The conservation equations for each of regimes are solved using an effective numerical technique. This technique makes it possible to determine distribution of the parameters of flows along the perimeter of fuel elements. Comparison of the calculated results with the experimental data shows that the computer code adequately describes complex processes in a channel with a rod bundle during accident.

  12. 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.

  13. Development of Design Technology on Thermal-Hydraulic Performance in Tight-Lattice Rod Bundles: I-Master Plan and Executive Summary

    Science.gov (United States)

    Ohnuki, Akira; Kureta, Masatoshi; Yoshida, Hiroyuki; Tamai, Hidesada; Liu, Wei; Misawa, Takeharu; Takase, Kazuyuki; Akimoto, Hajime

    R&D project to investigate thermal-hydraulic performance in tight-lattice rod bundles for Innovative Water Reactor for Flexible Fuel Cycle has been progressed at Japan Atomic Energy Agency in collaboration with power utilities, reactor vendors and universities since 2002. In this series-study, we will summarize the R&D achievements using large-scale test facility (37-rod bundle with full-height and full-pressure), model experiments and advanced numerical simulation technology. This first paper described the master plan for the development of design technology and showed an executive summary for this project up to FY2005. The thermal-hydraulic characteristics in the tight-lattice configuration were investigated and the feasibility was confirmed based on the experiments. We have developed the design technology including 3-D numerical simulation one to evaluate the effects of geometry/scale on the thermal-hydraulic behaviors.

  14. Test case specifications for coupled neutronics-thermal hydraulics calculation of Gas-cooled Fast Reactor

    Science.gov (United States)

    Osuský, F.; Bahdanovich, R.; Farkas, G.; Haščík, J.; Tikhomirov, G. V.

    2017-01-01

    The paper is focused on development of the coupled neutronics-thermal hydraulics model for the Gas-cooled Fast Reactor. It is necessary to carefully investigate coupled calculations of new concepts to avoid recriticality scenarios, as it is not possible to ensure sub-critical state for a fast reactor core under core disruptive accident conditions. Above mentioned calculations are also very suitable for development of new passive or inherent safety systems that can mitigate the occurrence of the recriticality scenarios. In the paper, the most promising fuel material compositions together with a geometry model are described for the Gas-cooled fast reactor. Seven fuel pin and fuel assembly geometry is proposed as a test case for coupled calculation with three different enrichments of fissile material in the form of Pu-UC. The reflective boundary condition is used in radial directions of the test case and vacuum boundary condition is used in axial directions. During these condition, the nuclear system is in super-critical state and to achieve a stable state (which is numerical representation of operational conditions) it is necessary to decrease the reactivity of the system. The iteration scheme is proposed, where SCALE code system is used for collapsing of a macroscopic cross-section into few group representation as input for coupled code NESTLE.

  15. Development of Design Technology on Thermal-Hydraulic Performance in Tight-Lattice Rod Bundles: II - Rod Bowing Effect on Boiling Transition under Transient Conditions

    Science.gov (United States)

    Liu, Wei; Tamai, Hidesada; Kureta, Masatoshi; Ohnuki, Akira; Akimoto, Hajime

    A thermal-hydraulic feasibility project for an Innovative Water Reactor for Flexible fuel cycle (FLWR) has been performed since 2002. In this R&D project, large-scale thermal-hydraulic tests, several model experiments and development of advanced numerical analysis codes have been carried out. In this paper, we describe the critical power characteristics in a 37-rod tight-lattice bundle with rod bowing under transient states. It is observed that transient Boiling Transition (BT) always occurs axially at exit elevation of upper high-heat-flux region and transversely in the central area of the bundle, which is same as that under steady state. For the postulated power increase and flow decrease cases that may be possibly met in a normal operation of the FLWR, it is confirmed that no BT occurs when Initial Critical Power Ratio (ICPR) is 1.3. Moreover, when the transients are run under severer ICPR that causes BT, the transient critical powers are generally same as the steady ones. The experiments are analyzed with a modified TRAC-BFI code, where Japan Atomic Energy Agency (JAEA) newest critical power correlation is implemented for the BT judgement. The code shows good prediction for the occurrence or the non occurrence of the BT and predicts the BT starting time conservatively. Traditional quasi-steady state prediction of the transient BT is confirmed being applicable for the postulated abnormal transient processes in the tight-lattice bundle with rod bowing.

  16. Development of multi-dimensional thermal hydraulic modeling using mixing factors for wire wrapped fuel pin bundles with inter-subassembly heat transfer in fast reactors

    Energy Technology Data Exchange (ETDEWEB)

    Nishimura, M.; Kamide, H.; Ohshima, H. [Power Reactor and Nuclear Fuel Development Corp., Oarai, Ibaraki (Japan). Oarai Engineering Center

    1996-10-01

    Temperature distributions in fuel subassemblies of fast reactors interactively affect heat transfer from center to outer region of the core (inter-subassembly heat transfer) and cooling capability of an inter-wrapper flow, as well as maximum cladding temperature. The prediction of temperature distribution in the sub-assembly is, therefore one of the important issues for the reactor safety assessment. To treat the complex phenomena in the core, a multi-dimensional thermal hydraulic analysis is the most promising method. From the studies on the multi-dimensional thermal hydraulic modeling for the fuel sub-assemblies, the modeling have been recommended through the analysis of sodium experiments using driver subassembly test rig PLANDTL-DHX and blanket subassembly test rig CCTL-CFR. Computations of steady states experiments in the test rigs using the above modeling showed quite good agreement to the experimental data. In the present study, the use of this modeling was extended to transient analyses, and its applicability was examined. Firstly, non-dimensional parameters used to determine the mixing factors were modified from the ones based on bundle-averaged values to the ones by local values. Secondly, a new threshold function was derived and introduced to cut off the mixing factor of thermal plumes under inertia force dominant conditions. In the results of this validation, the accuracy was comparable between the modeling and the experimental instrumentation. Thus the present modeling is capable of predicting the thermal hydraulic fields of the wire wrapped fuel pin bundles with inter-subassembly heat transfer under the conditions from rated steady operations to transitions toward natural circulation decay heat removal modes. (J.P.N.)

  17. Development of Design Technology on Thermal-Hydraulic Performance in Tight-Lattice Rod Bundles: III - Numerical Evaluation of Fluid Mixing Phenomena using Advanced Interface-Tracking Method -

    Science.gov (United States)

    Yoshida, Hiroyuki; Nagayoshi, Takuji; Takase, Kazuyuki; Akimoto, Hajime

    Thermal-hydraulic design of the current boiling water reactor (BWR) is performed by correlations with empirical results of actual-size tests. However, for the Innovative Water Reactor for Flexible Fuel Cycle (FLWR) core, an actual size test of an embodiment of its design is required to confirm or modify such correlations. Development of a method that enables the thermal-hydraulic design of nuclear reactors without these actual size tests is desired, because these tests take a long time and entail great cost. For this reason we developed an advanced thermal-hydraulic design method for FLWRs using innovative two-phase flow simulation technology. In this study, detailed Two-Phase Flow simulation code using advanced Interface Tracking method: TPFIT is developed to calculate the detailed information of the two-phase flow. We tried to verify the TPFIT code by comparing it with the 2-channel air-water and steam-water mixing experimental results. The predicted result agrees well the observed results and bubble dynamics through the gap and cross flow behavior could be effectively predicted by the TPFIT code, and pressure difference between fluid channels is responsible for the fluid mixing.

  18. Thermal-hydraulic Fortran program for steady-state calculations of plate-type fuel research reactors

    Directory of Open Access Journals (Sweden)

    Khedr Ahmed

    2008-01-01

    Full Text Available The safety assessment of research and power reactors is a continuous process covering their lifespan and requiring verified and validated codes. Power reactor codes all over the world are well established and qualified against real measuring data and qualified experimental facilities. These codes are usually sophisticated, require special skills and consume a lot of running time. On the other hand, most research reactor codes still require much more data for validation and qualification. It is, therefore, of benefit to any regulatory body to develop its own codes for the review and assessment of research reactors. The present paper introduces a simple, one-dimensional Fortran program called THDSN for steady-state thermal-hydraulic calculations of plate-type fuel research reactors. Besides calculating the fuel and coolant temperature distributions and pressure gradients in an average and hot channel, the program calculates the safety limits and margins against the critical phenomena encountered in research reactors, such as the onset of nucleate boiling, critical heat flux and flow instability. Well known thermal-hydraulic correlations for calculating the safety parameters and several formulas for the heat transfer coefficient have been used. The THDSN program was verified by comparing its results for 2 and 10 MW benchmark reactors with those published in IAEA publications and a good agreement was found. Also, the results of the program are compared with those published for other programs, such as the PARET and TERMIC.

  19. Development of additional module to neutron-physic and thermal-hydraulic computer codes for coolant acoustical characteristics calculation

    Energy Technology Data Exchange (ETDEWEB)

    Proskuryakov, K.N.; Bogomazov, D.N.; Poliakov, N. [Moscow Power Engineering Institute (Technical University), Moscow (Russian Federation)

    2007-07-01

    The new special module to neutron-physic and thermal-hydraulic computer codes for coolant acoustical characteristics calculation is worked out. The Russian computer code Rainbow has been selected for joint use with a developed module. This code system provides the possibility of EFOCP (Eigen Frequencies of Oscillations of the Coolant Pressure) calculations in any coolant acoustical elements of primary circuits of NPP. EFOCP values have been calculated for transient and for stationary operating. The calculated results for nominal operating were compared with results of measured EFOCP. For example, this comparison was provided for the system: 'pressurizer + surge line' of a WWER-1000 reactor. The calculated result 0.58 Hz practically coincides with the result of measurement (0.6 Hz). The EFOCP variations in transients are also shown. The presented results are intended to be useful for NPP vibration-acoustical certification. There are no serious difficulties for using this module with other computer codes.

  20. Mars/master coupled system calculation of the OECD MSLB benchmark exercise 3 with refined core thermal-hydraulic nodalization

    Energy Technology Data Exchange (ETDEWEB)

    Jeong, J.J.; Joo, H.G.; Cho, B.O.; Zee, S.Q.; Lee, W.J. [Korea Atomic Energy Research Inst., Daejeon (Korea, Republic of)

    2001-07-01

    To assess the performance of KAERI coupled multi-dimensional system thermal- hydraulics (T/H) and three-dimensional (3-D) kinetics code, MARS/MASTER, Exercise III of the OECD main steam line break benchmark problem is solved. The coupled code is capable of employing an individual flow channel for each fuel assembly as well as lumped ones. The basic analysis model of the reference plant consists of four major components: a 3-D core neutronics model, a 3-D thermal-hydraulic model for the reactor vessel employing lumped flow channels, a refined core T/H model and a 1-D T/H model for coolant system. Calculations were performed with and without the refined core T/H model. The results of the basic calculation performed without the refined core T/H model show that the core power distribution evolves to a highly localized shape due to the presence of a stuck rod, as well as asymmetric flow distribution in the reactor core. The results of the refined core T/H model indicate that the local peaking factor can be reduced by as much as 22 % through accurate representation of the local T/H feedback effects. Nonetheless, the global transient behaviors are not significantly affected. (author)

  1. Validation of Reactor Physics-Thermal hydraulics Calculations for Research Reactors Cooled by the Laminar Flow of Water

    Energy Technology Data Exchange (ETDEWEB)

    Jordan, K. A.; Schubring, D. [Univ. of Florida, Florida (United States); Girardin, G.; Pautz, A. [Swiss Federal Institute of Technology, Zuerich (Switzerland)

    2013-07-01

    A collaboration between the University of Florida and the Swiss Federal Institute of Technology, Lausanne (EPFL) has been formed to develop and validate detailed coupled multiphysics models of the zero-power (100 W) CROCUS reactor at EPFL and the 100 kW University of Florida Training Reactor, for the comprehensive analysis of the reactor behavior under transient (neutronic or thermal-hydraulic induced) conditions. These two reactors differ significantly in the core design and thermal power output, but share unique heat transfer and flow characteristics. They are characterized by single-phase laminar water flow at near-atmospheric pressures in complex geometries with the possibility of mechanically entrained air bubbles. Validation experiments will be designed to expand the validation domain of these existing models, computational codes and techniques. In this process, emphasis will be placed on validation of the coupled models developed to gain confidence in their applicability for safety analysis. EPFL is responsible for the design and implementation of transient experiments to generate a database of reactor parameters (flow distribution, power profile, and power evolution) to be used to validate against code predictions. The transient experiments performed at EPFL will be simulated on the basis of developed models for these tasks. Comparative analysis will be performed with SERPENT and MCNPX reference core models. UF focuses on the generation of the coupled neutron kinetics and thermal-hydraulic models, including implementation of a TRACE/PARCS reactor simulator model, a PARET model, and development of full-field computational fluid dynamics models (using OpenFOAM) for refined thermal-hydraulics physics treatments. In this subtask of the project, the aim is to verify by means of CFD the validity of TRACE predictions for near-atmospheric pressure water flow in the presence of mechanically entrained air bubbles. The scientific understanding of these multiphysics

  2. Progress of the DUPIC fuel compatibility analysis (II) - thermal-hydraulics

    Energy Technology Data Exchange (ETDEWEB)

    Park, Joo Hwan; Choi, Hang Bok

    2005-03-01

    Thermal-hydraulic compatibility of the DUPIC fuel bundle with a 713 MWe Canada deuterium uranium (CANDU-6) reactor was studied by using both the single channel and sub-channel analysis methods. The single channel analysis provides the fuel channel flow rate, pressure drop, critical channel power, and the channel exit quality, which are assessed against the thermal-hydraulic design requirements of the CANDU-6 reactor. The single channel analysis by the NUCIRC code showed that the thermal-hydraulic performance of the DUPIC fuel is not different from that of the standard CANDU fuel. Regarding the local flow characteristics, the sub-channel analysis also showed that the uncertainty of the critical channel power calculation for the DUPIC fuel channel is very small. As a result, both the single and sub-channel analyses showed that the key thermal-hydraulic parameters of the DUPIC fuel channel do not deteriorate compared to the standard CANDU fuel channel.

  3. 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)

  4. 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.

  5. COBRA-IV-I: an interim version of COBRA for thermal-hydraulic analysis of rod bundle nuclear fuel elements and cores

    Energy Technology Data Exchange (ETDEWEB)

    Wheeler, C.L.; Stewart, C.W.; Cena, R.J.; Rowe, D.S.; Sutey, A.M.

    1976-03-01

    The COBRA-IV-I computer code uses the subchannel analysis approach to determine the enthalpy and flow distribution in rod bundles for both steady-state and transient conditions. The steady-state and transient solution schemes used in COBRA-IIIC are still available in COBRA-IV-I as the implicit solution scheme option. In addition to these techniques, a new explicit solution scheme is now available which allows the calculation of severe transients involving flow reversals, recirculations, expulsion and reentry flows, with a pressure or flow boundary condition specified. Significant storage compaction and reduced running times have been achieved to allow the calculation of problems involving hundreds of subchannels.

  6. Development of Design Technology on Thermal-hydraulic Performance in Tight-lattice Rod Bundles: V-Estimation of Void Fraction

    Science.gov (United States)

    Kureta, Masatoshi; Tamai, Hidesada; Yoshida, Hiroyuki; Ohnuki, Akira; Akimoto, Hajime

    An estimation of the void fraction in a tight-lattice rod bundle was needed for the R&D of the Innovative Water Reactor for Flexible Fuel Cycle (FLWR). For this purpose, we measured the void fraction and studied the behaviors of boiling flow. The void fraction was measured by a neutron radiography, a quick-shut-valve technique, and an electro void fraction meter. The data were taken using the 7-, 14-, 19- and 37-rod bundle test sections with the rod gap of 1.0 or 1.3 mm under from atmospheric pressure to 7.2 MPa conditions. A spacer effect test was also carried out. The following estimations were conducted: (1) a similarity of the advanced analysis codes with the 3D void fraction data, (2) the comparisons of the TRAC-BF1 code and a drift-flux model with the 1D data. Followings were made clear: (a) The void fraction becomes lower at the peripheral and higher at the rod gap part of the lower core and at the center of the subchannel of the upper core, (b) the codes calculates the similar distribution to the data, and (c) the TRAC-BF1 and the drift-flux model tends to overestimate the void fraction at the lower quality region, on the other hand at the higher quality, those methods tend to same characteristics to the data. It was confirmed that several special features were existed in the tight-lattice rod bundle but the codes were applicable.

  7. COBRA-IV PC: A personal computer version of COBRA-IV-I for thermal-hydraulic analysis of rod bundle nuclear fuel elements and cores

    Energy Technology Data Exchange (ETDEWEB)

    Webb, B.J.

    1988-01-01

    COBRA-IV PC is a modified version of COBRA-IV-I, adapted for use with most IBM PC and PC-compatible desktop computers. Like COBRA-IV-I, COBRA-IV PC uses the subchannel analysis approach to determine the enthalpy and flow distribution in rod bundles for both steady-state and transient conditions. The steady-state and transient solution schemes used in COBRA-IIIC are still available in COBRA-IV PC as the implicit solution scheme option. An explicit solution scheme is also available, allowing the calculation of severe transients involving flow reversals, recirculations, expulsions, and reentry flows, with a pressure or flow boundary condition specified. In addition, several modifications have been incorporated into COBRA-IV PC to allow the code to run on the PC. These include a reduction in the array dimensions, the removal of the dump and restart options, and the inclusion of several code modifications by Oregon State University, most notably, a critical heat flux correlation for boiling water reactor fuel and a new solution scheme for cross-flow distribution calculations. 7 refs., 8 figs., 1 tab.

  8. Qualification of coupled 3D neutron kinetic/thermal hydraulic code systems by the calculation of a VVER-440 benchmark. Re-connection of an isolated loop

    Energy Technology Data Exchange (ETDEWEB)

    Kotsarev, Alexander; Lizorkin, Mikhail [National Research Centre ' Kurchatov Institute' , Moscow (Russian Federation); Bencik, Marek; Hadek, Jan [UJV Rez, a.s., Rez (Czech Republic); Kozmenkov, Yaroslav; Kliem, Soeren [Helmholtz-Zentrum Dresden-Rossendorf (HZDR) e.V., Dresden (Germany)

    2016-09-15

    The 7th AER dynamic benchmark is a continuation of the efforts to validate the codes systematically for the estimation of the transient behavior of VVER type nuclear power plants. The main part of the benchmark is the simulation of the re-connection of an isolated circulation loop with low temperature in a VVER-440 plant. This benchmark was calculated by the National Research Centre ''Kurchatov Institute'' (with the code ATHLET/BIPR-VVER), UJV Rez (with the code RELAP5-3D {sup copyright}) and HZDR (with the code DYN3D/ATHLET). The paper gives an overview of the behavior of the main thermal hydraulic and neutron kinetic parameters in the provided solutions.

  9. 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)

  10. CFX Analysis of the CANDU Moderator Thermal-Hydraulics in the Stern Lab. Test Facility

    Science.gov (United States)

    Kim, Hyoung Tae

    2014-06-01

    A numerical calculation with the commercial CFD code CFX is conducted for a test facility simulating the CANDU moderator thermal-hydraulics. Two kinds of moderator thermal-hydraulic tests at Stern Laboratories Inc. were performed in the full geometric configuration of the CANDU moderator circulating vessel, which is called a Calandria, housing a matrix of horizontal rod bundles simulating the Calandria tubes. The first of these tests is the pressure drop measurement of a cross flow in the horizontal rod bundles. The other is the local temperature measurement on the cross section of the horizontal cylinder vessel simulating the Calandria. In the present study the full geometric details of the Calandria are incorporated in the grid generation of the computational domain to which the boundary conditions for each experiment are applied. The numerical solutions are reviewed and compared with the available test data.

  11. Analyses of deformation and thermal-hydraulics within a wire-wrapped fuel subassembly in a liquid metal fast reactor by the coupled code system

    Energy Technology Data Exchange (ETDEWEB)

    Uwaba, Tomoyuki, E-mail: uwaba.tomoyuki@jaea.go.jp; Ohshima, Hiroyuki; Ito, Masahiro

    2017-06-15

    Highlights: • The coupled computational code system allowed for mechanical and thermal-hydraulic analyses in a fast reactor fuel subassembly. • In this system interactive calculations between flow area deformations and coolant temperature changes are repeated to their convergence state. • Effects on bundle-duct interaction on coolant temperature distributions were investigated by using the code system. - Abstract: The coupled numerical analysis of mechanical and thermal-hydraulic behaviors was performed for a wire-wrapped fuel pin bundle subassembly irradiated in a fast reactor. For the analysis, the fuel pin bundle deformation analysis code BAMBOO and the thermal-hydraulic analysis code ASFRE exchanged the deformation and temperature analysis results through the iterative calculations to attain convergence corresponding to the static balance between deformation and temperature. The analysis by the coupled code system showed that the radial distribution of coolant temperature in the subassembly tended to flatten as a result of the fuel pin bundle deformation governed by cladding void swelling and irradiation creep. Such flattening of temperature distribution was slightly observed as a result of fuel pin bowings due to the cladding-wire interaction even when no bundle-duct interaction occurred. The effect of the spacer wire-pitch on deformation and thermal-hydraulics was also investigated in this study.

  12. 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.

  13. 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.

  14. 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.

  15. Helical coil thermal hydraulic model

    Science.gov (United States)

    Caramello, M.; Bertani, C.; De Salve, M.; Panella, B.

    2014-11-01

    A model has been developed in Matlab environment for the thermal hydraulic analysis of helical coil and shell steam generators. The model considers the internal flow inside one helix and its associated control volume of water on the external side, both characterized by their inlet thermodynamic conditions and the characteristic geometry data. The model evaluates the behaviour of the thermal-hydraulic parameters of the two fluids, such as temperature, pressure, heat transfer coefficients, flow quality, void fraction and heat flux. The evaluation of the heat transfer coefficients as well as the pressure drops has been performed by means of the most validated literature correlations. The model has been applied to one of the steam generators of the IRIS modular reactor and a comparison has been performed with the RELAP5/Mod.3.3 code applied to an inclined straight pipe that has the same length and the same elevation change between inlet and outlet of the real helix. The predictions of the developed model and RELAP5/Mod.3.3 code are in fairly good agreement before the dryout region, while the dryout front inside the helical pipes is predicted at a lower distance from inlet by the model.

  16. Neutronic, thermal-hydraulics and accident analysis calculations for an irradiation device to be used in the qualification process of dispersion fuels in the IEA-R1 research reactor

    Energy Technology Data Exchange (ETDEWEB)

    Domingos, Douglas Borges; Silva, Antonio Teixeira e; Umbehaun, Pedro Ernesto; Silva, Jose Eduardo Rosa da; Conti, Thadeu das Neves; Yamaguchi, Mitsuo [Instituto de Pesquisas Energeticas e Nucleares (IPEN-CNEN/SP), Sao Paulo, SP (Brazil)], e-mail: douglasborgesdomingos@yahoo.com.br

    2009-07-01

    Neutronic, thermal-hydraulics and accident analysis calculations were developed to estimate the safety of an irradiation device placed in the IEA-R1 reactor core. The irradiation device will be used to receive miniplates of U{sub 3}O{sub 8}-Al e U{sub 3}Si{sub 2}-Al dispersion fuels, LEU type (19.9% of {sup 235}U), with uranium densities of, respectively, 3.0 gU/cm{sup 3} and 4.8gU/cm{sup 3}. The fuel miniplates will be irradiated to nominal {sup 235}U burnup levels of 50% and 80%, in order to qualify the above high-density dispersion fuels to be used in the Brazilian Multipurpose Reactor, now in the conception phase. For the neutronic calculation, the computer code CITATION was utilized. The computer code FLOW was used to calculate the coolant flow rate in the irradiation device, allowing the determination of the fuel miniplate temperatures with the computer model MTRCR-IEA-R1. A postulated Loss of Coolant Accident (LOCA) was analyzed with the computer codes LOSS and TEMPLOCA, allowing the calculation of the fuel miniplate temperatures after the reactor pool draining. The calculations showed that the irradiation of the fuel miniplates will happen without any adverse consequence in the IEA-R1 reactor. (author)

  17. 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)

    2000-03-15

    The main purpose of this study is to develop a thermal hydraulic auditing code for the CANDU reactor, modifying the model if 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.

  18. 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)

  19. 废液处理系统冷却器热工水力计算研究%The Research of Thermal-hydraulic Calculation for Waste Processing System Cooler

    Institute of Scientific and Technical Information of China (English)

    孙圣权; 刘宇昊; 高峰; 马贞钦; 陈先林; 徐江

    2016-01-01

    As one of the key equipment is cooler radioactive in waste processing system,which will determine whether the system is capable of stable operation. In this paper, a mathematical model based on thermal-hydraulic calculation,by calculation and verification,to ensure that the system can meet the key equipment cooler design and use requirements,this method is similar to the equipment design and testing provided useful experience.%冷却器作为放射性废液处理系统中的关键设备之一,直接决定了系统是否能够稳定运行,本文建立基于热工水力计算的数学模型,通过计算与校核,确保系统关键设备冷却器能够满足设计和使用要求,该方法为类似设备的设计和试验提供了可借鉴的经验。

  20. Thermal hydraulics development for CASL

    Energy Technology Data Exchange (ETDEWEB)

    Lowrie, Robert B [Los Alamos National Laboratory

    2010-12-07

    This talk will describe the technical direction of the Thermal-Hydraulics (T-H) Project within the Consortium for Advanced Simulation of Light Water Reactors (CASL) Department of Energy Innovation Hub. CASL is focused on developing a 'virtual reactor', that will simulate the physical processes that occur within a light-water reactor. These simulations will address several challenge problems, defined by laboratory, university, and industrial partners that make up CASL. CASL's T-H efforts are encompassed in two sub-projects: (1) Computational Fluid Dynamics (CFD), (2) Interface Treatment Methods (ITM). The CFD subproject will develop non-proprietary, scalable, verified and validated macroscale CFD simulation tools. These tools typically require closures for their turbulence and boiling models, which will be provided by the ITM sub-project, via experiments and microscale (such as DNS) simulation results. The near-term milestones and longer term plans of these two sub-projects will be discussed.

  1. The analysis of thermal-hydraulic models in MELCOR code

    Energy Technology Data Exchange (ETDEWEB)

    Kim, M. H.; Hur, C.; Kim, D. K.; Cho, H. J. [POhang Univ., of Science and TECHnology, Pohang (Korea, Republic of)

    1996-07-15

    The objective of the present work is to verify the prediction and analysis capability of MELCOR code about the progression of severe accidents in light water reactor and also to evaluate appropriateness of thermal-hydraulic models used in MELCOR code. Comparing the results of experiment and calculation with MELCOR code is carried out to achieve the above objective. Specially, the comparison between the CORA-13 experiment and the MELCOR code calculation was performed.

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

    OpenAIRE

    2013-01-01

    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 be...

  3. 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.

  4. 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.

  5. Development of realistic thermal-hydraulic system analysis codes ; development of thermal hydraulic test requirements for multidimensional flow modeling

    Energy Technology Data Exchange (ETDEWEB)

    Suh, Kune Yull; Yoon, Sang Hyuk; Noh, Sang Woo; Lee, Il Suk [Seoul National University, Seoul (Korea)

    2002-03-01

    This study is concerned with developing a multidimensional flow model required for the system analysis code MARS to more mechanistically simulate a variety of thermal hydraulic phenomena in the nuclear stem supply system. The capability of the MARS code as a thermal hydraulic analysis tool for optimized system design can be expanded by improving the current calculational methods and adding new models. In this study the relevant literature was surveyed on the multidimensional flow models that may potentially be applied to the multidimensional analysis code. Research items were critically reviewed and suggested to better predict the multidimensional thermal hydraulic behavior and to identify test requirements. A small-scale preliminary test was performed in the downcomer formed by two vertical plates to analyze multidimensional flow pattern in a simple geometry. The experimental result may be applied to the code for analysis of the fluid impingement to the reactor downcomer wall. Also, data were collected to find out the controlling parameters for the one-dimensional and multidimensional flow behavior. 22 refs., 40 figs., 7 tabs. (Author)

  6. Determination of thermal-hydraulic loads on reactor internals in a DBA-situation

    Energy Technology Data Exchange (ETDEWEB)

    Ville Lestinen; Timo Toppila [POB 10, 00048 FORTUM (Finland)

    2005-07-01

    Full text of publication follows: According to Finnish regulatory requirements, reactor internals have to stay intact in a design basis accident (DBA) situation, so that control rods can still penetrate into the core. To fulfill this demand some criteria must be followed in periodical in-service inspections. This is the motivation for studying and developing more detailed methods for analysis of thermal-hydraulic loads on reactor internals during the DBA-situation for the Loviisa NPP in Finland. The objective of this research program is to connect thermal-hydraulic and mechanical analysis methods with the goal to produce a reliable method for determination of thermal-hydraulic and mechanical loads on reactor internals in the accident situation. The tools studied are thermal-hydraulic system codes, computational fluid dynamics (CFD) codes and finite element analysis (FEA) codes. This paper concentrates mainly on thermal-hydraulic part of the research, but also the mechanical aspects are discussed. Firstly, the paper includes a short literary review of the available methods to analyse the described problem including both thermal-hydraulic and structural analysis parts. Secondly, different possibilities to carry out thermal-hydraulic analyses have been studied. The DBA-case includes complex physical phenomena and therefore modelling is difficult. The accident situation can be for example LLOCA. When the pipe has broken, the pressure decreases and water starts to evaporate, which consumes energy and that way limits the pressure decrease. After some period of time, the system reaches a new equilibrium state. To perform exact thermal-hydraulic analysis also two phase phenomena must be included. Therefore CFD codes are not capable of modelling the DBA situation very well, but the use of CFD codes requires that the effect of two phase flow must be added somehow. One method to calculate two phase phenomena with CFD codes is to use thermal-hydraulic system codes to calculate

  7. Thermal hydraulic analysis of the annular flow helium heater design

    Science.gov (United States)

    Chen, N. C.; Sanders, J. P.

    1982-05-01

    Core support performance test (CSPT) by use of an existing facility, components flow test loop (CFTL), as part of the high temperature gas cooled reactor (HTGR) application program were conducted. A major objective of the CSPT is to study accelerated corrosion of the core graphite support structure in helium at reactor conditions. Concentration of impurities will be adjusted so that a 6 month test represents the 30 year reactor life. Thermal hydraulic and structural integrity of the graphite specimen, will be studied at high pressure of 7.24 MPa (1050 psi) and high temperature of 1000 deg C in a test vessel. To achieve the required high temperature at the test section, a heater bundle has to be specially designed and properly manufactured. Performance characteristics of the heater which were determined from an analysis based on this design are presented.

  8. Steady thermal hydraulic analysis for a molten salt reactor

    Institute of Scientific and Technical Information of China (English)

    ZHANG Dalin; QIU Suizheng; LIU Changliang; SU Guanghui

    2008-01-01

    The Molten Salt Reactor (MSR) can meet the demand of transmutation and breeding. In this study, theoretical calculation of steady thermal hydraulic characteristics of a graphite-moderated channel type MSR is conducted. The DRAGON code is adopted to calculate the axial and radial power factor firstly. The flow and heat transfer model in the fuel salt and graphite are developed on basis of the fundamental mass, momentum and energy equations. The results show the detailed flow distribution in the core, and the temperature profiles of the fuel salt, inner and outer wall in the nine typical elements along the axial flow direction are also obtained.

  9. 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.

  10. 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 ...

  11. 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

  12. Thermal hydraulic characteristics study of prototype NET and CEA cable-in-conduit conductors (CICCs)

    Energy Technology Data Exchange (ETDEWEB)

    Maekawa, Ryuji

    1995-10-31

    The thermal hydraulic characteristics of low temperature helium in a Cable-in-Conduit Conductor (CICC) significantly affects the overall design and performance of the associated large scale superconducting magnet system. It is essential to understand the transient and steady state behavior of the helium in the conductor. Throughout the development of CICCs, the reduction of flow impedance has been one of the key factors to improving the overall pressure drop. The newly developed CICC for the ITER project has a hybrid cooling scheme: a central channel that is surrounded by bundles, for which the thermal hydraulic characteristics are not well understood. This thesis describes an experimental and analytical investigation of thermal hydraulic characteristics of low temperature helium in conventional and hybrid CICCS. Pressure drop measurements for both NET and CEA conductors have been conducted, using low temperature helium and liquid nitrogen to obtain a range of Reynolds numbers. The results are correlated with classical friction factor and Reynolds number analysis. The flow impedance reduction of the CEA conductor is described by measures of a developed flow model. Thermally induced flow in the CEA conductor has been studied with an inductive heating method. The induced velocity in the central channel is measured by a Pitot tube with steady state Reynolds number up to {approximately}7000. The transient pressure wave propagation has been recorded with pressure transducers placed equally along the conductor. The supercritical helium temperature in the central channel has been measured with the thermometer probe. However, the reduction of the central channel area significantly affects the overall thermal hydraulic characteristics of the conductor. The results suggest the importance of the central channel. A transient heat transfer experiment studied the.transverse heat transfer mechanism in the CEA conductor. The temperatures in the central channel and bundle region

  13. Thermal hydraulic characteristics study of prototype NET and CEA cable-in-conduit conductors (CICCs)

    Energy Technology Data Exchange (ETDEWEB)

    Maekawa, Ryuji

    1995-10-31

    The thermal hydraulic characteristics of low temperature helium in a Cable-in-Conduit Conductor (CICC) significantly affects the overall design and performance of the associated large scale superconducting magnet system. It is essential to understand the transient and steady state behavior of the helium in the conductor. Throughout the development of CICCs, the reduction of flow impedance has been one of the key factors to improving the overall pressure drop. The newly developed CICC for the ITER project has a hybrid cooling scheme: a central channel that is surrounded by bundles, for which the thermal hydraulic characteristics are not well understood. This thesis describes an experimental and analytical investigation of thermal hydraulic characteristics of low temperature helium in conventional and hybrid CICCS. Pressure drop measurements for both NET and CEA conductors have been conducted, using low temperature helium and liquid nitrogen to obtain a range of Reynolds numbers. The results are correlated with classical friction factor and Reynolds number analysis. The flow impedance reduction of the CEA conductor is described by measures of a developed flow model. Thermally induced flow in the CEA conductor has been studied with an inductive heating method. The induced velocity in the central channel is measured by a Pitot tube with steady state Reynolds number up to {approximately}7000. The transient pressure wave propagation has been recorded with pressure transducers placed equally along the conductor. The supercritical helium temperature in the central channel has been measured with the thermometer probe. However, the reduction of the central channel area significantly affects the overall thermal hydraulic characteristics of the conductor. The results suggest the importance of the central channel. A transient heat transfer experiment studied the.transverse heat transfer mechanism in the CEA conductor. The temperatures in the central channel and bundle region

  14. 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.

  15. Thermal Hydraulic Stability in a Coaxial Thermosyphon

    Institute of Scientific and Technical Information of China (English)

    YANG Jianhui; LU Wenqiang; LI Qing; LI Qiang; ZHOU Yuan

    2005-01-01

    The heat transfer and thermal hydraulic stability in a two-phase thermosyphon with coaxial riser and down-comer has been experimentally investigated and theoretically analyzed to facilitate its application in cold neutron source. The flow in a coaxial thermosyphon was studied experimentally for a variety of heating rates, transfer tube lengths, charge capacities, and area ratios. A numerical analysis of the hydraulic balance between the driving pressure head and the resistance loss has also been performed. The results show that the presented coaxial thermosyphon has dynamic performance advantages relative to natural circulation in a boiling water reactor.

  16. 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.

  17. Minerve: thermal-hydraulic phenomena simulation and virtual reality

    Energy Technology Data Exchange (ETDEWEB)

    Laffont, A.; Pentori, B. [EDF R and D, EDF SEPTEN Electricity of France - Research and Development, Department SINETICS, 92 - Clamart (France)

    2003-07-01

    MINERVE is a 3D interactive application representing the thermal-hydraulic phenomena happening in a nuclear plant. Therefore, the 3D geometric model of the French 900 MW PWR installations has been built. The users can interact in real time with this model to see at each step of the simulation what happens in the pipes. The thermal-hydraulic simulation is made by CATHARE-2, which calculates at every time step data on about one thousand meshes (the whole primary circuit, a part of the second circuit, and the Residual Heat Removal System). The simulation covers incidental and accidental cases on these systems. There are two main innovations in MINERVE: In the domain of nuclear plant's visualization, it is to introduce interactive 3D software mechanisms to visualize results of a physical simulation. In the domain of real-time 3D, it is to visualize fluids in a pipe, while they can have several configurations, like bubbles or single liquid phase. These mechanisms enable better comprehension and better visual representation of the possible phenomena. This paper describes the functionalities of MINERVE, and the difficulties to represent fluids with several characteristics like speed, configuration,..., in 3D. On the end, we talk about the future of MINERVE, and more widely of the possible futures of such an application in scientific visualization. (authors)

  18. 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.

  19. 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.

  20. Influence of Spacer Grid Outer Strap on Fuel Assembly Thermal Hydraulic Performance

    Directory of Open Access Journals (Sweden)

    Jingwen Yan

    2014-01-01

    Full Text Available The outer strap as a typical structure of a spacer grid enhances the mechanical strength, decreases hang-up susceptibility, and also influences thermal hydraulic performance, for example, pressure loss, mixing performance, and flow distribution. In the present study, a typical grid spacer with different outer strap designs is adopted to investigate the influence of outer strap design on fuel assembly thermal hydraulic performance by using a commercial computational fluid dynamics (CFD code, ANSYS CFX, and a subchannel analysis code, FLICA. To simulate the outer straps’ influence between fuel assemblies downstream, four quarter-bundles from neighboring fuel assemblies are constructed to form the computational domain. The results show that the outer strap design has a major impact on cross-flow between fuel assemblies and temperature distribution within the fuel assembly.

  1. Thermal-Hydraulic Simulations of Single Pin and Assembly Sector for IVG- 1M Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Kraus, A. [Argonne National Lab. (ANL), Argonne, IL (United States); Garner, P. [Argonne National Lab. (ANL), Argonne, IL (United States); Hanan, N. [Argonne National Lab. (ANL), Argonne, IL (United States)

    2015-01-15

    Thermal-hydraulic simulations have been performed using computational fluid dynamics (CFD) for the highly-enriched uranium (HEU) design of the IVG.1M reactor at the Institute of Atomic Energy (IAE) at the National Nuclear Center (NNC) in the Republic of Kazakhstan. Steady-state simulations were performed for both types of fuel assembly (FA), i.e. the FA in rows 1 & 2 and the FA in row 3, as well as for single pins in those FA (600 mm and 800 mm pins). Both single pin calculations and bundle sectors have been simulated for the most conservative operating conditions corresponding to the 10 MW output power, which corresponds to a pin unit cell Reynolds number of only about 7500. Simulations were performed using the commercial code STAR-CCM+ for the actual twisted pin geometry as well as a straight-pin approximation. Various Reynolds-Averaged Navier-Stokes (RANS) turbulence models gave different results, and so some validation runs with a higher-fidelity Large Eddy Simulation (LES) code were performed given the lack of experimental data. These singled out the Realizable Two-Layer k-ε as the most accurate turbulence model for estimating surface temperature. Single-pin results for the twisted case, based on the average flow rate per pin and peak pin power, were conservative for peak clad surface temperature compared to the bundle results. Also the straight-pin calculations were conservative as compared to the twisted pin simulations, as expected, but the single-pin straight case was not always conservative with regard to the straight-pin bundle. This was due to the straight-pin temperature distribution being strongly influenced by the pin orientation, particularly near the outer boundary. The straight-pin case also predicted the peak temperature to be in a different location than the twisted-pin case. This is a limitation of the straight-pin approach. The peak temperature pin was in a different location from the peak power pin in every case simulated, and occurred at an

  2. Theoretical calculations of thermophysical properties of single-wall carbon nanotube bundles

    Institute of Scientific and Technical Information of China (English)

    Miao Ting-Ting; Song Meng-Xuan; Ma Wei-Gang; Zhang Xing

    2011-01-01

    Carbon nanotube bundles are promising thermal interfacial materials due to their excellent thermal and mechanical characteristics. In this study, the phonon dispersion relations and density of states of the single-wall carbon nanotube bundles are calculated by using the force constant model. The calculation results show that the inter-tube interaction leads to a significant frequency raise of the low frequency modes. To verify the applied calculation method, the specific heat of a single single-wall carbon nanotube is calculated first based on the obtained phonon dispersion relations and the results coincide well with the experimental data. Moreover, the specific heat of the bundles is calculated and exhibits a slight reduction at low temperatures in comparison with that of the single tube. The thermal conductivity of the bundles at low temperatures is calculated by using the ballistic transport model. The calculation results indicate that the inter-tube interaction, i.e. van der Waals interaction, hinders heat transfer and cannot be neglected at extremely low temperatures. For (5, 5) bundles, the relative difference of the thermal conductivity caused by ignoring inter-tube effect reaches the maximum value of 26% around 17 K, which indicates the significant inter-tube interaction effect on the thermal conductivity at low temperatures.

  3. 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.

  4. 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...

  5. 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...

  6. Thermal hydraulic design and decay heat removal of a solid target for a spallation neutron source

    Energy Technology Data Exchange (ETDEWEB)

    Takenaka, N. [Department of Mechanical Engineering, Kobe University, Kobe (Japan)]. E-mail: takenaka@mech.kobe-u.ac.jp; Nio, D. [Hokkaido University, Sapporo (Japan); Kiyanagi, Y. [Hokkaido University, Sapporo (Japan); Mishima, K. [Kyoto University Research Reactor Institute, Kumatori (Japan); Kawai, M. [High Energy Accelerator Research Institute, Tsukuba (Japan); Furusaka, M. [High Energy Accelerator Research Institute, Tsukuba (Japan)

    2005-08-01

    Thermal hydraulic design and thermal stress calculations were conducted for a water-cooled solid target irradiated by a MW-class proton beam for a spallation neutron source. Plate type and rod bundle type targets were examined. The thickness of the plate and the diameter of the rod were determined based on the maximum and the wall surface temperature. The thermal stress distributions were calculated by a finite element method (FEM). The neutronics performance of the target is roughly proportional to its average density. The averaged densities of the designed targets were calculated for tungsten plates, tantalum-clad tungsten plates, tungsten rods sheathed by tantalum and Zircaloy and they were compared with mercury density. It was shown that the averaged density was highest for the tungsten plates and was high for the tantalum cladding tungsten plates, the tungsten rods sheathed by tantalum and Zircaloy in order. They were higher than or equal to that of mercury for the 1-2 MW proton beams. Tungsten target without the cladding or the sheath is not practical due to corrosion by water under irradiation condition. Therefore, the tantalum cladding tungsten plate already made successfully by HIP and the sheathed tungsten rod are the candidate of high performance solid targets. The decay heat of each target was calculated. It was low enough low compared to that of ISIS for the target without tantalum but was about four times as high as that of ISIS when the thickness of the tantalum cladding was 0.5 mm. Heat removal methods of the decay heat with tantalum were examined. It was shown that a special cooling system was required for the target exchange when tantalum was used for the target. It was concluded that the tungsten rod target sheathed with stainless steel or Zircaloy was the most reliable from the safety considerations and had similar neutronics performance to that of mercury.

  7. Thermal hydraulic design and decay heat removal of a solid target for a spallation neutron source

    Science.gov (United States)

    Takenaka, N.; Nio, D.; Kiyanagi, Y.; Mishima, K.; Kawai, M.; Furusaka, M.

    2005-08-01

    Thermal hydraulic design and thermal stress calculations were conducted for a water-cooled solid target irradiated by a MW-class proton beam for a spallation neutron source. Plate type and rod bundle type targets were examined. The thickness of the plate and the diameter of the rod were determined based on the maximum and the wall surface temperature. The thermal stress distributions were calculated by a finite element method (FEM). The neutronics performance of the target is roughly proportional to its average density. The averaged densities of the designed targets were calculated for tungsten plates, tantalum clad tungsten plates, tungsten rods sheathed by tantalum and Zircaloy and they were compared with mercury density. It was shown that the averaged density was highest for the tungsten plates and was high for the tantalum cladding tungsten plates, the tungsten rods sheathed by tantalum and Zircaloy in order. They were higher than or equal to that of mercury for the 1 2 MW proton beams. Tungsten target without the cladding or the sheath is not practical due to corrosion by water under irradiation condition. Therefore, the tantalum cladding tungsten plate already made successfully by HIP and the sheathed tungsten rod are the candidate of high performance solid targets. The decay heat of each target was calculated. It was low enough low compared to that of ISIS for the target without tantalum but was about four times as high as that of ISIS when the thickness of the tantalum cladding was 0.5 mm. Heat removal methods of the decay heat with tantalum were examined. It was shown that a special cooling system was required for the target exchange when tantalum was used for the target. It was concluded that the tungsten rod target sheathed with stainless steel or Zircaloy was the most reliable from the safety considerations and had similar neutronics performance to that of mercury.

  8. Steady-state thermal-hydraulic analysis of SCWR assembly

    Institute of Scientific and Technical Information of China (English)

    Xiaojing LIU; Xu CHENG

    2008-01-01

    Among the six gen-Ⅳ reactor concepts recom-mended by the gen-Ⅳ international forum (GIF), super-critical water-cooled reactor (SCWR), the only reactor with water as coolant, achieves a high thermal efficiency and, subsequently, has economic advantages over the existing reactors due to its high outlet temperature. A thermal-hydraulic analysis of the SCWR assembly is per-formed in this paper using the modified COBRA-Ⅳ code. Two approaches to reduce the hot channel factor are investigated: decreasing the moderator mass flow and increasing the thermal resistance between moderator channel and its adjacent sub-channels. It is shown that heat transfer deterioration cannot be avoided in SCWR fuel assembly. It is, therefore, highly required to calculate the cladding temperature accurately and to preserve the fuel rod cladding integrity under heat transfer deteriora-tion conditions.

  9. On-Line Core Thermal-Hydraulic Model Improvement

    Energy Technology Data Exchange (ETDEWEB)

    In, Wang Kee; Chun, Tae Hyun; Oh, Dong Seok; Shin, Chang Hwan; Hwang, Dae Hyun; Seo, Kyung Won

    2007-02-15

    The objective of this project is to implement a fast-running 4-channel based code CETOP-D in an advanced reactor core protection calculator system(RCOPS). The part required for the on-line calculation of DNBR were extracted from the source of the CETOP-D code based on analysis of the CETOP-D code. The CETOP-D code was revised to maintain the input and output variables which are the same as in CPC DNBR module. Since the DNBR module performs a complex calculation, it is divided into sub-modules per major calculation step. The functional design requirements for the DNBR module is documented and the values of the database(DB) constants were decided. This project also developed a Fortran module(BEST) of the RCOPS Fortran Simulator and a computer code RCOPS-SDNBR to independently calculate DNBR. A test was also conducted to verify the functional design and DB of thermal-hydraulic model which is necessary to calculate the DNBR on-line in RCOPS. The DNBR margin is expected to increase by 2%-3% once the CETOP-D code is used to calculate the RCOPS DNBR. It should be noted that the final DNBR margin improvement could be determined in the future based on overall uncertainty analysis of the RCOPS.

  10. 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.

  11. KUGEL: a thermal, hydraulic, fuel performance, and gaseous fission product release code for pebble bed reactor core analysis

    Energy Technology Data Exchange (ETDEWEB)

    Shamasundar, B.I.; Fehrenbach, M.E.

    1981-05-01

    The KUGEL computer code is designed to perform thermal/hydraulic analysis and coated-fuel particle performance calculations for axisymmetric pebble bed reactor (PBR) cores. This computer code was developed as part of a Department of Energy (DOE)-funded study designed to verify the published core performance data on PBRs. The KUGEL code is designed to interface directly with the 2DB code, a two-dimensional neutron diffusion code, to obtain distributions of thermal power, fission rate, fuel burnup, and fast neutron fluence, which are needed for thermal/hydraulic and fuel performance calculations. The code is variably dimensioned so that problem size can be easily varied. An interpolation routine allows variable mesh size to be used between the 2DB output and the two-dimensional thermal/hydraulic calculations.

  12. Thermal-hydraulic analysis of SWAMUP facility using ATHLET-SC code

    Directory of Open Access Journals (Sweden)

    Zidi eWang

    2015-03-01

    Full Text Available During the loss of coolant accident (LOCA of supercritical water cooled reactor (SCWR, the pressure in the reactor system will undergo a rapid decrease from the supercritical pressure to the subcritical condition. This process is called trans-critical transients, which is of crucial importance for the LOCA analysis of SCWR. In order to simulate the trans-critical transient, a number of system codes for SCWR have been developed up to date. However, the validation work for the trans-critical models in these codes is still missing. The test facility Supercritical WAter MUltiPurpose loop (SWAMUP with 2×2 rod bundle in Shanghai Jiao Tong University (SJTU will be applied to provide test data for code validation. Some pre-test calculations are important and necessary to show the feasibility of the experiment. In this study, trans-critical transient analysis is performed for the SWAMUP facility with the system code ATHLET-SC, which is modified in SJTU, for supercritical water system. This paper presents the system behavior e.g. system pressure, coolant mass flow, cladding temperature during the depressurization. The effects of some important parameters such as heating power, depressurization rate on the system characteristics are also investigated in this paper. Additionally, some sensitivities study of the code models, e.g. heat transfer coefficient, CHF correlation, are analyzed and discussed. The results indicate that the revised system code ATHLET-SC is capable of simulating thermal hydraulic behavior during the trans-critical transient. According to the results, the cladding temperature during the transient is kept at a low value. However, the pressure difference of the heat exchanger after depressurization could reach 6 MPa, which should be considered in the experiment.

  13. Validation of the thermal-hydraulic system code ATHLET based on selected pressure drop and void fraction BFBT tests

    Energy Technology Data Exchange (ETDEWEB)

    Di Marcello, Valentino, E-mail: valentino.marcello@kit.edu; Escalante, Javier Jimenez; Espinoza, Victor Sanchez

    2015-07-15

    Highlights: • Simulation of BFBT-BWR steady-state and transient tests with ATHLET. • Validation of thermal-hydraulic models based on pressure drops and void fraction measurements. • TRACE system code is used for the comparative study. • Predictions result in a good agreement with the experiments. • Discrepancies are smaller or comparable with respect to the measurements uncertainty. - Abstract: Validation and qualification of thermal-hydraulic system codes based on separate effect tests are essential for the reliability of numerical tools when applied to nuclear power plant analyses. To this purpose, the Institute for Neutron Physics and Reactor Technology (INR) at the Karlsruhe Institute of Technology (KIT) is involved in various validation and qualification activities of different CFD, sub-channel and system codes. In this paper, the capabilities of the thermal-hydraulic code ATHLET are assessed based on the experimental results provided within the NUPEC BFBT benchmark related to key Boiling Water Reactors (BWR) phenomena. Void fraction and pressure drops measurements in the BFBT bundle performed under steady-state and transient conditions which are representative for e.g. turbine trip and recirculation pump trip events, are compared with the numerical results of ATHLET. The comparison of code predictions with the BFBT data has shown good agreement given the experimental uncertainty and the results are consistent with the trends obtained with similar thermal-hydraulic codes.

  14. Prototypic Thermal-Hydraulic Experiment in NRU to Simulate Loss-of-Coolant Accidents

    Energy Technology Data Exchange (ETDEWEB)

    Mohr, C. L.; Hesson, G. M.; Russcher, G. E.; Marsh, R. K.; King, L. L.; Wildung, N. J.; Rausch, W. N.; Bennett, W. D.

    1981-04-01

    Quick-look test results are reported for the initial test series of the Loss-of-Coolant Accident (LOCA) Simulation in the National Research Universal {NRU) test program, conducted by Pacific Northwest Laboratory (PNL) for the U.S. Nuclear Regulatory Commission (NRC). This test was devoted to evaluating the thermal-hydraulic characteristics of a full-length light water reactor (LWR) fuel bundle during the heatup, reflood, and quench phases of a LOCA. Experimental results from 28 tests cover reflood rates of 0.74 in./sec to 11 in./sec and delay times to initiate reflood of 3 sec to 66 sec. The results indicate that current analysis methods can predict peak temperatures within 10% and measured quench times for the bundle were significantly less than predicted. For reflood rates of 1 in./sec where long quench times were predicted (>2000 sec}, measured quench times of 200 sec were found.

  15. Prototypic Thermal-Hydraulic Experiment in NRU to Simulate Loss-of-Coolant Accidents

    Energy Technology Data Exchange (ETDEWEB)

    Mohr, C. L.; Hesson, G. M.; Russcher, G. E.; Marsh, R. K.; King, L. L.; Wildung, N. J.; Rausch, W. N.; Bennett, W. D.

    1981-04-01

    Quick-look test results are reported for the initial test series of the Loss-of-Coolant Accident (LOCA) Simulation in the National Research Universal {NRU) test program, conducted by Pacific Northwest Laboratory (PNL) for the U.S. Nuclear Regulatory Commission (NRC). This test was devoted to evaluating the thermal-hydraulic characteristics of a full-length light water reactor (LWR) fuel bundle during the heatup, reflood, and quench phases of a LOCA. Experimental results from 28 tests cover reflood rates of 0.74 in./sec to 11 in./sec and delay times to initiate reflood of 3 sec to 66 sec. The results indicate that current analysis methods can predict peak temperatures within 10% and measured quench times for the bundle were significantly less than predicted. For reflood rates of 1 in./sec where long quench times were predicted (>2000 sec}, measured quench times of 200 sec were found.

  16. Thermal-hydraulic calculation methods for transients and accidents of the reactor cooling system under special consideration of multi-dimensional flows (ATHLET, FLUBOX, CFX). Final report; Thermohydraulische Rechenmethoden zu Transienten und Stoerfaellen im Reaktorkuehlkreislauf unter besonderer Beruecksichtigung mehrdimensionaler Stroemungen (ATHLET, FLUBOX, CFX). Abschlussbericht

    Energy Technology Data Exchange (ETDEWEB)

    Glaeser, H.; Graf, U.; Herb, J.; and others

    2012-02-15

    The project RS1184 „Thermal-hydraulic Calculation Methods for Transients and Accidents of the Reactor Cooling System Under Special Consideration of Multi-Dimensional Flows (ATHLET, FLUBOX, CFX)'' consists of four work packages: 1. Further development of the computer code ATHLET 2. Termination of FLUBOX development and development of an ATHLET-internal 3D module 3. Coupling of ATHLET and CFD code ANSYS CFX as well as CFX model development to simulate three-dimensional flows in the reactor coolant system 4. Prediction capability of computer code ATHLET. One of the superior objectives of the project is to improve the prediction capability of the thermal-hydraulic system code ATHLET, including the simulation of multi-dimensional flow in the reactor vessel. The constitutive equations in ATHLET, especially the momentum equations in ATHLET, are written in one-dimensional form. It was planned to develop the 2D/3D module FLUBOX further and couple it with ATHLET. Due to reasons given in chapter 3, the FLUBOX development was terminated. Instead, the decision was made to develop a fast running internal ATHLET-module. The 2D/3D equations for ATHLET have been derived and were implemented. That strategy allows using all ATHLET models and the ATHLET code structure. An additional advantage is that different numerical schemes of different codes, and consequently a loss of efficiency, will be avoided. A second possibility is the coupling of ATHLET with the CFD code ANSYS CFX. Such a coupled code system will be used in those cases when a part of the simulation area is needed to be calculated with high resolution. Such a detailed modelling cannot be provided by ATHLET-3D. A complete representation of the cooling system by a CFD code cannot be performed due to calculation time. In order to calculate the complete system behavior still with ATHLET, that part to be investigated in more detail, will be replaced by a CFX model. Several new models and improvements of existing

  17. 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.

  18. Condensation heat transfer coefficient with noncondensible gases for heat transfer in thermal hydraulic codes

    Energy Technology Data Exchange (ETDEWEB)

    Banerjee, S.; Hassan, Y.A. [Texas A& M Univ., College Station, TX (United States)

    1995-09-01

    Condensation in the presence of noncondensible gases plays an important role in the nuclear industry. The RELAP5/MOD3 thermal hydraulic code was used to study the ability of the code to predict this phenomenon. Two separate effects experiments were simulated using this code. These were the Massachusetts Institute of Technology`s (MIT) Pressurizer Experiment, the MIT Single Tube Experiment. A new iterative approach to calculate the interface temperature and the degraded heat transfer coefficient was developed and implemented in the RELAP5/MOD3 thermal hydraulic code. This model employs the heat transfer simultaneously. This model was found to perform much better than the reduction factor approach. The calculations using the new model were found to be in much better agreement with the experimental values.

  19. Study on Thermal-Hydraulic Behavior of an Integral Type Reactor under Heaving Condition

    OpenAIRE

    2014-01-01

    A self-developed program was used to study the thermal-hydraulic behavior of an integral type reactor under heaving condition. Comparison of calculated results with the data of experiments performed on a natural circulation loop designed with reference to an integral type reactor of Tsinghua University in inclination, heaving, and rolling motions was carried out. Characteristics of natural circulation in heaving motion and effect of motion parameters on natural circulation were investigated. ...

  20. 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.

  1. 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.; Lee, S. W. [Korea Automic Energy Research Institute, Taejon (Korea, Republic of)

    2004-02-15

    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 second step of the 3 year project, and the main researches were focused on the development of downcorner boiling model. During the current year, the bubble stream model of downcorner has been developed and installed in he auditing code. The model sensitivity analysis has been performed for APR1400 LBLOCA scenario using the modified code. The preliminary calculation has been performed for the experimental test facility using FLUENT and MARS code. The facility for air bubble experiment has been installed. The thermal hydraulic phenomena for VHTR and super critical reactor have been identified for the future application and model development.

  2. 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)

  3. 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.

  4. Sensitivity studies on thermal-hydraulic behavior associated with LOFT total loss-of-feedwater accident

    Energy Technology Data Exchange (ETDEWEB)

    Lee, K.B.; Lee, H.; Ko, H.J.; Seong, H.J.; Jeung, J.S. [Department of Nuclar Engineering, Korea Power Engineering Company, Inc., Gyeonggi-do (Korea, Republic of); Chang, H.K. [Seoul National University, Seoul (Korea, Republic of)

    2004-07-01

    The purpose of this paper is to study the thermal-hydraulic behaviors associated with the Loss-of-Fluid Test (LOFT) Experiment of total loss-of-feedwater accident and its recovery (Test No. L9-1/L3-3), and to perform sensitivity studies to identify parameters having significant effect on the thermal-hydraulic behaviors predicted by RELAP5/MOD3.3-Beta computer code. The Experiment L9-1/L3-3 was a simulation of the total loss-of-feedwater (TLOFW) accident and its recovery modes. Generally, the RELAP5/MOD3.3Beta calculation results were in good agreement with experimental data, even though some differences were also observed. Sensitivity studies showed that several parameters have significant effect on the predicted thermal-hydraulic behaviors. These parameters include the pressurizer spray system loss coefficient, nodalization of the steam generator secondary side, steam generator U-tube heat transfer area, heat transfer coefficient from the LOFT main components to the environment, and the power-operated relief valve (PORV) discharge coefficient. The first three parameters are significant during the short-term transient phase, while the last two are significant during the long-term transient phase. Therefore, the five parameters should be carefully modeled to obtain appropriate calculation results for transient types such as the total-loss-of feedwater accident and its recovery modes. (authors)

  5. 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.

  6. Assessment of uncertainties of the models used in thermal-hydraulic computer codes

    Science.gov (United States)

    Gricay, A. S.; Migrov, Yu. A.

    2015-09-01

    The article deals with matters concerned with the problem of determining the statistical characteristics of variable parameters (the variation range and distribution law) in analyzing the uncertainty and sensitivity of calculation results to uncertainty in input data. A comparative analysis of modern approaches to uncertainty in input data is presented. The need to develop an alternative method for estimating the uncertainty of model parameters used in thermal-hydraulic computer codes, in particular, in the closing correlations of the loop thermal hydraulics block, is shown. Such a method shall feature the minimal degree of subjectivism and must be based on objective quantitative assessment criteria. The method includes three sequential stages: selecting experimental data satisfying the specified criteria, identifying the key closing correlation using a sensitivity analysis, and carrying out case calculations followed by statistical processing of the results. By using the method, one can estimate the uncertainty range of a variable parameter and establish its distribution law in the above-mentioned range provided that the experimental information is sufficiently representative. Practical application of the method is demonstrated taking as an example the problem of estimating the uncertainty of a parameter appearing in the model describing transition to post-burnout heat transfer that is used in the thermal-hydraulic computer code KORSAR. The performed study revealed the need to narrow the previously established uncertainty range of this parameter and to replace the uniform distribution law in the above-mentioned range by the Gaussian distribution law. The proposed method can be applied to different thermal-hydraulic computer codes. In some cases, application of the method can make it possible to achieve a smaller degree of conservatism in the expert estimates of uncertainties pertinent to the model parameters used in computer codes.

  7. 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

  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. 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.

  10. Effect of fuel pin ballooning on the sub-channel thermal hydraulics during small break loca for Indian PHWRS

    Energy Technology Data Exchange (ETDEWEB)

    Mukhopadhyay, D.; Behera, G.H.; Bandopadhyay, S.K.; Gupta, S.K. [Bhabha Atomic Research Centre, Div. Reactor Safety, Bombay (India)

    2001-07-01

    Effect of fuel pin ballooning on the subchannel thermal-hydraulics during a small break (0.25%) located at the Reactor Inlet Feeder (RIF) has been studied for Indian PHWRs. The break leads to a low flow situation in the affected reactor channel along with delayed reactor trip. Higher power to flow ratio in the inner subchannels in comparison to outer subchannel of a 19 pin fuel bundle causes early 2-phase condition causing the flow to by pass from the inner ones to outer ones. This causes the fuel pins to experience different temperatures. Fuel pin ballooning causes reduction in the subchannel areas and further flow redistribution takes place. The transient subchannel thermal-hydraulic conditions along the reactor channel are very much different due to the power distribution and pressure drop. (authors)

  11. Thermal-hydraulic Analysis in the Pool of PGSFR including the Shielding

    Energy Technology Data Exchange (ETDEWEB)

    Yoon, Jung; Lee, Taeho [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2014-05-15

    Various design issues relate to this region, and one of them is thermal-hydraulic behavior when shielding exists inside the reactor vessel. The shielding is used for the blockage of the radiation emitted by the reactor core. The shielding is installed at the Intermediate Heat eXchanger (IHX), core shroud, and redan region at the top of core. However, this structure disturbs the normal flow path and heat transfer of the primary heat transfer system. In this study, the multi-dimensional thermal-hydraulic characteristics in the pool of PGSFR including the shielding are analyzed. Also these results are compared to a case in which no shielding is installed. A thermal-hydraulic analysis in the pool of the PGSFR considering the shielding structure are performed using STAR-CCM+. The internal major components of the pool inside are modeled, and calculations are performed with a normal operation condition. Also, these results are compared to a no shielding case. The flow and temperature changes owing to the shielding structure at a redan inside are shown, but the overall flow and temperature distributions in both cases are substantially similar. Also the physical properties such as the flow rate, temperature, and static pressure at each major point are almost the same. These results are utilized in the arrangement of the reactor internal structure and design of the shielding structure.

  12. 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)

  13. Thermal-hydraulic design and transient evaluation of a small long-life HTR

    Energy Technology Data Exchange (ETDEWEB)

    Ding, Ming [Harbin Engineering University, Nantong Street 145, 150001 Harbin (China); Delft University of Technology, Mekelweg 15, 2629 JB Delft (Netherlands); Kloosterman, Jan Leen, E-mail: j.l.kloosterman@tudelft.nl [Delft University of Technology, Mekelweg 15, 2629 JB Delft (Netherlands)

    2013-02-15

    Highlights: ► We present the thermal-hydraulic evaluations of a small, long-life and block-type HTR using the DALTON/THERMIX code system. ► A cross section generation methodology is developed and verified for the diffusion calculations of the small HTR. ► The thermal-hydraulic characteristics of the small HTR during pressurized loss of forced-cooling incidents are compared with depressurized loss of forced-cooling ones. ► The thermal-hydraulic characteristics of a cylindrical core are compared with an annular one. ► Thermal power limit of the small HTR is investigated based on depressurized loss of forced-cooling incidents. -- Abstract: Small long-life high temperature gas-cooled reactors (HTRs) may provide electricity or heat for remote areas or industrial users in developed and/or developing countries. Moreover, small HTRs have advantages over large nuclear reactors of demonstrated inherent safety, transportability, modular construction, and flexible site selection. This paper presents the thermal-hydraulic evaluations of the U-Battery, which is a small, long-life and block-type HTR using the DALTON/THERMIX code system. The thermal-hydraulic characteristics of a cylindrical design and an annular design of the U-Battery were evaluated for loss of forced-cooling (LOFC) incidents including depressurized LOFC (DLOFC) and pressurized LOFC (PLOFC) incidents. The calculations show that the stronger natural circulation during the PLOFC makes the reactor core cool faster than during the DLOFC, flattens the radial solid temperature distribution, and transfers more heat from the hot regions (bottom and center of the reactor core) to cold regions (top and periphery of the reactor core). Although the natural circulation in the reactor core is so weak that it is neglected during the DLOFC, the decay heat is removed passively by conduction without any violation of the temperature limits for the 20 MWth U-Battery. The comparisons of the cylindrical and annular reactor

  14. Thermal hydraulic and mechanical analysis of CH HCSB TBM

    Institute of Scientific and Technical Information of China (English)

    WANG Xiao-yu; FENG Kai-ming; ZHANG Guo-shu; YUAN Tao

    2006-01-01

    Based on the structure design and results of neutronics analysis of the CH HCSB TBM (Chinese helium cooled solid breeder test blanket module), thermal hydraulic and mechanical analyses have been carried out. Results show that the design of the CH HCSB TBM is reasonable and acceptable.

  15. 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.

  16. Inert matrix fuel neutronic, thermal-hydraulic, and transient behavior in a light water reactor

    Science.gov (United States)

    Carmack, W. J.; Todosow, M.; Meyer, M. K.; Pasamehmetoglu, K. O.

    2006-06-01

    Currently, commercial power reactors in the United States operate on a once-through or open cycle, with the spent nuclear fuel eventually destined for long-term storage in a geologic repository. Since the fissile and transuranic (TRU) elements in the spent nuclear fuel present a proliferation risk, limit the repository capacity, and are the major contributors to the long-term toxicity and dose from the repository, methods and systems are needed to reduce the amount of TRU that will eventually require long-term storage. An option to achieve a reduction in the amount, and modify the isotopic composition of TRU requiring geological disposal is 'burning' the TRU in commercial light water reactors (LWRs) and/or fast reactors. Fuel forms under consideration for TRU destruction in light water reactors (LWRs) include mixed-oxide (MOX), advanced mixed-oxide, and inert matrix fuels. Fertile-free inert matrix fuel (IMF) has been proposed for use in many forms and studied by several researchers. IMF offers several advantages relative to MOX, principally it provides a means for reducing the TRU in the fuel cycle by burning the fissile isotopes and transmuting the minor actinides while producing no new TRU elements from fertile isotopes. This paper will present and discuss the results of a four-bundle, neutronic, thermal-hydraulic, and transient analyses of proposed inert matrix materials in comparison with the results of similar analyses for reference UOX fuel bundles. The results of this work are to be used for screening purposes to identify the general feasibility of utilizing specific inert matrix fuel compositions in existing and future light water reactors. Compositions identified as feasible using the results of these analyses still require further detailed neutronic, thermal-hydraulic, and transient analysis study coupled with rigorous experimental testing and qualification.

  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. Determination of thermal hydraulic data of GHARR-1 under reactivity insertion transients using the PARET/ANL code

    Energy Technology Data Exchange (ETDEWEB)

    Adoo, N.A., E-mail: nanakwame10@hotmail.com [School of Nuclear and Allied Sciences, University of Ghana, Legon, P.O. Box AE 1, Atomic, Accra (Ghana); Nyarko, B.J.B.; Akaho, E.H.K. [School of Nuclear and Allied Sciences, University of Ghana, Legon, P.O. Box AE 1, Atomic, Accra (Ghana); National Nuclear Research Institute, Ghana Atomic Energy Commission, P.O. Box LG 80, Accra (Ghana); Alhassan, E.; Agbodemegbe, V.Y.; Bansah, C.Y.; Della, R. [School of Nuclear and Allied Sciences, University of Ghana, Legon, P.O. Box AE 1, Atomic, Accra (Ghana)

    2011-12-15

    The PARET/ANL code has been adapted by the IAEA for testing transient behaviour in research reactors since it provides a coupled thermal hydrodynamic and point kinetics capability for estimating thermal hydraulic margins. A two-channel power peaking profile of the Ghana Research Reactor-1 (GHARR-1) has been developed for the PARET/ANL (Version 7.3; 2007) using the Monte Carlo N-Particle code (MCNP) to determine the thermal hydraulic data for reactivity insertion transients in the range of 2.0 Multiplication-Sign 10{sup -3} {Delta}k/k to 5.5 Multiplication-Sign 10{sup -3} {Delta}k/k. Peak clad and coolant temperatures ranged from 59.18 Degree-Sign C to 112.36 Degree-Sign C and 42.95 Degree-Sign C to 79.42 Degree-Sign C respectively. Calculated safety margins (DNBR) satisfied the MNSR thermal hydraulic design criteria for which no boiling occurs in the reactor core. The generated thermal hydraulic data demonstrated a high inherent safety feature of GHARR-1 for which the high negative reactivity feedback of the moderator limits power excursion and consequently the escalation of the clad temperature.

  19. 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.

  20. 堆内辐照在线产氚氦冷陶瓷氚增殖剂包层球床组件热工流体力学计算分析%Thermal-hydraulic Calculation and Analysis on Helium Cooled Ceramic Breeder Pebble Bed Assembly for In-pile Irradiation & In-situ Tritium Extraction

    Institute of Scientific and Technical Information of China (English)

    郭春秋; 解家春; 刘兴民

    2013-01-01

    在研究堆内进行辐照在线产氚试验是ITER计划专项国内配套研究项目之一.本工作主要针对研究堆内辐照氦冷陶瓷氚增殖剂包层(简称陶瓷)球床组件的试验技术要求,评估辐照陶瓷球床组件设计方案的可行性.通过对不同的陶瓷球床组件结构参数和组件在堆内的不同辐照位置,进行热工流体力学设计计算,得到满足要求的入堆辐照陶瓷球床组件设计方案.%In-pile irradiation & in-situ tritium extraction experiment is one of associated domestic research projects in ITER special program.According to the technical requirements of in-pile irradiation experiment of helium cooled ceramic breeder (ceramic) pebble bed assembly in a research reactor,the feasibility of the design for the in-pile irradiation & in-situ tritium extraction experiment of ceramic pebble bed assembly was evaluated.By conducting thermal-hydraulic design calculation with different in-pile irradiation channels,locations and structure parameters for ceramic pebble bed assembly,a reasonable design scheme of ceramic pebble bed assembly satisfying the design requirements for in-pile irradiation was obtained.

  1. 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

  2. Horizontal steam generator thermal hydraulic simulation in typical steady and transient conditions

    Energy Technology Data Exchange (ETDEWEB)

    Rabiee, Ataollah, E-mail: rabiee@shirazu.ac.ir; Kamalinia, Amir Hossein; Haddad, Kamal

    2016-08-15

    Highlights: • Simulation of the horizontal steam generator with the available code in typical normal and transient operations. • Replacement of tube bundle with a porous media due to the complexity of the SG geometry. • Simulation of typical transient mode of the VVER 440 steam generator, loss of feed water accident. - Abstract: Thermal hydraulic analysis of the steam generators as one of the main components of the power cycle in pressurized water reactor (PWR) is crucial in the design and safety of the nuclear power plants. Two phase flow field simulation near the tube bundles is important in obtaining logical numerical results however the complexity of the tube bundles due to geometry and arrangement makes the numerical analysis complicated. In this research tube bundle has been assumed as the porous media and the outlet boundary condition as the one of the main challenge in these kind of simulations has been optimized according to similar researches. In order to adjust and tune the available computational fluid dynamic (CFD) code, pressure drop of the typical kettle reboiler tube bundle in two various heat fluxes and vapor volume fraction distribution in VVER 1000 steam generator in normal operation have been investigated. The typical transient mode of the VVER 440 steam generator, loss of feed water accident, has been studied eventually. It was observed that obtained vapor volume fraction can predict experimental data with more accuracy than the similar researches and would be increased with the elevation during the accident. On the other hand, pressure drop and level of the feed water value reduces through time and show good adoption with the measurements.

  3. 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)

  4. 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.

  5. 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.

  6. Portable Life Support Subsystem Thermal Hydraulic Performance Analysis

    Science.gov (United States)

    Barnes, Bruce; Pinckney, John; Conger, Bruce

    2010-01-01

    This paper presents the current state of the thermal hydraulic modeling efforts being conducted for the Constellation Space Suit Element (CSSE) Portable Life Support Subsystem (PLSS). The goal of these efforts is to provide realistic simulations of the PLSS under various modes of operation. The PLSS thermal hydraulic model simulates the thermal, pressure, flow characteristics, and human thermal comfort related to the PLSS performance. This paper presents modeling approaches and assumptions as well as component model descriptions. Results from the models are presented that show PLSS operations at steady-state and transient conditions. Finally, conclusions and recommendations are offered that summarize results, identify PLSS design weaknesses uncovered during review of the analysis results, and propose areas for improvement to increase model fidelity and accuracy.

  7. 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.

  8. 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.

  9. Calculation of Quad-Cities Central Bundle Documented by the U.S. in FY98 Using Russian Computer Codes

    Energy Technology Data Exchange (ETDEWEB)

    Pavlovichev, A.M.

    2001-06-19

    The report presents calculation results of isotopic composition of irradiated fuel performed for the Quad Cities-1 reactor bundle with UO{sub 2} and MOX fuel. The MCU-REA code was used for calculations. The code is developed in Kurchatov Institute, Russia. The MCU-REA results are compared with the experimental data and HELIOS code results.

  10. Primary system thermal hydraulics of future Indian fast reactors

    Energy Technology Data Exchange (ETDEWEB)

    Velusamy, K., E-mail: kvelu@igcar.gov.in [Thermal Hydraulics Section, Indira Gandhi Centre for Atomic Research, Kalpakkam 603102 (India); Natesan, K.; Maity, Ram Kumar; Asokkumar, M.; Baskar, R. Arul; Rajendrakumar, M.; Sarathy, U. Partha; Selvaraj, P.; Chellapandi, P. [Thermal Hydraulics Section, Indira Gandhi Centre for Atomic Research, Kalpakkam 603102 (India); Kumar, G. Senthil; Jebaraj, C. [AU-FRG Centre for CAD/CAM, Anna University, Chennai 600 025 (India)

    2015-12-01

    Highlights: • We present innovative design options proposed for future Indian fast reactor. • These options have been validated by extensive CFD simulations. • Hotspot factors in fuel subassembly are predicted by parallel CFD simulations. • Significant safety improvement in the thermal hydraulic design is quantified. - Abstract: As a follow-up to PFBR (Indian prototype fast breeder reactor), many FBRs of 500 MWe capacity are planned. The focus of these future FBRs is improved economy and enhanced safety. They are envisaged to have a twin-unit concept. Design and construction experiences gained from PFBR project have provided motivation to achieve an optimized design for future FBRs with significant design changes for many critical components. Some of the design changes include, (i) provision of four primary pipes per primary sodium pump, (ii) inner vessel with single torus lower part, (iii) dome shape roof slab supported on reactor vault, (iv) machined thick plate rotating plugs, (v) reduced main vessel diameter with narrow-gap cooling baffles and (vi) safety vessel integrated with reactor vault. This paper covers thermal hydraulic design validation of the chosen options with respect to hot and cold pool thermal hydraulics, flow requirement for main vessel cooling, inner vessel temperature distribution, safety analysis of primary pipe rupture event, adequacy of decay heat removal capacity by natural convection cooling, cold pool transient thermal loads and thermal management of top shield and reactor vault.

  11. Sensitivity theory for reactor thermal-hydraulics problems

    Energy Technology Data Exchange (ETDEWEB)

    Oblow, E. M.

    1978-07-01

    A sensitivity theory based on reactor physics experience was successfully developed for a reactor thermal-hydraulics problem. The new theory is derived for the case of non-linear, transient heat and mass transfer in a typical reactor subassembly. Suitable adjoint equations for heat and fluid flow are presented along with methods for deriving the sources and boundary and final conditions for these equations. Expressions for the sensitivity of any integral temperature response to problem input data are also presented. The theory is applied to a sample problem describing the steady-state thermal-hydraulic conditions in a CRBR fuel channel. For this case, sensitivity coefficients are derived for several thermal response functions (i.e., peak clad and peak fuel temperature) for all physical input data (i.e., the heat transfer coefficient, thermal conductivities, etc.). A typical uncertainty analysis for peak clad and peak fuel temperature was also performed using uncertainty information about the physical data. Conclusions are drawn about the applicability of this approach to more general problems and the procedures for its implementation in conjunction with large safety or thermal-hydraulics codes are outlined. The method is also compared with currently used response surface techniques.

  12. Thermal-Hydraulic Performance of Scrubbing Nozzle Used for CFVS

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Hyun Chul; Lee, Doo Yong; Jung, Woo Young; Lee, Jong Chan; Kim, Gyu Tae [FNC TECH, Yongin (Korea, Republic of)

    2016-05-15

    A Containment Filtered Venting System (CFVS) is the most interested device to mitigate a threat against containment integrity under the severe accident of nuclear power plant by venting with the filtration of the fission products. FNC technology and partners have been developed the self-priming scrubbing nozzle used for the CFVS which is based on the venturi effect. The thermal-hydraulic performances such as passive scrubbing water suction as well as pressure drop across the nozzle have been tested under various thermal-hydraulic conditions. The two types of test section have been built for testing the thermal-hydraulic performance of the self-priming scrubbing nozzle. Through the visualization loop, the liquid suction performance through the slit, pressure drop across the nozzle are measured. The passive water suction flow through the suction slit at the throat is important parameter to define the scrubbing performance of the self-priming scrubbing nozzle. The water suction flow is increased with the increase of the overhead water level at the same inlet gas flow. It is not so much changed with the change of inlet gas flow at the overhead water level.

  13. 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.

  14. VISTA : thermal-hydraulic integral test facility for SMART reactor

    Energy Technology Data Exchange (ETDEWEB)

    Choi, K. Y.; Park, H. S.; Cho, S.; Park, C. K.; Lee, S. J.; Song, C. H.; Chung, M. K. [KAERI, Taejon (Korea, Republic of)

    2003-07-01

    Preliminary performance tests were carried out using the thermal-hydraulic integral test facility, VISTA (Experimental Verification by Integral Simulation of Transients and Accidents), which has been constructed to simulate the SMART-P. The VISTA facility is an integral test facility including the primary and secondary systems as well as safety-related Passive Residual Heat Removal (PRHR) systems. Its scaled ratio with respect to the SMART-P is 1/1 in height and 1/96 in volume and heater power. Several steady states and power changing tests have been carried out to verify the overall thermal hydraulic primary and secondary characteristics in the range of 10% to 100% power operation. As for the preliminary results, the steady state conditions were found to coincide with the expected design values of the SMART-P. But the major thermal hydraulic parameters are greatly affected by the initial water level and the nitrogen pressure in the reactor's upper annular cavity. The power step/ramp changing tests are successfully carried out and the system responses are observed. The primary natural circulation operation is achieved, but advanced control logics need to be developed to reach the natural circulation mode without pressure excursion. In the PRHR transient tests, the natural circulation flow rate through the PRHR system was found to be about 10 percent in the early phases of PRHR operation.

  15. A new thermal hydraulics code coupled to agent for light water reactor analysis

    Science.gov (United States)

    Eklund, Matthew Deric

    A new numerical model for coupling a thermal hydraulics method based on the Drift Flux and Homogeneous Equilibrium Mixture (HEM) models, with a deterministic neutronics code system AGENT (Arbitrary Geometry Neutron Transport), is developed. Named the TH thermal hydraulics code, it is based on the mass continuity, momentum, and energy equations integrated with appropriate relations for liquid and vapor phasic velocities. The modified conservation equations are then evaluated in one-dimensional (1D) steady-state conditions for LWR coolant subchannel in the axial direction. This permits faster computation times without sacrificing significant accuracy, as compared to other three-dimensional (3D) codes such as RELAP5/TRACE. AGENT is a deterministic neutronics code system based on the Method of Characteristics to solve the 2D/3D neutron transport equation in current and future reactor systems. The coupling scheme between the TH and AGENT codes is accomplished by computing the normalized fission rate profile in the LWR fuel elements by AGENT. The normalized fission rate profile is then transferred to the TH thermal hydraulics code for computing the reactor coolant properties. In conjunction with the 1D axial TH code, a separate 1D radial heat transfer model within the TH code is used to determine the average fuel temperature at each node where coolant properties are calculated. These properties then are entered into Scale 6.1, a criticality analysis code, to recalculate fuel pin neutron interaction cross sections based on thermal feedback. With updated fuel neutron interaction cross sections, the fission rate profile is recalculated in AGENT, and the cycle continues until convergence is reached. The TH code and coupled AGENT-TH code are benchmarked against the TRACE reactor analysis software, showing required agreement in evaluating the basic reactor parameters.

  16. 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.

  17. Bootstrap and Order Statistics for Quantifying Thermal-Hydraulic Code Uncertainties in the Estimation of Safety Margins

    Directory of Open Access Journals (Sweden)

    Enrico Zio

    2008-01-01

    Full Text Available In the present work, the uncertainties affecting the safety margins estimated from thermal-hydraulic code calculations are captured quantitatively by resorting to the order statistics and the bootstrap technique. The proposed framework of analysis is applied to the estimation of the safety margin, with its confidence interval, of the maximum fuel cladding temperature reached during a complete group distribution blockage scenario in a RBMK-1500 nuclear reactor.

  18. Uniform versus Nonuniform Axial Power Distribution in Rod Bundle CHF Experiments

    Directory of Open Access Journals (Sweden)

    Baowen Yang

    2014-01-01

    Full Text Available Rod bundle experiments with axially uniform and nonuniform heat fluxes are examined to explore the potential limitations of using uniform rod bundle CHF data for CHF correlation development of light water reactors with nonuniform axial power distribution (APD. The case of upstream burnout is presented as an example of unique phenomena associated with nonuniform rod bundle CHF experiments. It is a result from combined effect of axial nonuniform power shape and different interchannel mixing mechanisms. In addition, several key parameters are investigated with respect to their potential impacts on the thermal-hydraulic behaviors between rod bundles with uniform and nonuniform APDs. This type of misrepresentation cannot be amended or compensated through the use of correction factors due to the lack of critical information in the uniform rod bundle CHF testing as well as the fundamental difference in the underlining driving mechanisms. Other potential issues involved with the use of uniform rod bundle CHF data for nonuniform APD system applications also present strong evidence concerning the limitations and inadequacy of using uniform rod bundle CHF data for the correlation, prediction, and design limit calculation for safety analysis.

  19. Project W-320 thermal hydraulic model benchmarking and baselining

    Energy Technology Data Exchange (ETDEWEB)

    Sathyanarayana, K.

    1998-09-28

    Project W-320 will be retrieving waste from Tank 241-C-106 and transferring the waste to Tank 241-AY-102. Waste in both tanks must be maintained below applicable thermal limits during and following the waste transfer. Thermal hydraulic process control models will be used for process control of the thermal limits. This report documents the process control models and presents a benchmarking of the models with data from Tanks 241-C-106 and 241-AY-102. Revision 1 of this report will provide a baselining of the models in preparation for the initiation of sluicing.

  20. First wall thermal hydraulic models for fusion blankets

    Energy Technology Data Exchange (ETDEWEB)

    Fillo, J A

    1980-01-01

    Subject to normal and off-normal reactor conditions, thermal hydraulic models of first walls, e.g., a thermal mass barrier, a tubular shield, and a radiating liner are reviewed. Under normal operation the plasma behaves as expected in a predicted way for transient and steady-state conditions. The most severe thermal loading on the first wall occurs when the plasma becomes unstable and dumps its energy on the wall in a very short period of time (milliseconds). Depending on the plasma dump time and area over which the energy is deposited may result in melting of the first wall surface, and if the temperature is high enough, vaporization.

  1. The Thermal Hydraulic Test of SFR Irradiation Test Capsule

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Chul Yong; Park, Sung Jae; Choo, Kee Nam [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2010-10-15

    Irradiation test using a capsule has been performed for fuel or material performance test in the test reactor. Irradiation capsule for the HANARO reactor must satisfy the hydraulic conditions and structural integrity from a out-pile test. Items of a out-pile test are pressure drop, flow-induced vibration test and endurance test. The results of test must show that a test capsule satisfy the HANARO operational requirement with sufficient margin. This paper describes the devices of thermal hydraulic test and the results of SFR test capsule

  2. 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.

  3. 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)

  4. Specifications for a coupled neutronics thermal-hydraulics SFR test case

    Science.gov (United States)

    Tassone, A.; Smirnov, A. D.; Tikhomirov, G. V.

    2017-01-01

    Coupling neutronics/thermal-hydraulics calculations for the design of nuclear reactors are a growing trend in the scientific community. This approach allows to properly represent the mutual feedbacks between the neutronic distribution and the thermal-hydraulics properties of the materials composing the reactor, details which are often lost when separate analysis are performed. In this work, a test case for a generation IV sodium-cooled fast reactor (SFR), based on the ASTRID concept developed by CEA, is proposed. Two sub-assemblies (SA) characterized by different fuel enrichment and layout are considered. Specifications for the test case are provided including geometrical data, material compositions, thermo-physical properties and coupling scheme details. Serpent and ANSYS-CFX are used as reference in the description of suitable inputs for the performing of the benchmark, but the use of other code combinations for the purpose of validation of the results is encouraged. The expected outcome of the test case are the axial distribution of volumetric power generation term (q‴), density and temperature for the fuel, the cladding and the coolant.

  5. 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.

  6. Measurement of basic thermal-hydraulic characteristics under the test facility and reactor conditions

    Energy Technology Data Exchange (ETDEWEB)

    Eduard A Boltenko; Victor P Sharov [Elektrogorsk Research and Engineering Center, EREC, Bezimyannaja Street, 6, Elektrogorsk, Moscow Region, 142530 (Russian Federation); Dmitriy E Boltenko [State Scientific Center of Russian Federation IPPE, Bondarenko Square, Obhinsk, Kaluga Region, 249020 (Russian Federation)

    2005-07-01

    Full text of publication follows: The nuclear power of Russia is based on the reactors of two types: water-water - WWER and uranium - graphite channel RBMK. The nuclear power development is possible with performance of the basic condition - level of nuclear power plants (NPP) safety should satisfy the rigid requirements. The calculated proof of NPPs safety made by means of thermal-hydraulic codes of improved estimation, verified on experimental data is the characteristic of this level. The data for code verification can be obtained at the integral facilities simulating a circulation circuit of NPP with the basic units and intended for investigation of circuit behaviour in transient and accident conditions. For verification of mathematical models in transient and accident conditions, development of physically reasonable methods for definition of the various characteristics of two-phase flow the experimental data, as the integrated characteristics of a flow, and data on the local characteristics and structure of a flow is necessary. For safety assurance of NPP it is necessary to monitor and determine the basic thermalhydraulic characteristics of reactor facility (RF). It is possible to refer coolant flow-rate, core input and output water temperature, heat-power. The description of the EREC works in the field completion and adaptation of certain methods with reference to measurements in dynamic modes of test facility conditions and development of methods for measurements of basic thermal-hydraulic characteristics of reactor facilities is presented in the paper. (authors)

  7. Fundamental approaches for analysis thermal hydraulic parameter for Puspati Research Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Hashim, Zaredah, E-mail: zaredah@nm.gov.my; Lanyau, Tonny Anak, E-mail: tonny@nm.gov.my; Farid, Mohamad Fairus Abdul; Kassim, Mohammad Suhaimi [Reactor Technology Centre, Technical Support Division, Malaysia Nuclear Agency, Ministry of Science, Technology and Innovation, Bangi, 43000, Kajang, Selangor Darul Ehsan (Malaysia); Azhar, Noraishah Syahirah [Universiti Teknologi Malaysia, 80350, Johor Bahru, Johor Darul Takzim (Malaysia)

    2016-01-22

    The 1-MW PUSPATI Research Reactor (RTP) is the one and only nuclear pool type research reactor developed by General Atomic (GA) in Malaysia. It was installed at Malaysian Nuclear Agency and has reached the first criticality on 8 June 1982. Based on the initial core which comprised of 80 standard TRIGA fuel elements, the very fundamental thermal hydraulic model was investigated during steady state operation using the PARET-code. The main objective of this paper is to determine the variation of temperature profiles and Departure of Nucleate Boiling Ratio (DNBR) of RTP at full power operation. The second objective is to confirm that the values obtained from PARET-code are in agreement with Safety Analysis Report (SAR) for RTP. The code was employed for the hot and average channels in the core in order to calculate of fuel’s center and surface, cladding, coolant temperatures as well as DNBR’s values. In this study, it was found that the results obtained from the PARET-code showed that the thermal hydraulic parameters related to safety for initial core which was cooled by natural convection was in agreement with the designed values and safety limit in SAR.

  8. FX2-TH: a two-dimensional nuclear reactor kinetics code with thermal-hydraulic feedback

    Energy Technology Data Exchange (ETDEWEB)

    Shober, R.A.; Daly, T.A.; Ferguson, D.R.

    1978-10-01

    FX2-TH is a two-dimensional, time-dependent nuclear reactor kinetics program with thermal and hydraulic feedback. The neutronics model used is multigroup neutron diffusion theory. The following geometry options are available: x, r, x-y, r-z, theta-r, and triangular. FX2-TH contains two basic thermal and hydraulic models: a simple adiabatic fuel temperature calculation, and a more detailed model consisting of an explicit representation of a fuel pin, gap, clad, and coolant. FX2-TH allows feedback effects from both fuel temperature (Doppler) and coolant temperature (density) changes. FX2-TH will calculate a consistent set of steady state conditions by iterating between the neutronics and thermal-hydraulics until convergence is reached. The time-dependent calculation is performed by the use of the improved quasistatic method. A disk editing capability is available. FX2-TH is operational on IBM system 360 or 370 computers and on the CDC 7600.

  9. Calculation of relative permeability in reservoir engineering using an interacting triangular tube bundle model

    Institute of Scientific and Technical Information of China (English)

    Jinxun Wang; Mingzhe Dong; Jun Yao

    2012-01-01

    Analytical expressions of relative permeability are derived for an interacting cylindrical tube bundle model.Equations for determining relative permeability curves from both the interacting uniform and interacting serial types of triangular tube bundle models are presented.Model parameters affecting the trend of relative permeability curves are discussed.Interacting triangular tube bundle models are used to history-match laboratory displacement experiments to determine the relative permeability curves of actual core samples.By adjusting model parameters to match the history of oil production and pressure drop,the estimated relative permeability curves provide a connection between the macroscopic flow behavior and the pore-scale characteristics of core samples.

  10. 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.

  11. Thermal Hydraulic Analysis on Containment Filtered Venting System

    Energy Technology Data Exchange (ETDEWEB)

    Bang, Young Suk; Park, Tong Kyu; Lee, Doo Yong; Lee, Byung Chul [FNC Technology Co. Ltd., Yongin (Korea, Republic of); Lee, Sang Won; Kim, Hyeong Taek [KHNP-Central Research Institute, Daejeon (Korea, Republic of)

    2014-05-15

    In this study, the thermal hydraulic conditions (e. g. pressure and flow rate) at each component have been examined and the sensitivity analysis on CFVS design parameters (e. g. water inventory, volumetric flow rate). The purpose is to know the possible range of flow conditions at each component to determine the optimum size of filtration system. GOTHIC code has been used to simulate the thermal-hydraulic behavior inside of CFVS. The behavior of flows in the CFVS has been investigated. The vessel water level and the flow rates during the CFVS operation are examined. It was observed that the vessel water level would be changed significantly due to steam condensation/thermal expansion and steam evaporation. Therefore, the vessel size and the initial water inventory should be carefully determined to keep the minimum water level required for filtration components and not to flood the components in the upper side of the vessel. It has been also observed that the volumetric flow rate is maintained during the CFVS operation, which is beneficial for pool scrubbing units. However, regarding the significant variations at the orifice downstream, careful design would be necessary.

  12. Teaching Thermal Hydraulics & Numerical Methods: An Introductory Control Volume Primer

    Energy Technology Data Exchange (ETDEWEB)

    D. S. Lucas

    2004-10-01

    A graduate level course for Thermal Hydraulics (T/H) was taught through Idaho State University in the spring of 2004. A numerical approach was taken for the content of this course since the students were employed at the Idaho National Laboratory and had been users of T/H codes. The majority of the students had expressed an interest in learning about the Courant Limit, mass error, semi-implicit and implicit numerical integration schemes in the context of a computer code. Since no introductory text was found the author developed notes taught from his own research and courses taught for Westinghouse on the subject. The course started with a primer on control volume methods and the construction of a Homogeneous Equilibrium Model (HEM) (T/H) code. The primer was valuable for giving the students the basics behind such codes and their evolution to more complex codes for Thermal Hydraulics and Computational Fluid Dynamics (CFD). The course covered additional material including the Finite Element Method and non-equilibrium (T/H). The control volume primer and the construction of a three-equation (mass, momentum and energy) HEM code are the subject of this paper . The Fortran version of the code covered in this paper is elementary compared to its descendants. The steam tables used are less accurate than the available commercial version written in C Coupled to a Graphical User Interface (GUI). The Fortran version and input files can be downloaded at www.microfusionlab.com.

  13. Numerical Modeling of a Thermal-Hydraulic Loop and Test Section Design for Heat Transfer Studies in Supercritical Fluids

    Science.gov (United States)

    McGuire, Daniel

    A numerical tool for the simulation of the thermal dynamics of pipe networks with heat transfer has been developed with the novel capability of modeling supercritical fluids. The tool was developed to support the design and deployment of two thermal-hydraulic loops at Carleton University for the purpose of heat transfer studies in supercritical and near-critical fluids. First, the system was characterized based on its defining features; the characteristic length of the flow path is orders of magnitude larger than the other characteristic lengths that define the system's geometry; the behaviour of the working fluid in the supercritical thermodynamic state. An analysis of the transient thermal behaviour of the model's domains is then performed to determine the accuracy and range of validity of the modeling approach for simulating the transient thermal behaviour of a thermal-hydraulic loop. Preliminary designs of three test section geometries, for the purpose of heat transfer studies, are presented in support of the overall design of the Carleton supercritical thermal-hydraulic loops. A 7-rod-bundle, annular and tubular geometries are developed with support from the new numerical tool. Materials capable of meeting the experimental requirements while operating in supercritical water are determined. The necessary geometries to satisfy the experimental goals are then developed based on the material characteristics and predicted heat transfer behaviour from previous simulation results. An initial safety analysis is performed on the test section designs, where they are evaluated against the ASME Boiler, Pressure Vessel, and Pressure Piping Code standard, required for safe operation and certification.

  14. Implementation of CFD module in the KORSAR thermal-hydraulic system code

    Energy Technology Data Exchange (ETDEWEB)

    Yudov, Yury V.; Danilov, Ilia G.; Chepilko, Stepan S. [Alexandrov Research Inst. of Technology (NITI), Sosnovy Bor (Russian Federation)

    2015-09-15

    The Russian KORSAR/GP (hereinafter KORSAR) computer code was developed by a joint team from Alexandrov NITI and OKB ''Gidropress'' for VVER safety analysis and certified by the Rostechnadzor of Russia in 2009. The code functionality is based on a 1D two-fluid model for calculation of two-phase flows. A 3D CFD module in the KORSAR computer code is being developed by Alexandrov NITI for representing 3D effects in the downcomer and lower plenum during asymmetrical loop operation. The CFD module uses Cartesian grid method with cut cell approach. The paper presents a numerical algorithm for coupling 1D and 3D thermal- hydraulic modules in the KORSAR code. The combined pressure field is calculated by the multigrid method. The performance efficiency of the algorithm for coupling 1D and 3D modules was demonstrated by solving the benchmark problem of mixing cold and hot flows in a T-junction.

  15. Thermal-hydraulic tests of a recirculation cooling installation for the Rostov nuclear power station

    Science.gov (United States)

    Balunov, B. F.; Balashov, V. A.; Il'in, V. A.; Krayushnikov, V. V.; Lychakov, V. D.; Meshalkin, V. V.; Ustinov, A. N.; Shcheglov, A. A.

    2013-09-01

    Results obtained from thermal-hydraulic tests of the recirculation cooling installation used as part of the air cooling system under the containments of the Rostov nuclear power station Units 3 and 4 are presented. The operating modes of the installation during normal operation (air cooling on the surface of finned tubes), under the conditions of anticipated operational occurrences (air cooling and steam condensation from a steam-air mixture), and during an accident (condensation of pure steam) are considered. Agreement is obtained between the results of tests and calculations carried out according to the recommendations given in the relevant regulatory documents. A procedure of carrying out thermal calculation for the case of steam condensation from a steam-air mixture on the surface of fins is proposed. The possibility of efficient use of the recirculation cooling installation in the system for reducing emergency pressure under the containment of a nuclear power station is demonstrated.

  16. Coupled neutronics/thermal-hydraulics and safety characteristics of liquid-fueled molten salt reactors

    Energy Technology Data Exchange (ETDEWEB)

    Qiu, Suizheng; Zhang, Dalin; Liu, Minghao; Liu, Limin; Xu, Rongshuan; Gong, Cheng; Su, Guanghui [Xi' an Jiaotong Univ. (China). State Key Laboratory of Multiphase Flow in Power Engineering

    2016-05-15

    Molten salt reactor (MSR) as one candidate of the Generation IV advanced nuclear power systems is attracted more attention in China due to its top ranked fuel cycle and thorium utilization. The MSRs are characterized by using liquid-fuel, which offers complicated coupling problem of neutronics and thermal hydraulics. In this paper, the fundamental model and numerical method are established to calculate and analyze the safety characteristics for liquid-fuel MSRs. The theories and methodologies are applied to the MOSART concept. The liquid-fuel flow effects on neutronics, reactivity coefficients and three operation parameters' influences at steady state are obtained, which provide the basic information for safety analysis. The unprotected loss of flow transient is calculated, the results of which shows the inherent safety characteristics of MOSART due to its strong negative reactivity feedbacks.

  17. 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.

  18. 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.

  19. 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.

  20. 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.

  1. Reflooding and boil-off experiments in a VVER-440 like rod bundle and analyses with the CATHARE code

    Energy Technology Data Exchange (ETDEWEB)

    Korteniemi, V.; Haapalehto, T. [Lappeenranta Univ. of Technology (Finland); Puustinen, M. [VTT Energy, Lappeenranta (Finland)

    1995-09-01

    Several experiments were performed with the VEERA facility to simulate reflooding and boil-off phenomena in a VVER-440 like rod bundle. The objective of these experiments was to get experience of a full-scale bundle behavior and to create a database for verification of VVER type core models used with modern thermal-hydraulic codes. The VEERA facility used in the experiments is a scaled-down model of the Russian VVER-440 type pressurized water reactors used in Loviisa, Finland. The test section of the facility consists of one full-scale copy of a VVER-440 reactor rod bundle with 126 full-length electrically heated rod simulators. Bottom and top-down reflooding, different modes of emergency core cooling (ECC) injection and the effect of heating power on the heat-up of the rods was studied. In this paper the results of calculations simulating two reflood and one boil-off experiment with the French CATHARE2 thermal-hydraulic code are also presented. Especially the performance of the recently implemented top-down reflood model of the code was studied.

  2. 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...

  3. Development of a steady thermal-hydraulic analysis code for the China Advanced Research Reactor

    Institute of Scientific and Technical Information of China (English)

    TIAN Wenxi; QIU Suizheng; GUO Yun; SU Guanghui; JIA Dounan; LIU Tiancai; ZHANG Jianwei

    2007-01-01

    A multi-channel model steady-state thermalhydraulic analysis code was developed for the China Advanced Research Reactor (CARR). By simulating the whole reactor core, the detailed mass flow distribution in the core was obtained. The result shows that structure size plays the most important role in mass flow distribution, and the influence of core power could be neglected under singlephase flow. The temperature field of the fuel element under unsymmetrical cooling condition was also obtained, which is necessary for further study such as stress analysis, etc. Of the fuel element. At the same time, considering the hot channel effect including engineering factor and nuclear factor, calculation of the mean and hot channel was carried out and it is proved that all thermal-hydraulic parameters satisfy the "Safety design regulation of CARR".

  4. 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)

  5. Study on Thermal-Hydraulic Behavior of an Integral Type Reactor under Heaving Condition

    Directory of Open Access Journals (Sweden)

    Beibei Feng

    2014-01-01

    Full Text Available A self-developed program was used to study the thermal-hydraulic behavior of an integral type reactor under heaving condition. Comparison of calculated results with the data of experiments performed on a natural circulation loop designed with reference to an integral type reactor of Tsinghua University in inclination, heaving, and rolling motions was carried out. Characteristics of natural circulation in heaving motion and effect of motion parameters on natural circulation were investigated. Results indicated that: (1 long-period heaving motion would lead to more significant influence than inclination and rolling motion; (2 it was an alternating force field which consisted of gravity and an additional force that decided the flow temperature and density difference of natural circulation; (3 effect of strength k and cycle T of heaving motion on flow fluctuation of natural circulation and condensate depression of heating section outlet was performed.

  6. 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.

  7. 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.

  8. 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.

  9. Integral Circulation Experiment: Thermal-hydraulic simulator of a heavy liquid metal reactor

    Energy Technology Data Exchange (ETDEWEB)

    Tarantino, M., E-mail: mariano.tarantino@enea.it [ENEA UTIS, C.R. Brasimone, 40032 Camugnano, BO (Italy); Agostini, P.; Benamati, G.; Coccoluto, G.; Gaggini, P.; Labanti, V.; Venturi, G. [ENEA UTIS, C.R. Brasimone, 40032 Camugnano, BO (Italy); Class, A.; Liftin, K. [KIT, Forschungszentrum Karlsruhe, IKET, P.O. Box 3640, D-76021 Karlsruhe (Germany); Forgione, N. [Universita di Pisa, DIMNP, Via Diotisalvi 2, 56126 Pisa (Italy); Moreau, V. [CRS4, Loc. Piscina Manna, Edificio 1, 09010 Pula (Italy)

    2011-08-31

    In the frame of the IP-EUROTRANS (6th Framework Program EU), domain DEMETRA, ENEA was involved in the Work Package 4.5 'Large Scale Integral Test', devoted to characterize a relevant portion of a sub-critical ADS reactor block (core, internals, heat exchanger, cladding for fuel elements) in steady state, transient and accidental conditions. More in details ENEA assumed the commitment to perform an integral experiment aiming to reproduce the primary flow path of the 'European Transmutation Demonstrator (ETD)' pool-type nuclear reactor, cooled by Lead Bismuth Eutectics (LBE). This experimental activity, called 'Integral Circulation Experiment (ICE)', has been implemented merging the efforts of several research institutes, among which, besides ENEA, FZK, CRS4 and University of Pisa, allowing to design an appropriate test section to be installed in the CIRCE facility. The goal of the experiments is therefore to demonstrate the technological feasibility of a heavy liquid metal (HLM) nuclear system pool-type in a relevant scale (1 MW), investigating the related thermal-hydraulic behaviour (heat source and heat exchanger coupling, primary system and downcomer coupling, gas trapping into the main stream, thermal stratification in the pool, forced and mixed convection in rod bundle) under both steady state and transient conditions. Moreover the preliminary as well as the planned experiments aims to address performance and reliability tests of some prototypical components, such as heat source, heat exchanger, chemistry control system. The paper reports a detailed description of the experiment, the design performed for the test section and its main components as well as the preliminary experimental results carried out in the first experimental campaign run on the CIRCE pool, which consists of a full power steady state test. The preliminary experimental results carried out have demonstrate the proper design of the test section trough the

  10. Integral Circulation Experiment: Thermal-hydraulic simulator of a heavy liquid metal reactor

    Science.gov (United States)

    Tarantino, M.; Agostini, P.; Benamati, G.; Coccoluto, G.; Gaggini, P.; Labanti, V.; Venturi, G.; Class, A.; Liftin, K.; Forgione, N.; Moreau, V.

    2011-08-01

    In the frame of the IP-EUROTRANS (6th Framework Program EU), domain DEMETRA, ENEA was involved in the Work Package 4.5 " Large Scale Integral Test", devoted to characterize a relevant portion of a sub-critical ADS reactor block (core, internals, heat exchanger, cladding for fuel elements) in steady state, transient and accidental conditions. More in details ENEA assumed the commitment to perform an integral experiment aiming to reproduce the primary flow path of the " European Transmutation Demonstrator (ETD)" pool-type nuclear reactor, cooled by Lead Bismuth Eutectics (LBE). This experimental activity, called " Integral Circulation Experiment (ICE)", has been implemented merging the efforts of several research institutes, among which, besides ENEA, FZK, CRS4 and University of Pisa, allowing to design an appropriate test section to be installed in the CIRCE facility. The goal of the experiments is therefore to demonstrate the technological feasibility of a heavy liquid metal (HLM) nuclear system pool-type in a relevant scale (1 MW), investigating the related thermal-hydraulic behaviour (heat source and heat exchanger coupling, primary system and downcomer coupling, gas trapping into the main stream, thermal stratification in the pool, forced and mixed convection in rod bundle) under both steady state and transient conditions. Moreover the preliminary as well as the planned experiments aims to address performance and reliability tests of some prototypical components, such as heat source, heat exchanger, chemistry control system. The paper reports a detailed description of the experiment, the design performed for the test section and its main components as well as the preliminary experimental results carried out in the first experimental campaign run on the CIRCE pool, which consists of a full power steady state test. The preliminary experimental results carried out have demonstrate the proper design of the test section trough the experiment goals as well as the HLM

  11. 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

  12. THE THERMAL-HYDRAULICS ANALYSIS ON RADIAL AND AXIAL POWER FLUCTUATION FOR AP1000 REACTOR

    Directory of Open Access Journals (Sweden)

    Muh. Darwis Isnaini

    2015-06-01

    Full Text Available ABSTRACT THE THERMAL-HYDRAULICS ANALYSIS ON RADIAL AND AXIAL POWER FLUCTUATION FOR AP1000 REACTOR. The reduction of fissile material during reactor operation affects reactivity reduction. Therefore, in order to keep the reactor operating at fixed power, it must be compensated by slowly withdrawing the control-rod up. However, it will change the shape of the horizontal/axial power distribution and safety margin as well. The research carries out the calculations of the core thermal-hydraulics to determine the effect of the fluctuations of the power distribution on the thermal-hydraulic AP1000’s parameters and study their impacts on the safety margin. The calculation is done using the COBRA-EN code and the result shows that the maximum heat flux at the Beginning of Cycle (BOC is 1624.02 kW/m2. This heat flux will then decrease by 22.75% at the Middle of Cycle (MOC and by 0.29% at the End of Cycle (EOC. The peak fuel centerline temperature at the BOC, MOC and EOC, are 1608.15°C, 1232.15°C, and 1301.75°C, respectively. These temperature differences are caused by the heat flux effects on sub-cooled boiling regions in the cladding surface. Moreover, the value of MDNBRs at the MOC and EOC are 3.23 and 3.00, which are higher than the MDNBR at the BOC of 2.49. It could be concluded that the operating cycle of the AP1000 reactor should be operated in the MOC and the EOC, which will be more safely than be operated in the BOC. Keywords: Core thermal-hydraulics, AP1000, fluctuation of power distribution, COBRA-EN.   ABSTRAK ANALISIS TERMOHIDRAULIKA PADA FLUKTUASI DAYA AXIAL DAN RADIAL UNTUK REAKTOR AP1000. Berkurangnya material fisil selama operasi reaktor, mengakibatkan reaktivitas berkurang. Oleh karena itu, agar reaktor tetap beroperasi pada daya yang tetap, maka harus dikompensasi dengan menarik batang kendali ke atas sedikit demi sedikit. Akan tetapi, hal ini akan berakibat pada berubahnya bentuk distribusi daya ke arah horisontal/aksial dan

  13. 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)

  14. Icare/Cathare coupling: three-dimensional thermal hydraulics of severe LWR accidents

    Energy Technology Data Exchange (ETDEWEB)

    Guillard, V.; Fichot, F. [CEA Fontenay aux Roses, Inst. de Protection et de Surete Nucleaire, Dept. de Recherches en Securite, DRS, 92 (France); Boudier, P.; Parent, M. [CEA Grenoble, Dir. des Reacteurs Nucleaires, DRN, 38 (France); Roser, R. [Communication et Systemes Systemes d' Information, CS SI, 38 - Fontaine (France)

    2001-07-01

    In the phenomenology of severe LWR accidents considered in safety studies, the accidental sequences can be divided into three phases: the initial phase, where no severe damage of fuel or control rods and structures occurs; the early core degradation phase, where limited material melting and relocation takes place; and the late core degradation phase during which substantial material relocation happens, molten pools and debris beds can form and corium may fall into the lower plenum and, in case of vessel failure, come into the containment. The CATHARE2 code is a system code which has been developed by CEA for IPSN, EDF and FRAMATOME to describe the thermal-hydraulics behavior of a whole PWR circuit during the first of these three phases, with a core degradation model limited to clad rupture. The ICARE2 code, developed by IPSN, allows the complete description of early and late core degradation phases, with a thermal-hydraulics model limited to the vessel, initial and boundary conditions being provided by a system code. The aim of this paper is to present the main features of the new version of the coupling, ICARE/CATHARE V2. First, the general characteristics of ICARE2 V3mod1 and CATHARE2 V1.5 standard codes, dealing with physical models and numerical aspects, are described. Second, the technical features of the coupling between the two codes are detailed. At last, some results of ICARE/CATHARE V2 calculations are presented which demonstrate the ability of the code to simulate a severe accident in a PWR and notably to describe multi-dimensional effects occurring in the core during the LOCA and degradation phases. (authors)

  15. 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.

  16. 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

  17. Relevant thermal-hydraulic aspects in the design of the RRR (Replacement Research Reactor)

    Energy Technology Data Exchange (ETDEWEB)

    Doval, Alicia S.; Mazufri, Claudio M. [INVAP SE, Bariloche (Argentina)

    2002-07-01

    A description of the main thermal-hydraulic features and challenges of the Replacement Research Reactor, for the Australian Nuclear Science and Technology Organization (ANSTO), is presented. Different hydraulic and thermal-hydraulic aspects are considered, core cooling during full power operation and the way it affects the design, design criteria, engineered safety features and computational tools, amongst others. A special section is devoted to the thermal-hydraulic aspects inside the reflector tank, as well as the cooling of irradiation facilities, particularly, the Molybdenum production facility. (author)

  18. TRAC-PF1/MOD1 thermal-hydraulic predictions of JAERI Slab Core Test Facility gravity-feed tests

    Energy Technology Data Exchange (ETDEWEB)

    Gilbert, J.S.; Lin, J.C.

    1985-12-01

    The Transient Reactor Analysis Code, TRAC-PF1/MOD1, was used to analyze the Slab Core Test Facility gravity-feed tests (Runs 604, 605, 611, and 613) performed by the Japan Atomic Energy Research Institute. The objectives of the TRAC analysis are to compare the TRAC predictions with the test results and to assess the TRAC capability for simulating the core thermal-hydraulic behavior during the reflood phase of a large loss-of-coolant accident. In general, the TRAC-calculated results agree well with the data.

  19. Verification of the computer code KORSAR taking into account the effect of nondensables on thermal-hydraulic processes

    Science.gov (United States)

    Gudoshnikov, A. N.; Migrov, Yu. A.

    2008-11-01

    Calculations to verify the Russian computer code KORSAR were carried out for the B4.1 experimental operating conditions, in which nitrogen was supplied to the reactor coolant (primary) circuit of a reactor plant model, and which were simulated at the PKL III integral test facility. It is shown that dissolution of gases in coolant has an essential effect on the thermal-hydraulic processes during long-term passive removal of heat from the primary to secondary coolant circuit of the reactor plant model under the conditions of natural circulation.

  20. Thermal-hydraulic experiments for the PCHE type steam generator

    Energy Technology Data Exchange (ETDEWEB)

    Shin, C. W.; No, H. C. [KAIST, Daejeon (Korea, Republic of)

    2015-05-15

    Printed circuit heat exchanger (PCHE) manufactured by HEATRIC is a compact type of the mini-channel heat exchanger. The PCHE is manufactured by diffusion bonding of the chemically-etched plates, and has high heat transfer rate due to a large surface. Therefore, the size of heat exchanger can be reduced by 1/5 - 1/6 and PCHE can be operated under high pressure, high temperature and multi-phase flow. Under such merits, it is used as heat exchanger with various purposes of gas cycle and water cycle. Recently, it is newly suggested as an application of a steam generator. IRIS of MIT and FASES of KAIST conceptually adopted PCHE as a steam generator. When using boiling condition of micro-channel, flow instability is one of the critical issues. Instability may cause unstable mass flow rate, sudden temperature change and system control failure. However instability tests of micro channels using water are very limited because the previous studies were focused on a single tube or other fluid instead of water. In KAIST, we construct the test facility to study the thermal hydraulics and fluid dynamics of the heat exchanger, especially occurrence of instability. By inducing the pressure drop of inlet water, amplitude of oscillation declined by 90%. Finally, the throttling effect was experimentally confirmed that PCHE could be utilized as a steam generator.

  1. 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)

  2. Commissioning of the ATLAS thermal-hydraulic integral test facility

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Yeon-Sik [Thermal Hydraulics Safety Research Division, Korea Atomic Energy Research Institute, 1045 Daedeokdaero, Yuseong-gu, Daejeon 305-353 (Korea, Republic of)], E-mail: yskim3@kaeri.re.kr; Choi, Ki-Yong; Park, Hyeon-Sik; Cho, Seok; Kim, Bok-Deug; Choi, Nam-Hyeon; Baek, Won-Pil [Thermal Hydraulics Safety Research Division, Korea Atomic Energy Research Institute, 1045 Daedeokdaero, Yuseong-gu, Daejeon 305-353 (Korea, Republic of)

    2008-10-15

    KAERI recently constructed a new thermal-hydraulic integral test facility for advanced pressurized water reactors (PWRs) - ATLAS. The ATLAS facility has the following characteristics: (a) 1/2-height and length, 1/288-volume, and full pressure simulation of APR1400, (b) maintaining a geometrical similarity with APR1400 including 2(hot legs) x 4(cold legs) reactor coolant loops, direct vessel injection (DVI) of emergency core cooling water, integrated annular downcomer, etc., (c) incorporation of specific design characteristics of OPR1000 such as cold leg injection and low-pressure safety injection pumps, (d) maximum 10% of the scaled nominal core power. The ATLAS will mainly be used to simulate various accident and transient scenarios for evolutionary PWRs, OPR1000 and APR1400: the simulation capability of broad scenarios including the reflood phase of a large-break loss-of-coolant accident (LOCA), small-break LOCA scenarios including DVI line breaks, a steam generator tube rupture, a main steam line break, a feed line break, a mid-loop operation, etc. The ATLAS is now in operation after an extensive series of commissioning tests in 2006.

  3. Thermal hydraulics analysis of the Advanced High Temperature Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Wang, Dean, E-mail: Dean_Wang@uml.edu [University of Massachusetts Lowell, One University Avenue, Lowell, MA 01854 (United States); Yoder, Graydon L.; Pointer, David W.; Holcomb, David E. [Oak Ridge National Laboratory, 1 Bethel Valley RD #6167, Oak Ridge, TN 37831 (United States)

    2015-12-01

    Highlights: • The TRACE AHTR model was developed and used to define and size the DRACS and the PHX. • A LOFF transient was simulated to evaluate the reactor performance during the transient. • Some recommendations for modifying FHR reactor system component designs are discussed. - Abstract: The Advanced High Temperature Reactor (AHTR) is a liquid salt-cooled nuclear reactor design concept, featuring low-pressure molten fluoride salt coolant, a carbon composite fuel form with embedded coated particle fuel, passively triggered negative reactivity insertion mechanisms, and fully passive decay heat rejection. This paper describes an AHTR system model developed using the Nuclear Regulatory Commission (NRC) thermal hydraulic transient code TRAC/RELAP Advanced Computational Engine (TRACE). The TRACE model includes all of the primary components: the core, downcomer, hot legs, cold legs, pumps, direct reactor auxiliary cooling system (DRACS), the primary heat exchangers (PHXs), etc. The TRACE model was used to help define and size systems such as the DRACS and the PHX. A loss of flow transient was also simulated to evaluate the performance of the reactor during an anticipated transient event. Some initial recommendations for modifying system component designs are also discussed. The TRACE model will be used as the basis for developing more detailed designs and ultimately will be used to perform transient safety analysis for the reactor.

  4. 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.

  5. Assessment on 900–1300 MWe PWRs of the ASTEC-based simulation tool of SGTR thermal-hydraulics for the IRSN Emergency Technical Centre

    Energy Technology Data Exchange (ETDEWEB)

    Foucher, L., E-mail: laurent.foucher@irsn.fr [Institut de Radioprotection et de Sûreté Nucléaire (IRSN), PSN-RES/SAG, Cadarache, Saint-Paul-lez-Durance 13115 (France); Cousin, F.; Fleurot, J. [Institut de Radioprotection et de Sûreté Nucléaire (IRSN), PSN-RES/SAG, Cadarache, Saint-Paul-lez-Durance 13115 (France); Brethes, S. [Institut de Radioprotection et de Sûreté Nucléaire (IRSN), PRP-CRI/SESUC, Cadarache, Saint-Paul-lez-Durance 13115 (France)

    2014-06-01

    In the event of an accident occurring in a nuclear power plant (NPP), being able to predict the amount of released radioactive substances in the environment is of prime importance. Depending on the severity of the accident, it can be necessary to quickly and efficiently protect the population and the surrounding environment from the associated radiological consequences. In France, the IRSN Emergency Technical Centre provides a technical support in decision making in case of a nuclear accident. The main objectives are to evaluate and predict the plant behaviour and radioactive releases during the accident. Different types of complementary tools are used: expert assessments, pre-calculated databases, simulation tools, etc. In the case of Steam Generator Tube Rupture (SGTR) accidents that may lead to significant radioactive releases to the atmosphere through the steam generator relief valves, IRSN is currently improving the simulation tools for diagnosis in crisis management. The objective is to adapt the thermal-hydraulic and FP behaviour modules of the severe accident integral code ASTEC V2.0, jointly developed by IRSN and its German counterpart GRS, to crisis management requirements. These requirements impose a fast running, highly reliable (accurate physical results), flexible and simple tool. This paper summarizes the results of the benchmarks between the ASTEC V2.0 thermal-hydraulic module and the CATHARE 2 (V2.5) French reference thermal-hydraulics code on several SGTR scenarios both for PWR 900 and 1300 MWe, with a particular emphasis on the computational time and physical models assessment. The overall agreement between both codes is good on the primary and secondary circuit thermal-hydraulic parameters. Moreover, the reliability and fast computational time of the thermal-hydraulic module of ASTEC V2.0 code appeared very satisfactory and in accordance with the requirements of an emergency tool.

  6. 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

  7. 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...

  8. 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.

  9. 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.

  10. Thermal-hydraulic Optimization of Water-cooled Center Conductor Post for Spherical Tokamaks Reactor

    Institute of Scientific and Technical Information of China (English)

    柯严; 吴宜灿; 黄群英; 郑善良

    2002-01-01

    This paper proposes a conceptual structure of segmental water-cooled Center Con ductor Post (CCP) to be flexible in installment and replacement. Thermal-hydraulic optimization and sensitivity analysis of key parameters are performed based on a reference fusion transmutation system with 100 MW fusion power. Numerical simulation by using a commercial code PHOEN]CS has been carried out to be close to the thermal-hydraulic analytical results of the CCP mid-part.

  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. 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.).

  13. 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.

  14. Thermal Hydraulics of the Very High Temperature Gas Cooled Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Chang Oh; Eung Kim; Richard Schultz; Mike Patterson; Davie Petti

    2009-10-01

    The U.S Department of Energy (DOE) is conducting research on the Very High Temperature Reactor (VHTR) design concept for the Next Generation Nuclear Plant (NGNP) Project. The reactor design will be a graphite moderated, thermal neutron spectrum reactor that will produce electricity and hydrogen in a highly efficient manner. The NGNP reactor core will be either a prismatic graphite block type core or a pebble bed core. The NGNP will use very high-burnup, low-enriched uranium, TRISO-coated fuel, and have a projected plant design service life of 60 years. The VHTR concept is considered to be the nearest-term reactor design that has the capability to efficiently produce hydrogen. The plant size, reactor thermal power, and core configuration will ensure passive decay heat removal without fuel damage or radioactive material releases during reactor core-accidents. The objectives of the NGNP Project are to: Demonstrate a full-scale prototype VHTR that is commercially licensed by the U.S. Nuclear Regulatory Commission, and Demonstrate safe and economical nuclear-assisted production of hydrogen and electricity. The DOE laboratories, led by the INL, perform research and development (R&D) that will be critical to the success of the NGNP, primarily in the areas of: • High temperature gas reactor fuels behavior • High temperature materials qualification • Design methods development and validation • Hydrogen production technologies • Energy conversion. This paper presents current R&D work that addresses fundamental thermal hydraulics issues that are relevant to a variety of possible NGNP designs.

  15. Analysis of Thermal-Hydraulic Gravity/ Buoyancy Effects in the Testing of the ITER Poloidal Field Full Size Joint Sample (PF-FSJS)

    Science.gov (United States)

    Zanino, R.; Bruzzone, P.; Ciazynski, D.; Ciotti, M.; Gislon, P.; Nicollet, S.; Savoldi Richard, L.

    2004-06-01

    The PF-FSJS is a full-size joint sample, based on the NbTi dual-channel cable-in-conduit conductor (CICC) design currently foreseen for the International Thermonuclear Experimental Reactor (ITER) Poloidal Field coil system. It was tested during the summer of 2002 in the Sultan facility of CRPP at a background peak magnetic field of typically 6 T. It includes about 3 m of two jointed conductor sections, using different strands but with identical layout. The sample was cooled by supercritical helium at nominal 4.5-5.0 K and 0.9-1.0 MPa, in forced convection from the top to the bottom of the vertical configuration. A pulsed coil was used to test AC losses in the two legs resulting, above a certain input power threshold, in bundle helium backflow from the heated region. Here we study the thermal-hydraulics of the phenomenon with the M&M code, with particular emphasis on the effects of buoyancy on the helium dynamics, as well as on the thermal-hydraulic coupling between the wrapped bundles of strands in the annular cable region and the central cooling channel. Both issues are ITER relevant, as they affect the more general question of the heat removal capability of the helium in this type of conductors.

  16. 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)

  17. A novel methodology of internal assessment of uncertainty for three-dimensional neutronics/thermal-hydraulics system codes

    Energy Technology Data Exchange (ETDEWEB)

    Petruzzi, A.; Auria, F. [Pisa Universita, Dipartimento di Ingegneria Meccanica, Nucleare e della Produzione, Pisa (Italy); Ivanov, K. [Pennsylvania State University, Department of Mechanical and Nuclear Engineering, PA (Italy)

    2003-07-01

    The paper stresses how the internal assessment of uncertainty is a desirable capability for thermal-hydraulic system codes. This consists of the possibility of obtaining proper uncertainty bands each time a nuclear plant transient scenario is calculated. A methodology suitable for introducing such a capability into a system code is discussed. At the basis of the derivation of the code with (the capability of) internal assessment of uncertainty (CIAU), there is the uncertainty methodology based on the accuracy extrapolation (UMAE), previously proposed by the University of Pisa, although other uncertainty methodologies can be used for the same purpose. The idea of the CIAU is the identification and the characterization of standard plant statuses and the association of uncertainty to each status. One hypercube and one time interval identify the plant status. Quantity and time uncertainties are combined for each plant status. The RELAP5/MOD3.2 system code has been used inside the CIAU to show the applicability of the proposed method. The derivation of the methodology is discussed, and reference results of pressurized water reactor plant transients are shown bounded by the CIAU calculated uncertainty bands. Recently, a new activity has been started with the aim to extend the CIAU to the 3D neutronics/thermal-hydraulics coupled codes. (authors)

  18. Computational simulation of thermal hydraulic processes in the model LMFBR fuel assembly

    Science.gov (United States)

    Bayaskhalanov, M. V.; Merinov, I. G.; Korsun, A. S.; Vlasov, M. N.

    2017-01-01

    The aim of this study was to verify a developed software module on the experimental fuel assembly with partial blockage of the flow section. The developed software module for simulation of thermal hydraulic processes in liquid metal coolant is based on theory of anisotropic porous media with specially developed integral turbulence model for coefficients determination. The finite element method is used for numerical solution. Experimental data for hexahedral assembly with electrically heated smooth cylindrical rods cooled by liquid sodium are considered. The results of calculation obtained with developed software module for a case of corner blockade are presented. The calculated distribution of coolant velocities showed the presence of the vortex flow behind the blockade. Features vortex region are in a good quantitative and qualitative agreement with experimental data. This demonstrates the efficiency of the hydrodynamic unit for developed software module. But obtained radial coolant temperature profiles differ significantly from the experimental in the vortex flow region. The possible reasons for this discrepancy were analyzed.

  19. Thermal Hydraulic Analysis of a Passive Residual Heat Removal System for an Integral Pressurized Water Reactor

    Directory of Open Access Journals (Sweden)

    Junli Gou

    2009-01-01

    Full Text Available A theoretical investigation on the thermal hydraulic characteristics of a new type of passive residual heat removal system (PRHRS, which is connected to the reactor coolant system via the secondary side of the steam generator, for an integral pressurized water reactor is presented in this paper. Three-interknited natural circulation loops are adopted by this PRHRS to remove the residual heat of the reactor core after a reactor trip. Based on the one-dimensional model and a simulation code (SCPRHRS, the transient behaviors of the PRHRS as well as the effects of the height difference between the steam generator and the heat exchanger and the heat transfer area of the heat exchanger are studied in detail. Through the calculation analysis, it is found that the calculated parameter variation trends are reasonable. The higher height difference between the steam generator and the residual heat exchanger and the larger heat transfer area of the residual heat exchanger are favorable to the passive residual heat removal system.

  20. Multidimensional Thermal-Hydraulic Analysis for Decay Heat Exchanger of PGSFR

    Energy Technology Data Exchange (ETDEWEB)

    Hong, Jonggan; Yoon, Jung; Kim, Dehee; Lee, Tae-Ho [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-05-15

    The decay heat exchanger (DHX) of PGSFR is a shell-and-tube type counter-current flow sodium heat exchanger, and each unit is designed for the rated thermal power of 1.0 MWt, which is corresponding to the nominal design capacity of a single passive decay heat removal system (PDHRS) and active decay heat removal system (ADHRS) loops. The DHX unit is fully immersed in the cold sodium pool region and removes the system heat load sufficiently and reliably during the temperature transient. In this work, a multidimensional thermal-hydraulic analysis for the DHX was carried out numerically and the numerical results were compared with the calculated results of the 1-D DHX design code to verify the reliability of the design code. In addition, an influence of the cold pool sodium which flows into the shell-side of the DHX through the shell outlet was evaluated. The SHXSA code was conservative in calculating the pressure drop of the shell-side which is our major concern in designing the natural circulation of the decay heat removal system. It was revealed that the buffer region is needed to reduce the thermal stress in the lower tubesheet by the inflow of the cold pool sodium.

  1. 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.

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

    Science.gov (United States)

    Borsuk, Grzegorz; Dobrowolski, Bolesław; Nowosielski, Grzegorz; Wydrych, Jacek; Duda, Jerzy

    2016-03-01

    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. Evaluation of the RELAP4/MOD6 thermal-hydraulic code. [PWR

    Energy Technology Data Exchange (ETDEWEB)

    Haigh, W.S.; Margolis, S.G.; Rice, R.E.

    1978-01-01

    The NRC RELAP4/MOD6 computer code was recently released to the public for use in thermal-hydraulic analysis. This code has a unique new capability permitting analysis of both the blowdown and reflood portions of a postulated pressurized water reactor (PWR) loss-of-coolant accident (LOCA). A principal code evaluation objective is to assess the accuracy of the code for computing LOCA behavior over a wide range of system sizes and scaling concepts. The scales of interest include all LOCA experiments and will ultimately encompass full-sized PWR systems for which no experiments or data are available. Quantitative assessment of the accuracy of the code when it is applied to large PWR systems is still in the future. With RELAP4/MOD6, however, a technique has been demonstrated for using results derived from small-scale blowdown and reflood experiments to predict the accuracy of calculations for similar experiments of significantly different scale or component size. This demonstration is considered a first step in establishing confidence levels for the accuracy of calculations of a postulated LOCA.

  4. 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)

  5. 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)

  6. 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)

  7. 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.

  8. Thermal hydraulic analysis of reactivity accidents in MTR research reactors using RELAP5

    Energy Technology Data Exchange (ETDEWEB)

    El-Sahlamy, N.; Khedr, A. [Nuclear and Radiological Regulatory Authority (NRRA), Cairo (Egypt); D' Auria, F.D. [Pisa Univ. (Italy). Facolta di Ingegneria

    2015-12-15

    The present paper comes in the line with the international approach which use the best estimate codes, instead of conservative codes, to get more realistic prediction of system behavior under off-normal reactor conditions. The aim of the current work is to apply this approach using the thermal-hydraulic system code RELAP5/Mod3.3 in a reassessment of safety of the IAEA benchmark 10 MW Research Reactor. The assessment is performed for both slow and fast reactivity insertion transients at initial power of 1.0 W. The reactor power is calculated using the RELA5 point kinetic model. The reactivity feedback terms are considered in two steps. In the first step the feedback from changes in water density and fuel temperature (Doppler effects) are considered. In the second step the feedback from the water temperature changes is added. The results from the first step are compared with that published in IAEA-TECDOC-643 benchmarks. The comparison shows that RELAP5 over predicts the peak power and consequently the fuel, clad and coolant temperatures in case of fast reactivity insertion. The results from the second step show unjustified values for reactor power. Therefore, the model of reactivity feedback from water temperature changes in the RELAP5 code may have to be reviewed.

  9. TRAC analysis of upper plenum thermal-hydraulic phenomena in the slab core test facility

    Energy Technology Data Exchange (ETDEWEB)

    Shire, P.; Boyack, B.

    1986-01-01

    The Transient Reactor Analysis Code, TRAC-PF1/MOD1, was used to analyze an upper-plenum model of the Slab Core Test Facility (SCTF). The SCTF is a two-dimensional thermal-hydraulic model of a pressurized water reactor used for core-reflood simulations by the Japan Atomic Energy Research Institute. The purpose of this study was to evaluate the effects of code input-model refinements on the comparison between TRAC calculations and test data. Of particular interest were the comparisons of upper-plenum liquid levels and of the distributions of liquid radially across the upper-plenum. The upper-plenum region was selected for study. The test data indicated that the liquid level responded to the onset of emergency core cooling (ECC) by rising in the upper-plenum immediately when injection occurred. However, the early TRAC results indicated no significant liquid level increase until approx.300 s after the injection. Test data also indicated a liquid gradient rising toward the hot-leg entrance, but none was observed with TRAC.

  10. 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.

  11. Steady state thermal hydraulic analysis of LMR core using COBRA-K code

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Eui Kwang; Kim, Young Gyun; Kim Young In; Kim Young Cheol

    1997-02-01

    A thermal hydraulics analysis code COBRA-K is being developed by the KAERI LMR core design technology development team. COBRA-K is a part of the integrated computation system for LMR core design and analysis, the K-CORE system. COBRA-K is supposed to predict the flow and temperature distributions in LMR core. COBRA-K is an extension of the previously published COBRA-IV-I code with several functional improvements. Specially COBRA-K has been improved to analyze single and multi-assembly, and whole-core in the transient condition. This report describes the overall features of COBRA-K and gives general input descriptions. The 19 pin assembly experimental data of ORNL were used to verify the accuracy of this code for the steady state analysis. The comparative results show good agreements between the calculated and the measured data. And COBRA-K can be used to predict flow and temperature distributions for the LMR core design. (author). 7 refs., 6 tabs., 13 figs.

  12. Comparative study of Thermal Hydraulic Analysis Codes for Pressurized Water Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Yang Hoon; Jang, Mi Suk; Han, Kee Soo [Nuclear Engineering Service and Solution Co. Ltd., Daejeon (Korea, Republic of)

    2015-05-15

    Various codes are used for the thermal hydraulic analysis of nuclear reactors. The use of some codes among these is limited by user and some codes are not even open to general person. Thus, the use of alternative code is considered for some analysis. In this study, simple thermal hydraulic behaviors are analyzed using three codes to show that alternative codes are possible for the analysis of nuclear reactors. We established three models of the simple u-tube manometer using three different codes. RELAP5 (Reactor Excursion and Leak Analysis Program), SPACE (Safety and Performance Analysis CodE for nuclear power Plants), GOTHIC (Generation of Thermal Hydraulic Information for Containments) are selected for this analysis. RELAP5 is widely used codes for the analysis of system behavior of PWRs. SPACE has been developed based on RELAP5 for the analysis of system behavior of PWRs and licensing of the code is in progress. And GOTHIC code also has been widely used for the analysis of thermal hydraulic behavior in the containment system. The internal behavior of u-tube manometer was analyzed by RELAP5, SPACE and GOTHIC codes. The general transient behavior was similar among 3 codes. However, the stabilized status of the transient period analyzed by REPAP5 was different from the other codes. It would be resulted from the different physical models used in the other codes, which is specialized for the multi-phase thermal hydraulic behavior analysis.

  13. 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.

  14. 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)

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

    Directory of Open Access Journals (Sweden)

    F. Reventós

    2008-01-01

    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.

  16. Numerical Study of Thermal Hydraulics for Secondary side of Steam Generator by CUPID/MARS Coupled Simulation

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Jae Ryong; Yoon, Han Young [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-10-15

    As a thermal-hydraulic behavior in the secondary side of steam generator such as two-phase boiling flow, flow-induce vibration of U-tubes is quite complicated, the importance to numerically investigate the flow behavior has been arisen. Recently, multi-scale analyses have been developed to take into account the primary side as well. In this study, the coupled CUPID and MARS code was used for the simulation of boiler side of the PWR steam generator. Calculation results are compared with the existing code quantitatively. Coupled CUPID/MARS code was applied for the simulation of the steam generator. The primary side of the steam generator and other RCS was simulated by MARS and the secondary side was calculated by CUPID with porous media approach.

  17. Thermal-hydraulic analysis of a heavy-water reactor moderator tank using the CUPID Code

    Energy Technology Data Exchange (ETDEWEB)

    Choi, Su Ryong; Jeong, Jae Jun [Pusan National Univ., Busan (Korea, Republic of); Kim, Hyoung Tae; Yoon, Han Young [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2014-05-15

    In this study, a preliminary analysis is performed for the CANDU moderator tank. The calculation results using the basic case input showed a unrealistic, thermal stratification in the upper region, which was caused by the lack of the momentum of the cooling water from the inlet nozzle. To increase the flow momentum from the inlet nozzle, the cross-section area of each inlet nozzle was reduced by half and, then, the calculation showed very realistic results. It is clear that the modeling of the inlet nozzle affects the calculation result significantly. Further studies are needed for a realistic and efficient simulation of the flow in the Calandria tank. When the core cooling system fails to remove the decay heat from the fuel channels during a loss of coolant accident (LOCA), the pressure tube (PT) could strain to contact its surrounding Calandria tube (CT), which leads to sustained CTs dry out, finally resulting in damages to nuclear fuel. This situation can occur when the degree of the subcooling of the moderator inside the Calandria vessel is insufficient. In this regard, to estimate the local subcooling of the moderator inside the Calandria vessel is very important. However, the local temperature is measured at the inlet and outlet of the vessel only. Therefore, we need to accurately predict the local temperature inside the Calandria vessel.In this study, the thermal-hydraulic analysis of the real-scale heavy-water reactor moderator is carried out using the CUPID code. The applicability of the CUPID code to the analysis of the flow in the Calandria vessel has been assessed in the previous studies.

  18. Characteristics and performance analysis report of the major thermal hydraulic components in the high temperature/high pressure thermal hydraulic test facility (VISTA)

    Energy Technology Data Exchange (ETDEWEB)

    Choi, Ki Yong; Park, Hyun Sik; Cho, Seok; Lee, Sung Jae; Song, Chul Hwa; Park, Chun Kyong; Chung, Moon Ki

    2003-12-01

    The VISTA (Experimental Verification by Integral Simulation of Transients and Accidents) is an experimental facility to verify the performance and safety issues of the SMART-P (Pilot plant of the System-integrated Modular Advanced Reactor). The basic design of the SMART-P has been completed by KAERI. The present report describes the characteristics and performance of the major thermal hydraulic components in the VISTA Facility.

  19. Simulation of the passive condensation cooling tank of the PASCAL test facility using the component thermal-hydraulic analysis code CUPID

    Energy Technology Data Exchange (ETDEWEB)

    Cho, H. K.; Lee, S. J.; Kang, K. H.; Yoon, H. Y. [Korea Atomic Energy Research Inst., 1045 Daeduk-daero, Daejeon (Korea, Republic of)

    2012-07-01

    For the analysis of transient two-phase flows in nuclear reactor components, a three-dimensional thermal hydraulics code, named CUPID, has been being developed. In the present study, the CUPID code was applied for the simulation of the PASCAL (PAFS Condensing Heat Removal Assessment Loop) test facility constructed with an aim of validating the cooling and operational performance of the PAFS (Passive Auxiliary Feedwater System). The PAFS is one of the advanced safety features adopted in the APR+ (Advanced Power Reactor +), which is intended to completely replace the conventional active auxiliary feedwater system. This paper presents the preliminary simulation results of the PASCAL facility performed with the CUPID code in order to verify its applicability to the thermal-hydraulic phenomena inside the system. A standalone calculation for the passive condensation cooling tank was performed by imposing a heat source boundary condition and the transient thermal-hydraulic behaviors inside the system, such as the water level, temperature and velocity, were qualitatively investigated. The simulation results verified that the natural circulation and boiling phenomena in the water pool can be well reproduced by the CUPID code. (authors)

  20. Simulation of the Passive Condensation Cooling Tank of the PASCAL Test Facility using the Component Thermal-hydraulic Analysis Code CUPID

    Energy Technology Data Exchange (ETDEWEB)

    Cho, Hyoung Kyu; Lee, Seung Jun; Yoon, Han Young [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2012-05-15

    The need for a multi-dimensional analysis of transient thermal hydraulic phenomena in a component of a nuclear reactor is increasing with the advanced design features, such as a direct vessel injection system, a gravity-driven safety injection system, and a passive cooling system. Motivated by this, the development of a new thermal-hydraulic analysis code, named CUPID, is in progress at KAERI (Korea Atomic Energy Research Institute). Its numerical solver and two-phase flow models have been verified against standard conceptual problems of single and two-phase flows and validated for thermal-hydraulic experiments in our previous studies. The simulation of the passive secondary cooling system, PAFS (Passive Auxiliary Feedwater System) has been considered as one of the practical applications of CUPID. In the present study, the PCCT (Passive Condensation Cooling Tank) of the PASCAL test facility was analyzed with CUPID prior to simulating the prototype PAFS system. The objectives of the PASCAL simulation were to validate physical models of CUPID and its applicability to the PAFS analysis. This paper presents the two-dimensional transient calculation results and the comparisons with the experimental data

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

    OpenAIRE

    2007-01-01

    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 togeth...

  2. Mitigation method of thermal transient stress by a total analysis of thermal hydraulic and structural phenomena

    Energy Technology Data Exchange (ETDEWEB)

    Kasahara, Naoto [Japan Nuclear Cycle Development Inst., Oarai, Ibaraki (Japan). Oarai Engineering Center; Jinbo, Masakazu [Toshiba Co., Tokyo (Japan); Hosogai, Hiromi [Joyo Industry Co., Ltd., Tokai, Ibaraki (Japan)

    2002-09-01

    This study proposes a mitigation method of thermal transient loads in fast reactor components by utilizing relationships among plant system parameters and resulting thermal stresses. Conventional design procedure against thermal transient loads has two independent steps: thermal hydraulic analysis to determine conservative thermal transient conditions considering variation of the system parameters and structural analysis to check structural integrity under given conditions. On the other hand, a total analysis procedure of thermal hydraulic and structural phenomena can grasp the relationship among system parameters and thermal stresses. It enables the mitigation of thermal transient loads by adjusting system parameters. (author)

  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. A review on the thermal hydraulic characteristics of the air-cooled heat exchangers in forced convection

    Indian Academy of Sciences (India)

    Ankur Kumar; Jyeshtharaj B Joshi; Arun K Nayak; Pallippattu K Vijayan

    2015-05-01

    In this paper, a review is presented on the experimental investigations and the numerical simulations performed to analyze the thermal-hydraulic performance of the air-cooled heat exchangers. The air-cooled heat exchangers mostly consist of the finned-tube bundles. The primary role of the extended surfaces (fins) is to provide more heat transfer area to enhance the rate of heat transfer on the air side. The secondary role of the fins is to generate vortices, which help in enhancing the mixing and the heat transfer coefficient. In this study, the annular and plate fins are considered, the annular fins are further divided into four categories: (1) plane annular fins, (2) serrated fins, (3) crimped spiral fins, (4) perforated fins, and similarly for the plate fins, the fin types are: (1) plain plate fins, (2) wavy plate fins, (3) plate fins with DWP, and (4) slit and strip fins. In Section 4, the performance of the various types of fins is presented with respect to the parameters: (1) Reynolds number, (2) fin pitch, (3) fin height, (4) fin thickness, (5) tube diameter, (6) tube pitch, (7) tube type, (8) number of tube rows, and (9) effect of dehumidifying conditions. In Section 5, the conclusions and the recommendations for the future work have been given.

  5. 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.

  6. Three dimensional neutronic/thermal-hydraulic coupled simulation of MSR in steady state condition

    Energy Technology Data Exchange (ETDEWEB)

    Zhou, Jianjun; Wang, Chenglong [School of Nuclear Science and Technology, Xi’an Jiaotong University, Xi’an, Shaanxi 710049 (China); An, Hongzhen [Nuclear and Radiation Safety Center, Ministry of Environmental Protection of the People' s Republic of China, Beijing 100082 (China); Zhang, Daling, E-mail: dlzhang@mail.xjtu.edu.cn [School of Nuclear Science and Technology, Xi’an Jiaotong University, Xi’an, Shaanxi 710049 (China); Qiu, Suizheng; Su, Guanghui; Tian, Wenxi; Wu, Yingwei [School of Nuclear Science and Technology, Xi’an Jiaotong University, Xi’an, Shaanxi 710049 (China)

    2014-02-15

    Highlights: • Developed a three dimensional coupled code for MSR. • Investigated the neutron distribution and thermal-hydraulic characters of the core under steady state condition. • Analyzed the influence of inlet temperature and inlet velocity to thermal-hydraulics characteristics of the reactor. - Abstract: MSR (molten salt reactor) uses liquid molten salt as the coolant and fuel solvent, making it the only liquid reactor among the six generation IV reactor types. As a liquid reactor the physical properties of the reactor are significantly influenced by the fuel salt flow therefore conventional analysis methods applied in solid fuel reactors are not applicable for this type of reactors. The present work developed a three dimensional neutronic/thermal-hydraulic coupled code and applied it to investigate the thermal-hydraulic characteristics of the core in steady state condition based on neutron diffusion theory and numerical heat transfer. The code consists of two group neutron diffusion equations for fast and thermal neutron fluxes and six group balance equations for delayed neutron precursors. The temperature distribution, neutron fluxes and delayed neutron precursors distribution of the core in steady state conditions was studied, and the result analyzed when inlet temperature and velocity were changed. From simulation it was found that the inlet temperature has little influence to neutron distribution however inlet velocity affects the delayed neutron distribution in steady state condition. The results provide some valuable information in design and research of this kind of reactor.

  7. Evaluation of thermal-hydraulic parameter uncertainties in a TRIGA research reactor

    Energy Technology Data Exchange (ETDEWEB)

    Mesquita, Amir Z.; Costa, Antonio C.L.; Ladeira, Luiz C.D.; Rezende, Hugo C., E-mail: amir@cdtn.br, E-mail: aclc@cdtn.br, E-mail: lcdl@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 [Comissao Nacional de Energia Nuclear (CNEN), Rio de Janeiro, RJ (Brazil)

    2015-07-01

    Experimental studies had been performed in the TRIGA Research Nuclear Reactor of CDTN/CNEN to find out the its thermal hydraulic parameters. Fuel to coolant heat transfer patterns must be evaluated as function of the reactor power in order to assess the thermal hydraulic performance of the core. The heat generated by nuclear fission in the reactor core is transferred from fuel elements to the cooling system through the fuel-cladding (gap) and the cladding to coolant interfaces. As the reactor core power increases the heat transfer regime from the fuel cladding to the coolant changes from single-phase natural convection to subcooled nucleate boiling. This paper presents the uncertainty analysis in the results of the thermal hydraulics experiments performed. The methodology used to evaluate the propagation of uncertainty in the results was done based on the pioneering article of Kline and McClintock, with the propagation of uncertainties based on the specification of uncertainties in various primary measurements. The uncertainty analysis on thermal hydraulics parameters of the CDTN TRIGA fuel element is determined, basically, by the uncertainty of the reactor's thermal power. (author)

  8. 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.

  9. Thermal-hydraulic assessment of concrete storage cubicle with horizontal 3013 canisters

    Energy Technology Data Exchange (ETDEWEB)

    HEARD, F.J.

    1999-04-08

    The FIDAP computer code was used to perform a series of analyses to assess the thermal-hydraulic performance characteristics of the concrete plutonium storage cubicles, as modified for the horizontal placement of 3013 canisters. Four separate models were developed ranging from a full height model of the storage cubicle to a very detailed standalone model of a horizontal 3013 canister.

  10. Development of thermal-hydraulic analysis methodology for multiple modules of water-cooled breeder blanket in fusion DEMO reactor

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Geon-Woo; Lee, Jeong-Hun [Department of Nuclear Engineering, Seoul National University 1 Gwanak-ro, Gwanak-gu, Seoul 151-744 (Korea, Republic of); Cho, Hyoung-Kyu, E-mail: chohk@snu.ac.kr [Department of Nuclear Engineering, Seoul National University 1 Gwanak-ro, Gwanak-gu, Seoul 151-744 (Korea, Republic of); Park, Goon-Cherl [Department of Nuclear Engineering, Seoul National University 1 Gwanak-ro, Gwanak-gu, Seoul 151-744 (Korea, Republic of); Im, Kihak [National Fusion Research Institute, 169-148, Yuseong-gu, Daejeon 305-806 (Korea, Republic of)

    2016-02-15

    Highlights: • A methodology to simulate the K-DEMO blanket system was proposed. • The results were compared with the CFD, to verify the prediction capability of MARS. • 46 Blankets in a single sector in K-DEMO were simulated using MARS-KS. • Supervisor program was devised to handle each blanket module individually. • The calculation results showed the flow rates, pressure drops, and temperatures. - Abstract: According to the conceptual design of the fusion DEMO reactor proposed by the National Fusion Research Institute of Korea, the water-cooled breeding blanket system incorporates a total of 736 blanket modules. The heat flux and neutron wall loading to each blanket module vary along their poloidal direction, and hence, thermal analysis for at least one blanket sector is required to confirm that the temperature limitations of the materials are satisfied in all the blanket modules. The present paper proposes a methodology of thermal analysis for multiple modules of the blanket system using a nuclear reactor thermal-hydraulic analysis code, MARS-KS. In order to overcome the limitations of the code, caused by the restriction on the number of computational nodes, a supervisor program was devised, which handles each blanket module separately at first, and then corrects the flow rate, considering pressure drops that occur in each module. For a feasibility test of the proposed methodology, 46 blankets in a single sector were simulated; the calculation results of the parameters, such as mass flow, pressure drops, and temperature distribution in the multiple blanket modules showed that the multi-module analysis method can be used for efficient thermal-hydraulic analysis of the fusion DEMO reactor.

  11. 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

  12. Thermal hydraulic similarity analysis of the integral effect test facility for main steam line break events

    Energy Technology Data Exchange (ETDEWEB)

    Choi, K.Y.; Park, H.S.; Euh, D.J.; Kwon, T.S.; Baek, W.P. [Thermal Hydraulic Safety Research Division Korea Atomic Energy Research Institute 150 Dukjin-Dong, Yusong-Gu, Daejeon 305-353 (Korea, Republic of)

    2005-07-01

    Full text of publication follows: A thermal-hydraulic integral effect test facility, ATLAS (Advanced Thermal-hydraulic Test Loop for Accident Simulation), is being constructed at Korea Atomic Energy Research Institute (KAERI). The ATLAS is a 1/2 reduced height and 1/288 volume scaled test facility based on the design features of the APR1400, an evolutionary pressurized water reactor developed by Korean industry. The ATLAS will be used to get more realistic understanding of the thermal hydraulic phenomena following postulated events and to carry out performance evaluation and safety analysis of the reference plants. The MSLB (Main Steam Line Break) event is one of the representative non-LOCA events and thermalhydraulic phenomena following the event are to be investigated in the ATLAS. In this paper, thermal hydraulic similarity for MSLB events between the ATLAS and the prototype plant, APR1400 is assessed by using the MARS code, which is a multi-dimensional best-estimate thermal hydraulic code being developed by KAERI. Several cases including SLBFPLOOP and SLBFP are taken into account for similarity analysis in this paper. The neutronic effects such as moderator temperature coefficients and doppler reactivity in APR1400 are not considered in this study. The same control logics for the major sequence of events such as reactor trip, turbine trip, valve opening and actuation of the emergency cooling system are applied to the ATLAS and the APR1400. The present investigation is focused on the scaling and the reduced power effects on thermal hydraulic similarity after initiation of MSLB events. It is found that the ATLAS facility has the similar thermal hydraulic responses against the MSLB events. However, the initial high secondary pressure before the MSLB initiation resulted in different primary pressure and temperature progression from the APR1400. The break flow from the main steam line is found to be one of the most dominating parameters governing the transient

  13. 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

  14. Thermal-hydraulic modeling of the Pennsylvania State University Breazeale Nuclear Reactor (PSBR)

    Science.gov (United States)

    Chang, Jong E.

    2005-11-01

    Earlier experiments determined that the Pennsylvania State University Breazeale Nuclear Reactor (PSBR) core is cooled, not by an axial flow, but rather by a strong cross flow due to the thermal expansion of the coolant. To further complicate the flow field, a nitrogen-16 (N-16) pump was installed above the PSBR core to mix the exiting core buoyant thermal plume in order to delay the rapid release of radioactive N-16 to the PSBR pool surface. Thus, the interaction between the N-16 jet flow and the buoyancy driven flow complicates the analysis of the flow distribution in the PSBR pool. The main objectives of this study is to model the thermal-hydraulic behavior of the PSBR core and pool. During this study four major things were performed including the Computational Fluid Dynamics (CFD) model for the PSBR pool, the stand-alone fuel rod model for a PSBR fuel rod, the velocity measurements in and around the PSBR core, and the temperature measurements in the PSBR pool. Once the flow field was predicted by the CFD model, the measurement devices were manufactured and calibrated based on the CFD results. The major contribution of this study is to understand and to explain the flow behavior in the PSBR subchannels and pool using the FLOW3D model. The stand-alone dynamic fuel rod model was developed to determine the temperature distribution inside a PSBR fuel rod. The stand-alone fuel rod model was coupled to the FLOW3D model and used to predict the temperature behavior during steady-state and pulsing. The heat transfer models in the stand-alone fuel rod code are used in order to overcome the disadvantage of the CFD code, which does not calculate the mechanical stress, the gap conductance, and the two phase heat transfer. (Abstract shortened by UMI.)

  15. Thermal-hydraulics of LLCB TBM under different ITER operational conditions

    Energy Technology Data Exchange (ETDEWEB)

    Chaudhuri, Paritosh, E-mail: paritc@gmail.com; Ranjithkumar, S.; Sharma, Deepak; Danani, Chandan; Kumar, E. Rajendra

    2016-11-01

    Highlights: • Thermal analysis of LLCB TBM has been performed for different ITER operations during normal and power excursion scenarios. • It is observed that in all ITER operations, the temperatures in all components of LLCB TBM are well within the limits. • The maximum temperature of CB-1 in steady state is 920 °C, which is 72% of the maximum temperature obtained in ITER pulse. - Abstract: Lead–lithium cooled ceramic breeder (LLCB) TBM is the Indian DEMO relevant blanket module, which will be tested in one-half of a designated ITER port from day 1 operation of ITER. LLCB TBM will be tested in various ITER operational phases. Different TBM may be tested corresponding to the four different phases of plasma operation. The paper present a review of thermal analyses based on the various heat loads on the LLCB TBM obtained from neutronics calculations to confirm heat removal and structural integrity under different ITER operational conditions. This includes the normal plasma operation, fusion power excursions, surface heat flux due to MARFES, disruption heat loads during thermal and current quench etc., in inductive and non-inductive plasma operations. Transient events, like plasma disruption, vertical displacement events (VDEs) and edge localized modes (ELMs) deliver considerable heat flux (high power densities) for short durations onto the first wall of TBM. Detailed thermal-hydraulic simulation studies have been performed using ANSYS. These analyses, therefore, will provide important information about the temperature distribution in different materials used in TBM and their temperature-time histories.

  16. Investigation of Thermal Hydraulics of a Nuclear Reactor Moderator

    Science.gov (United States)

    Sarchami, Araz

    A three-dimensional numerical modeling of the thermo hydraulics of Canadian Deuterium Uranium (CANDU) nuclear reactor is conducted. The moderator tank is a Pressurized heavy water reactor which uses heavy water as moderator in a cylindrical tank. The main use of the tank is to bring the fast neutrons to the thermal neutron energy levels. The moderator tank compromises of several bundled tubes containing nuclear rods immersed inside the heavy water. It is important to keep the water temperature in the moderator at sub-cooled conditions, to prevent potential failure due to overheating of the tubes. Because of difficulties in measuring flow characteristics and temperature conditions inside a real reactor moderator, tests are conducted using a scaled moderator in moderator test facility (MTF) by Chalk River Laboratories of Atomic Energy of Canada Limited (CRL, AECL). MTF tests are conducted using heating elements to heat tube surfaces. This is different than the real reactor where nuclear radiation is the source of heating which results in a volumetric heating of the heavy water. The data recorded inside the MTF tank have shown levels of fluctuations in the moderator temperatures and requires in depth investigation of causes and effects. The purpose of the current investigation is to determine the causes for, and the nature of the moderator temperature fluctuations using three-dimensional simulation of MTF with both (surface heating and volumetric heating) modes. In addition, three dimensional simulation of full scale actual moderator tank with volumetric heating is conducted to investigate the effects of scaling on the temperature distribution. The numerical simulations are performed on a 24-processor cluster using parallel version of the FLUENT 12. During the transient simulation, 55 points of interest inside the tank are monitored for their temperature and velocity fluctuations with time.

  17. Analytical Algorithms for the Blend Ratios by Fibre-bundle Tensile Curves Part Ⅱ:Calculations of the Modulus Method and the Percentage Method

    Institute of Scientific and Technical Information of China (English)

    YU Wei-dong(于伟东); Ron Postle; YAN Hao-jing(严灏景)

    2004-01-01

    The principles for the modulus method and the percentage method are established and discussed in the part following Part Ⅰ of the series papers, in which we proposed the various algorithms of the strength method and the work method. The samples of Wool/PET blended fibre bundles,the method of fibre-bundle tensile tests and the typical specific stress-extension curves from the fibre bundles with different blend ratios are the same as in Part Ⅰ. Itcan be found that the theoretical results estimated by the modulus and percentage methods accord with the experimental values highly though the calculations of the two methods are slightly more complex than those of the strength and work methods. Especially, using the modulus method can not only avoid the influence of the error caused by the determination of the tensile curve of no fibre breaking in stretching, Y (e), but also need not to know the tensile curves of mono-component fibre bundles in certaincalculation. The latter advantage of the modulus method exists in the percentage method too, but it should adopt the improved calculation of ones.

  18. 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.

  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, 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)

  20. Thermal-hydraulic modeling and analysis of hydraulic system by pseudo-bond graph

    Institute of Scientific and Technical Information of China (English)

    胡均平; 李科军

    2015-01-01

    To increase the efficiency and reliability of the thermodynamics analysis of the hydraulic system, the method based on pseudo-bond graph is introduced. According to the working mechanism of hydraulic components, they can be separated into two categories: capacitive components and resistive components. Then, the thermal-hydraulic pseudo-bond graphs of capacitive C element and resistance R element were developed, based on the conservation of mass and energy. Subsequently, the connection rule for the pseudo-bond graph elements and the method to construct the complete thermal-hydraulic system model were proposed. On the basis of heat transfer analysis of a typical hydraulic circuit containing a piston pump, the lumped parameter mathematical model of the system was given. The good agreement between the simulation results and experimental data demonstrates the validity of the modeling method.

  1. Study on thermal-hydraulics of natural circulation operation for decay heat removal of LMFBR

    Energy Technology Data Exchange (ETDEWEB)

    Koga, Tomonari [Central Research Inst. of Electric Power Industry, Abiko, Chiba (Japan). Abiko Research Lab

    2000-05-01

    Thermal-hydraulic characteristics of the decay heat removal system (DHRS) has been evaluated by both a water test and its computational analysis. The direct reactor auxiliary cooling system (DRACS) was studied as a representative DHRS in the LMFBR design. A natural circulation flow of the DRACS is originated from the buoyancy force balanced with the pressure loss within the active core. Hence it appears that the performance of the DRACS could be fully estimated by the thermal-hydraulic study of one fuel sub-assembly under the condition of low flow rate. The inter-wrapper flow occurring in the gap between core sub-assemblies under the natural circulation operation might have a large capacity to cool the core. An inverse flow occurring at the outer region of the core was due to the cooling effect of the inter-wrapper flow. (author)

  2. Thermal-hydraulic characteristics of natural circulation operation for decay heat removal of LMFBR

    Energy Technology Data Exchange (ETDEWEB)

    Koga, Tomonari [Central Research Inst. of Electric Power Industry, Abiko, Chiba (Japan); Watanabe, Osamu [Advanced Reactor Technology Co., Ltd., Tokyo (Japan)

    2000-10-01

    Thermal-hydraulic characteristics of the decay heat removal system (DHRS) has been evaluated by both a water test and its computational analysis. The direct reactor auxiliary cooling system (DRACS) was studied as a representative DHRS in the LMFBR design. A natural circulation flow of the DRACS is originated from the buoyancy force balanced with the pressure loss within the active core. Hence it appears that the performance of the DRACS could be fully estimated by the thermal-hydraulic study of one fuel sub-assembly under the condition of low flow rate. The inter-wrapper flow occurring in the gap between core sub-assemblies under the natural circulation operation might have a large capacity to cool the core. An inverse flow occurring at the outer region of the core was due to the cooling effect of the inter-wrapper flow. (author)

  3. 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

  4. THERMAL HYDRAULIC ANALYSIS OF A LIQUID-METAL-COOLED NEUTRON SPALLATION TARGET

    Energy Technology Data Exchange (ETDEWEB)

    W. GREGORY; R. MARTIN; T. VALACHOVIC

    2000-07-01

    We have carried out numerical simulations of the thermal hydraulic behavior of a neutron spallation target where liquid metal lead-bismuth serves as both coolant and as a neutron spallation source. The target is one of three designs provided by the Institute of Physics and Power Engineering (IPPE) in Russia. This type of target is proposed for Accelerator-driven Transmutation of Waste (ATW) to eliminate plutonium from hazardous fission products. The thermal hydraulic behavior was simulated by use of a commercial CFD computer code called CFX. Maximum temperatures in the diaphragm window and in the liquid lead were determined. In addition the total pressure drop through the target was predicted. The results of the CFX analysis were close to those results predicted by IPPE in their preliminary analysis.

  5. 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.

  6. 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.

  7. 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.

  8. 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.

  9. Thermal-hydraulic Analysis of New Zirconium Alloys Assembly Irradiated in CARR

    Institute of Scientific and Technical Information of China (English)

    YIN; Hao; ZHAO; Shou-zhi; LIU; Xing-min

    2013-01-01

    This article is mainly about the thermal-hydraulic analysis of the new zirconium alloys assembly on irradiation test of China Advanced Research Reactor(CARR),so as to provide security assessment throughout the design.CFD software was used for three-dimensional simulation.Firstly,the geometric model,mesh,specified boundary condition types and region types were constructed.Then importing the

  10. Thermal Hydraulic Analysis of a Passive Residual Heat Removal System for an Integral Pressurized Water Reactor

    OpenAIRE

    2009-01-01

    A theoretical investigation on the thermal hydraulic characteristics of a new type of passive residual heat removal system (PRHRS), which is connected to the reactor coolant system via the secondary side of the steam generator, for an integral pressurized water reactor is presented in this paper. Three-interknited natural circulation loops are adopted by this PRHRS to remove the residual heat of the reactor core after a reactor trip. Based on the one-dimensional model and a simulation code (S...

  11. Thermal-hydraulics Analysis of a Radioisotope-powered Mars Hopper Propulsion System

    Energy Technology Data Exchange (ETDEWEB)

    Robert C. O' Brien; Andrew C. Klein; William T. Taitano; Justice Gibson; Brian Myers; Steven D. Howe

    2011-02-01

    Thermal-hydraulics analyses results produced using a combined suite of computational design and analysis codes are presented for the preliminary design of a concept Radioisotope Thermal Rocket (RTR) propulsion system. Modeling of the transient heating and steady state temperatures of the system is presented. Simulation results for propellant blow down during impulsive operation are also presented. The results from this study validate the feasibility of a practical thermally capacitive RTR propulsion system.

  12. Neutron Tomography Using Mobile Neutron Generators for Assessment of Void Distributions in Thermal Hydraulic Test Loops

    OpenAIRE

    Andersson, Peter; Bjelkenstedt, Tom; Andersson Sundén, Erik; Sjöstrand, Henrik; Jacobsson, Staffan

    2015-01-01

    Detailed knowledge of the lateral distribution of steam (void) and water in a nuclear fuel assembly is of great value for nuclear reactor operators and fuel manufacturers, with consequences for both reactor safety and economy of operation. Therefore, nuclear relevant two-phase flows are being studied at dedicated thermal-hydraulic test loop, using twophase flow systems ranging from simplified geometries such as heated circular pipes to full scale mock-ups of nuclear fuel assemblies. Neutron t...

  13. Test program element II blanket and shield thermal-hydraulic and thermomechanical testing, experimental facility survey

    Energy Technology Data Exchange (ETDEWEB)

    Ware, A.G.; Longhurst, G.R.

    1981-12-01

    This report presents results of a survey conducted by EG and G Idaho to determine facilities available to conduct thermal-hydraulic and thermomechanical testing for the Department of Energy Office of Fusion Energy First Wall/Blanket/Shield Engineering Test Program. In response to EG and G queries, twelve organizations (in addition to EG and G and General Atomic) expressed interest in providing experimental facilities. A variety of methods of supplying heat is available.

  14. Three dimensional neutronic/thermal-hydraulic coupled simulation of MSR in transient state condition

    Energy Technology Data Exchange (ETDEWEB)

    Zhou, Jianjun [School of Nuclear Science and Technology, Xi’an Jiaotong University, Xianning Road, 28, Xi’an 710049, Shaanxi (China); College of Mechanical and Power Engineering, China Three Gorges University, No 8, Daxue road, Yichang, Hubei 443002 (China); Zhang, Daling, E-mail: dlzhang@mail.xjtu.edu.cn [School of Nuclear Science and Technology, Xi’an Jiaotong University, Xianning Road, 28, Xi’an 710049, Shaanxi (China); Qiu, Suizheng; Su, Guanghui; Tian, Wenxi; Wu, Yingwei [School of Nuclear Science and Technology, Xi’an Jiaotong University, Xianning Road, 28, Xi’an 710049, Shaanxi (China)

    2015-02-15

    Highlights: • Developed a three dimensional neutronic/thermal-hydraulic coupled transient analysis code for MSR. • Investigated the neutron distribution and thermal-hydraulic characters of the core under transient condition. • Analyzed three different transient conditions of inlet temperature drop, reactivity jump and pump coastdown. - Abstract: MSR (molten salt reactor) use liquid molten salt as coolant and fuel solvent, which was the only one liquid reactor of six Generation IV reactor types. As a liquid reactor the physical property of reactor was significantly influenced by fuel salt flow and the conventional analysis methods applied in solid fuel reactors are not applicable for this type of reactors. The present work developed a three dimensional neutronic/thermal-hydraulic coupled code investigated the neutronics and thermo-hydraulics characteristics of the core in transient condition based on neutron diffusion theory and numerical heat transfer. The code consists of two group neutron diffusion equations for fast and thermal neutron fluxes and six group balance equations for delayed neutron precursors. The code was separately validated by neutron benchmark and flow and heat transfer benchmark. Three different transient conditions was analyzed with inlet temperature drop, reactivity jump and pump coastdown. The results provide some valuable information in design and research this kind of reactor.

  15. 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.

  16. 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.

  17. 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)

  18. Lead Coolant Test Facility Systems Design, Thermal Hydraulic Analysis and Cost Estimate

    Energy Technology Data Exchange (ETDEWEB)

    Soli Khericha; Edwin Harvego; John Svoboda; Ryan Dalling

    2012-01-01

    The Idaho National Laboratory prepared a preliminary technical and functional requirements (T&FR), thermal hydraulic design and cost estimate for a lead coolant test facility. The purpose of this small scale facility is to simulate lead coolant fast reactor (LFR) coolant flow in an open lattice geometry core using seven electrical rods and liquid lead or lead-bismuth eutectic coolant. Based on review of current world lead or lead-bismuth test facilities and research needs listed in the Generation IV Roadmap, five broad areas of requirements were identified as listed: (1) Develop and Demonstrate Feasibility of Submerged Heat Exchanger; (2) Develop and Demonstrate Open-lattice Flow in Electrically Heated Core; (3) Develop and Demonstrate Chemistry Control; (4) Demonstrate Safe Operation; and (5) Provision for Future Testing. This paper discusses the preliminary design of systems, thermal hydraulic analysis, and simplified cost estimate. The facility thermal hydraulic design is based on the maximum simulated core power using seven electrical heater rods of 420 kW; average linear heat generation rate of 300 W/cm. The core inlet temperature for liquid lead or Pb/Bi eutectic is 4200 C. The design includes approximately seventy-five data measurements such as pressure, temperature, and flow rates. The preliminary estimated cost of construction of the facility is $3.7M (in 2006 $). It is also estimated that the facility will require two years to be constructed and ready for operation.

  19. Development of scaling laws on thermal-hydraulic effect test facility for CANDU-6 moderator

    Energy Technology Data Exchange (ETDEWEB)

    Lee, J. Y.; Jung, Y. S.; Kim, N. S. [Handong University, Pohang (Korea, Republic of); Kim, M. W.; Kim, H. J. [Korea Institute of Nuclear Safety, Taejon (Korea, Republic of)

    2004-07-01

    The scaling laws on thermal-hydraulic effect test facility for CANDU-6 moderator (HGU-KINS) have been investigated and manufactured. The basic laws are the satisfaction of energy conservation and dimensionless number, Ar and Re, for the similarities of thermal-hydraulic properties. And then the thermal-hydraulic scaling analyses of test facilities, SPEL(1/10 scale) and STERN(1/4 scale), have been identified by the present method. As a result, in the case of SPEL, the energy conservation is confirmed, but the similarities of Ar and the heat density are not considered. In the case of STERN, the energy conservation and the characteristics of Ar were well defined. But the similarity of the heat density is unsatisfied, either. Therefore the present method was applied with 1/8 length scale. For the performance test, CFD analysis has been accomplished by CFX5. The results of flow pattern certifications and variation of axial properties with CANDU show that the present scaling method is acceptable.

  20. Lead coolant test facility systems design, thermal hydraulic analysis and cost estimate

    Energy Technology Data Exchange (ETDEWEB)

    Khericha, Soli, E-mail: slk2@inel.gov [Battelle Energy Alliance, LLC, Idaho National Laboratory, Idaho Falls, ID 83415 (United States); Harvego, Edwin; Svoboda, John; Evans, Robert [Battelle Energy Alliance, LLC, Idaho National Laboratory, Idaho Falls, ID 83415 (United States); Dalling, Ryan [ExxonMobil Gas and Power Marketing, Houston, TX 77069 (United States)

    2012-01-15

    The Idaho National Laboratory prepared a preliminary technical and functional requirements (T and FR), thermal hydraulic design and cost estimate for a lead coolant test facility. The purpose of this small scale facility is to simulate lead coolant fast reactor (LFR) coolant flow in an open lattice geometry core using seven electrical rods and liquid lead or lead-bismuth eutectic coolant. Based on review of current world lead or lead-bismuth test facilities and research needs listed in the Generation IV Roadmap, five broad areas of requirements were identified as listed below: Bullet Develop and demonstrate feasibility of submerged heat exchanger. Bullet Develop and demonstrate open-lattice flow in electrically heated core. Bullet Develop and demonstrate chemistry control. Bullet Demonstrate safe operation. Bullet Provision for future testing. This paper discusses the preliminary design of systems, thermal hydraulic analysis, and simplified cost estimated. The facility thermal hydraulic design is based on the maximum simulated core power using seven electrical heater rods of 420 kW; average linear heat generation rate of 300 W/cm. The core inlet temperature for liquid lead or Pb/Bi eutectic is 4200 Degree-Sign C. The design includes approximately seventy-five data measurements such as pressure, temperature, and flow rates. The preliminary estimated cost of construction of the facility is $3.7M (in 2006 $). It is also estimated that the facility will require two years to be constructed and ready for operation.

  1. 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.

  2. Nuclear-coupled thermal-hydraulic stability analysis of boiling water reactors

    Science.gov (United States)

    Karve, Atul A.

    We have studied the nuclear-coupled thermal-hydraulic stability of boiling water reactors (BWRs) using a model we developed from: the space-time modal neutron kinetics equations based on spatial omega-modes, the equations for two-phase flow in parallel boiling channels, the fuel rod heat conduction equations, and a simple model for the recirculation loop. The model is represented as a dynamical system comprised of time-dependent nonlinear ordinary differential equations, and it is studied using stability analysis, modern bifurcation theory, and numerical simulations. We first determine the stability boundary (SB) in the most relevant parameter plane, the inlet-subcooling-number/external-pressure-drop plane, for a fixed control rod induced external reactivity equal to the 100% rod line value and then transform the SB to the practical power-flow map. Using this SB, we show that the normal operating point at 100% power is very stable, stability of points on the 100% rod line decreases as the flow rate is reduced, and that points are least stable in the low-flow/high-power region. We also determine the SB when the modal kinetics is replaced by simple point reactor kinetics and show that the first harmonic mode has no significant effect on the SB. Later we carry out the relevant numerical simulations where we first show that the Hopf bifurcation, that occurs as a parameter is varied across the SB is subcritical, and that, in the important low-flow/high-power region, growing oscillations can result following small finite perturbations of stable steady-states on the 100% rod line. Hence, a point on the 100% rod line in the low-flow/high-power region, although stable, may nevertheless be a point at which a BWR should not be operated. Numerical simulations are then done to calculate the decay ratios (DRs) and frequencies of oscillations for various points on the 100% rod line. It is determined that the NRC requirement of DR loop model that we develop is studied by carrying

  3. 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

  4. 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

  5. Thermal-hydraulic behavior of physical quantities at critical velocities in a nuclear research reactor core channel using plate type fuel

    Directory of Open Access Journals (Sweden)

    Sidi Ali Kamel

    2012-01-01

    Full Text Available The thermal-hydraulic study presented here relates to a channel of a nuclear reactor core. This channel is defined as being the space between two fuel plates where a coolant fluid flows. The flow velocity of this coolant should not generate vibrations in fuel plates. The aim of this study is to know the distribution of the temperature in the fuel plates, in the cladding and in the coolant fluid at the critical velocities of Miller, of Wambsganss, and of Cekirge and Ural. The velocity expressions given by these authors are function of the geometry of the fuel plate, the mechanical characteristics of the fuel plate’s material and the thermal characteristics of the coolant fluid. The thermal-hydraulic study is made under steady-state; the equation set-up of the thermal problem is made according to El Wakil and to Delhaye. Once the equation set-up is validated, the three critical velocities are calculated and then used in the calculations of the different temperature profiles. The average heat flux and the critical heat flux are evaluated for each critical velocity and their ratio reported. The recommended critical velocity to be used in nuclear channel calculations is that of Wambsganss. The mathematical model used is more precise and all the physical quantities, when using this critical velocity, stay in safe margins.

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

    Science.gov (United States)

    Ohshima, Hiroyuki; Uwaba, Tomoyuki; Hashimoto, Akihiko; Imai, Yasutomo; Ito, Masahiro

    2015-12-01

    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.

  7. 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.

  8. Review of fuel assembly and pool thermal hydraulics for fast reactors

    Energy Technology Data Exchange (ETDEWEB)

    Roelofs, Ferry, E-mail: roelofs@nrg.eu; Gopala, Vinay R.; Jayaraju, Santhosh; Shams, Afaque; Komen, Ed

    2013-12-15

    Highlights: • Literature review of fuel assembly and pool thermal hydraulics for fast reactors. • Experiments and state-of-the-art simulations. • For wire wrapped fuel assemblies RANS for complete fuel assembly is state-of-the-art, LES serves reference. • For pool thermal hydraulics, typically 5 to 20 million computational volumes are used in RANS simulations. • Gas entrainment analyses are extremely demanding as in addition they request multiphase modelling. -- Abstract: Liquid metal cooled reactors are envisaged to play an important role in the future of nuclear energy production because of their possible efficient use of uranium and the possibility to reduce the volume and lifetime of nuclear waste. Thermal-hydraulics is recognized as a key scientific subject in the development of such reactors. Two important challenges for the design of liquid metal fast reactors (LMFRs) are fuel assembly and pool thermal hydraulics. The heart of every nuclear reactor is the core, where the nuclear chain reaction takes place. Heat is produced in the nuclear fuel and transported to the coolant. LMFR core designs consist of many fuel assemblies which in turn consist of a large number of fuel rods. Wire wraps are commonly envisaged as spacer design in LMFR fuel assemblies. For the design and safety analyses of such reactors, simulations of the heat transport within the core are essential. The flow exiting the core is made up of the outlets of many different fuel assemblies. The liquid metal in these assemblies may be heated up to different temperatures. This leads to temperature fluctuations on various above core structures. As these temperature fluctuations may lead to thermal fatigue damage of the structures, an accurate characterization of the liquid metal flow field in the above core region is very important. This paper will provide an overview of state-of-the-art evaluations of fuel assembly and pool thermal hydraulics for LMFRs. It will show the tight interaction

  9. Prediction of Flow Regimes and Thermal Hydraulic Parameters in Two-Phase Natural Circulation by RELAP5 and TRACE Codes

    Directory of Open Access Journals (Sweden)

    Viet-Anh Phung

    2015-01-01

    Full Text Available In earlier study we have demonstrated that RELAP5 can predict flow instability parameters (flow rate, oscillation period, temperature, and pressure in single channel tests in CIRCUS-IV facility. The main goals of this work are to (i validate RELAP5 and TRACE capabilities in prediction of two-phase flow instability and flow regimes and (ii assess the effect of improvement in flow regime identification on code predictions. Most of the results of RELAP5 and TRACE calculation are in reasonable agreement with experimental data from CIRCUS-IV. However, both codes misidentified instantaneous flow regimes which were observed in the test with high speed camera. One of the reasons for the incorrect identification of the flow regimes is the small tube flow regime transition model in RELAP5 and the combined bubbly-slug flow regime in TRACE. We found that calculation results are sensitive to flow regime boundaries of RELAP5 which were modified in order to match the experimental data on flow regimes. Although the flow regime became closer to the experimental one, other predicted thermal hydraulic parameters showed larger discrepancy with the experimental data than with the base case calculations where flow regimes were misidentified.

  10. Numerical modeling of the thermal-hydraulic behavior of wire-on-tube condensers operating with HFC-134a using homogeneous equilibrium model: evaluation of some void fraction correlations

    Science.gov (United States)

    Guzella, Matheus dos Santos; Cabezas-Gómez, Luben; da Silva, José Antônio; Maia, Cristiana Brasil; Hanriot, Sérgio de Morais

    2016-02-01

    This study presents a numerical evaluation of the influence of some void fraction correlations over the thermal-hydraulic behavior of wire-on-tube condensers operating with HFC-134a. The numerical model is based on finite volume method considering the homogeneous equilibrium model. Empirical correlations are applied to provide closure relations. Results show that the choice of void fraction correlation influences the refrigerant charge and pressure drop calculations, while no influences the heat transfer rate.

  11. 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.

  12. Thermal-hydraulic modeling of the steady-state operating conditions of a fire-tube boiler

    Directory of Open Access Journals (Sweden)

    Rahmani Ahmed

    2009-01-01

    Full Text Available In this work, we are interested to simulate the thermal-hydraulic behavior of three-pass type fire-tube boiler. The plant is designed to produce 4.5 tons per hour of saturated steam at 8 bar destined principally for heating applications. A calculation program is developed in order to simulate the boiler operation under several steady-state operating conditions. This program is based upon heat transfer laws between hot gases and the fire-tube internal walls. In the boiler combustion chamber, the heat transfer has been simulated using the well-stirred furnace model. In the convection section, heat balance has been carried out to estimate the heat exchanges between the hot gases and the tube banks. The obtained results are compared to the steady-state operating data of the considered plant. A comparative analysis shows that the calculation results are in good agreement with the boiler operating data. Furthermore, a sensitivity study has been carried out to assess the effects of input parameters, namely the fuel flow rate, air excess, ambient temperature, and operating pressure, upon the boiler thermal performances.

  13. Development of Thermal-hydraulic Analysis Methodology for Multi-module Breeding Blankets in K-DEMO

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Geon-Woo; Lee, Jeong-Hun; Park, Goon-Cherl; Cho, Hyoung-Kyu [Seoul National University, Seoul (Korea, Republic of); Im, Kihak [National Fusion Research Institute, Daejeon (Korea, Republic of)

    2015-05-15

    In this paper, the purpose of the analyses is to extend the capability of MARS-KS to the entire blanket system which includes a few hundreds of single blanket modules. Afterwards, the plan for the whole blanket system analysis using MARS-KS is introduced and the result of the multiple blanket module analysis is summarized. A thermal-hydraulic analysis code for a nuclear reactor safety, MARS-KS, was applied for the conceptual design of the K-DEMO breeding blanket thermal analysis. Then, a methodology to simulate multiple blanket modules was proposed, which uses a supervisor program to handle each blanket module individually at first and then distribute the flow rate considering pressure drops arises in each module. For a feasibility test of the proposed methodology, 10 outboard blankets in a toroidal field sector were simulated, which are connected with each other through the inlet and outlet common headers. The calculation results of flow rates, pressure drops, and temperatures showed the validity of the calculation and thanks to the parallelization using MPI, almost linear speed-up could be obtained.

  14. Coolant mixing in LMFBR rod bundles and outlet plenum mixing transients

    Science.gov (United States)

    Todreas, N. E.; Cheng, S. K.; Basehore, K.

    1984-08-01

    The thermal hydraulic performance of wire wrapped fuel bundles of LMFBR configuration was investigated. Results obtained included phenomenological models for friction factors, flow split and mixing characteristics; correlations for predicting these characteristics suitable for insertion in design codes; numerical codes for analyzing bundle behavior both of the lumped subchannel and distributed parameter categories and experimental techniques for pressure velocity, flow split, salt conductivity and temperature measurement in water cooled mockups of bundles and subchannels. Flow regimes investigated included laminar, transition and turbulent flow under forced convection and mixed convection conditions. Forced convections conditions are emphasized. Outlet plenum behavior is also investigated.

  15. Test Facility Construction for Flow Visualization on Mixing Flow inside Subchannels of PWR Rod Bundle

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Seok; Jeon, Byong-Guk; Youn, Young-Jung; Choi, Hae-Seob; Euh, Dong-Jin [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2016-10-15

    Flow inside rod bundles has a similarity with flow in porous media. To ensure thermal performance of a nuclear reactor, detailed information of the heat transfer and turbulent mixing flow phenomena taking place within the subchannels is required. The subchannel analysis is one of the key thermal-hydraulic calculations in the safety analysis of the nuclear reactor core. At present, subchannel computer codes are employed to simulate fuel elements of nuclear reactor cores and predict the performance of cores under normal operating and hypothetical accident conditions. The ability of these subchannels codes to predict both the flow and enthalpy distribution in fuel assemblies is very important in the design of nuclear reactors. Recently, according to the modern tend of the safety analysis for the nuclear reactor, a new component scale analysis code, named CUPID, and has been developed in KAERI. The CUPID code is based on a two-fluid and three-field model, and both the open and porous media approaches are incorporated. The PRIUS experiment has addressed many key topics related to flow behaviour in a rod bundle. These issues are related to the flow conditions inside a nuclear fuel element during normal operation of the plant or in accident scenarios. From the second half of 2016, flow visualization will be performed by using a high speed camera and image analysis technique, from which detailed information for the two-dimensional movement of single phase flow is quantified.

  16. Strategic Need for Multi-Purpose Thermal Hydraulic Loop for Support of Advanced Reactor Technologies

    Energy Technology Data Exchange (ETDEWEB)

    James E. O' Brien; Piyush Sabharwall; Su-Jong Yoon; Gregory K. Housley

    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

  17. Resolution of thermal-hydraulic safety and licensing issues for the system 80+{sup {trademark}} design

    Energy Technology Data Exchange (ETDEWEB)

    Carpentino, S.E.; Ritterbusch, S.E.; Schneider, R.E. [ABB-Combustion Engineering, Windsor, CT (United States)] [and others

    1995-09-01

    The System 80+{sup {trademark}} Standard Design is an evolutionary Advanced Light Water Reactor (ALWR) with a generating capacity of 3931 MWt (1350 MWe). The Final Design Approval (FDA) for this design was issued by the Nuclear Regulatory Commission (NRC) in July 1994. The design certification by the NRC is anticipated by the end of 1995 or early 1996. NRC review of the System 80+ design has involved several new safety issues never before addressed in a regulatory atmosphere. In addition, conformance with the Electric Power Research Institute (EPRI) ALWR Utility Requirements Document (URD) required that the System 80+ plant address nuclear industry concerns with regard to design, construction, operation and maintenance of nuclear power plants. A large number of these issues/concerns deals with previously unresolved generic thermal-hydraulic safety issues and severe accident prevention and mitigation. This paper discusses the thermal-hydraulic analyses and evaluations performed for the System 80+ design to resolve safety and licensing issues relevant to both the Nuclear Stream Supply System (NSSS) and containment designs. For the NSSS design, the Safety Depressurization System mitigation capability and resolution of the boron dilution concern are described. Examples of containment design issues dealing with containment shell strength, robustness of the reactor cavity walls and hydrogen mixing under severe accident conditions are also provided. Finally, the overall approach used in the application of NRC`s new (NUREG-1465) radiological source term for System 80+ evaluation is described. The robustness of the System 80+ containment design to withstand severe accident consequences was demonstrated through detailed thermal-hydraulic analyses and evaluations. This advanced design to shown to meet NRC severe accident policy goals and ALWR URD requirements without any special design features and unnecessary costs.

  18. Scaling analysis of the thermal-hydraulic test facility for the large break LOCA of KNGR

    Energy Technology Data Exchange (ETDEWEB)

    Yun, Byong Jo; Kwon, Tae Soon; Song, Chul Hwa; Euh, Dong Jin; Chu, In Cheol; Cho, Hyoung Kyu; Park, Jong Kyun

    2001-03-01

    Korea Next Generation Reactor(KNGR) adopts a Direct Vessel Injection (DVI) system instead of conventional Cold Leg Injection (CLI) system. In this report, a scaling analysis for the steam-water test facility of KNGR with DVI under reflood phase of Loss of Coolant Accident(LBLOCA) is carried out. The major objectives of the test facility are to clarify the thermal hydraulics phenomena in the upper downcomer region and to provide experimental data for evaluating or validating relevant thermal hydraulic models and correlations of the best estimate codes. The test facility should be designed based on the appropriate scaling law so that the same thermal hydraulics phenomena is happened as in the case of prototype. For these, the investigations of previous scaling laws are carried out. And, in the present study, a new scaling approach, named the modified linear scaling, is developed for the design of a scaled-down experimental facility. Its velocity is scaled by a Wallis-type parameter and an aspect ratio of experimental facility is preserved with that of a prototype. The test facility is designed primarily by a volume scaling law and the area ratio of test facility is set to be 1/24.3. However, additional DVI nozzles are also installed at the elevation which is determined by the modified linear scaling law. It is for the scaling analysis of ECC bypass fraction. The cold leg, hot leg and DVI nozzles are additionally attached in the upper annulus downcomer region so that the UPTF counterpart test is possible.

  19. 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

  20. Thermal-hydraulic analysis of the semiscale Mod-1 blowdown heat transfer test series. [PWR

    Energy Technology Data Exchange (ETDEWEB)

    Cozzuol, J.M.

    1976-06-01

    Selected experimental thermal-hydraulic data from the recent Semiscale Mod-1 blowdown heat transfer test series are analyzed from an experimental viewpoint with emphasis on explaining those phenomena which influence core fluid behavior. Comparisons are made between the trends measured by the system instrumentation and the trends predicted by the RELAP4 computer code to aid in obtaining an understanding of the interactions between phenomena occurring in different parts of the system. The analyses presented in this report are valuable for evaluating the adequacy and improving the predictive capability of analytical models developed to predict the system response of a pressurized water reactor during a postulated loss-of-coolant accident (LOCA).

  1. CFD Analysis on the Air-Side Thermal-Hydraulic Performance of Multi-Louvered Fin Heat Exchangers at Low Reynolds Numbers

    Directory of Open Access Journals (Sweden)

    Arslan Saleem

    2017-06-01

    Full Text Available The air-side thermal-hydraulic performance of multi-louvered aluminium fin heat exchangers is investigated. A systematic numerical study has been performed to analyze the air-sde thermal hydraulic characteristics over a wide range of Reynolds number i.e., from 30 to 500. Air-side heat transfer coefficient and pressure drop were calculated and validated over the mentioned band of Reynolds numbers. The critical Reynolds number was determined numerically; and also the variation of flow pattern along with the air-side heat transfer coefficient and pressure drop in a multi-louvered heat exchanger associated with R e c r i has been reported. Moreover, a parametric study of the multi-louvered aluminium fin heat exchangers was also performed for 36 heat exchanger configurations with the louver angles (19–31°; fin pitches (1.0, 1.2, 1.4 mm and flow depths (16, 20, 24 mm; and the geometric configuration exhibiting the highest air-side heat transfer coefficient was reported. The air-side heat transfer coefficient and pressure drop results for different geometrical configurations were presented in terms of Colburn j factor and Fanning friction factor f; as a function of Reynolds number based on louver pitch.

  2. Neutronic and Thermal-Hydraulic Safety Analysis for the Optimization of the Uranium Foil Target in the RSG-GAS Reactor

    Directory of Open Access Journals (Sweden)

    S. Pinem

    2016-12-01

    Full Text Available The G. A. Siwabessy Multipurpose Reactor (Reaktor Serba Guna G.A. Siwabessy, RSG-GAS has an average thermal neutron flux of 2×1014 neutron/(cm2 sec at the nominal power of 30 MW. With such a high thermal neutron flux, the reactor is suitable for the production of Mo-99 which is widely used as a medical diagnostic radioisotope. This paper describes a safety analysis to determine the optimum LEU foil target by using a coupled neutronic and thermal-hydraulic code, MTR-DYN. The code has been developed based on the three-dimensional multigroup neutron diffusion theory. The best estimated results can be achieved by using a coupled neutronic and thermal-hydraulic code. The calculation results show that the optimum LEU foil target is 54 g corresponding to the reactivity change of less than the limit value of 500 pcm. From the safety analysis for the case when the primary flow rate decreased by 15% from its nominal value, it was found that the peak temperatures of the coolant and cladding are 69.5°C and 127.9°C, respectively. It can be concluded that the optimum LEU foil target can be irradiated safely without exceeding the limit value.

  3. Test problem for thermal-hydraulics and neutronic coupled calculation fore ALFREAD reactor core

    Science.gov (United States)

    Filip, A.; Darie, G.; Saldikov, I. S.; Smirnov, A. D.; Tikhomirov, G. V.

    2017-01-01

    The beginning of a new era of nuclear reactor requires technological advances and also multiples studies. The European Liquid metal cooled Fast breeder Reactor is one of the designs for the generation IV nuclear reactor, selected by ENEA. A pioneer of its time, ELFR needs a demonstrator in order to prove the feasibility of this project and to acquire more data and experience in operating a LFR. For this reason the ALFRED project was started and it is expected to be under operation by the year 2030. This paper has the objective of analyzing the neutronic and thermohydraulics of the ALFRED core by the means of a coupled scheme. The selected code for neutronic simulation is MCNP and the selected code for thermohydraulics is ANSYS.

  4. 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.).

  5. Post-test thermal-hydraulic analysis of two intermediate LOCA tests at the ROSA facility including uncertainty evaluation

    Energy Technology Data Exchange (ETDEWEB)

    Freixa, J., E-mail: jordi@freixa.net [Paul Scherrer Institut (PSI) 5232 Villigen PSI (Switzerland); Kim, T.-W. [Paul Scherrer Institut (PSI) 5232 Villigen PSI (Switzerland); Manera, A. [University of Michigan, Ann Arbor, MI 48109 (United States)

    2013-11-15

    The OECD/NEA ROSA-2 project aims at addressing thermal-hydraulic safety issues relevant for light water reactors by building up an experimental database at the ROSA Large Scale Test Facility (LSTF). The ROSA facility simulates a PWR Westinghouse design with a four-loop configuration and a nominal power of 3423 MWth. Two intermediate break loss-of-coolant-accident (LOCA) experiments (Tests 1 and 2) have been carried out during 2010. The two tests were analyzed by using the US-NRC TRACE best estimate code, employing the same nodalization previously used for the simulation of small-break LOCA experiments of the ROSA-1 programme. A post-test calculation was performed for each test along with uncertainty analysis providing uncertainty bands for each relevant time trend. Uncertainties in the code modelling capabilities as well as in the initial and boundary conditions were taken into account, following the guidelines and lessons learnt through participation in the OECD/NEA BEMUSE programme. Two different versions of the TRACE code were used in the analysis, providing a qualitatively good prediction of the tests. However, the uncertainty analysis revealed differences between the performances of some models in the two versions. The most relevant parameters of the two experimental tests were falling within the computed uncertainty bands.

  6. Post-test thermal-hydraulic analysis of two intermediate LOCA tests at the ROSA facility including uncertainty evaluation

    Energy Technology Data Exchange (ETDEWEB)

    Freixa, J.; Kim, T-W.; Manera, A. [Paul Scherrer Inst., Villigen (Switzerland)

    2011-07-01

    The OECD/NEA ROSA-2 project aims at addressing thermal-hydraulic safety issues relevant for light water reactors by building up an experimental database at the ROSA Large Scale Test Facility (LSTF). The ROSA facility simulates a PWR Westinghouse design with a four-loop configuration and a nominal power of 3423 MWth. Two intermediate break loss-of-coolant-accident (LOCA) experiments (Test 1 and 2) have been carried out during 2010. The two tests were analyzed by using the US-NRC TRACE best estimate code, employing the same nodalization previously used for the simulation of small-break LOCA experiments of the ROSA-1 program. A post-test calculation was performed for each test along with uncertainty analysis providing uncertainty bands for each relevant time trend. Uncertainties in the code modeling capabilities as well as in the initial and boundary conditions were taken into account, following the guidelines and lessons learnt through participation in the OECD/NEA BEMUSE program. Two different versions of the TRACE code were used in the analysis, providing a qualitatively good prediction of the tests. However, both versions showed deficiencies that need to be addressed. The most relevant parameters of the two experimental tests were falling within the computed uncertainty bands. (author)

  7. COBRA-SFS (Spent Fuel Storage): A thermal-hydraulic analysis computer code: Volume 2, User's manual

    Energy Technology Data Exchange (ETDEWEB)

    Rector, D.R.; Cuta, J.M.; Lombardo, N.J.; Michener, T.E.; Wheeler, C.L.

    1986-11-01

    COBRA-SFS (Spent Fuel Storage) is a general thermal-hydraulic analysis computer code used to predict temperatures and velocities in a wide variety of systems. The code was refined and specialized for spent fuel storage system analyses for the US Department of Energy's Commercial Spent Fuel Management Program. The finite-volume equations governing mass, momentum, and energy conservation are written for an incompressible, single-phase fluid. The flow equations model a wide range of conditions including natural circulation. The energy equations include the effects of solid and fluid conduction, natural convection, and thermal radiation. The COBRA-SFS code is structured to perform both steady-state and transient calculations; however, the transient capability has not yet been validated. This volume contains the input instructions for COBRA-SFS and an auxiliary radiation exchange factor code, RADX-1. It is intended to aid the user in becoming familiar with the capabilities and modeling conventions of the code.

  8. 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)

  9. 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)

  10. Thermal hydraulic feasibility assessment of the hot conditioning system and process

    Energy Technology Data Exchange (ETDEWEB)

    Heard, F.J.

    1996-10-10

    The Spent Nuclear Fuel Project 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. A series of analyses have been completed investigating the thermal-hydraulic performance and feasibility of the proposed Hot Conditioning System and process for the Spent Nuclear Fuel Project. The analyses were performed using a series of thermal-hydraulic models that could respond to all process and safety-related issues that may arise pertaining to the Hot Conditioning System. The subject efforts focus on independently investigating, quantifying, and establishing the governing heat production and removal mechanisms, flow distributions within the multi-canister overpack, and performing process simulations for various purge gases under consideration for the Hot Conditioning System, as well as obtaining preliminary results for comparison with and verification of other analyses, and providing technology- based recommendations for consideration and incorporation into the Hot Conditioning System design bases.

  11. 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)

  12. Thermal-Hydraulic Issues in the ITER Toroidal Field Model Coil (TFMC) Test and Analysis

    Science.gov (United States)

    Zanino, R.; Bagnasco, M.; Fillunger, H.; Heller, R.; Savoldi Richard, L.; Suesser, M.; Zahn, G.

    2004-06-01

    The International Thermonuclear Experimental Reactor (ITER) Toroidal Field Model Coil (TFMC) was tested in the Toska facility of Forschungszentrum Karlsruhe during 2001 (standalone) and 2002 (in the background magnetic field of the LCT coil). The TFMC is a racetrack coil wound in five double pancakes on stainless steel radial plates using Nb3Sn dual-channel cable-in-conduit conductor (CICC) with a thin circular SS jacket. The coil was cooled by supercritical helium in forced convection at nominal 4.5 K and 0.5 MPa. Instrumentation, all outside the coil, included voltage taps, pressure and temperature sensors, as well as flow meters. Additionally, differential pressure drop measurement was available on the two pancakes DP1.1 and DP1.2, equipped with heaters. Two major thermal-hydraulic issues in the TFMC tests will be addressed here: 1) the pressure drop along heated pancakes and the comparison with friction factor correlations; 2) the quench initiation and propagation. Other thermal-hydraulic issues like heat generation and exchange in joints, radial plates, coil case, or the effects of the resistive heaters on the helium dynamics, have been already addressed elsewhere.

  13. 3D thermal-hydraulic analysis of two irregular field joints for the ITER vacuum vessel

    Energy Technology Data Exchange (ETDEWEB)

    Savoldi, Laura, E-mail: laura.savoldi@polito.it [Dipartimento Energia, Politecnico di Torino, I-10129 Torino (Italy); Bonifetto, Roberto [Dipartimento Energia, Politecnico di Torino, I-10129 Torino (Italy); Izquierdo, Jesus [Fusion for Energy, ES-08019 Barcelona (Spain); Le Barbier, Robin; Utin, Yuri [ITER Organization, Route de Vinon sur Verdon, CS 90 046, 13067 Saint Paul Lez Durance Cedex (France); Zanino, Roberto [Dipartimento Energia, Politecnico di Torino, I-10129 Torino (Italy)

    2015-10-15

    Highlights: • ITER vacuum vessel irregular field joints analyzed with ANSYS-FLUENT. • Steady-state thermal-hydraulic performance evaluated with 3D analysis. • Pressure drop comparable to that of regular field joints. • Acceptable hot spots on the inner shell and ports. • Heat transfer coefficient above the target value of 500 W/m{sup 2} K on the inner shell. - Abstract: In ITER, so-called “Irregular” Field Joints (IFJs) are foreseen at the interface between irregular sectors of the Vacuum Vessel (VV), which is located inside the cryostat and houses the in-vessel components. In the IFJs, a peculiar design of the equatorial port, with respect to that adopted in the Regular Field Joints (RFJs), accommodates the irregularities of the adjacent VV sectors. The IFJs are subject to nuclear heating and actively cooled by sub-cooled pressurized water flowing in a dedicated hydraulic loop, which includes the space left open by the borated In-Wall Shielding (IWS). Here we perform the 3D steady state thermal-hydraulic analysis of two different IFJs using the Computational Fluid Dynamics (CFD) software ANSYS-FLUENT{sup ®}. The water flow field, the pressure drop and the temperature maps are computed. The thermal performance of the IFJs in nominal operation is compared to that of an RFJ and it is shown that also in this case enough cooling capability is available to avoid hot spots above the design limits, while the pressure drop remains acceptably low.

  14. Analysis of Thermal-Hydraulic Behavior of CMT in the SMART-ITL Facility

    Energy Technology Data Exchange (ETDEWEB)

    Jeon, Byong Guk; Bae, Hwang; Ryu, Sung-Uk; Ryu, Hyobong; Byun, Sun-Joon; Yi, Sung-Jae; Park, Hyun-Sik [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-10-15

    SMART, an integral small modular reactor, received a standard design approval in 2012 and now extends its safety features through replacing active safety injection pumps by passive safety injection systems: core makeup tanks (CMT) and safety injection tanks (SIT). SMART-ITL has been built in a full height scale and 1/49 area and power scale. One train of CMT and SIT has been installed and their thermal-hydraulic behaviors have been identified through a series of tests. In this paper, initial condensation characteristics as well as force balance around the CMT will be discussed for a representative test. PSIS are added into SMART for better treatment of accidents with prolonged station blackout. In the SMART-ITL, the CMT and SIT are installed to evaluate their performance and a series of tests have been conducted. In this paper, the thermal-hydraulic behavior of CMT is addressed based on the experimental data, especially focusing on the issues of fierce condensation after opening of the isolation valve and driving force balance around the CMT.

  15. Thermal-hydraulic performance analysis for the conceptual design of Korean HCCR TBMset

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Dong Won; Jin, Hyung Gon; Lee, Eo Hwak; Yoon, Jae Sung; Kim, Suk Kwon [KAERI, Daejeon (Korea, Republic of); Shin, Kyu In [Gentec Co., Daejeon (Korea, Republic of); Cho, Seungyon [National Fusion Research Institute, Daejeon (Korea, Republic of)

    2015-05-15

    The purpose of this document is to provide the thermal-hydraulic (TH) analyses results of the Helium Cooled Ceramic Reflector (HCCR) Test Blanket Module (TBM) including TBM-shield, which is called TBM-set. The analyses were performed for Electro- Magnetic Module (EM-TBM) and INTegral Module (INT-TBM) including TBM-shield, respectively, with the same model and meshes according to the ITER operation conditions of H/He and D-T phases, respectively. Thermal-hydraulic performance of the EM- and INTTBM- sets were analysed using the fixed CATIA model for CDR. Fine mesh with 15.9 million elements for solid and 44.7 million elements for fluid was used for ANSYS-CFX 14.5 simulation and coarse mesh with 7.6 million elements for solid is prepared for the thermomechanical analysis. The boundary conditions such as heat flux, nuclear heating, and coolant conditions were determined considering the ITER operation condition and designed cooling scheme. The analysis results and conclusions are as follows; (1) It is confirmed that both EM- and INT-TBM performance results meet the design requirements, which were determined by the material characteristics. (2) The temperature results with fine mesh of both EMand INT-TBM-sets were successfully transferred to those of coarse mesh for the thermo-mechanical analysis.

  16. Numerical Methods for an Analysis of Hydrogen Behaviors Coupled with Thermal Hydraulics in a NPP Containment

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-10-15

    In a containment safety analysis, multi-dimensional characteristics in thermal hydraulics are very important because the flow paths are not confined in a large free volume of the containment. The analysis is difficult because of a difference in length scales between a characteristic length of the flow and representative length of the containment. In order to simulate hydrogen and steam behaviors in a containment during postulated severe accidents, the GASFLOW code as a multi-dimensional analysis tool for NPP containment has been used for years because of its computational efficiency. Though GASFLOW is well developed for a real NPP containment analysis, there exist shortcomings in nodalization, two-phase and turbulence models. It is based on a Cartesian or cylindrical coordinate mesh, so it is impractical to refine a mesh locally in a region with a physical or geometrical complication. In this paper, the importance of the hydrogen safety in an NPP containment and requirements of the analysis tool was described. And physical models necessary for the hydrogen safety analysis code were listed. As a member of international collaborative project HYMERES for containment thermal hydraulics, KAERI is actively participating in an analytic working group. As an analysis tool for blind benchmarkes, the analysis code described in this paper was used. From the blind benchmark analyses, it was found that the code is very promising for hydrogen safety analysis. Currently, it is proposed to develop the code collaboratively in a hydrogen safety community based on an open-source strategy.

  17. Thermal-hydraulic performance of novel louvered fin using flat tube cross-flow heat exchanger

    Institute of Scientific and Technical Information of China (English)

    Junqi DONG; Jiangping CHEN; Zhijiu CHEN

    2008-01-01

    Experimental studies were conducted to investigate the air-side heat transfer and pressure drop characteristics of a novel louvered fins and flat tube heat exchangers. A series of tests were conducted for 9 heat exchangers with different fin space and fin length, at a constant tube-side water flow rate of 2.8 m/h. The air side thermal performance data were analyzed using the effectiveness-NTU method. Results were presented as plot of Colburn j factor and friction factor f against the Reynolds number in the range of 500-6500. The characteristics of the heat transfer and pressure drop of different fin space and fin length were analyzed and compared. In addition, the curves of the heat transfer coefficients vs. pumping power per unit heat transfer area were plotted. Finally, the area optimization factor was used to evaluate the thermal hydraulic performance of the louvered fins with differential geometries. The results showed that the j and ffactors increase with the decrease of the fin space and fin length, and the fin space has more obvious effect on the thermal hydraulic characteristics of the novel louvered fins.

  18. Development of the Real-time Core and Thermal-Hydraulic Models for Kori-1 Simulator

    Energy Technology Data Exchange (ETDEWEB)

    Hong, Jin Hyuk; Lee, Myeong Soo; Hwang, Do Hyun; Byon, Soo Jin [KEPRI, Daejeon (Korea, Republic of)

    2010-10-15

    The operation of the Kori-Unit 1 (1723.5MWt) is expanded to additional 10 years with upgrades of the Main Control Room (MCR). Therefore, the revision of the procedures, performance tests and works related with the exchange of the Main Control Board (MCB) are currently carried out. And as a part of it, the fullscope simulator for the Kori-1 is being developed for the purpose of the pre-operation and emergence response capability for the operators. The purpose of this paper is to report on the performance of the developed neutronics and thermal-hydraulic (TH) models of Kori Unit 1 simulator. The neutronics model is based on the NESTLE code and TH model based on the RELAP5/MOD3 thermal-hydraulics analysis code which was funded as FY-93 LDRD Project 7201 and is running on the commercial simulator environment tool (the 3KeyMaster{sup TM} of the WSC). As some examples for the verification of the developed neutronics and TH models, some figures are provided. The outputs of the developed neutronics and TH models are in accord with the Nuclear Design Report (NDR) and Final Safety Analysis Report (FSAR) of the reference plant

  19. 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.

  20. 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.

  1. 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)

  2. Thermal-hydraulic modeling and analysis of spool valve with sloping U-shape notch by bond graph

    Institute of Scientific and Technical Information of China (English)

    娄磊; 吴万荣; 王兆强; 梁向京

    2015-01-01

    To increase the efficiency and reliability of the thermodynamics analysis of the spool valve, the precise function expression of the flow area for the sloping U-shape notch orifice versus the spool stroke and thermal-hydraulic bond graph based on the conservation of mass and energy were introduced. Subsequently, the connection rule for the bond graph elements and the method to construct the complete thermal-hydraulic system model were proposed. On the basis of heat transfer analysis of a typical hydraulic circuit containing the spool valve, the lumped parameter for mathematical model of the system was given. At last, the reliability of the mathematical model of the flow area and the thermal-hydraulic system for the sloping U-shape notch orifice on the spool were demonstrated by the test. The good agreement between the simulation results and experimental data demonstrates the validity of the modeling method.

  3. 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.

  4. Assessment of the thermal-hydraulic technology of the transition phase of a core-disruptive accident in a LMFBR

    Energy Technology Data Exchange (ETDEWEB)

    Greene, G.A.; Ginsberg, T.; Kazimi, M.S.

    1982-11-01

    The technology of thermal hydraulic aspects of the transition phase accident sequence in liquid metal fast breeder reactors has been reviewed. Previous analyses of the transition phase accident sequence have been reviewed and the current understanding of major thermal hydraulic phenomenology has been assessed. As a result of the foregoing, together with a scoping analysis of the transition phase accident sequence, major transition phase issues have been defined and research needs have been identified. The major conclusion of transition phase scoping analysis is that fuel dispersal cannot be relied upon to rule out the possibility of recriticalities during this stage of the accident.

  5. 热工水力子通道分析程序ATHAS的稳态验证%Steady-state Verification of Thermal-hydraulics Sub-channel Analysis Code ATHAS

    Institute of Scientific and Technical Information of China (English)

    刘伟; 朱元兵; 白宁; 单建强; 张博; 苟军利; 厉井钢

    2014-01-01

    GE3×3 test bundle experiments were simulated with sub-channel analysis code ATHAS.Comparisons of the obtained results by ATHAS code with the experimental measurements and other sub-channel codes show that ATHAS is capable to predict thermal-hydraulic parameters distribution in GE3 ×3 components accurately.All of this demonstrates the reasonable physical models and powerful application functions of ATHAS.The work of this thesis can be taken example by the design and development of thermal-hydraulic program of nuclear power plant in China.%利用具有自主知识产权的子通道程序 ATHAS对 GE3×3组件进行稳态计算,并将 ATHAS的预测值与实验测量值及其他子通道程序的预测值进行了对比分析,结果表明:ATHAS 能够准确预测GE3×3组件内的热工水力参数分布,展示了 ATHAS可靠的物理模型。本文对 ATHAS 进行稳态验证的思路和方法,对我国核电站热工水力软件自主化的设计开发具有借鉴意义。

  6. Advanced neutron source reactor thermal-hydraulic test loop facility description

    Energy Technology Data Exchange (ETDEWEB)

    Felde, D.K.; Farquharson, G.; Hardy, J.H.; King, J.F.; McFee, M.T.; Montgomery, B.H.; Pawel, R.E.; Power, B.H.; Shourbaji, A.A.; Siman-Tov, M.; Wood, R.J.; Yoder, G.L.

    1994-02-01

    The Thermal-Hydraulic Test Loop (THTL) is a facility for experiments constructed to support the development of the Advanced Neutron Source Reactor (ANSR) at Oak Ridge National Laboratory. The ANSR is both cooled and moderated by heavy water and uses uranium silicide fuel. The core is composed of two coaxial fuel-element annuli, each of different diameter. There are 684 parallel aluminum-clad fuel plates (252 in the inner-lower core and 432 in the outer-upper core) arranged in an involute geometry that effectively creates an array of thin rectangular flow channels. Both the fuel plates and the coolant channels are 1.27 mm thick, with a span of 87 mm (lower core), 70 mm (upper core), and 507-mm heated length. The coolant flows vertically upwards at a mass flux of 27 Mg/m{sup 2}s (inlet velocity of 25 m/s) with an inlet temperature of 45{degrees}C and inlet pressure of 3.2 MPa. The average and peak heat fluxes are approximately 6 and 12 MW/m{sup 2}, respectively. The availability of experimental data for both flow excursion (FE) and true critical heat flux (CHF) at the conditions applicable to the ANSR is very limited. The THTL was designed and built to simulate a full-length coolant subchannel of the core, allowing experimental determination of thermal limits under the expected ANSR thermal-hydraulic conditions. For these experimental studies, the involute-shaped fuel plates of the ANSR core with the narrow 1.27-mm flow gap are represented by a narrow rectangular channel. Tests in the THTL will provide both single- and two-phase thermal-hydraulic information. The specific phenomena that are to be examined are (1) single-phase heat-transfer coefficients and friction factors, (2) the point of incipient boiling, (3) nucleate boiling heat-transfer coefficients, (4) two-phase pressure-drop characteristics in the nucleate boiling regime, (5) flow instability limits, and (6) CHF limits.

  7. 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.

  8. Coupling of Thermal-Hydraulic-Mechanical Processes for Geothermal Reservoir Modelling

    Institute of Scientific and Technical Information of China (English)

    Ali Karrechl Oussama Beltaief; Ruyan Vincec; Thomas Poulet; Klaus Regenauer-Lieb

    2015-01-01

    This paper uses a fully coupled framework of thermal-hydraulic-mechanical processes to investigate how the injection and extraction of fluid within a geothermal reservoir impacts on the dis-tributions of temperature, pore pressure, and deformation within the rock formations. Based on this formulation, a numerical model is developed in light of the thermodynamics of porous materials. The proposed procedure relies on the derivation of dissipative flow rules by postulating proper storage and dissipation functions. This approach opens new horizons for several resource engineering applications. Since it allows for full coupling, this formulation can play a key role in predicting risks when used for reservoir simulation. The results indicate that the injection-extraction process and temperature change have a definite impact on altering the in-situ properties of the reservoir.

  9. 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)

  10. Thermal-hydraulic performance of oval tubes in a cross-flow of air

    Science.gov (United States)

    Hasan, Ala

    2005-06-01

    The thermal-hydraulic performance of five oval tubes is experimentally investigated and compared with that for a circular tube in a cross-flow of air. The range of Reynolds numbers ReD is approximately between 1,000 and 11,000. The nominal axis ratios R (major axis/minor axis) for three of the investigated oval tubes are 2, 3, and 4. Two other configurations of oval tubes are also tested, an oval tube R=3 with two wires soldered on its upper and lower top positions, and a cut-oval tube. The performance of the tubes is corrected for the effects of area blockage and turbulence intensity. The measurement results show that the mean Nusselt numbers NuD for the oval tubes are close to that for the circular tube for ReDthermal-hydraulic performance is indicated by the ratio NuD/Cd, which shows a better combined performance for the oval tubes.

  11. A comparison of the CHF between tubes and annuli under PWR thermal-hydraulic conditions

    Energy Technology Data Exchange (ETDEWEB)

    Herer, C. [RRAMATOME EP/TC, Paris (France); Souyri, A. [EdF DER/RNE/TTA, Chatou (France); Garnier, J. [CEA DRN/DTP/STR/LETC, Grenoble (France)

    1995-09-01

    Critical Heat Flux (CHF) tests were carried out in three tubes with inside diameters of 8, 13, and 19.2 mm and in two annuli with an inner tube of 9.5 mm and an outer tube of 13 or 19.2 mm. All axial heat flux distributions in the test sections were uniform. The coolant fluid was Refrigerant 12 (Freon-12) under PWR thermal-hydraulic conditions (equivalent water conditions - Pressure: 7 to 20 MPa, Mass Velocity: 1000 to 6000 kg/m2/s, Local Quality: -75% to +45%). The effect of tube diameter is correlated for qualities under 15%. The change from the tube to the annulus configuration is correctly taken into account by the equivalent hydraulic diameter. Useful information is also provided concerning the effect of a cold wall in an annulus.

  12. Numerical analysis on thermal hydraulic performance of a flat plate heat pipe with wick column

    Science.gov (United States)

    Lu, Longsheng; Liao, Huosheng; Liu, Xiaokang; Tang, Yong

    2015-08-01

    A simplified thermal hydraulic model is developed to investigate the influence of wick column on the performance of a flat plate heat pipe (FPHP). The governing equations of the FPHP are solved by using the computational fluid dynamics package FLUENT. The temperature, velocity and pressure fields are obtained. The validity of the model is confirmed by comparing the present solutions with the open literature data. The numerical results show that with the increase of the wick column size, the maximum velocity of the liquid and vapor decreases while the total thermal resistance and capillary heat transfer limit of the FPHP increases gradually. The performance of the FPHP may degrade if the wick column is placed inside the vapor core asymmetrically.

  13. Thermal-hydraulic characteristics in a tokamak vacuum vessel of fusion reactor after transient events occurred

    Energy Technology Data Exchange (ETDEWEB)

    Takase, Kazuyuki; Kunugi, Tomoaki [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Seki, Yasushi

    1997-12-31

    The thermal-hydraulic characteristics in a vacuum vessel (VV) of fusion reactor under the ingress-of-coolant-event (ICE) or loss-of-vacuum-event (LOVA) condition were carried out to investigate experimentally the thermofluid safety in the International Thermonuclear Experimental Reactor (ITER) under transient events. In the ICE experiments, the pressure rise and wall temperatures in the VV were measured and the performance of a suppression tank was confirmed. In the LOVA experiments, the exchange time inside the VV from the vacuum to be the atmospheric pressure was measured for various breach size and the exchange flow rates through the breaches of the VV under the atmospheric pressure conditions were clarified. (author)

  14. Numerical discretization analysis of a HTR steam generator model for the thermal-hydraulics code trace

    Directory of Open Access Journals (Sweden)

    Esch Markus

    2014-01-01

    Full Text Available For future high temperature reactor projects, e. g., for electricity production or nuclear process heat applications, the steam generator is a crucial component. A typical design is a helical coil steam generator consisting of several tubes connected in parallel forming cylinders of different diameters. This type of steam generator was a significant component used at the thorium high temperature reactor. In the work presented the temperature profile is being analyzed by the nodal thermal hydraulics code TRACE for the thorium high temperature reactor steam generator. The influence of the nodalization is being investigated within the scope of this study and compared to experimental results from the past. The results of the standard TRACE code are compared to results using a modified Nusselt number for the primary side. The implemented heat transfer correlation was developed within the past German HTR program. This study shows that both TRACE versions are stable and provides a discussion of the nodalization requirements.

  15. 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.

  16. 4C code analysis of thermal-hydraulic transients in the KSTAR PF1 superconducting coil

    Science.gov (United States)

    Savoldi Richard, L.; Bonifetto, R.; Chu, Y.; Kholia, A.; Park, S. H.; Lee, H. J.; Zanino, R.

    2013-01-01

    The KSTAR tokamak, in operation since 2008 at the National Fusion Research Institute in Korea, is equipped with a full superconducting magnet system including the central solenoid (CS), which is made of four symmetric pairs of coils PF1L/U-PF4L/U. Each of the CS coils is pancake wound using Nb3Sn cable-in-conduit conductors with a square Incoloy jacket. The coils are cooled with supercritical He in forced circulation at nominal 4.5 K and 5.5 bar inlet conditions. During different test campaigns the measured temperature increase due to AC losses turned out to be higher than expected, which motivates the present study. The 4C code, already validated against and applied to different types of thermal-hydraulic transients in different superconducting coils, is applied here to the thermal-hydraulic analysis of a full set of trapezoidal current pulses in the PF1 coils, with different ramp rates. We find the value of the coupling time constant nτ that best fits, at each current ramp rate, the temperature increase up to the end of the heating at the coil outlet. The agreement between computed results and the whole set of measured data, including temperatures, pressures and mass flow rates, is then shown to be very good both at the inlet and at the outlet of the coil. The nτ values needed to explain the experimental results decrease at increasing current ramp rates, consistently with the results found in the literature.

  17. Thermal Hydraulic Analysis of K-DEMO Single Blanket Module for Preliminary Accident Analysis using MELCOR

    Energy Technology Data Exchange (ETDEWEB)

    Moon, Sung Bo; Bang, In Cheol [UNIST, Ulsan (Korea, Republic of)

    2016-05-15

    To develop the Korean fusion commercial reactor, preliminary design concept for K-DEMO (Korean fusion demonstration reactor) has been announced by NFRI (National Fusion Research Institute). This pre-conceptual study of K-DEMO has been introduced to identify technical details of a fusion power plant for the future commercialization of fusion reactor in Korea. Before this consideration, to build the K-DEMO, accident analysis is essential. Since the Fukushima accident, which is severe accident from unexpected disaster, safety analysis of nuclear power plant has become important. The safety analysis of both fission and fusion reactors is deemed crucial in demonstrating the low radiological effect of these reactors on the environment, during severe accidents. A risk analysis of K-DEMO should be performed, as a prerequisite for the construction of a fusion reactor. In this research, thermal-hydraulic analysis of single blanket module of K-DEMO is conducted for preliminary accident analysis for K-DEMO. Further study about effect of flow distributer is conducted. The normal K-DEMO operation condition is applied to the boundary condition and simulated to verify the material temperature limit using MELCOR. MELCOR is fully integrated, relatively fast-running code developed by Sandia National Laboratories. MELCOR had been used for Light Water Reactors and fusion reactor version of MELCOR was developed for ITER accident analysis. This study shows the result of thermal-hydraulic simulation of single blanket module with MELCOR which is severe accident code for nuclear fusion safety analysis. The difference of mass flow rate for each coolant channel with or without flow distributer is presented. With flow distributer, advantage of broadening temperature gradient in the K-DEMO blanket module and increase mass flow toward first wall is obtained. This can enhance the safety of K-DEMO blanket module. Most 13 .deg. C temperature difference in blanket module is obtained.

  18. FY 1993 progress report on the ANS thermal-hydraulic test loop operation and results

    Energy Technology Data Exchange (ETDEWEB)

    Siman-Tov, M.; Felde, D.K.; Farquharson, G. [and others

    1994-07-01

    The Thermal-Hydraulic Test Loop (THTL) is an experimental facility constructed to support the development of the Advanced Neutron Source Reactor (ANSR) at Oak Ridge National Laboratory (ORNL). Highly subcooled heavy-water coolant flows vertically upward at a very high mass flux of almost 27 MG/m{sup 2}-s. In a parallel fuel plate configuration as in the ANSR, the flow is subject to a potential excursive static-flow instability that can very rapidly lead to flow starvation and departure from nucleate boiling (DNB) in the ``hot channel``. The current correlations and experimental data bases for flow excursion (FE) and critical heat flux (CHF) seldom evaluate the specific combination of ANSR operating parameters. The THTL facility was designed and built to provide known thermal-hydraulic (T/H) conditions for a simulated full-length coolant subchannel of the ANS reactor core, thus facilitating experimental determination of FE and CHF thermal limits under expected ANSR T/H conditions. A series of FE tests with water flowing vertically upward was completed over a nominal heat flux range of 6 to 17 MW/m{sup 2}, a mass flux range of 8 to 28 Mg/m{sup 2}-s, an exit pressure range of 1.4 to 2.1 MPa, and an inlet temperature range of 40 to 50 C. FE experiments were also conducted using as ``soft`` a system as possible to secure a true FE phenomena (actual secondary burnout). True DNB experiments under similar conditions were also conducted. To the author`s knowledge, no other FE data have been reported in the literature to date that dover such a combination of conditions of high mass flux, high heat flux, and moderately high pressure.

  19. Neutron Tomography Using Mobile Neutron Generators for Assessment of Void Distributions in Thermal Hydraulic Test Loops

    Science.gov (United States)

    Andersson, P.; Bjelkenstedt, T.; Sundén, E. Andersson; Sjöstrand, H.; Jacobsson-Svärd, S.

    Detailed knowledge of the lateral distribution of steam (void) and water in a nuclear fuel assembly is of great value for nuclear reactor operators and fuel manufacturers, with consequences for both reactor safety and economy of operation. Therefore, nuclear relevant two-phase flows are being studied at dedicated thermal-hydraulic test loop, using two-phase flow systems ranging from simplified geometries such as heated circular pipes to full scale mock-ups of nuclear fuel assemblies. Neutron tomography (NT) has been suggested for assessment of the lateral distribution of steam and water in such test loops, motivated by a good ability of neutrons to penetrate the metallic structures of metal pipes and nuclear fuel rod mock-ups, as compared to e.g. conventional X-rays, while the liquid water simultaneously gives comparatively good contrast. However, these stationary test loops require the measurement setup to be mobile, which is often not the case for NT setups. Here, it is acknowledged that fast neutrons of 14 MeV from mobile neutron generators constitute a viable option for a mobile NT system. We present details of the development of neutron tomography for this purpose at the division of Applied Nuclear Physics at Uppsala University. Our concept contains a portable neutron generator, exploiting the fusion reaction of deuterium and tritium, and a detector with plastic scintillator elements designed to achieveadequate spatial and energy resolution, all mounted in a light-weight frame without collimators or bulky moderation to allow for a mobile instrument that can be moved about the stationary thermal hydraulic test sections. The detector system stores event-to-event pulse-height information to allow for discrimination based on the energy deposition in the scintillator elements.

  20. Thermal hydraulic characteristics during ingress of coolant and loss of vacuum events in fusion reactors

    Science.gov (United States)

    Takase, K.; Kunugi, T.; Seki, Y.; Akimoto, H.

    2000-03-01

    The thermal hydraulic characteristics in the vacuum vessel (VV) of a fusion reactor under an ingress of coolant event (ICE) and a loss of vacuum event (LOVA) were investigated quantitatively using preliminary experimental apparatuses. In the ICE experiments, pressure rise characteristics in the VV were clarified for experimental parameters of the wall temperature and water temperature and for cases with and without a blowdown tank. In addition, the functional performance of a blowdown tank with and without a water cooling system was examined and it was confirmed that the blowdown tank with a water cooling system is effective for suppressing the pressure rise during the ICE. In the LOVA experiments, the saturation time in the VV from vacuum to atmosphere was investigated for various breach sizes and it was found that the saturation time is in inverse proportion to the breach size. In addition, the characteristics of exchange flow through breaches were clarified for the different breach positions on the VV. It was proven from the experimental results that the exchange flow became a counter-current flow when the breach was positioned on the top of the VV and a stratified flow when it was formed on the side wall of the VV, and that the exchange flow under the stratified flow condition was smoother than that of counter-current flow. On the basis of these results, the severest breach condition in ITER was changed from the top-break case to the side-break case. To predict with high accuracy the thermal hydraulic characteristics during ICEs and LOVAs under ITER conditions, a large scale test facility will be necessary. The current conceptual design of the combined ICE-LOVA test facility with a scaling factor of 1/1000 in comparison with the ITER volume is presented.

  1. 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.

  2. 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

    Science.gov (United States)

    Mushtaq, Ahmad; Iqbal, Masood; Bokhari, Ishtiaq Hussain; Mahmood, Tayyab; Muhammad, Atta

    2012-09-01

    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% 235U) 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; uranium is needed. LEU aluminum uranium dispersion target has been developed, which may replace existing HEU aluminum/uranium alloy targets for production of 99Mo using basic dissolution technique. Neutronic and thermal hydraulic calculations were performed for safe irradiation of targets in the core of PARR-1.

  3. Evaluation of TASS/SMR with steady state analysis of high temperature/high pressure thermal-hydraulic test facility (VISTA)

    Energy Technology Data Exchange (ETDEWEB)

    Jang, Dong Ju; Choi, Yong Won; Park, Chang Hwan; Lee, Un Chul [Seoul National University, Seoul (Korea, Republic of); Hwang, Young Dong; Lee, Kyu Hyung; Kim, Hee Chul [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    2005-07-01

    The TASS/SMR code is the revised version of the TASS code, which is the result of code development effort of KAERI since 1997, for safety analysis of NPP coolant system. Lately, it is scheduled to evaluate thermal-hydraulic phenomena during several transient periods of SMART-P with TASS/SMR. To establish the pertinence of the calculative results of TASS/SMR, there should be a process of validation and verification of TASS/SMR. The objective of this study is validating the numerical capability and reliability of TASS/SMR with steady state analysis of VISTA (Experimental Verification by Integral Simulation of Transient and Accidents) that was designed to simulate SMART-P.

  4. Analysing the scalability of thermal hydraulic test facility data to reactor scale with a computer code; Vertailuanalyysin kaeyttoe termohydraulisten koelaitteistojen tulosten laitosmittakaavaan skaalautumisen tutkimisessa

    Energy Technology Data Exchange (ETDEWEB)

    Suikkanen, P.

    2009-01-15

    The objective of the Masters thesis was to study guidelines and procedures for scaling of thermal hydraulic test facilities and to compare results from two test facility models and from EPR model. Aim was to get an impression of how well the studied test facilities describe the behaviour in power plant scale during accident scenarios with computer codes. Models were used to determine the influence of primary circuit mass inventory on the behaviour of the circuit. The data from test facility models represent the same phenomena as the data from EPR model. The results calculated with PKL model were also compared against PKL test facility data. They showed good agreement. Test facility data is used to validate computer codes, which are used in nuclear safety analysis. The scale of the facility has effect on the behaviour of the phenomena and therefore special care must be taken in using the data. (orig.)

  5. 75 FR 69140 - NUREG-1953, Confirmatory Thermal-Hydraulic Analysis To Support Specific Success Criteria in the...

    Science.gov (United States)

    2010-11-10

    ... COMMISSION NUREG-1953, Confirmatory Thermal-Hydraulic Analysis To Support Specific Success Criteria in the... Regulatory Commission has issued for public comment a document entitled: NUREG-1953, ``Confirmatory Thermal...-4209, 301-415-4737, or by e-mail to pdr.resource@nrc.gov . NUREG-1953 is available electronically...

  6. 75 FR 80544 - NUREG-1953, Confirmatory Thermal-Hydraulic Analysis To Support Specific Success Criteria in the...

    Science.gov (United States)

    2010-12-22

    ... COMMISSION NUREG-1953, Confirmatory Thermal-Hydraulic Analysis To Support Specific Success Criteria in the... Regulatory Commission is re-opening the public comment period for the document entitled: NUREG-1953....gov . NUREG-1953 is available electronically under ADAMS Accession Number ML102940233....

  7. Solving thermal-hydraulic tasks in the context of structure-mechanical analysis; Loesung thermohydraulischer Aufgaben im Rahmen strukturmechanischer Analysen

    Energy Technology Data Exchange (ETDEWEB)

    Hermsmeyer, S.

    1999-06-01

    The thermomechanical analysis of fluid-cooled structures depends critically on local coolant temperatures. This is particularly true for transient analyses of fusion reactor blankets that are exposed to large radial and temporal power gradients in power cycling reactors and see large thermal-mechanical loads. This report is concerned with the issue of thermal-hydraulic modelling and analysis that is an integral part of structure-mechanical analyses yet has to be treated separately because of differing needs regarding the finite element code and structure discretisation. This report presents a simplified thermal no-momentum fluid model that poses acceptable icomputational cost even for extended and branched cooling systems. The implementation of this model in the finite element codes FIDAP and ABAQUS is described. A comparison of the codes finds advantages for thermal-hydraulic modelling in FIDAP, stressing however, that unified computer-aided-design-based grid generation would be desirable. Two examples serve the purpose of demonstrating the methodology of the thermal-hydraulic analysis. Key parts of this methodology are the use of symmetry conditions when modelling a representative blanket section, the partitioning into submodels, the simplification of geometrical model features and the feeding thermal-hydraulic results into the structure-mechanic analysis. Part of the appendix is a description and manual for a computer code that has been written to simplify the design of three-dimensional FIDAP models. The code automates the successive rotation and/or translation of surfaces defined in FIDAP.

  8. Analytical Algorithms for the Blend Ratios by Fibre-bundle Tensile Curves Part Ⅲ:The Definition Method and the Relationships between the Blend Ratio and Tensile Parameters Calculated from Different Ways

    Institute of Scientific and Technical Information of China (English)

    YU Wei-dong; Ron Postle; YAN Hao-jing

    2004-01-01

    The algorithms of the strength, work, modulus and percentage methods have been introduced and discussed in Part Ⅰ and Part Ⅱ of the series papers. As the following, the percentage-function calculation and the frequency-function calculation of the definition method are defined and developed in detail. Meanwhile, the procedures to find the tensile parameters are analysed experimentally and involve "ED" being the average values of the corresponding tensile parameters of each bundle tensile curve measured; "EC" being the tensile parameters obtained from the averaged experimental tensile curve; and "TC" being the parameters of the theoretical tensile curve according to the corresponding nominal blend ratio. The theoretical and experimental results indicate that the blend ratio estimated by the definition method from wool/polyester blended fibre bundles has a high correlation with the actual blend ratio though the variables to be known in this method are the most of all the methods and there exist the difficulty to solve fibre bundle tensile behaviour without fibre broken in accuracy and that there are high correlations between the EC and TC data and the tensile parameters should be calculated using the EC method rather than the ED method with the exception of the bundle initial modulus. The relationships between blend ratios and the tensile properties of blended fibre bundles can be used to select the optimum blend ratio of blended tops or yarns.

  9. RDS - A systematic approach towards system thermal hydraulics input code development for a comprehensive deterministic safety analysis

    Energy Technology Data Exchange (ETDEWEB)

    Salim, Mohd Faiz, E-mail: mohdfaizs@tnb.com.my [Nuclear Energy Department, Tenaga Nasional Berhad, Level 32, Dua Sentral, 50470 Kuala Lumpur (Malaysia); Roslan, Ridha [Nuclear Installation Division, Atomic Energy Licensing Board, Batu 24, Jalan Dengkil, 43800 Dengkil, Selangor (Malaysia); Ibrahim, Mohd Rizal Mamat [Technical Support Division, Malaysian Nuclear Agency, Bangi, 43000 Kajang, Selangor (Malaysia)

    2014-02-12

    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.

  10. 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.

  11. E-SCAPE: A scale facility for liquid-metal, pool-type reactor thermal hydraulic investigations

    Energy Technology Data Exchange (ETDEWEB)

    Van Tichelen, Katrien, E-mail: kvtichel@sckcen.be [SCK-CEN, Boeretang 200, 2400 Mol (Belgium); Mirelli, Fabio, E-mail: fmirelli@sckcen.be [SCK-CEN, Boeretang 200, 2400 Mol (Belgium); Greco, Matteo, E-mail: mgreco@sckcen.be [SCK-CEN, Boeretang 200, 2400 Mol (Belgium); Viviani, Giorgia, E-mail: giorgiaviviani@gmail.com [University of Pisa, Lungarno Pacinotti 43, 56126 Pisa (Italy)

    2015-08-15

    Highlights: • The E-SCAPE facility is a thermal hydraulic scale model of the MYRRHA fast reactor. • The focus is on mixing and stratification in liquid-metal pool-type reactors. • Forced convection, natural convection and the transition are investigated. • Extensive instrumentation allows validation of computational models. • System thermal hydraulic and CFD models have been used for facility design. - Abstract: MYRRHA (Multi-purpose hYbrid Research Reactor for High-tech Applications) is a flexible fast-spectrum research reactor under design at SCK·CEN. MYRRHA is a pool-type reactor with lead bismuth eutectic (LBE) as primary coolant. The proper understanding of the thermal hydraulic phenomena occurring in the reactor pool is an important issue in the design and licensing of the MYRRHA system and liquid-metal cooled reactors by extension. Model experiments are necessary for understanding the physics, for validating experimental tools and to qualify the design for the licensing. The E-SCAPE (European SCAled Pool Experiment) facility at SCK·CEN is a thermal hydraulic 1/6-scale model of the MYRRHA reactor, with an electrical core simulator, cooled by LBE. It provides experimental feedback to the designers on the forced and natural circulation flow patterns. Moreover, it enables to validate the computational methods for their use with LBE. The paper will elaborate on the design of the E-SCAPE facility and its main parameters. Also the experimental matrix and the pre-test analysis using computational fluid dynamics (CFD) and system thermal hydraulics codes will be described.

  12. 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

  13. Research and development program for PWR safety at the CEA reactor thermal hydraulics laboratories

    Energy Technology Data Exchange (ETDEWEB)

    Bernard, M. [CEA, Grenoble (France)

    1995-04-15

    Since the start of the French electronuclear program, the three partners Fermate, EDF and Cea (DRN and IPSN) have devoted considerable effort to research and development for safety issues. In particular an important program on thermal hydraulics was initiated at the beginning of the seventies. It is illustrated by the development of the CATHARE thermalhydraulic safety code which includes physical models derived from a large experimental support program and the construction of the BETHSY integral facility which is aimed to assess both the CATHARE code and the physical relevance of the accident management procedures to be applied on reactors. The state of the art on this program is described with particular emphasis on the capabilities and the assessment of the last version of CATHARE and the lessons drawn from 50 BETHSY tests performed so far. The future plans for safety research cover the following strategy: - to solve the few problems identified on present computing tools and extend the assessment - to solve the few problems identified on present computing tools and extend the assessment - to perform safety studies on the basis of plant operation feedback - to contribute to treating the safety issues related to the future reactors and in particular the case of severe accidents which have to be taken into account from the design stage. The program on severe accidents is aimed to support the design studies performed by the industrial partners and to provide computing tools which model the various phases of severe accidents and will be validated on experiments performed with real and simulating materials. The main lines of the program are: - the development of the TOLBIAC 3D code for the thermal hydraulics of core melt pools, which will be validated against the Bali experiment presently under construction - the Sultan experiment, to study the capability of cooling by external flooding of the reactor vessel - the development of the MC-3D code for core melt

  14. Proceedings of the 7th International Meeting on Nuclear Reactor Thermal-Hydraulics NURETH-7. Volume 1, Sessions 1-5

    Energy Technology Data Exchange (ETDEWEB)

    Block, R.C.; Feiner, F. [comps.] [American Nuclear Society, La Grange Park, IL (United States)

    1995-09-01

    This document, Volume 1, includes papers presented at the 7th International Meeting on Nuclear Reactor Thermal-Hydraulics (NURETH-7) September 10--15, 1995 at Saratoga Springs, N.Y. The following subjects are discussed: Progress in analytical and experimental work on the fundamentals of nuclear thermal-hydraulics, the development of advanced mathematical and numerical methods, and the application of advancements in the field in the development of novel reactor concepts. Also combined issues of thermal-hydraulics and reactor/power-plant safety, core neutronics and/or radiation. Selected papers are indexed separately for inclusion in the Energy Science and Technology Database.

  15. Thermal-hydraulic characteristics of a Westinghouse Model 51 steam generator. Volume 2. Appendix A, numerical results. Interim report. [CALIPSOS code numerical data

    Energy Technology Data Exchange (ETDEWEB)

    Fanselau, R.W.; Thakkar, J.G.; Hiestand, J.W.; Cassell, D.

    1981-03-01

    The Comparative Thermal-Hydraulic Evaluation of Steam Generators program represents an analytical investigation of the thermal-hydraulic characteristics of four PWR steam generators. The analytical tool utilized in this investigation is the CALIPSOS code, a three-dimensional flow distribution code. This report presents the steady state thermal-hydraulic characteristics on the secondary side of a Westinghouse Model 51 steam generator. Details of the CALIPSOS model with accompanying assumptions, operating parameters, and transport correlations are identified. Comprehensive graphical and numerical results are presented to facilitate the desired comparison with other steam generators analyzed by the same flow distribution code.

  16. 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.

  17. Incorporating Artificial Neural Networks in the dynamic thermal-hydraulic model of a controlled cryogenic circuit

    Science.gov (United States)

    Carli, S.; Bonifetto, R.; Savoldi, L.; Zanino, R.

    2015-09-01

    A model based on Artificial Neural Networks (ANNs) is developed for the heated line portion of a cryogenic circuit, where supercritical helium (SHe) flows and that also includes a cold circulator, valves, pipes/cryolines and heat exchangers between the main loop and a saturated liquid helium (LHe) bath. The heated line mimics the heat load coming from the superconducting magnets to their cryogenic cooling circuits during the operation of a tokamak fusion reactor. An ANN is trained, using the output from simulations of the circuit performed with the 4C thermal-hydraulic (TH) code, to reproduce the dynamic behavior of the heated line, including for the first time also scenarios where different types of controls act on the circuit. The ANN is then implemented in the 4C circuit model as a new component, which substitutes the original 4C heated line model. For different operational scenarios and control strategies, a good agreement is shown between the simplified ANN model results and the original 4C results, as well as with experimental data from the HELIOS facility confirming the suitability of this new approach which, extended to an entire magnet systems, can lead to real-time control of the cooling loops and fast assessment of control strategies for heat load smoothing to the cryoplant.

  18. Thermal hydraulics in the hot pool of Fast Breeder Test Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Padmakumar, G. [Fast Reactor Technology Group, Indira Gandhi Centre for Atomic Research, Kalpakkam, TN 603 102 (India)], E-mail: gpk@igcar.gov.in; Pandey, G.K.; Vaidyanathan, G. [Fast Reactor Technology Group, Indira Gandhi Centre for Atomic Research, Kalpakkam, TN 603 102 (India)

    2009-06-15

    Sodium cooled Fast Breeder Test Reactor (FBTR) of 40 MWt/13 MWe capacity is in operation at Kalpakkam, near Chennai. Presently it is operating with a core of 10.5 MWt. Knowledge of temperatures and flow pattern in the hot pool of FBTR is essential to assess the thermal stresses in the hot pool. While theoretical analysis of the hot pool has been conducted by a three-dimensional code to access the temperature profile, it involves tuning due to complex geometry, thermal stresses and vibration. With this in view, an experimental model was fabricated in 1/4 scale using acrylic material and tests were conducted in water. Initially hydraulic studies were conducted with ambient water maintaining Froude number similarity. After that thermal studies were conducted using hot and cold water maintaining Richardson similitude. In both cases Euler similarity was also maintained. Studies were conducted simulating both low and full power operating conditions. This paper discusses the model simulation, similarity criteria, the various thermal hydraulic studies that were carried out, the results obtained and the comparison with the prototype measurements.

  19. CFD analysis of the ITER first wall 06 panel. Part II: Thermal-hydraulics

    Energy Technology Data Exchange (ETDEWEB)

    Zanino, R.; Bonifetto, R. [Dipartimento Energia, Politecnico di Torino, 10129 Torino (Italy); Cau, F.; Portone, A. [Fusion for Energy, 08019 Barcelona (Spain); Savoldi Richard, L., E-mail: laura.savoldi@polito.it [Dipartimento Energia, Politecnico di Torino, 10129 Torino (Italy)

    2014-04-15

    The computational fluid dynamics (CFD) analysis of the FW06 panel of the ITER shielding blanket is presented in two companion papers. In this Part II we concentrate on the thermal-hydraulics of the water coolant, driven by the nuclear volumetric and plasma surface heat loads discussed in Part I. Both the detailed steady state analysis of a single cooling channel and the coarse transient analysis of the whole panel are considered. The compatibility of the hot spots with the maximum recommended temperatures for the different materials is confirmed. The heat transfer coefficient between coolant and walls is obtained post-processing the results of the simulation and compared with the results of available correlations, which may be used for simpler analyses: in the fully developed flow regions of the cooling pipes, it turns out to be well approximated by the Sieder–Tate correlation. The operation margin with respect to the critical heat flux is also computed and turns out to be sufficiently large compared with the design limit.

  20. Thermal hydraulic investigations of primary coolant pipe rupture in an LMFBR

    Energy Technology Data Exchange (ETDEWEB)

    Natesan, K. [Reactor Engineering Group, Indira Gandhi Centre for Atomic Research, Kalpakkam 603 102 (India)]. E-mail: natesan@igcar.ernet.in; Kasinathan, N. [Reactor Engineering Group, Indira Gandhi Centre for Atomic Research, Kalpakkam 603 102 (India); Velusamy, K. [Reactor Engineering Group, Indira Gandhi Centre for Atomic Research, Kalpakkam 603 102 (India); Selvaraj, P. [Reactor Engineering Group, Indira Gandhi Centre for Atomic Research, Kalpakkam 603 102 (India); Chellapandi, P. [Reactor Engineering Group, Indira Gandhi Centre for Atomic Research, Kalpakkam 603 102 (India); Chetal, S.C. [Reactor Engineering Group, Indira Gandhi Centre for Atomic Research, Kalpakkam 603 102 (India)

    2006-06-15

    High quality for primary coolant pipes in fast reactors is ensured through utmost care taken in the design and manufacture. Demonstration of high structural reliability of them by extensive experimental and theoretical studies renders the double-ended guillotine rupture (DEGR) of a primary pipe a highly improbable event. However, as a defense in depth approach instantaneous DEGR of one of the pipes has been considered in design. Thermal hydraulic analyses of this event in a typical liquid metal cooled fast breeder have been carried out to study its consequences and to establish the availability of safety margins. Various uncertainties relevant to the event have been analysed to evaluate the sensitivity of each parameter. For this purpose, one-dimensional plant dynamics studies using thermal and hydraulic models of core subassemblies and primary sodium circuit have been performed. Validity of the assumptions made in the one-dimensional model like, uniform flow through all subassemblies in core under pipe ruptured condition and non possibility of sodium boiling by flashing have also been investigated through detailed three-dimensional and pressure transient studies. Analyses indicate the availability of good margins against the design safety limits in all the parametric cases analysed.

  1. Thermal hydraulic studies of lead–bismuth eutectic spallation target of CIADS

    Energy Technology Data Exchange (ETDEWEB)

    Chen, Kang [Institute of Modern Physics, Chinese Academy of Sciences, 509 Nanchang Rd., Lanzhou, Gansu 730000 (China); University of Chinese Academy of Sciences, 19 A Yuquan Rd, Shijingshan District, Beijing 100049 (China); Yang, Yongwei, E-mail: yangyongwei@impcas.ac.cn [Institute of Modern Physics, Chinese Academy of Sciences, 509 Nanchang Rd., Lanzhou, Gansu 730000 (China); Fan, Deliang; Gao, Yucui [Institute of Modern Physics, Chinese Academy of Sciences, 509 Nanchang Rd., Lanzhou, Gansu 730000 (China)

    2016-08-15

    Highlights: • A 3-D fluid-solid coupling thermal-hydraulic analysis is made for the LBE target. • The reactor and the spallation target is coupled in thermal process. • The heat transfer between the inlet and outlet of the spallation target is taken into account. - Abstract: For the China Initiative Accelerator Driven System (CIADS), it includes three sub-systems: accelerator, spallation target, and sub-critical reactor. The sub-system of spallation target is an important component of the CIADS, which is coupled with the other two sub-systems. The proton beam from the accelerator with an energy of 250 MeV and a current intensity of 2 mA reacts with the nuclei of the lead–bismuth eutectic (LBE), approximately 0.5 MW of heat is deposited in the target zone, which must be removed by circulating the LBE. To reach the goal, we carried out the study by using the Computational Fluid Dynamics (CFD) software FLUENT, to study the flow patterns and temperature distribution in the target zone. For these simulations, the heat transferred from the sub-critical reactor was taken into account. The results indicated that the heat deposited in the target zone can be removed safely.

  2. Thermal-hydraulics and safety analysis of sectored compact reactor for lunar surface power

    Energy Technology Data Exchange (ETDEWEB)

    Schriener, T. M. [Inst. for Space and Nuclear Power Studies, Univ. of New Mexico, Albuquerque, NM (United States); Chemical and Nuclear Engineering Dept., Univ. of New Mexico, Albuquerque, NM (United States); El-Genk, M. S. [Inst. for Space and Nuclear Power Studies, Univ. of New Mexico, Albuquerque, NM (United States); Chemical and Nuclear Engineering Dept., Univ. of New Mexico, Albuquerque, NM (United States); Mechanical Engineering Dept., Univ. of New Mexico, Albuquerque, NM (United States)

    2012-07-01

    The liquid NaK-cooled, fast-neutron spectrum, Sectored Compact Reactor (SCoRe-N 5) concept has been developed at the Univ. of New Mexico for lunar surface power applications. It is loaded with highly enriched UN fuel pins in a triangular lattice, and nominally operates at exit and inlet coolant temperatures of 850 K and 900 K. This long-life reactor generates up to 1 MWth continuously for {>=} 20 years. To avoid a single point failure in reactor cooling, the core is divided into 6 sectors that are neutronically and thermally coupled, but hydraulically independent. This paper performs a 3-D the thermal-hydraulic analysis of SCoRe--N 5 at nominal operation temperatures and a power level of 1 MWth. In addition, the paper investigates the potential of continuing reactor operation at a lower power in the unlikely event that one sector in the core experiences a loss of coolant (LOC). Redesigning the core with a contiguous steel matrix enhances the cooling of the sector experiencing a LOC. Results show that with a core sector experiencing a LOC, SCORE-N 5 could continue operating safely at a reduced power of 166.6 kWth. (authors)

  3. An anisotropic numerical model for thermal hydraulic analyses: application to liquid metal flow in fuel assemblies

    Science.gov (United States)

    Vitillo, F.; Vitale Di Maio, D.; Galati, C.; Caruso, G.

    2015-11-01

    A CFD analysis has been carried out to study the thermal-hydraulic behavior of liquid metal coolant in a fuel assembly of triangular lattice. In order to obtain fast and accurate results, the isotropic two-equation RANS approach is often used in nuclear engineering applications. A different approach is provided by Non-Linear Eddy Viscosity Models (NLEVM), which try to take into account anisotropic effects by a nonlinear formulation of the Reynolds stress tensor. This approach is very promising, as it results in a very good numerical behavior and in a potentially better fluid flow description than classical isotropic models. An Anisotropic Shear Stress Transport (ASST) model, implemented into a commercial software, has been applied in previous studies, showing very trustful results for a large variety of flows and applications. In the paper, the ASST model has been used to perform an analysis of the fluid flow inside the fuel assembly of the ALFRED lead cooled fast reactor. Then, a comparison between the results of wall-resolved conjugated heat transfer computations and the results of a decoupled analysis using a suitable thermal wall-function previously implemented into the solver has been performed and presented.

  4. 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.

  5. 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.

  6. CFD investigation the thermal-hydraulic behavior behind the flow blockage in SFR

    Energy Technology Data Exchange (ETDEWEB)

    Yoo, Jin; Jeong, J. H.; Chang, W. P; Ha, K. S. [KAERI, Daejeon (Korea, Republic of)

    2015-05-15

    It is evident that the flow blockage is basically a local phenomenon, and the main issue to investigate is the thermal-hydraulic behavior of the region downstream from the obstacle because it determines the clad temperature peak. For this reason, a local detailed CFD analysis has been carried out in order to assess the impacts of a flow blockage. The flow blockage events are classified into two types, internal and external blockage, depending on their locations. The objective of this paper is to investigate the influence caused by a flow blockage. A CFD analysis using fully resolved RANS simulations has been carried on the fluid flow and heat transfer in the case of a flow blockage for fuel assemblies in a PGSFR. A fuel assembly with 91 pins instead of all 217 pins was considered for this study. Two main effects can be distinguished in a flow blockage: a locally lower mass flow rate in the wake/recirculation region downstream of the blockage, and the peak temperature behind the blockage. Both of them are closely related. The recirculation region exists within a short distance downstream from the blockage, and it has an effect on the cladding integrity. The maximum cladding temperature is about 1000 .deg. C and is located in the central pins of the blockage region. It could lead to a rupture of the cladding. From these analysis results, the axial blockage size may have a significant impact on the clad integrity.

  7. 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.

  8. Preliminary Thermal Hydraulic Analyses of the Conceptual Core Models with Tubular Type Fuel Assemblies

    Energy Technology Data Exchange (ETDEWEB)

    Chae, Hee Taek; Park, Jong Hark; Park, Cheol

    2006-11-15

    A new research reactor (AHR, Advanced HANARO Reactor) based on the HANARO has being conceptually developed for the future needs of research reactors. A tubular type fuel was considered as one of the fuel options of the AHR. A tubular type fuel assembly has several curved fuel plates arranged with a constant small gap to build up cooling channels, which is very similar to an annulus pipe with many layers. This report presents the preliminary analysis of thermal hydraulic characteristics and safety margins for three conceptual core models using tubular fuel assemblies. Four design criteria, which are the fuel temperature, ONB (Onset of Nucleate Boiling) margin, minimum DNBR (Departure from Nucleate Boiling Ratio) and OFIR (Onset of Flow Instability Ratio), were investigated along with various core flow velocities in the normal operating conditions. And the primary coolant flow rate based a conceptual core model was suggested as a design information for the process design of the primary cooling system. The computational fluid dynamics analysis was also carried out to evaluate the coolant velocity distributions between tubular channels and the pressure drop characteristics of the tubular fuel assembly.

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

    Energy Technology Data Exchange (ETDEWEB)

    Peltonen, Joanna

    2009-09-15

    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 fine grid NK domain is necessary. However, improper mappings may result in loss of valuable information, thus causing inaccurate prediction of safety parameters. The purpose of this thesis is to study the sensitivity of spatial coupling (channel refinement and spatial mapping) and develop recommendations for NK-TH mapping in simulation of safety transients - Control Rod Drop, Turbine Trip, Feedwater Transient combined with stability performance (minimum pump speed of recirculation pumps). The research methodology consists of spatial coupling convergence study, as increasing number of TH channels and different mapping approach the reference case. The reference case consists of one TH channel per one fuel assembly. The comparison of results has been done under steady-state and transient conditions

  10. Development and validation of the ASTEC-Na thermal-hydraulic models

    Energy Technology Data Exchange (ETDEWEB)

    Herranz, L. E.; Perez, S.; Bandini, G.; Jacq, F.; Parisi, C.; Berna, C.

    2014-07-01

    Last years the interest in sodium-cooled fast reactors (SFR) has been fostered worldwide by the search for higher nuclear energy sustainability. This has been reflected in the various international initiatives like GEN-IV International Forum, INPRO or ESNII platforms. At the same time, innovative nuclear reactor designs, particularly SFR, are aiming at even higher safety standards than current LWRs. A proof of it is the consideration of severe accidents since the earliest stages of reactor design. commonalities of LWR and SFR severe accident scenarios suggest that some of the knowledge achieved in the LWR arena might be applicable to some extent to SFRs. This is the spirit underneath of the EU-JASMIN project, which generic goal is developing the ASTEC-Na code from the LWR ASTEC platform. This will entail to t extend and adapt some existing models as well as to implement new ones in all the areas covered, from neutronics and pin thermo-mechanics and pin thermo-mechanics to the in-containment source term behavior by these, going through the indispensable Na thermal-hydraulics. (Author)

  11. Preliminary Design of Large Scale Sodium Thermal-Hydraulic Test Facility

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Tae Ho; Kim, Tae Joon; Eoh, Jae Hyuk; Lee, Hyeong Yeon; Lee, Jae Han; Jeong, Ji Young; Park, Su Ki; Han, Ji Woong; Yoo, Yong Hwan; Lee, Yong Bum [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2009-10-15

    A large scale sodium thermal-hydraulic test facility is being designed for verification of the advanced design concept of the passive decay heat removal circuit (PDRC) in a medium- or large-sized pool-type SFR. In the test, its cooling capability during the long- and short-term periods after the reactor trip will be evaluated, and also the produced experimental data will be utilized for the assessment and verification of the safety and performance analysis codes. Starting with the preliminary design of the test facility this year using KALIMER-600 as a reference reactor, the basic and the detailed designs will be made through 2011-2012 based on the demonstration reactor which is intended to be constructed by 2028 according to a long-term national SFR development plan. The installation is scheduled to be completed by the end of 2013, and the main experiments will commence from 2015 after the startup test in 2014. This paper briefly introduces the preliminary design features which were produced as a first step to assess the appropriateness of the facility design methodology.

  12. Use of laser flow visualization techniques in reactor component thermal-hydraulic studies

    Energy Technology Data Exchange (ETDEWEB)

    Oras, J.J.; Kasza, K.E.

    1984-01-01

    To properly design reactor components, an understanding of the various thermal hydraulic phenomena, i.e., thermal stratification flow channeling, recirculation regions, shear layers, etc., is necessary. In the liquid metal breeder reactor program, water is commonly used to replace sodium in experimental testing to facilitate the investigations, (i.e., reduce cost and allow fluid velocity measurement or flow pattern study). After water testing, limited sodium tests can be conducted to validate the extrapolation of the water results to sodium. This paper describes a novel laser flow visualization technique being utilized at ANL together with various examples of its use and plans for further development. A 3-watt argon-ion laser, in conjunction with a cylindrical opticallens, has been used to create a thin (approx. 1-mm) intense plane of laser light for the illuminiation of various flow tracers in precisely defined regions of interest within a test article having windows. Both fluorescing dyes tuned to the wavelength of the laser light (to maximize brightness and sharpness of flow image) and small (< 0.038-mm, 0.0015-in. dia.) opaque, nearly neutrally buoyant polystyrene spheres (to ensure that the particles trace out the fluid motion) have been used as flow tracers.

  13. 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)

  14. Development of An Automatic Verification Program for Thermal-hydraulic System Codes

    Energy Technology Data Exchange (ETDEWEB)

    Lee, J. Y.; Ahn, K. T.; Ko, S. H.; Kim, Y. S.; Kim, D. W. [Pusan National University, Busan (Korea, Republic of); Suh, J. S.; Cho, Y. S.; Jeong, J. J. [System Engineering and Technology Co., Daejeon (Korea, Republic of)

    2012-05-15

    As a project activity of the capstone design competitive exhibition, supported by the Education Center for Green Industry-friendly Fusion Technology (GIFT), we have developed a computer program which can automatically perform non-regression test, which is needed repeatedly during a developmental process of a thermal-hydraulic system code, such as the SPACE code. A non-regression test (NRT) is an approach to software testing. The purpose of the non-regression testing is to verify whether, after updating a given software application (in this case, the code), previous software functions have not been compromised. The goal is to prevent software regression, whereby adding new features results in software bugs. As the NRT is performed repeatedly, a lot of time and human resources will be needed during the development period of a code. It may cause development period delay. To reduce the cost and the human resources and to prevent wasting time, non-regression tests need to be automatized. As a tool to develop an automatic verification program, we have used Visual Basic for Application (VBA). VBA is an implementation of Microsoft's event-driven programming language Visual Basic 6 and its associated integrated development environment, which are built into most Microsoft Office applications (In this case, Excel)

  15. Numerical Study of Thermal Hydraulic behavior of Pressurizer for PLCS Scenario by CUPID Code

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Jae Ryong; Yoon, Han Young [KAERI, Daejeon (Korea, Republic of); Yoon, Bo Kam; Kim, Jeong Ju; Park, Jong Cheol; Lee, Gyu Cheon [KEPCO, Daejeon (Korea, Republic of)

    2016-05-15

    For a malfunction of a pressurizer level control system, a chemical and volume control system (CVCS) charging flowrate becomes a maximum level and a letdown flowrate a minimum level as well. Consequently, a water level and pressure of pressurizer is abnormally increased, which causes a pilot operated relief valve (POSRV) opened. It becomes important to investigate that a mixture from the POSRV becomes single-phase gas or two-phase mixture. In this study, the three-dimensional thermal-hydraulic behavior inside the pressurizer is numerically investigated by the CUPID code. The flow fields highly depend on some parameters such as subcooling of the stored water, interfacial drag model and POSRV opening. Thus, these parameters are examined in this study. These parameters are examined in this study. Less subcooling temperature makes the flow behavior unstable and flashing occur. The two-phase mixture is discharged through the POSRV. Moreover, less flow area delays a discharging flow rate. A sensitivity for the other parameters such critical flow model should be examined for the future work.

  16. Experimental and analytical study on thermal hydraulics in reduced-moderation water reactors

    Energy Technology Data Exchange (ETDEWEB)

    Akimoto, Hajime; Araya, Fumimasa; Ohnuki, Akira; Yoshida, Hiroyuki; Kureta, Masatoshi [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    2000-06-01

    Study and development of reduced-moderation spectrum water reactor proceeds as a option of the future type reactor in Japan Atomic Energy Research Institute (JAERI). The reduced-moderation spectrum in which a neutron has higher energy than the conventional water reactors is achieved by decreasing moderator-to-fuel ratio in the lattice core of the reactor. Conversion ratio in the reduced-moderation water reactor can be more than 1.0. High burnup and long term cycle operation of the reactor are expected. A type of heavy water cooled PWR and three types of BWR are discussed as follows; For the PWR, (1) critical heat flux experiments in hexagonal tight lattice core, (2) evaluation of cooling limit at a nominal power operation, and (3) analysis of rewetting cooling behavior at loss of coolant accident following with large scale pipe rupture. For the BWR, analyses of cooling limit at a nominal power operation of, (1) no blanket BWR, (2) long term cycle operation BWR, and (3) high conversion ratio BWR. The experiments and the analyses proved that the basic thermal hydraulic characteristics of these reduced-moderation water reactors satisfy the essential points of the safety requirements. (Suetake, M.)

  17. 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).

  18. 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.

  19. 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)

  20. Thermal Hydraulic Design and Analysis of a Water-Cooled Ceramic Breeder Blanket with Superheated Steam for CFETR

    Science.gov (United States)

    Cheng, Xiaoman; Ma, Xuebin; Jiang, Kecheng; Chen, Lei; Huang, Kai; Liu, Songlin

    2015-09-01

    The water-cooled ceramic breeder blanket (WCCB) is one of the blanket candidates for China fusion engineering test reactor (CFETR). In order to improve power generation efficiency and tritium breeding ratio, WCCB with superheated steam is under development. The thermal-hydraulic design is the key to achieve the purpose of safe heat removal and efficient power generation under normal and partial loading operation conditions. In this paper, the coolant flow scheme was designed and one self-developed analytical program was developed, based on a theoretical heat transfer model and empirical correlations. Employing this program, the design and analysis of related thermal-hydraulic parameters were performed under different fusion power conditions. The results indicated that the superheated steam water-cooled blanket is feasible. supported by the National Special Project for Magnetic Confined Nuclear Fusion Energy of China (Nos. 2013GB108004, 2014GB122000 and 2014GB119000), and National Natural Science Foundation of China (No. 11175207)

  1. Validation methodology for the evaluation of thermal-hydraulic sub-channel codes devoted to LOCA simulations

    Energy Technology Data Exchange (ETDEWEB)

    Seiler, N.; Ruyer, P.; Biton, B., E-mail: nathalie.seiler@irsn.fr, E-mail: pierre.ruyer@irsn.fr [IRSN/DPAM/SEMCA/LEMAR, CE Cadarache, Saint Paul lez Durance (France)

    2011-07-01

    This study focuses on thermal-hydraulic simulations, at sub-channel scale, of a damaged PWR reactor core during a Loss Of Coolant Accident (LOCA). The aim of this study is to accurately simulate the thermal-hydraulics to provide the thermal-mechanical code DRACCAR with an accurate wall heat transfer law. This latter code is developed by the French Safety Institute “Institut de Radioprotection et de Surete Nucleaire” (IRSN) to evaluate the thermics and deformations of fuel assemblies within the core. The present paper first describes the methodology considered to evaluate the capabilities of existing codes CATHARE-3 and CESAR to simulate dispersed droplet flows at a sub-channel scale and then provides some first evaluations of them. (author)

  2. A numerical study on improving the thermal hydraulic performance of printed circuit heat exchanger using the supercritical carbon dioxide

    Energy Technology Data Exchange (ETDEWEB)

    Park, Bo Guen; Kim, Dae Hyun; Chung, Jin Taek [Dept. of Mechanical Engineering, Korea University, Seoul (Korea, Republic of)

    2015-10-15

    The objective of this study is to propose a new channel shape that improves thermal-hydraulic performance. The existing Zigzag channel has high pressure loss due to flow separation and reverse flow. To improve this disadvantage, partial straight channel is inserted into bended points. Also, the effects of straight channel's length change on heat transfer and pressure loss are analyzed. Thermal-hydraulic performance of the new shape and existing Zigzag channel are quantitatively compared in terms of Goodness Factor. Mass flow rate was changed from 1.41 x 10{sup -4} kg/s to 2.48 x 10{sup -4} kg/s . The average volume goodness factor of 1mm straight channel shape was increased by 25% compared to the Zigzag channel.

  3. 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.

  4. Numerical analysis of thermal-hydraulic behavior of supercritical water in vertical upward/downward flow channels

    Institute of Scientific and Technical Information of China (English)

    GU Hanyang; YU Yiqi; CHENG Xu; LIU Xiaojing

    2008-01-01

    Investigations on the thermal-hydraulic behavior in the SCWR fuel assembly have obtained a significant attention in the international SCWR community. However, there is still a lack of understanding of the heat transfer behavior of supercritical fluids. In this paper, the numerical analysis is carried out to study the thermal-hydraulic behaviour in vertical sub-channels cooled by supercritical water. Remarkable differences in characteristics of secondary flow are found, especially in square lattice, between the upward flow and downward flow. The turbulence mixing across sub-channel gap for downward flow is much stronger than that for upward flow in wide lattice when the bulk temperature is lower than pseudo-critical point temperature. For downward flow, heat transfer deterioration phenomenon is suppressed with respect to the case of upward flow at the same conditions.

  5. Program ELM: A tool for rapid thermal-hydraulic analysis of solid-core nuclear rocket fuel elements

    Science.gov (United States)

    Walton, James T.

    1992-01-01

    This report reviews the state of the art of thermal-hydraulic analysis codes and presents a new code, Program ELM, for analysis of fuel elements. ELM is a concise computational tool for modeling the steady-state thermal-hydraulics of propellant flow through fuel element coolant channels in a nuclear thermal rocket reactor with axial coolant passages. The program was developed as a tool to swiftly evaluate various heat transfer coefficient and friction factor correlations generated for turbulent pipe flow with heat addition which have been used in previous programs. Thus, a consistent comparison of these correlations was performed, as well as a comparison with data from the NRX reactor experiments from the Nuclear Engine for Rocket Vehicle Applications (NERVA) project. This report describes the ELM Program algorithm, input/output, and validation efforts and provides a listing of the code.

  6. 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

  7. 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.

  8. Thermal-hydraulics, physical chemistry, and technology at nuclear power stations equipped with fast-neutron sodium-cooled reactors

    Science.gov (United States)

    Alekseev, V. V.; Efanov, A. D.; Kozlov, F. A.; Sorokin, A. P.

    2007-12-01

    Main results of investigations aimed at developing a verified system of computer codes that take into account the interrelation among nuclear-physical, thermal-hydraulic, physicochemical, thermal-mechanical, mass-transfer, and technological processes in nuclear power installations and at substantiating the models used as the core of these codes are presented together with the results of tests carried out to obtain data for verifying the codes.

  9. 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.

  10. RELAP5 model to simulate the thermal-hydraulic effects of grid spacers and cladding rupture during reflood

    Energy Technology Data Exchange (ETDEWEB)

    Nithianandan, C.K.; Klingenfus, J.A.; Reilly, S.S. [B& W Nuclear Technologies, Lynchburg, VA (United States)

    1995-09-01

    Droplet breakup at spacer grids and a cladding swelled and ruptured locations plays an important role in the cooling of nuclear fuel rods during the reflooding period of a loss-of-coolant accident (LOCA) in a pressurized water reactor (PWR). During the reflood phase, a spacer grid affects the thermal-hydraulic system behavior through increased turbulence, droplet breakup due to impact on grid straps, grid rewetting, and liquid holdup due to grid form losses. Recently, models to simulate spacer grid effects and blockage and rupture effects on system thermal hydraulics were added to the B&W Nuclear Technologies (BWNT) version of the RELAP5/MOD2 computer code. Several FLECHT-SEASET forced reflood tests, CCTF Tests C1-19 and C2-6, SCTF Test S3-15, and G2 Test 561 were simulated using RELAP5/MOD2-B&W to verify the applicability of the model at the cladding swelled and rupture locations. The results demonstrate the importance of modeling the thermal-hydraulic effects due to grids, and clad swelling and rupture to correctly predict the clad temperature response during the reflood phase of large break LOCA. The RELAP5 models and the test results are described in this paper.

  11. 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.

  12. 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)

  13. A coupled neutronic/thermal-hydraulic scheme between COBAYA3 and SUBCHANFLOW within the NURESIM simulation platform

    Energy Technology Data Exchange (ETDEWEB)

    Calleja, M.; Stieglitz, R.; Sanchez, V.; Jimenez, J.; Imke, U. [Karlsruhe Inst. of Technology KIT, Inst. for Neutron Physics and Reactor Technology INR, Hermann-von-Helmholtz-Platz 1, D-76344 Eggenstein-Leopoldshafen (Germany)

    2012-07-01

    Multi-scale, multi-physics problems reveal significant challenges while dealing with coupled neutronic/thermal-hydraulic solutions. Current generation of codes applied to Light Water Reactors (LWR) are based on 3D neutronic nodal methods coupled with one or two phase flow thermal-hydraulic system or sub-channel codes. In addition, spatial meshing and temporal schemes are crucial for the proper description of the non-symmetrical core behavior in case of transient and accidents e.g. reactivity insertion accidents. This paper describes the coupling approach between the 3D neutron diffusion code COBAYA3 and the sub-channel code SUBCHANFLOW within SALOME. The coupling is done inside the SALOME open source platform that is characterized by a powerful pre- and post-processing capabilities and a novel functionality for mapping of the neutronic and thermal hydraulic domains. The peculiar functionalities of SALOME and the steps required for the code integration and coupling are presented. The validation of the coupled codes is done based on two benchmarks the PWR MOX/UO{sub 2} RIA and the TMI-1 MSLB benchmark. A discussion of the prediction capability of COBAYA3/SUBCHANFLOW compared to other coupled solutions will be provided too. (authors)

  14. Thermal-mechanical and thermal-hydraulic integrated study of the Helium-Cooled Lithium Lead Test Blanket Module

    Energy Technology Data Exchange (ETDEWEB)

    Chiovaro, P., E-mail: pg.chiovaro@din.unipa.it [Dipartimento di Ingegneria Nucleare, Universita di Palermo, Palermo (Italy); Di Maio, P.A.; Giammusso, R.; Lupo, Q.; Vella, G. [Dipartimento di Ingegneria Nucleare, Universita di Palermo, Palermo (Italy)

    2010-12-15

    The Helium-Cooled Lithium Lead Test Blanket Module (HCLL-TBM) is one of the two TBM to be installed in an ITER equatorial port since day 1 of operation, with the specific aim to investigate the main concept functionalities and issues such as high efficiency helium cooling, resistance to thermo-mechanical stresses, manufacturing techniques, as well as tritium transport, magneto-hydrodynamics effects and corrosion. In particular, in order to show a DEMO-relevant thermo-mechanical and thermal-hydraulic behavior, the HCLL-TBM has to meet several requirements especially as far as its coolant thermofluid-dynamic conditions and its thermal-mechanical field are concerned. The present paper is focused on the assessment of the HCLL-TBM thermal-mechanical performances under both nominal and accidental load conditions, by adopting a computational approach based on the Finite Element Method. A realistic 3D finite element model of the whole HCLL-TBM, in the horizontal first wall design has been set up, consisting of about 597,000 elements and 767,000 nodes. In particular, since the thermal fields of both the module and the coolant are strictly coupled, the helium flow domain has been modeled too and a thermal contact model has been set up to properly simulate the convective heat transfer between the structure wall and the coolant. Pure conductive heat transfer has been assumed within the Pb-Li eutectic alloy of the breeder units. The volumetric density of the nuclear deposited power, recently calculated at Department of Nuclear Engineering of the University of Palermo by the MCNP 4C code, has been applied as distributed thermal load in order to assess the potential influence on the module thermo-mechanical performances of the markedly non-uniform poloidal and toroidal distributions that have been predicted within the Segment Box. Different loading scenarios have been considered as to the heat flux onto the module First Wall. Steady state and transient thermal-mechanical analyses

  15. 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.

  16. 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.

  17. Thermal-hydraulic instabilities in natural circulation flow loops under supercritical conditions

    Science.gov (United States)

    Jain, Rachna

    In recent years, a growing interest has been generated in investigating the thermal hydraulics and flow stability phenomenon in supercritical natural circulation loops. These flow conditions are relevant to some of the innovative passive safety designs proposed for the Gen-IV Supercritical Water Reactor (SCWR) concepts. A computational model has been developed at UW Madison which provides a good basic simulation tool for the steady state and transient analysis of one dimensional natural circulation flow, and can be applied to conduct stability analysis. Several modifications and improvements were incorporated in an earlier numerical scheme before applying it to investigate the transient behavior of two experimental loops, namely, the supercritical water loop at UW-Madison and the supercritical carbon-dioxide (SCCO2) loop at Argonne National Laboratories. Although the model predicted development of instabilities for both SCW and SCCO2 loop which agrees with some previous work, the experiments conducted at SCCO2 loop exhibited stable behavior under similar conditions. To distinguish between numerical effects and physical processes, a linear stability approach has also been developed to investigate the stability characteristics associated with the natural circulation loop systems for various inlet conditions, input powers and geometries. The linear stability results for the SCW and SCCO2 loops exhibited differences with the corresponding transient simulations. This linear model also predicted the presence of instability in the SCCO 2 loop for certain high input powers contradictory to the experimental findings. Dimensionless parameters were proposed which would generalize the stability characteristics of the natural circulation flow loops under supercritical conditions.

  18. Supplemental Thermal-Hydraulic Transient Analyses of BR2 in Support of Conversion to LEU Fuel

    Energy Technology Data Exchange (ETDEWEB)

    Licht, J. [Argonne National Lab. (ANL), Argonne, IL (United States); Dionne, B. [Argonne National Lab. (ANL), Argonne, IL (United States); Sikik, E. [Belgian Nuclear Research Center (SCK-CEN), Mol (Belgium); Van den Branden, G. [Belgian Nuclear Research Center (SCK-CEN), Mol (Belgium); Koonen, E. [Belgian Nuclear Research Center (SCK-CEN), Mol (Belgium)

    2016-01-01

    Belgian Reactor 2 (BR2) is a research and test reactor located in Mol, Belgium and is primarily used for radioisotope production and materials testing. The Materials Management and Minimization (M3) Reactor Conversion Program of the National Nuclear Security Administration (NNSA) is supporting the conversion of the BR2 reactor from Highly Enriched Uranium (HEU) fuel to Low Enriched Uranium (LEU) fuel. The RELAP5/Mod 3.3 code has been used to perform transient thermal-hydraulic safety analyses of the BR2 reactor to support reactor conversion. A RELAP5 model of BR2 has been validated against select transient BR2 reactor experiments performed in 1963 by showing agreement with measured cladding temperatures. Following the validation, the RELAP5 model was then updated to represent the current use of the reactor; taking into account core configuration, neutronic parameters, trip settings, component changes, etc. Simulations of the 1963 experiments were repeated with this updated model to re-evaluate the boiling risks associated with the currently allowed maximum heat flux limit of 470 W/cm2 and temporary heat flux limit of 600 W/cm2. This document provides analysis of additional transient simulations that are required as part of a modern BR2 safety analysis report (SAR). The additional simulations included in this report are effect of pool temperature, reduced steady-state flow rate, in-pool loss of coolant accidents, and loss of external cooling. The simulations described in this document have been performed for both an HEU- and LEU-fueled core.

  19. 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)

  20. The effects of complexity, of simplicity and of scaling in thermal-hydraulics

    Energy Technology Data Exchange (ETDEWEB)

    Zuber, Novak

    2001-02-01

    This lecture has a twofold purpose. First, we will assess the state of the art and the trends in thermal-hydraulics (T-H) technology, within the context of replicating and non-replicating information systems. Four T-H examples are used to illustrate that an ever-increasing complexity in formulating and analyzing problems leads to inefficiency, obsolescence and evolutionary failure. By contrast, simplicity, which allows for parsimony, synthesis and clarity of information, ensures efficiency, survival and replication. This comparison (complexity versus simplicity) also provides the requirements and guidance for a success path in T-H development. The second objective of this paper is to demonstrate that scaling provides the means to process information in an efficient manner, as required by competitive (and, thereby, replicating) systems. To this end, the lecture summarizes the essential features of the Fractional Change, Scaling and Analysis approach, which offers a general paradigm for quantifying the effects that an agent of change has on a given information system. The paper will further demonstrate that a single concept and a single method may be used to scale and analyze all transport processes in a given field of interest (fluid mechanics, heat transfer, etc.) and/or across fields and disciplines (mechanics, biology, etc.) Therefore, the paradigm: (1) ensures economy and efficiency in addressing and resolving technical or scientific problems; and (2) enables a 'cultural cross-pollination' between different information systems (disciplines). By means of a simple example in the Appendix, we shall: (1) demonstrate the efficiency to be gained through scaling; and (2) illustrate the inefficiency and wastefulness of computer-based safety studies as presently conducted.

  1. 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.

  2. 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.

  3. 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

  4. 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

  5. Numerical study of two equation turbulence models for subchannel thermal hydraulics

    Energy Technology Data Exchange (ETDEWEB)

    Nazififard, Mohammad; Suha, Kune Y. [Seoul National Univ., Seoul (Korea, Republic of)

    2012-10-15

    The need for more accurate computational methods for the analysis of nuclear reactor systems has generated rising interests for computational fluid dynamics (CFD) and growing range of applications of commercial CFD software. This study presents results of the sensitivity analysis using the two equation turbulence models for several grid configurations. The Turbulence Enhanced Mixing Analysis (TEMA) result contributes further to turbulent convective heat transfer mechanisms in a subchannel of a square array rod bundle.

  6. Thermal hydraulic test for reactor safety system - Critical heat flux experiment and development of prediction models

    Energy Technology Data Exchange (ETDEWEB)

    Chang, Soon Heung; Baek, Won Pil; Yang, Soo Hyung; No, Chang Hyun [Korea Advanced Institute of Science and Technology, Taejon (Korea)

    2000-04-01

    To acquire CHF data through the experiments and develop prediction models, research was conducted. Final objectives of research are as follows: 1) Production of tube CHF data for low and middle pressure and mass flux and Flow Boiling Visualization. 2) Modification and suggestion of tube CHF prediction models. 3) Development of fuel bundle CHF prediction methodology base on tube CHF prediction models. The major results of research are as follows: 1) Production of the CHF data for low and middle pressure and mass flux. - Acquisition of CHF data (764) for low and middle pressure and flow conditions - Analysis of CHF trends based on the CHF data - Assessment of existing CHF prediction methods with the CHF data 2) Modification and suggestion of tube CHF prediction models. - Development of a unified CHF model applicable for a wide parametric range - Development of a threshold length correlation - Improvement of CHF look-up table using the threshold length correlation 3) Development of fuel bundle CHF prediction methodology base on tube CHF prediction models. - Development of bundle CHF prediction methodology using correction factor. 11 refs., 134 figs., 25 tabs. (Author)

  7. Validation and Application of the Thermal Hydraulic System Code TRACE for Analysis of BWR Transients

    Directory of Open Access Journals (Sweden)

    V. H. Sánchez

    2012-01-01

    Full Text Available The Karlsruhe Institute of Technology (KIT is participating on (Code Applications and Maintenance Program CAMP of the US Nuclear Regulatory Commission (NRC to validate TRACE code for LWR transient analysis. The application of TRACE for the safety assessment of BWR requires a throughout verification and validation using experimental data from separate effect and integral tests but also using plant data. The validation process is normally focused on safety-relevant phenomena for example, pressure drop, void fraction, heat transfer, and critical power models. The purpose of this paper is to validate selected BWR-relevant TRACE-models using both data of bundle tests such as the (Boiling Water Reactor Full-Size Fine-Mesh Bundle Test BFBT and plant data recorded during a turbine trip event (TUSA occurred in a Type-72 German BWR plant. For the validation, TRACE models of the BFBT bundle and of the BWR plant were developed. The performed investigations have shown that the TRACE code is appropriate to describe main BWR-safety-relevant phenomena (pressure drop, void fraction, and critical power with acceptable accuracy. The comparison of the predicted global BWR plant parameters for the TUSA event with the measured plant data indicates that the code predictions are following the main trends of the measured parameters such as dome pressure and reactor power.

  8. Thermal-hydraulic analysis of NSSS and containment response during extended station blackout for Maanshan PWR plant

    Energy Technology Data Exchange (ETDEWEB)

    Yuann, Yng-Ruey, E-mail: ryyuann@iner.gov.tw; Hsu, Keng-Hsien, E-mail: hardlycampus@iner.gov.tw; Lin, Chin-Tsu, E-mail: jtling@iner.gov.tw

    2015-07-15

    Highlights: • Calculate NSSS and containment transient response during extended SBO of 24 h. • RELAP5-3D and GOTHIC models are developed for Maanshan PWR plant. • Reactor coolant pump seal leakage is specifically modeled for each loop. • Analyses are performed with and without secondary-side depressurization, respectively. • Considering different total available time for turbine driven auxiliary feedwater system. - Abstract: A thermal-hydraulic analysis has been performed with respect to the response of the nuclear steam supply system (NSSS) and the containment during an extended station blackout (SBO) duration of 24 h in Maanshan PWR plant. Maanshan plant is a Westinghouse three-loop PWR design with rated core thermal power of 2822 MWt. The analyses in the NSSS and the containment are based on the RELAP5-3D and GOTHIC models, respectively. Important design features of the plant in response to SBO are considered in the respective models, e.g., the steam generator PORVs, turbine driven auxiliary feedwater system (TDAFWS), accumulators, reactor coolant pump (RCP) seal design, various heat structures in the containment, etc. In the analysis it is assumed that the shaft seal in each RCP failed due to loss of seal cooling and the RCS fluid flows to the containment directly. Some parameters calculated from the RELPA5-3D model are input to the containment GOTHIC model, including the RCS average temperature and the RCP seal leakage flow and enthalpy. The RCS average temperature is used to drive the sensible heat transfer to the containment. It is found that the severity of the event depends mainly on whether the secondary side is depressurized or not. If the secondary side is depressurized in time (within 1 h after SBO) and the TDAFWS is available greater than 19 h, then the reactor core will be covered with water throughout the SBO duration, which ensures the integrity of the reactor core. On the contrary, if the secondary side is not depressurized, then the RCS

  9. 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

  10. 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.

  11. 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.

  12. Development and qualification of a thermal-hydraulic nodalization for modeling station blackout accident in PSB-VVER test facility

    Energy Technology Data Exchange (ETDEWEB)

    Saghafi, Mahdi [Department of Energy Engineering, Sharif University of Technology, Azadi Avenue, Tehran (Iran, Islamic Republic of); Ghofrani, Mohammad Bagher, E-mail: ghofrani@sharif.edu [Department of Energy Engineering, Sharif University of Technology, Azadi Avenue, Tehran (Iran, Islamic Republic of); D’Auria, Francesco [San Piero a Grado Nuclear Research Group (GRNSPG), University of Pisa, Via Livornese 1291, San Piero a Grado, Pisa (Italy)

    2016-07-15

    Highlights: • A thermal-hydraulic nodalization for PSB-VVER test facility has been developed. • Station blackout accident is modeled with the developed nodalization in MELCOR code. • The developed nodalization is qualified at both steady state and transient levels. • MELCOR predictions are qualitatively and quantitatively in acceptable range. • Fast Fourier Transform Base Method is used to quantify accuracy of code predictions. - Abstract: This paper deals with the development of a qualified thermal-hydraulic nodalization for modeling Station Black-Out (SBO) accident in PSB-VVER Integral Test Facility (ITF). This study has been performed in the framework of a research project, aiming to develop an appropriate accident management support tool for Bushehr nuclear power plant. In this regard, a nodalization has been developed for thermal-hydraulic modeling of the PSB-VVER ITF by MELCOR integrated code. The nodalization is qualitatively and quantitatively qualified at both steady-state and transient levels. The accuracy of the MELCOR predictions is quantified in the transient level using the Fast Fourier Transform Base Method (FFTBM). FFTBM provides an integral representation for quantification of the code accuracy in the frequency domain. It was observed that MELCOR predictions are qualitatively and quantitatively in the acceptable range. In addition, the influence of different nodalizations on MELCOR predictions was evaluated and quantified using FFTBM by developing 8 sensitivity cases with different numbers of control volumes and heat structures in the core region and steam generator U-tubes. The most appropriate case, which provided results with minimum deviations from the experimental data, was then considered as the qualified nodalization for analysis of SBO accident in the PSB-VVER ITF. This qualified nodalization can be used for modeling of VVER-1000 nuclear power plants when performing SBO accident analysis by MELCOR code.

  13. DEVELOPMENT OF A COMPUTER PROGRAM TO SUPPORT AN EFFICIENT NON-REGRESSION TEST OF A THERMAL-HYDRAULIC SYSTEM CODE

    Directory of Open Access Journals (Sweden)

    JUN YEOB LEE

    2014-10-01

    Full Text Available During the development process of a thermal-hydraulic system code, a non-regression test (NRT must be performed repeatedly in order to prevent software regression. The NRT process, however, is time-consuming and labor-intensive. Thus, automation of this process is an ideal solution. In this study, we have developed a program to support an efficient NRT for the SPACE code and demonstrated its usability. This results in a high degree of efficiency for code development. The program was developed using the Visual Basic for Applications and designed so that it can be easily customized for the NRT of other computer codes.

  14. Comparison Study on Thermal-Hydraulic Analysis Depending on Liquid Relief Valve Response for an Station Blackout in CANDU-6

    Energy Technology Data Exchange (ETDEWEB)

    Kim, S. M.; Kho, D. W. [KHNP-CRI, Daejeon (Korea, Republic of); Choi, S. H.; Moon, B. J.; Kim, S. R. [Nuclear Engineering Service and Solution Co., Daejeon (Korea, Republic of)

    2014-10-15

    The purpose of this analysis is to compare the results of thermal-hydraulic analysis depending on liquid relief valve response during a station black out (SBO) events in CANDU-6. The primary heat transport system (PHTS) behavior following the postulated SBO is analyzed using CATHENA code. In the paper, analysis was performed to evaluate the effect on coolant system where LRVs are assumed to be opened or opened according to normal open characteristics in the condition of SBO. The result showed that the primary pressure boundary is extended from LRV to DCT and the effects on primary system behavior were neglectable.

  15. 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.

  16. 反转压水反应堆热工水力特性初步研究%The Preliminary Research of Thermal-Hydraulic Behavior of an Inverted Pressurized Water Reactor

    Institute of Scientific and Technical Information of China (English)

    刘杰; 于涛; 谢金森; 曾正魁; 秦勉

    2012-01-01

    In this paper, CFD analysis is carried out to study the single fuel element and the coolant channel flow field of the Inverted Pressurized Water Reactor (IPWR) using commercial CFD code FLUENT,which analyses and compares the thermal-hydraulic char- acteristics of different grid size. The calculation results show that the dimensions of the IPWR fuel cell has greater influence on the temperature and heat transfer characteristics of coolant, and the study provides preliminary reference and basis for the next design of the IPWR fuel cell,fuel assembly, reactor core and the thermal-hydraulic analysis.%采用CFD软件FLUENT对反转压水反应堆(IPWR:Inverted PressurizedWaterReactor)单个燃料元件及冷却剂通道流场进行了数值模拟计算,分析比较了不同栅格尺寸情况下的热工水力特性.计算结果表明,栅格尺寸对IPWR燃料元件温度及冷却剂流动传热特性有较大影响,为今后IPWR燃料栅元、组件、堆芯设计和热工水力分析提供了初步参考和依据.

  17. Large-scale Flow Pulsation in Tight Square Arrayed Rod Bundles of Nuclear Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Tae Hwan; Kim, Kyung Min; Cho, Hyung Hee [Yonsei University, Seoul (Korea, Republic of); Shin, Chang Hwan; In, Wang Kee [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2011-05-15

    As a major component of modern nuclear reactor, the nuclear fuel rod bundles with liquid coolant have been studied by a lot of researchers to understand the flow structure between the fuel rods. Recently, rod arrays with much small pitch-to-diameter ratio have been being tried to increase performance of the nuclear reactor. The liquid coolant flowing axially through these small spaces between the rods is known to show some peculiar phenomena including large-scale, quasi-periodic flow pulsation. These flow pulsation phenomena dominate mixing process in the subchannels. Thus, precise understating of the flow structure is essential to predict thermal-hydraulic phenomena in nuclear rod bundles. In this present paper, the turbulent flow in tight square arrayed rod bundles is investigated with Hot-wire anemometry. Then, the measured velocity data are analyzed by using Fast Fourier Transform analysis to find characteristic frequency of the pulsation

  18. Coolant mixing in LMFBR rod bundles and outlet plenum mixing transients. Final report

    Energy Technology Data Exchange (ETDEWEB)

    Todreas, N.E.; Cheng, S.K.; Basehore, K.

    1984-08-01

    This project principally undertook the investigation of the thermal hydraulic performance of wire wrapped fuel bundles of LMFBR configuration. Results obtained included phenomenological models for friction factors, flow split and mixing characteristics; correlations for predicting these characteristics suitable for insertion in design codes; numerical codes for analyzing bundle behavior both of the lumped subchannel and distributed parameter categories and experimental techniques for pressure velocity, flow split, salt conductivity and temperature measurement in water cooled mockups of bundles and subchannels. Flow regimes investigated included laminar, transition and turbulent flow under forced convection and mixed convection conditions. Forced convections conditions were emphasized. Continuing efforts are underway at MIT to complete the investigation of the mixed convection regime initiated here. A number of investigations on outlet plenum behavior were also made. The reports of these investigations are identified.

  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. On Double Vector Bundles

    Institute of Scientific and Technical Information of China (English)

    Zhuo CHEN; Zhang Ju LIU; Yun He SHENG

    2014-01-01

    In this paper, we construct a category of short exact sequences of vector bundles and prove that it is equivalent to the category of double vector bundles. Moreover, operations on double vector bundles can be transferred to operations on the corresponding short exact sequences. In particular, we study the duality theory of double vector bundles in term of the corresponding short exact sequences. Examples including the jet bundle and the Atiyah algebroid are discussed.

  1. On Double Vector Bundles

    OpenAIRE

    Chen, Zhuo; Liu, Zhangju; Sheng, Yunhe

    2011-01-01

    In this paper, we construct a category of short exact sequences of vector bundles and prove that it is equivalent to the category of double vector bundles. Moreover, operations on double vector bundles can be transferred to operations on the corresponding short exact sequences. In particular, we study the duality theory of double vector bundles in term of the corresponding short exact sequences. Examples including the jet bundle and the Atiyah algebroid are discussed.

  2. 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.

  3. The Thermal-Hydraulic model for the pebble bed modular reactor (PBMR) plant operator training simulator system

    Energy Technology Data Exchange (ETDEWEB)

    Dudley, Trevor [Pebble Bed Modular Reactor (Proprietary) Limited, Die Anker Building, Centurion 0046 (South Africa)], E-mail: trevor.dudley@pbmr.co.za; Bouwer, Werner; Villiers, Piet de [Pebble Bed Modular Reactor (Proprietary) Limited, Die Anker Building, Centurion 0046 (South Africa); Wang Zen [GSE Systems, Inc., 7133 Rutherford Suite 200, Baltimore, MD 21244 (United States)

    2008-11-15

    This paper provides a discussion of the model development status and verification efforts for the Reactor Core Thermal-Hydraulic model developed for the full-scope plant Operator Training Simulator System of the Pebble Bed Modular Reactor (PBMR). Due to the First of a Kind Engineering nature and lack of reference plant data, model verification has mainly been focused on benchmarking the model configurations against test cases performed by PBMR design analysis codes, i.e. TINTE, VSOP and FLOWNEX. As a first step, due to the symmetrical physical nature of the PBMR core, a two-dimensional (2D) model configuration in radial and axial directions (axial-symmetry) was developed. The design was subsequently extended to a three-dimensional (3D) configuration. Through the use of cross-flow and cross-conduction links, three nearly identical 2D configurations were glued together to form this 3D model configuration. To date, the 3D configuration represents the most comprehensive model to simulate the PBMR core thermo-hydraulics. This paper concludes with the verification of thermodynamic and heat-transfer properties of two steady state (100% and 40% power) conditions between the 3D Reactor Core Thermal-Hydraulic model and the available FLOWNEX and TINTE design code analysis. The transient operations between these two power levels are also discussed.

  4. Dynamic thermal-hydraulic modeling and stack flow pattern analysis for all-vanadium redox flow battery

    Science.gov (United States)

    Wei, Zhongbao; Zhao, Jiyun; Skyllas-Kazacos, Maria; Xiong, Binyu

    2014-08-01

    The present study focuses on dynamic thermal-hydraulic modeling for the all-vanadium flow battery and investigations on the impact of stack flow patterns on battery performance. The inhomogeneity of flow rate distribution and reversible entropic heat are included in the thermal-hydraulic model. The electrolyte temperature in tanks is modeled with the finite element modeling (FEM) technique considering the possible non-uniform distribution of electrolyte temperature. Results show that the established model predicts electrolyte temperature accurately under various ambient temperatures and current densities. Significant temperature gradients exist in the battery system at extremely low flow rates, while the electrolyte temperature tends to be the same in different components under relatively high flow rates. Three stack flow patterns including flow without distribution channels and two cases of flow with distribution channels are compared to investigate their effects on battery performance. It is found that the flow rates are not uniformly distributed in cells especially when the stack is not well designed, while adding distribution channels alleviates the inhomogeneous phenomenon. By comparing the three flow patterns, it is found that the serpentine-parallel pattern is preferable and effectively controls the uniformity of flow rates, pressure drop and electrolyte temperature all at expected levels.

  5. Principal noncommutative torus bundles

    DEFF Research Database (Denmark)

    Echterhoff, Siegfried; Nest, Ryszard; Oyono-Oyono, Herve

    2008-01-01

    In this paper we study continuous bundles of C*-algebras which are non-commutative analogues of principal torus bundles. We show that all such bundles, although in general being very far away from being locally trivial bundles, are at least locally trivial with respect to a suitable bundle version...... of bivariant K-theory (denoted RKK-theory) due to Kasparov. Using earlier results of Echterhoff and Williams, we shall give a complete classification of principal non-commutative torus bundles up to equivariant Morita equivalence. We then study these bundles as topological fibrations (forgetting the group...... action) and give necessary and sufficient conditions for any non-commutative principal torus bundle being RKK-equivalent to a commutative one. As an application of our methods we shall also give a K-theoretic characterization of those principal torus-bundles with H-flux, as studied by Mathai...

  6. Development of Thermal-Hydraulic Steady-State Analysis Program for Primary Loop of China Experimental Fast Reactor%中国实验快堆一回路热工水力稳态计算程序开发

    Institute of Scientific and Technical Information of China (English)

    饶彧先; 崔满满; 郭赟

    2012-01-01

    针对中国实验快堆(CEFR)的具体结构和稳态运行特点,利用Fortran语言开发了CEFR一回路热工水力稳态计算程序.重点开发了有关钠的多种物性的子程序、适应不同工况的钠的流动与换热计算子程序,并对关系式进行了对比分析,最后建立了稳态计算模型并开发了程序.在此基础上,对CEFR的一回路系统在满功率下的稳态热工水力特性进行了计算分析,所获得的结果同设计参数吻合,证明了所开发的子程序及稳态程序的正确性.%According to the characteristics of structure and steady-state for primary loop of China Experimental Fast Reactor (CEFR), a thermal-hydraulic steady-state analysis program was developed by using Fortran language. This paper focused on the development of a set of subroutine of physical properties of sodium and the sodium flow and heat transfer correlations for different operation conditions. And the difference among these correlations was compared. The calculation program was developed based on the steady model. At last, the thermal-hydraulic characteristics of steady-state of the primary loop of CEFR at full power were calculated. The calculation results are consistent with the design parameters and the correctness of the developed subroutines and steady-state calculation program was proved.

  7. Analysis report of the thermal-hydraulic characteristics of the high temperature/high pressure thermal-hydraulic test facility (VISTA) in steady state conditions

    Energy Technology Data Exchange (ETDEWEB)

    Park, Hyun Sik; Choi, Ki Yong; Cho, Seok; Lee, Sung Jae; Song, Chul Hwa; Park, Chun Kyong; Chung, Moon Ki

    2004-01-01

    The VISTA (Experimental Verification by Integral Simulation of Transients and Accidents) is an experimental facility to verify the performance and safety issues of the SMART-P (Pilot plant of the system-integrated modular advanced reactor). The basic design of the SMART-P has been completed by KAERI. The present report describes the experimental results on the water inventory distribution, the pressure distribution, and the differential pressure characteristics of the VISTA facility and on the heat transfer characteristics of the core simulating heater and the steam generator of the VISTA facility. There were little differences of their water inventories between the designed and the measured data. The pressure of the VISTA primary system kept near the steady-state operating pressure of 147 bar, and the differential pressures through the primary and secondary systems increased with the increase of their flow rates. Also the surface temperatures of core simulating heaters were measured, and the overall heat transfer coefficient of the VISTA steam generator was calculated to show a little higher values than that of the SMART-P.

  8. Thermal-hydraulic Analysis for TVS-2M Type Pilot Fuel Assembly at Tianwan NPS%田湾核电站 TVS-2M 先导组件堆芯热工水力分析

    Institute of Scientific and Technical Information of China (English)

    姚进国; 李旭东; 杨晓强; 李载鹏; 杨高升

    2014-01-01

    Six TVS-2M type pilot fuel assemblies were loaded starting from the 5th fuel cycle at Tianwan NPS Unit 1 ,and would be under 4-year in-core fuel operation lifetime test from the 5th cycle to the 8th fuel cycle .Based on in-core thermal-hydraulic analy-sis ,as well as the thermal-hydraulic test measurement and fuel assembly deformation inspection for the TVS-2M type pilot fuel assemblies ,it is verified that the calculation mode of the thermal-hydraulic design code is reasonable and the calculation results are in good agreement with test results .The results show that the TVS-2M type fuel assembly and the AFA fuel assembly have a good compatibility during reactor operation ,and the core operation reliability and safety can also be ensured during the transition fuel cycles .%田湾核电站一号机组于第5燃料循环装入6组T VS-2M先导燃料组件,并将经历从第5燃料循环到第8燃料循环4年的堆内运行。本文通过对先导燃料组件堆芯热工水力分析,堆芯运行实际试验测量以及组件变形检查,验证了热工水力设计程序计算模型的合理性以及计算结果与试验结果的符合性。结果表明,TVS-2M燃料组件与AFA燃料组件具有良好的相容性,从而证实了过渡循环条件下反应堆运行的安全性和可靠性。

  9. Multiple Module Simulation of Water Cooled Breeding Blankets in K-DEMO Using Thermal-Hydraulic Analysis Code MARS-KS

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Geon-Woo; Lee, Jeong-Hun; Park, Goon-Cherl; Cho, Hyoung-Kyu [Seoul National University, Seoul (Korea, Republic of); Im, Kihak [National Fusion Research Institute, Daejeon (Korea, Republic of)

    2015-10-15

    A preliminary concept for the Korean fusion demonstration reactor (K-DEMO) has been studied by the National Fusion Research Institute (NFRI) based on the National Fusion Roadmap of Korea. The feasibility studies have been performed in order to establish the conceptual design guidelines of the breeding blanket. As a part of the NFRI research, Seoul National University (SNU) is conducting thermal design, evaluation and validation of the water-cooled breeding blanket for the K-DEMO reactor. The purpose of this study is to extend the capability of MARS-KS to the overall blanket system analysis which includes 736 blanket modules in total. The strategy for the multi-module blanket system analysis using MARS-KS is introduced and the analysis result of the 46 blanket modules of single sector was summarized. A thermal-hydraulic analysis code for a nuclear reactor safety, MARS-KS, was applied for thermal analysis of the conceptual design of the K-DEMO breeding blanket. Then, a methodology to simulate multiple blanket modules was proposed, which uses a supervisor program to handle each blanket module individually at first and then distribute the flow rate considering the pressure drop that occurs in each module. For a feasibility test of the proposed methodology, 46 blankets in a sector, which are connected with each other through the common headers for the sector inlet and outlet, were simulated. The calculation results of flow rates, pressure drops, and temperatures showed the validity of the calculation. Because of parallelization using the MPI system, the computational time could be reduced significantly. In future, this methodology will be extended to an efficient simulation of multiple sectors, and further validation for transient simulation will be carried out for more practical applications.

  10. 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)

  11. Development of TUF-ELOCA - a software tool for integrated single-channel thermal-hydraulic and fuel element analyses

    Energy Technology Data Exchange (ETDEWEB)

    Popescu, A.I.; Wu, E.; Yousef, W.W.; Pascoe, J. [Nuclear Safety Solutions Ltd., Toronto, Ontario (Canada); Parlatan, Y. [Ontario Power Generation, Toronto, Ontario (Canada); Kwee, M. [Bruce Power, Tiverton, Ontario (Canada)

    2006-07-01

    The TUF-ELOCA tool couples the TUF and ELOCA codes to enable an integrated thermal-hydraulic and fuel element analysis for a single channel during transient conditions. The coupled architecture is based on TUF as the parent process controlling multiple ELOCA executions that simulate the fuel elements behaviour and is scalable to different fuel channel designs. The coupling ensures a proper feedback between the coolant conditions and fuel elements response, eliminates model duplications, and constitutes an improvement from the prediction accuracy point of view. The communication interfaces are based on PVM and allow parallelization of the fuel element simulations. Developmental testing results are presented showing realistic predictions for the fuel channel behaviour during a transient. (author)

  12. EXPERIMENTAL VERIFICATION OF THE THREE-DIMENSIONAL THERMAL-HYDRAULIC MODELS IN THE BEST-ESTIMATE CODE BAGIRA.

    Energy Technology Data Exchange (ETDEWEB)

    KALINICHENKO,S.D.KROSHILIN,A.E.KROSHILIN,V.E.SMIRNOV,A.V.KOHUT,P.

    2004-03-15

    In this paper we present verification results of the BAGIRA code that was performed using data from integral thermal-hydraulic experimental test facilities as well as data obtained from operating nuclear power plants. BAGIRA is a three-dimensional numerical best-estimate code that includes non-homogeneous modeling. Special consideration was given to the recently completed experimental data from the PSB-VVER integral test facility (EREC, Electrogorsk, Russia)--a new Russian large-scale four-loop unit, which has been designed to model the primary circuits of VVER-1000 type reactors. It is demonstrated that the code BAGIRA can be used to analyze nuclear reactor behavior under normal and accident conditions.

  13. Three dimensional considerations in thermal-hydraulics of helical cruciform fuel rods for LWR power uprates

    Energy Technology Data Exchange (ETDEWEB)

    Shirvan, Koroush, E-mail: kshirvan@mit.edu; Kazimi, Mujid S.

    2014-04-01

    Highlights: • We benchmarked the 4 × 4 helical cruciform fuel (HCF) bundle pressure drop experimental data with CFD. • We also benchmarked the 4 × 4 HCF mixing experimental data with CFD. • We derived new friction factors for PWR and BWR designs at PWR and BWR operating conditions from CFD. • We showed the importance of modeling the 3D conduction in HCF in steady state and transient conditions. - Abstract: In order to increase the power density of current and new light water reactor designs, the helical cruciform fuel (HCF) rods have been proposed. The HCF rod is equivalent to a thin cylindrical rod, with 4 fuel containing vanes, wrapped around it. The HCF rods increase the surface area to volume ratio of the fuel and enhance the inter-subchannel mixing due to their helical shape. The rods do not need supporting grids, as they are packed to periodically contact their neighbors along the flow direction, enabling a higher power density in the core. The HCF rods were reported to have the potential to uprate existing PWRs by 45% and BWRs by 20%. In order to quantify the mixing behavior of the HCF rods based on their twist pitch, experiments were previously performed at atmospheric pressures with single phase water in a 4 by 4 HCF and cylindrical rod bundles. In this paper, the experimental results on pressure drop and mixing are benchmarked with computational fluid dynamic (CFD) using steady state the Reynolds average Navier–Stokes (RANS) turbulence model. The sensitivity of the CFD approach to computational domain, mesh size, mesh shape and RANS turbulence models are examined against the experimental conditions. Due to the refined radial velocity profile from the HCF rods twist, the turbulence models showed little sensitivity to the domain. Based on the CFD simulations, the total pressure drops under the PWR and BWR conditions are expected to be about 10% higher than the values previously reported solely from an empirical correlation based on the

  14. Predictive uncertainty reduction in coupled neutron-kinetics/thermal hydraulics modeling of the BWR-TT2 benchmark

    Energy Technology Data Exchange (ETDEWEB)

    Badea, Aurelian F., E-mail: aurelian.badea@kit.edu [Karlsruhe Institute of Technology, Vincenz-Prießnitz-Str. 3, 76131 Karlsruhe (Germany); Cacuci, Dan G. [Center for Nuclear Science and Energy/Dept. of ME, University of South Carolina, 300 Main Street, Columbia, SC 29208 (United States)

    2017-03-15

    Highlights: • BWR Turbine Trip 2 (BWR-TT2) benchmark. • Substantial (up to 50%) reduction of uncertainties in the predicted transient power. • 6660 uncertain model parameters were calibrated. - Abstract: By applying a comprehensive predictive modeling methodology, this work demonstrates a substantial (up to 50%) reduction of uncertainties in the predicted total transient power in the BWR Turbine Trip 2 (BWR-TT2) benchmark while calibrating the numerical simulation of this benchmark, comprising 6090 macroscopic cross sections, and 570 thermal-hydraulics parameters involved in modeling the phase-slip correlation, transient outlet pressure, and total mass flow. The BWR-TT2 benchmark is based on an experiment that was carried out in 1977 in the NPP Peach Bottom 2, involving the closure of the turbine stop valve which caused a pressure wave that propagated with attenuation into the reactor core. The condensation of the steam in the reactor core caused by the pressure increase led to a positive reactivity insertion. The subsequent rise of power was limited by the feedback and the insertion of the control rods. The BWR-TT2 benchmark was modeled with the three-dimensional reactor physics code system DYN3D, by coupling neutron kinetics with two-phase thermal-hydraulics. All 6660 DYN3D model parameters were calibrated by applying a predictive modeling methodology that combines experimental and computational information to produce optimally predicted best-estimate results with reduced predicted uncertainties. Simultaneously, the predictive modeling methodology yields optimally predicted values for the BWR total transient power while reducing significantly the accompanying predicted standard deviations.

  15. Numerical simulation of the transient thermal-hydraulic behaviour of the ITER blanket cooling system under the draining operational procedure

    Energy Technology Data Exchange (ETDEWEB)

    Di Maio, P.A. [Dipartimento di Energia, Ingegneria dell’Informazione e Modelli Matematici, Università di Palermo Viale delle Scienze, 90128 Palermo (Italy); Dell’Orco, G.; Furmanek, A. [ITER Organization, Route de Vinon-sur-Verdon, CS 90 046, 13067 St Paul Lez Durance Cedex (France); Garitta, S. [Dipartimento di Energia, Ingegneria dell’Informazione e Modelli Matematici, Università di Palermo Viale delle Scienze, 90128 Palermo (Italy); Merola, M.; Mitteau, R.; Raffray, R. [ITER Organization, Route de Vinon-sur-Verdon, CS 90 046, 13067 St Paul Lez Durance Cedex (France); Spagnuolo, G.A. [Dipartimento di Energia, Ingegneria dell’Informazione e Modelli Matematici, Università di Palermo Viale delle Scienze, 90128 Palermo (Italy); Vallone, E., E-mail: eug.vallone@gmail.com [Dipartimento di Energia, Ingegneria dell’Informazione e Modelli Matematici, Università di Palermo Viale delle Scienze, 90128 Palermo (Italy)

    2015-10-15

    Highlights: • ITER blanket cooling system hydraulic behaviour is studied under draining transient. • A computational approach based on the finite volume method has been followed. • Draining efficiency has been assessed in term of transient duration and residual water. • Transient duration ranges from ∼40 to 50 s, under the reference draining scenario. • Residual water is predicted to range from few tens of gram up to few kilograms. - Abstract: Within the framework of the research and development activities supported by the ITER Organization on the blanket system issues, an intense analysis campaign has been performed at the University of Palermo with the aim to investigate the thermal-hydraulic behaviour of the cooling system of a standard 20° sector of ITER blanket during the draining transient operational procedure. The analysis has been carried out following a theoretical-computational approach based on the finite volume method and adopting the RELAP5 system code. In a first phase, attention has been focused on the development and validation of the finite volume models of the cooling circuits of the most demanding modules belonging to the standard blanket sector. In later phase, attention has been put to the numerical simulation of the thermal-hydraulic transient behaviour of each cooling circuit during the draining operational procedure. The draining procedure efficiency has been assessed in terms of both transient duration and residual amount of coolant inside the circuit, observing that the former ranges typically between 40 and 120 s and the latter reaches at most ∼8 kg, in the case of the cooling circuit of twinned modules #6–7. Potential variations to operational parameters and/or to circuit lay-out have been proposed and investigated to optimize the circuit draining performances. In this paper, the set-up of the finite volume models is briefly described and the key results are summarized and critically discussed.

  16. A review of modern advances in analyses and applications of single-phase natural circulation loop in nuclear thermal hydraulics

    Energy Technology Data Exchange (ETDEWEB)

    Basu, Dipankar N., E-mail: dipankar.n.basu@gmail.com [Department of Mechanical Engineering, Indian Institute of Technology Guwahati, Guwahati 781039 (India); Bhattacharyya, Souvik; Das, P.K. [Department of Mechanical Engineering, Indian Institute of Technology Kharagpur, Kharagpur 721302 (India)

    2014-12-15

    Highlights: • Comprehensive review of state-of-the-art on single-phase natural circulation loops. • Detailed discussion on growth in solar thermal system and nuclear thermal hydraulics. • Systematic development in scaling methodologies for fabrication of test facilities. • Importance of numerical modeling schemes for stability assessment using 1-D codes. • Appraisal of current trend of research and possible future directions. - Abstract: A comprehensive review of single-phase natural circulation loop (NCL) is presented here. Relevant literature reported since the later part of 1980s has been meticulously surveyed, with occasional obligatory reference to a few pioneering studies originating prior to that period, summarizing the key observations and the present trend of research. Development in the concept of buoyancy-induced flow is discussed, with introduction to flow initiation in an NCL due to instability. Detailed discussion on modern advancement in important application areas like solar thermal systems and nuclear thermal hydraulics are presented, with separate analysis for various reactor designs working on natural circulation. Identification of scaling criteria for designing lab-scale experimental facilities has gone through a series of modification. A systematic analysis of the same is presented, considering the state-of-the-art knowledge base. Different approaches have been followed for modeling single-phase NCLs, including simplified Lorenz system mostly for toroidal loops, 1-D computational modeling for both steady-state and stability characterization and 3-D commercial system codes to have a better flow visualization. Methodical review of the relevant studies is presented following a systematic approach, to assess the gradual progression in understanding of the practical system. Brief appraisal of current research interest is reported, including the use of nanofluids for fluid property augmentation, marine reactors subjected to rolling waves

  17. Core thermal hydraulic behavior during the reflood phase of cold-leg LBLOCA experiments using the ATLAS test facility

    Energy Technology Data Exchange (ETDEWEB)

    Cho, Seok; Park, Hyun Sik; Choi, Ki Yong; Kang, Kyoung Ho; Baek, Won Pil; Kim, Yeon Sik [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2009-12-15

    Several experimental tests to simulate a reflood phase of a cold-leg LBLOCA of the APR1400 have been performed using the ATLAS facility. This paper describes the related experimental results with respect to the thermal-hydraulic behavior in the core and the system-core interactions during the reflood phase of the cold-leg LBLOCA conditions. The present descriptions will be focused on the LB-CL-09, LB-CL-11, LB-CL-14, and LB-CL-15 tests performed using the ATLAS. The LB-CL-09 is an integral effect test with conservative boundary condition; the LB-CL-11 and -14 are integral effect tests with realistic boundary conditions, and the LB-CL-15 is a separated effect test. The objectives of these tests are to investigate the thermal-hydraulic behavior during an entire reflood phase and to provide reliable experimental data for validating the LBLOCA analysis methodology for the APR1400. The initial and boundary conditions were obtained by applying scaling ratios to the MARS simulation results for the LBLOCA scenario of the APR1400. The ECC water flow rate from the safety injection tanks and the decay heat were simulated from the start of the reflood phase. The simulated core power was controlled to be 1.2 times that of the ANS-73 decay heat curve for LB-CL-09 and 1.02 times that of the ANS-79 decay curve for LB-CL-11, -14, and -15. The simulated ECC water flow rate from the high pressure safety injection pump was 0.32 kg/s. The present experimental data showed that the cladding temperature behavior is closely related to the collapsed water level in the core and the downcomer

  18. Thermal hydraulic evaluation for an experimental facility to investigate pressurized thermal shock (PTS) in CDTN/CNEN; Avaliacao termo-hidraulica da montagem experimental de choque termico pressurizado do CDTN/CNEN

    Energy Technology Data Exchange (ETDEWEB)

    Palmieri, Elcio T.; Navarro, Moyses A.; Aronne, Ivam D.; Terra, Jose L. [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN), Belo Horizonte, MG (Brazil)

    2002-07-01

    The goal of the work presented in this paper is to provide necessary thermal hydraulics information to the design of an experimental installation to investigate the Pressurized Thermal Shock (PTS) to be implemented at Centro de Desenvolvimento da Tecnologia Nuclear (CDTN/CNEN). The envisaged installation has a test section that represents, in a small scale, a pressure vessel of a nuclear reactor. This test section will be heated and then exposed to a PTS in order to evaluate the appearance and development of cracks. To verify the behavior of the temperatures of the pressure vessel after a sudden flood through the annulus, calculations were made using the RELAP5/MOD 3.2.2 gamma code. Different outer radiuses were studied for the annular region. The results showed that the smaller annulus spacing (20 mm) anticipates the wetting of the surface and produces a higher cooling of the external surface, which stays completely wet for a longer time. (author)

  19. 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.

  20. Experimental studies on dynamic system characteristics of the high temperature/high pressure thermal-hydraulic test facility(VISTA) for the power variation

    Energy Technology Data Exchange (ETDEWEB)

    Choi, K. Y.; Park, H. S.; Joe, S.; Park, C. K.; Lee, S. J.; Song, C. W.; Jeong, M. K. [KAERI, Taejon (Korea, Republic of)

    2003-10-01

    Dynamic system characteristics tests were carried out for the power variation by using the high temperature/high pressure thermal-hydraulic test facility, VISTA(Experimental Verification by Integral Simulation of Transient and Accidents), which had been constructed to simulate the SMART-P by KAERI. Experimental tests have been performed to investigate the thermal-hydraulic dynamic characteristics of the primary and the secondary systems in the range of 5% to 85% power. Automatic PID control logics were developed and installed to the VISTA facility to control the major thermal hydraulic parameters. Power was changed with either a step or a ramp changing method from the reference power of 10%, 25%, 50% and 75% to 5% or 10% higher power. It was found that there is no noticeable difference in the responses between a step and a ramp changing method. When unique constants of P, I, and D were used in the range of 5% to 85% power, it was found to be liable to lose the system control. Further studies are required to quantify the controllability and the time constants of the major thermal hydraulic parameters.

  1. Reflood completion report: Volume 1. A phenomenological thermal-hydraulic model of hot rod bundles experiencing simultaneous bottom and top quenching and an optimization methodology for closure development

    Energy Technology Data Exchange (ETDEWEB)

    Nelson, R.A. Jr.; Pimentel, D.A.; Jolly-Woodruff, S.; Spore, J.

    1998-04-01

    In this report, a phenomenological model of simultaneous bottom-up and top-down quenching is developed and discussed. The model was implemented in the TRAC-PF1/MOD2 computer code. Two sets of closure relationships were compared within the study, the Absolute set and the Conditional set. The Absolute set of correlations is frequently viewed as the pure set because the correlations is frequently viewed as the pure set because the correlations utilize their original coefficients as suggested by the developer. The Conditional set is a modified set of correlations with changes to the correlation coefficient only. Results for these two sets indicate quite similar results. This report also summarizes initial results of an effort to investigate nonlinear optimization techniques applied to the closure model development. Results suggest that such techniques can provide advantages for future model development work, but that extensive expertise is required to utilize such techniques (i.e., the model developer must fully understand both the physics of the process being represented and the computational techniques being employed). The computer may then be used to improve the correlation of computational results with experiments.

  2. 高温气冷堆堆芯实时热工水力模型%Real Time Thermal Hydraulic Model for High Temperature Gas-cooled Reactor Core

    Institute of Scientific and Technical Information of China (English)

    眭喆; 张瑞鹏; 孙俊; 马远乐

    2013-01-01

    A real-time thermal hydraulic model of the reactor core was described and integrated into the simulation system for the high temperature gas-cooled pebble bed reactor nuclear power plant, which was developed in the vPower platform, a new simulation environment for nuclear and fossil power plants. In the thermal hydraulic model, the helium flow paths were established by the flow network tools in order to obtain the flow rates and pressure distributions. Meanwhile, the heat structures, representing all the solid heat transfer elements in the pebble bed, graphite reflectors and carbon bricks, were connected by the heat transfer network in order to solve the temperature distributions in the reactor core. The flow network and heat transfer network were coupled and calculated in real time. Two steady states (100% and 50% full power) and two transients (inlet temperature step and flow step) were tested that the quantitative comparisons of the steady results with design data and qualitative analysis of the transients showed the good applicability of the present thermal hydraulic model.%为建立适用于球床式高温气冷堆核电厂的模拟机,采用一体化仿真支撑平台vPower建立高温气冷堆堆芯的实时热工水力模型,利用流体网络求解氦气流道的流量与压力分布及传热网络求解球床燃料区、石墨反射层区与碳砖区的温度分布,实现整个氦气流场与固相温度场的实时、耦合计算.模拟100%额定负荷和50%额定负荷2个稳态工况和入口温度阶跃和流量阶跃2个动态过程.稳态工况与设计参数的定量对比以及动态过程的定性分析表明,该模型具有较好的适用性.

  3. 应用于反应堆热工水力程序的核态沸腾传热关系式评价%Assessment of Nucleate Boiling Correlations Applied to Thermal-hydraulic Code of Reactor

    Institute of Scientific and Technical Information of China (English)

    李美琳; 林萌; 杨燕华; 张昊; 龚湛

    2015-01-01

    This article is set in development of reactor thermal-hydraulic analysis code-COSINE,studies variation characteristic of results of nucleate boiling correlations usually used for thermal-hydraulic analysis codes with variable changing,compares difference degree of different correlations and sensitivity in different ranges based on fuel rod wall heat transfer and external reactor vessel coolant conditions,in order to provide advice on setting of code user options and further experiment researches. It draws a conclusion that the condition of high superheat degree is mostly needed experimental demonstration and Chen,Schrock-Grossman1,Wright and Schrock-Grossman2 correlations are more suitable for calculating these conditions in reactor thermal-hydraulic analysis code.%本文以反应堆热工水力分析程序 COSINE 开发为背景,针对燃料棒和冷却剂换热及压力容器外部冷却时的核态沸腾两种特殊的工况,研究常用于计算热工水力程序的核态沸腾传热关系式的计算结果随影响参数的变化关系,比较不同范围内各关系式计算结果的差异程度和敏感性,为程序中用户选项的设置和进一步实验验证提供参考意见,研究表明高过热度工况最需进行实验验证,反应堆热工水力分析程序计算这两种工况下的核态沸腾传热更适宜选用 Chen、Schrock-Grossman1、Wright 和 Schrock-Grossman2公式。

  4. 钠冷快堆燃料组件热工水力特性数值模拟与分析%Numerical Simulation and Analysis on Thermal-hydraulic Behavior of Fuel Assembly for Sodium-cooled Fast Reactor

    Institute of Scientific and Technical Information of China (English)

    刘洋; 喻宏; 周志伟

    2014-01-01

    The thermal-hydraulic behavior of triangular arranged fuel bundle with wrapped wire spacer of fuel assembly for sodium-cooled fast reactor was investigated by employing CFD code CFX ,and the results were compared with subchannel analysis code SuperEnergy .Fuel bundles composed of 7,19 ,37 and 61 fuel rods were analyzed sepa-rately .The axial velocity ,cross flow mixing effect ,and temperature rise along axial direction for different subchannels of the fuel bundle were discussed ,and the effect of wrapped wire spacer was carefully investigated .The results show that the wrapped wire spacer plays an important role on the cross flow effect and axial velocity distribution as well as the temperature rise in different subchannels .Moreover ,with the increase of fuel rods ,the flow in fuel bundle becomes more complicated ,and the non-uniformity of the axial flow also shows a tendency to enhance .%利用CFD程序CFX ,分别对7、19、37、61根棒组成的三角形排列螺旋绕丝定位的钠冷快堆燃料组件棒束通道进行了热工水力特性的分析研究,并将结果与子通道程序SuperEnergy进行了对比验证。重点考察了棒束通道轴向流动分布、横向流交混效应及子通道轴向温升,分析了定位绕丝的影响。结果表明,绕丝对棒束通道的横向流交混效应、轴向流动分布及子通道温升有着重要影响,且随棒束的增多,通道内的流动趋向复杂化,轴向流动不均匀性有升高趋势。

  5. 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.

  6. RTDP方法在大型先进压水堆热工设计中的应用初步研究%Preliminary Research on RTDP Methodology for Advanced LPP Thermal-hydraulic Design

    Institute of Scientific and Technical Information of China (English)

    杨萍; 贾红轶; 王喆

    2013-01-01

    Departure from nucleate boiling (DNB) design basis is one of the most important basis for reactor core thermal-hydraulic design.In order to evaluate whether the DNB design basis meets the demand of thermal-hydraulic design,the departure from nucleate boiling ratio (DNBR) design limit should be determined first.The RTDP methodology was described detailedly,in which the uncertainties of operating parameters and nuclear design parameters were statistically combined.Then the RTDP methodology and a reactor subchannel code were applied to calculate the DNBR design limit and quality limit for LPP.The conclusions were presented to provide the key acceptable criterion for DNBR design basis.%偏离泡核沸腾(DNB)设计基准是反应堆热工水力设计中的重要基准之一,为评价该设计基准是否满足热工水力设计要求,首先需确定堆芯偏离泡核沸腾比(DNBR)设计限值.本工作详细论述了使用统计学方法确定运行参数及核设计参数等不确定性的RTDP原理,并应用该方法和堆芯子通道分析程序对大型先进压水堆DNBR设计限值及含汽率限值进行计算并给出结论,为DNBR设计基准的验证提供了关键判据.

  7. Impact of uranium concentration reduction in side plates of the fuel elements of IEA-R1 reactor on neutronic and thermal hydraulic analyses; Impacto da reducao na concentracao de uranio nas placas laterais dos elementos combustiveis do reator IEA-R1 nas analises neutronica e termo-hidraulica

    Energy Technology Data Exchange (ETDEWEB)

    Rios, Ilka Antonia

    2013-09-01

    This master thesis presents a study to verify the impact of the uranium concentration reduction in the side plates of the reactor IEA-R1 fuel elements on the neutronic and thermal-hydraulic analyses. To develop such study, a previous IPEN-CNEN/SP research was reproduced by simulating the fuel elements burn-up, with side plate uranium density reduced to 50, 60 and 70% of the standard fuel element plates. This research begins with the neutronic analysis using the computer code HAMMER and the first step consists in the calculation of the cross section of all materials presented at the reactor core, with their initial concentration; the second step consists in the calculation of the fast and thermal neutron group fluxes and power densities for fuel elements using the computer code CITATION. HAMMER output data is used as input data. Once the neutronic analysis is finished and the most critical fuel elements with highest power density have been defined, the thermal-hydraulics analysis begins. This analysis uses MCTR-IEA-R1 thermal-hydraulics model, which equations are solved by commercial code EES. Thermalhydraulics analysis input is the power density data calculated by CITATION: it is considered the highest power density on each fuel element, where there is a higher energy release and, consequently, higher temperatures. This data is used on energy balance equations to calculate temperatures on critical fuel element regions. Reactor operation comparison for three different uranium densities on fuel side plates is presented. Uranium density reduction contributes to the cladding surface temperature to remain below the established limit, as reactor operation safety requirement and it does not affect significantly fuel element final burn-up nor reactor reactivity. The reduction of uranium in the side plates of the fuel elements of the IEA-R1 showed to be a viable option to avoid corrosion problems due to high temperatures. (author)

  8. Development of a Neutron Radiography Three-Dimensional Computed Tomography System for Void Fraction Measurement of Boiling Flow in Tight Lattice Rod Bundles

    Science.gov (United States)

    Kureta, Masatoshi

    A neutron radiography three-dimensional computed tomography (NR3DCT) system was developed to visualize the void fraction distribution of boiling flow in tight lattice heated-rod bundles. This paper chiefly reports on the data processing and the error estimation method of NR3DCT. Practical γ-ray noise reduction and image correction techniques were studied to improve the reliability of the experimental data. Using the system and a directly heated 14-rod bundle test section, the behavior of boiling flow in a tight lattice rod bundle was clearly visualized. The effect of each data processing step on the result was also discussed. By this development, the three-dimensional vapor distribution of boiling flow in a heated bundle is made clear, and void fraction databases can be provided for verification of a thermal-hydraulic simulation code.

  9. 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

  10. Neutronic and thermal-hydraulic analysis of devices for irradiation of LEU targets type of UAL{sub x}-Al and U-Ni to production of {sup 99}Mo in reactor IEA-R1 and RMB; Analises neutronicas e termo-hidraulica de dispositivos para irradiacao de alvos tipo LEU de UAL{sub x}-Al e U-Ni para producao de {sup 99}Mo nos reatores IEA-R1 e RMB

    Energy Technology Data Exchange (ETDEWEB)

    Domingos, Douglas Borges

    2014-07-01

    In this work neutronic and thermal-hydraulic analyses were made to compare three types of targets (UAl{sub 2}-Al, U-Ni cylindrical and U-Ni plate) used for the production of {sup 99}Mo by fission of {sup 235}U. Some experiments were conducted to validate the neutronic and thermal-hydraulics methodologies used in this work. For the neutronic calculations the computational programs NJOY99.0, AMPX-II and HAMMERTECHNION were used to generate the cross sections. SCALE 6.0 and CITATION computational programs were used for three-dimensional calculations of the reactor cores, fuel burning and the production of {sup 99}Mo. The computational programs MTRCR-IEAR1 and ANSYS CFX were used to calculate the thermal and hydraulic parameters of the irradiation devices and for comparing them to limits and design criteria. First were performed neutronic and thermal-hydraulic analyzes for the reactor IEA-R1 with the targets of UAl{sub 2}-Al (10 mini plates). Analyses have shown that the total activity obtained for {sup 99}Mo on the mini plates does not meet the demand of Brazilian hospitals (450 Ci/week) and that no limit of thermo-hydraulic design is overtaken. Next, the same calculations were performed for the three target types in Multipurpose Brazilian Reactor (MBR). The neutronic analyzes demonstrated that the three targets meet the demand of Brazilian hospitals. The thermal hydraulic analysis shows that a minimum speed of 7 m/s for the target UAl{sub 2}-Al, 8 m/s for the cylindrical target U-Ni and 9 m/s for the target U-Ni plate will be necessary in the irradiation device to not exceed the design limits. Were performed experiments using a test bench for validate the methodologies for the thermal-hydraulic calculation. The experiments performed to validate the neutronic calculations were made in the reactor IPEN/MB-01. All experiments were simulated with the methodologies described above and the results compared. The simulations results showed good agreement with experimental

  11. Parabolic k-ample bundles

    CERN Document Server

    Biswas, Indranil

    2011-01-01

    We construct projectivization of a parabolic vector bundle and a tautological line bundle over it. It is shown that a parabolic vector bundle is ample if and only if the tautological line bundle is ample. This allows us to generalize the notion of a k-ample bundle, introduced by Sommese, to the context of parabolic bundles. A parabolic vector bundle $E_*$ is defined to be k-ample if the tautological line bundle ${\\mathcal O}_{{\\mathbb P}(E_*)}(1)$ is $k$--ample. We establish some properties of parabolic k-ample bundles.

  12. CFD Simulation of Thermal-Hydraulic Benchmark V1000CT-2 Using ANSYS CFX

    OpenAIRE

    2009-01-01

    Plant measured data from VVER-1000 coolant mixing experiments were used within the OECD/NEA and AER coupled code benchmarks for light water reactors to test and validate computational fluid dynamic (CFD) codes. The task is to compare the various calculations with measured data, using specified boundary conditions and core power distributions. The experiments, which are provided for CFD validation, include single loop cooling down or heating-up by disturbing the heat transfer in the steam gene...

  13. KARATE - a code for VVER-440 core calculation

    Energy Technology Data Exchange (ETDEWEB)

    Gado, J.; Hegedus, Cs.J.; Hegyi, Gy.; Kereszturi, A.; Makai, M.; Maraczi, Cs.; Telbisz, M.

    1994-12-31

    A modular calculation system has been elaborated at the KFKI Atomic Energy Research Institute for VVER-440 cores. The purpose of KARATE is the calculation of neutron physical and thermal-hydraulic processes in the core at normal, startup, and slow transient conditions. KARATE is under validation and verification (V&V) against mathematical, experimental, and operational data.

  14. CFD Simulation of Thermal-Hydraulic Benchmark V1000CT-2 Using ANSYS CFX

    Directory of Open Access Journals (Sweden)

    Thomas Höhne

    2009-01-01

    Full Text Available Plant measured data from VVER-1000 coolant mixing experiments were used within the OECD/NEA and AER coupled code benchmarks for light water reactors to test and validate computational fluid dynamic (CFD codes. The task is to compare the various calculations with measured data, using specified boundary conditions and core power distributions. The experiments, which are provided for CFD validation, include single loop cooling down or heating-up by disturbing the heat transfer in the steam generator through the steam valves at low reactor power and with all main coolant pumps in operation. CFD calculations have been performed using a numerical grid model of 4.7 million tetrahedral elements. The Best Practice Guidelines in using CFD in nuclear reactor safety applications has been used. Different advanced turbulence models were utilized in the numerical simulation. The results show a clear sector formation of the affected loop at the downcomer, lower plenum and core inlet, which corresponds to the measured values. The maximum local values of the relative temperature rise in the calculation are in the same range of the experiment. Due to this result, it is now possible to improve the mixing models which are usually used in system codes.

  15. 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.

  16. 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.

  17. 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)

  18. 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)

  19. 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

  20. 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.

  1. An analysis of the proposed MITR-III core to establish thermal-hydraulic limits at 10 MW. Final report

    Energy Technology Data Exchange (ETDEWEB)

    Harling, O.K.; Lanning, D.D.; Bernard, J.A.; Meyer, J.E.; Henry, A.F.

    1997-06-01

    The 5 MW Massachusetts Institute of Technology Research Reactor (MITR-II) is expected to operate under a new license beginning in 1999. Among the options being considered is an upgrade in the heat removal system to allow operation at 10 MW. The purpose of this study is to predict the Limiting Safety System Settings and Safety Limits for the upgraded reactor (MITR-III). The MITR Multi-Channel Analysis Code was written to analyze the response of the MITR system to a series of anticipated transients in order to determine the Limiting Safety System Settings and Safety Limits under various operating conditions. The MIT Multi-Channel Analysis Code models the primary and secondary systems, with special emphasis placed on analyzing the thermal-hydraulic conditions in the core. The code models each MITR fuel element explicitly in order to predict the behavior of the system during flow instabilities. The results of the code are compared to experimental data from MITR-II and other sources. New definitions are suggested for the Limiting Safety System Settings and Safety Limits. MITR Limit Diagrams are included for three different heat removal system configurations. It is concluded that safe, year-round operating at 10 MW is possible, given that the primary and secondary flow rates are both increased by approximately 40%.

  2. Implicit treatment of technical specification and thermal hydraulic parameter uncertainties in Gaussian process model to estimate safety margin

    Energy Technology Data Exchange (ETDEWEB)

    Fynan, Douglas A.; Ahn, Kwang Il [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2016-06-15

    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.

  3. The Thermal-hydraulic Performance Test Report for the Non-instrumented Irradiation Test Rig of Annular Fuel

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Dae Ho; Lee, Kang Hee; Shin, Chang Hwan

    2008-09-15

    This report presents the results of pressure drop test, vibration test and endurance test for the non-instrumented rig using the irradiation test in HANARO of the double cooled annular fuel which were designed and fabricated by KAERI. From the out-pile thermal hydraulic tests, corresponding to the pressure drop of 200 kPa is measured to be about 9.72 kg/sec. Vibration frequency for the non-instrumented rig ranges from 5.0 to 10.7 kg/s. RMS(Root Mean Square) displacement for non-instrumented rig is less than 11.73 m, and the maximum displacement is less than 54.87m. The flow rate for endurance test were 10.5 kg/s, which was 110% of 9.72 kg/s. And the endurance test was carried out for 3 days. The test results found not to the wear and satisfied to the limits of pressure drop, flow rate, vibration and wear in the non-instrumented rig. This test was performed at the FIVPET facility.

  4. Thermal hydraulic performance analysis of the printed circuit heat exchanger using a helium test facility and CFD simulations

    Energy Technology Data Exchange (ETDEWEB)

    Kim, In Hun [Department of Nuclear and Quantum Engineering, Korea Advanced Institute of Science and Technology, 373-1, Guseong-dong, Yuseong-gu, Daejeon 305-701 (Korea, Republic of); No, Hee Cheon [Department of Nuclear and Quantum Engineering, Korea Advanced Institute of Science and Technology, 373-1, Guseong-dong, Yuseong-gu, Daejeon 305-701 (Korea, Republic of)], E-mail: hcno@kaist.ac.kr; Lee, Jeong Ik; Jeon, Byong Guk [Department of Nuclear and Quantum Engineering, Korea Advanced Institute of Science and Technology, 373-1, Guseong-dong, Yuseong-gu, Daejeon 305-701 (Korea, Republic of)

    2009-11-15

    The thermal-hydraulic performance of the PCHE was investigated using the KAIST helium test loop. Experiments were performed in the helium laminar region with 350 < Re < 1200. The hot/cold side inlet conditions were 25-550 {sup o}C/25-100 {sup o}C over the operating pressure of 1.5-1.9 MPa, respectively. Mass flow rates were controlled in the range of 40-100 kg/h. Pressure drop and temperature difference were measured at the inlet and outlet of the hot and cold sides. A global Fanning factor correlation and a global Nusselt number correlation were proposed using information only at the inlet and outlet of the hot and cold sides. A three-dimensional (3-D) numerical simulation was performed using FLUENT, a commercial computational fluid dynamics (CFD) code, to compare simulation results to the KAIST helium test data and to obtain the local Nusselt number in the PCHE. CFD predictions showed good agreement with experimental data. A local pitch-averaged Nusselt number correlation was proposed using local temperature, pressure, surface heat fluxes, and properties provided by CFD simulations. The system analysis code, GAMMA, was also utilized to identify which correlation was more applicable for system analysis. It turns out that the proposed local pitch-averaged Nusselt number correlation from CFD simulations is more appropriate than the global Nusselt number correlation developed from experimental data.

  5. A three-dimensional transient neutronics routine for the TRAC-PF1 reactor thermal hydraulic computer code

    Energy Technology Data Exchange (ETDEWEB)

    Bandini, B.R. [Los Alamos National Lab., NM (United States)

    1990-05-01

    No present light water reactor accident analysis code employs both high state of the art neutronics and thermal-hydraulics computational algorithms. Adding a modern three-dimensional neutron kinetics model to the present TRAC-PFI/MOD2 code would create a fully up to date pressurized water reactor accident evaluation code. After reviewing several options, it was decided that the Nodal Expansion Method would best provide the basis for this multidimensional transient neutronic analysis capability. Steady-state and transient versions of the Nodal Expansion Method were coded in both three-dimensional Cartesian and cylindrical geometries. In stand-alone form this method of solving the few group neutron diffusion equations was shown to yield efficient and accurate results for a variety of steady-state and transient benchmark problems. The Nodal Expansion Method was then incorporated into TRAC-PFl/MOD2. The combined NEM/TRAC code results agreed well with the EPRI-ARROTTA core-only transient analysis code when modelling a severe PWR control rod ejection accident.

  6. Simulation of A Main Steam Line Break Accident Using the Coupled 'System Thermal-Hydraulics, 3D reactor Kinetics, and Hot Channel' Analysis Capability of MARS 3.0

    Energy Technology Data Exchange (ETDEWEB)

    Jeong, Jae Jun; Chung, Bub Dong

    2005-09-15

    For realistic analysis of thermal-hydraulics (T-H) transients in light water reactors, KAERI has developed the best-estimate T-H system code, MARS. The code has been improved from the consolidated version of the RELAP5/MOD3 and COBRA-TF codes. Then, the MARS code was coupled with a three-dimensional (3-D) reactor kinetics code, MASTER. This coupled calculation feature, in conjunction with the existing hot channel analysis capabilities of the MARS and MASTER codes, allows for more realistic simulations of nuclear system transients. In this work, a main steam line break (MSLB) accident is simulated using the coupled 'system T-H, 3-D reactor kinetics, and hot channel analysis' feature of the MARS code. Two coupled calculations are performed for demonstration. First, a coupled calculation of the 'system T-H and 3-D reactor kinetics' with a refined core T-H nodalization is carried out to obtain global core power and local departure from nucleate boiling (DNB) ratio (DNBR) behaviors. Next, for a more accurate DNBR prediction, another coupled calculation with subchannel meshes for the hot channels is performed. The results of the coupled calculations are very reasonable and consistent so that these can be used to remove the excessive conservatism in the conventional safety analysis.

  7. Application of the thermal-hydraulic codes in VVER-440 steam generators modelling

    Energy Technology Data Exchange (ETDEWEB)

    Matejovic, P.; Vranca, L.; Vaclav, E. [Nuclear Power Plant Research Inst. VUJE (Slovakia)

    1995-12-31

    Performances with the CATHARE2 V1.3U and RELAP5/MOD3.0 application to the VVER-440 SG modelling during normal conditions and during transient with secondary water lowering are described. Similar recirculation model was chosen for both codes. In the CATHARE calculation, no special measures were taken with the aim to optimize artificially flow rate distribution coefficients for the junction between SG riser and steam dome. Contrary to RELAP code, the CATHARE code is able to predict reasonable the secondary swell level in nominal conditions. Both codes are able to model properly natural phase separation on the SG water level. 6 refs.

  8. 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.

  9. Experimental studies on heat transfer characteristics and natural circulation performance of PRHRS of the high temperature and high pressure thermal-hydraulic test facility

    Energy Technology Data Exchange (ETDEWEB)

    Park, H. S.; Choi, K. Y.; Joe, S.; Park, C. K.; Lee, S. J.; Song, C. W.; Jeong, M. K. [KAERI, Taejon (Korea, Republic of)

    2003-10-01

    Several experiments are performed to investigate the heat transfer characteristics and natural circulation performance of passive residual removal system (PRHRS) of the high temperature and high pressure thermal-hydraulic test facility. Especially the natural circulation performance of PRHRS, the heat transfer characteristics of PRHRS heat exchangers and emergency cooldown tank (ECT), and the thermal-hydraulic behavior of the primary loop are investigated in detail. The coolant flows steadily in the natural circulation loop which is composed of the steam generator (SG) primary side, the secondary system, and the PRHRS. The heat transfers through the PRHRS heat exchanger and ECT are sufficient enough to enable the natural circulation of the coolant. Also the experimental results show that the core decay heat are sufficiently removed with the operation of the PRHRS.

  10. Development of a preliminary PIRT (Phenomena Identification and Ranking Table) of thermal-hydraulic phenomena for 330MWt SMART integral reactor

    Energy Technology Data Exchange (ETDEWEB)

    Chung, B. D.; Lee, W. J.; Sim, S. K.; Song, J. H.; Kim, H. C.

    1997-09-01

    The work reported in this document identifies the thermal-hydraulic phenomena that are expected to occur during a number of key transients in a 330 MWt SMART integral reactor which is under development at KAERI. The result of this efforts 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 the consensus of KAERI expert panelists and AHP (Analytical Hierarchy Process). Preliminary PIRT developed in this report is intended for use to identify and integrate development areas of further experimental tests needed and thermal-hydraulic models and correlations and code improvements for the safety analysis of the SMART integral reactor. (author). 7 refs., 21 tabs., 22 figs.

  11. Preliminary study of CANDU moderator thermal hydraulics using the CUPID code

    Energy Technology Data Exchange (ETDEWEB)

    Park, Sang Gi; Jeong Jae Jun [Pusan National Univ., Busan (Korea, Republic of); Lee, Jae Ryong; Kim, Hyoung Tae [KAERI, Daejeon (Korea, Republic of)

    2012-10-15

    When the moderator cooling system fails, moderator may act as to remove decay heat which occurs in fuel. During loss of coolant accident (LOCA), the film boiling occurs in the Calandria tube (CT) because the hot pressure tube would deform into contacting with the calandria tube. And lower subcooling would decrease the margin of the CT to dryout. So, it is important to estimate a local subcooling of the moderator inside the Calandria vessel. However, in order to predict the internal temperature the study of empirical experiments and calculations are needed because only the inlet/outlet temperature can be measured in real reactor. In this study, the internal flow of the moderator was predicted by using the CUPID code, which has been developed in KAERI. The CUPID adopts three dimensional, transient, two phase and three field model, and includes various physical models and correlations of the interfacial mass, momentum and energy transfer for the closure relations of the two fluid model. The CUPID code shows single phase and two phase flow through two phase flow calculations of virtual can be applied.

  12. Modelling of thermal hydraulics in a KAROLINA calorimeter for its calibration methodology validation

    Directory of Open Access Journals (Sweden)

    Luks Aleksandra

    2016-12-01

    Full Text Available Results of numerical calculations of heat exchange in a nuclear heating detector for nuclear reactors are presented in this paper. The gamma radiation is generated in nuclear reactor during fission and radiative capture reactions as well as radioactive decay of its products. A single-cell calorimeter has been designed for application in the MARIA research reactor in the National Centre for Nuclear Research (NCBJ in Świerk near Warsaw, Poland, and can also be used in the Jules Horowitz Reactor (JHR, which is under construction in the research centre in Cadarache, France. It consists of a cylindrical sample, which is surrounded by a gas layer, contained in a cylindrical housing. Additional calculations had to be performed before its insertion into the reactor. Within this analysis, modern computational fluid dynamics (CFD methods have been used for assessing important parameters, for example, mean surface temperature, mean volume temperature, and maximum sample (calorimeter core temperature. Results of an experiment performed at a dedicated out-of-pile calibration bench and results of numerical modelling validation are also included in this paper.

  13. Study of a particle method for thermal-hydraulic analysis. 2

    Energy Technology Data Exchange (ETDEWEB)

    Oka, Yoshiaki; Koshizuka, Seiichi [Tokyo Univ., Tokai, Ibaraki (Japan). Nuclear Engineering Research Lab.

    1997-03-01

    In liquid metal fast breeder reactors (LMFBRs), liquid metal of sodium is used as the coolant under the atmospheric pressure. Thus, the coolant system has free surfaces in the components. In addition, the structures should be thin enough to reduce thermal stresses because the coolant is used in a wide range of temperature. Therefore, troubles may take place due to the sloshing, its interactions with structures and fluid-structure coupling vibration induced by flows. However, there have been no numerical methods to analyze large deformations of free surfaces and structures. Moving Particle Semi-implicit (MPS) method can be applied to topological change as well as large deformations of continuum since the calculation is based on macroscopic particles. We have developed an algorithm for incompressible flow analysis and flows with wave breaking on a free surface were successfully calculated. The objectives of the present study are development of the MPS method to analyze fluid-structure interactions and analysis of sloshing in a tank made of elastic walls. As a conclusion , a numerical method for fluid-structure interactions with large deformations of free surfaces and structures is developed based on the MPS method in the present study. (J.P.N.)

  14. Bundling of harvesting residues and whole-trees and the treatment of bundles; Hakkuutaehteiden ja kokopuiden niputus ja nippujen kaesittely

    Energy Technology Data Exchange (ETDEWEB)

    Kaipainen, H.; Seppaenen, V.; Rinne, S.

    1996-12-31

    The conditions on which the bundling of the harvesting residues from spruce regeneration fellings would become profitable were studied. The calculations showed that one of the most important features was sufficient compaction of the bundle, so that the portion of the wood in the unit volume of the bundle has to be more than 40 %. The tests showed that the timber grab loader of farm tractor was insufficient for production of dense bundles. The feeding and compression device of the prototype bundler was constructed in the research and with this device the required density was obtained.The rate of compaction of the dry spruce felling residues was about 40 % and that of the fresh residues was more than 50 %. The comparison between the bundles showed that the calorific value of the fresh bundle per unit volume was nearly 30 % higher than that of the dry bundle. This means that the treatment of the bundles should be done of fresh felling residues. Drying of the bundles succeeded well, and the crushing and chipping tests showed that the processing of the bundles at the plant is possible. The treatability of the bundles was also excellent. By using the prototype, developed in the research, it was possible to produce a bundle of the fresh spruce harvesting residues, the diameter of which was about 50 cm and the length about 3 m, and the rate of compaction over 50 %. By these values the reduction target of the costs is obtainable

  15. Application of thermal-hydraulic codes in the nuclear sector; Aplicaciones de los codigos termo-hidraulicos en el sector nuclear espanol

    Energy Technology Data Exchange (ETDEWEB)

    Queral, C.; Coriso, M.; Garcia Sedano, P. J.; Ruiz, J. A.; Posada, J. M.; Jimenez Varas, G.; Sol, I.; Herranz, L. E.

    2011-07-01

    Use of thermal-hydraulic codes is extended all over many different aspects of nuclear engineering. This article groups and briefly describes the main features of some of the well known codes as an introduction to their recent applications in the Spain nuclear sector. the broad range and quality of applications highlight the maturity achieved both in industry and research organizations and universities within the Spanish nuclear sector. (Author)

  16. Influence of Nodalization on Major Thermal-Hydraulic Variables Governing Mark-I Containment Response to an SBO Sequence: Plant Assessment

    Energy Technology Data Exchange (ETDEWEB)

    Herranz, L. E.; Fontanet, J.; Fernandez, E.; Lopez, C.

    2014-07-01

    Simulating a severe accident with lumped parameter codes, like MELCOR 2.1, requires defining a suitable nodalization of the plant. In particular, the number of nodes the wet-well (WW) is divided into and the way these nodes are interconnected have an important effect on thermal-hydraulics estimates of the containment. Therefore, a quantitative assessment of the effect of WW nodalization on accident sequence is needed when performing plant analyses. (Author)

  17. 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)

  18. Screening and metamodeling of computer experiments with functional outputs. Application to thermal-hydraulic computations

    CERN Document Server

    Auder, Benjamin; Iooss, Bertrand; Marques, Michel

    2010-01-01

    To perform uncertainty, sensitivity or optimization analysis on scalar variables calculated by a cpu time expensive computer code, a widely accepted methodology consists in first identifying the most influential uncertain inputs (by screening techniques), and then in replacing the cpu time expensive model by a cpu inexpensive mathematical function, called a metamodel. This paper extends this methodology to the functional output case, for instance when the model output variables are curves. The screening approach is based on the analysis of variance and principal component analysis of output curves. The functional metamodeling consists in a curve classification step, a dimension reduction step, then a classical metamodeling step. An industrial nuclear reactor application (dealing with uncertainties in the pressurized thermal shock analysis) illustrates all these steps.

  19. Methodology of a PWR containment analysis during a thermal-hydraulic accident

    Energy Technology Data Exchange (ETDEWEB)

    Silva, Dayane F.; Sabundjian, Gaiane; Lima, Ana Cecilia S., E-mail: dayane.silva@usp.br, E-mail: gdjian@ipen.br, E-mail: aclima@ipen.br [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil)

    2015-07-01

    The aim of this work is to present the methodology of calculation to Angra 2 reactor containment during accidents of the type Loss of Coolant Accident (LOCA). This study will be possible to ensure the safety of the population of the surroundings upon the occurrence of accidents. One of the programs used to analyze containment of a nuclear plant is the CONTAIN. This computer code is an analysis tool used for predicting the physical conditions and distributions of radionuclides inside a containment building following the release of material from the primary system in a light-water reactor during an accident. The containment of the type PWR plant is a concrete building covered internally by metallic material and has limits of design pressure. The methodology of containment analysis must estimate the limits of pressure during a LOCA. The boundary conditions for the simulation are obtained from RELAP5 code. (author)

  20. Statistical Safety Evaluation of BWR Turbine Trip Scenario Using Coupled Neutron Kinetics and Thermal Hydraulics Analysis Code SKETCH-INS/TRACE5.0

    Science.gov (United States)

    Ichikawa, Ryoko; Masuhara, Yasuhiro; Kasahara, Fumio

    The Best Estimate Plus Uncertainty (BEPU) method has been prepared for the regulatory cross-check analysis at Japan Nuclear Energy Safety Organization (JNES) on base of the three-dimensional neutron-kinetics/thermal- hydraulics coupled code SKETCH-INS/TRACE5.0. In the preparation, TRACE5.0 is verified against the large-scale thermal-hydraulic tests carried out with NUPEC facility. These tests were focused on the pressure drop of steam-liquid two phase flow and void fraction distribution. From the comparison of the experimental data with other codes (RELAP5/MOD3.3 and TRAC-BF1), TRACE5.0 was judged better than other codes. It was confirmed that TRACE5.0 has high reliability for thermal hydraulics behavior and are used as a best-estimate code for the statistical safety evaluation. Next, the coupled code SKETCH-INS/TRACE5.0 was applied to turbine trip tests performed at the Peach Bottom-2 BWR4 Plant. The turbine trip event shows the rapid power peak due to the voids collapse with the pressure increase. The analyzed peak value of core power is better simulated than the previous version SKETCH-INS/TRAC-BF1. And the statistical safety evaluation using SKETCH-INS/TRACE5.0 was applied to the loss of load transient for examining the influence of the choice of sampling method.

  1. Preliminary experimental results using the thermal-hydraulic integral test facility (VISTA) for the pilot plant of the system integrated modular advanced reactor, SMART-P

    Energy Technology Data Exchange (ETDEWEB)

    Choi, Ki Yong; Pak, Hyun Sik; Cho, Seok; Pak, Choon Kyung; Lee, Sung Jae; Song, Chul Hwa; Chung, Moon Ki [KAERI, Taejon (Korea, Republic of)

    2003-07-01

    Preliminary experimental tests were carried out using the thermal-hydraulic integral test facility, VISTA (Experimental Verification by Integral Simulation of Transients and Accidents), which has been constructed to simulate the SMART-P. The VISTA facility is an integral test facility including the primary and secondary systems as well as safety-related Passive Residual heat removal (PRHR) systems. Its scaled ratio with respect to the SMART-P is 1/1 in height and 1/96 in volume and heater power. So far, several steady states and transient tests have been carried out to verify the overall thermal hydraulic primary and secondary characteristics in a range of 10% to 100% power operation. As results of preliminary results, the steady state conditions were found to coincide with the expected design values of the SMART-P. But the major thermal hydraulic parameters are greatly affected by the initial water level and the nitrogen pressure in the reactor upper annular cavity. In the PRHR transient tests, the steam inlet temperature of the PRHR system is found to drop suddenly from a superheated condition to a saturated condition at the end period of PRHR operation.

  2. Thermal properties for the thermal-hydraulics analyses of the BR2 maximum nominal heat flux.

    Energy Technology Data Exchange (ETDEWEB)

    Dionne, B.; Kim, Y. S.; Hofman, G. L. (Nuclear Engineering Division)

    2011-05-23

    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 {sup 235}U) to LEU (19.75% enriched in {sup 235}U) 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. This section is regrouping all of the thermal property tables. Section 2 provides a summary of the thermal properties in form of tables while the following sections 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 shows the evolution of the BR2 maximum heat flux with burnup.

  3. Thermal-hydraulic instabilities in pressure tube graphite - moderated boiling water reactors

    Energy Technology Data Exchange (ETDEWEB)

    Tsiklauri, G.; Schmitt, B.

    1995-09-01

    Thermally induced two-phase instabilities in non-uniformly heated boiling channels in RBMK-1000 reactor have been analyzed using RELAP5/MOD3 code. The RELAP5 model of a RBMK-1000 reactor was developed to investigate low flow in a distribution group header (DGH) supplying 44 fuel pressure tubes. The model was evaluated against experimental data. The results of the calculations indicate that the period of oscillation for the high power tube varied from 3.1s to 2.6s, over the power range of 2.0 MW to 3.0 MW, respectively. The amplitude of the flow oscillation for the high powered tube varied from +100% to -150% of the tube average flow. Reverse flow did not occur in the lower power tubes. The amplitude of oscillation in the subcooled region at the inlet to the fuel region is higher than in the saturated region at the outlet. In the upper fuel region and outlet connectors the flow oscillations are dissipated. The threshold of flow instability for the high powered tubes of a RBMK reactor is compared to Japanese data and appears to be in good agreement.

  4. Characteristics of thermal hydraulic stability in a HYPER system with enhanced natural circulation potential

    Energy Technology Data Exchange (ETDEWEB)

    Tak, Nam Il; Park, Won S.; Han, Seok Jung

    1999-06-01

    Pb-Bi eutectic chosen as a coolant of HYPER is an excellent heat transfer medium but requires relatively large pumping power. Thus the mixed cooling concept to increase economy and safety is being considered for HYPER. In this cooling concept, a large fraction of total thermal power is carried by natural circulation. However, the mixed cooling concept has been considered for conceptual designs only an it has never been applied to real reactors. The purpose of the present study is to provide simple tools to analyze mixed flow and to examine fundamental stability characteristics of mixed flow. Conventional one-dimensional approaches using mass, momentum, and energy conservation are used to describe a forced circulating flow affected by a large buoyancy force. The results of simple analysis using preliminary design parameters of HYPER show that cooling by mixed flow is possible only when the total pressure loss of system is sufficiently low. The stability behavior of mixed flow in a simple rectangular loop has been studied using numerical solutions of the governing equations. As in the case of natural circulation, three types of flow regions, such as stable, neutrally stable, and unstable regions, were found. The stability map of mixed flow has been obtained using the results of calculations. Forced flow due to the pump is found to increase the stability of the loop, since the stable portion of the stability map is increased. However, the unstable region of the mixed flow does not completely disappear, even though the pump exists. (author). 37 refs., 4 tabs., 23 figs.

  5. 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)

  6. A study of return to saturation oscillations in the OSU APEX thermal hydraulic testing facility

    Science.gov (United States)

    Franz, Scott Cameron

    The purpose of this paper is to describe the flow oscillations which occur in the AP600 long term cooling test facility at Oregon State University. The AP600 system is an advanced pressurized water reactor design utilizing passive emergency cooling systems. A few hours after the initiation of a cold leg break, the passive cooling systems inject gravity fed cold water at a rate allowing steam production in the reactor vessel. Steam production in the core causes the pressure in the upper head to increase leading to flow oscillations in all the connecting reactor systems. This paper will show that the oscillations have a definite region of onset and termination for specific conditions in the APEX testing facility. Tests performed at high powers, high elevation breaks, and small break sizes do not exhibit oscillations. The APOS (Advanced Plant Oscillation Simulator) computer code has been developed using a quasi-steady state analysis for flows and a transient analysis for the core node energy balance. The pressure in the reactor head is calculated using a modified perfect gas analysis. For tank liquid inventories, a simple conservation of mass analysis is used to estimate the tank elevations. Simulation logic gleaned from APEX data and photographic evidence have been incorporated into the code to predict termination of the oscillations. Areas which would make the work more complete include a better understanding of two-phase fluid behavior for a top offtake on a pipe, more instrumentation in the core region of the APEX testing facility, and a clearer understanding of fluid conditions in the reactor barrel. Scaling of the oscillations onset and pressure amplitude are relatively straightforward, but termination and period are difficult to scale to the full AP600 plant. Differences in the core power profile and other geometrical differences between the testing facility and the actual plant make the scaling of this phenomenon to the actual plant conditions very difficult.

  7. Interface requirements for coupling a containment code to a reactor system thermal hydraulic codes

    Energy Technology Data Exchange (ETDEWEB)

    Baratta, A.J.

    1997-07-01

    To perform a complete analysis of a reactor transient, not only the primary system response but the containment response must also be accounted for. Such transients and accidents as a loss of coolant accident in both pressurized water and boiling water reactors and inadvertent operation of safety relief valves all challenge the containment and may influence flows because of containment feedback. More recently, the advanced reactor designs put forth by General Electric and Westinghouse in the US and by Framatome and Seimens in Europe rely on the containment to act as the ultimate heat sink. Techniques used by analysts and engineers to analyze the interaction of the containment and the primary system were usually iterative in nature. Codes such as RELAP or RETRAN were used to analyze the primary system response and CONTAIN or CONTEMPT the containment response. The analysis was performed by first running the system code and representing the containment as a fixed pressure boundary condition. The flows were usually from the primary system to the containment initially and generally under choked conditions. Once the mass flows and timing are determined from the system codes, these conditions were input into the containment code. The resulting pressures and temperatures were then calculated and the containment performance analyzed. The disadvantage of this approach becomes evident when one performs an analysis of a rapid depressurization or a long term accident sequence in which feedback from the containment can occur. For example, in a BWR main steam line break transient, the containment heats up and becomes a source of energy for the primary system. Recent advances in programming and computer technology are available to provide an alternative approach. The author and other researchers have developed linkage codes capable of transferring data between codes at each time step allowing discrete codes to be coupled together.

  8. The ABCDEF Implementation Bundle

    Directory of Open Access Journals (Sweden)

    Annachiara Marra

    2016-08-01

    Full Text Available Long-term morbidity, long-term cognitive impairment and hospitalization-associated disability are common occurrence in the survivors of critical illness, with significant consequences for patients and for the caregivers. The ABCDEF bundle represents an evidence-based guide for clinicians to approach the organizational changes needed for optimizing ICU patient recovery and outcomes. The ABCDEF bundle includes: Assess, Prevent, and Manage Pain, Both Spontaneous Awakening Trials (SAT and Spontaneous Breathing Trials (SBT, Choice of analgesia and sedation, Delirium: Assess, Prevent, and Manage, Early mobility and Exercise, and Family engagement. The purpose of this review is to describe the core features of the ABCDEF bundle.

  9. Idaho National Engineering Laboratory (INEL) technical review of YGN 3 and 4 thermal-hydraulic relative size effects

    Energy Technology Data Exchange (ETDEWEB)

    Ward, L.W.; Fineman, C.P.; Gruen, G.E.

    1989-08-01

    Combustion Engineering, Inc., (CE) and the Korean Advanced Energy Research Institute (KAERI) are jointly designing two 2825 MW{sub t} System 80 nuclear steam supply systems for construction in Korea. The two 2825 MW{sub t} plants are similar in design to the larger System 80 class of plants but are reduced in size from 3817 MW{sub t}. These plants will be operated by the Korean Electric Power Company and have been designated as Yonggwang Nuclear Units 3 and 4. The Idaho National Engineering Laboratory (INEL) was selected by CE to perform a third party independent technical review of the thermal-hydraulic safety analyses for Yonggwang Units 3 and 4. The purpose of the review is to establish the acceptability of the safety analyses addressing the differences in size between the 2825 and 3817 MW{sub t} CE designed System 80 plants. The analysis methods used by Combustion Engineering, Inc. were also reviewed to assure that only United States Nuclear Regulatory Commission approved methods were used for the Yonggwang Units 3 and 4 safety analyses and that the methods were applied in a manner consistent with that for the Palo Verde System 80 plants, currently in operation in the US. In general, it was found that the differences between 3817 and 2825 MW{sub t} units led to increased margins except for the large break LOCA (LBLOCA) and boron dilution transient. For the LBLOCA, use of improved models enhanced performance which allowed an increase in peak linear heat generation rate relative to that for the 3817 MW{sub t} plant. For the boron dilution event, an increase in the shutdown margin was necessary to assure the same time to criticality as that for the 3817 MW{sub t} plant. 39 refs., 9 figs., 4 tabs.

  10. CATHARE-3: A new system code for thermal-hydraulics in the context of the NEPTUNE project

    Energy Technology Data Exchange (ETDEWEB)

    Emonot, P., E-mail: philippe.emonot@cea.fr [CEA DEN/DER/SSTH, 17 rue des Martyrs, 38054 Grenoble Cedex 9 (France); Souyri, A., E-mail: annick.souyri@edf.fr [EDF R and D/MFEE, 6 Quai Watier, 78401 Chatou Cedex (France); Gandrille, J.L., E-mail: jeanluc.gandrille@areva.com [AREVA-NP, Tour Areva, 92084 Paris La Defense Cedex (France); Barre, F., E-mail: francois.barre@irsn.fr [IRSN DPAM, BP 3, 13115 Saint-Paul-Lez-Durance Cedex (France)

    2011-11-15

    After a thorough analysis of the industrial needs and of the limitations of current simulation tools, EDF and CEA (Commissariat a l'Energie Atomique) launched the NEPTUNE Project in 2001 (see) with the support of AREVA-NP and IRSN. The NEPTUNE activities include software development, research in physical modeling and numerical methods, development of advanced instrumentation techniques and new experimental programs. Four different simulation scales were addressed including DNS (Direct Numerical Simulation), CFD in open medium (Computational Fluid Dynamics), component (subchannel-type analysis) and system (reactor modeling) scales. In 2006 CEA, EDF, AREVA-NP and IRSN defined the strategy for the system scale of NEPTUNE and the CATHARE-3 development was launched. The main objectives are: Bullet advanced physical modeling of two-phases flows, mainly by using multi-field and turbulence models, Bullet improved 3D modeling by the use of fine and non conforming structured meshes, Bullet generalized coupling capabilities with other thermal-hydraulic scales and with other disciplines (core physics, structural mechanics, Horizontal-Ellipsis), Bullet extension of the applicability to new Gen IV reactors (Sodium Cooled Fast Breeder Reactors, Gas Cooled Reactors, Supercritical Light Water Reactors), Bullet a true object-oriented code architecture. At the same time CATHARE-3 is in continuity with the CATHARE-2 code which is the current industrial version of CATHARE and internationally used for nuclear power plant safety analysis, in simulators and in coupled simulation tools. The road map of these two codes will allow a smooth transition from CATHARE-2 to CATHARE-3 for all users. This paper gives an overview of the choices made for the development of CATHARE-3 including new physical models, validation strategy and experimental programs, numerical improvements, enhanced coupling capability and software architecture evolution. The current status of the project as well as the

  11. 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.

  12. Acquired experience on organizing 3D S.UN.COP: international course to support nuclear license by user training in the areas of scaling, uncertainty, and 3D thermal-hydraulics/neutron-kinetics coupled codes

    Energy Technology Data Exchange (ETDEWEB)

    Petruzzi, Alessandro; D' Auria, Francesco [University of Pisa, San Piero a Grado (Italy). Nuclear Research Group San Piero a Grado (GRNSPG); Galetti, Regina, E-mail: regina@cnen.gov.b [National Commission for Nuclear Energy (CNEN), Rio de Janeiro, RJ (Brazil); Bajs, Tomislav [University of Zagreb (Croatia). Fac. of Electrical Engineering and Computing. Dept. of Power Systems; Reventos, Francesc [Technical University of Catalonia, Barcelona (Spain). Dept. of Physics and Nuclear Engineering

    2011-07-01

    Thermal-hydraulic system computer codes are extensively used worldwide for analysis of nuclear facilities by utilities, regulatory bodies, nuclear power plant designers, vendors, and research organizations. Computer code user represents a source of uncertainty that may significantly affect the results of system code calculations. Code user training and qualification represent an effective means for reducing the variation of results caused by the application of the codes by different users. This paper describes the experience in applying a systematic approach to training code users who, upon completion of the training, should be able to perform calculations making the best possible use of the capabilities of best estimate codes. In addition, this paper presents the organization and the main features of the 3D S.UN.COP (scaling, uncertainty, and 3D coupled code calculations) seminars during which particular emphasis is given to practical applications in connection with the licensing process of best estimate plus uncertainty methodologies, showing the designer, utility and regulatory approaches. (author)

  13. Development of improved thermal hydraulics and fuel performance technology; development of turbulence model for flow analysis in nuclear fuel assembly

    Energy Technology Data Exchange (ETDEWEB)

    Joo, W. K.; Kong, D. W.; Park, H. Z. [Yonsei University, Seoul (Korea)

    2001-04-01

    The flow through a nuclear rod bundle with mixing vanes are very complex and required a suitable turbulence model to be predicted accurately. Subchannel flow in a nuclear bundle having vanes to mix flow appears complex turbulent flow. Objective of this study is to develop turbulence model which can predict complex flow. Also, the module will be produced, which can implement the developed turbulence model in the CFX code. The selected turbulence models are k-epsilon model, non-linear k-epsilon model, Reynolds stress model and modified Reynolds stress model to test their performance in the prediction of the flow in nuclear assembly. These models are tested for a 2-D backwise step flow, square duct flow, rod bundle flow and subchannel flow using CFX. The modules, which can implement Reynolds stress model and non-linear k-epsilon odel in CFX code, are produced. The advantages and disadvantages for these turbulence models are described and the limitation of implementation of non-linear model in CFX code is discussed. The results obtained from the research would give a help for the development of turbulence model which can accurately predict the flow through the rod bundles with mixing vanes. 18 refs., 37 figs., 8 tabs. (Author)

  14. Reflood experiments in rod bundles with flow blockages due to clad ballooning

    Energy Technology Data Exchange (ETDEWEB)

    Moon, S.K.; Kim, J.; Kim, K.; Kim, B.J.; Park, J.K.; Youn, Y.J.; Choi, H.S.; Song, C.H. [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2016-07-15

    Clad ballooning and the resulting partial flow blockage are one of the major thermal-hydraulic concerns associated with the coolability of partially blocked cores during a loss-of-coolant accident (LOCA). Several in-pile tests have shown that fuel relocation causes a local power accumulation and a high thermal coupling between the clad and fuel debris in the ballooned regions. However, previous experiments in the 1980s did not take into account the fuel relocation phenomena and resulting local power increase in the ballooned regions. The present paper presents the results of systematic investigations on the coolability of rod bundles with flow blockages. The experiments were mainly performed in 5 x 5 rod bundles, 2 x 2 rod bundles and other test facilities. The experiments include a reflood heat transfer, single-phase convective heat transfer, flow redistributions phenomena, and droplet break-up behavior. The effects of the fuel relocation and resulting local power increase were investigated using a 5 x 5 rod bundle. The fuel relocation phenomena increase the peak cladding temperature.

  15. Measurements of Flow Mixing at Subchannels in a Wire-Wrapped 61-Rod Bundle for a Sodium Cooled Fast Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Dong Won; Kim, Hyungmo; Ko, Yung Joo; Choi, Hae Seob; Euh, Dong-Jin; Jeong, Ji-Young; Lee, Hyeong-Yeon [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-05-15

    For a safety analysis in a core thermal design of a sodium-cooled fast reactor (SFR), flow mixing characteristics at subchannels in a wire-wrapped rod bundle are crucial factor for the design code verification and validation. Wrapped wires make a cross flow in a circumference of the fuel rod, and this effect lets flow be mixed. Therefore the sub-channel analysis method is commonly used for thermal hydraulic analysis of a SFR, a wire wrapped sub-channel type. To measure flow mixing characteristics, a wire mesh sensing technique can be useful method. A wire mesh sensor has been traditionally used to measure the void fraction of a two-phase flow field, i.e. gas and liquid. However, the recent reports that the wire mesh sensor can be used successfully to recognize the flow field in liquid phase by injecting a tracing liquid with a different level of electric conductivity. The subchannel flow characteristics analysis method is commonly used for the thermal hydraulic analysis of a SFR, a wire wrapped subchannel type. In this study, mixing experiments were conducted successfully at a hexagonally arrayed 61-pin wire-wrapped fuel rod bundle test section. Wire mesh sensor was used to measure flow mixing characteristics. The developed post-processing method has its own merits, and flow mixing results were reasonable.

  16. Subtleties Concerning Conformal Tractor Bundles

    CERN Document Server

    Graham, C Robin

    2012-01-01

    The realization of tractor bundles as associated bundles in conformal geometry is studied. It is shown that different natural choices of principal bundle with normal Cartan connection corresponding to a given conformal manifold can give rise to topologically distinct associated tractor bundles for the same inducing representation. Consequences for homogeneous models and conformal holonomy are described. A careful presentation is made of background material concerning standard tractor bundles and equivalence between parabolic geometries and underlying structures.

  17. 中国HCCB-TBM氚增殖球床热工水力学特性数值模拟%Numerical simulation on thermal hydraulics characteristics of breeder pebble bed for China HCCB-TBM

    Institute of Scientific and Technical Information of China (English)

    汪卫华; 程德胜; 冯开明; 邓海飞

    2014-01-01

    基于CFD软件平台,针对中国HCCB-TBM氚增殖区球床热工水力学特性开展3维数值模拟研究。依据ITER实际运行工况给出吹氚氦气和结构冷却剂氦气在硅酸锂球床内的流动与传热特性,获取球床内详细的速度分布、温度分布和压力降。计算结果表明:圆球的排列方式影响球床内氦气流场和球床的最高温度;ITER运行工况下HCCB-TBM增殖区硅酸锂小球及其壁面的最高温度不会超过设计温度。研究结果为增殖区热工水力学方案的设计验证和下一步开展实验提供参考。%Based on the CFD software platform FLUENT, three-dimensional numerical simulation was carried out for thermal hydraulics characteristics of the breeder zone pebble bed for China helium cooled ceramic breeder-test blanket module (HCCB-TBM). According the actual operating conditions of ITER, such as the nuclear heat distributions in the breeder zone and structure wall, the inlet velocities and temperatures of purge gas helium and coolant helium, the flow field and heat transfer characteristics of the purge gas helium carrying the tritium flowing over the lithium silicate pebble bed were obtained, and the temperature distribution and pressure drop were presented as well. The calculation results show that the permutation mode of the lithium silicate spheres in the pebble bed have an influence on the flow field and the maximum temperature of the pebble-bed. The maximum temperature of the pebble-bed and the structure wall does not exceed the designed temperature under the ITER operating conditions. The results will be benefit for the design verification of the breeder zone thermal hydraulics scheme and improving the following experiments of the purge gas.

  18. 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

  19. Enthalpy and void distributions in subchannels of PHWR fuel bundles

    Energy Technology Data Exchange (ETDEWEB)

    Park, J. W.; Choi, H.; Rhee, B. W. [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1998-12-31

    Two different types of the CANDU fuel bundles have been modeled for the ASSERT-IV code subchannel analysis. From calculated values of mixture enthalpy and void fraction distribution in the fuel bundles, it is found that net buoyancy effect is pronounced in the central region of the DUPIC fuel bundle when compared with the standard CANDU fuel bundle. It is also found that the central region of the DUPIC fuel bundle can be cooled more efficiently than that of the standard fuel bundle. From the calculated mixture enthalpy distribution at the exit of the fuel channel, it is found that the mixture enthalpy and void fraction can be highest in the peripheral region of the DUPIC fuel bundle. On the other hand, the enthalpy and the void fraction were found to be highest in the central region of the standard CANDU fuel bundle at the exit of the fuel channel. This study shows that the subchannel analysis is very useful in assessing thermal behavior of the fuel bundle that could be used in CANDU reactors. 10 refs., 4 figs., 2 tabs. (Author)

  20. Right bundle branch block

    DEFF Research Database (Denmark)

    Bussink, Barbara E; Holst, Anders Gaarsdal; Jespersen, Lasse

    2013-01-01

    AimsTo determine the prevalence, predictors of newly acquired, and the prognostic value of right bundle branch block (RBBB) and incomplete RBBB (IRBBB) on a resting 12-lead electrocardiogram in men and women from the general population.Methods and resultsWe followed 18 441 participants included.......5%/2.3% in women, P Right bundle branch block was associated with significantly...... increased all-cause and cardiovascular mortality in both genders with age-adjusted hazard ratios (HR) of 1.31 [95% confidence interval (CI), 1.11-1.54] and 1.87 (95% CI, 1.48-2.36) in the gender pooled analysis with little attenuation after multiple adjustment. Right bundle branch block was associated...

  1. Thermal hydraulic analysis of the IPR-R1 TRIGA reactor; Analise termo-hidraulica do reator TRIGA IPR-R1

    Energy Technology Data Exchange (ETDEWEB)

    Veloso, Marcelo Antonio [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN), Belo Horizonte, MG (Brazil); Fortini, Maria Auxiliadora [Minas Gerais Univ., Belo Horizonte, MG (Brazil). Dept. de Engenharia Nuclear

    2002-07-01

    The subchannel approach, normally employed for the analysis of power reactor cores that work under forced convection, have been used for the thermal hydraulic evaluation of a TRIGA Mark I reactor, named IPR-R1, at 250 kW power level. This was accomplished by using the PANTERA-1P subchannel code, which has been conveniently adapted to the characteristics of natural convection of TRIGA reactors. The analysis of results indicates that the steady state operation of IPR-R1 at 250 kW do not imply risks to installations, workers and public. (author)

  2. Bundles of Banach algebras

    Directory of Open Access Journals (Sweden)

    J. W. Kitchen

    1994-01-01

    Full Text Available We study bundles of Banach algebras π:A→X, where each fiber Ax=π−1({x} is a Banach algebra and X is a compact Hausdorff space. In the case where all fibers are commutative, we investigate how the Gelfand representation of the section space algebra Γ(π relates to the Gelfand representation of the fibers. In the general case, we investigate how adjoining an identity to the bundle π:A→X relates to the standard adjunction of identities to the fibers.

  3. Principal -bundles on Nodal Curves

    Indian Academy of Sciences (India)

    Usha N Bhosle

    2001-08-01

    Let be a connected semisimple affine algebraic group defined over . We study the relation between stable, semistable -bundles on a nodal curve and representations of the fundamental group of . This study is done by extending the notion of (generalized) parabolic vector bundles to principal -bundles on the desingularization of and using the correspondence between them and principal -bundles on . We give an isomorphism of the stack of generalized parabolic bundles on with a quotient stack associated to loop groups. We show that if is simple and simply connected then the Picard group of the stack of principal -bundles on is isomorphic to ⊕ , being the number of components of .

  4. On projective space bundle with nef normalized tautological line bundle

    CERN Document Server

    Yasutake, Kazunori

    2011-01-01

    In this paper, we study the structure of projective space bundles whose relative anti-canonical line bundle is nef. As an application, we get a characterization of abelian varieties up to finite etale covering.

  5. Meromorphic Higgs bundles And Related Geometries

    OpenAIRE

    Dalakov, Peter

    2016-01-01

    The present note is mostly a survey on the generalised Hitchin integrable system and moduli spaces of meromorphic Higgs bundles. We also fill minor gaps in the existing literature, outline a calculation of the infinitesimal period map and review briefly some related geometries.

  6. Meromorphic Higgs bundles and related geometries

    Science.gov (United States)

    Dalakov, Peter

    2016-11-01

    The present note is mostly a survey on the generalised Hitchin integrable system and moduli spaces of meromorphic G-Higgs bundles. We also fill minor gaps in the existing literature, outline a calculation of the infinitesimal period map and review some related geometries.

  7. Preliminary thermal-hydraulic design and simulation for hybrid breeder blanket%聚变-快裂变增殖堆包层初步热工水力学设计分析

    Institute of Scientific and Technical Information of China (English)

    王小勇; 栗再新; 赵奉超; 赵周; 武兴华; 王琦杰

    2014-01-01

    Thermal-hydraulic design and analysis for the new conceptual design of fusion-fission breeding reactor using casing pipes for fuel assembly was done. Based on typical thermal-hydraulic design parameters, preliminary thermal-hydraulic design for the blanket was proposed. The corresponding temperature distribution and pressure distribution were obtained using thermal-hydraulic codes, CFX. The simulation results showed that maximum temperature of the materials were all below their corresponding temperature limits, coolant temperature at the outlet was higher than 773℃, and pressure drop of the coolant could satisfy engineering requirement. The reasonability of this thermal-hydraulic design was preliminarily verified.%对新提出的套管结构聚变-快裂变增殖堆包层概念设计方案进行了热工水力学分析和设计,给出了典型的热工设计参数,并结合大型热工水力学软件CFX对其进行了温度场和压力分布的模拟分析。分析结果表明,材料温度均已低于许用温度,冷却剂出口温度高于773K,冷却剂压降也符合工程上的要求,初步验证了增殖堆包层设计的合理性。

  8. Quillen Bundle and Geometric Prequantization of Non-Abelian Vortices on a Riemann Surface

    Indian Academy of Sciences (India)

    Rukmini Dey; Samir K Paul

    2011-02-01

    In this paper we prequantize the moduli space of non-abelian vortices. We explicitly calculate the symplectic form arising from 2 metric and we construct a prequantum line bundle whose curvature is proportional to this symplectic form. The prequantum line bundle turns out to be Quillen’s determinant line bundle with a modified Quillen metric. Next, as in the case of abelian vortices, we construct line bundles over the moduli space whose curvatures form a family of symplectic forms which are parametrized by $\\Psi_0$, a section of a certain bundle. The equivalence of these prequantum bundles are discussed.

  9. Kernel bundle EPDiff

    DEFF Research Database (Denmark)

    Sommer, Stefan Horst; Lauze, Francois Bernard; Nielsen, Mads

    2011-01-01

    In the LDDMM framework, optimal warps for image registration are found as end-points of critical paths for an energy functional, and the EPDiff equations describe the evolution along such paths. The Large Deformation Diffeomorphic Kernel Bundle Mapping (LDDKBM) extension of LDDMM allows scale space...

  10. Universal Lagrangian bundles

    NARCIS (Netherlands)

    Sepe, D.

    2013-01-01

    The obstruction to construct a Lagrangian bundle over a fixed integral affine manifold was constructed by Dazord and Delzant (J Differ Geom 26:223–251, 1987) and shown to be given by ‘twisted’ cup products in Sepe (Differ GeomAppl 29(6): 787–800, 2011). This paper uses the topology of universal Lagr

  11. ALUMINUM BOX BUNDLING PRESS

    Directory of Open Access Journals (Sweden)

    Iosif DUMITRESCU

    2015-05-01

    Full Text Available In municipal solid waste, aluminum is the main nonferrous metal, approximately 80- 85% of the total nonferrous metals. The income per ton gained from aluminum recuperation is 20 times higher than from glass, steel boxes or paper recuperation. The object of this paper is the design of a 300 kN press for aluminum box bundling.

  12. Thermal-Hydraulic System Study of the Helium Cooled Pebble Bed (HCPB) Test Blanket Module (TBM) for ITER Using System Code RELAP5

    Science.gov (United States)

    Jin, Xuezhou; R, Meyder

    2005-04-01

    The HCPB concept has been a European DEMO reference concept for nearly one decade. Detailed thermal-hydraulic study on the control behavior of the whole system is one of the important parts of this development. The thermal-hydraulic effect of the TBM-combined cooling circuit during a cyclic operation in ITER has been studied using the system code RELAP5. The RELAP5 is based on an one-dimensional, transient two-fluid model for the flow of a two-phase steam-water mixture that can contain noncondensable components like Helium. The RELAP5-models are modified to take the cyclic operation of the circulator, heat exchanger, bypass, valves etc in to account. A sequence of operational phases is investigated, starting from the cold state through the heating phase that brings the system to a stand-by condition, followed by typical power cycles applied in ITER. The results show that the implemented control mechanisms keep the inlet temperature to the TBM and the total mass flow rate at the required values through all phases.

  13. Thermal-Hydraulic System Study of the Helium Cooled Pebble Bed (HCPB) Test Blanket Module (TBM) for ITER Using System Code RELAP5

    Institute of Scientific and Technical Information of China (English)

    Jin Xuezhou; R. Meyder

    2005-01-01

    The HCPB concept has been a European DEMO reference concept for nearly one decade. Detailed thermal-hydraulic study on the control behavior of the whole system is one of the important parts of this development. The thermal-hydraulic effect of the TBM-combined cooling circuit during a cyclic operation in ITER has been studied using the system code RELAP5. The RELAP5 is based on an one-dimensional, transient two-fluid model for the flow of a two-phase steam-water mixture that can contain noncondensable components like Helium. The RELAP5-models are modified to take the cyclic operation of the circulator, heat exchanger, bypass, valves etc in to account. A sequence of operational phases is investigated, starting from the cold state through the heating phase that brings the system to a stand-by condition, followed by typical power cycles applied in ITER. The results show that the implemented control mechanisms keep the inlet temperature to the TBM and the total mass flow rate at the required values through all phases.

  14. Benchmarking of thermal hydraulic loop models for Lead-Alloy Cooled Advanced Nuclear Energy System (LACANES), phase-I: Isothermal steady state forced convection

    Science.gov (United States)

    Cho, Jae Hyun; Batta, A.; Casamassima, V.; Cheng, X.; Choi, Yong Joon; Hwang, Il Soon; Lim, Jun; Meloni, P.; Nitti, F. S.; Dedul, V.; Kuznetsov, V.; Komlev, O.; Jaeger, W.; Sedov, A.; Kim, Ji Hak; Puspitarini, D.

    2011-08-01

    As highly promising coolant for new generation nuclear reactors, liquid Lead-Bismuth Eutectic has been extensively worldwide investigated. With high expectation about this advanced coolant, a multi-national systematic study on LBE was proposed in 2007, which covers benchmarking of thermal hydraulic prediction models for Lead-Alloy Cooled Advanced Nuclear Energy System (LACANES). This international collaboration has been organized by OECD/NEA, and nine organizations - ENEA, ERSE, GIDROPRESS, IAEA, IPPE, KIT/IKET, KIT/INR, NUTRECK, and RRC KI - contribute their efforts to LACANES benchmarking. To produce experimental data for LACANES benchmarking, thermal-hydraulic tests were conducted by using a 12-m tall LBE integral test facility, named as Heavy Eutectic liquid metal loop for integral test of Operability and Safety of PEACER (HELIOS) which has been constructed in 2005 at the Seoul National University in the Republic of Korea. LACANES benchmark campaigns consist of a forced convection (phase-I) and a natural circulation (phase-II). In the forced convection case, the predictions of pressure losses based on handbook correlations and that obtained by Computational Fluid Dynamics code simulation were compared with the measured data for various components of the HELIOS test facility. Based on comparative analyses of the predictions and the measured data, recommendations for the prediction methods of a pressure loss in LACANES were obtained. In this paper, results for the forced convection case (phase-I) of LACANES benchmarking are described.

  15. 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.

  16. 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