WorldWideScience

Sample records for thermal hydraulic investigations

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

    Science.gov (United States)

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

    2017-07-01

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

  2. Characteristic thermal-hydraulic problems in NHRs: Overview of experimental investigations and computer codes

    Energy Technology Data Exchange (ETDEWEB)

    Falikov, A A; Vakhrushev, V V; Kuul, V S; Samoilov, O B; Tarasov, G I [OKBM, Nizhny Novgorod (Russian Federation)

    1997-09-01

    The paper briefly reviews the specific thermal-hydraulic problems for AST-type NHRs, the experimental investigations that have been carried out in the RF, and the design procedures and computer codes used for AST-500 thermohydraulic characteristics and safety validation. (author). 13 refs, 10 figs, 1 tab.

  3. Thermal-hydraulic transient characteristics of ship-propulsion reactor investigated through safety analysis

    International Nuclear Information System (INIS)

    Fujiki, Kazuo; Asaka, Hideaki; Ishida, Toshihisa

    1986-01-01

    Thermal-hydraulic behaviors in the reactor of Nuclear Ship ''Mutsu'' were investigated through safety evaluation of operational transients by using RETRAN and COBRA-IV codes. The results were compared to the transient behaviors of typical commercial PWR and the characteristics of transient thermal-hydraulic behaviors in ship-loaded reactor were figured out. ''Mutsu'' reactor has larger thermal margin than commercial PWR because it is designed to be used as ship-propulsion power source in the load-following operation mode. This margin makes transient behavior in general milder than in commercial PWR but high opening pressure set point of main-steam safety valves leads poor heat-sink condition after reactor trip. The effects of other small-sized components are also investigated. The findings in the paper will be helpful in the design of future advanced reactor for nuclear ship. (author)

  4. Investigation of coupling scheme for neutronic and thermal-hydraulic codes

    International Nuclear Information System (INIS)

    Wang Guoli; Yu Jianfeng; Pen Muzhang; Zhang Yuman.

    1988-01-01

    Recently, a number of coupled neutronics/thermal-hydraulics codes have been used in reaction design and safty analysis, which have been obtained by coupling previous neutronic and thermal-hydraulic codes. The different coupling schemes affect computer time and accuracy of calculation results. Numberical experiments of several different coupling schemes and some heuristic results are described

  5. Thermally Actuated Hydraulic Pumps

    Science.gov (United States)

    Jones, Jack; Ross, Ronald; Chao, Yi

    2008-01-01

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

  6. Thermal-hydraulics investigations for the Liquid Lead-Bismuth Target of the SINQ spallation source

    International Nuclear Information System (INIS)

    Sigg, B.; Dury, T.; Hudina, M.; Smith, B.

    1991-01-01

    The paper contains a discussion of the thermal-hydraulic problems of the target which require detailed analysis by means of a two- or three-dimensional space- and in part also time-dependent fluid dynamics code. There follows a short description of the general-purpose code ASTEC, which is being used for these investigations, and examples of the target modelling, including results. The final part of the paper is devoted to a short discussion of experiments against which this application of the code has to be validated. (author)

  7. 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)

  8. Thermal hydraulic evaluation for an experimental facility to investigate pressurized thermal shock (PTS) in CDTN/CNEN

    International Nuclear Information System (INIS)

    Palmieri, Elcio T.; Navarro, Moyses A.; Aronne, Ivam D.; Terra, Jose L.

    2002-01-01

    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)

  9. Numerical investigation of thermal-hydraulic performance of channel with protrusions by turbulent cross flow jet

    Science.gov (United States)

    Sahu, M. K.; Pandey, K. M.; Chatterjee, S.

    2018-05-01

    In this two dimensional numerical investigation, small rectangular channel with right angled triangular protrusions in the bottom wall of test section is considered. A slot nozzle is placed at the middle of top wall of channel which impinges air normal to the protruded surface. A duct flow and nozzle flow combined to form cross flow which is investigated for heat transfer enhancement of protruded channel. The governing equations for continuity, momentum, energy along with SST k-ω turbulence model are solved with finite volume based Computational fluid dynamics code ANSYS FLUENT 14.0. The range of duct Reynolds number considered for this analysis is 8357 to 51760. The ratios of pitch of protrusion to height of duct considered are 0.5, 0.64 and 0.82. The ratios of height of protrusion to height of duct considered are 0.14, 0.23 and 0.29. The effect of duct Reynolds number, pitch and height of protrusion on thermal-hydraulic performance is studied under cross flow condition. It is found that heat transfer rate is more at relatively larger pitch and small pressure drop is found in case of low height of protrusion.

  10. Thermal-hydraulic investigations on the CEA-ENEA DEMO relevant helium cooled poloidal blanket

    International Nuclear Information System (INIS)

    Dell'Orco, G.; Polazzi, G.; Vallette, F.; Proust, E.; Eid, M.

    1994-01-01

    The CEA-ENEA design of an Helium Cooled Solid Breeder Blanket (HCSBB) for the DEMO reactor, with a breeder in tube (BIT) poloidal arrangement, is based on the use of lithium ceramic pellets, the ENEA γ-LiAlO 2 or the CEA Li 2 ZrO 3 . Due to the geometry of the DEMO reactor plasma chamber, these breeder bundles are adapted to the Vacuum Vessel with a strong poloidal curvature. This curvature influences the thermal-hydraulic behaviour of the coolant flowing inside the bundle. The paper presents the CEA-ENEA first results of the experimental and theoretical programme, aiming at optimizing the breeder module thermal hydraulic design. (author) 6 refs.; 7 figs.; 1 tab

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2013-05-15

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2017-09-15

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

  13. Thermal hydraulic numerical investigation of the heavy liquid metal free surface of MYRRHA spallation target experimental

    International Nuclear Information System (INIS)

    Batta, A.; Class, A.

    2015-01-01

    The first advanced design of accelerator-driven systems (ADS) is currently being built in SCK-CEN (Mol, Belgium): MYRRHA (Multi-purpose hybrid research reactor for high-tech applications). The experiment investigates the free surface design of the MYRRHA target. The free surface lead-bismuth eutectic (LBE) liquid metal experiment is a full-scale model of the concentric MYRRHA target. The design of the target is combined with CFD simulations using a volume of fluid method accounting for mass transfer across the free surface. The model used has been validated with water experimental results. The design of the target enables a high fluid velocity and a stable surface at the beam entry. In the current work, we present numerical results of Star- CD simulations employing a high-resolution interface-capturing scheme in conjunction with the cavitation model for the nominal operation conditions. Thermal hydraulic of the target is considered for the nominal flow rate and nominal heat load. Results show that the target has a very stable free surface configuration for the considered flow rate and heat load

  14. 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.

  15. Numerical investigation on thermal-hydraulic performance of new printed circuit heat exchanger model

    International Nuclear Information System (INIS)

    Kim, Dong Eok; Kim, Moo Hwan; Cha, Jae Eun; Kim, Seong O.

    2008-01-01

    Three-dimensional numerical analysis was performed to investigate heat transfer and pressure drop characteristics of supercritical CO 2 flow in new Printed Circuit Heat Exchanger (PCHE) model using commercial CFD code, Fluent 6.3. First, numerical analysis for conventional zigzag channel PCHE model was performed and compared with previous experimental data. Maximum deviation of in-outlet temperature difference and pressure drop from experimental data is about 10%. A new PCHE model has been designed to optimize thermal-hydraulic performance of PCHE. The new PCHE model has several airfoil shape fins (NACA 0020 model), which are designed to streamlined shape. Simulation results showed that in the airfoil shape fin PCHE, total heat transfer rate per unit volume was almost same with zigzag channel PCHE and the pressure drop was reduced to one-twentieth of that in zigzag channel PCHE. In airfoil shape fin PCHE model, the enhancement of heat transfer area and the uniform flow configuration contributed to obtain the same heat transfer performance with zigzag channel PCHE model. And the reduction of pressure drop in airfoil shape fin PCHE model was caused by suppressing generation of separated flow owing to streamlined shape of airfoil fins

  16. Investigation of the possibility to use a fine-mesh solver for resolving coupled neutronics and thermal-hydraulics

    International Nuclear Information System (INIS)

    Jareteg, K.; Vinai, P.; Demaziere, C.

    2013-01-01

    The development of a fine-mesh coupled neutronic/thermal-hydraulic solver is touched upon in this paper. The reported work investigates the feasibility of using finite volume techniques to discretize a set of conservation equations modeling neutron transport, fluid dynamics, and heat transfer within a single numerical tool. With the long-term objective of developing fine-mesh computing capabilities for a few selected fuel assemblies in a nuclear core, this preliminary study considers an infinite array of a single fuel assembly having a finite height. Thermal-hydraulic conditions close to the ones existing in PWRs are taken as a first test case. The neutronic modeling relies on the diffusion approximation in a multi-energy group formalism, with cross-sections pre-calculated and tabulated at the sub-pin level using a Monte Carlo technique. The thermal-hydraulics is based on the Navier-Stokes equations, complemented by an energy conservation equation. The non-linear coupling terms between the different conservation equations are fully resolved using classical iteration techniques. Early tests demonstrate that the numerical tool provides an unprecedented level of details of the coupled solution estimated within the same numerical tool and thus avoiding any external data transfer, using fully consistent models between the neutronics and the thermal-hydraulics. (authors)

  17. CFD investigation of thermal-hydraulic characteristics in a PBR core using different contact treatments between pebbles

    International Nuclear Information System (INIS)

    Ferng, Y.M.; Lin, K.Y.

    2014-01-01

    Highlights: • It is important to study thermal-hydraulic characteristics in a PBR for a HTGR. • A CFD model is proposed to simulate flow and heat transfer in a segment of pebbles. • Area and point contact treatments for adjacent pebbles are adopted in this study. • Predicted dependences of Nu and friction factor agree with the correlations. - Abstract: A high temperature gas cooled reactor (HTGR) with a pebble bed core (PBR) can be considered as one of the possible energy generation sources in the incoming future due to its inherently safe performance, lower power density, and higher conversion efficiency, etc. It is important to study the thermal-hydraulic characteristics in a PBR for optimum design and safe operation of a HTGR. In this paper, a computational fluid dynamics (CFD) methodology is proposed to investigate the thermal-hydraulic behavior in a segment of pebbles representing the central region of a PBR. Two kinds of contact modeling between adjacent pebbles are adopted, namely area and point contact treatments. The former contact treatment is a geometric approximation modeling. Based on the comparisons of thermal-hydraulic characteristics in the pebbles predicted by both contact treatments, no significant difference is revealed except for the near-wall secondary flow pattern. In addition, compared with the calculated results from the well-known correlations, the present predicted dependence of Nu number and friction factor on the particle Reynolds number shows good agreement qualitatively and quantitatively

  18. Investigation of coolant thermal mixing within 28-element CANDU fuel bundles using the ASSERT-PV thermal hydraulics code

    International Nuclear Information System (INIS)

    Lightston, M.F.; Rock, R.

    1996-01-01

    This paper presents the results of a study of the thermal mixing of single-phase coolant in 28-element CANDU fuel bundles under steady-state conditions. The study, which is based on simulations performed using the ASSERT-PV thermal hydraulic code, consists of two main parts. In the first part the various physical mechanisms that contribute to coolant mixing are identified and their impact is isolated via ASSERT-PV simulations. The second part is concerned with development of a preliminary model suitable for use in the fuel and fuel channel code FACTAR to predict the thermal mixing that occurs between flow annuli. (author)

  19. 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.

  20. Investigation of Two-Phase Flow Regime Maps for Development of Thermal-Hydraulic Analysis Codes

    International Nuclear Information System (INIS)

    Kim, Kyung Doo; Kim, Byoung Jae; Lee, Seong Wook

    2010-04-01

    This reports is a literature survey on models and correlations for determining flow pattern that are used to simulate thermal-hydraulics in nuclear reactors. Determination of flow patterns are a basis for obtaining physical values of wall/interfacial friction, wall/interfacial heat transfer, and droplet entrainment/de-entrainment. Not only existing system codes, such as RELAP5-3D, TRAC-M, MARS, TRACE, CATHARE) but also up-to-date researches were reviewed to find models and correlations

  1. PETER loop. Multifunctional test facility for thermal hydraulic investigations of PWR fuel elements

    International Nuclear Information System (INIS)

    Ganzmann, I.; Hille, D.; Staude, U.

    2009-01-01

    The reliable fuel element behavior during the complete fuel cycle is one of the fundamental prerequisites of a safe and efficient nuclear power plant operation. The fuel element behavior with respect to pressure drop and vibration impact cannot be simulated by means of fluid-structure interaction codes. Therefore it is necessary to perform tests using fuel element mock-ups (1:1). AREVA NP has constructed the test facility PETER (PWR fuel element tests in Erlangen) loop. The modular construction allows maximum flexibility for any type of fuel elements. Modern measuring instrumentation for flow, pressure and vibration characterization allows the analysis of cause and consequences of thermal hydraulic phenomena. PETER loop is the standard test facility for the qualification of dynamic fuel element behavior in flowing fluid and is used for failure mode analysis.

  2. Steam generator thermal-hydraulics

    International Nuclear Information System (INIS)

    Inch, W.W.; Scott, D.A.; Carver, M.B.

    1980-01-01

    This paper discusses a code for detailed numerical modelling of steam generator thermal-hydraulics, and describes related experimental programs designed to promote in-depth understanding of three-dimensional two-phase flow. (auth)

  3. Summary report of NEPTUN investigations into transient thermal hydraulics of the passive decay heat removal

    International Nuclear Information System (INIS)

    Weinberg, D.; Hoffmann, H.; Rust, K.; Frey, H.H.; Hain, K.; Leiling, W.; Hayafune, H.

    1995-12-01

    The results corroborate the findings of tests with the RAMONA model. With the core power reduction at scram and the start of the decay heat exchangers operation cold fluid is delivered into the prevailing upper plenum. A temperature stratification develops with distinct large temperature gradients. The onset of natural convection is mainly influenced by two effects, namely, the temperature increase on the intermediate heat exchangers primary sides as a result of which the downward pressures are reduced, and the startup of the decay heat exchangers which leads to a decrease of the buoyancy forces in the core. The temperatures of the upper plenum are systematically reduced as soon as the decay heat exchangers are in operation. Then mixed fluid in the hot plenum reaches the intermediate heat exchangers inlet windows and causes an increase in the core flow rate. The primary pump coastdown curve influences the primary system thermal hydraulics only during the first thousand seconds after scram. The longer the pumps operate the more cold fluid is delivered via the core to the upper plenum. The delay of the start of the decay heat exchangers operation separates the two effects which influence the core mass flow, namely the heatup of the intermediate heat exchangers as well as the formation of the stratification in the upper plenum. Increasing the power as well as the operation of only half of the available decay heat exchangers increase the system temperatures. A permeable above core structure produces a temperature stratification along the total upper plenum, and therefore a lower temperature gradient in the region between core outlet and lower edge of the above core structure, in comparison to the impermeable design. A complete flow path blockage of the primary fluid through the intermediate heat exchangers leads to an enhanced cooling effect of the interstitial flow and gives rise to a thermosiphon effect inside the core elements. (orig./GL) [de

  4. Horizontal steam generator thermal-hydraulics

    Energy Technology Data Exchange (ETDEWEB)

    Ubra, O. [SKODA Praha Company, Prague (Czechoslovakia); Doubek, M. [Czech Technical Univ., Prague (Czechoslovakia)

    1995-09-01

    Horizontal steam generators are typical components of nuclear power plants with pressure water reactor type VVER. Thermal-hydraulic behavior of horizontal steam generators is very different from the vertical U-tube steam generator, which has been extensively studied for several years. To contribute to the understanding of the horizontal steam generator thermal-hydraulics a computer program for 3-D steady state analysis of the PGV-1000 steam generator has been developed. By means of this computer program, a detailed thermal-hydraulic and thermodynamic study of the horizontal steam generator PGV-1000 has been carried out and a set of important steam generator characteristics has been obtained. The 3-D distribution of the void fraction and 3-D level profile as functions of load and secondary side pressure have been investigated and secondary side volumes and masses as functions of load and pressure have been evaluated. Some of the interesting results of calculations are presented in the paper.

  5. GCFR thermal-hydraulic experiments

    International Nuclear Information System (INIS)

    Schlueter, G.; Baxi, C.B.; Dalle Donne, M.; Gat, U.; Fenech, H.; Hanson, D.; Hudina, M.

    1980-01-01

    The thermal-hydraulic experimental studies performed and planned for the Gas-Cooled Fast Reactor (GCFR) core assemblies are described. The experiments consist of basic studies performed to obtain correlations, and bundle experiments which provide input for code validation and design verification. These studies have been performed and are planned at European laboratories, US national laboratories, Universities in the US, and at General Atomic Company

  6. Investigation of pool thermal hydraulics and temperature distribution in inner vessel under mechanical seal leakage

    International Nuclear Information System (INIS)

    Abraham, Juby; Velusamy, K.; Selvaraj, P.

    2015-01-01

    The primary heat sink of prototype fast breeder reactor is a sodium pool which is partitioned into cold pool and hot pool. The inner vessel which separates the cold and hot pools is having penetrations for intermediate heat exchangers. The hot sodium from hot pool leaks into the cold pool through these penetrations and to reduce the leakage, mechanical seals are provided. Leakage of hot sodium into cold pool can lead to thermal stratification in the cold pool and also will affect the temperature distribution in inner vessel. 3-D CFD studies were performed focusing these features as a function of sodium leakage. The analyses indicate that the maximum temperature difference across the IV thickness is 65°C without any leakage of sodium. The temperature difference is found to decrease with increase in leakage through the seals. It is seen that a leakage of 2.5% is acceptable. (author)

  7. A HISTORICAL PERSPECTIVE OF NUCLEAR THERMAL HYDRAULICS

    Energy Technology Data Exchange (ETDEWEB)

    D’Auria, F; Rohatgi, Upendra S.

    2017-01-12

    The nuclear thermal-hydraulics discipline was developed following the needs for nuclear power plants (NPPs) and, to a more limited extent, research reactors (RR) design and safety. As in all other fields where analytical methods are involved, nuclear thermal-hydraulics took benefit of the development of computers. Thermodynamics, rather than fluid dynamics, is at the basis of the development of nuclear thermal-hydraulics together with the experiments in complex two-phase situations, namely, geometry, high thermal density, and pressure.

  8. Summary report of NEPTUN investigations into the steady state thermal hydraulics of the passive decay heat removal

    International Nuclear Information System (INIS)

    Rust, K.; Weinberg, D.; Hoffmann, H.; Frey, H.H.; Baumann, W.; Hain, K.; Leiling, W.; Hayafune, H.; Ohira, H.

    1995-12-01

    During the course of steady state NEPTUN investigations, the effects of different design and operating parameters were studied; in particular: The shell design of the above core sturcture, the core power, the number of decay heat exchangers put in operation, the complete flow path blockage at the primary side of the intermediate heat exchangers, and the fluid level in the primary vessel. The findings of the NEPTUN experiments indicate that the decay heat can be safely removed by natural convection. The interwrapper flow makes an essential contribution to that behavior. The decay heat exchangers installed in the upper plenum cause a thermal stratification associated with a pronounced gradient. The vertical extent of the stratification and the quantity of the gradient are depending on the fact whether a permeable or an impermeable shell covers the above core structure. An increase of the core power or a reduction of the number of decay heat exchangers being in operation leads to a higher temperature level in the primary system but does not alter the global temperature distribution. In the case that no coolant enters the inlet windows at the primary side of the intermediate and decay heat exchangers, the core remains coolable as far as the primary vessel is filled with fluid up to a minimum level. Cold water penetrates from the upper plenum into the core and removes the decay heat. The thermal hydraulic computer code FLUTAN was applied for the three-dimensional numerical simulation of the majority of NEPTUN tests reported here. The comparison of computed against experimental data indicates a qualitatively and quantitatively satisfying agreement of the findings with respect to the field of isotherms as well as the temperature profiles in the upper plenum and within the core region of very complex geometry. (orig./HP) [de

  9. CFD investigation of flow inversion in typical MTR research reactor undergoing thermal-hydraulic transients

    International Nuclear Information System (INIS)

    Salama, Amgad

    2011-01-01

    Highlights: → The 3D, CFD simulation of FLOFA accident in the generic IAEA 10 MW research reactor is carried out. → The different flow and heat transfer mechanisms involved in this process were elucidated. → The transition between these mechanisms during the course of FLOFA is discussed and investigated. → The interesting inversion process upon the transition from downward flow to upward flow is shown. → The temperature field and the friction coefficient during the whole transient process were shown. - Abstract: Three dimensional CFD full simulations of the fast loss of flow accident (FLOFA) of the IAEA 10 MW generic MTR research reactor are conducted. In this system the flow is initially downward. The transient scenario starts when the pump coasts down exponentially with a time constant of 1 s. As a result the temperatures of the heating element, the clad, and the coolant rise. When the flow reaches 85% of its nominal value the control rod system scrams and the power drops sharply resulting in the temperatures of the different components to drop. As the coolant flow continues to drop, the decay heat causes the temperatures to increase at a slower rate in the beginning. When the flow becomes laminar, the rate of temperature increase becomes larger and when the pumps completely stop a flow inversion occurs because of natural convection. The temperature will continue to rise at even higher rates until natural convection is established, that is when the temperatures settle off. The interesting 3D patterns of the flow during the inversion process are shown and investigated. The temperature history is also reported and is compared with those estimated by one-dimensional codes. Generally, very good agreement is achieved which provides confidence in the modeling approach.

  10. Reactor Thermal Hydraulic Numerical Calculation And Modeling

    International Nuclear Information System (INIS)

    Duong Ngoc Hai; Dang The Ba

    2008-01-01

    In the paper the results of analysis of thermal hydraulic state models using the numerical codes such as COOLOD, EUREKA and RELAP5 for simulation of the reactor thermal hydraulic states are presented. The calculations, analyses of reactor thermal hydraulic state and safety were implemented using different codes. The received numerical results, which were compared each to other, to experiment measurement of Dalat (Vietnam) research reactor and published results, show their appropriateness and capacity for analyses of different appropriate cases. (author)

  11. Investigating thermal-hydraulic characteristic of molten fluoride salt in a circular pipe using a CFD methodology

    International Nuclear Information System (INIS)

    Chi Chenwei; Ferng Yuhming; Pei Baushei; Liang Jenqhorng

    2011-01-01

    In recent years, the molten salt reactor (MSR) has attracted increasing attention and become one of the most important 'Generation IV reactor' designs. In particular, the fact that molten fluoride salts are utilized as liquid fuel and coolant constitutes the main feature of the reactor. Furthermore, since the molten fluoride salt has a high Prandtl number and contains quite different behaviors to those of ordinary water and gas, an in-depth investigation of molten fluoride salt is thus highly demanded. Hence, it is the central objective of this study to examine the thermal-hydraulic characteristics of molten salt especially for the optimal design of reactor core and its safety operation. In this study, the dependence of pressure drop, Nusselt number and entrance length on the inlet Reynolds number for a molten fluoride salt (LiF(46.5)-NaF(11.5)-KF(42)) are computed using a comprehensive computational fluid dynamics (CFD) methodology. The methodology employs the continuity equation, momentum equation, energy equation, and standard k - ε turbulence model to conduct fluid dynamics simulation. For simplicity, the geometry employed in this study is a circular tube. The simulated results indicated that the pressure drop and Nusselt number and entrance length increase as the inlet Reynolds number increases. And the computed pressure drop corresponds well to theoretical value. It is also given a new correlation of computed entrance length in this paper. In addition, two well-known Nusselt number correlations such as, Hausen, Gnielinski, are employed to make comparisons with the computed results. It is also found that the computed Nusselt numbers overestimate the Hausen ones in the high Reynolds number region. However, the computed Nusselt numbers correspond well to the Gnielinski ones in all the Reynolds numbers region. Also notice that an experimental setup is currently in progress in order to validate the present CFD simulation. (author)

  12. Transient thermal hydraulic analysis of the IAEA 10 MW MTR reactor during Loss of Flow Accident to investigate the flow inversion

    International Nuclear Information System (INIS)

    AL-Yahia, Omar S.; Albati, Mohammad A.; Park, Jonghark; Chae, Heetaek; Jo, Daeseong

    2013-01-01

    Highlights: • Transient analyses of a slow and fast LOFA were investigated. • A reactor kinetic and thermal hydraulic coupled model was developed. • Based on force balance, the flow rate during flow inversion was determined. • Flow inversion in a hot channel occurred earlier than in an average channel. • Two temperature peaks were observed during both slow and fast LOFA. - Abstract: Transient analyses of the IAEA 10 MW MTR reactor are investigated during a fast and slow Loss of Flow Accident (LOFA) with a neutron kinetic and thermal hydraulic coupling model. A spatial-dependent thermal hydraulic technique is adopted for analyzing the local thermal hydraulic parameters and hotspot location during a flow inversion. The flow rate through the channel is determined in terms of a balance between driving and preventing forces. Friction and buoyancy forces act as resistance of the flow before a flow inversion while buoyancy force becomes the driving force after a flow inversion. By taking into account the buoyancy effect to determine the flow rate, the difference in the flow inversion time between hot and average channels is investigated: a flow inversion occurs earlier in the hot channel than in an average channel. Furthermore, the movement of the hotspot location before and after a flow inversion is investigated for a slow and fast LOFA. During a flow inversion, two temperature peaks are observed: (1) the first temperature peak is at the initiation of the LOFA, and (2) the second temperature peak is when a flow inversion occurs. The maximum temperature of the cladding is found at the second temperature peak for both LOFA analyses, and is lower than the saturation temperature

  13. HANARO thermal hydraulic accident analysis

    Energy Technology Data Exchange (ETDEWEB)

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

    1996-06-01

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

  14. Liquid metal thermal-hydraulics

    International Nuclear Information System (INIS)

    Kottowski-Duemenil, H.M.

    1994-01-01

    This textbook is a report of the 26 years activity of the Liquid Metal Boiling Working Group (LMBWG). It summarizes the state of the art of liquid metal thermo-hydraulics achieved through the collaboration of scientists concerned with the development of the Fast Breeder Reactor. The first chapter entitled ''Liquid Metal Boiling Behaviour'', presents the background and boiling mechanisms. This section gives the reader a brief but thorough survey on the superheat phenomena in liquid metals. The second chapter of the text, ''A Review of Single and Two-Phase Flow Pressure Drop Studies and Application to Flow Stability Analysis of Boiling Liquid Metal Systems'' summarizes the difficulty of pressure drop simulation of boiling sodium in core bundles. The third chapter ''Liquid Metal Dry-Out Data for Flow in Tubes and Bundles'' describes the conditions of critical heat flux which limits the coolability of the reactor core. The fourth chapter dealing with the LMFBR specific topic of ''Natural Convection Cooling of Liquid Metal Systems''. This chapter gives a review of both plant experiments and out-of-pile experiments and shows the advances in the development of computing power over the past decade of mathematical modelling ''Subassembly Blockages Suties'' are discussed in chapter five. Chapter six is entitled ''A Review of the Methods and Codes Available for the Calculation on Thermal-Hydraulics in Rod-Cluster and other Geometries, Steady state and Transient Boiling Flow Regimes, and the Validation achieves''. Codes available for the calculation of thermal-hydraulics in rod-clusters and other geometries are reviewed. Chapter seven, ''Comparative Studies of Thermohydraulic Computer Code Simulations of Sodium Boiling under Loss of Flow Conditions'', represents one of the key activities of the LMBWG. Several benchmark exercises were performed with the aim of transient sodium boiling simulation in single channels and bundle blockages under steady state conditions and loss of

  15. Thermal-hydraulic unreliability of passive systems

    International Nuclear Information System (INIS)

    Tzanos, C.P.; Saltos, N.T.

    1995-01-01

    Advanced light water reactor designs like AP600 and the simplified boiling water reactor (SBWR) use passive safety systems for accident prevention and mitigation. Because these systems rely on natural forces for their operation, their unavailability due to hardware failures and human error is significantly smaller than that of active systems. However, the coolant flows predicted to be delivered by these systems can be subject to significant uncertainties, which in turn can lead to a significant uncertainty in the predicted thermal-hydraulic performance of the plant under accident conditions. Because of these uncertainties, there is a probability that an accident sequence for which a best estimate thermal-hydraulic analysis predicts no core damage (success sequence) may actually lead to core damage. For brevity, this probability will be called thermal-hydraulic unreliability. The assessment of this unreliability for all the success sequences requires very expensive computations. Moreover, the computational cost increases drastically as the required thermal-hydraulic reliability increases. The required computational effort can be greatly reduced if a bounding approach can be used that either eliminates the need to compute thermal-hydraulic unreliabilities, or it leads to the analysis of a few bounding sequences for which the required thermal-hydraulic reliability is relatively small. The objective of this paper is to present such an approach and determine the order of magnitude of the thermal-hydraulic unreliabilities that may have to be computed

  16. Thermal hydraulic feasibility assessment of the spent nuclear fuel project

    International Nuclear Information System (INIS)

    Heard, F.J.

    1996-01-01

    A series of analyses have been completed investigating the thermal-hydraulic performance and feasibility of the Spent Nuclear Fuel Project (SNFP) Integrated Process Strategy (IPS). The goal was to develop a series of thermal-hydraulic models that could respond to all process and safety related issues that may arise pertaining to the SNFP, as well as provide a basis for validation of the results. Results show that there is a reasonable envelope for process conditions and requirements that are thermally and hydraulically acceptable

  17. Experimental thermal hydraulics in support of FBR

    International Nuclear Information System (INIS)

    Padmakumar, G.; Anand Babu, C.; Kalyanasundaram, P.; Vaidyanathan, G.

    2009-01-01

    The thermal hydraulic design plays a crucial role for the safe and economical deployment of Liquid Metal Cooled Fast Breeder Reactor (LMFBR). Robust experimental programmes are required in support of LMFBR thermal hydraulics design. The philosophy of testing has been to construct small scale models to understand the physical behaviour and to build larger scale models to optimize the component design. The experiments are conducted either in sodium or using a simulant like water/air. The paper gives a brief account of the various thermal hydraulic experiments carried out in support of the design of Prototype Fast Breeder Reactor (PFBR). (author)

  18. 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.

  19. Applied mathematical methods in nuclear thermal hydraulics

    International Nuclear Information System (INIS)

    Ransom, V.H.; Trapp, J.A.

    1983-01-01

    Applied mathematical methods are used extensively in modeling of nuclear reactor thermal-hydraulic behavior. This application has required significant extension to the state-of-the-art. The problems encountered in modeling of two-phase fluid transients and the development of associated numerical solution methods are reviewed and quantified using results from a numerical study of an analogous linear system of differential equations. In particular, some possible approaches for formulating a well-posed numerical problem for an ill-posed differential model are investigated and discussed. The need for closer attention to numerical fidelity is indicated

  20. Review of computational thermal-hydraulic modeling

    International Nuclear Information System (INIS)

    Keefer, R.H.; Keeton, L.W.

    1995-01-01

    Corrosion of heat transfer tubing in nuclear steam generators has been a persistent problem in the power generation industry, assuming many different forms over the years depending on chemistry and operating conditions. Whatever the corrosion mechanism, a fundamental understanding of the process is essential to establish effective management strategies. To gain this fundamental understanding requires an integrated investigative approach that merges technology from many diverse scientific disciplines. An important aspect of an integrated approach is characterization of the corrosive environment at high temperature. This begins with a thorough understanding of local thermal-hydraulic conditions, since they affect deposit formation, chemical concentration, and ultimately corrosion. Computational Fluid Dynamics (CFD) can and should play an important role in characterizing the thermal-hydraulic environment and in predicting the consequences of that environment,. The evolution of CFD technology now allows accurate calculation of steam generator thermal-hydraulic conditions and the resulting sludge deposit profiles. Similar calculations are also possible for model boilers, so that tests can be designed to be prototypic of the heat exchanger environment they are supposed to simulate. This paper illustrates the utility of CFD technology by way of examples in each of these two areas. This technology can be further extended to produce more detailed local calculations of the chemical environment in support plate crevices, beneath thick deposits on tubes, and deep in tubesheet sludge piles. Knowledge of this local chemical environment will provide the foundation for development of mechanistic corrosion models, which can be used to optimize inspection and cleaning schedules and focus the search for a viable fix

  1. Investigation of 3D spatial effect on point kinetics estimation of the thermal hydraulics code RELAP for the analysis of MSLB accident of KK-NP

    International Nuclear Information System (INIS)

    Bera, S.; Pradhan, S.K.; Dubey, S.K.; Gupta, S.K.

    2011-01-01

    In general safety analyses of design basis accident of NPPs are being carried out using system thermal hydraulics code like RELAP. In RELAP, power is calculated based on point kinetics approximation, which virtually ignores the space and energy dependence of neutron flux. To include the space and energy dependence of neutron flux, three-dimensional neutronics code TRIHEXFA has been externally coupled with RELAP through interface program, TRIHEXFA-RELAP Interface Program (TRIP). Calculation methodology of TRIP program is based on adiabatic approximation. In the adiabatic approximation the neutron flux is being factored into spatial and amplitude part. Spatial part of flux is slowly varying with time whereas amplitude part is strongly varying function. The RELAP controls the transient time steps. Transient time is divided into several major and minor time steps. Minor time step is the sub-step of major time step. Thermal hydraulics and neutronics data are exchanged at each major time step. Spatial part of neutron flux has been updated at each major time step using TRIHEXFA code. But amplitude part of the neutron flux is calculated at each minor time step using RELAP code. Convergence of results of the coupled code, TRIP has been checked through coupling time step descritization study. This study determines the minimum coupling time step. Transient concerning VVER-1000 Main Steam Line Break, MSLB has been considered to investigate the space-time effect on point kinetics. MSLB occurs as a consequence of the rupture of one steam line upstream of main steam line isolation valves. Reference design and data from Kudankulam Nuclear Power Plant (KK-NPP) are used for the analysis. From this investigation it is found that TRIP significantly overestimates the maximum reactor power against uncoupled RELAP result. The time of scram also occur six seconds earlier in TRIP calculation compared to the RELAP. This exercise has also shown a proof of principle that coupling 3D

  2. Pore Scale Thermal Hydraulics Investigations of Molten Salt Cooled Pebble Bed High Temperature Reactor with BCC and FCC Configurations

    Directory of Open Access Journals (Sweden)

    Shixiong Song

    2014-01-01

    CFD results and empirical correlations’ predictions of pressure drop and local Nusselt numbers. Local pebble surface temperature distributions in several default conditions are investigated. Thermal removal capacities of molten salt are confirmed in the case of nominal condition; the pebble surface temperature under the condition of local power distortion shows the tolerance of pebble in extreme neutron dose exposure. The numerical experiments of local pebble insufficient cooling indicate that in the molten salt cooled pebble bed reactor, the pebble surface temperature is not very sensitive to loss of partial coolant. The methods and results of this paper would be useful for optimum designs and safety analysis of molten salt cooled pebble bed reactors.

  3. Thermal hydraulics and mechanics core design programs

    International Nuclear Information System (INIS)

    Heinecke, J.

    1992-10-01

    The report documents the work performed within the Research and Development Task T hermal hydraulics and mechanics core design programs , funded by the German government. It contains the development of new codes, the extension of existing codes, the qualification and verification of codes and the development of a code library. The overall goal of this work was to adapt the system of thermal hydraulics and mechanics codes to the permanently growing requirements of the status of science and technology

  4. Proceedings of the third nuclear thermal hydraulics meeting

    International Nuclear Information System (INIS)

    Anon.

    1987-01-01

    This book contains the proceedings of the Thermal Hydraulics Division of the American Nuclear Society. The papers presented include: Simulator qualification using engineering codes and Development of thermal hydraulic analysis capabilities for Oyster Creek

  5. Thermal-hydraulics of helium cooled First Wall channels and scoping investigations on performance improvement by application of ribs and mixing devices

    Energy Technology Data Exchange (ETDEWEB)

    Arbeiter, Frederik, E-mail: frederik.arbeiter@kit.edu [Karlsruhe Institute of Technology, Institute of Neutron Physics and Reactor Technology, Hermann-von-Helmholtz-Platz 1, 76344 Eggenstein-Leopoldshafen (Germany); Bachmann, Christian [EUROfusion – Programme Management Unit, Garching (Germany); Chen, Yuming; Ilić, Milica; Schwab, Florian [Karlsruhe Institute of Technology, Institute of Neutron Physics and Reactor Technology, Hermann-von-Helmholtz-Platz 1, 76344 Eggenstein-Leopoldshafen (Germany); Sieglin, Bernhard [Max-Planck-Institut für Plasmaphysik, Garching (Germany); Wenninger, Ronald [EUROfusion – Programme Management Unit, Garching (Germany)

    2016-11-01

    Highlights: • Existing first wall designs and expected plasma heat loads are reviewed. • Heat transfer enhancement methods are investigated by CFD. • The results for heat transfer and friction are given, compared and explained. • Relations for needed pumping power and gained thermal heat are shown. • A range for the maximum permissible heat loads from the plasma is estimated. - Abstract: The first wall (FW) of DEMO is a component with high thermal loads. The cooling of the FW has to comply with the material's upper and lower temperature limits and requirements from stress assessment, like low temperature gradients. Also, the cooling has to be integrated into the balance-of-plant, in a sense to deliver exergy to the power cycle and require a limited pumping power for coolant circulation. This paper deals with the basics of FW cooling and proposes optimization approaches. The effectiveness of several heat transfer enhancement techniques is investigated for the use in helium cooled FW designs for DEMO. Among these are wall-mounted ribs, large scale mixing devices and modified hydraulic diameter. Their performance is assessed by computational fluid dynamics (CFD), and heat transfer coefficients and pressure drop are compared. Based on the results, an extrapolation to high heat fluxes is tried to estimate the higher limits of cooling capabilities.

  6. Thermal hydraulic parametric investigation of decay heat removal from degraded core of a sodium cooled fast Breeder reactor

    Energy Technology Data Exchange (ETDEWEB)

    Verma, Lokesh [Department of Physics and Astrophysics, University of Delhi, Delhi 110007 (India); Kumar Sharma, Anil, E-mail: aksharma@igcar.gov.in [Reactor Design Group, Indira Gandhi Centre for Atomic Research, HBNI, Kalpakkam (India); Velusamy, K. [Reactor Design Group, Indira Gandhi Centre for Atomic Research, HBNI, Kalpakkam (India)

    2017-03-15

    Highlights: • Decay heat removal from degraded core of a typical SFR is highlighted. • Influence of number of DHXs in operation on PAHR is analyzed. • Investigations on structural integrity of the inner vessel and core catcher. • Feasibility study for retention of a part of debris in upper pool of SFR. - Abstract: Ensuring post accident decay heat removal with high degree of reliability following a Core Disruptive Accident (CDA) is very important in the design of sodium cooled fast reactors (SFR). In the recent past, a lot of research has been done towards the design of an in-vessel core catcher below the grid plate to prevent the core debris reaching the main vessel in a pool type SFR. However, during an energetic CDA, the entire core debris is unlikely to reach the core catcher. A significant part of the debris is likely to settle in core periphery between radial shielding subassemblies and the inner vessel. Failure of inner vessel due to the decay heat can lead to core debris reaching the main vessel and threatening its integrity. On the other hand, retention of a part of debris in core periphery can reduce the load on main core catcher. Towards achieving an optimum design of SFR and safety evaluation, it is essential to quantify the amount of heat generating core debris that can be retained safely within the primary vessel. This has been performed by a mathematical simulation comprising solution of 2-D transient form of the governing equations of turbulent sodium flow and heat transfer with Boussinesq approximations. The conjugate conduction-convection model adopted for this purpose is validated against in-house experimental data. Transient evolutions of natural convection in the pools and structural temperatures in critical components have been predicted. It is found that 50% of the core debris can be safely accommodated in the gap between radial shielding subassemblies and inner vessel without exceeding structural temperature limit. It is also

  7. Thermal-Hydraulic Tests for Reactor Core Safety

    International Nuclear Information System (INIS)

    Chun, Se Young; Chung, Moon Ki; Baek, Won Pil and others

    2005-04-01

    The reflood experiments for single rod annulus geometry have been performed to investigate the effect of spacer grid on thermal-hydraulics under reflood conditions. The reflood experimental loop for 6x6 rod bundle with a spacer grid developed in Korea has been provided. About 8000 data points for Post-CHF heat transfer have been obtained from the experiments About 1400 CHF data points for 3x3 Water and 5x5 Freon rod bundles have been obtained. The existing evaluation methodology for core safety under return-to-power conditions has been investigated using KAERI low flow CHF database. The hydraulic tests for turbulence mixing characteristics in subchannel of 5x5 rod bundle have been carried out using advanced measurement technique, LVD and the database for various spacer grids have been provided. In order to measure the turbulence mixing characteristics in details, the hydraulic loop with a magnified 5x5 rod bundle has been prepared. The database which was constructed through a systematic thermal hydraulic tests for the reflood phenomenon, CHF, Post-CHF is surely to be useful to the industry field, the regulation body and the development of thermal-hydraulic analysis code

  8. Thermal-hydraulic design of the 200 MW NHR

    Energy Technology Data Exchange (ETDEWEB)

    Jincai, Li; Zuying, Gao; Baocheng, Xu; Junxiao, He [Institute of Nuclear Energy and Technology, Tsingua Univ., Beijing (China)

    1997-09-01

    The main problems regarding the AST-500 NHR thermal-hydraulics are considered. Basic thermal data of the reactor plant are given and peculiarities of coolant parameters at natural convection in the primary circuit are discussed. The in-reactor instrumentation system is briefly describes, as well as the results of natural-convective flow characteristics investigations using reactor test models. (author). 4 refs, 5 figs.

  9. Analysis of uncertainties of thermal hydraulic calculations

    International Nuclear Information System (INIS)

    Macek, J.; Vavrin, J.

    2002-12-01

    In 1993-1997 it was proposed, within OECD projects, that a common program should be set up for uncertainty analysis by a probabilistic method based on a non-parametric statistical approach for system computer codes such as RELAP, ATHLET and CATHARE and that a method should be developed for statistical analysis of experimental databases for the preparation of the input deck and statistical analysis of the output calculation results. Software for such statistical analyses would then have to be processed as individual tools independent of the computer codes used for the thermal hydraulic analysis and programs for uncertainty analysis. In this context, a method for estimation of a thermal hydraulic calculation is outlined and selected methods of statistical analysis of uncertainties are described, including methods for prediction accuracy assessment based on the discrete Fourier transformation principle. (author)

  10. Thermal hydraulics in undergraduate nuclear engineering education

    International Nuclear Information System (INIS)

    Theofanous, T.G.

    1986-01-01

    The intense safety-related research efforts of the seventies in reactor thermal hydraulics have brought about the recognition of the subject as one of the cornerstones of nuclear engineering. Many nuclear engineering departments responded by building up research programs in this area, and mostly as a consequence, educational programs, too. Whether thermal hydraulics has fully permeated the conscience of nuclear engineering, however, remains yet to be seen. The lean years that lie immediately ahead will provide the test. The purpose of this presentation is to discuss the author's own educational activity in undergraduate nuclear engineering education over the past 10 yr or so. All this activity took place at Purdue's School of Nuclear Engineering. He was well satisfied with the results and expects to implement something similar at the University of California in Santa Barbara in the near future

  11. Thermal-hydraulic analysis of nuclear reactors

    CERN Document Server

    Zohuri, Bahman

    2015-01-01

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

  12. Virginia Power thermal-hydraulics methods

    International Nuclear Information System (INIS)

    Anderson, R.C.; Basehore, K.L.; Harrell, J.R.

    1987-01-01

    Virginia Power's nuclear safety analysis group is responsible for the safety analysis of reload cores for the Surry and North Anna power stations, including the area of core thermal-hydraulics. Postulated accidents are evaluated for potential departure from nucleate boiling violations. In support of these tasks, Virginia Power has employed the COBRA code and the W-3 and WRB-1 DNB correlations. A statistical DNBR methodology has also been developed. The code, correlations and statistical methodology are discussed

  13. Thermal-Hydraulic Experiment Facility (THEF)

    International Nuclear Information System (INIS)

    Martinell, J.S.

    1982-01-01

    This paper provides an overview of the Thermal-Hydraulic Experiment Facility (THEF) at the Idaho National Engineering Laboratory (INEL). The overview describes the major test systems, measurements, and data acquisition system, and presents objectives, facility configuration, and results for major experimental projects recently conducted at the THEF. Plans for future projects are also discussed. The THEF is located in the Water Reactor Research Test Facility (WRRTF) area at the INEL

  14. Thermal-hydraulics of actinide burner reactors

    International Nuclear Information System (INIS)

    Takizuka, Takakazu; Mukaiyama, Takehiko; Takano, Hideki; Ogawa, Toru; Osakabe, Masahiro.

    1989-07-01

    As a part of conceptual study of actinide burner reactors, core thermal-hydraulic analyses were conducted for two types of reactor concepts, namely (1) sodium-cooled actinide alloy fuel reactor, and (2) helium-cooled particle-bed reactor, to examine the feasibility of high power-density cores for efficient transmutation of actinides within the maximum allowable temperature limits of fuel and cladding. In addition, calculations were made on cooling of actinide fuel assembly. (author)

  15. An improved numerical model for the investigation of thermal hydraulic phenomena with applications to LMR reactor components

    International Nuclear Information System (INIS)

    Chan, B.C.; Kennett, R.J.; Van Tuyle, G.J.

    1992-01-01

    A basic limited scope, fast-running computer model is presented for the solution of single phase two-dimensional transients in thermally coupled incompressible fluid flow problems. The governing equations and the two-equation transport model (k-ε) of turbulence are reduced to a set of linear algebraic equations in an implicit finite difference scheme, based on the control volume approach. These equations are solved iteratively in a line-by-line procedure using the tri-diagonal matrix algorithm. The numerical formulation and general calculational procedure are described in detail. The calculations show good agreement when compared with experimental data and other independent analyses

  16. 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)

  17. Theoretical thermal-hydraulic investigations of the operational and safety features of the cold D2-moderator for SINQ

    International Nuclear Information System (INIS)

    Skala, K.; Spitzer, H.; Bauer, G.S.

    1991-01-01

    In order to be able to investigate the behaviour of the gas-liquid mixture in the rather complex vessel and pipe system of the cold D 2 -moderator with heat exchanger and ballast volume, a computational model of the system has been set up. The model was first benchmarked against experimental data from the cold moderator mockup for the second cold source at the ILL Grenoble. It was then used to predict operational parameters such as pressure, velocity, moisture content and others of the system under normal conditions with the anticipated distribution of heat influx. Finally, the case of total loss of insulating vacuum was investigated. (author)

  18. Development of regulatory technology for thermal-hydraulic safety analysis

    International Nuclear Information System (INIS)

    Bang, Young Seok; Lee, S. H.; Ryu, Y. H.

    2001-02-01

    The present study aims to develop the regulation capability in thermal-hydraulic safety analysis which was required for the reasonable safety regulation in the current NPP, the next generation reactors, and the future-type reactors. The fourth fiscal year of the first phase of the research was focused on the following research topics: Investigation on the current status of the thermal-hydraulic safety analysis technology outside and inside of the country; Review on the improved features of the thermal-hydraulic safety analysis regulatory audit code, RELAP5/MOD3; Assessments of code with LOFT L9-3 ATWS experiment and LSTF SB-SG-10 multiple SGTR experiment; Application of the RELAP5/CANDU code to analyses of SLB and LBLOCA and evaluation of its effect on safety; Application of the code to IAEA PHWR ISP analysis; Assessments of RELAP5 and TRAC with UPTF downcomer injection test and Analysis of LBLOCA with RELAP5 for the performance evaluation of KNGR DVI; Setup of a coupled 3-D kinetics and thermal-hydraulics and application it to a reactivity accident analysis; and Extension of database and improvement of plant input decks. For supporting the resolution of safety issues, loss of RHR event during midloop operation was analyzed for Kori Unit 3, issues on high burnup fuel were reviewed and performance of FRAPCON-3 assessed. Also MSLB was analyzed to figure out the sensitivity of downcomer temperature supporting the PTS risk evaluation of Kori Unit 1. Thermal stratification in pipe was analyzed using the method proposed. And a method predicting the thermal-hydraulic performance of IRWST of KNGR was explored. The PWR ECCS performance criteria was issued as a MOST Article 200-19.and a regulatory guide on evaluation methodology was improved to cover concerns raised from the related licensing review process

  19. Experimental Investigation on Effect of Fin Shape on the Thermal-Hydraulic Performance of Compact Fin-and-Tube Heat Exchangers

    Science.gov (United States)

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

    2018-03-01

    The aim of this paper is to investigate the effect of fin shapes on the performance of compact finned flat tube heat exchangers. Three types of fin shapes namely plain, wavy, and rectangular grooved fins attached to three by three arrays of flat tube banks were considered. Moreover, the tubes were deployed in in-line and staggered arrangements. In addition to the fin shapes, the air velocity and the tube inclination angles were varied and the thermal-hydraulic performance was analysed. On the other hand, the temperatures at the tube surfaces were kept constant to produce constant heat flux throughout the study. The results showed that as flowrate increases, the heat transfer increases, however, the friction factor decreases. Staggered arrangement produces higher heat transfer and friction factor than inline fin. Moreover, the rectangular fin is the best in terms of high heat transfer however the drawback of high friction factor leads the fin to have the least efficiency of all. On the other hand, plain fin had the least heat transfer performance however the highest efficiency was achieved. Therefore, plain fin should be used when efficiency is prioritized and rectangular fin when high heat transfer is desired.

  20. 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.

  1. Investigating the thermal hydraulic performance of spacer grid with mixing vanes using STAR-CCM+ and MATRA

    International Nuclear Information System (INIS)

    Agbodemegbe, V. Y.

    2014-07-01

    Enhancement of heat transfer for flow through rod bundles is linked to the extent and sustainability of mixing in the flow geometry. Spacer grids used as support for rod bundles in nuclear reactors, when attached with ditferent designs of mixing vanes promote turbulent mixing by inducing swirl or forced lateral convection that improves mixing within or between sub-channels. The improved turbulent mixing raises the margin of the onset of critical heat flux in light water reactors (LWR) and also ensures a higher fuel cycle economy. To optimize design of mixing vanes and performance of spacer grids with mixing vanes, computational fluid dynamic simulations arc carried out on new designs and validated experimentally prior to industrial application. In the present study, computational fluid dynamic simulation using STAR-CCM+ and sub-channel analysis with MATRA were performed for flow of water through a 5 X 5 rod bundle geometry for which the rod to rod pitch to diameter ratio was 1.33 and the wall to rod pitch to diameter ratio was 0.74. The two layer k-epsilon turbulence model with an all- y + automatic wall treatment function in STAR-CCM+ were adopted for an isothermal single phase flow through the geometry with imposed cyclic periodic and non-cyclic periodic interface boundary conditions. The objective was to primarily investigate the detail flow behavior in rod bundle in the presence of spacer grids with and without attached mixing vanes. Furthermore, the present study also investigated the extent of turbulent mixing and lateral mass flux induced by the mixing vanes through comparative analysis of parametric trends for spacer grid with and without mixing vanes. New models and empirical correlations for describing the mixing vanes effects were also developed. These semi-empirical correlations improved the prediction for lateral mass flux due to turbulence, fraction of flow diverted through gaps and cross-flow resistance coefficients. Validation of simulation results

  2. Experimental investigation of the thermal hydraulics in lead bismuth eutectic-helium experimental loop of an accelerator-driven system

    Energy Technology Data Exchange (ETDEWEB)

    Xi, Wenxuan; Wang, Yong Wei; Li, Xun Feng; Huai, Xiulan; Cal, Jun [Institute of Engineering Thermophysics, Chinese Academy of Sciences, Beijing (China)

    2016-10-15

    The heat transfer characteristics between liquid lead bismuth eutectic (LBE) and helium are of great significance for the two-loop cooling system based on an accelerator-driven system (ADS). This paper presents an experimental study on the resistance characteristics and heat transfer performance in a LBE-helium experimental loop of ADS. Pressure drops in the LBE loop, the main heat transfer, and the coupled heat transfer characteristics between LBE and helium are investigated experimentally. The temperature of LBE has a significant effect on the LBE thermo-physical properties, and is therefore considered in the prediction of pressure drops. The results show that the overall heat transfer coefficient increases with the increasing helium flow rate and the decreasing inlet temperature of helium. Increasing the LBE Reynolds number and LBE inlet temperature promotes the heat transfer performance of main heat transfer and thus the overall heat transfer coefficient. The experimental results give an insight into the flow and heat transfer properties in a LBE-helium heat exchanger and are helpful for the optimization of an ADS system design.

  3. Preliminary investigation of the potential for transient vapor release events during in situ vitrification based on thermal- hydraulic modeling

    International Nuclear Information System (INIS)

    Roberts, J.S.; Woosley, S.L.; Lessor, D.L.; Strachan, C.

    1992-07-01

    This study investigates a possible cause of molten glass displacements that occurred during two recent in situ vitrification (ISV) tests. The study was conducted for the US Department of Energy by Pacific Northwest Laboratory. It is hypothesized that these glass displacements are caused by large gas bubbles rising up through the ISV melt and bursting at its surface. These bubbles cause the molten surface to upwell and possibly overflow. When the bubbles burst, molten glass is thrown from the melt surface and the volume of gas contained in the bubble is released into the hood. Both of these phenomena are undesirable because the molten soil ejected from the melt is dangerous to operating personnel and can damage equipment. The sudden gas release can cause a temporary pressurization of the hood, allowing potentially contaminated gas to escape to the atmosphere. This study attempts to explain the conditions necessary for formation of large gas bubbles in the melt so that future glass displacements can be avoided

  4. Preliminary investigation of the potential for transient vapor release events during in situ vitrification based on thermal- hydraulic modeling

    Energy Technology Data Exchange (ETDEWEB)

    Roberts, J.S.; Woosley, S.L.; Lessor, D.L.; Strachan, C.

    1992-07-01

    This study investigates a possible cause of molten glass displacements that occurred during two recent in situ vitrification (ISV) tests. The study was conducted for the US Department of Energy by Pacific Northwest Laboratory. It is hypothesized that these glass displacements are caused by large gas bubbles rising up through the ISV melt and bursting at its surface. These bubbles cause the molten surface to upwell and possibly overflow. When the bubbles burst, molten glass is thrown from the melt surface and the volume of gas contained in the bubble is released into the hood. Both of these phenomena are undesirable because the molten soil ejected from the melt is dangerous to operating personnel and can damage equipment. The sudden gas release can cause a temporary pressurization of the hood, allowing potentially contaminated gas to escape to the atmosphere. This study attempts to explain the conditions necessary for formation of large gas bubbles in the melt so that future glass displacements can be avoided.

  5. Experimental investigation of thermal-hydraulic performance of PCCS with horizontal tube heat exchangers: single U-tube test

    International Nuclear Information System (INIS)

    Nakamura, Hideo; Anoda, Yoshinari; Arai, Kenji; Kurita, Tomohisa

    2000-01-01

    JAERI and JAPC started a cooperative study to verify performance of a PCCS (Passive Containment Cooling System) using horizontal heat exchanger for next-generation BWR in 1998. A test facility with a horizontal single U-tube was constructed in JAERI in 1999 to investigate fundamental condensation behavior under influences of non-condensable gas. Preliminary pre-test analyses were performed using RELAP5/ MOD3.2.1.2 code to expect the experimental outcomes by incorporating a correlation for condensation degradation because of non-condensable gas by Ueno et al. for better prediction. Preliminary results from both experiments (shakedown) and pre-test analyses indicated that the PCCS using horizontal U-tube heat exchanger is promising. Steam generated under assumed severe accident conditions; steam generation rate approx. = 1% core power, non-condensable gas concentration of 1% and simulated containment vessel pressure of 0.7 MPa, was totally condensed with a small differential pressure across inlet and outlet plenum. Experimental data will be accumulated to develop models and correlations for a better prediction of responses of the PCCS using horizontal heat exchanger during postulated severe accidents. (author)

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

    International Nuclear Information System (INIS)

    Heard, F.J.; Cramer, E.R.; Beaver, T.R.; Thurgood, M.J.

    1996-01-01

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

  7. SBWR core thermal hydraulic analysis during startup

    International Nuclear Information System (INIS)

    Lin, J.H.; Huang, R.L.; Sawyer, C.D.

    1993-01-01

    This paper reports on a thermal hydraulic analysis of the SIMPLIFIED BOILING WATER REACTOR (SBWR) during startup. The potential instability during a SBWR startup has drawn the attention of designers, researchers, and engineers. It has not been a concern for a Boiling Water Reactor (BWR) with forced recirculation; however, for SBWR with natural circulation the concern exists. The concern is about the possibility of a geysering mode oscillation during SBWR startup from a cold temperature and a low system pressure with a low natural circulation flow rate. A thermal hydraulic analysis of the SBWR is performed in simulation of the startup using the TRACG computer code. The temperature, pressure, and reactor power profiles of SBWR during the startup are presented. The results are compared with the data of a natural circulation boiling water reactor, the DODEWAARD plant, in which no instabilities have been observed during many startups. It is shown that a SBWR startup which follows proper procedures, geysering and other modes of oscillations can be avoided

  8. Thermal hydraulic design of PFBR core

    International Nuclear Information System (INIS)

    Roychowdhury, D.G.; Vinayagam, P.P.; Ravichandar, S.C.

    2000-01-01

    The thermal-hydraulic design of core is important in respecting temperature limits while achieving higher outlet temperature. This paper deals with the analytical process developed and implemented for analysing steady state thermal-hydraulics of PFBR core. A computer code FLONE has been developed for optimisation of flow allocation through the subassemblies (SA). By calibrating β n (ratio between the maximum channel temperature rise and SA average temperature rise) values with SUPERENERGY code and using these values in FLONE code, prediction of average and maximum coolant temperature distribution is found to be reasonably accurate. Hence, FLONE code is very powerful design tool for core design. A computer code SAPD has been developed to calculate the pressure drop of fuel and blanket SA. Selection of spacer wire pitch depends on the pressure drop, flow-induced vibration and the mixing characteristics. A parametric study was made for optimisation of spacer wire pitch for the fuel SA. Experimental programme with 19 pin-bundle has been undertaken to find the flow-induced vibration characteristics of fuel SA. Also, experimental programme has been undertaken on a full-scale model to find the pressure drop characteristics in unorificed SA, orifices and the lifting force on the SA. (author)

  9. 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.

  10. Investigation of break location effects on thermal-hydraulics during intermediate break loss-of-coolant accident experiments at ROSA-III

    International Nuclear Information System (INIS)

    Koizumi, Yasuo; Tasaka, Kanji

    1986-01-01

    The rig of safety assessment (ROSA)-III facility is a volumetrically scaled (1/424) boiling water reactor (BWR/6) system with an electrically heated core designed for integral loss-of-coolant accident (LOCA) and emergency core cooling system (ECCS) tests. Break location effects on thermal-hydraulics during intermediate LOCAs were investigated by using four experiments at the ROSA-III, the 15 and 25 % main recirculation pump suction line break (MRPS-B) experiments, the 21 % single-ended jet pump drive line break (JPD-B) experiment and the 15 % main steam line break (MSL-B) experiment. Water injection from the high pressure core spray (HPCS) was not used in any of the experiments. Failure of ECCS actuation by the high containment pressure was also assumed in the tests. In the MRPS-B experiments, the discharge flow turned from low quality fluid to high quality fluid when the downcomer water level dropped to the main recirculation line outlet elevation, which suppressed coolant loss from the vessel and the core. In the JPD-B experiment, the jet pump drive nozzle was covered with low quality fluid and low quality fluid discharge continued even after the downcomer water level reached the jet pump suction elevation. Low quality fluid discharge ceased after the ADS actuation. It suggestes that the JPD-B LOCA has the possibility of causing larger and more severe core dryout and cladding temperature excursion than the MRPS-B LOCA. The MSL-B LOCA was characterized by mixture level swell in the downcomer and the core. The core mixture level swell resulted in the much later core dryout initiation than that in the MRPS-B LOCA, however, ECCS actuation was also delayed because of slow downcomer water level drop. (author)

  11. BWR 9 X 9 Fuel Assembly Thermal-Hydraulic Tests (2): Hydraulic Vibration Test

    International Nuclear Information System (INIS)

    Yoshiaki Tsukuda; Katsuichiro Kamimura; Toshiitsu Hattori; Akira Tanabe; Noboru Saito; Masahiko Warashina; Yuji Nishino

    2002-01-01

    Nuclear Power Engineering Corporation (NUPEC) conducted thermal-hydraulic projects for verification of thermal-hydraulic design reliability for BWR high-burnup 8 x 8 and 9 x 9 fuel assemblies, entrusted by the Ministry of Economy, Trade and Industry (METI). As a part of the NUPEC thermal-hydraulic projects, hydraulic vibration tests using full-scale test assemblies simulating 9 x 9 fuel assemblies were carried out to evaluate BWR fuel integrity. The test data were applied to development of a new correlation for the estimation of fuel rod vibration amplitude. (authors)

  12. Thermal hydraulic model validation for HOR mixed core fuel management

    International Nuclear Information System (INIS)

    Gibcus, H.P.M.; Vries, J.W. de; Leege, P.F.A. de

    1997-01-01

    A thermal-hydraulic core management model has been developed for the Hoger Onderwijsreactor (HOR), a 2 MW pool-type university research reactor. The model was adopted for safety analysis purposes in the framework of HEU/LEU core conversion studies. It is applied in the thermal-hydraulic computer code SHORT (Steady-state HOR Thermal-hydraulics) which is presently in use in designing core configurations and for in-core fuel management. An elaborate measurement program was performed for establishing the core hydraulic characteristics for a variety of conditions. The hydraulic data were obtained with a dummy fuel element with special equipment allowing a.o. direct measurement of the true core flow rate. Using these data the thermal-hydraulic model was validated experimentally. The model, experimental tests, and model validation are discussed. (author)

  13. 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...

  14. 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...

  15. Thermal hydraulic reactor safety analyses and experiments

    International Nuclear Information System (INIS)

    Holmstroem, H.; Eerikaeinen, L.; Kervinen, T.; Kilpi, K.; Mattila, L.; Miettinen, J.; Yrjoelae, V.

    1989-04-01

    The report introduces the results of the thermal hydraulic reactor safety research performed in the Nuclear Engineering Laboratory of the Technical Research Centre of Finland (VTT) during the years 1972-1987. Also practical applications i.e. analyses for the safety authorities and power companies are presented. The emphasis is on description of the state-of-the-art know how. The report describes VTT's most important computer codes, both those of foreign origin and those developed at VTT, and their assessment work, VTT's own experimental research, as well as international experimental projects and other forms of cooperation VTT has participated in. Appendix 8 contains a comprehensive list of the most important publications and technical reports produced. They present the content and results of the research in detail.(orig.)

  16. An overview on rod-bundle thermal-hydraulic analyses

    International Nuclear Information System (INIS)

    Sha, W.T.

    1980-01-01

    Three methods used in rod-bundle thermal-hydraulic analysis are summarized. These methods are: (1) subchannel analysis, (2) porous medium formulation with volume porosity, surface permeability, distributed resistance and distributed heat source (sink) and, (3) bench-mark rod-bundle thermal-hydraulic analysis using a boundary-fitted coordinate system. Basic limitations and merits of each method are delineated. (orig.)

  17. 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

  18. 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.

  19. Thermal-hydraulic design of the 200 MW NHR

    International Nuclear Information System (INIS)

    Li Jincai; Gao Zuying; Xu Baocheng; He Junxiao

    1997-01-01

    The thermal hydraulic design of the 200-MW Nuclear Heating Reactor (NHR), design criteria, design methods, important characteristics and some development results are presented in this paper. (author). 5 refs, 8 figs, 2 tabs

  20. Thermal-hydraulic design of the 200 MW NHR

    Energy Technology Data Exchange (ETDEWEB)

    Jincai, Li; Zuying, Gao; Baocheng, Xu; Junxiao, He [Institute of Nuclear Energy and Technology, Tsingua Univ., Beijing (China)

    1997-09-01

    The thermal hydraulic design of the 200-MW Nuclear Heating Reactor (NHR), design criteria, design methods, important characteristics and some development results are presented in this paper. (author). 5 refs, 8 figs, 2 tabs.

  1. Thermal-hydraulic characteristic of the PGV-1000 steam generator

    International Nuclear Information System (INIS)

    Ubra, O.; Doubek, M.

    1995-01-01

    Horizontal steam generators are typical parts of nuclear power plants with pressure water reactor type VVER. By means of this computer program, a detailed thermal-hydraulic study of the horizontal steam generator PGV-1000 has been carried out and a special attention has been paid to the thermal-hydraulics of the secondary side. A set of important steam generator characteristics has been obtained and analyzed. Some of the interesting results of the analysis are presented in the paper. (author)

  2. Process management using component thermal-hydraulic function classes

    Science.gov (United States)

    Morman, J.A.; Wei, T.Y.C.; Reifman, J.

    1999-07-27

    A process management expert system where following malfunctioning of a component, such as a pump, for determining system realignment procedures such as for by-passing the malfunctioning component with on-line speeds to maintain operation of the process at full or partial capacity or to provide safe shut down of the system while isolating the malfunctioning component. The expert system uses thermal-hydraulic function classes at the component level for analyzing unanticipated as well as anticipated component malfunctions to provide recommended sequences of operator actions. Each component is classified according to its thermal-hydraulic function, and the generic and component-specific characteristics for that function. Using the diagnosis of the malfunctioning component and its thermal hydraulic class, the expert system analysis is carried out using generic thermal-hydraulic first principles. One aspect of the invention employs a qualitative physics-based forward search directed primarily downstream from the malfunctioning component in combination with a subsequent backward search directed primarily upstream from the serviced component. Generic classes of components are defined in the knowledge base according to the three thermal-hydraulic functions of mass, momentum and energy transfer and are used to determine possible realignment of component configurations in response to thermal-hydraulic function imbalance caused by the malfunctioning component. Each realignment to a new configuration produces the accompanying sequence of recommended operator actions. All possible new configurations are examined and a prioritized list of acceptable solutions is produced. 5 figs.

  3. Process management using component thermal-hydraulic function classes

    Science.gov (United States)

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

    1999-01-01

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

  4. Hydraulic and thermal design of a gas microchannel heat exchanger

    International Nuclear Information System (INIS)

    Yang Yahui; Brandner, Juergen J; Morini, Gian Luca

    2012-01-01

    In this paper investigations on the design of a gas flow microchannel heat exchanger are described in terms of hydrodynamic and thermal aspects. The optimal choice for thermal conductivity of the solid material is discussed by analysis of its influences on the thermal performance of a micro heat exchanger. Two numerical models are built by means of a commercial CFD code (Fluent). The simulation results provide the distribution of mass flow rate, inlet pressure and pressure loss, outlet pressure and pressure loss, subjected to various feeding pressure values. Based on the thermal and hydrodynamic analysis, a micro heat exchanger made of polymer (PEEK) is designed and manufactured for flow and heat transfer measurements in air flows. Sensors are integrated into the micro heat exchanger in order to measure the local pressure and temperature in an accurate way. Finally, combined with numerical simulation, an operating range is suggested for the present micro heat exchanger in order to guarantee uniform flow distribution and best thermal and hydraulic performances.

  5. Thermal-hydraulic modeling of porous bed reactors

    International Nuclear Information System (INIS)

    Araj, K.J.; Nourbakhsh, H.P.

    1987-01-01

    Optimum design of nuclear reactor cores requires an iterative approach between the thermal-hydraulic, neutronic, and operational analysis. This paper will concentrate on the thermal-hydraulic behavior of a hydrogen-cooled small particle bed reactor (PBR). The PBR core modeled here consists of a hexagonal array of fuel elements embedded in a moderator matrix. The fuel elements are annular packed beds of fuel particles held between two porous cylindrical frits. These particles, 500 to 600 μm in diameter, have a uranium carbide core, which is coated by two layers of graphite and an outer coating of zirconium carbide. Coolant flows, radially inward, from the cold frit through the packed bed and hot frit and axially out the channel, formed by the hot frit to a common plenum. A fast running one-dimensional lumped-parameter steady-state code (FTHP) was developed to evaluate the effects of design changes in fuel assembly and power distribution. Another objective for the code was to investigate various methods of coolant control to minimize hot channel effects and maximize outlet temperatures

  6. Thermal hydraulic analysis of BWR containment venting system

    International Nuclear Information System (INIS)

    Baburajan, P.K.; Sharma, Prashant; Paul, U.K.; Gaikwad, Avinash

    2015-01-01

    Installation of additional containment filtered venting system (CFVS) is necessary to depressurize the containment to maintain its mechanical integrity due to over pressurization during severe accident condition. A typical venting system for BWR is modelled using RELAP5 and analysed to investigate the effect of various thermal hydraulic parameters on the operational parameters of the venting system. The venting system consists of piping from the containment to the scrubber tank and exit line from the scrubber tank. The scrubber tank is partially filled with water to enable the scrubbing action to remove the particulate radionuclides from the incoming containment air. The pipe line from the containment is connected to the venturi inlet and the throat of the venturi is open to the scrubber tank water inventory at designed submergence level. The exit of the venturi is open to scrubber tank water. Filters are used in the upper air space of the scrubber tank as mist separator before venting out the air into the atmosphere through the exit vent line. The effect of thermal hydraulic parameters such as inlet fluid temperature, inlet steam content and venturi submergence in the scrubber tank on the venting flow rate, exit steam content, scrubber tank inventory, overflow line and siphon breaker flow rate is analysed. Results show that inlet steam content and the venturi nozzle submergence influence the venting system parameters. (author)

  7. 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.

  8. Thermal hydraulic issues and challenges for current and new generation FBRs

    Energy Technology Data Exchange (ETDEWEB)

    Chellapandi, P.; Velusamy, K., E-mail: kvelu@igcar.gov.in

    2015-12-01

    Highlights: • We present challenges in thermal hydraulic design of sodium cooled fast reactors. • We present roadmap of Indian fast reactor program and innovative design concepts. • Analysis methodology for thermal striping and thermal stratification are highlighted. • Design solutions for gas entrainment are presented. • Experimental approaches for normal and post accident decay heat removal are highlighted. - Abstract: Pool type sodium cooled fast reactors pose several design challenges and among them, certain thermal hydraulics and structural mechanics issues are special. High frequency temperature fluctuations due to thermal striping, thermal stratifications and sodium free level fluctuations at the liquid–cover gas interfaces are to be investigated carefully to eliminate high cycle thermal fatigue of structures. Solutions to address the core thermal hydraulics call for high power computing. Innovative concepts and methods are developed to carry out plant dynamics and safety studies. Particularly, extensive numerical and experimental simulation techniques are needed for understanding and solving the gas entrainment mechanisms and its effects on core safety. Though decay heat removal through natural convection is achievable in a pool type SFR, demonstration of design solutions conceived in the reactor and performance of diverse systems under all operating conditions, especially over prolonged station blackout situations needs advanced CFD computations and should be validated by relatively large scale simulated experiments. These issues are addressed in this paper under five broad topics: special thermal hydraulic issues to be addressed in SFR, thermal hydraulic design and analysis, plant dynamics studies, safety studies and evolving thermal hydraulic studies for the future FBRs. The 500 MWe Prototype Fast Breeder Reactor (PFBR) is taken as the reference design for addressing the issues. Indian fast reactor programme is highlighted in the introduction

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

    International Nuclear Information System (INIS)

    Song, C. H.; Baek, W. P.; Chung, M. K.

    2007-06-01

    The objectives of the project are to study thermal hydraulic characteristics of advanced nuclear reactor system for evaluating key thermal-hydraulic phenomena relevant to new safety concepts. 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. The Followings are main research topics: - Multi-dimensional Phenomena in a Reactor Vessel Downcomer - Condensation-induced Thermal Mixing in a Pool - Development of Thermal-Hydraulic Models for Two-Phase Flow - Construction of T-H Data Base

  10. 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.

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

    International Nuclear Information System (INIS)

    Ubra, O.; Doubek, M.

    1995-01-01

    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.)

  12. 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)

    1996-12-31

    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.

  13. Thermal Hydraulic Integral Effect Tests for Pressurized Water Reactors

    International Nuclear Information System (INIS)

    Baek, Won Pil; Song, C. H.; Kim, Y. S.

    2007-02-01

    The objectives of the project are to construct a thermal-hydraulic integral effect test facility and to perform the tests for design, operation, and safety regulation of pressurized water reactors. In the first phase of this project (1997.8∼2002.3), the basic technology for thermal-hydraulic integral effect tests was established and the basic design of the test facility was accomplished. In the second phase (2002.4∼2005.2), an optimized design of the ATLAS (Advanced Thermal-hydraulic Test Loop for Accident Simulation) was established and the construction of the facility was almost completed. In the third phase (2005.3∼2007.2), the construction and commission tests of the ATLAS are to be completed and some first-phase tests are to be conducted

  14. 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.

  15. HYTRAN: hydraulic transient code for investigating channel flow stability

    International Nuclear Information System (INIS)

    Kao, H.S.; Cardwell, W.R.; Morgan, C.D.

    1976-01-01

    HYTRAN is an analytical program used to investigate the possibility of hydraulic oscillations occurring in a reactor flow channel. The single channel studied is ordinarily the hot channel in the reactor core, which is parallel to other channels and is assumed to share a constant pressure drop with other channels. Since the channel of highest thermal state is studied, provision is made for two-phase flow that can cause a flow instability in the channel. HYTRAN uses the CHATA(1) program to establish a steady-state condition. A heat flux perturbation is then imposed on the channel, and the flow transient is calculated as a function of time

  16. Hydraulic modeling of thermal discharges into shallow, tidal affected streams

    International Nuclear Information System (INIS)

    Copp, H.W.; Shashidhara, N.S.

    1981-01-01

    A two-unit nuclear fired power plant is being constructed in western Washington state. Blowdown water from cooling towers will be discharged into the Chehalis River nearby. The location of a diffuser is some 21 miles upriver from Grays Harbor on the Pacific Ocean. Because the Chehalis River is classified as an excellent stream from the standpoint of water quality, State regulatory agencies required demonstration that thermal discharges would maintain water quality standards within fairly strict limits. A hydraulic model investigation used a 1:12 scale, undistorted model of a 1300-foot river reach in the vicinity of the diffuser. The model scale was selected to insure fully turbulent flows both in the stream and from the diffuser (Reynolds similitude). Model operation followed the densimetric Froude similitude. Thermistors were employed to measure temperatures in the model; measurements were taken by computer command and such measurements at some 250 positions were effected in about 2.5 seconds

  17. Thermal hydraulic and neutronic interaction in the rotating bed reactor

    International Nuclear Information System (INIS)

    Lee, C.C.

    1986-01-01

    Power transient characteristics in a rotating fluidized bed reactor (RBR) are investigated theoretically. A propellant flow perturbation is assumed to occur in an initially equilibrium state of the core. Transfer functions representing quasi-one-dimensional mutual feedback between thermal hydraulics and neutronics are developed and analyzed in the frequency domain. Neutronic responses are determined by Fermi-age theory for slowing down of fast neutrons and diffusion theory for thermal neutron distribution. Neutron leakage through the exhaust nozzle is accounted for by applying diffuse view factors similar to those applied in radiative heat transfer. The bed expansion behavior is described by a kinematic wave equation derived from the continuity of the gas phase. The drift flux approach is used to determine the yield fractions in the equilibrium bed. Thermal responses of fuel are evaluated by dividing it into several volume-averaged zones to better account for the transient effects over single zone models. Sample calculations are undertaken for the various operation conditions and design parameters of the RBR based on 250 MW/sub t/, 1000 MW/sub t/, and 5000 MW/sub t/ power reactors. The results show that power transients are dependent on the parametric changes of optical thickness and view factors

  18. 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.

  19. Thermal hydraulic considerations in liquid-metal-cooled components of tokamak fusion reactors

    International Nuclear Information System (INIS)

    Picologlou, B.F.; Reed, C.B.; Hua, T.Q.

    1989-01-01

    The basic considerations of MHD thermal hydraulics for liquid-metal-cooled blankets and first walls of tokamak fusion reactors are discussed. The liquid-metal MHD program of Argonne National Laboratory (ANL) dedicated to analytical and experimental investigations of reactor relevant MHD flows and development of relevant thermal hydraulic design tools is presented. The status of the experimental program and examples of local velocity measurements are given. An account of the MHD codes developed to date at ANL is also presented as is an example of a 3-D thermal hydraulic analysis carried out with such codes. Finally, near term plans for experimental investigations and code development are outlined. 20 refs., 8 figs., 1 tab

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

    Energy Technology Data Exchange (ETDEWEB)

    Kim, K.

    1977-06-15

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

  1. Thermal-hydraulic methods in fast reactor safety

    International Nuclear Information System (INIS)

    Weber, D.P.; Briggs, L.L.

    1985-01-01

    Methods for the solution of thermal-hydraulic problems in liquid metal fast breeder reactors (LMFBRs) arising primarily from transient accident analysis are reviewed. Principal emphasis is given to the important phenomenological issues of sodium boiling and fuel motion. Descriptions of representative phenomenological and mathematical models, computational algorithms, advantages and limitations of the approaches, and current research needs and directions are provided

  2. Development of thermal hydraulic evaluation code for CANDU reactors

    International Nuclear Information System (INIS)

    Kim, Man Woong; Yu, Seon Oh; Choi, Yong Seog; Shin, Chull; Hwang, Soo Hyun

    2004-02-01

    To enhance the safety of operating CANDU reactors, the establishment of the safety analysis codes system for CANDU reactors is in progress. As for the development of thermal-hydraulic analysis code for CANDU system, the studies for improvement of evaluation model inside RELAP/CANDU code and the development of safety assessment methodology for GSI (Generic Safety Issues) are in progress as a part of establishment of CANDU safety assessment system. To develop the 3-D thermal-hydraulic analysis code for moderator system, the CFD models for analyzing the CANDU-6 moderator circulation are developed. One model uses a structured grid system with the porous media approach for the 380 Calandria tubes in the core region. The other uses a unstructured grid system on the real geometry of 380 Calandria tubes, so that the detailed fluid flow between the Calandria tubes can be observed. As to the development of thermal-hydraulic analysis code for containment, the study on the applicability of CONTAIN 2.0 code to a CANDU containment was conducted and a simulation of the thermal-hydraulic phenomena during the accident was performed. Besides, the model comparison of ESFs (Engineered Safety Features) inside CONTAIN 2.0 code and PRESCON code has also conducted

  3. Design and thermal-hydraulic calculation for EAST PFCs' baking

    International Nuclear Information System (INIS)

    Wan Xiaogang; Yao Damao

    2006-01-01

    According to the vacuum requirements for fusion in a tokamak device, the authors adopted a kind of gas flow baking technique in EAST. This paper presented the sketch design for EAST PFCs' baking, selected the specifications for the working gas. Calculated the hydraulic and thermal conditions in PFCs under baking, and simulated the results. (authors)

  4. Thermal-hydraulic research plan for Babcock and Wilcox plants

    International Nuclear Information System (INIS)

    Lee, R.Y.

    1988-02-01

    This document presents a plan for thermal-hydraulic research for Babcock and Wilcox designed reactor systems. It describes the technical issues, regulatory needs, and the research necessary to address these needs. The plan also discusses the relationship between current and proposed research, and provides a tentative schedule to complete the required work

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

    International Nuclear Information System (INIS)

    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)

  6. Thermal-hydraulic analysis for wire-wrapped PWR cores

    Energy Technology Data Exchange (ETDEWEB)

    Diller, P. [General Electric Company, 3901 Castle Hayne Rd., Wilmington, NC 28401 (United States)], E-mail: pdiller@gmail.com; Todreas, N. [Massachusetts Institute of Technology, Cambridge, MA 02139 (United States)], E-mail: todreas@mit.edu; Hejzlar, P. [Massachusetts Institute of Technology, Cambridge, MA 02139 (United States)

    2009-08-15

    This work focuses on the steady-state and transient thermal-hydraulic analyses for PWR cores using wire wraps in a hexagonal array with either U (45% w/o)-ZrH{sub 1.6} (referred to as U-ZrH{sub 1.6}) or UO{sub 2} fuels. Equivalences (thermal-hydraulic and neutronic) were created between grid spacer and wire wrap designs, and were used to apply results calculated for grid spacers to wire wrap designs. Design limits were placed on the pressure drop, critical heat flux (CHF), fuel and cladding temperature and vibrations. The vibrations limits were imposed for flow-induced vibrations (FIV) and thermal-hydraulic vibrations (THV). The transient analysis examined an overpower accident, loss of coolant accident (LOCA) and loss of flow accident (LOFA). The thermal-hydraulic performance of U-ZrH{sub 1.6} and UO{sub 2} were found very similar. Relative to grid spacer designs, wire wrap designs were found to have smaller fretting wear, substantially lower pressure drop and higher CHF. As a result, wire wrap cores were found to offer substantially higher maximum powers than grid spacer cores, allowing for a 25% power increase relative to the grid spacer uprate [Shuffler, C.A., Malen, J.A., Trant, J.M., Todreas, N.E., 2009a. Thermal-hydraulic analysis for grid supported and inverted fueled PWR cores. Nuclear Technology (this special issue devoted to hydride fuel in LWRs)] and a 58% power increase relative to the reference core.

  7. Nuclear power plant thermal-hydraulic performance research program plan

    International Nuclear Information System (INIS)

    1988-07-01

    The purpose of this program plan is to present a more detailed description of the thermal-hydraulic research program than that provided in the NRC Five-Year Plan so that the research plan and objectives can be better understood and evaluated by the offices concerned. The plan is prepared by the Office of Nuclear Regulatory Research (RES) with input from the Office of Nuclear Reactor Regulation (NRR) and updated periodically. The plan covers the research sponsored by the Reactor and Plant Systems Branch and defines the major issues (related to thermal-hydraulic behavior in nuclear power plants) the NRC is seeking to resolve and provides plans for their resolution; relates the proposed research to these issues; defines the products needed to resolve these issues; provides a context that shows both the historical perspective and the relationship of individual projects to the overall objectives; and defines major interfaces with other disciplines (e.g., structural, risk, human factors, accident management, severe accident) needed for total resolution of some issues. This plan addresses the types of thermal-hydraulic transients that are normally considered in the regulatory process of licensing the current generation of light water reactors. This process is influenced by the regulatory requirements imposed by NRC and the consequent need for technical information that is supplied by RES through its contractors. Thus, most contractor programmatic work is administered by RES. Regulatory requirements involve the normal review of industry analyses of design basis accidents, as well as the understanding of abnormal occurrences in operating reactors. Since such transients often involve complex thermal-hydraulic interactions, a well-planned thermal-hydraulic research plan is needed

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

    International Nuclear Information System (INIS)

    Chung, Moon Ki; Park, Chun Kyeong; Yang, Seon Kyu; Chung, Chang Hwan; Chun, Shee Yeong; Song, Cheol Hwa; Chun, Hyeong Gil; Chang, Seok Kyu; Chung, Heung Joon; Won, Soon Yeon; Cho, Yeong Ro; Kim, Bok Deuk; Min, Kyeong Ho

    1994-07-01

    The present research aims at the development of the thermal hydraulic verification test technology for the reactor safety system of the conventional and advanced nuclear power plant and the development of the advanced thermal hydraulic measuring techniques. (Author)

  9. 78 FR 8202 - Meeting of the Joint ACRS Subcommittees on Thermal Hydraulic Phenomena and Materials, Metallurgy...

    Science.gov (United States)

    2013-02-05

    ... NUCLEAR REGULATORY COMMISSION Advisory Committee on Reactor Safeguards (ACRS) Meeting of the Joint ACRS Subcommittees on Thermal Hydraulic Phenomena and Materials, Metallurgy and Reactor Fuels; Notice of Meeting The Joint ACRS Subcommittees on Thermal Hydraulic Phenomena and Materials, Metallurgy and...

  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. Study of thermal - hydraulic sensors signal fluctuations in PWR

    International Nuclear Information System (INIS)

    Hennion, F.

    1987-10-01

    This thesis deals with signal fluctuations of thermal-hydraulic sensors in the main coolant primary of a pressurized water reactor. The aim of this work is to give a first response about the potentiality of use of these noise signals for the functionning monitoring. Two aspects have been studied: - the modelisation of temperature fluctuations of core thermocouples, by a Monte-Carlo method, gives the main characteristics of these signals and their domain of application. - the determination of eigenfrequency in the primary by an acoustic representation could permit the monitoring of local and global thermo-hydraulic conditions [fr

  12. Investigation of hydraulic transmission noise sources

    Science.gov (United States)

    Klop, Richard J.

    Advanced hydrostatic transmissions and hydraulic hybrids show potential in new market segments such as commercial vehicles and passenger cars. Such new applications regard low noise generation as a high priority, thus, demanding new quiet hydrostatic transmission designs. In this thesis, the aim is to investigate noise sources of hydrostatic transmissions to discover strategies for designing compact and quiet solutions. A model has been developed to capture the interaction of a pump and motor working in a hydrostatic transmission and to predict overall noise sources. This model allows a designer to compare noise sources for various configurations and to design compact and inherently quiet solutions. The model describes dynamics of the system by coupling lumped parameter pump and motor models with a one-dimensional unsteady compressible transmission line model. The model has been verified with dynamic pressure measurements in the line over a wide operating range for several system structures. Simulation studies were performed illustrating sensitivities of several design variables and the potential of the model to design transmissions with minimal noise sources. A semi-anechoic chamber has been designed and constructed suitable for sound intensity measurements that can be used to derive sound power. Measurements proved the potential to reduce audible noise by predicting and reducing both noise sources. Sound power measurements were conducted on a series hybrid transmission test bench to validate the model and compare predicted noise sources with sound power.

  13. First wall thermal hydraulic models for fusion blankets

    International Nuclear Information System (INIS)

    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

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

    International Nuclear Information System (INIS)

    Sathyanarayana, K.

    1998-01-01

    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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-01-15

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

  16. Thermal hydraulic analyses of two fusion reactor first wall/blanket concepts

    International Nuclear Information System (INIS)

    Misra, B.; Maroni, V.A.

    1978-01-01

    A comparative study has been made of the thermal hydraulic performance of two liquid lithium blanket concepts for tokamak-type reactors. In one concept lithium is circulated through 60-cm deep cylindrical modules oriented so that the module axis is parallel to the reactor minor radius. In the other concept helium carrying channels oriented parallel to the first wall are used to cool a 60-cm thick stagnant lithium blanket. Paralleling studies were carried out wherein the thermal and structural properties of the construction materials were based on those projected for either solution-annealed 316-stainless steel or vanadium-base alloys. The effects of limitations on allowable peak structural temperature, material strength, thermal stress, coolant inlet temperature, and pumping power/thermal power ratio were evaluated. Consequences to thermal hydraulic performance resulting from the presence of or absence of a divertor were also investigated

  17. Thermal hydraulic analyses of two fusion reactor first wall/blanket concepts

    International Nuclear Information System (INIS)

    Misra, B.; Maroni, V.A.

    1977-01-01

    A comparative study has been made of the thermal hydraulic performance of two liquid lithium blanket concepts for tokamak-type reactors. In one concept lithium is circulated through 60-cm deep cylindrical modules oriented so that the module axis is parallel to the reactor minor radius. In the other concept helium carrying channels oriented parallel to the first wall are used to cool a 60-cm thick stagnant lithium blanket. Paralleling studies were carried out wherein the thermal and structural properties of the construction materials were based on those projected for either solution-annealed 316-stainless steel or vanadium-base alloys. The effects of limitations on allowable peak structural temperature, material strength, thermal stress, coolant inlet temperature, and pumping power/thermal power ratio were evaluated. Consequences to thermal hydraulic performance resulting from the presence of or absence of a divertor were also investigated

  18. Hierarchic modeling of heat exchanger thermal hydraulics

    International Nuclear Information System (INIS)

    Horvat, A.; Koncar, B.

    2002-01-01

    Volume Averaging Technique (VAT) is employed in order to model the heat exchanger cross-flow as a porous media flow. As the averaging of the transport equations lead to a closure problem, separate relations are introduced to model interphase momentum and heat transfer between fluid flow and the solid structure. The hierarchic modeling is used to calculate the local drag coefficient C d as a function of Reynolds number Re h . For that purpose a separate model of REV is built and DNS of flow through REV is performed. The local values of heat transfer coefficient h are obtained from available literature. The geometry of the simulation domain and boundary conditions follow the geometry of the experimental test section used at U.C.L.A. The calculated temperature fields reveal that the geometry with denser pin-fins arrangement (HX1) heats fluid flow faster. The temperature field in the HX2 exhibits the formation of thermal boundary layer between pin-fins, which has a significant role in overall thermal performance of the heat exchanger. Although presented discrepancies of the whole-section drag coefficient C d are large, we believe that hierarchic modeling is an appropriate strategy for calculation of complex transport phenomena in heat exchanger geometries.(author)

  19. Thermal-hydraulics associated with nuclear education and research

    International Nuclear Information System (INIS)

    Yokobori, Seiichi

    2011-01-01

    This article was the rerecording of the author's lecture at the fourth 'Future Energy Forum' (aiming at improving nuclear safety and economics) held in December 2010. The lecture focused on (1) importance of thermal hydraulics associated with nuclear education and research (critical heat flux, two-phase flow and multiphase flow), (2) emerging trend of maintenance engineering (fluid induced vibration, flow accelerated corrosion and stress corrosion cracks), (3) fostering sensible nuclear engineer with common engineering sense, (4) balanced curriculum of basics and advanced research, (5) computerized simulation and fluid mechanics, (6) crucial point of thermo hydraulics education (viscosity, flux, steam and power generation), (7) safety education and human resources development (indispensable technologies such as defence in depth) and (8) topics of thermo hydraulics research (vortices of curbed pipes and visualization of two-phase flow). (T. Tanaka)

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

    Indian Academy of Sciences (India)

    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 ...

  1. Thermal-hydraulic characteristics of double flat core HCLWR

    International Nuclear Information System (INIS)

    Sugimoto, Jun; Iwamura, Takamichi; Okubo, Tsutomu; Murao, Yoshio

    1989-02-01

    A thermal-hydraulic characteristics of double flat core high conversion light water reactor (HCLWR) is described. The concept of flat core proposed by Ishiguro et al. is to achieve negative void reactivity coefficient in tight lattice core, and at the same time, high conversion ratio and high burnup can be obtainable. The proposed double flat core HCLWR, based on these physical advantages and the consideration of safety assurance, aims at efficient use of the pressure vessel space to produce comparable thermal output as current 3-loop PWRs. The present work revealed the following items concerning the thermalhydraulic feasibility of the double flat core HCLWR: (1) Main thermal-hydraulic parameters of the plant can be almost the same as current PWRs, showing the use of PWR standard components without major modifications except in core region. (2) Heat removal from the fuel rod in a steady operational condition has enough margin to the critical heat flux (CHF) limit, which is evaluated with the existing CHF correlations. (3) The calculation by REFLA code shows that the maximum cladding temperature in LOCA-reflood is estimated to be far lower than the licensing criteria. It is therefore considered that the proposed double flat core HCLWR is feasible from the point of thermal-hydraulics. Since the available data base has certain applicational limit to the very short core as the present double flat core HCLWR, further detailed assessment is required. (author)

  2. VHTR core modeling: coupling between neutronic and thermal-hydraulics

    International Nuclear Information System (INIS)

    Limaiem, I.; Damian, F.; Raepsaet, X.; Studer, E.

    2005-01-01

    Following the present interest in the next generation nuclear power plan (NGNP), Cea is deploying special effort to develop new models and qualify its research tools for this next generation reactors core. In this framework, the Very High Temperature Reactor concept (VHTR) has an increasing place in the actual research program. In such type of core, a strong interaction exists between neutronic and thermal-hydraulics. Consequently, the global core modelling requires accounting for the temperature feedback in the neutronic models. The purpose of this paper is to present the new neutronic and thermal-hydraulics coupling model dedicated to the High Temperature Reactors (HTR). The coupling model integrates a new version of the neutronic scheme calculation developed in collaboration between Cea and Framatome-ANP. The neutronic calculations are performed using a specific calculation processes based on the APOLLO2 transport code and CRONOS2 diffusion code which are part of the French reactor physics code system SAPHYR. The thermal-hydraulics model is characterised by an equivalent porous media and 1-D fluid/3-D thermal model implemented in the CAST3M/ARCTURUS code. The porous media approach involves the definition of both homogenous and heterogeneous models to ensure a correct temperature feedback. This study highlights the sensitivity of the coupling system's parameters (radial/axial meshing and data exchange strategy between neutronic and thermal-hydraulics code). The parameters sensitivity study leads to the definition of an optimal coupling system specification for the VHTR. Besides, this work presents the first physical analysis of the VHTR core in steady-state condition. The analysis gives information about the 3-D power peaking and the temperature coefficient. Indeed, it covers different core configurations with different helium distribution in the core bypass. (authors)

  3. Thermal hydraulics of accelerator driven system windowless targets

    Directory of Open Access Journals (Sweden)

    Bruno ePanella

    2015-07-01

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

  4. THERMAL HYDRAULIC ANALYSIS OF FIRE DIVERTOR

    International Nuclear Information System (INIS)

    C.B. bAXI; M.A. ULRICKSON; D.E. DRIMEYER; P. HEITZENROEDER

    2000-01-01

    The Fusion Ignition Research Experiment (FIRE) is being designed as a next step in the US magnetic fusion program. The FIRE tokamak has a major radius of 2 m, a minor radius of 0.525 m, and liquid nitrogen cooled copper coils. The aim is to produce a pulse length of 20 s with a plasma current of 6.6 MA and with alpha dominated heating. The outer divertor and baffle of FIRE are water cooled. The worst thermal condition for the outer divertor and baffle is the baseline D-T operating mode (10 T, 6.6 MA, 20 s) with a plasma exhaust power of 67 MW and a peak heat flux of 20 MW/m 2 . A swirl tape (ST) heat transfer enhancement method is used in the outer divertor cooling channels to increase the heat transfer coefficient and the critical heat flux (CHF). The plasma-facing surface consists of tungsten brush. The finite element (FE) analysis shows that for an inlet water temperature of 30 C, inlet pressure of 1.5 MPa and a flow velocity of 10 m/s, the incident critical heat flux is greater than 30 MW/m 2 . The peak copper temperature is 490 C, peak tungsten temperature is 1560 C, and the pressure drop is less than 0.5 MPa. All these results fulfill the design requirements

  5. 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.

  6. Thermal-hydraulic modeling of porous bed reactors

    International Nuclear Information System (INIS)

    Araj, K.J.; Nourbakhsh, H.P.

    1987-01-01

    Optimum design of nuclear reactor core requires an iterative approach between the thermal-hydraulic, neutronic and operational analysis. This paper concentrates on the thermal-hydraulic behavior of a hydrogen cooled, small particle bed reactor (PBR). The PBR core, modeled here, consists of a hexagonal array of fuel elements embedded in a moderator matrix. The fuel elements are annular packed beds of fuel particles held between two porous cylindrical frits. These particles, 500 to 600 μm in diameter, have a uranium carbide core, which is coated by two layers of graphite and an outer coating of zirconium carbide. Coolant flow, radially inward, from the cold frit through the packed bed and hot frit and axially out the channel, formed by the hot frit, to a common plenum. 5 refs., 1 fig., 2 tabs

  7. 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.

  8. Simulation of Thermal Hydraulic at Supercritical Pressures with APROS

    Energy Technology Data Exchange (ETDEWEB)

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

    2008-07-01

    The proposed concepts for the fourth generation of nuclear reactors include a reactor operating with water at thermodynamically supercritical state, the Supercritical Water Reactor (SCWR). For the design and safety demonstrations of such a reactor, the possibility to accurately simulate the thermal hydraulics of the supercritical coolant is an absolute prerequisite. For this purpose, the one-dimensional two-phase thermal hydraulics solution of APROS process simulation software was developed to function at the supercritical pressure region. Software modifications included the redefinition of some parameters that have physical significance only at the subcritical pressures, improvement of the steam tables, and addition of heat transfer and friction correlations suitable for the supercritical pressure region. (author)

  9. Thermal hydraulic simulation of the CANDU nuclear reactor

    Energy Technology Data Exchange (ETDEWEB)

    Carvalho, Athos M.S.S. de; Ramos, Mario C.; Costa, Antonella L.; Fernandes, Gustavo H.N., E-mail: athos1495@yahoo.com.br [Universidade Federal de Minas Gerais (UFMG), Belo Horizonte, MG (Brazil). Departamento de Engenharia Nuclear; Instituto Nacional de Ciência e Tecnologia de Reatores Nucleares Inovadores (INCT/CNPq), Rio de janeiro, RJ (Brazil)

    2017-07-01

    The CANDU (Canada Deuterium Uranium) is a Canadian-designed power reactor of PHWR type (Pressurized Heavy Water Reactor) that uses heavy water (deuterium oxide) for moderator and coolant, and natural uranium for fuel. There are about 47 reactors of this type in operation around the world generating more than 23 GWe, highlighting the importance of this kind of device. In this way, the main purpose of this study is to develop a thermal hydraulic model for a CANDU reactor to aggregate knowledge in this line of research. In this way, a core modeling was performed using RELAP5-3D code. Results were compared with reference data to verify the model behavior in steady state operation. Thermal hydraulic parameters as temperature, pressure and mass flow rate were verified and the results are in good agreement with reference data, as it is being presented in this work. (author)

  10. Thermal-hydraulic tests for reactor safety system

    International Nuclear Information System (INIS)

    Chun, Se Young; Chung, Moon Ki; Baek, Won Pil

    2002-05-01

    Tests for the safety depressurization system, Sparger adopted for the Korean next generation reactor, APR1400 are carried out for several geometries with the B and C (Blowdown and Condensation) facility in the condition of high temperature and pressure and with a small test facility in the condition of atmospheric temperature and pressure. Tests for the critical heat flux are performed with the RCS(Reactor Coolant System) facility as well as with the Freon CHF Loop in the condition of high temperature and pressure. The atmospheric temperature and pressure facility is utilized for development of the high standard thermal hydraulic measurement technology. The optical method is developed to measure the local thermal-hydraulic behavior for the single and two-phase boiling phenomena

  11. Equipping simulators with an advanced thermal hydraulics model EDF's experience

    International Nuclear Information System (INIS)

    Soldermann, R.; Poizat, F.; Sekri, A.; Faydide, B.; Dumas, J.M.

    1997-01-01

    The development of an accelerated version of the advanced CATHARe-1 thermal hydraulics code designed for EDF training simulators (CATHARE-SIMU) was successfully completed as early as 1991. Its successful integration as the principal model of the SIPA Post-Accident Simulator meant that its use could be extended to full-scale simulators as part of the renovation of the stock of existing simulators. In order to further extend the field of application to accidents occurring in shutdown states requiring action and to catch up with developments in respect of the CATHARE code, EDF initiated the SCAR Project designed to adapt CATHARE-2 to simulator requirements (acceleration, parallelization of the computation and extension of the simulation range). In other respects, the installation of SIPA on workstations means that the authors can envisage the application of this remarkable training facility to the understanding of thermal hydraulics accident phenomena

  12. A study on the thermal hydraulics in rod bundles

    International Nuclear Information System (INIS)

    Chung, Moon Ki; Yang, Sun Kyu

    1989-03-01

    In order to improve the thermal hydraulic characteristics of the nuclear reactor core, it is necessary to obtain better understanding of the coolant flow and the enthalpy distribution in complex rod bundle geometries. The purpose of this report is to obtain a comprehensive survey on the thermal hydraulic in rod bundles from both experimental and numerical point of view. From references on experimental study, measurement methods and results of the flow velocity and the pressure drop in the subchannels of rod bundles are expressed. The microscopic flow characteristics of the subchannels and spacer grid effect on the flow structure are described. Physical phenomena and measurement methods of the secondary flow are also described. From references on the numerical study, general numerical methods are expressed. Numerical studies on the laminar flow and turbulent flow such as 1-equation and 2-equation model are reviewed.(Author)

  13. Outage Risk Assessment and Management (ORAM) thermal-hydraulics toolkit

    International Nuclear Information System (INIS)

    Denny, V.E.; Wassel, A.T.; Issacci, F.; Pal Kalra, S.

    2004-01-01

    A PC-based thermal-hydraulic toolkit for use in support of outage optimization, management and risk assessment has been developed. This mechanistic toolkit incorporates simple models of key thermal-hydraulic processes which occur during an outage, such as recovery from or mitigation of outage upsets; this includes heat-up of water pools following loss of shutdown cooling, inadvertent drain down of the RCS, boiloff of coolant inventory, heatup of the uncovered core, and reflux cooling. This paper provides a list of key toolkit elements, briefly describes the technical basis and presents illustrative results for RCS transient behavior during reflux cooling, peak clad temperatures for an uncovered core and RCS response to loss of shutdown cooling. (author)

  14. State of the art of thermal-hydraulics of BWRs

    International Nuclear Information System (INIS)

    Rouhani, Z.

    1980-10-01

    The present report is a summary review of the developments in the field of thermal-hydraulics of Boiling Water Reactors. It covers briefly the development of BWR systems, including some comparison of the main features of the modern BWRs that are marketed by different vendors. The analytical aspects of BWR are also covered briefly with some remarks on the problem areas and limitations in this field. (author)

  15. Application of thermal-hydraulic codes in the nuclear sector

    International Nuclear Information System (INIS)

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

    2011-01-01

    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. 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.

  17. 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.

  18. Thermal-hydraulic interfacing code modules for CANDU reactors

    International Nuclear Information System (INIS)

    Liu, W.S.; Gold, M.; Sills, H.

    1997-01-01

    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

  19. CFD studies on thermal hydraulics of spallation targets

    International Nuclear Information System (INIS)

    Tak, N.I.; Batta, A.; Cheng, X.

    2005-01-01

    Full text of publication follows: Due to the fast advances in computer hardware as well as software in recent years, more and more interests have been aroused to use computational fluid dynamics (CFD) technology in nuclear engineering and designs. During recent many years, Forschungszentrum Karlsruhe (FZK) has been actively involved in the thermal hydraulic analysis and design of spallation targets. To understand the thermal hydraulic behaviors of spallation targets very detailed simulations are necessary because of their complex geometries, complicated boundary conditions such as spallation heat distributions, and very strict design limits. A CFD simulation is believed to be the best for this purpose even though the validation of CFD codes are not perfectly completed yet in specific topics like liquid metal heat transfer. The research activities on three spallation targets (i.e., MEGAPIE, TRADE, and XADS targets) are currently very active in Europe in order to consolidate the European ADS road-map. In the thermal hydraulics point of view, two kinds of the research activities, i.e., (1) numerical design and (2) experimental work, are required to achieve the objectives of these targets. It should be noted that CFD studies play important role on both kinds of two activities. A preliminary design of a target can be achieved by sophisticated CFD analysis and pre-and-post analyses of an experimental work using a CFD code help the design of the test section of the experiment as well as the analysis of the experimental results. The present paper gives an overview about the recent CFD studies relating to thermal hydraulics of the spallation targets recently involved in FZK. It covers numerical design studies as well as CFD studies to support experimental works. The CFX code has been adopted for the studies. Main recent results for the selected examples performed by FZK are presented and discussed with their specific lessons learned. (authors)

  20. 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.

  1. 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.

  2. Views on the future of thermal hydraulic modeling

    International Nuclear Information System (INIS)

    Ishii, M.

    1997-01-01

    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

  3. 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)

  4. CFD thermal-hydraulic analysis of a CANDU fuel channel

    International Nuclear Information System (INIS)

    Catana, A.; Prisecaru, I.; Dupleac, D.; Danila, N.

    2009-01-01

    This paper presents the numerical investigation of a CANDU fuel channel using CFD (Computational fluid dynamics) methodology approach. Limited computer power available at Bucharest University POLITEHNICA forced the authors to analyse only segments of fuel channel namely the significant ones: fuel bundle junctions with adjacent segments, fuel bundle spacer planes with adjacent segments, regular segments of fuel bundles. The computer code used is FLUENT. Fuel bundles contained in pressure tubes forms a complex flow domain. The flow is characterized by high turbulence and in some parts of fuel channel also by multi-phase flow. The flow in the fuel channel has been simulated by solving the equations for conservation of mass and momentum. For turbulence modelling the standard k-e model is employed although other turbulence models can be used as well. In this paper we do not consider heat generation and heat transfer capabilities of CFD methods. Since we consider only some relatively short segments of a CANDU fuel channel we can assume, for this starting stage, that heat transfer is not very important for these short segments of fuel channel. The boundary conditions for CFD analysis are provided by system and sub-channel analysis. In this paper the discussion is focused on some flow parameters behaviour at the bundle junction, spacer's plane configuration, etc. In this paper we present results for Standard CANDU 6 Fuel Bundles as a basis for CFD thermal-hydraulic analysis of INR proposed SEU43 and other new nuclear fuels. (authors)

  5. Local thermal-hydraulic behaviour in tight 7-rod bundles

    International Nuclear Information System (INIS)

    Cheng, X.; Yu, Y.Q.

    2009-01-01

    Advanced water-cooled reactor concepts with tight lattices have been proposed worldwide to improve the fuel utilization and the economic competitiveness. In the present work, experimental investigations were performed on thermal-hydraulic behaviour in tight hexagonal 7-rod bundles under both single-phase and two-phase conditions. Freon-12 was used as working fluid due to its convenient operating parameters. Tests were carried out under both single-phase and two-phase flow conditions. Rod surface temperatures are measured at a fixed axial elevation and in various circumferential positions. Test data with different radial power distributions are analyzed. Measured surface temperatures of unheated rods are used for the assessment of and comparison with numerical codes. In addition, numerical simulation using sub-channel analysis code MATRA and the computational fluid dynamics (CFD) code ANSYS-10 is carried out to understand the experimental data and to assess the validity of these codes in the prediction of flow and heat transfer behaviour in tight rod bundle geometries. Numerical results are compared with experimental data. A good agreement between the measured temperatures on the unheated rod surface and the CFD calculation is obtained. Both sub-channel analysis and CFD calculation indicates that the turbulent mixing in the tight rod bundle is significantly stronger than that computed with a well established correlation.

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

    International Nuclear Information System (INIS)

    Zhou, Jianjun; Zhang, Daling; Qiu, Suizheng; Su, Guanghui; Tian, Wenxi; Wu, Yingwei

    2015-01-01

    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

  7. 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.

  8. 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

  9. BEPU-FSAR: establishing a background for extension of nuclear thermal hydraulic principles to non thermal-hydraulic codes

    Energy Technology Data Exchange (ETDEWEB)

    Menzel, Francine; Sabundjian, Gaianê, E-mail: franmenzel@gmail.com, E-mail: gdjian@ipen.br [Instituto de Pesquisas Energéticas e Nucleares (IPEN/CNEN-SP), São Paulo, SP (Brazil); D’Auria, Francesco, E-mail: f.dauria@ing.unipi.it [University of Pisa, San Piero a Grado Nuclear Research Group (Italy)

    2017-07-01

    Nuclear thermal hydraulic and accident analysis are based in three pillar activities, which consists in: Scaling, Coupling and V and V. Each of them are established technology, with key documents to describe and widely used. The final goal of this work is to apply the BEPU methodology in all parts of FSAR where analytical techniques are needed (BEPU-FSAR) and for that the crucial step is the transfer of the BEPU concepts into the other areas. In this sense, the issue is how to adapt to other disciplines the pillar activities presented in the thermal hydraulic area. For that we need to identify which elements can be applied in the other areas, to show that the proposed methodology is feasible. This work aims to discuss the first steps towards a BEPU-FSAR methodology and to show that the Scaling, Coupling and V and V elements, currently done for thermal-hydraulic codes, can be also done for different codes, which are used to perform different analysis included on a FSAR of a generic plant. (author)

  10. 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)

  11. 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.

  12. Transient thermal-hydraulic/neutronic analysis in a VVER-1000 reactor core

    International Nuclear Information System (INIS)

    Seyed khalil Mousavian; Mohammad Mohsen Ertejaei; Majid Shahabfar

    2005-01-01

    Full text of publication follows: Nowadays, coupled thermal-hydraulic and three-dimensional neutronic codes in order to consider different feedback effects is state of the art subject in nuclear engineering researches. In this study, RELAP5/COBRA and WIMS/CITATION codes are implemented to investigate the VVER-1000 reactor core parameters during Large Break Loss of Coolant Accident (LB-LOCA). In a LB-LOCA, the primary side pressure, coolant density and fuel temperature strongly decrease but the cladding temperature experiences a strong peak. For this purpose, the RELAP5 Best Estimate (BE) system code is used to simulate the LB-LOCA analysis in VVER-1000 nuclear thermal-hydraulic loops. Also, the modified COBRA-IIIc software as a sub-channel analysis code is applied for modeling of VVER-1000 reactor core. Moreover, WIMS and CITATION as a cross section and 3-D neutron flux codes are coupled with thermal-hydraulic codes with the aim of consider the spatial effects through the reactor core. For this reason, suitable software is developed to link and speed up the coupled thermalhydraulic and three-dimensional neutronic calculations. This software utilizes of external coupling concept in order to integrate thermal-hydraulic and neutronic calculations. (authors)

  13. Analysis of molten salt thermal-hydraulics using computational fluid dynamics

    International Nuclear Information System (INIS)

    Yamaji, B.; Csom, G.; Aszodi, A.

    2003-01-01

    To give a good solution for the problem of high level radioactive waste partitioning and transmutation is expected to be a pro missing option. Application of this technology also could extend the possibilities of nuclear energy. Large number of liquid-fuelled reactor concepts or accelerator driven subcritical systems was proposed as transmutors. Several of these consider fluoride based molten salts as the liquid fuel and coolant medium. The thermal-hydraulic behaviour of these systems is expected to be fundamentally different than the behaviour of widely used water-cooled reactors with solid fuel. Considering large flow domains three-dimensional thermal-hydraulic analysis is the method seeming to be applicable. Since the fuel is the coolant medium as well, one can expect a strong coupling between neutronics and thermal-hydraulics too. In the present paper the application of Computational Fluid Dynamics for three-dimensional thermal-hydraulics simulations of molten salt reactor concepts is introduced. In our past and recent works several calculations were carried out to investigate the capabilities of Computational Fluid Dynamics through the analysis of different molten salt reactor concepts. Homogenous single region molten salt reactor concept is studied and optimised. Another single region reactor concept is introduced also. This concept has internal heat exchanges in the flow domain and the molten salt is circulated by natural convection. The analysis of the MSRE experiment is also a part of our work since it may form a good background from the validation point of view. In the paper the results of the Computational Fluid Dynamics calculations with these concepts are presented. In the further work our objective is to investigate the thermal-hydraulics of the multi-region molten salt reactor (Authors)

  14. Alternative solution algorithm for coupled thermal-hydraulic problems

    International Nuclear Information System (INIS)

    Farnsworth, D.A.; Rice, J.G.

    1986-01-01

    A thermal-hydraulic system involves flow of a fluid for which a combined solution of the continuity, momentum, and energy equations is required. When the solutions of the energy and momentum fields are dependent on each other, the system is said to be thermally coupled. A common problem encountered in the numerical solution of strongly coupled thermal-hydraulic problems is a very slow rate of convergence or a complete lack of convergence. Many times this degradation in convergence is due to the lack of true coupling between the energy and momentum fields during the iteration process. In the most widely used solution algorithms - such as the SIMPLE algorithm and its many variants - a sequential solution technique is required. That is, the solution process alternates between the flow and energy fields until a converged solution is obtained. This approach allows only implicit energy-momentum coupling. To improve the convergence rate for strongly coupled problems, a practical solution algorithm that can accommodate true energy-momentum coupling terms was developed. A complete simultaneous (versus sequential) solution of the governing conservation equations utilizing a line-by-line solution was developed and direct coupling terms between the momentum and energy fields were added utilizing a modified Newton-Raphson technique

  15. Applications for coupled core neutronics and thermal-hydraulic models

    International Nuclear Information System (INIS)

    Eller, J.

    1996-01-01

    The unprecedented increases in computing capacity that have occurred during the last decade have affected our sciences, and thus our lives, to an extent that is difficult to overstate. All indications are that this trend will continue for years to come. Nuclear reactor systems analysis is one of many areas of engineering that has changed dramatically as a result of this evolution. Our ability to model the various mechanical and physical systems in greater and greater detail has allowed significant improvements in operational efficiency in spite of increasing regulatory requirements. Many of these efficiencies result from the use of more complex and geometrically detailed computer modeling, which is used to justify a reduction or elimination of some of the conservatisms required by earlier, less sophisticated analyses. And more recently, as our industries open-quotes downsize,close quotes efforts are being made to find ways to use the ever-increasing computing capacity to design systems that accomplish more work, in less time, and with fewer people. The balance of this paper discusses some of the visions that Duke Power Company feels would most benefit their particular methodologies. One of the concepts receiving a lot of attention involves an automated coupling of a thermal-hydraulic plant systems analysis model to a three-dimensional core neutronics program. The thermal-hydraulic analysis of several postulated system transients incorporates large conservatisms because of limited ability to model complex time-dependent asymmetric heat sources in adequate geometric detail. For these transients, the core behavior is closely coupled with the thermal-hydraulic behavior of the total plant system and vice versa. Steam-line break, uncontrolled rod withdrawal, and rod drop anayses are likely to benefit most from this type of linked process

  16. Thermal-hydraulic investigations of fuel elements

    International Nuclear Information System (INIS)

    Rehme, K.; Weinberg, D.

    1983-01-01

    Extensive fluid-dynamic examining of flow distribution and turbulent flow distribution was done to control and safeguard calculation methods allowing the determination of three-dimensional flow distribution in fuel elements. Results show that the flow distribution greatly depends on the frequency of pulse exchange between subchannels in narrow rod grids. The comparison of these measured values to VELASCO's results shows that the calculation methods need to be considerably improved. The subchannel analysis proved to be very suitable to calculate mean flow temperatures conforming with the subchannel analysis principle. However, this does not include statements on wall temperatures occurring in the structures. Mean wall temperatures can be determined by empirical interrelationships for Nusseltnumbers. On the other hand, the calculation of detailed wall temperature distributions is not possible with the subchannel analysis unless it can be further improved due to more detailed measurement results. (orig.) [de

  17. Computerized hydraulic and thermal laboratory investigations

    International Nuclear Information System (INIS)

    Lepetit, J.P.; Manoha, B.

    1981-02-01

    The paper presents the principles and capabilities of a computer controlled tank performed by the L.N.H. to assume an automatic reproduction of non steady currents. It gives a description of the tank and its equipment for model control, measurement and treatment of data; it shows an example of results obtained on the site of the Flamanville nuclear plant [fr

  18. Thermal hydraulic behavior evaluation of tank A-101

    International Nuclear Information System (INIS)

    Ogden, D.M.

    1996-01-01

    This report describes a new evaluation conducted to help understand the thermal-hydraulic behavior of tank A-101. Prior analysis of temperature data indicated that the dome space and upper waste layer was slowly increasing in temperature increases are due to increasing ambient temperatures and termination of forced ventilation. However, this analysis also indicates that other dome cooling processes are slowly decreasing, or some slow increase in heating is occurring at the waste surface. Dome temperatures are not decreasing at the rate expected as a forced ventilation termination effects are accounted for

  19. Thermal-hydraulic analysis of PWR cores in transient condition

    International Nuclear Information System (INIS)

    Silva Galetti, M.R. da.

    1984-01-01

    A calculational methodology for thermal - hydraulic analysis of PWR cores under steady-state and transient condition was selected and made available to users. An evaluation of the COBRA-IIIP/MIT code, used for subchannel analysis, was done through comparison of the code results with experimental data on steady state and transient conditions. As a result, a comparison study allowing spatial and temporal localization of critical heat flux was obtained. A sensitivity study of the simulation model to variations in some empirically determined parameter is also presented. Two transient cases from Angra I FSAR were analysed, showing the evolution of minimum DNBR with time. (Author) [pt

  20. Current Development and Trends in Thermal-Hydraulics

    International Nuclear Information System (INIS)

    Toth, I.

    2008-01-01

    A review of CSNI activities during the last two decades in the field of thermal-hydraulics and related topics has been extensively presented in sessions 2. to 9. New activities are in progress or planned partly based on recommendations of the CSNI Operating Plan and the CSNI SESAR SFEAR report, but also on requests coming from the member states. These activities are performed in the frame of the CSNI Working Group on the Analysis and Management of Accidents (GAMA) or in the frame of CSNI Projects. These actions are summarized in this paper.

  1. Thermal-hydraulic considerations for particle bed reactors

    Science.gov (United States)

    Benenati, R.; Araj, K. J.; Horn, F.

    In the design of particle bed reactor (PBR) cores, consideration must be given to the gas coolant channels and their configuration. Neutronics analysis provides the relative volume fractions of the component materials, but these must be arranged in such a manner as to allow proper cooling of all components by the gas flow at relatively low pressure drops. The thermal hydraulic aspects of this problem are addressed. A description of the computer model used in the analysis of the steady state condition is also included. Blowdown tests on hot particle bed fuel elements were carried out and are described.

  2. Thermal hydraulic behavior of SCWR sliding pressure startup

    International Nuclear Information System (INIS)

    Fu Shengwei; Zhou Chong; Xu Zhihong; Yang Yanhua

    2011-01-01

    The modification to ATHLET-SC code is introduced in this paper, which realizes the simulation of trans-critical transients using two-phase model. With the modified code, the thermal-hydraulic dynamic behavior of the mixed SCWR core during the startup process is simulated. The startup process is similar to the design of SCLWR-H sliding pressure startup. The results show that maximum temperature of cladding-surface does not exceed 650℃ in the whole startup process, and the sudden change of water properties in the trans-critical transients will not cause harmful influence to the heat transfer of the fuel cladding. (authors)

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

    Directory of Open Access Journals (Sweden)

    Reza Akbari

    2017-08-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1998-03-01

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

  5. Cross-cutting european thermal-hydraulics research for innovative nuclear systems

    International Nuclear Information System (INIS)

    Roelofs, F.; Class, A.; Cheng, X.; Meloni, P.; Van Tichelen, K.; Boudier, P.; Prasser, M.

    2010-01-01

    Thermal-hydraulics is recognized as a key scientific subject in the development of different innovative nuclear reactor systems. From the thermal-hydraulic point of view, different innovative reactors are mainly characterized by their coolants (gas, water, liquid metals and molten salt). This results in different micro- and macroscopic behavior of flow and heat transfer and requires specific models and advanced analysis tools. However, many common thermal-hydraulic issues are identified among various innovative nuclear systems. In Europe, such cross-cutting thermal-hydraulic issues are the subject of the 7. framework programme THINS (Thermal-Hydraulics of Innovative Nuclear Systems) project which runs from 2010 until 2014. This paper will describe the activities in this project which address the main identified thermal hydraulics issues for innovative nuclear systems. (authors)

  6. Thermal hydraulic evaluation of advanced wire-wrapped assemblies

    International Nuclear Information System (INIS)

    Wei, J.P.

    1975-01-01

    The thermal-hydraulic analyses presented in this report are based on application of the subchannel concept in association with the use of bulk parameters for coolant velocity and coolant temperature within a subchannel. The interactions between subchannels are due to turbulent interchange, pressure-induced diversion crossflow, directed sweeping crossflow induced by the helical wire wrap, and transverse thermal conduction. The FULMIX-II computer program was successfully developed to perform the steady-state temperature predictions for LMFBR fuel assemblies with the reference straight-start design and the advanced wire-wrap designs. Predicted steady-state temperature profiles are presented for a typical CRBRP 217-rod wire-wrapped assembly with the selected wire-wrap designs

  7. Spent nuclear fuel storage pool thermal-hydraulic analysis

    International Nuclear Information System (INIS)

    Gay, R.R.

    1984-01-01

    Storage methods and requirements for spent nuclear fuel at U.S. commercial light water reactors are reviewed in Section 1. Methods of increasing current at-reactor storage capabilities are also outlined. In Section 2 the development of analytical methods for the thermal-hydraulic analysis of spent fuel pools is chronicled, leading up to a discussion of the GFLOW code which is described in Section 3. In Section 4 the verification of GFLOW by comparisons of the code's predictions to experimental data taken inside the fuel storage pool at the Maine Yankee nuclear power plant is presented. The predictions of GFLOW using 72, 224, and 1584 node models of the storage pool are compared to each other and to the experimental data. An example of thermal licensing analysis for Maine Yankee using the GFLOW code is given in Section 5. The GFLOW licensing analysis is compared to previous licensing analysis performed by Yankee Atomic using the RELAP-4 computer code

  8. Thermal-hydraulic modeling needs for passive reactors

    International Nuclear Information System (INIS)

    Kelly, J.M.

    1997-01-01

    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

  9. Thermal hydraulics analysis of LIBRA-SP target chamber

    International Nuclear Information System (INIS)

    Mogahed, E.A.

    1996-01-01

    LIBRA-SP is a conceptual design study of an inertially confined 1000 MWe fusion power reactor utilizing self-pinched light ion beams. There are 24 ion beams which are arranged around the reactor cavity. The reaction chamber is an upright cylinder with an inverted conical roof resembling a mushroom, and a pool floor. The vertical sides of the cylinder are occupied by a blanket zone consisting of many perforated rigid HT-9 ferritic steel tubes called PERITs (PEr-forated RIgid Tube). The breeding/cooling material, liquid lead-lithium, flows through the PERITs, providing protection to the reflector/vacuum chamber so as to make it a lifetime component. The neutronics analysis and cavity hydrodynamics calculations are performed to account for the neutron heating and also to determine the effects of vaporization/condensation processes on the surface heat flux. The steady state nuclear heating distribution at the midplane is used for thermal hydraulics calculations. The maximum surface temperature of the HT-9 is chosen to not exceed 625 degree C to avoid drastic deterioration of the metal's mechanical properties. This choice restricts the thermal hydraulics performance of the reaction cavity. The inlet first surface coolant bulk temperature is 370 degree C, and the heat exchanger inlet coolant bulk temperature is 502 degree C. 4 refs., 6 figs., 2 tabs

  10. Thermal hydraulic and safety analyses for Pakistan Research Reactor-1

    International Nuclear Information System (INIS)

    Bokhari, I.H.; Israr, M.; Pervez, S.

    1999-01-01

    Thermal hydraulic and safety analysis of Pakistan Research Reactor-1 (PARR-1) utilizing low enriched uranium (LEU) fuel have been performed using computer code PARET. The present core comprises of 29 standard and 5 control fuel elements. Results of the thermal hydraulic analysis show that the core can be operated at a steady-state power level of 10 MW for a flow rate of 950 m 3 /h, with sufficient safety margins against ONB (onset of nucleate boiling) and DNB (departure from nucleate boiling). Safety analysis has been carried out for various modes of reactivity insertions. The events studied include: start-up accident; accidental drop of a fuel element in the core; flooding of a beam tube with water; removal of an in-pile experiment during reactor operation etc. For each of these transients, time histories of reactor power, energy released and clad surface temperature etc. were calculated. The results indicate that the peak clad temperatures remain well below the clad melting temperature during these accidents. It is therefore concluded that the reactor can be safely operated at 10 MW without compromising safety. (author)

  11. 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.

  12. LWR containment thermal hydraulic codes benchmark demona B3 exercise

    International Nuclear Information System (INIS)

    Della Loggia, E.; Gauvain, J.

    1988-01-01

    Recent discussion about the aerosol codes currently used for the analysis of containment retention capabilities have revealed a number of questions concerning the reliabilities and verifications of the thermal-hydraulic modules of these codes with respect to the validity of implemented physical models and the stability and effectiveness of numerical schemes. Since these codes are used for the calculation of the Source Term for the assessment of radiological consequences of severe accidents, they are an important part of reactor safety evaluation. For this reason the Commission of European Communities (CEC), following the recommendation mode by experts from Member Stades, is promoting research in this field with the aim also of establishing and increasing collaboration among Research Organisations of member countries. In view of the results of the studies, the CEC has decided to carry out a Benchmark exercise for severe accident containment thermal hydraulics codes. This exercise is based on experiment B3 in the DEMONA programme. The main objective of the benchmark exercise has been to assess the ability of the participating codes to predict atmosphere saturation levels and bulk condensation rates under conditions similar to those predicted to follow a severe accident in a PWR. This exercise follows logically on from the LA-4 exercise, which, is related to an experiment with a simpler internal geometry. We present here the results obtained so far and from them preliminary conclusions are drawn, concerning condensation temperature, pressure, flow rates, in the reactor containment

  13. Finite volume thermal-hydraulics and neutronics coupled calculations - 15300

    International Nuclear Information System (INIS)

    Araujo Silva, V.; Campagnole dos Santos, A.A.; Mesquit, A.Z.; Bernal, A.; Miro, R.; Verdu, G.; Pereira, C.

    2015-01-01

    The computational power available nowadays allows the coupling of neutronics and thermal-hydraulics codes for reactor studies. The present methodology foresees at least one constraint to the separated codes in order to perform coupled calculations: both codes must use the same geometry, however, meshes can be different for each code as long as the internal surfaces stays the same. Using the finite volume technique, a 3D diffusion nodal code was implemented to deal with neutron transport. This code can handle non-structured meshes which allows for complicated geometries calculations and therefore more flexibility. A computational fluid dynamics (CFD) code was used in order to obtain the same level of details for the thermal hydraulics calculations. The chosen code is OpenFOAM, an open-source CFD tool. Changes in OpenFOAM allow simple coupled calculations of a PWR fuel rod with neutron transport code. OpenFOAM sends coolant density information and fuel temperature to the neutron transport code that sends back power information. A mapping function is used to average values when one node in one side corresponds to many nodes in the other side. Data is exchanged between codes by library calls. As the results of a fuel rod calculations progress, more complicated and processing demanding geometries will be simulated, aiming to the simulation of a real scale PWR fuel assembly

  14. Advanced modelling and numerical strategies in nuclear thermal-hydraulics

    International Nuclear Information System (INIS)

    Staedtke, H.

    2001-01-01

    The first part of the lecture gives a brief review of the current status of nuclear thermal hydraulics as it forms the basis of established system codes like TRAC, RELAP5, CATHARE or ATHLET. Specific emphasis is given to the capabilities and limitations of the underlying physical modelling and numerical solution strategies with regard to the description of complex transient two-phase flow and heat transfer conditions as expected to occur in PWR reactors during off-normal and accident conditions. The second part of the lecture focuses on new challenges and future needs in nuclear thermal-hydraulics which might arise with regard to re-licensing of old plants using bestestimate methodologies or the design and safety analysis of Advanced Light Water Reactors relying largely on passive safety systems. In order to meet these new requirements various advanced modelling and numerical techniques will be discussed including extended wellposed (hyperbolic) two-fluid models, explicit modelling of interfacial area transport or higher order numerical schemes allowing a high resolution of local multi-dimensional flow processes.(author)

  15. Scaling of Thermal-Hydraulic Phenomena and System Code Assessment

    International Nuclear Information System (INIS)

    Wolfert, K.

    2008-01-01

    In the last five decades large efforts have been undertaken to provide reliable thermal-hydraulic system codes for the analyses of transients and accidents in nuclear power plants. Many separate effects tests and integral system tests were carried out to establish a data base for code development and code validation. In this context the question has to be answered, to what extent the results of down-scaled test facilities represent the thermal-hydraulic behaviour expected in a full-scale nuclear reactor under accidental conditions. Scaling principles, developed by many scientists and engineers, present a scientific technical basis and give a valuable orientation for the design of test facilities. However, it is impossible for a down-scaled facility to reproduce all physical phenomena in the correct temporal sequence and in the kind and strength of their occurrence. The designer needs to optimize a down-scaled facility for the processes of primary interest. This leads compulsorily to scaling distortions of other processes with less importance. Taking into account these weak points, a goal oriented code validation strategy is required, based on the analyses of separate effects tests and integral system tests as well as transients occurred in full-scale nuclear reactors. The CSNI validation matrices are an excellent basis for the fulfilling of this task. Separate effects tests in full scale play here an important role.

  16. 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.

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

    International Nuclear Information System (INIS)

    Laffont, A.; Pentori, B.

    2003-01-01

    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. 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)

  19. PLUGM: a coupled thermal-hydraulic computer model for freezing melt flow in a channel

    International Nuclear Information System (INIS)

    Pilch, M.

    1982-01-01

    PLUGM is a coupled thermal-hydraulic computer model for freezing liquid flow and plugging in a cold channel. PLUGM is being developed at Sandia National Laboratories for applications in Sandia's ex-vessel Core Retention Concept Assessment Program and in Sandia's LMFBR Transition Phase Program. The purpose of this paper is to introduce PLUGM and demonstrate how it can be used in the analysis of two of the core retention concepts under investigation at Sandia: refractory brick crucibles and particle beds

  20. Thermal-hydraulic codes validation for safety analysis of NPPs with RBMK

    International Nuclear Information System (INIS)

    Brus, N.A.; Ioussoupov, O.E.

    2000-01-01

    This work is devoted to validation of western thermal-hydraulic codes (RELAP5/MOD3 .2 and ATHLET 1.1 Cycle C) in application to Russian designed light water reactors. Such validation is needed due to features of RBMK reactor design and thermal-hydraulics in comparison with PWR and BWR reactors, for which these codes were developed and validated. These validation studies are concluded with a comparison of calculation results of modeling with the thermal-hydraulics codes with the experiments performed earlier using the thermal-hydraulics test facilities with the experimental data. (authors)

  1. Proceedings of the 8. Brazilian Meeting on Reactor Physics and Thermal Hydraulics

    International Nuclear Information System (INIS)

    1991-01-01

    Some papers about pressurized light water reactors, fast reactors, accident analysis, transients, research reactors, nuclear data collection, thermal hydraulics, reactor monitoring, neutronics are presented. (E.G.)

  2. Regulatory requirements to the thermal-hydraulic and thermal-mechanical computer codes

    International Nuclear Information System (INIS)

    Vitkova, M.; Kalchev, B.; Stefanova, S.

    2006-01-01

    The paper presents an overview of the regulatory requirements to the thermal-hydraulic and thermal-mechanical computer codes, which are used for safety assessment of the fuel design and the fuel utilization. Some requirements to the model development, verification and validation of the codes and analysis of code uncertainties are also define. Questions concerning Quality Assurance during development and implementation of the codes as well as preparation of a detailed verification and validation plan are briefly discussed

  3. Transitioning from interpretive to predictive in thermal hydraulic codes

    International Nuclear Information System (INIS)

    Mousseau, V.A.

    2004-01-01

    The current thermal hydraulic codes in use in the US, RELAP and TRAC, where originally written in the mid to late 1970's. At that time computers were slow, expensive, and had small memories. Because of these constraints, sacrifices had to be made, both in physics and numerical methods, which resulted in limitations on the accuracy of the solutions. Significant changes have occurred that induce very different requirements for the thermal hydraulic codes to be used for the future GEN-IV nuclear reactors. First, computers speed and memory grow at an exponential rate while the costs hold constant or decrease. Second, passive safety systems in modern designs stretch the length of relevant transients to many days. Finally, costs of experiments have grown very rapidly. Because of these new constraints, modern thermal hydraulic codes will be relied on for a significantly larger portion of bringing a nuclear reactor on line. Simulation codes will have to define in which part of state space experiments will be run. They will then have to be able to extend the small number of experiments to cover the large state space in which the reactors will operate. This data extrapolation mode will be referred to as 'predictive'. One of the keys to analyzing the accuracy of a simulation is to consider the entire domain being simulated. For example, in a reactor design where the containment is coupled to the reactor cooling system through radiative heat transfer, the accuracy of a transient includes the containment, the radiation heat transfer, the fluid flow in the cooling system, the thermal conduction in the solid, and the neutron transport in the reactor. All of this physics is coupled together in one nonlinear system through material properties, cross sections, heat transfer coefficients, and other mechanisms that exchange mass, momentum, and energy. Traditionally, these different physical domains, (containment, cooling system, nuclear fuel, etc.) have been solved in different

  4. Engineered Barrier System Thermal-Hydraulic-Chemical Column Test Report

    International Nuclear Information System (INIS)

    W.E. Lowry

    2001-01-01

    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 and 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. A new approach to designing reduced scale thermal-hydraulic experiments

    International Nuclear Information System (INIS)

    Lapa, Celso M.F.; Sampaio, Paulo A.B. de; Pereira, Claudio M.N.A.

    2004-01-01

    Reduced scale experiments are often employed in engineering because they are much cheaper than real scale testing. Unfortunately, though, it is difficult to design a thermal-hydraulic circuit or equipment in reduced scale capable of reproducing, both accurately and simultaneously, all the physical phenomena that occur in real scale and operating conditions. This paper presents a methodology to designing thermal-hydraulic experiments in reduced scale based on setting up a constrained optimization problem that is solved using genetic algorithms (GAs). In order to demonstrate the application of the methodology proposed, we performed some investigations in the design of a heater aimed to simulate the transport of heat and momentum in the core of a pressurized water reactor (PWR) at 100% of nominal power and non-accident operating conditions. The results obtained show that the proposed methodology is a promising approach for designing reduced scale experiments

  6. 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.

  7. Thermal hydraulics of sodium-cooled fast reactors - key issues and highlights

    International Nuclear Information System (INIS)

    Ninokata, H.; Kamide, H.

    2011-01-01

    In this paper key issues and highlighted topics in thermal hydraulics are discussed in connection to the current Japan's sodium-cooled fast reactor development efforts. In particular, design study and related researches of the Japan Sodium-cooled Fast Reactor (JSFR) are focused. Several innovative technologies, e.g., compact reactor vessel, two-loop system, fully natural circulation decay heat removal, and recriticality free core, have been investigated in order to reduce construction cost and to achieve higher level of reactor safety. Preliminary evaluations of innovative technologies to be applied to JSFR are on-going. Here, progress of design study is introduced. Then, research and development activities on the thermal hydraulics related to the innovative technologies are briefly reviewed. (author)

  8. Neutronics and thermal hydraulics coupling scheme for design improvement of liquid metal fast systems

    International Nuclear Information System (INIS)

    Sanchez-Espinoza, V.H.; Jaeger, W.; Travleev, A.; Monti, L.; Doern, R.

    2009-01-01

    Many advanced reactor concepts are nowadays under investigations within the Generation IV international initiative as well as in European research programs including subcritical and critical fast reactor systems cooled by liquid metal, gas and supercritical water. The Institute of Neutron Physics and Reactor Technology (INR) at the Forschungszentrum Karlsruhe GmbH is involved in different European projects like IP EUROTRANS, ELSY, ESFR. The main goal of these projects is, among others, to assess the technical feasibility of proposed concepts regarding safety, economics and transmutation requirements. In view of increased computer capabilities, improved computational schemes, where the neutronic and the thermal hydraulic solution is iteratively coupled, become practicable. The codes ERANOS2.1 and TRACE are being coupled to analyze fuel assembly or core designs of lead-cooled fast reactors (LFR). The neutronic solution obtained with the coupled system for a LFR fuel assembly was compared with the MCNP5 solution. It was shown that the coupled system is predicting physically sound results. The iterative coupling scheme was realized using Perlscripts and auxiliary Fortran programs to ensure that the mapping between the neutronic and the thermal hydraulic part is consistent. The coupled scheme is very flexible and appropriate for the neutron physical and thermal hydraulic investigation of fuel assemblies and of cores of lead cooled fast reactors. The developed methods and the obtained results will be presented and discussed. (author)

  9. Evaluation on thermal-hydraulic characteristics for passive safety device of APR1400

    Energy Technology Data Exchange (ETDEWEB)

    Yoo, Seong Yeon; Lee, S. H.; Son, M. K. [Korea Association for Nuclear Technology, Taejon (Korea, Republic of); Jee, M. S.; Chung, M. H. [Chungnam National Univ., Taejon (Korea, Republic of)

    2001-07-15

    To establish evaluation and verification guideline for the APR1400, thermal-hydraulic characteristics for fuel rod bundle, reactor vessel and fluidic device is analyzed using FLUENT. Scope and major results of research are as follows : Thermal-hydraulic characteristics for nuclear fuel rod bundle: design data for nuclear fuel rod bundle and structure are surveyed, and 3 x 3 sub-channel model is adopted to investigate the fluid flow and heat transfer characteristics in fuel rod bundle. Computational results are compared with the heat transfer data measured by naphthalene sublimation method, and numerical analysis and evaluation are performed at various design conditions and flow conditions. Thermal-hydraulic characteristics for reactor vessel: reactor vessel design data are surveyed to develop numerical model. Porous media model is applied for fuel rod bundle, and full-scale, three dimensional simulation is performed at actual operating conditions. Distributions of velocity, pressure and temperature are discussed. Flow characteristics for fluidic device: three dimensional numerical model for fluidic device is developed, and numerical results are compared with experimental data obtained at KAERI in order to verify numerical simulation. In addition, variation of flow rate is investigated at various elapsed times after valve operating, and flow characteristics is analyzed at low and high flow rate conditions, respectively.

  10. Validation of containment thermal hydraulic computer codes for VVER reactor

    Energy Technology Data Exchange (ETDEWEB)

    Jiri Macek; Lubomir Denk [Nuclear Research Institute Rez plc Thermal-Hydraulic Analyses Department CZ 250 68 Husinec-Rez (Czech Republic)

    2005-07-01

    Full text of publication follows: The Czech Republic operates 4 VVER-440 units, two VVER-1000 units are being finalized (one of them is undergoing commissioning). Thermal-hydraulics Department of the Nuclear Research Institute Rez performs accident analyses for these plants using a number of computer codes. To model the primary and secondary circuits behaviour the system codes ATHLET, CATHARE, RELAP, TRAC are applied. Containment and pressure-suppression system are modelled with COCOSYS and MELCOR codes, the reactor power calculations (point and space-neutron kinetics) are made with DYN3D, NESTLE and CDF codes (FLUENT, TRIO) are used for some specific problems.An integral part of the current Czech project 'New Energy Sources' is selection of a new nuclear source. Within this and the preceding projects financed by the Czech Ministry of Industry and Trade and the EU PHARE, the Department carries and has carried out the systematic validation of thermal-hydraulic and reactor physics computer codes applying data obtained on several experimental facilities as well as the real operational data. One of the important components of the VVER 440/213 NPP is its containment with pressure suppression system (bubble condenser). For safety analyses of this system, computer codes of the type MELCOR and COCOSYS are used in the Czech Republic. These codes were developed for containments of classic PWRs or BWRs. In order to apply these codes for VVER 440 systems, their validation on experimental facilities must be performed.The paper provides concise information on these activities of the NRI and its Thermal-hydraulics Department. The containment system of the VVER 440/213, its functions and approaches to solution of its safety is described with definition of acceptance criteria. A detailed example of the containment code validation on EREC Test facility (LOCA and MSLB) and the consequent utilisation of the results for a real NPP purposes is included. An approach to

  11. Study of thermal-hydraulic characteristics in an LMFBR intermediate plenum

    International Nuclear Information System (INIS)

    Uotani, M.; Naohara, N.; Kinoshita, I.

    1985-01-01

    Experimental studies using water and liquid metal were conducted in order to investigate the thermal-hydraulic characteristics of an LMFBR intermediate plenum. The present study is an attempt to evaluate the effect of natural convection on the temperature field and to validate the prediction method of temperature profile in a thermally stratified cavity. The experimental results indicated that the effect of the natural convection on flow velocity and heat transfer in the cavity is reduced with increasing the modified stratification parameter. The calculation by FEM code and a simple 1-D model are effective to predict the temperature profile in the cavity

  12. Thermal hydraulic feasibility analysis of the IBED PHTS for ITER

    Energy Technology Data Exchange (ETDEWEB)

    Carloni, Dario, E-mail: dacarloni@gmail.com [Dipartimento di Ingegneria Meccanica, Nucleare e della Produzione, Pisa University, Via Diotisalvi 2, 56126 Pisa (Italy); Dell’Orco, Giovanni; Babulal, Gopalapillai; Somboli, Fabio; Serio, Luigi [ITER Organisation, Route de Vinon sur Verdon, 13115 Saint Paul Lez Durance (France); Paci, Sandro [Dipartimento di Ingegneria Meccanica, Nucleare e della Produzione, Pisa University, Via Diotisalvi 2, 56126 Pisa (Italy)

    2013-10-15

    One of the main challenges of the ITER fusion reactor is to effectively remove large amount of heat deposited to the surface of the plasma facing components. The tokamak cooling water system (TCWS) will accomplish the objective of removing about 1 GW of peak heat load from in-vessel components while maintaining pressures and temperatures of the coolant within acceptable and safe limits during different operational scenarios. A study of feasibility has been launched for the IBED PHTS (Integrated Blanket, Edge localized mode coils (ELMs) and Divertor Primary Heat Transfer System; it consists of five independent cooling trains (four operational and one in stand-by), one steam pressurizer, supply and return headers, ring manifolds and connections to the all in-vessel components (i.e. First Wall Blanket, Divertor, ELM, Diagnostics and other Ports clients). The dynamic behaviour of the IBED PHTS has been investigated by means of RELAP5{sup ®} code to simulate the response of the system during plasma pulse and baking operations. Due to the plasma heat deposition on the surfaces of the in-vessel components and subsequent increase in hot leg temperature, a large amount of water volume is transferred from the hot legs of the circuit to the surge-line of the pressurizer during each burn cycle. This causes rapid increase of pressure and temperature of the system and the following actions are proposed to counteract these variations: spray injection in the upper dome of the pressurizer from the Chemical and Volume Control System (CVCS) to reduce the pressure and active control of flow rates through heat exchangers and their bypass loops to regulate the heat transfer from the primary system to the environment via secondary and tertiary loops. This paper focuses on the prediction of the thermal hydraulic behaviour of the IBED PHTS during plasma pulses and baking scenarios, describing the various activity of the analysis, the geometrical assessment of the circuit and the modelling

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

    International Nuclear Information System (INIS)

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

    2005-04-01

    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

  14. 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.

  15. Thermal-hydraulic experiments and analyses on cold moderator

    International Nuclear Information System (INIS)

    Aso, Tomokazu; Kaminaga, Masanori; Terada, Atsushi; Hino, Ryutaro

    2001-01-01

    A cold moderator using supercritical hydrogen is one of the key components in a MW-scale spallation target system, which directly affects the neutronic performance both in intensity and resolution. Since a hydrogen temperature rise in the moderator vessel affects the neutronic performance, it is necessary to suppress the local temperature rise within 3 K. In order to develop the conceptual design of the moderator structure in progress, the flow patterns were measured using a PIV (Particle Image Velocimeter) system under water flow conditions using a flat model that simulated a moderator vessel. From these results, the flow patterns (such as recirculation flows, stagnant flows etc.) were clarified. The hydraulic analytical results obtained using the STAR-CD code agreed well with experimental results. Thermal-hydraulic analyses in the moderator vessel were carried out using the STAR-CD code. Based on these results, we clarified the possibility of suppressing the local temperature rise to within 3 K under 2 MW operating conditions. In order to achieve the cost decreasing of the hydrogen loop, it is necessary to operate it reducing the hydrogen flow rate and the whole hydrogen mass. Then improved moderator concept using blowholes and a twisted tape was proposed, and we have tried to examine the effect of the blowing flow from the inlet pipe. From the experimental and analytical results, the blowing flow could be feasible for the suppression of the stagnant region. (author)

  16. Development of thermal hydraulic analysis code for IHX of FBR

    International Nuclear Information System (INIS)

    Kumagai, Hiromichi; Naohara, Nobuyuki

    1991-01-01

    In order to obtain flow resistance correlations for thermal-hydrauric analysis code concerned with an intermediate heat exchanger (IHX) of FBR, the hydraulic experiment by air was carried out through a bundle of tubes arranged in an in-line and staggard fashion. The main results are summarized as follows. (1) On pressure loss per unit length of a tube bundle, which is densely a regular triangle arrangement, the in-line fashion is almost the same as the staggard one. (2) In case of 30deg sector model for IHX tube bundle, pressure loss is 1/3 in comparison with the in-line or staggard arrangement. (3) By this experimental data, flow resistance correlations for thermalhydrauric analysis code are obtained. (author)

  17. Evolution of thermal-hydraulics testing in EBR-II

    International Nuclear Information System (INIS)

    Golden, G.H.; Planchon, H.P.; Sackett, J.I.; Singer, R.M.

    1987-01-01

    A thermal-hydraulics testing and modeling program has been underway at the Experimental Breeder Reactor-II (EBR-II) for 12 years. This work culminated in two tests of historical importance to commercial nuclear power, a loss of flow without scram and a loss of heat sink wihout scram, both from 100% initial power. These tests showed that natural processes will shut EBR-II down and maintain cooling without automatic control rod action or operator intervention. Supporting analyses indicate that these results are characteristic of a range of sizes of liquid metal cooled reactors (LMRs), if these reactors use metal driver fuel. This type of fuel is being developed as part of the Integral Fast Reactor Program at Argonne National Laboratory. Work is now underway at EBR-II to exploit the inherent safety of metal-fueled LMRs with regard to development of improved plant control strategies. (orig.)

  18. 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...

  19. Thermal hydraulic stability in a pressure tube nuclear reactor

    International Nuclear Information System (INIS)

    Villani, A.; Ravetta, R.; Mansani, L.

    1986-01-01

    The CIRENE plant which will undergo preoperational tests in the near future is equipped with a 40 MW(e) Heavy Water moderated Boiling Light Water cooled Reactor (HWBLWR); at the start-up and up to about 30 % of nominal power, the necessary low coolant density is obtained injecting into the core a mixture of liquid and steam. To verify the thermal-hydraulic stability of the plant in this situation, tests have been carried out in a facility simulating two full scale power channels; the system stability has been confirmed in the reference conditions, and is not reduced by even a significant reduction of the liquid flowrate, where a decrease in liquid temperature has some negative effect and steam flowrate has a small influence. (author)

  20. GNPS 18-months fuel cycles core thermal hydraulic design

    International Nuclear Information System (INIS)

    Liu Changwen; Zhou Zhou

    2002-01-01

    GNPS begins to implement the 18-month fuel cycles from the initial annual reload at cycle 9, thus the initial core thermal hydraulic design is not valid any more. The new critical heat flux (CHF) correlation, FC, which is developed by Framatome, is used in the design, and the generalized statistical methodology (GSM) instead of the initial deterministic methodology is used to determine the DNBR design limit. As the AFA 2G and AFA 3G are mixed loaded in the transition cycle, it will result that the minimum DNBR in the mixed core is less than that of AFA 3G homogenous core, the envelop mixed core DNBR penalty is given. Consequently the core physical limit for mixed core and equilibrium cycles, and the new over temperature ΔT overpower ΔT are determined

  1. Report on the thermal-hydraulics computational component

    International Nuclear Information System (INIS)

    Laughton, T.; Jones, B.G.

    1996-01-01

    The nodal methods computer code utilizing hexagonal geometry, which is being developed as part of this DOE contract, is called THMZ. The computational objective of the code is to calculate the steady-state thermal-hydraulic conditions in a hexagonal geometry reactor core given the appropriate initial conditions and the axial neutron flux profile. The latter is given by a companion nodal neutronics code which was developed in an earlier part of the contact. The joining of these two codes to provide a coupled analysis tool for hexagonal lattice cores is the ultimate objective of the contract and its follow-on work. The remaining part of this report presents the current status of the development and the results which have been obtained to date. These will appear in the MS thesis of Mr. Terrill Laughton in the Department of Nuclear Engineering which is currently in preparation

  2. Thermal hydraulic analysis of the encapsulated nuclear heat source

    Energy Technology Data Exchange (ETDEWEB)

    Sienicki, J.J.; Wade, D.C. [Argonne National Lab., IL (United States)

    2001-07-01

    An analysis has been carried out of the steady state thermal hydraulic performance of the Encapsulated Nuclear Heat Source (ENHS) 125 MWt, heavy liquid metal coolant (HLMC) reactor concept at nominal operating power and shutdown decay heat levels. The analysis includes the development and application of correlation-type analytical solutions based upon first principles modeling of the ENHS concept that encompass both pure as well as gas injection augmented natural circulation conditions, and primary-to-intermediate coolant heat transfer. The results indicate that natural circulation of the primary coolant is effective in removing heat from the core and transferring it to the intermediate coolant without the attainment of excessive coolant temperatures. (authors)

  3. Validation of the TEXSAN thermal-hydraulic analysis program

    International Nuclear Information System (INIS)

    Burns, S.P.; Klein, D.E.

    1992-01-01

    The TEXSAN thermal-hydraulic analysis program has been developed by the University of Texas at Austin (UT) to simulate buoyancy driven fluid flow and heat transfer in spent fuel and high level nuclear waste (HLW) shipping applications. As part of the TEXSAN software quality assurance program, the software has been subjected to a series of test cases intended to validate its capabilities. The validation tests include many physical phenomena which arise in spent fuel and HLW shipping applications. This paper describes some of the principal results of the TEXSAN validation tests and compares them to solutions available in the open literature. The TEXSAN validation effort has shown that the TEXSAN program is stable and consistent under a range of operating conditions and provides accuracy comparable with other heat transfer programs and evaluation techniques. The modeling capabilities and the interactive user interface employed by the TEXSAN program should make it a useful tool in HLW transportation analysis

  4. Scaling in nuclear reactor system thermal-hydraulics

    International Nuclear Information System (INIS)

    D'Auria, F.; Galassi, G.M.

    2010-01-01

    Scaling is a reference 'key-word' in engineering and in physics. The relevance of scaling in the water cooled nuclear reactor technology constitutes the motivation for the present paper. The origin of the scaling-issue, i.e. the impossibility to get access to measured data in case of accident in nuclear reactors, is discussed at first. The so-called 'scaling-controversy' constitutes an outcome. Then, a critical survey (or 'scaling state-of-art';) is given of the attempts and of the approaches to provide a solution to the scaling-issue in the area of Nuclear Reactor System Thermal-Hydraulics (NRSTH): dimensionless design factors for Integral Test Facilities (ITF) are distinguished from scaling factors. The last part of the paper has a two-fold nature: (a) classifying the information about achievements in the area of thermal-hydraulics which are relevant to scaling: the concepts of 'scaling-pyramid' and the related 'scaling bridges' are introduced; (b) establishing a logical path across the scaling achievements (represented as a 'scaling puzzle'). In this context, the 'roadmap for scaling' is proposed: the objective is addressing the scaling issue when demonstrating the applicability of system codes in the licensing process of nuclear power plants. The code itself is referred hereafter as the 'key-to-scaling'. The database from the operation of properly scaled ITF and the availability of qualified system codes are identified as main achievements in NRSTH connected with scaling. The 'roadmap to scaling' constitutes a unified approach to scaling which aims at solving the 'scaling puzzle' created by researches performed during a half-a-century period.

  5. Scaling in nuclear reactor system thermal-hydraulics

    Energy Technology Data Exchange (ETDEWEB)

    D' Auria, F., E-mail: dauria@ing.unipi.i [Dipartimento di Ingegneria Meccanica, Nucleare e della Produzione, University of Pisa, Via Diotisalvi 2, 56126 Pisa (Italy); Galassi, G.M. [Dipartimento di Ingegneria Meccanica, Nucleare e della Produzione, University of Pisa, Via Diotisalvi 2, 56126 Pisa (Italy)

    2010-10-15

    Scaling is a reference 'key-word' in engineering and in physics. The relevance of scaling in the water cooled nuclear reactor technology constitutes the motivation for the present paper. The origin of the scaling-issue, i.e. the impossibility to get access to measured data in case of accident in nuclear reactors, is discussed at first. The so-called 'scaling-controversy' constitutes an outcome. Then, a critical survey (or 'scaling state-of-art';) is given of the attempts and of the approaches to provide a solution to the scaling-issue in the area of Nuclear Reactor System Thermal-Hydraulics (NRSTH): dimensionless design factors for Integral Test Facilities (ITF) are distinguished from scaling factors. The last part of the paper has a two-fold nature: (a) classifying the information about achievements in the area of thermal-hydraulics which are relevant to scaling: the concepts of 'scaling-pyramid' and the related 'scaling bridges' are introduced; (b) establishing a logical path across the scaling achievements (represented as a 'scaling puzzle'). In this context, the 'roadmap for scaling' is proposed: the objective is addressing the scaling issue when demonstrating the applicability of system codes in the licensing process of nuclear power plants. The code itself is referred hereafter as the 'key-to-scaling'. The database from the operation of properly scaled ITF and the availability of qualified system codes are identified as main achievements in NRSTH connected with scaling. The 'roadmap to scaling' constitutes a unified approach to scaling which aims at solving the 'scaling puzzle' created by researches performed during a half-a-century period.

  6. COOLOD, Steady-State Thermal Hydraulics of Research Reactors

    International Nuclear Information System (INIS)

    Kaminaga, Masanori

    1997-01-01

    1 - Description of program or function: The COOLOD-N2 code provides a capability for the analyses of the steady-state thermal-hydraulics of research reactors. This code is a revised version of the COOLOD-N code, and is applicable not only for research reactors in which plate-type fuel is adopted, but also for research reactors in which rod-type fuel is adopted. In the code, subroutines to calculate temperature distribution in rod-type fuel have been newly added to the COOLOD-N code. The COOLOD-N2 code can calculate fuel temperatures under both forced convection cooling mode and natural convection cooling mode. A 'Heat Transfer package' is used for calculating heat transfer coefficient, DNB heat flux etc. The 'Heat Transfer package' is a subroutine program and is especially developed for research reactors in which plate-type fuel is adopted. In case of rod-type fuel, DNB heat flux is calculated by both the 'Heat Transfer package' and Lund DNB heat flux correlation which is popular for TRIGA reactor. The COOLOD-N2 code also has a capability of calculating ONB temperature, the heat flux at onset of flow instability as well as DNB heat flux. 2 - Method of solution: The 'Heat Transfer Package' is a subprogram for calculating heat transfer coefficients, ONB temperature, heat flux at onset of flow instability and DNB heat flux. The 'Heat transfer package' was especially developed for research reactors which are operated under low pressure and low temperature conditions using plate-type fuel, just like the JRR-3M. Heat transfer correlations adopted in the 'Heat Transfer Package' were obtained or estimated based on the heat transfer experiments in which thermal-hydraulic features of the upgraded JRR-3 core were properly reflected. The 'Heat Transfer Package' is applicable to upward and downward flow

  7. ATLAS program for advanced thermal-hydraulic safety research

    International Nuclear Information System (INIS)

    Song, Chul-Hwa; Choi, Ki-Yong; Kang, Kyoung-Ho

    2015-01-01

    Highlights: • Major achievements of the ATLAS program are highlighted in conjunction with both developing advanced light water reactor technologies and enhancing the nuclear safety. • The ATLAS data was shown to be useful for the development and licensing of new reactors and safety analysis codes, and also for nuclear safety enhancement through domestic and international cooperative programs. • A future plan for the ATLAS testing is introduced, covering recently emerging safety issues and some generic thermal-hydraulic concerns. - Abstract: This paper highlights the major achievements of the ATLAS program, which is an integral effect test program for both developing advanced light water reactor technologies and contributing to enhancing nuclear safety. The ATLAS program is closely related with the development of the APR1400 and APR"+ reactors, and the SPACE code, which is a best-estimate system-scale code for a safety analysis of nuclear reactors. The multiple roles of ATLAS testing are emphasized in very close conjunction with the development, licensing, and commercial deployment of these reactors and their safety analysis codes. The role of ATLAS for nuclear safety enhancement is also introduced by taking some examples of its contributions to voluntarily lead to multi-body cooperative programs such as domestic and international standard problems. Finally, a future plan for the utilization of ATLAS testing is introduced, which aims at tackling recently emerging safety issues such as a prolonged station blackout accident and medium-size break LOCA, and some generic thermal-hydraulic concerns as to how to figure out multi-dimensional phenomena and the scaling issue.

  8. Study on the thermal-hydraulic stability of high burn up STEP III fuel in Japan

    International Nuclear Information System (INIS)

    Ishikawa, M.; Kitamura, H.; Toba, A.; Omoto, A.

    2004-01-01

    Japanese BWR utilities have performed a joint study of the Thermal Hydraulic Stability of High Burn up STEP III Fuel. In this study, the parametric dependency of thermal hydraulic stability threshold was obtained. It was confirmed through experiments that the STEP III Fuel has sufficient stability characteristics. (author)

  9. Preliminary thermal/hydraulic sizing calculations for duplex tube evaporator/superheater (interchangeable units). Revision 1

    International Nuclear Information System (INIS)

    Waszink, R.P.; Hwang, J.Y.; Efferding, L.E.

    1974-06-01

    This is a preliminry thermal/hydraulic report reflecting work under Subtask 6.2 of Ref. 1.1. This report is an extension of the previous thermal/hydraulic design report. Parts of this report have been transmitted to GE. The detailed design basis, listed by source, is given. Additional details are discussed

  10. Thermal-Hydraulic Performance of a Corrugated Cooling Fin with Louvered Surfaces

    DEFF Research Database (Denmark)

    Sønderby, Simon Kaltoft; Hosseini, Seyed Mojtaba Mir; Rezaniakolaei, Alireza

    2017-01-01

    The main objective of the article is to investigate thermal-hydraulic performance of a corrugated cooling fin with louvered surfaces. The investigation is carried out using the fin geometry of one most commonly used liquid-to-air heat exchangers. The investigation was carried out by numerically...... simulating the airflow with louvered fin geometry. The simulation model was verified by comparing simulated j- and f-factors with the corresponding values of several experimental correlations. The j-factors deviated less than 10.7 % from two of the experimental correlations, whereas deviations ranging...

  11. 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)

    1999-12-31

    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)

  12. Effect of the inlet throttling on the thermal-hydraulic instability of the natural circulation BWR

    International Nuclear Information System (INIS)

    Furuya, Masahiro; Inada, Fumio; Yoneda, Kimitoshi

    1997-01-01

    Although it is well-established that inlet restriction has a stabilizing for forced circulation BWR, the effect of inlet on the thermal-hydraulic stability of natural circulation BWR remains unknown since increasing inlet restriction affect thermal-hydraulic stability due to reduction of the recirculation flow rate. Therefore experiments have been conducted to investigate the effect of inlet restriction on the thermal-hydraulic stability. A test facility used in this experiments was designed and constructed to have non-dimensional values which are nearly equal to those of natural circulation BWR. Experimental results showed that driving force of the natural circulation was described as a function of heat flux and inlet subcooling independent of inlet restriction. Stability maps in reference to the channel inlet subcooling, heat flux were presented for various inlet restriction which were carried out by an analysis based on the homogeneous flow various using this function. Instability region during the inlet subcooling shifted to the higher inlet subcooling with increasing inlet restriction and became larger with increasing heat flux. (author)

  13. 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)

  14. Thermal-hydraulic and neutronic considerations for designing a lithium-cooled tokamak blanket

    International Nuclear Information System (INIS)

    Chao, J.; Mikic, B.; Todreas, N.

    1978-12-01

    A methodology for the design of lithium cooled blankets is developed. The thermal-hydraulics, neutronics and interactions between them are extensively investigated. In thermal hydraulics, two models illustrate the methodology used to obtain the acceptable ranges for a set of design parameters. The methodology can be used to identify the limiting constraints for a particular design. A complete neutronic scheme is set up for the calculations of the volumetric heating rate as a function of the distance from the first wall, the breeding ratio as a function of the amount of structural material in the blanket, and the radiation damage in terms of atom displacements and gas production rate. Different values of the volume percent of Type-316 stainless steel are assigned in four breeding zones to represent a nonuniformly distributed structural material which satisfies various thermal-hydraulic requirements. The role that the radiation damage plays in the overall design methodology is described. The product of the first wall lifetime and neutron loading is limited by the radiation damage which degrades the mechanical properties of the material

  15. Measurement and flow visualization research of thermal hydraulic characteristics for the SFR reactor Vessel

    International Nuclear Information System (INIS)

    Cha, J. E.; Kim, S. O.; Choi, H. L.; Kim, H. B.; Kim, H. W.; Lee, S. H.

    2012-01-01

    In this report, the thermal hydraulic and flow visualization experiment was described for the KALIMER-600 water-scaled model. In order to investigate a thermal hydraulic characteristics for the SFR KALIMER-600, which has been conceptually designed in the KAERI, a water-scaled 1/10 reactor vessel model was designed and prepared through the scaling analysis during three-years research. In this research, SFR Photos system, which has inherently very complicated the internal structures, was fabricated with a transparent vessel. It was shown that a serious of thermal hydraulic test was conducted within a short period if modeled with water than sodium. Natural circulation test was successfully performed with the modeled heater assembly and heat exchanger system coupled with cooling system. The water-scaled RSV experimental facility made in this research could be used to study the USA development for the future SFR system and utilized to analyze the flow characteristics before changing a main internal part of Photos system. It could also be used to test a pool-inspection study and a sensor selection study before large scale sodium experiment. The PCV system prepared in this research could be utilized to test other TSH experiment and temperature field measurement

  16. 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)

  17. Development of realistic thermal hydraulic system analysis code

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Won Jae; Chung, B. D; Kim, K. D. [and others

    2002-05-01

    The realistic safety analysis system is essential for nuclear safety research, advanced reactor development, safety analysis in nuclear industry and 'in-house' plant design capability development. In this project, we have developed a best-estimate multi-dimensional thermal-hydraulic system code, MARS, which is based on the integrated version of the RELAP5 and COBRA-TF codes. To improve the realistic analysis capability, we have improved the models for multi-dimensional two-phase flow phenomena and for advanced two-phase flow modeling. In addition, the GUI (Graphic User Interface) feature were developed to enhance the user's convenience. To develop the coupled analysis capability, the MARS code were linked with the three-dimensional reactor kinetics code (MASTER), the core thermal analysis code (COBRA-III/CP), and the best-estimate containment analysis code (CONTEMPT), resulting in MARS/MASTER/COBRA/CONTEMPT. Currently, the MARS code system has been distributed to 18 domestic organizations, including research, industrial, regulatory organizations and universities. The MARS has been being widely used for the safety research of existing PWRs, advanced PWR, CANDU and research reactor, the pre-test analysis of TH experiments, and others.

  18. Development of realistic thermal hydraulic system analysis code

    International Nuclear Information System (INIS)

    Lee, Won Jae; Chung, B. D; Kim, K. D.

    2002-05-01

    The realistic safety analysis system is essential for nuclear safety research, advanced reactor development, safety analysis in nuclear industry and 'in-house' plant design capability development. In this project, we have developed a best-estimate multi-dimensional thermal-hydraulic system code, MARS, which is based on the integrated version of the RELAP5 and COBRA-TF codes. To improve the realistic analysis capability, we have improved the models for multi-dimensional two-phase flow phenomena and for advanced two-phase flow modeling. In addition, the GUI (Graphic User Interface) feature were developed to enhance the user's convenience. To develop the coupled analysis capability, the MARS code were linked with the three-dimensional reactor kinetics code (MASTER), the core thermal analysis code (COBRA-III/CP), and the best-estimate containment analysis code (CONTEMPT), resulting in MARS/MASTER/COBRA/CONTEMPT. Currently, the MARS code system has been distributed to 18 domestic organizations, including research, industrial, regulatory organizations and universities. The MARS has been being widely used for the safety research of existing PWRs, advanced PWR, CANDU and research reactor, the pre-test analysis of TH experiments, and others

  19. Compatibility analysis of DUPIC fuel(4) - thermal hydraulic analysis

    Energy Technology Data Exchange (ETDEWEB)

    Park, Jee Won; Chae, Kyung Myung; Choi, Hang Bok

    2000-07-01

    Thermal-hydraulic compatibility of the DUPIC fuel bundle in the CANDU reactor has been studied. The critical channel power, the critical power ratio, the channel exit quality and the channel flow are calculated for the DUPIC and the standard fuels by using the NUCIRC code. The physical models and associated parametric values for the NUCIRC analysis of the fuels are also presented. Based upon the slave channel analysis, the critical channel power and the critical power ratios have been found to be very similar for the two fuel types. The same dryout model is used in this study for the standard and the DUPIC fuel bundles. To assess the dryout characteristics of the DUPIC fuel bundle, the ASSERT-PV code has been used for the subchannel analysis. Based upon the results of the subchannel analysis, it is found that the dryout location and the power for the two fuel types are indeed very similar. This study shows that thermal performance of the DUPIC fuel is not significantly different from that of the standard fuel.

  20. Thermal Hydraulic Analysis of RPV Support Cooling System for HTGR

    International Nuclear Information System (INIS)

    Min Qi; Wu Xinxin; Li Xiaowei; Zhang Li; He Shuyan

    2014-01-01

    Passive safety is now of great interest for future generation reactors because of its reduction of human interaction and avoidance of failures of active components. reactor pressure vessel (RPV) support cooling system (SCS) for high temperature gas-cooled reactor (HTGR) is a passive safety system and is used to cool the concrete seats for the four RPV supports at its bottom. The SCS should have enough cooling capacity to ensure the temperature of the concrete seats for the supports not exceeding the limit temperature. The SCS system is composed of a natural circulation water loop and an air cooling tower. In the water loop, there is a heat exchanger embedded in the concrete seat, heat is transferred by thermal conduction and convection to the cooling water. Then the water is cooled by the air cooler mounted in the air cooling tower. The driving forces for water and air are offered by the density differences caused by the temperature differences. In this paper, the thermal hydraulic analysis for this system was presented. Methods for decoupling the natural circulation and heat transfer between the water loop and air flow were introduced. The operating parameters for different working conditions and environment temperatures were calculated. (author)

  1. Hydraulic performance of compacted clay liners under simulated daily thermal cycles.

    Science.gov (United States)

    Aldaeef, A A; Rayhani, M T

    2015-10-01

    Compacted clay liners (CCLs) are commonly used as hydraulic barriers in several landfill applications to isolate contaminants from the surrounding environment and minimize the escape of leachate from the landfill. Prior to waste placement in landfills, CCLs are often exposed to temperature fluctuations which can affect the hydraulic performance of the liner. Experimental research was carried out to evaluate the effects of daily thermal cycles on the hydraulic performance of CCLs under simulated landfill conditions. Hydraulic conductivity tests were conducted on different soil specimens after being exposed to various thermal and dehydration cycles. An increase in the CCL hydraulic conductivity of up to one order of magnitude was recorded after 30 thermal cycles for soils with low plasticity index (PI = 9.5%). However, medium (PI = 25%) and high (PI = 37.2%) plasticity soils did not show significant hydraulic deviation due to their self-healing potential. Overlaying the CCL with a cover layer minimized the effects of daily thermal cycles, and maintained stable hydraulic performance in the CCLs even after exposure to 60 thermal cycles. Wet-dry cycles had a significant impact on the hydraulic aspect of low plasticity CCLs. However, medium and high plasticity CCLs maintained constant hydraulic performance throughout the test intervals. The study underscores the importance of protecting the CCL from exposure to atmosphere through covering it by a layer of geomembrane or an interim soil layer. Copyright © 2015 Elsevier Ltd. All rights reserved.

  2. COMMIX analysis of four constant flow thermal upramp experiments performed in a thermal hydraulic model of an advanced LMR

    International Nuclear Information System (INIS)

    Yarlagadda, B.S.

    1989-04-01

    The three-dimensional thermal hydraulics computer code COMMIX-1AR was used to analyze four constant flow thermal upramp experiments performed in the thermal hydraulic model of an advanced LMR. An objective of these analyses was the validation of COMMIX-1AR for buoyancy affected flows. The COMMIX calculated temperature histories of some thermocouples in the model were compared with the corresponding measured data. The conclusions of this work are presented. 3 refs., 5 figs

  3. Investigation on the effect of seawater to hydraulic property and wetting process of bentonite

    International Nuclear Information System (INIS)

    Hasegawa, Takuma

    2004-01-01

    On high-level waste disposal, bentonite is one of the most promising material for buffer and backfill material. The hydraulic properties and wetting process of bentonite are important not only for barrier performance assessment but also for prediction of waste disposal environment, such as resaturation time and thermal distribution. In Japan, we should consider the effect of seawater for bentonite, because radioactive waste will be disposed of in coastal area and in marine sediment where seawater remained. However, it is not enough to understand the effect of seawater. Therefore, experimental study was conducted to investigate the effect of seawater on the hydraulic conductivity and wetting process of bentonite. The effect of seawater on hydraulic conductivity is significant for Na-bentonite, the hydraulic conductivity of Na-bentonite in seawater is one order to magnitude higher than that in distilled water. On the other hand, the hydraulic conductivity of Ca-bentonite is not influenced by seawater. The hydraulic conductivity of bentonite decreases as effective montmorillonite density increases. The effective montmorillonite density is ratio between the weight of montmorillonite and volume of porosity and montmorillonite. The hydraulic conductivity of bentonite is close related to swelling property since the hydraulic conductivity decrease as the swelling pressure increase. Wetting process of compacted bentonite could be evaluated by diffusion phenomena since infiltration rate and change of saturation rate and represented by diffusion equation. The effect of seawater on water diffusivity is significant for Na-type bentonite with low effective montmorillonite density. Except for that condition, the water diffusivity of bentonite is almost constant and is not influenced by effective montmorillonite density and seawater. (author)

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

    Directory of Open Access Journals (Sweden)

    W.Z. Chen

    2017-06-01

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

  5. PWR Blowdown Heat Transfer Separate-Effects Program. Thermal-Hydraulic Test Facility experimental data report for test 166S

    International Nuclear Information System (INIS)

    Clemons, V.D.; White, M.D.; Hedrick, R.A.

    1978-01-01

    Reduced instrument responses are presented for Thermal-Hydraulic Test Facility (THTF) test 166S, which is part of the ORNL Pressurized-Water Reactor (PWR) Blowdown Heat Transfer Separate-Effects Program. The objective of the program is to investigate the thermal-hydraulic phenomenon governing the energy transfer and transport processes that occur during a loss-of-coolant accident in a PWR system. Test 166S was conducted to obtain thermal-hydraulic and CHF information in THTF bundle 1 with an intact hot leg. The primary purpose of this report is to make the reduced instrument responses during tests 166S available. These are presented in graphical form in engineering units and have been analyzed only to the extent necessary to ensure reasonableness and consistency

  6. PWR blowdown heat transfer separate-effects program: thermal-hydraulic test facility experimental data report for test 104

    International Nuclear Information System (INIS)

    Leon, D.M.; White, M.D.; Moore, P.A.; Hedrick, R.A.

    1978-01-01

    Reduced instrument responses are presented for Thermal-Hydraulic Test Facility (THTF) test 104, which is part of the ORNL Pressurized-Water Reactor (PWR) Blowdown Heat Transfer Separate-Effects Program. The objective of the program is to investigate the thermal-hydraulic phenomenon governing the energy transfer and transport processes that occur during a loss-of-coolant accident in the PWR system. Test 104 was conducted to obtain CHF in bundle 1 under blowdown conditions. The primary purpose of this report is to make the reduced instrument responses during test 104 available

  7. Thermal-hydraulics for space power, propulsion, and thermal management system design

    International Nuclear Information System (INIS)

    Krotiuk, W.J.

    1990-01-01

    The present volume discusses thermal-hydraulic aspects of current space projects, Space Station thermal management systems, the thermal design of the Space Station Free-Flying Platforms, the SP-100 Space Reactor Power System, advanced multi-MW space nuclear power concepts, chemical and electric propulsion systems, and such aspects of the Space Station two-phase thermal management system as its mechanical pumped loop and its capillary pumped loop's supporting technology. Also discussed are the startup thaw concept for the SP-100 Space Reactor Power System, calculational methods and experimental data for microgravity conditions, an isothermal gas-liquid flow at reduced gravity, low-gravity flow boiling, computations of Space Shuttle high pressure cryogenic turbopump ball bearing two-phase coolant flow, and reduced-gravity condensation

  8. Investigation of thermal hydraulic mixing mechanism in T-junction pipe with a 90-degree bend in upstream side for mitigation and controlling of thermal-striping phenomena (Joint research)

    International Nuclear Information System (INIS)

    Yuki, Kazuhisa; Hashizume, Hidetoshi; Tanaka, Masaaki; Muramatsu, Toshiharu

    2006-03-01

    In T-junction pipe, where two fluids of different temperatures mix together, temperature fluctuation is induced due to unstable fluid mixing. Since there is a possibility that high cycle thermal fatigue in structural materials is caused by this temperature fluctuation in fluid, the development of the mitigation and control techniques for the thermal fatigue is one of the most important issues in the future plant design. If a 90-degree bend exists in the upstream of T-junction pipe, a secondary flow formed in the bend makes the fluid mixing phenomena even more complex. This study aims to clarify the effects of curvature ratio of the bend on the temperature fluctuation in T-junction pipe, by the flow visualization with PIV (Particle Image Velocimetry) and the measurement of fluid temperature in the vicinity of wall with thermo couples. From the visualization, it is clarified that a jet from branch pipe swings and sways near the wall and they brings strong temperature fluctuation in comparison of the case without the 90-degree bend. In the case of the short curvature bend, in which there exists flow separation, the fluid mixing and the temperature fluctuation characteristics are completely different from those in the case of the long curvature bend. Furthermore, in the case of long curvature bend, in which the flow separation doesn't occur, there are cautionary conditions that the temperature fluctuation is maximized in a transition regime of a stratified flow and a turn-jet flow. It seems that the cause for this phenomena is the repetition of generation and disappearance of a circulating flow formed behind the jet, due to the oscillation of jet caused by an interaction between unsteady behavior of a secondary flow and the wakes formed behind the jet. In the experiments, experimental correlation equations to predict the temperature fluctuation were obtained in each flow pattern. By using these equations, it's possible to predict the suitable flow condition and piping

  9. The Phebus FP thermal-hydraulic analysis with Melcor

    International Nuclear Information System (INIS)

    Akgane, Kikuo; Kiso, Yoshihiro; Fukahori, Takanori; Yoshino, Mamoru

    1995-01-01

    The severe accident analysis code MELCOR, version 1.8.2, has been applied for thermal-hydraulic pre-test analysis of the first test of the Phebus FP program (test FPT-0) to study the best test parameters and the applicability of the code. The Phebus FP program is an in-pile test program which has been planned by the French Commissariate a L'Energie Atomique and the Commission of the European Union. The experiments are being conducted by an international collaboration to study the release and transport of fission products (FPs) under conditions assumed to be the most representative of those that would occur in a severe accident. The Phebus FP test apparatus simulates a test bundle of an in-pile section, the circuit including the steam generator U-tubes and the containment. The FPT-0 test was designed to simulate the heat-up and subsequent fuel bundle degradation after a loss of coolant severe accident, using fresh fuel. Two options for fuel degradation models in MELCOR have been applied to fuel degradation behavior. the first model assumes that fuel debris will be formed immediately after the fuel support fails by cladding relocation due to the candling process. The other is the uncollapsed bare fuel pellets option, in which the fuel pellets remain standing in a columnar shape until the fuel reaches its melting point, even if the cladding has been relocated by candling. The thermal-hydraulic behaviors in the circuit and containment of Phebus FP are discussed herein. Flow velocities in the Phebus FP circuit are high in order to produce turbulent flow in a small diameter test pipe. The MELCOR calculation has shown that the length of the hot leg and steam generator are adequate to attain steam temperatures or 700 degrees C and 150 degrees C in the respective outlets. The containment atmosphere temperature and humidity derived by once through integral system calculation show that objective test conditions would be satisfied in the Phebus FP experiment

  10. The Phebus FP thermal-hydraulic analysis with Melcor

    Energy Technology Data Exchange (ETDEWEB)

    Akgane, Kikuo; Kiso, Yoshihiro [Nuclear Power Engineering Corporation, Tokyo (Japan); Fukahori, Takanori [Hitachi Engineering Company, Ltd., Hitachi-shi Ibaraki-ken (Japan); Yoshino, Mamoru [Nuclear Engineering Ltd., Tosabori Nishi-ku (Japan)

    1995-09-01

    The severe accident analysis code MELCOR, version 1.8.2, has been applied for thermal-hydraulic pre-test analysis of the first test of the Phebus FP program (test FPT-0) to study the best test parameters and the applicability of the code. The Phebus FP program is an in-pile test program which has been planned by the French Commissariate a L`Energie Atomique and the Commission of the European Union. The experiments are being conducted by an international collaboration to study the release and transport of fission products (FPs) under conditions assumed to be the most representative of those that would occur in a severe accident. The Phebus FP test apparatus simulates a test bundle of an in-pile section, the circuit including the steam generator U-tubes and the containment. The FPT-0 test was designed to simulate the heat-up and subsequent fuel bundle degradation after a loss of coolant severe accident, using fresh fuel. Two options for fuel degradation models in MELCOR have been applied to fuel degradation behavior. the first model assumes that fuel debris will be formed immediately after the fuel support fails by cladding relocation due to the candling process. The other is the uncollapsed bare fuel pellets option, in which the fuel pellets remain standing in a columnar shape until the fuel reaches its melting point, even if the cladding has been relocated by candling. The thermal-hydraulic behaviors in the circuit and containment of Phebus FP are discussed herein. Flow velocities in the Phebus FP circuit are high in order to produce turbulent flow in a small diameter test pipe. The MELCOR calculation has shown that the length of the hot leg and steam generator are adequate to attain steam temperatures or 700{degrees}C and 150{degrees}C in the respective outlets. The containment atmosphere temperature and humidity derived by once through integral system calculation show that objective test conditions would be satisfied in the Phebus FP experiment.

  11. R and D on thermal hydraulics of core and core-bottom structure

    International Nuclear Information System (INIS)

    Inagaki, Yoshiyuki; Hino, Ryutaro; Kunitomi, Kazuhiko; Takase, Kazuyuki; Ioka, Ikuo; Maruyama, So

    2004-01-01

    Thermal hydraulic tests on the core and core-bottom structure of the high-temperature engineering test reactor (HTTR) were carried out with the helium engineering demonstration loop (HENDEL) under simulated reactor operating conditions. The HENDEL was composed of helium gas circulation loops, mother sections (M 1 and M 2 ) and adaptor section (A), and two test sections, i.e. the fuel stack test section (T 1 ) and in-core structure test section (T 2 ). In the T 1 test section simulating a fuel stack of the core, thermal and hydraulic performances of helium gas flowing through a fuel block were investigated for thermal design of the HTTR core. In the T 2 test section simulating the core-bottom structure, demonstration tests were performed to verify the structural integrity of graphite and metal components, seal performance against helium gas leakage among the graphite permanent blocks and thermal mixing performance of helium gas. The above test results in the T 1 and T 2 test sections were applied to the detailed design and licensing works of the HTTR and the HENDEL-loop was dismantled in 1999

  12. On-Line Core Thermal-Hydraulic Model Improvement

    International Nuclear Information System (INIS)

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

    2007-02-01

    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

  13. VIPRE-01: A thermal-hydraulic code for reactor cores

    International Nuclear Information System (INIS)

    Cuta, J.M.; Koontz, A.S.; Stewart, C.W.; Montgomery, S.D.; Nomura, K.K.

    1989-08-01

    The VIPRE-01 thermal hydraulics code for PWR and BWR analysis has undergone significant modifications and error correction. This manual for the updated code, designated as VIPRE-01 Mod-02, describes improvements that eliminate problems of slow convergence with the drift flux model in transient simulation. To update the VIPRE-01 code and its documentation the drift flux model of two-phase flow was implemented and error corrections developed during VIPRE-01 application were included. The project team modified the existing VIPRE-01 equations into drift flux model equations by developing additional terms. They also developed and implemented corrections for the errors identified during the last four years. They then validated the modified code against standard test data using selected test cases. The project team prepared documentation revisions reflecting code improvements and corrections to replace the corresponding sections in the original VIPRE documents. The revised VIPRE code, designated VIPRE-01 Mod-02, incorporates improvements that eliminate many shortcomings of the previous version. During the validation, the code produced satisfactory output compared with test data. The revised documentation is in the form of binder pages to replace existing pages in three of the original manuals

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

    International Nuclear Information System (INIS)

    Shin, C. W.; No, H. C.

    2015-01-01

    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

  15. Thermal hydraulics model for Sandia's annular core research reactor

    International Nuclear Information System (INIS)

    Rao, Dasari V.; El-Genk, Mohamed S.; Rubio, Reuben A.; Bryson, James W.; Foushee, Fabian C.

    1988-01-01

    A thermal hydraulics model was developed for the Annular Core Research Reactor (ACRR) at Sandia National Laboratories. The coupled mass, momentum and energy equations for the core were solved simultaneously using an explicit forward marching numerical technique. The model predictions of the temperature rise across the central channel of the ACRR core were within ± 10 percent agreement with the in-core temperature measurements. The model was then used to estimate the coolant mass flow rate and the axial distribution of the cladding surface temperature in the central and average channels as functions of the operating power and the water inlet subcooling. Results indicated that subcooled boiling occurs at the cladding surface in the central channels of the ACRR at power levels in excess of 0.5 MW. However, the high heat transfer coefficient due to subcooled boiling causes the cladding temperature along most of the active fuel rod region to be quite uniform and to increase very little with the reactor power. (author)

  16. 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.

  17. Nuclear reactor thermal hydraulics safety analysis and thoughts on FUKUSHIMA

    International Nuclear Information System (INIS)

    Ninokata, Hisashi

    2012-01-01

    The first part of this article is to show my thoughts on the accident at Fukushima Daiichi Nuclear Power Station. It is cited from a summary of my lecture talk in Indonesia, in the beginning of the last December, 2011. This talk was based on my previous lecture and seminar talks including those delivered at MIT, June 16, at the ANS Annual Meeting in Hollywood, Florida, June 28 at NURETH-13 in Toronto, September 27, and others. The content is based on the open and latest information available to date in Japan. It may contain some erroneous or uncertain information. I tried to minimize it to my best capability. Also I tried to eliminate any critical issues or opinions that may jeopardize some people who were involved in. The latter half of this article will be excerpts of my recent R and D activities related to the safety-by-design for sodium cooled fast reactors and light water reactors, thermal hydraulics analysis focusing on the simulation-based technology, in particular subchannel analysis and computational fluid dynamics. (J.P.N.)

  18. Parallelization methods study of thermal-hydraulics codes

    International Nuclear Information System (INIS)

    Gaudart, Catherine

    2000-01-01

    The variety of parallelization methods and machines leads to a wide selection for programmers. In this study we suggest, in an industrial context, some solutions from the experience acquired through different parallelization methods. The study is about several scientific codes which simulate a large variety of thermal-hydraulics phenomena. A bibliography on parallelization methods and a first analysis of the codes showed the difficulty of our process on the whole applications to study. Therefore, it would be necessary to identify and extract a representative part of these applications and parallelization methods. The linear solver part of the codes forced itself. On this particular part several parallelization methods had been used. From these developments one could estimate the necessary work for a non initiate programmer to parallelize his application, and the impact of the development constraints. The different methods of parallelization tested are the numerical library PETSc, the parallelizer PAF, the language HPF, the formalism PEI and the communications library MPI and PYM. In order to test several methods on different applications and to follow the constraint of minimization of the modifications in codes, a tool called SPS (Server of Parallel Solvers) had be developed. We propose to describe the different constraints about the optimization of codes in an industrial context, to present the solutions given by the tool SPS, to show the development of the linear solver part with the tested parallelization methods and lastly to compare the results against the imposed criteria. (author) [fr

  19. One-dimensional two-phase thermal hydraulics (ENSTA course)

    International Nuclear Information System (INIS)

    Olive, J.

    1995-11-01

    This course is part of the ENSTA 3rd year thermal hydraulics program (nuclear power option). Its purpose is to provide the theoretical basis and main physical notions pertaining to two-phase flow, mainly focussed on water-steam flows. The introduction describes the physical specificities of these flows, emphasizing their complexity. The mathematical bases are then presented (partial derivative equations), leading to a one-dimensional type, simplified description. Balances drawn up for a pipe length volume are used to introduce the mass conservation. motion and energy equations for each phase. Various postulates used to simplify two-phase models are presented, culminating in homogeneous model definitions and equations, several common examples of which are given. The model is then applied to the calculation of pressure drops in two-phase flows. This involves presenting the models most frequently used to represent pressure drops by friction or due to pipe irregularities, without giving details (numerical values of parameters). This chapter terminates with a brief description of static and dynamic instabilities in two-phase flows. Finally, heat transfer conditions frequently encountered in liquid-steam flows are described, still in the context of a 1D model. This chapter notably includes reference to under-saturated boiling conditions and the various forms of DNB. The empirical heat transfer laws are not discussed in detail. Additional material is appended, some of which is in the form of corrected exercises. (author). 6 appends

  20. Development of fuel performance and thermal hydraulic technology

    International Nuclear Information System (INIS)

    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)

  1. 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.

  2. Thermal Hydraulic Analysis Of Thorium-Based Annular Fuel Assemblies

    Energy Technology Data Exchange (ETDEWEB)

    Han, Kyu Hyun [Korea Institute of Nuclear Safety, 19, Guseong-dong, Yuseong-gu, Daejeon, 305-338 (Korea, Republic of)

    2008-07-01

    Thermal hydraulic characteristics of thorium-based fuel assemblies loaded with annular seed pins have been analyzed using AMAP combined with MATRA, and compared with those of the existing thorium-based assemblies. MATRA and AMAP showed good agreements for the pressure drops at the internal sub-channels. The pressure drop generally increased in the cases of the assemblies loaded with annular seed pins due to the larger wetted perimeter, but an exception existed. In the inner sub-channels of the seed pins, mass fluxes were high due to the grid form losses in the outer sub-channels. About 43% of the heat generated from the seed pin flowed into the inner sub-channel and the rest into the outer sub-channel, which implies the inner to outer wall heat flux ratio was approximately 1.2. The maximum temperatures of the annular seed pins were slightly above 500 deg. C. The MDNBRs of the assemblies loaded with annular seed pins were higher than those of the existing assemblies. Due to the fact that inter-channel mixing cannot occur in the inner sub-channels, temperatures and enthalpies were higher in the inner sub-channels. (author)

  3. Evaluation of hot spot factors for thermal and hydraulic design of HTTR

    International Nuclear Information System (INIS)

    Maruyama, So; Yamashita, Kiyonobu; Fujimoto, Nozomu; Murata, Isao; Sudo, Yukio; Murakami, Tomoyuki; Fujii, Sadao.

    1993-01-01

    High Temperature Engineering Test Reactor (HTTR) is a graphite-moderated and helium gas-cooled reactor with 30 MW in thermal power and 950degC in reactor outlet coolant temperature. One of the major items in thermal and hydraulic design of the HTTR is to evaluate the maximum fuel temperature with a sufficient margin from a viewpoint of integrity of coated fuel particles. Hot spot factors are considered in the thermal and hydraulic design to evaluate the fuel temperature not only under the normal operation condition but also under any transient condition conservatively. This report summarizes the items of hot spot factors selected in the thermal and hydraulic design and their estimated values, and also presents evaluation results of the thermal and hydraulic characteristics of the HTTR briefly. (author)

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

    Science.gov (United States)

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

    2016-04-01

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

  5. Experimental studies on thermal hydraulic responses for transient operations of the SMART-P

    International Nuclear Information System (INIS)

    Choi, K.Y.; Park, H.S.; Cho, S.; Park, C.K.; Lee, S.J.; Song, C.H.; Chung, M.K.

    2005-01-01

    Full text of publication follows: Thermal hydraulic responses for transient operations of the SMART-P are experimentally investigated by using a integral effect test facility. This test facility (VISTA) has been constructed to simulate the SMART-P, which is a pilot plant of the SMART. The SMART-P is an advanced modular integral type pressurized water reactor (65 MWt) whose major RCS components, such as main coolant pumps, helical-coiled tube bundle steam generators and pressurizers, are contained in a reactor vessel. This integral design approach eliminates the large coolant loop piping, thus eliminates the occurrence of a large break LOCA. Passive Residual Heat Removal System (PRHRS) is installed to prevent overheating and over-pressurization of the primary system during accidental conditions. The PRHRS of the SMART-P removes the core decay heat by natural circulation of the two-phase fluid. The VISTA facility is a full height and 1/96 volume scaled test facility with respect to the SMART-P and will be used to understand the thermal-hydraulic responses following transients and finally to verify the system design of the SMART-P. The experimental data from the VISTA facility will be essential to system designers to resolve open issues relevant to the design of the SMART-P. The full functional control logics are implanted into the VISTA facility to cope with abnormal transients. The core of the facility can be selectively controlled by either a T-control or a T+N control method. The T-control method is a control method to adjust the core power according to the core exit coolant temperature and is designed to be used for high primary coolant flow conditions. On the other hand, the T+N control method is for low primary coolant flow conditions and it uses core exit temperature as well as core power itself as control inputs. The thermal hydraulic responses are carefully investigated according to different core control methods. Several experiments have been performed to

  6. Thermal-hydraulic behavior on break simulation of steam generator U-tube

    International Nuclear Information System (INIS)

    Seul, Kwang Won; Bang, Young Seok; Lee, Sukho; Kim, Hho Jung

    1995-01-01

    The thermal-hydraulic behavior depending on the break simulation in a steam generator U-tube was investigated and identified the code predictability on plant responses during SGTR accident. The calculated results were compared and assessed with LSTF SB-SG-06 test data. The RELAP5/MOD3.1 code well predicted the sequence of events and the significant phenomena, such as the asymmetric loop behavior, the RCS cooldown and heat transfer by the natural circulation, and system depressurization, even though there were some differences from the experimental data. The break flowrate was found to be sensitive to the break model and affected the system behavior

  7. Dynamic force profile in hydraulic hybrid vehicles: a numerical investigation

    Science.gov (United States)

    Mohaghegh-Motlagh, Amin; Elahinia, Mohammad H.

    2010-04-01

    A hybrid hydraulic vehicle (HHV) combines a hydraulic sub-system with the conventional drivetrain in order to improve fuel economy for heavy vehicles. The added hydraulic module manages the storage and release of fluid power necessary to assist the motion of the vehicle. The power collected by a pump/motor (P/M) from the regenerative braking phase is stored in a high-pressure accumulator and then released by the P/M to the driveshaft during the acceleration phase. This technology is effective in significantly improving fuel-economy for heavy-class vehicles with frequent stop-and-go drive schedules. Despite improved fuel economy and higher vehicle acceleration, noise and vibrations are one of the main problems of these vehicles. The dual function P/Ms are the main source of noise and vibration in a HHV. This study investigates the dynamics of a P/M and particularly the profile and frequency-dependence of the dynamic forces generated by a bent-axis P/M unit. To this end, the fluid dynamics side of the problem has been simplified for investigating the system from a dynamics perspective. A mathematical model of a bent axis P/M has been developed to investigate the cause of vibration and noise in HHVs. The forces are calculated in time and frequency domains. The results of this work can be used to study the vibration response of the chassis and to design effective vibration isolation systems for HHVs.

  8. NEPTUNE: A new software platform for advanced nuclear thermal hydraulics

    International Nuclear Information System (INIS)

    Guelfi, A.; Boucker, M.; Herard, J.M.; Peturaud, P.; Bestion, D.; Boudier, P.; Hervieu, E.; Fillion, P.; Grandotto, M.

    2007-01-01

    The NEPTUNE project constitutes the thermal-hydraulic part of the long-term Electricite de France and Commissariat a l'Energie Atomique joint research and development program for the next generation of nuclear reactor simulation tools. This program is also financially supported by the Institut de Radioprotection et Surete Nucleaire and AREVA NP. The project aims at developing a new software platform for advanced two-phase flow thermal hydraulics covering the whole range of modeling scales and allowing easy multi-scale and multidisciplinary calculations. NEPTUNE is a fully integrated project that covers the following fields: software development, research in physical modeling and numerical methods, development of advanced instrumentation techniques, and performance of new experimental programs. The analysis of the industrial needs points out that three main simulation scales are involved. The system scale is dedicated to the overall description of the reactor. The component or subchannel scale allows three-dimensional computations of the main components of the reactors: cores, steam generators, condensers, and heat exchangers. The current generation of system and component codes has reached a very high level of maturity for industrial applications. The third scale, computational fluid dynamics (CFD) in open medium, allows one to go beyond the limits of the component scale for a finer description of the flows. This scale opens promising perspectives for industrial simulations, and the development and validation of the NEPTUNE CFD module have been a priority since the beginning of the project. It is based on advanced physical models (two-fluid or multi field model combined with interfacial area transport and two-phase turbulence) and modern numerical methods (fully unstructured finite volume solvers). For the system and component scales, prototype developments have also started, including new physical models and numerical methods. In addition to scale

  9. Proceedings of transient thermal-hydraulics and coupled vessel and piping system responses 1991

    International Nuclear Information System (INIS)

    Wang, G.Y.; Shin, Y.W.; Moody, F.J.

    1991-01-01

    This book reports on transient thermal-hydraulics and coupled vessel and piping system responses. Topics covered include: nuclear power plant containment designs; analysis of control rods; gate closure of hydraulic turbines; and shock wave solutions for steam water mixtures in piping systems

  10. Unsolved issues related to thermal-hydraulics in the suppression chamber during Fukushima Daiichi accident progressions

    International Nuclear Information System (INIS)

    Mizokami, Shinya; Yamada, Daichi; Honda, Takeshi; Yamauchi, Daisuke; Yamanaka, Yasunori

    2016-01-01

    On 11 March 2011, the Great East Japan Earthquake and Tsunami hit the Fukushima Daiichi Nuclear Power Station. The Fukushima Daiichi Units 1-3 lost all DC and AC power supplies, which set in motion a chain of events that led to releases of radioactivity to the environment. Since then, TEPCO has made many efforts to investigate the accident progressions and the status of the reactors and containment vessels. However, there still exist several tens of unsolved issues to be investigated for the fully understanding of the accident. In this paper, we introduce the unsolved issues related to thermal-hydraulics in the suppression chamber during the Fukushima Daiichi accident progressions. Especially, in Units 2 and 3, there are possibilities that thermal stratification inside their suppression chambers played an important role. It is important that these phenomena are addressed following both theoretical and experimental approaches as support to severe accident simulations. (author)

  11. Influence of sinusoidal flow on the thermal and hydraulic performance of microchannel heat sink

    International Nuclear Information System (INIS)

    Om, N I; Gunnasegaran, P; Rajasegaran, S

    2013-01-01

    In this paper, the effect of sinusoidal flow on the thermal and hydraulic performance of microchannel heat sink (MCHS) is numerically investigated. This investigation covers Reynolds number in the range of 100 ≤ Re ≤ 1000 and pure water is used as a working fluid. The three-dimensional steady, laminar flow and heat transfer governing equations are solved using finite volume method (FVM). The water flow field and heat transfer performance inside the sinusoidal microchannels is simulated and the results are compared with the straight microchannels. The effect of using sinusoidal microchannels on temperature distribution, Nusselt number, friction factor and thermal resistance is presented in this paper. It is found that with same rectangular cross-section, sinusoidal microchannels have a better heat transfer performance compared to the straight microchannels.

  12. Characteristics of Core Thermal-Hydraulic Design of SMART-P

    International Nuclear Information System (INIS)

    Hwang, Dae-Hyun; Seo, Kyong-Won; Kim, Tae-Wan; Lee, Chung-Chan

    2006-01-01

    The SMART (System-Integrated Modular Advanced ReacTor) is an integral-type advanced light water reactor which is purposed to be utilized as an energy source for sea water desalination as well as a small scale power generation. A prototype of this reactor, named SMART-P, has been studied at KAERI in order to demonstrate the relevant technologies incorporated in the SMART design. Due to the closed-channel type fuel assemblies and low mass velocity in the reactor core, the thermal hydraulic design features of SMART-P revealed fairly different characteristics in comparison with existing PWRs. The allowable operating region of the core, from the aspect of the thermal integrity of the fuel, should be primarily limited by two design parameters; critical heat flux (CHF) and fuel temperature. The occurrence of CHF may cause a sudden increase of the cladding temperature which eventually results in the fuel failure. The fuel temperature limit is relevant to a fuel failure mechanism such as a fuel centerline melting or a phase change of metallic fuels. Two phase flow instability is also an important design parameter since a flow oscillation may trigger a CHF or mechanical vibration of the channel. The characteristics of important thermal-hydraulic design parameters have been investigated for the SMART-P core with the closed-channel type fuel assemblies which contained non-square arrayed SSF (Self-sustained Square Finned) fuel rods

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

    International Nuclear Information System (INIS)

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

    1998-01-01

    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

  14. 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.

  15. 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.

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

    International Nuclear Information System (INIS)

    Chung, B. D.; Song, C. H.; Lee, Y. J.; Kwon, T. S.

    2003-04-01

    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

  17. Proceedings of the 10. Meeting on Reactor Physics and Thermal Hydraulics

    International Nuclear Information System (INIS)

    Santos Bastos, W. dos

    1995-01-01

    These proceedings presents all the Meeting papers emphasizing specific aspects on reactor physics method, criticality, fuel management, nuclear data, safety analysis, simulation and shielding, neutronics, thermal hydraulics, reactor operation and computational methods

  18. Computational features of the MELT-III neutronics, thermal-hydraulics computer code system

    International Nuclear Information System (INIS)

    Wilburn, N.P.; Waltar, A.E.

    1976-01-01

    A multichannel, thermal-hydraulics, neutronic accident analysis program for simulating fast reactor behavior from a hypothetical accident inception to the start of core disassembly or to reactor shutdown is described

  19. Study of thermal-hydraulic analyses with CIP method

    International Nuclear Information System (INIS)

    Doi, Yoshihiro

    1996-09-01

    New type of numerical scheme CIP has been proposed for solving hyperbolic type equations and the CIP is focused on as a less numerical diffusive scheme. C-CUP method with the CIP scheme is adopted to numerical simulations that treat compressible and incompressible fluids, phase change phenomena and Mixture fluids. To evaluate applicabilities of the CIP scheme and C-CUP method for thermal hydraulic analyses related to Fast Breeder Reactors (FBRs), the scheme and the method were reviewed. Feature of the CIP scheme and procedure of the C-CUP method were presented. The CIP scheme is used to solve linear hyperbolic type equations for advection term in basic equations of fluids. Key issues of the scheme is that profile between grid points is described to solve the equation by cubic polynomial and spatial derivatives of the polynomial. The scheme can capture steep change of solution and suppress numerical error. In the C-CUP method, the basic equations of fluids are divided into advection terms and the other terms. The advection terms is solved with CIP scheme and the other terms is solved with difference method. The C-CUP method is robust for numerical instability, but mass of fluid will be in unfair preservation with nonconservative equations for fluids. Numerical analyses with the CIP scheme and the C-CUP method has been performed for phase change, mixture and moving object. These analyses are depend on characteristics of that the scheme and the method are robust for steep change of density and useful for interface tracking. (author)

  20. Validation of thermal hydraulic codes for fusion reactors safety

    International Nuclear Information System (INIS)

    Sardain, P.; Gulden, W.; Massaut, V.; Takase, K.; Merill, B.; Caruso, G.

    2006-01-01

    A significant effort has been done worldwide on the validation of thermal hydraulic codes, which can be used for the safety assessment of fusion reactors. This work is an item of an implementing agreement under the umbrella of the International Energy Agency. The European part is supported by EFDA. Several programmes related to transient analysis in water-cooled fusion reactors were run in order to assess the capabilities of the codes to treat the main physical phenomena governing the accidental sequences related to water/steam discharge into the vacuum vessel or the cryostat. The typical phenomena are namely the pressurization of a volume at low initial pressure, the critical flow, the flashing, the relief into an expansion volume, the condensation of vapor in a pressure suppression system, the formation of ice on a cryogenic structure, the heat transfer between walls and fluid in various thermodynamic conditions. · A benchmark exercise has been done involving different types of codes, from homogeneous equilibrium to six equations non-equilibrium models. Several cases were defined, each one focusing on a particular phenomenon. · The ICE (Ingress of Coolant Event) facility has been operated in Japan. It has simulated an in-vessel LOCA and the discharge of steam into a pressure suppression system. · The EVITA (European Vacuum Impingement Test Apparatus) facility has been operated in France. It has simulated ingress of coolant into the cryostat, i.e. into a volume at low initial pressure containing surfaces at cryogenic temperature. This paper gives the main lessons gained from these programs, in particular the possibilities for the improvement of the computer codes, extending their capabilities. For example, the water properties have been extended below the triple point. Ice formation models have been implemented. Work has also been done on condensation models. The remaining needs for R-and-D are also highlighted. (author)

  1. VIPRE-01. a thermal-hydraulic analysis code for reactor cores. Volume 1. Mathematical modeling

    International Nuclear Information System (INIS)

    Stewart, C.W.; Cuta, J.M.; Koontz, A.S.; Kelly, J.M.; Basehore, K.L.; George, T.L.; Rowe, D.S.

    1983-04-01

    VIPRE (Versatile Internals and Component Program for Reactors; EPRI) has been developed for nuclear power utility thermal-hydraulic analysis applications. It is designed to help evaluate nuclear reactor core safety limits including minimum departure from nucleate boiling ratio (MDNBR), critical power ratio (CPR), fuel and clad temperatures, and coolant state in normal operation and assumed accident conditions. This volume (Volume 1: Mathematical Modeling) explains the major thermal hydraulic models and supporting correlations in detail

  2. 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.

  3. 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.).

  4. Visual and intelligent transients and accidents analyzer based on thermal-hydraulic system code

    International Nuclear Information System (INIS)

    Meng Lin; Rui Hu; Yun Su; Ronghua Zhang; Yanhua Yang

    2005-01-01

    Full text of publication follows: Many thermal-hydraulic system codes were developed in the past twenty years, such as RELAP5, RETRAN, ATHLET, etc. Because of their general and advanced features in thermal-hydraulic computation, they are widely used in the world to analyze transients and accidents. But there are following disadvantages for most of these original thermal-hydraulic system codes. Firstly, because models are built through input decks, so the input files are complex and non-figurative, and the style of input decks is various for different users and models. Secondly, results are shown in off-line data file form. It is not convenient for analysts who may pay more attention to dynamic parameters trend and changing. Thirdly, there are few interfaces with other program in these original thermal-hydraulic system codes. This restricts the codes expanding. The subject of this paper is to develop a powerful analyzer based on these thermal-hydraulic system codes to analyze transients and accidents more simply, accurately and fleetly. Firstly, modeling is visual and intelligent. Users build the thermalhydraulic system model using component objects according to their needs, and it is not necessary for them to face bald input decks. The style of input decks created automatically by the analyzer is unified and can be accepted easily by other people. Secondly, parameters concerned by analyst can be dynamically communicated to show or even change. Thirdly, the analyzer provide interface with other programs for the thermal-hydraulic system code. Thus parallel computation between thermal-hydraulic system code and other programs become possible. In conclusion, through visual and intelligent method, the analyzer based on general and advanced thermal-hydraulic system codes can be used to analysis transients and accidents more effectively. The main purpose of this paper is to present developmental activities, assessment and application results of the visual and intelligent

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

    International Nuclear Information System (INIS)

    Akimoto, Hajime; Kukita; Ohnuki, Akira

    1997-01-01

    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

  6. 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).

  7. A flexible coupling scheme for Monte Carlo and thermal-hydraulics codes

    Energy Technology Data Exchange (ETDEWEB)

    Hoogenboom, J. Eduard, E-mail: J.E.Hoogenboom@tudelft.nl [Delft University of Technology (Netherlands); Ivanov, Aleksandar; Sanchez, Victor, E-mail: Aleksandar.Ivanov@kit.edu, E-mail: Victor.Sanchez@kit.edu [Karlsruhe Institute of Technology, Institute of Neutron Physics and Reactor Technology, Eggenstein-Leopoldshafen (Germany); Diop, Cheikh, E-mail: Cheikh.Diop@cea.fr [CEA/DEN/DANS/DM2S/SERMA, Commissariat a l' Energie Atomique, Gif-sur-Yvette (France)

    2011-07-01

    A coupling scheme between a Monte Carlo code and a thermal-hydraulics code is being developed within the European NURISP project for comprehensive and validated reactor analysis. The scheme is flexible as it allows different Monte Carlo codes and different thermal-hydraulics codes to be used. At present the MCNP and TRIPOLI4 Monte Carlo codes can be used and the FLICA4 and SubChanFlow thermal-hydraulics codes. For all these codes only an original executable is necessary. A Python script drives the iterations between Monte Carlo and thermal-hydraulics calculations. It also calls a conversion program to merge a master input file for the Monte Carlo code with the appropriate temperature and coolant density data from the thermal-hydraulics calculation. Likewise it calls another conversion program to merge a master input file for the thermal-hydraulics code with the power distribution data from the Monte Carlo calculation. Special attention is given to the neutron cross section data for the various required temperatures in the Monte Carlo calculation. Results are shown for an infinite lattice of PWR fuel pin cells and a 3 x 3 fuel BWR pin cell cluster. Various possibilities for further improvement and optimization of the coupling system are discussed. (author)

  8. Development of MARS for multi-dimensional and multi-purpose thermal-hydraulic system analysis

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Won Jae; Chung, Bub Dong; Kim, Kyung Doo; Hwang, Moon Kyu; Jeong, Jae Jun; Ha, Kwi Seok; Joo, Han Gyu [Korea Atomic Energy Research Institute, T/H Safety Research Team, Yusung, Daejeon (Korea)

    2000-10-01

    MARS (Multi-dimensional Analysis of Reactor Safety) code is being developed by KAERI for the realistic thermal-hydraulic simulation of light water reactor system transients. MARS 1.4 has been developed as a final version of basic code frame for the multi-dimensional analysis of system thermal-hydraulics. Since MARS 1.3, MARS 1.4 has been improved to have the enhanced code capability and user friendliness through the unification of input/output features, code models and code functions, and through the code modernization. Further improvements of thermal-hydraulic models, numerical method and user friendliness are being carried out for the enhanced code accuracy. As a multi-purpose safety analysis code system, a coupled analysis system, MARS/MASTER/CONTEMPT, has been developed using multiple DLL (Dynamic Link Library) techniques of Windows system. This code system enables the coupled, that is, more realistic analysis of multi-dimensional thermal-hydraulics (MARS 2.0), three-dimensional core kinetics (MASTER) and containment thermal-hydraulics (CONTEMPT). This paper discusses the MARS development program, and the developmental progress of the MARS 1.4 and the MARS/MASTER/CONTEMPT focusing on major features of the codes and their verification. It also discusses thermal hydraulic models and new code features under development. (author)

  9. Proceedings of the fourth international topical meeting on nuclear thermal hydraulics, operations and safety. Vol. 1

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2004-07-01

    More than 100 papers were presented. The meeting was divided in 56 sessions and covered the following topics: Plant Operation, Retrofitting and Maintenance Experience; Steam Generator Operation and Maintenance; Artificial Intelligence and Expert Systems; Seismic Technologies for Plant Design and Operations; Aging Management and Life Extension; Two-Phase Flow Modeling and Applications; Severe Accidents and Degraded Core Thermal Hydraulics; Plant Simulators, Analyzers, and Workstations; Advanced Nuclear Fuel Challenges; Recent Nuclear Power Station Decommissioning Experiences in the USA; Application of Probabilistic risk assessment/Probabilistic safety assessment (PRA/PSA) in Design and Modification; Numerical Modeling in Thermal Hydraulics; General Thermal Hydraulics; Severe Accident Management; Licensing and Regulatory Requirements; Advanced Light Water Reactor Designs to Support Reduced Emergency Planning; Best Estimate loss-of-coolant (LOCA) Methodologies; Plant Instrumentation and Control; LWR Fuel Designs for Improved Thermal Hydraulic Performance; Performance Assessment of Radioactive Waste Disposal; Thermal Hydraulics in Passive Reactor Systems; Advances in Man-Machine Interface Design and the Related Human Factors Engineering; Advances in Measurements and Instrumentation; Computer Aided Technology for non-destructive evaluation (NDE) and Plant Maintenance Plant Uprating; Flow-Accelerated Corrosion in Nuclear Power Plants; Advances in Radiological Measurement and Analysis Risk Management and Assessment; Stability in Thermal Hydraulic Systems; Critical heat flux (CHF) and Post Dryout Heat Transfer; Plant Transient and Accident Modeling.

  10. Proceedings of the fourth international topical meeting on nuclear thermal hydraulics, operations and safety. Vol. 1

    International Nuclear Information System (INIS)

    2004-01-01

    More than 100 papers were presented. The meeting was divided in 56 sessions and covered the following topics: Plant Operation, Retrofitting and Maintenance Experience; Steam Generator Operation and Maintenance; Artificial Intelligence and Expert Systems; Seismic Technologies for Plant Design and Operations; Aging Management and Life Extension; Two-Phase Flow Modeling and Applications; Severe Accidents and Degraded Core Thermal Hydraulics; Plant Simulators, Analyzers, and Workstations; Advanced Nuclear Fuel Challenges; Recent Nuclear Power Station Decommissioning Experiences in the USA; Application of Probabilistic risk assessment/Probabilistic safety assessment (PRA/PSA) in Design and Modification; Numerical Modeling in Thermal Hydraulics; General Thermal Hydraulics; Severe Accident Management; Licensing and Regulatory Requirements; Advanced Light Water Reactor Designs to Support Reduced Emergency Planning; Best Estimate loss-of-coolant (LOCA) Methodologies; Plant Instrumentation and Control; LWR Fuel Designs for Improved Thermal Hydraulic Performance; Performance Assessment of Radioactive Waste Disposal; Thermal Hydraulics in Passive Reactor Systems; Advances in Man-Machine Interface Design and the Related Human Factors Engineering; Advances in Measurements and Instrumentation; Computer Aided Technology for non-destructive evaluation (NDE) and Plant Maintenance Plant Uprating; Flow-Accelerated Corrosion in Nuclear Power Plants; Advances in Radiological Measurement and Analysis Risk Management and Assessment; Stability in Thermal Hydraulic Systems; Critical heat flux (CHF) and Post Dryout Heat Transfer; Plant Transient and Accident Modeling

  11. Proceedings of the fourth international topical meeting on nuclear thermal hydraulics, operations and safety. Vol. 2

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2004-07-01

    More than 100 papers presented at the meeting were divided in 56 sessions and covered the following topics: Plant Operation, Retrofitting and Maintenance Experience; Steam Generator Operation and Maintenance; Artificial Intelligence and Expert Systems; Seismic Technologies for Plant Design and Operations; Aging Management and Life Extension; Two-Phase Flow Modeling and Applications; Severe Accidents and Degraded Core Thermal Hydraulics; Plant Simulators, Analyzers, and Workstations; Advanced Nuclear Fuel Challenges; Recent Nuclear Power Station Decommissioning Experiences in the USA; Application of Probabilistic risk assessment/Probabilistic safety assessment (PRA/PSA) in Design and Modification; Numerical Modeling in Thermal Hydraulics; General Thermal Hydraulics; Severe Accident Management; Licensing and Regulatory Requirements; Advanced Light Water Reactor Designs to Support Reduced Emergency Planning; Best Estimate loss-of-coolant (LOCA) Methodologies; Plant Instrumentation and Control; LWR Fuel Designs for Improved Thermal Hydraulic Performance; Performance Assessment of Radioactive Waste Disposal; Thermal Hydraulics in Passive Reactor Systems; Advances in Man-Machine Interface Design and the Related Human Factors Engineering; Advances in Measurements and Instrumentation; Computer Aided Technology for non-destructive evaluation (NDE) and Plant Maintenance Plant Uprating; Flow-Accelerated Corrosion in Nuclear Power Plants; Advances in Radiological Measurement and Analysis Risk Management and Assessment; Stability in Thermal Hydraulic Systems; Critical heat flux (CHF) and Post Dryout Heat Transfer; Plant Transient and Accident Modeling.

  12. Proceedings of the fourth international topical meeting on nuclear thermal hydraulics, operations and safety. Vol. 2

    International Nuclear Information System (INIS)

    2004-01-01

    More than 100 papers presented at the meeting were divided in 56 sessions and covered the following topics: Plant Operation, Retrofitting and Maintenance Experience; Steam Generator Operation and Maintenance; Artificial Intelligence and Expert Systems; Seismic Technologies for Plant Design and Operations; Aging Management and Life Extension; Two-Phase Flow Modeling and Applications; Severe Accidents and Degraded Core Thermal Hydraulics; Plant Simulators, Analyzers, and Workstations; Advanced Nuclear Fuel Challenges; Recent Nuclear Power Station Decommissioning Experiences in the USA; Application of Probabilistic risk assessment/Probabilistic safety assessment (PRA/PSA) in Design and Modification; Numerical Modeling in Thermal Hydraulics; General Thermal Hydraulics; Severe Accident Management; Licensing and Regulatory Requirements; Advanced Light Water Reactor Designs to Support Reduced Emergency Planning; Best Estimate loss-of-coolant (LOCA) Methodologies; Plant Instrumentation and Control; LWR Fuel Designs for Improved Thermal Hydraulic Performance; Performance Assessment of Radioactive Waste Disposal; Thermal Hydraulics in Passive Reactor Systems; Advances in Man-Machine Interface Design and the Related Human Factors Engineering; Advances in Measurements and Instrumentation; Computer Aided Technology for non-destructive evaluation (NDE) and Plant Maintenance Plant Uprating; Flow-Accelerated Corrosion in Nuclear Power Plants; Advances in Radiological Measurement and Analysis Risk Management and Assessment; Stability in Thermal Hydraulic Systems; Critical heat flux (CHF) and Post Dryout Heat Transfer; Plant Transient and Accident Modeling

  13. A flexible coupling scheme for Monte Carlo and thermal-hydraulics codes

    International Nuclear Information System (INIS)

    Hoogenboom, J. Eduard; Ivanov, Aleksandar; Sanchez, Victor; Diop, Cheikh

    2011-01-01

    A coupling scheme between a Monte Carlo code and a thermal-hydraulics code is being developed within the European NURISP project for comprehensive and validated reactor analysis. The scheme is flexible as it allows different Monte Carlo codes and different thermal-hydraulics codes to be used. At present the MCNP and TRIPOLI4 Monte Carlo codes can be used and the FLICA4 and SubChanFlow thermal-hydraulics codes. For all these codes only an original executable is necessary. A Python script drives the iterations between Monte Carlo and thermal-hydraulics calculations. It also calls a conversion program to merge a master input file for the Monte Carlo code with the appropriate temperature and coolant density data from the thermal-hydraulics calculation. Likewise it calls another conversion program to merge a master input file for the thermal-hydraulics code with the power distribution data from the Monte Carlo calculation. Special attention is given to the neutron cross section data for the various required temperatures in the Monte Carlo calculation. Results are shown for an infinite lattice of PWR fuel pin cells and a 3 x 3 fuel BWR pin cell cluster. Various possibilities for further improvement and optimization of the coupling system are discussed. (author)

  14. Development of MARS for multi-dimensional and multi-purpose thermal-hydraulic system analysis

    International Nuclear Information System (INIS)

    Lee, Won Jae; Chung, Bub Dong; Kim, Kyung Doo; Hwang, Moon Kyu; Jeong, Jae Jun; Ha, Kwi Seok; Joo, Han Gyu

    2000-01-01

    MARS (Multi-dimensional Analysis of Reactor Safety) code is being developed by KAERI for the realistic thermal-hydraulic simulation of light water reactor system transients. MARS 1.4 has been developed as a final version of basic code frame for the multi-dimensional analysis of system thermal-hydraulics. Since MARS 1.3, MARS 1.4 has been improved to have the enhanced code capability and user friendliness through the unification of input/output features, code models and code functions, and through the code modernization. Further improvements of thermal-hydraulic models, numerical method and user friendliness are being carried out for the enhanced code accuracy. As a multi-purpose safety analysis code system, a coupled analysis system, MARS/MASTER/CONTEMPT, has been developed using multiple DLL (Dynamic Link Library) techniques of Windows system. This code system enables the coupled, that is, more realistic analysis of multi-dimensional thermal-hydraulics (MARS 2.0), three-dimensional core kinetics (MASTER) and containment thermal-hydraulics (CONTEMPT). This paper discusses the MARS development program, and the developmental progress of the MARS 1.4 and the MARS/MASTER/CONTEMPT focusing on major features of the codes and their verification. It also discusses thermal hydraulic models and new code features under development. (author)

  15. Thermal-hydraulic development a small, simplified, proliferation-resistant reactor

    International Nuclear Information System (INIS)

    Farmer, M. T.; Hill, D. J.; Sienicki, J. J.; Spencer, B. W.; Wade, D. C.

    1999-01-01

    This paper addresses thermal-hydraulics related criteria and preliminary concepts for a small (300 MWt), proliferation-resistant, liquid-metal-cooled reactor system. A main objective is to assess what extent of simplification is achievable in the concepts with the primary purpose of regaining economic competitiveness. The approach investigated features lead-bismuth eutectic (LBE) and a low power density core for ultra-long core lifetime (goal 15 years) with cartridge core replacement at end of life. This potentially introduces extensive simplifications resulting in capital cost and operating cost savings including: (1) compact, modular, pool-type configuration for factory fabrication, (2) 100+% natural circulation heat transport with the possibility of eliminating the main coolant pumps, (3) steam generator modules immersed directly in the primary coolant pool for elimination of the intermediate heat transport system, and (4) elimination of on-site fuel handling and storage provisions including rotating plug. Stage 1 natural circulation model and results are presented. Results suggest that 100+% natural circulation heat transport is readily achievable using LBE coolant and the long-life cartridge core approach; moreover, it is achievable in a compact pool configuration considerably smaller than PRISM A (for overland transportability) and with peak cladding temperature within the existing database range for ferritic steel with oxide layer surface passivation. Stage 2 analysis follows iteration with core designers. Other thermal hydraulic investigations are underway addressing passive, auxiliary heat removal by air cooling of the reactor vessel and the effects of steam generator tube rupture

  16. 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)

  17. Thermal hydraulic and neutron kinetic coupled simulation of the IPR-R1 Triga reactor

    Energy Technology Data Exchange (ETDEWEB)

    Reis, Patricia A.L.; Costa, Antonella L.; Pereira, Claubia; Silva, Clarysson A.M. da; Veloso, Maria Auxiliadora F.; Soares, Humbero V., E-mail: patricialire@yahoo.com.br, E-mail: antonella@nuclear.ufmg.br, E-mail: claubia@nuclear.ufmg.br, E-mail: clarysson@nuclear.ufmg.br, E-mail: dora@nuclear.ufmg.br, E-mail: betovitor@ig.com.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 Rede), Rio de Janeiro, RJ (Brazil)

    2013-07-01

    The nuclear industry and the scientific community have turned the attention for the development of coupled 3D neutron kinetics (NK) and thermal-hydraulic (TH) system codes to investigate specific nuclear reactor transients. Improving in theoretical investigations of complex phenomena in nuclear reactor technology have been increased thanks to numerical methods and computational resources incorporated in nuclear codes. This paper presents a model for the IPR-R1 TRIGA research reactor using the RELAP5-3D 3.0 code. The development and the assessment of the thermal-hydraulic RELAP5 code model for the IPR-R1 have been validated for steady state and transient situations and the results were published in preceding works. Results of RELAP5-3D steady state and a transient case presented in this paper show good agreement with experimental data, validating then this model for point kinetic calculations. To supply adequate cross sections to the NK code, the WIMSD5 is being used. First results of steady state calculation using the 3D neutron modeling are being presented in this paper. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1998-12-31

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2004-02-15

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

  20. Optimised design and thermal-hydraulic analysis of the IFMIF/HFTM test section

    Energy Technology Data Exchange (ETDEWEB)

    Gordeev, S.; Heinzel, V.; Lang, K.H.; Moeslang, A.; Schleisiek, K.; Slobodtchouk, V.; Stratmanns, E.

    2003-10-01

    On the basis of previous concepts, analyses and experiments, the high flux test module (HFTM) for the International Fusion Materials Irradiation Facility (IFMIF) was further optimised. The work focused on the design and the thermal hydraulic analysis of the HFTM section containing the material specimens to be irradiated, the ''test section'', with the main objective to improve the concept with respect to the optimum use of the available irradiation volume and to the temperature of the specimens. Particular emphasis was laid on the application of design principles which assure stable and reproducible thermal conditions. The present work has confirmed the feasibility and suitability of the optimised design of the HFTM test section with chocolate plate like shaped rigs. In particular it has been shown that the envisaged irradiation temperatures can be reached with acceptable temperature differences inside the specimen stack. The latter can be achieved only by additional electrical heating of the axial ends of the capsules. Division of the heater in three sections with separate power supply and control units is necessary. Maintaining of the temperatures during beam-off periods likewise requires electrical heating. The required electrical heaters - mineral isolated wires - are commercially available. The potential of the CFD code STAR-CD for the thermal hydraulic analysis of complex systems like the HFTM was confirmed. Nevertheless, experimental confirmation is desirable. Suitable experiments are under preparation. To verify the assumptions made on the thermal conductivity of the contact faces and layers between the two shells of the rig, dedicated experiments are suggested. The present work must be complemented by a thermal mechanical analysis of the module. Most critical component in this respect seems to be the rig wall. Furthermore, it will be necessary to investigate the response of the HFTM to power transients, and to determine the requirements

  1. Optimised design and thermal-hydraulic analysis of the IFMIF/HFTM test section

    International Nuclear Information System (INIS)

    Gordeev, S.; Heinzel, V.; Lang, K.H.; Moeslang, A.; Schleisiek, K.; Slobodtchouk, V.; Stratmanns, E.

    2003-10-01

    On the basis of previous concepts, analyses and experiments, the high flux test module (HFTM) for the International Fusion Materials Irradiation Facility (IFMIF) was further optimised. The work focused on the design and the thermal hydraulic analysis of the HFTM section containing the material specimens to be irradiated, the ''test section'', with the main objective to improve the concept with respect to the optimum use of the available irradiation volume and to the temperature of the specimens. Particular emphasis was laid on the application of design principles which assure stable and reproducible thermal conditions. The present work has confirmed the feasibility and suitability of the optimised design of the HFTM test section with chocolate plate like shaped rigs. In particular it has been shown that the envisaged irradiation temperatures can be reached with acceptable temperature differences inside the specimen stack. The latter can be achieved only by additional electrical heating of the axial ends of the capsules. Division of the heater in three sections with separate power supply and control units is necessary. Maintaining of the temperatures during beam-off periods likewise requires electrical heating. The required electrical heaters - mineral isolated wires - are commercially available. The potential of the CFD code STAR-CD for the thermal hydraulic analysis of complex systems like the HFTM was confirmed. Nevertheless, experimental confirmation is desirable. Suitable experiments are under preparation. To verify the assumptions made on the thermal conductivity of the contact faces and layers between the two shells of the rig, dedicated experiments are suggested. The present work must be complemented by a thermal mechanical analysis of the module. Most critical component in this respect seems to be the rig wall. Furthermore, it will be necessary to investigate the response of the HFTM to power transients, and to determine the requirements on the electrical

  2. Neutronics and thermal-hydraulics analysis of KUHFR

    Energy Technology Data Exchange (ETDEWEB)

    Woodruff, W L [Argonne National Laboratory, Argonne, IL (United States); Mishima, K [KURRI, Osaka (Japan)

    1983-08-01

    control rod worth with reduced enrichment has not yet determined, but only a small decrease in worth is expected. These BOL boron poisoned fuels are also used as the fresh fuel feed for the equilibrium fuel cycle studies contained in this report. The first three cases shown have matching cycle lengths in the equilibrium cycle, while the last case has a considerably longer cycle length. These results are similarly reflected in the 'Maximum Cycle Lengths' shown for unpoisoned BOL cores. Thus, the first three case can be considered comparable. The last case might be considered as an option for an extended cycle length design. The cycle length for this case is increased by about 21%. Obviously, by decreasing the uranium density in the fuel meat (to 2.7 g/cm{sup 3}), the cycle length for this design could be reduced to match that of the other cases. Thermal-hydraulic calculations have been carried out in order to study the safety aspects of the use of reduced enrichment uranium fuel for the KUHFR. The calculations were based on what is outlined in the Safety Analysis Report for the KUHFR and also the IAEA Guidebook for the RERTR program. Only a few combinations of hydraulic parameters have been tested because the reactor safety cannot be discussed without any nuclear physics considerations. For example, any variations in fuel coolant channels may change not only flow velocities but also power peaking factors which may affect the assessment of reactor safety. For this reason, the thermal-hydraulic calculations were carried out only for those specific cases on which neutronics analysis has been already performed. Low enriched uranium (LEU) cases are also included in this study as initial feasibility studies for potential conversion. The computer code PLTEMP has been developed to calculate the flow distribution in the core, fuel plate temperatures and DNB heat fluxes.

  3. Development of Mitsubishi high thermal performance grid 1 - CFD applicability for thermal hydraulic design

    International Nuclear Information System (INIS)

    Ikeda, K.; Hoshi, M.

    2001-01-01

    Mitsubishi applied the Computational Fluid Dynamics (CFD) evaluation method for designing of the new lower pressure loss and higher DNB performance grid spacer. Reduction of pressure loss of the grid has been estimated by CFD. Also, CFD has been developed as a design tool to predict the coolant mixing ability of vane structures, that is to compare the relative peak spot temperatures around fuel rods at the same heat flux condition. These evaluations have been reflected to the new grid spacer design. The prototype grid was manufactured and some flow tests were performed to examine the thermal hydraulic performance, which were predicted by CFD. The experimental data of pressure loss was in good agreement with CFD prediction. The CFD prediction of flow behaviors at downstream of the mixing vanes was verified by detail cross-flow measurements at rod gaps by the rod LDV system. It is concluded that the applicability of the CFD evaluation method for the thermal hydraulic design of the grid is confirmed. (authors)

  4. Visualization and measurement by image processing of thermal hydraulic phenomena by neutron radiography

    International Nuclear Information System (INIS)

    Takenaka, Nobuyuki

    1996-01-01

    Neutron Radiography was applied to visualization of thermal hydraulic phenomena and measurement was carried out by image processing the visualized images. Since attenuation of thermal neutron rays is high in ordinary liquids like water and organic fluid while it is low in most of metals, liquid flow behaviors can be visualized through a metallic wall by neutron radiography. Measurement of void fraction and flow vector field which is important to study thermal hydraulic phenomena can be carried out by image processing the images obtained by the visualization. Various two-phase and liquid metal flows were visualized by a JRR-3M thermal neutron radiography system in the present study. Multi-dimensional void fraction distributions in two-phase flows and flow vector fields in liquid metals, which are difficult to measure by the other methods, were successfully measured by image processing. It was shown that neutron radiography was efficiently applicable to study thermal hydraulic phenomena. (author)

  5. Thermal-hydraulics of the Loviisa reactor pressure vessel overcooling transients

    International Nuclear Information System (INIS)

    Tuomisto, Harri.

    1987-06-01

    In the Loviisa reactor pressure vessel safety analyses, the thermal-hydraulics of various overcooling transients has been evaluated to give pertinent initial data for fracture-mechanics calculations. The thermal-hydraulic simulations of the developed overcooling scenarios have been performed using best-estimate thermal-hydraulic computer codes. Experimental programs have been carried out to study phenomena related to natural circulation interruptions in the reactor coolant system. These experiments include buoyancy-induced phenomena such as thermal mixing and stratification of cold high-pressure safety injection water in the cold legs and the downcomer, and oscillations of the single-phase natural circulation. In the probabilistic pressurized thermal shock study, the Loviisa training simulator and the advanced system code RELAP5/MOD2 were utilized to simulate selected sequences. Flow stagnation cases were separately calculated with the REMIX computer program. The methods employed were assessed for these calculations against the plant data and own experiments

  6. Full vessel CFD analysis on thermal-hydraulic characteristics of CPR1000 PWR

    International Nuclear Information System (INIS)

    Chao Yanmeng; Yang Lixin; Zhang Mingqian

    2014-01-01

    To obtain flow distributions and thermal-hydraulic properties in a full vessel PWR under limited computation ability and time, a full vessel simulation model of CPR1000 was built based on two simplification methods. One simplified the inner geometry of the control rod guide tubes using equivalent flow area. Another substituted the core by a porous domain to maintain the pressure drop and temperature rise. After the computation, global and localized flow distributions, hydraulic loads of some main assemblies were obtained, as well as other thermal-hydraulic properties. The results indicate the flow distribution in the full vessel is asymmetrical. Therefore it is essential to use the full vessel model to simulate. The calculated thermal-hydraulic characteristics agree well with the operation statistics, providing the reference data for the reactor safety operation. (authors)

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

    International Nuclear Information System (INIS)

    Ville Lestinen; Timo Toppila

    2005-01-01

    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

  8. Thermal-hydraulic analysis of spent fuel storage systems

    International Nuclear Information System (INIS)

    Rector, D.R.; Wheeler, C.L.; Lombardo, N.J.

    1987-01-01

    This paper describes the COBRA-SFS (Spent Fuel Storage) computer code, which is designed to predict flow and temperature distributions in spent nuclear fuel storage and transportation systems. The decay heat generated by spent fuel in a dry storage cask is removed through a combination of conduction, natural convection, and thermal radiation. One major advantage of COBRA-SFS is that fluid recirculation within the cask is computed directly by solving the mass and momentum conservation equations. In addition, thermal radiation heat transfer is modeled using detailed radiation exchange factors based on quarter-rod segments. The equations governing mass, momentum, and energy conservation for incompressible flows are presented, and the semi-implicit solution method is described. COBRA-SFS predictions are compared to temperature data from a spent fuel storage cask test and the effect of different fill media on the cladding temperature distribution is discussed. The effect of spent fuel consolidation on cask thermal performance is also investigated. 16 refs., 6 figs., 2 tabs

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

    International Nuclear Information System (INIS)

    Hashim, Zaredah; Lanyau, Tonny Anak; Farid, Mohamad Fairus Abdul; Kassim, Mohammad Suhaimi; Azhar, Noraishah Syahirah

    2016-01-01

    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

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

    Science.gov (United States)

    Hashim, Zaredah; Lanyau, Tonny Anak; Farid, Mohamad Fairus Abdul; Kassim, Mohammad Suhaimi; Azhar, Noraishah Syahirah

    2016-01-01

    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.

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

    International Nuclear Information System (INIS)

    Hwnag, M.

    2001-04-01

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

    Science.gov (United States)

    Domalapally, Phani; Di Caro, Marco

    2018-05-01

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

  13. Sensitivity analysis of thermal hydraulic response in containment at core meltdown accident

    International Nuclear Information System (INIS)

    Kobayashi, Kensuke; Ishigami, Tsutomu; Horii, Hideo; Chiba, Takemi.

    1985-01-01

    A sensitivity analysis of thermal hydraulic response in a containment during a 'station blackout' (the loss of all AC power) accident at Browns Ferry unit one plant was performed with the computer code MARCH 1.0. In the analysis, the plant station batteries were assumed to be available for 4h after the initiation of the accident. The thermal hydraulic response in the containment was calculated by varying several input data for MARCH 1.0 independently and the deviation among calculated results were investigated. The sensitivity analysis showed that (a) the containment would fail due to the overtemperature without any operator actions for plant recovery, which would be strongly dependent on the model of the debris-concrete interaction and the input parameters for specifying the containment failure modes in MARCH 1.0, (b) a core melting temperature and an amount of water left in a primary system at the end of the meltdown were identified as important parameters which influenced the time of the containment failure, and (c) experimental works regarding the parameters mentioned above could be recommended. (author)

  14. Regulatory support activities of JNES by thermal-hydraulic and safety analyses

    International Nuclear Information System (INIS)

    Kasahara, Fumio

    2008-01-01

    Current status and some related topics on regulatory support activities of Japan Nuclear Energy Safety Organization (JNES) by thermal-hydraulic and safety analyses are reported. The safety of nuclear facilities is secured primarily by plant owners and operators. However, the regulatory body NISA (Nuclear and Industrial Safety Agency) has conducted a strict regulation to confirm the adequacy of the site condition as well as the basic and detailed design. The JNES has been conducting independent analyses from applicants (audit analyses, etc.) by direction of NISA and supporting its review. In addition to the audit analysis, thermal-hydraulic and safety analyses are used in such areas as analytical evaluation for investigation of causes of accidents and troubles, level 2 PSA for risk informed regulation, etc. Recent activities of audit analyses are for the application of Tsuruga 3 and 4 (APWR), the spent fuel storage facility for the establishment, and the LMFBR Monju for the core change. For the trouble event evaluation, the criticality accident analysis of Sika1 was carried out and the evaluation of effectiveness of accident management (AM) measure for Tomari 3 (PWR) and Monju was performed. The analytical codes for these evaluations are continuously revised by reflecting the state-of-art technical information and validated using the information provided by the data from JAEA, OECD project, etc. (author)

  15. 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.

  16. Thermal-hydraulic modeling of flow inversion in a research reactor

    International Nuclear Information System (INIS)

    Kazeminejad, H.

    2008-01-01

    The course of loss of flow accident and flow inversion in a pool type research reactor, with scram enabled under natural circulation condition is numerically investigated. The analyses were performed by a lumped parameters approach for the coupled kinetic-thermal-hydraulics, with continuous feedback due to coolant and fuel temperature effects. A modified Runge-Kutta method was adopted for a better solution to the set of stiff differential equations. Transient thermal-hydraulics during the process of flow inversion and establishment of natural circulation were considered for a 10-MW IAEA research reactor. Some important parameters such as the peak temperatures for the hot channel were obtained for both high-enriched and low enriched fuel. The model prediction is also verified through comparison with other computer code results reported in the literature for detailed simulations of loss of flow accidents (LOFA) and the agreement between the results for the peak clad temperatures and key parameters has been satisfactory. It was found that the flow inversion and subsequent establishment of natural circulation keep the peak cladding surface temperature below the saturation temperature to avoid the escalation of clad temperature to the level of onset of nucleate boiling and sub-cooled void formation to ensure the safe operation of the reactor

  17. Experimental investigation of the hydraulic characteristics of a counter flow wet cooling tower

    International Nuclear Information System (INIS)

    Lemouari, M.; Boumaza, M.; Kaabi, A.

    2011-01-01

    Thermal and nuclear electric power plants as well as several industrial processes invariably discharge considerable energy to their surrounding by heat transfer. Although water drawn from a nearby river or lake can be employed to carry away this energy, cooling towers offer an excellent alternative particularly in locations where sufficient cooling water cannot be easily obtained from natural sources or where concern for the environment imposes some limits on the temperature at which cooling water can be returned to the surrounding. This paper concerns an experimental investigation of the hydraulic characteristics of a counter flow wet cooling tower. The tower contains a 'VGA.' (Vertical Grid Apparatus) type packing which is 0.42 m high and consists of four (04) galvanised sheets having a zigzag form, between which are disposed three (03) metallic vertical grids in parallel with a cross sectional test area of 0.15 m x 0.148 m. The present investigation is focused mainly on the effect of the air and water flow rates on the hydraulic characteristics of the cooling tower, for different inlet water temperatures. The two hydrodynamic operating regimes which were observed during the air/water contact operation within the tower, namely the Pellicular Regime (PR) and the Bubble and Dispersion Regime (BDR) have enabled to distinguish two different states of pressure drop characteristics. The first regime is characterized by low pressure drop values, while in the second regime, the pressure drop values are relatively much higher than those observed in the first one. The dependence between the pressure drop characteristics and the combined heat and mass transport (air-water) through the packing inside the cooling tower is also highlighted. The obtained results indicate that this type of tower possesses relatively good hydraulic characteristics. This leads to the saving of energy. -- Highlights: → Cooling towers are widely used to reject waste heat from thermal and nuclear

  18. Thermal-hydraulic analysis of an annular fuel element: The Achilles' heel of the particle bed reactor

    International Nuclear Information System (INIS)

    Dibben, M.J.; Tuttle, R.F.

    1993-01-01

    The low pressure nuclear thermal propulsion (LPNTP) concept offers significant improvements in rocket engine specific impulse over rockets employment chemical propulsion. This study investigated a parametric thermal-hydraulic analysis of an annular fueld element, also referred to as a fuel pipe, using the computer code ATHENA (Advanced Thermal Hydraulic Energy Network Analyzer). The fuelpipe is an annular particle bed fuel element of the reactor with radially inward flow of hydrogen through the element. In this study, the outlet temperature of the hydrogen is parametrically related to key effects, including the reactor power at two different pressure drops, the effect of power coupling for in-core testing, and the effect of hydrogen flow rates. Results show that the temperature is linearly related to the reactor power, but not to pressure drop, and that cross flow inside the fuelpipe occurs at approximately 0.3 percent of the radial flow rates

  19. FY 1995 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.; McDuffee, J.L.; McFee, M.T.; Ruggles, A.E.; Wendel, M.W.; Yoder, G.L.

    1997-07-01

    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). 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. Special consideration was given to allow operation of the system in a stiff mode (constant flow) and in a soft mode (constant pressure drop) for proper implementation of true FE and DNB experiments. The facility is also designed to examine other T/H phenomena, including onset of incipient boiling (IB), single-phase heat transfer coefficients and friction factors, and two-phase heat transfer and pressure drop characteristics. Tests will also be conducted that are representative of decay heat levels at both high pressure and low pressure as well as other quasi-equilibrium conditions encountered during transient scenarios. A total of 22 FE tests and 2 CHF tests were performed during FY 1994 and FY 1995 with water flowing vertically upward. Comparison of these data as well as extensive data from other investigators led to a proposed modification to the Saha and Zuber correlation for onset of significant void (OSV), applied to FE prediction. The modification takes into account a demonstrated dependence of the OSV or FE thermal limits on subcooling levels, especially in the low subcooling regime.

  20. FY 1995 progress report on the ANS thermal-hydraulic test loop operation and results

    International Nuclear Information System (INIS)

    Siman-Tov, M.; Felde, D.K.; Farquharson, G.; McDuffee, J.L.; McFee, M.T.; Ruggles, A.E.; Wendel, M.W.; Yoder, G.L.

    1997-07-01

    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). 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. Special consideration was given to allow operation of the system in a stiff mode (constant flow) and in a soft mode (constant pressure drop) for proper implementation of true FE and DNB experiments. The facility is also designed to examine other T/H phenomena, including onset of incipient boiling (IB), single-phase heat transfer coefficients and friction factors, and two-phase heat transfer and pressure drop characteristics. Tests will also be conducted that are representative of decay heat levels at both high pressure and low pressure as well as other quasi-equilibrium conditions encountered during transient scenarios. A total of 22 FE tests and 2 CHF tests were performed during FY 1994 and FY 1995 with water flowing vertically upward. Comparison of these data as well as extensive data from other investigators led to a proposed modification to the Saha and Zuber correlation for onset of significant void (OSV), applied to FE prediction. The modification takes into account a demonstrated dependence of the OSV or FE thermal limits on subcooling levels, especially in the low subcooling regime

  1. FFTF thermal-hydraulic testing results affecting piping and vessel component design in LMFBR's

    International Nuclear Information System (INIS)

    Stover, R.L.; Beaver, T.R.; Chang, S.C.

    1983-01-01

    The Fast Flux Test Facility completed four years of pre-operational testing in April 1982. This paper describes thermal-hydraulic testing results from this period which impact piping and vessel component design in LMFBRs. Data discussed are piping flow oscillations, piping thermal stratification and vessel upper plenum stratification. Results from testing verified that plant design limits were met

  2. COBRA-3M: a digital computer code for analyzing thermal-hydraulic behavior in pin bundles

    International Nuclear Information System (INIS)

    Marr, W.W.

    1975-03-01

    The COBRA-3M computer program is a modification of the thermal-hydraulic subchannel-analysis program COBRA-III. It includes detailed thermal models of fuel pin and duct wall. It is especially suitable for analyzing small pin bundles used in in-reactor or out-of-reactor experiments. (U.S.)

  3. Comparison for the interfacial and wall friction models in thermal-hydraulic system analysis codes

    International Nuclear Information System (INIS)

    Hwang, Moon Kyu; Park, Jee Won; Chung, Bub Dong; Kim, Soo Hyung; Kim, See Dal

    2007-07-01

    The average equations employed in the current thermal hydraulic analysis codes need to be closed with the appropriate models and correlations to specify the interphase phenomena along with fluid/structure interactions. This includes both thermal and mechanical interactions. Among the closure laws, an interfacial and wall frictions, which are included in the momentum equations, not only affect pressure drops along the fluid flow, but also have great effects for the numerical stability of the codes. In this study, the interfacial and wall frictions are reviewed for the commonly applied thermal-hydraulic system analysis codes, i.e. RELAP5-3D, MARS-3D, TRAC-M, and CATHARE

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

    International Nuclear Information System (INIS)

    Mur, J.; Meignin, J.C.

    1997-07-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1997-07-01

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

  6. Parametric study on thermal-hydraulic characteristics of high conversion light water reactor

    International Nuclear Information System (INIS)

    Mori, Takamasa; Nakagawa, Masayuki; Fujii, Sadao.

    1988-11-01

    To assess the feasibility of high conversion light water reactors (HCLWRs) from the thermal-hydraulic viewpoint, parametric study on thermal-hydraulic characteristics of HCLWR has been carried out by using a unit cell model. It is assumed that a HCLWR core is contained in a current 1000 MWe PWR plant. At the present study, reactor core parameters such as fuel pin diameter, pitch, core height and linear heat rate are widely and parametrically changed to survey the relation between these parameters and the basic thermal-hydraulic characteristics, i.e. maximum fuel temperature, minimum DNBR, reduction of reactor thermal output and so on. The validity of the unit cell model used has been ensured by comparison with the result of a subchannel analysis carried out for a whole core. (author)

  7. Thermal-hydraulic of partially blocked fuel subassembly with porous media

    International Nuclear Information System (INIS)

    Nagata, Takemitsu; Ohshima, Hiroyuki

    2000-10-01

    The analysis code for investigations of local subassembly phenomena, which has been recognized as an issue of local subassembly accidents, has been required and developed at JNC. It is desirable for the analysis code to be applicable to various blockage conditions and random position of the blockage formation and to evaluate conservatively on the safety assessment with high accuracy. In this study, for the purpose of verifying the application and issues of the subchannel analysis code ASFRE-IV which evaluates thermal hydraulic phenomena in the porous blockage regions, the ASFRE-IV validation analysis was carried out on the basis of the data of an experiment investigation on a local porous blockage in a fuel subassembly performed by Reactor Engineering Groop, O-arai Engineering Center, JNC. Calculational results indicated that ASFRE-IV could reproduce the coolant temperature profile in a fuel subassembly and the peak temperature in the local subchannel conservatively. (author)

  8. 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.

  9. CCP Sensitivity Analysis by Variation of Thermal-Hydraulic Parameters of Wolsong-3, 4

    Energy Technology Data Exchange (ETDEWEB)

    You, Sung Chang [KHNP, Daejeon (Korea, Republic of)

    2016-10-15

    The PHWRs are tendency that ROPT(Regional Overpower Protection Trip) setpoint is decreased with reduction of CCP(Critical Channel Power) due to aging effects. For this reason, Wolsong unit 3 and 4 has been operated less than 100% power due to the result of ROPT setpoint evaluation. Typically CCP for ROPT evaluation is derived at 100% PHTS(Primary Heat Transport System) boundary conditions - inlet header temperature, header to header different pressure and outlet header pressure. Therefore boundary conditions at 100% power were estimated to calculate the thermal-hydraulic model at 100% power condition. Actually thermal-hydraulic boundary condition data for Wolsong-3 and 4 cannot be taken at 100% power condition at aged reactor condition. Therefore, to create a single-phase thermal-hydraulic model with 80% data, the validity of the model was confirmed at 93.8%(W3), 94.2%(W4, in the two-phase). And thermal-hydraulic boundary conditions at 100% power were calculated to use this model. For this reason, the sensitivities by varying thermal-hydraulic parameters for CCP calculation were evaluated for Wolsong unit 3 and 4. For confirming the uncertainties by variation PHTS model, sensitivity calculations were performed by varying of pressure tube roughness, orifice degradation factor and SG fouling factor, etc. In conclusion, sensitivity calculation results were very similar and the linearity was constant.

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

    Directory of Open Access Journals (Sweden)

    L. Batet

    2007-11-01

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

  11. European liquid metal thermal-hydraulics R and D: present and future

    International Nuclear Information System (INIS)

    Roelofs, F.; Batta, A.; Bandini, G.; Van Tichelen, K.; Gerschenfeld, A.; Cheng, X.

    2014-01-01

    A large role is attributed in the future within the European Sustainable Nuclear Energy Technology Platform (SNE-TP) and especially the underlying European Sustainable Nuclear Industry Initiative (ESNII) to the application of fast reactors for sustainable nuclear energy production. Specifically, fast reactors are considered attractive because of their possibility to use natural resources efficiently and to reduce the volume and lifetime of nuclear waste. Currently four demonstration projects have a promising outlook in Europe, i.e. the ASTRID project in France, the MYRRHA project in Belgium, the ALFRED project developed in Europe and to be built in Romania, and the ELECTRA project in Sweden. Sodium and lead(-alloys) are envisaged as coolants for these reactors. Obviously, in the development of these reactors, thermal-hydraulics is recognized as a key challenge with emphasis on safety issues. This paper will discuss the present development status of liquid metal cooled reactor thermal-hydraulics as an outcome of the European 7. framework programme THINS (Thermal-Hydraulics for Innovative Nuclear Systems) project. The main project results with respect to liquid metal cooled reactors will be summarized, i.e. turbulence heat transfer model development, fuel assembly analysis, pool thermal-hydraulics, system behaviour, multi-phase physics, and multiscale thermal-hydraulics simulation. In conclusion, the main challenges for future developments will be indicated. Emphasis will be put on the important experimental and numerical challenges. (authors)

  12. PWR blowdown heat transfer separate-effects program: Thermal-Hydraulic Test Facility experimental data report for test 100

    International Nuclear Information System (INIS)

    White, M.D.; Hedrick, R.A.

    1977-01-01

    Reduced instrument responses are presented for Thermal-Hydraulic Test Facility (THTF) test 100, which is part of the ORNL Pressurized-Water-Reactor (PWR) Blowdown Heat Transfer Separate-Effects Program. The objective of the program is to investigate the thermal-hydraulic phenomenon governing the energy transfer and transport processes that occur during a loss-of-coolant accident in a PWR system. Test 100 was conducted to investigate the response of heater rod bundle 1 and instrumented spool pieces with flow homogenizing screens to a double-ended rupture with equal break areas at the test section inlet and outlet. The primary purpose of this report is to make the reduced instrument responses during test 100 available. The responses are presented in graphical form in engineering units and have been analyzed only to the extent necessary to assure reasonableness and consistency

  13. Thermal modeling of a hydraulic hybrid vehicle transmission based on thermodynamic analysis

    International Nuclear Information System (INIS)

    Kwon, Hyukjoon; Sprengel, Michael; Ivantysynova, Monika

    2016-01-01

    Hybrid vehicles have become a popular alternative to conventional powertrain architectures by offering improved fuel efficiency along with a range of environmental benefits. Hydraulic Hybrid Vehicles (HHV) offer one approach to hybridization with many benefits over competing technologies. Among these benefits are lower component costs, more environmentally friendly construction materials, and the ability to recover a greater quantity of energy during regenerative braking which make HHVs partially well suited to urban environments. In order to further the knowledge base regarding HHVs, this paper explores the thermodynamic characteristics of such a system. A system model is detailed for both the hydraulic and thermal components of a closed circuit hydraulic hybrid transmission following the FTP-72 driving cycle. Among the new techniques proposed in this paper is a novel method for capturing rapid thermal transients. This paper concludes by comparing the results of this model with experimental data gathered on a Hardware-in-the-Loop (HIL) transmission dynamometer possessing the same architecture, components, and driving cycle used within the simulation model. This approach can be used for several applications such as thermal stability analysis of HHVs, optimal thermal management, and analysis of the system's thermodynamic efficiency. - Highlights: • Thermal modeling for HHVs is introduced. • A model for the hydraulic and thermal system is developed for HHVs. • A novel method for capturing rapid thermal transients is proposed. • The thermodynamic system diagram of a series HHV is predicted.

  14. Effects of non-linearity of material properties on the coupled mechanical-hydraulic-thermal behavior in rock mass

    International Nuclear Information System (INIS)

    Kobayashi, Akira; Ohnishi, Yuzo

    1986-01-01

    The nonlinearity of material properties used in the coupled mechanical-hydraulic-thermal analysis is investigated from the past literatures. Some nonlinearity that is respectively effective for the system is introduced into our computer code for analysis such a coupling problem by using finite element method. And the effects of nonlinearity of each material property on the coupled behavior in rock mass are examined for simple model and Stripa project model with the computer code. (author)

  15. Neutronics - thermal-hydraulics coupling: application to the helium-cooled fast reactor

    International Nuclear Information System (INIS)

    Vaiana, F.

    2009-11-01

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

  16. CFD analysis of thermal-hydraulic behavior in SCWR typical flow channels

    International Nuclear Information System (INIS)

    Gu, H.Y.; Cheng, X.; Yang, Y.H.

    2008-01-01

    Investigations on thermal-hydraulic behavior in SCWR fuel assembly have obtained a significant attention in the international SCWR community. However, there is still a lack of understanding and ability to predict the heat transfer behavior of supercritical water. In this paper, CFD analysis is carried out to study the flow and heat transfer behavior of supercritical water in sub-channels of both square and triangular rod bundles. Effect of various parameters, e.g. thermal boundary conditions and pitch-to-diameter ratio on the thermal-hydraulic behavior is investigated. Two boundary conditions, i.e., constant heat flux at the outer surface of cladding and constant heat density in the fuel pin are applied. The results show that the structure of the secondary flow mainly depends on the rod bundle configuration as well as the pitch-to-diameter ratio, whereas, the amplitude of the secondary flow is affected by the thermal boundary conditions, as well. The secondary flow is much stronger in a square lattice than that in a triangular lattice. The turbulence behavior is similar in both square and triangular lattices. The dependence of the amplitude of the turbulent velocity fluctuation across the gap on Reynolds number becomes prominent in both lattices as the pitch-to-diameter ratio increases. The effect of thermal boundary conditions on turbulent velocity fluctuation is negligibly small. For both lattices with small pitch-to-diameter ratios (P/D < 1.3), the mixing coefficient is about 0.022. Both secondary flow and turbulent mixing show unusual behavior in the vicinity of the pseudo-critical point. Further investigation is needed. A strong circumferential non-uniformity of wall temperature and heat transfer is observed in tight lattices at constant heat flux boundary conditions, especially in square lattices. In the case with constant heat density of fuel pin, the circumferential conductive heat transfer significantly reduces the non-uniformity of circumferential

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2013-04-15

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

  18. 11. international topical meeting on nuclear reactor thermal-hydraulics (NURETH-11)

    International Nuclear Information System (INIS)

    Lemonnier, H.

    2005-01-01

    The main topics covered by the NURETH 11 meeting are the thermal-hydraulics of existing and future nuclear power plants as foreseen by the Generation IV worldwide initiative. Normal operation and accidental situations are also relevant topics of the Conference. The topics cover modeling, experiments, instrumentation and numerical simulations related to flow and heat transfer in nuclear reactors with a special emphasis on the advances of multiphase CFD methods. The first part of this Book of Abstracts enumerates the Organizing Scientific Societies, the Sponsors of the Conference, the Conference Chairs, and the members of the Steering Committee and of the Technical Program Committee. The second part of this Book of Abstracts contains the list of the titles of the contributed papers. Each item includes the log number of the paper, the abstract of which can therefore be easily located in the next section of this book. The titles of the papers have been sorted out by topics to provide a synthetic view of the contributions in a selected domain. The last section of this Book includes an index of authors and co-authors with a reference to the log number(s) of their contributed paper(s). Finally, the CD-Rom of the Conference Proceedings containing the full-length papers is inserted at the inside back cover. Sessions content: A - two-phase flow and heat transfer fundamentals: computational and mathematical techniques (numerical schemes, LBM, BEM, mesh-less, etc.); contact angle and wettability phenomena; experiments and data bases for the assessment and the verification of 3D models; flow regime identification and modelling; heat transfer near critical pressure and supercritical water reactors; interfacial area (data base, modeling, measurement techniques); instrumentation techniques; micro-scale basic phenomena, fluid flow and heat transfer; scaling methods; counter current flow; B - code developments: containment analysis; core thermal-hydraulics and subchannel analysis

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

    International Nuclear Information System (INIS)

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

    2011-01-01

    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.

  20. Transient analysis and thermal hydraulic margins of GHARR-1 using the PARET/NAL code

    International Nuclear Information System (INIS)

    Adoo, N.A.

    2009-06-01

    The PARET code has been adapted by the IAEA for testing transient behaviour in research reactors. The PARET code provides a coupled thermal hydrodynamic and point kinetics capability with a continuous reactivity feedback and an optional voiding model that estimates the voiding produced by the subcooled boiling. The present version of the PARET/ANL 73 code provides a convenient means of assessing the various models and correlations proposed for the use in the analysis of research reactor behaviour. The Monte Carlo N-Particle code (MCNP) has been used to obtain power peaking profile for a two channel PARET/ANL model. A PARET model with the corresponding neutronics and thermal hydraulic characteristics for the miniature neutron source reactor (MNSR) has been used to simulate reactivity accidents for the Ghana Research Reactor - 1(GHARR-1) under the MNSR operation conditions of natural circulation, normal operation and reactivity insertion accidents. The simulation results via the insertion of large reactivity demonstrated the high inherent safety features of the MNSR for which the high negative reactivity feedback of moderator temperature limits power excursion and avoids consequently the escalation of clad temperature to the level of onset of sub-cooled void formation. The hot channel peaking factors for both radial and axial were found to be 1.17 and 1.44 respectively. Thermal hydraulic performance characteristics were investigated and the safety margins determined. The peak clad and coolant temperatures ranged from 59.18 0 C to 106.75 0 C and 42.95 0 C to 178.44 0 C respectively at which nucleate boiling will occur within the flow channels of the core. (au)

  1. Test results of the new NSSS thermal-hydraulics program of the KNPEC-2 simulator

    International Nuclear Information System (INIS)

    Jeong, J. Z.; Kim, K. D.; Lee, M. S.; Hong, J. H.; Lee, Y. K.; Seo, J. S.; Kweon, K. J.; Lee, S. W.

    2001-01-01

    As a part of the KNPEC-2 Simulator Upgrade Project, KEPRI and KAERI have developed a new NSSS thermal-hydraulics program, which is based on the best-estimate system code, RETRAN. The RETRAN code was originally developed for realistic simulation of thermal-hydraulic transient in power plant systems. The capability of 'real-time simulation' and robustness' should be first developed before being implemented in full-scope simulators. For this purpose, we have modified the RETRAN code by (i) eliminating the correlations' discontinuities between flow regime maps, (ii) simplifying physical correlations, (iii) correcting errors in the original program, and (iv) others. This paper briefly presents the test results fo the new NSSS thermal-hydraulics program

  2. Thermal-hydraulic analysis of PWR small assembly for irradiation test of CARR

    International Nuclear Information System (INIS)

    Yin Hao; Zou Yao; Liu Xingmin

    2015-01-01

    The thermal-hydraulic behaviors of the PWR 4 × 4 small assembly tested in the high temperature and high pressure loop of China Advanced Research Reactor were analyzed. The CFD method was used to carry out 3D simulation of the model, thus detailed thermal-hydraulic parameters were obtained. Firstly, the simplified model was simulated to give the 3D temperature and velocity distributions and analyze the heat transfer process. Then the whole scale small assembly model was simulated and the simulation results were compared with those of simplified rod bundle model. Its flow behavior was studied and flow mixing characteristics of the grids were analyzed, and the mixing factor of the grid was calculated and can be used for further thermal-hydraulic study. It is shown that the highest temperature of the fuel rod meets the design limit and the mixing effect of the grid is obvious. (authors)

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

    International Nuclear Information System (INIS)

    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

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

    Science.gov (United States)

    Jansen, Gunnar; Miller, Stephen A.

    2017-04-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2000-03-01

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

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

    International Nuclear Information System (INIS)

    Ferandes, Gustavo H.N.; Ramos, Mário C.; Carvalho, Athos M.S.S.; Cabrera, Carlos E.V.; Costa, Antonella L.; Pereira, Claubia

    2017-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2017-07-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Geffray, Clotaire Clement

    2017-03-20

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

  9. The status of studies on fast reactor core thermal hydraulics at PNC

    International Nuclear Information System (INIS)

    Nishimura, M.; Ohshima, H.; Kamide, H.; Yamaguchi, K.; Yamaguchi, A.

    2000-01-01

    An outlook was addressed on investigative activities of the fast reactor core thermal-hydraulics at Power Reactor and Nuclear Fuel Development Corporation. Firstly, a computational modeling to predict flow field under natural circulation decay heat removal condition using multi-dimensional codes and its validation were presented. The validation was carried out through calculations of sodium experiments on an inter-subassembly heat transfer, a transient from forced to natural circulation and an inter-wrapper flow. Secondly, experimental and computational studies were expressed on local blockage with porous media in a fuel subassembly. Lastly, information was presented on an advanced computational code based on a subchannel analysis code. The code is under the development and extended to perform whole core simulation. (author)

  10. Coupled 3D neutronics/thermal hydraulics modeling of the SAFARI-1 MTR

    International Nuclear Information System (INIS)

    Rosenkrantz, Adam; Avramova, Maria; Ivanov, Kostadin; Prinsloo, Rian; Botes, Danniëll; Elsakhawy, Khalid

    2014-01-01

    Highlights: • Development of 3D coupled neutronics/thermal–hydraulic model of SAFARI-1. • Verification of 3D steady-state NEM based neutronics model for SAFARI-1. • Verification of 3D COBRA-TF based thermal–hydraulic model of SAFARI-1. • Quantification of the effect of correct modeling of thermal–hydraulic feedback. - Abstract: The purpose of this study was to develop a coupled accurate multi-physics model of the SAFARI-1 Material Testing Reactor (MTR), a facility that is used for both research and the production of medical isotopes. The model was developed as part of the SAFARI-1 benchmarking project as a cooperative effort between the Pennsylvania State University (PSU) and the South African Nuclear Energy Corporation (Necsa). It was created using a multi-physics coupling of state of the art nuclear reactor simulation tools, consisting of a neutronics code and a thermal hydraulics code. The neutronics tool used was the PSU code NEM, and the results from this component were verified using the Necsa neutronics code OSCAR-4, which is utilized for SAFARI-1 core design and fuel management. On average, the multiplication factors of the neutronics models agreed to within 5 pcm and the radial assembly-averaged powers agreed to within 0.2%. The thermal hydraulics tool used was the PSU version of COBRA-TF (CTF) sub-channel code, and the results of this component were verified against another thermal hydraulics code, the RELAP5-3D system code, used at Necsa for thermal–hydraulics analysis of SAFARI-1. Although only assembly-averaged results from RELAP5-3D were available, they fell within the range of values for the corresponding assemblies in the comprehensive CTF solution. This comparison allows for the first time to perform a quantification of steady-state errors for a low-powered MTR with an advanced thermal–hydraulic code such as CTF on a per-channel basis as compared to simpler and coarser-mesh RELAP5-3D modeling. Additionally, a new cross section

  11. European activities on crosscutting thermal-hydraulic phenomena for innovative nuclear systems

    Energy Technology Data Exchange (ETDEWEB)

    Cheng, X., E-mail: xu.cheng@kit.edu [Karlsruhe Institute of Technology (KIT) (Germany); Batta, A. [Karlsruhe Institute of Technology (KIT) (Germany); Bandini, G. [Italian National Agency for New Technologies, Energy and Sustainable Economic Development (ENEA) (Italy); Roelofs, F. [Nuclear Research and Consultancy Group (NRG) (Netherlands); Van Tichelen, K. [Studiecentrum voor Kernenergie – Centre d’étude de l’Energie Nucléaire (SCK-CEN) (Belgium); Gerschenfeld, A. [Commissariat à l’Energie Atomique (CEA) (France); Prasser, M. [Paul Scherrer Institute (PSI) (Switzerland); Papukchiev, A. [Gesellschaft für Anlagen- und Reaktorsicherheit mbH (GRS) (Germany); Hampel, U. [Helmholtz-Zentrum Dresden-Rossendorf e.V. (HZDR) (Germany); Ma, W.M. [Kungliga Tekniska Högskolan (KTH) (Sweden)

    2015-08-15

    Highlights: • This paper serves as a guidance of the special issue. • The technical tasks and methodologies applied to achieve the objectives have been described. • Main results achieved so far are summarized. - Abstract: Thermal-hydraulics is recognized as a key scientific subject in the development of innovative reactor systems. In Europe, a consortium is established consisting of 24 institutions of universities, research centers and nuclear industries with the main objectives to identify and to perform research activities on important crosscutting thermal-hydraulic issues encountered in various innovative nuclear systems. For this purpose the large-scale integrated research project THINS (Thermal-Hydraulics of Innovative Nuclear Systems) is launched in the 7th Framework Programme FP7 of European Union. The main topics considered in the THINS project are (a) advanced reactor core thermal-hydraulics, (b) single phase mixed convection, (c) single phase turbulence, (d) multiphase flow, and (e) numerical code coupling and qualification. The main objectives of the project are: • Generation of a data base for the development and validation of new models and codes describing the selected crosscutting thermal-hydraulic phenomena. • Development of new physical models and modeling approaches for more accurate description of the crosscutting thermal-hydraulic phenomena. • Improvement of the numerical engineering tools for the design analysis of the innovative nuclear systems. This paper describes the technical tasks and methodologies applied to achieve the objectives. Main results achieved so far are summarized. This paper serves also as a guidance of this special issue.

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

    International Nuclear Information System (INIS)

    Alves, Sabrina P.; Mesquita, Amir Z.; Rezende, Hugo C.; Palma, Daniel A.P.

    2017-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2017-07-01

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

  14. The status of thermal-hydraulic studies on the decay heat removal by natural convection using RAMONA and NEPTUN models

    International Nuclear Information System (INIS)

    Hoffmann, H.; Hain, K.; Marten, K.; Rust, K.; Weinberg, D.; Ohira, H.

    2004-01-01

    Thermal-hydraulic experiments were performed with water in order to simulate the decay heat removal by natural convection in a pool-type sodium-cooled reactor. Two test rigs of different scales were used, namely RAMONA (1:20) and NEPTUN (1:5). RAMONA served to study the transition from nominal operation by forced convection to decay heat removal operation by natural convection. Steady-state similarity tests were carried out in both facilities. The investigations cover nominal and non-nominal operation conditions. These data provide a broad basis for the verification of computer programs. Numerical analyses performed with the three-dimensional FLUTAN code indicated that the thermal-hydraulic processes can be quantitatively simulated even for the very complex geometry of the NEPTUN test rig. (author)

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

    International Nuclear Information System (INIS)

    Davis, C.B.; Shieh, A. S.

    2000-01-01

    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

  16. 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.

  17. Thermal hydraulic test of advanced fuel bundle with spectral shift rod (SSR) for BWR. Effect of thermal hydraulic parameters on steady state characteristics

    International Nuclear Information System (INIS)

    Kondo, Takao; Kitou, Kazuaki; Chaki, Masao; Ohga, Yukiharu; Makigami, Takeshi

    2011-01-01

    Japanese national project of next generation light water reactor (LWR) development started in 2008. Under this project, spectral shift rod (SSR) is being developed. SSR, which replaces conventional water rod (WR) of boiling water reactor (BWR) fuel bundle, was invented to enhance the BWR's merit, spectral shift effect for uranium saving. In SSR, water boils by neutron and gamma-ray direct heating and water level is formed as a boundary of the upper steam region and the lower water region. This SSR water level can be controlled by core flow rate, which amplifies the change of average core void fraction, resulting in the amplified spectral shift effect. This paper presents the steady state test results of the base geometry case in SSR thermal hydraulic test, which was conducted under the national project of next generation LWR. In the test, thermal hydraulic parameters, such as flow rate, pressure, inlet subcooling and heater rod power are changed to evaluate these effects on SSR water level and other SSR characteristics. In the test results, SSR water level rose as flow rate rose, which showed controllability of SSR water level by flow rate. The sensitivities of other thermal hydraulic parameters on SSR water level were also evaluated. The obtained data of parameter's sensitivities is various enough for the further analytical evaluation. The fluctuation of SSR water level was also measured to be small enough. As a result, it was confirmed that SSR's steady state performance was as planned and that SSR design concept is feasible. (author)

  18. Simulation of Thermal-hydraulic Process in Reactor of HTR-PM

    International Nuclear Information System (INIS)

    Zhou Kefeng; Zhou Yangping; Sui Zhe; Ma Yuanle

    2014-01-01

    This paper provides the physical process in the reactor of High Temperature Gas-cooled Reactor Pebble-bed Module (HTR-PM) and introduces the standard operation conditions. The FORTRAN code developed for the thermal hydraulic module of Full-Scale Simulator (FSS) of HTR-PM is used to simulate two typical operation transients including cold startup process and cold shutdown process. And the results were compared to the safety analysis code, namely TINTE. The good agreement indicates that the code is applicable for simulating the thermal-hydraulic process in reactor of HTR-PM. And for long time transient process, the code shows good stability and convergence. (author)

  19. TRAC-B thermal-hydraulic analysis of the Black Fox boiling water reactor

    International Nuclear Information System (INIS)

    Martin, R.P.

    1993-05-01

    Thermal-hydraulic analyses of six hypothetical accident scenarios for the General Electric Black Fox Nuclear Project boiling water reactor were performed using the TRAC-BF1 computer code. This work is sponsored by the US Nuclear Regulatory Commission and is being done in conjunction with future analysis work at the US Nuclear Regulatory Commission Technical Training Center in Chattanooga, Tennessee. These accident scenarios were chosen to assess and benchmark the thermal-hydraulic capabilities of the Black Fox Nuclear Project simulator at the Technical Training Center to model abnormal transient conditions

  20. Light-water-reactor coupled neutronic and thermal-hydraulic codes

    International Nuclear Information System (INIS)

    Diamond, D.J.

    1982-01-01

    An overview is presented of computer codes that model light water reactor cores with coupled neutronics and thermal-hydraulics. This includes codes for transient analysis and codes for steady state analysis which include fuel depletion and fission product buildup. Applications in nuclear design, reactor operations and safety analysis are given and the major codes in use in the USA are identified. The neutronic and thermal-hydraulic methodologies and other code features are outlined for three steady state codes (PDQ7, NODE-P/B and SIMULATE) and four dynamic codes (BNL-TWIGL, MEKIN, RAMONA-3B, RETRAN-02). Speculation as to future trends with such codes is also presented

  1. 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.

  2. CHF predictor derived from a 3D thermal-hydraulic code and an advanced statistical method

    International Nuclear Information System (INIS)

    Banner, D.; Aubry, S.

    2004-01-01

    A rod bundle CHF predictor has been determined by using a 3D code (THYC) to compute local thermal-hydraulic conditions at the boiling crisis location. These local parameters have been correlated to the critical heat flux by using an advanced statistical method based on spline functions. The main characteristics of the predictor are presented in conjunction with a detailed analysis of predictions (P/M ratio) in order to prove that the usual safety methodology can be applied with such a predictor. A thermal-hydraulic design criterion is obtained (1.13) and the predictor is compared with the WRB-1 correlation. (author)

  3. 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.

  4. Thermal - hydraulic analysis of pressurizer water reactors using the model of open lateral boundary

    International Nuclear Information System (INIS)

    Borges, R.C.

    1980-10-01

    A computational method is developed for thermal-hydraulic analysis, where the channel may be analysed by more than one independent steps of calculation. This is made possible by the incorporation of the model of open lateral boundary in the code COBRA-IIIP, which permits the determination of the subchannel of an open lattice PWR core in a multi-step calculation. The thermal-hydraulic code COBRA-IIIP, developed at the Massachusetts Institute of Technology, is used as the basic model for this study. (Author) [pt

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2008-02-15

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

  6. Perturbative methods applied for sensitive coefficients calculations in thermal-hydraulic systems

    International Nuclear Information System (INIS)

    Andrade Lima, F.R. de

    1993-01-01

    The differential formalism and the Generalized Perturbation Theory (GPT) are applied to sensitivity analysis of thermal-hydraulics problems related to pressurized water reactor cores. The equations describing the thermal-hydraulic behavior of these reactors cores, used in COBRA-IV-I code, are conveniently written. The importance function related to the response of interest and the sensitivity coefficient of this response with respect to various selected parameters are obtained by using Differential and Generalized Perturbation Theory. The comparison among the results obtained with the application of these perturbative methods and those obtained directly with the model developed in COBRA-IV-I code shows a very good agreement. (author)

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

    International Nuclear Information System (INIS)

    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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2017-02-15

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

  9. Preliminary thermal-hydraulic and safety analysis of China DFLL-TBM system

    Energy Technology Data Exchange (ETDEWEB)

    Li, Wei [School of Nuclear Science and Technology, Xi’an Jiaotong University, No. 28, Xianning West Road, Xi’an, Shanxi 710049 (China); Tian, Wenxi, E-mail: wxtian@mail.xjtu.edu.cn [School of Nuclear Science and Technology, Xi’an Jiaotong University, No. 28, Xianning West Road, Xi’an, Shanxi 710049 (China); Qiu, Suizheng; Su, Guanghui; Jiao, Hong [School of Nuclear Science and Technology, Xi’an Jiaotong University, No. 28, Xianning West Road, Xi’an, Shanxi 710049 (China); Bai, Yunqing; Chen, Hongli [Institute of Nuclear Energy Safety Technology, Chinese Academy of Sciences, Hefei, Anhui 230031 (China); Wu, Yican, E-mail: yican.Wu@Fds.Org.Cn [Institute of Nuclear Energy Safety Technology, Chinese Academy of Sciences, Hefei, Anhui 230031 (China)

    2013-06-15

    Highlights: • Thermal-hydraulic and safety analysis on DFLL-TBM system is performed. • The TBM FW maximum temperature is 541 °C under steady state condition. • The TBM FW maximum temperature does not exceed the melt point of CLAM steel 1500 °C. • Neither the VV pressurization nor vault pressure build-up goes beyond 0.2 MPa. -- Abstract: China has proposed the dual-functional lithium-lead (DFLL) tritium breeding blanket concept for testing in ITER as a test blanket module (TBM), to demonstrate the technologies of tritium self-sufficiency, high-grade heat extraction and efficient electricity production which are needed for DEMO and fusion power plant. Safety assessment of the TBM and its auxiliary system should be conducted to deal with ITER safety issues directly caused by the TBM system failure during the design process. In this work, three potential initial events (PIEs) – in-vessel loss of helium (He) coolant and ex-vessel loss of He coolant and loss of flow without scram (LOFWS) – were analyzed for the TBM system with a modified version of the RELAP5/MOD3 code containing liquid lithium-lead eutectic (LiPb). The code also comprised an empirical expression for MHD pressure drop relevant to three-dimensional (3D) effect, the Lubarsky–Kaufman convective heat transfer correlation for LiPb flow and the Gnielinski convective heat transfer correlation for He flow. Since both LiPb and He serve as TBM coolants, the LiPb and He ancillary cooling systems were modeled to investigate the thermal-hydraulic characteristic of the TBM system and its influence on ITER safety under those accident conditions. The TBM components and the coolants flow within the TBM were simulated with one-dimensional heat structures and their associated hydrodynamic components. ITER enclosures including vacuum vessel (VV), port cell and TCWS vault were also covered in the model for accident analyses. Through this best estimate approach, the calculation indicated that the current

  10. Steady state and transient thermal-hydraulic characterization of full-scale ITER divertor plasma facing components

    International Nuclear Information System (INIS)

    Tincani, A.; Malavasi, A.; Ricapito, I.; Riccardi, B.; Di Maio, P.A.; Vella, G.

    2007-01-01

    In the frame of the activities related to ITER divertor R and D, ENEA CR Brasimone was charged by EFDA (European Fusion Design Agreement) to investigate the thermal-hydraulic behaviour of the full-scale divertor plasma facing components, i.e. Inner Vertical Target, Dome Liner and Outer Vertical Target, both in steady state and during draining and drying transient. More in detail, for each PFC, the first phase of the work is the steady state hydraulic characterization which consists of: - measurements of pressure drops at different temperatures; - determination of the velocity distribution in the internal channels; - check the possible insurgence of cavitation. The subsequent phase of the thermal-hydraulic characterization foresees a testing campaign of draining and drying procedure by means of a suitable gas flow. The objective of this experimental procedure is to eliminate in the most efficient way the residual amount of water after gravity discharge. In order to accomplish this experimental campaign a significant modification of CEF1 loop has been designed and realized. This paper presents, first of all, the experimental set-up, the agreed test matrix and the achieved results for both steady state and transient tests. Moreover, the level of the implementation of a predictive hydraulic model, based on RELAP 5 code, as well as its results are described, discussed and compared with the experimental ones. (orig.)

  11. Thermal-hydraulic process for cooling, heating and power production with low-grade heat sources in residential sector

    International Nuclear Information System (INIS)

    Borgogno, R.; Mauran, S.; Stitou, D.; Marck, G.

    2017-01-01

    Highlights: • Assessment of solar thermal-hydraulic process for tri-generation application. • Choice of the most suitable working fluid pair (R1234yf/R1233zd). • Evaluation of the global annual performance in Mediterranean climate. • Global annual COP and heat amplification achieving 0.24 and 1.2 respectively. • Global annual performance achieving an electric efficiency of 3.7%. - Abstract: A new process based on thermal-hydraulic conversion actuated by low-grade thermal energy is investigated. Input thermal energy can be provided by the means of solar collectors, as well as other low temperature energy sources. In the following article, “thermo-hydraulic” term refers to a process involving an incompressible fluid used as an intermediate medium to transfer work hydraulically between different thermal operated components or sub-systems. The system aims at providing trigeneration energy features for the residential sector, that is providing heating, cooling and electrical power for meeting the energy needs of domestic houses. This innovative system is made of two dithermal processes (working at two different levels of temperatures) and featuring two different working fluids. The first process is able to directly supply either electrical energy generated by an hydraulic turbine or drives the second process thanks to the incompressible fluid, which is similar to a heat pump effect for heating or cooling purposes. The innovative aspect of this process relies on the use of an hydraulic transfer fluid to transfer the work between each sub-system and therefore simplifying the conversion chain. A model, assuming steady-state operation, is developed to assess the energy performances of different variants of this thermo-hydraulic process with various heat source temperatures (80–110 °C) or heat sinks (0–30 °C), as well as various pairs of working fluids. For instance, in the frame of a single-family home, located in the Mediterranean region, the working

  12. Thermal-hydraulic Experiments for Advanced Physical Model Development

    International Nuclear Information System (INIS)

    Song, Chul Hwa; Baek, W. P.; Yoon, B. J.

    2010-04-01

    The improvement of prediction models is needed to enhance the safety analysis capability through the fine measurements of local phenomena. To improve the two-phase interfacial area transport model, the various experiments were carried out used SUBO and DOBO. 2x2 and 6x6 rod bundle test facilities were used for the experiment on the droplet behavior. The experiments on the droplet behavior inside a heated rod bundle were focused on the break-up of droplets induced by a spacer grid in a rod bundle geometry. The experiments used GIRLS and JICO and CFD analysis were carried out to comprehend the local condensation of steam jet, turbulent jet induced by condensation and the thermal mixing in a pool. An experimental database of the CHF (Critical Heat Flux) and PDO (Post-dryout) had been constructed. The mechanism of the heat transfer enhancement by surface modifications in nano-fluid was investigated in boiling mode and rapid quenching mode. The special measurement techniques were developed. They are Double -sensor optical void probe, Optic Rod, PIV technique and UBIM system

  13. Thermal-hydraulic Experiments for Advanced Physical Model Development

    International Nuclear Information System (INIS)

    Song, Chulhwa

    2012-04-01

    The improvement of prediction models is needed to enhance the safety analysis capability through experimental database of local phenomena. To improve the two-phase interfacial area transport model, the various experiments were carried out with local two-phase interfacial structure test facilities. 2 Χ 2 and 6 Χ 6 rod bundle test facilities were used for the experiment on the droplet behavior. The experiments on the droplet behavior inside a heated rod bundle geometry. The experiments used GIRLS and JICO and CFD analysis were carried out to comprehend the local condensation of steam jet, turbulent jet induced by condensation and the thermal mixing in a pool. In order to develop a model for key phenomena of newly adapted safety system, experiments for boiling inside a pool and condensation in horizontal channel have been performed. An experimental database of the CHF (Critical Heat Flux) and PDO (Post-dryout) was constructed. The mechanism of the heat transfer enhancement by surface modifications in nano-fluid was investigated in boiling mode and rapid quenching mode. The special measurement techniques were developed. They are Double-sensor optical void probe, Optic Rod, PIV technique and UBIM system

  14. Thermal hydraulic study of a corium molten pool

    International Nuclear Information System (INIS)

    Pigny, S.; Grand, D.; Seiler, J.M.; Durin, M.

    1993-01-01

    The thermohydraulic behaviour of a mass of molten core is investigated, in the frame of PWR severe accidents studies. The corium may be located in the vessel lower head or in an external core-catcher. It is assumed to be present in the container instantaneously. Its motion is described by one velocity field. It may be homogeneous or made of two stratified fluids. The residual power is assumed to be constant and uniform in the UO 2 phase. The radiative losses and the external water-cooling are taken into account. The thermal resistance of a peripheral crust is considered. The influence of the crust on the pool geometry may be studied. The wall behaviour is analysed by a conduction calculation. The interest of a sacrificial layer is underlined, so as the necessity of a multicomponent multiphase model to study the behaviour of a core catcher. It is also concluded that some experiments are needed for code validation about volume heated natural convection and multiphase flows. (author). 14 figs., 3 refs

  15. Thermal-hydraulic and neutron-physical characteristics of a new SCWR fuel assembly

    International Nuclear Information System (INIS)

    Liu, X.J.; Cheng, X.

    2009-01-01

    A new fuel assembly design for a thermal supercritical water cooled reactor (SCWR) core is proposed. Compared to the existing fuel assemblies, the present fuel assembly has two-rows of fuel rods between the moderator channels, to achieve a more uniform moderation for all fuel rod cells, and subsequently, a more uniform radial power distribution. In addition, a neutron-kinetics/thermal-hydraulics coupling method is developed, to analyze the neutron-physical and thermal-hydraulic behavior of the fuel assembly designs. This coupling method is based on the sub-channel analysis code COBRA-IV for thermal-hydraulics and the neutron-kinetics code SKETCH-N for neutron-physics. Both the COBRA-IV code and the SKETCH-N code are accordingly modified. An interface is established for the data transfer between these two codes. This coupling method is applied to both the one-row fuel assemblies (previous design) and the two-row fuel assemblies (present design). The performance of the two types of fuel assemblies is compared. The results show clearly that the two-row fuel assembly has more favorable neutron-physical and thermal-hydraulic characteristics than the one-row fuel assembly. The effect of various parameters on the fuel assembly performance is discussed. The coupling method is proven to be well suitable for further applications to SCWR fuel assembly design analysis

  16. Core Thermal-Hydraulic Conceptual Design for the Advanced SFR Design Concepts

    International Nuclear Information System (INIS)

    Cho, Chung Ho; Chang, Jin Wook; Yoo, Jae Woon; Song, Hoon; Choi, Sun Rock; Park, Won Seok; Kim, Sang Ji

    2010-01-01

    The Korea Atomic Energy Research Institute (KAERI) has developed the advanced SFR design concepts from 2007 to 2009 under the National longterm Nuclear R and D Program. Two types of core designs, 1,200 MWe breakeven and 600 MWe TRU burner core have been proposed and evaluated whether they meet the design requirements for the Gen IV technology goals of sustainability, safety and reliability, economics, proliferation resistance, and physical protection. In generally, the core thermal hydraulic design is performed during the conceptual design phase to efficiently extract the core thermal power by distributing the appropriate sodium coolant flow according to the power of each assembly because the conventional SFR core is composed of hundreds of ducted assemblies with hundreds of fuel rods. In carrying out the thermal and hydraulic design, special attention has to be paid to several performance parameters in order to assure proper performance and safety of fuel and core; the coolant boiling, fuel melting, structural integrity of the components, fuel-cladding eutectic melting, etc. The overall conceptual design procedure for core thermal and hydraulic conceptual design, i.e., flow grouping and peak pin temperature calculations, pressure drop calculations, steady-state and detailed sub-channel analysis is shown Figure 1. In the conceptual design phase, results of core thermal-hydraulic design for advanced design concepts, the core flow grouping, peak pin cladding mid-wall temperature, and pressure drop calculations, are summarized in this study

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

    International Nuclear Information System (INIS)

    Mesquita, Amir Z.; Costa, Antonio C.L.; Ladeira, Luiz C.D.; Rezende, Hugo C.; Palma, Daniel A.P.

    2015-01-01

    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)

  18. BWR transient analysis using neutronic / thermal hydraulic coupled codes including uncertainty quantification

    International Nuclear Information System (INIS)

    Hartmann, C.; Sanchez, V.; Tietsch, W.; Stieglitz, R.

    2012-01-01

    The KIT is involved in the development and qualification of best estimate methodologies for BWR transient analysis in cooperation with industrial partners. The goal is to establish the most advanced thermal hydraulic system codes coupled with 3D reactor dynamic codes to be able to perform a more realistic evaluation of the BWR behavior under accidental conditions. For this purpose a computational chain based on the lattice code (SCALE6/GenPMAXS), the coupled neutronic/thermal hydraulic code (TRACE/PARCS) as well as a Monte Carlo based uncertainty and sensitivity package (SUSA) has been established and applied to different kind of transients of a Boiling Water Reactor (BWR). This paper will describe the multidimensional models of the plant elaborated for TRACE and PARCS to perform the investigations mentioned before. For the uncertainty quantification of the coupled code TRACE/PARCS and specifically to take into account the influence of the kinetics parameters in such studies, the PARCS code has been extended to facilitate the change of model parameters in such a way that the SUSA package can be used in connection with TRACE/PARCS for the U and S studies. This approach will be presented in detail. The results obtained for a rod drop transient with TRACE/PARCS using the SUSA-methodology showed clearly the importance of some kinetic parameters on the transient progression demonstrating that the coupling of a best-estimate coupled codes with uncertainty and sensitivity tools is very promising and of great importance for the safety assessment of nuclear reactors. (authors)

  19. Engineering and thermal-hydraulic design of water cooled PFC for SST-1 tokamak

    International Nuclear Information System (INIS)

    Paritosh Chaudhuri; Santra, P.; Rabi Prakash, N.; Khirwadkar, S.; Arun Prakash, A.; Ramash, G.; Dubey, S.; Chenna Reddy, D.; Saxena, Y.C.

    2005-01-01

    plate is necessary for the efficient heat transfer from the tube to the back plate. The contact at the brazed joint of the tube to the backplate/heat sink is critical for the above application. The manufactured modules need to be evaluated for the quality of brazed joint. Using an infra-red-camera, spatial and temporal evaluation of the temperature profile has been studied under various flow parameters. Design considerations included 2-D steady state and transient tile temperature distribution and resulting thermal loads in PFC during cooling, coolant parameters necessary to maintain optimum thermal-hydraulic design, erosion rate of the tiles, and tile fitting mechanism. A 2-D Finite Difference code has been developed to study of flow behavior and thermal response of PFC during cooling. The temperature distribution results for different PFC obtained by code were assessed by comparison with 2-D Finite Element (FE) method (using ANSYS). FE models have been developed to conduct the heat transfer and stress analyses of the PFC to understand its thermal and mechanical behaviors. The result of the calculation led to a good understanding of the flow behavior and the temperature distribution in the tube wall and the different parts of the PFC. Thermal response on divertors has been performed both in steady state and transient case. Stress analyses also have been performed by ANSYS to investigate the thermal stress on different PFC during cooling. In this paper an optimized thermal-hydraulic design of PFC cooling and their thermal response will be discussed in detail. (authors)

  20. 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.

  1. 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.

  2. Uncertainty Evaluation of the SFR Subchannel Thermal-Hydraulic Modeling Using a Hot Channel Factors Analysis

    International Nuclear Information System (INIS)

    Choi, Sun Rock; Cho, Chung Ho; Kim, Sang Ji

    2011-01-01

    In an SFR core analysis, a hot channel factors (HCF) method is most commonly used to evaluate uncertainty. It was employed to the early design such as the CRBRP and IFR. In other ways, the improved thermal design procedure (ITDP) is able to calculate the overall uncertainty based on the Root Sum Square technique and sensitivity analyses of each design parameters. The Monte Carlo method (MCM) is also employed to estimate the uncertainties. In this method, all the input uncertainties are randomly sampled according to their probability density functions and the resulting distribution for the output quantity is analyzed. Since an uncertainty analysis is basically calculated from the temperature distribution in a subassembly, the core thermal-hydraulic modeling greatly affects the resulting uncertainty. At KAERI, the SLTHEN and MATRA-LMR codes have been utilized to analyze the SFR core thermal-hydraulics. The SLTHEN (steady-state LMR core thermal hydraulics analysis code based on the ENERGY model) code is a modified version of the SUPERENERGY2 code, which conducts a multi-assembly, steady state calculation based on a simplified ENERGY model. The detailed subchannel analysis code MATRA-LMR (Multichannel Analyzer for Steady-State and Transients in Rod Arrays for Liquid Metal Reactors), an LMR version of MATRA, was also developed specifically for the SFR core thermal-hydraulic analysis. This paper describes comparative studies for core thermal-hydraulic models. The subchannel analysis and a hot channel factors based uncertainty evaluation system is established to estimate the core thermofluidic uncertainties using the MATRA-LMR code and the results are compared to those of the SLTHEN code

  3. Thermal-hydraulic analysis of SMART steam generator tube rupture using TASS/SMR-S code

    International Nuclear Information System (INIS)

    Kim, Hee-Kyung; Kim, Soo Hyoung; Chung, Young-Jong; Kim, Hyeon-Soo

    2013-01-01

    Highlights: ► The analysis was performed from the viewpoint of primary coolant leakage. ► The thermal hydraulic responses and the maximum leakage have been identified. ► There is no direct release into the atmosphere caused by an SGTR accident. ► SMART safety system works well against an SGTR accident. - Abstract: A steam generator tube rupture (SGTR) accident analysis for SMART was performed using the TASS/SMR-S code. SMART with a rated thermal power of 330 MWt has been developed at the Korea Atomic Energy Research Institute. The TASS/SMR-S code can analyze the thermal hydraulic phenomena of SMART in a full range of reactor operating conditions. An SGTR is one of the most important accidents from a thermal hydraulic and radiological viewpoint. A conservative analysis against a SMART SGTR was performed. The major concern of this analysis is to find the thermal hydraulic responses and maximum leakage amount from a primary to a secondary side caused by an SGTR accident. A sensitivity study searching for the conservative thermal hydraulic conditions, break locations, reactivity and other conditions was performed. The dominant parameters related with the integral leak are the high RCS pressure, low core inlet coolant temperature and low break location of the SG cassette. The largest integral leak comes to 28 tons in the most conservative case during 1 h. But there is no direct release into the atmosphere because the secondary system pressure is maintained with a sufficient margin for the design pressure. All leaks go to the condenser. The analysis results show that the primary and secondary system pressures are maintained below the design pressure and the SMART safety system is working well against an SGTR accident

  4. Feasibility study on thermal-hydraulic performance in tight-lattice rod bundles for reduced-moderation water reactors

    International Nuclear Information System (INIS)

    Ohnuki, A.; Kureta, M.; Liu, W.; Tamai, H.; Akimoto, H.

    2004-01-01

    Research and development project for investigating thermal-hydraulic performance in tight-lattice rod bundles for Reduced-Moderation Water Reactor (RMWR) started at Japan Atomic Energy Research Institute (JAERI) in 2002. The RMWR can attain the favorable characteristics such as effective utilization of uranium resources, multiple recycling of plutonium, high burn-up and long operation cycle, based on matured light-water reactor technologies. MOX fuel assemblies with tight lattice arrangement are used to increase the conversion ratio by reducing the moderation of neutron. Increasing the in-core void fraction also contributes to the reduction of neutron moderation. The confirmation of thermal-hydraulic feasibility is one of the most important issues for the RMWR because of the tight-lattice configuration. The project has mainly consisted of a large-scale thermal-hydraulic test and development of analytical methods named modeling engineering. In the large-scale test, 37-rod bundle experiments can be performed. Steady-state critical power experiments have been achieved in the test facility and the experimental data reveal the feasibility of RMWR

  5. Thermal hydraulic analysis of the IPR-R1 TRIGA research reactor using a RELAP5 model

    International Nuclear Information System (INIS)

    Costa, Antonella L.; Reis, Patricia Amelia L.; Pereira, Claubia; Veloso, Maria Auxiliadora F.; Mesquita, Amir Z.; Soares, Humberto V.

    2010-01-01

    The RELAP5 code is widely used for thermal hydraulic studies of commercial nuclear power plants. Current investigations and code adaptations have demonstrated that the RELAP5 code can be also applied for thermal hydraulic analysis of nuclear research reactors with good predictions. Therefore, as a contribution to the assessment of RELAP5/MOD3.3 for research reactors analysis, this work presents steady-state and transient calculation results performed using a RELAP5 model to simulate the IPR-R1 TRIGA research reactor at 50 kilowatts (kW) of power operation. The reactor is located in the Nuclear Technology Development Center (CDTN), Brazil. It is a 250 kW, light water moderated and cooled, graphite-reflected, open pool type research reactor. The development and the assessment of a RELAP5 model for the IPR-R1 TRIGA are presented. Experimental data were considered in the process of the RELAP5 model validation. The RELAP5 results were also compared with calculated data from the STHIRP-1 (Research Reactors Thermal Hydraulic Simulation) code. The results obtained have shown that the RELAP5 model for the IPR-R1 TRIGA reproduces the actual steady-state reactor behavior in good agreement with the available data.

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

    Science.gov (United States)

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

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

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

    International Nuclear Information System (INIS)

    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

  8. VIPRE-01: a thermal-hydraulic code for reactor cores. Volume 3: programmer's manual (Revision 2)

    International Nuclear Information System (INIS)

    Stewart, C.W.; Koontz, A.S.; Cuta, J.M.; Montgomery, S.D.

    1985-07-01

    The VIPRE thermal-hydraulic computer code for PWR and BWR core analysis has undergone a detailed design review by a committee of experts. A new version of the code, incorporating the committee's recommendations, has been submitted for NRC review and issuance of a safety evaluation report. The changes in the programmers's manual are given

  9. Thermal-hydraulic feedback model to calculate the neutronic cross-section in PWR reactions

    International Nuclear Information System (INIS)

    Santiago, Daniela Maiolino Norberto

    2011-01-01

    In neutronic codes,it is important to have a thermal-hydraulic feedback module. This module calculates the thermal-hydraulic feedback of the fuel, that feeds the neutronic cross sections. In the neutronic co de developed at PEN / COPPE / UFRJ, the fuel temperature is obtained through an empirical model. This work presents a physical model to calculate this temperature. We used the finite volume technique of discretized the equation of temperature distribution, while calculation the moderator coefficient of heat transfer, was carried out using the ASME table, and using some of their routines to our program. The model allows one to calculate an average radial temperature per node, since the thermal-hydraulic feedback must follow the conditions imposed by the neutronic code. The results were compared with to the empirical model. Our results show that for the fuel elements near periphery, the empirical model overestimates the temperature in the fuel, as compared to our model, which may indicate that the physical model is more appropriate to calculate the thermal-hydraulic feedback temperatures. The proposed model was validated by the neutronic simulator developed in the PEN / COPPE / UFRJ for analysis of PWR reactors. (author)

  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. 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.

  12. Validation of a thermal-hydraulic system code on a simple example

    International Nuclear Information System (INIS)

    Kopecek, Vit; Zacha, Pavel

    2014-01-01

    A mathematical model of a U tube was set up and the analytical solution was calculated and used in the assessment of the numerical solutions obtained by using the RELAP5 mod3.3 and TRACE V5 thermal hydraulics codes. A good agreement between the 2 types of calculation was obtained.

  13. Trend analysis of troubles caused by thermal-hydraulic phenomena at nuclear power plants

    International Nuclear Information System (INIS)

    Komatsu, Teruo

    2010-01-01

    The Institute of Nuclear Safety System (INSS) is promoting researches to improve the safety and reliability of nuclear power plants. In the present study, our attention was focused on troubles attributed to thermal-hydraulic phenomena in particular, trend analysis were carried out to learn lessons from these troubles and to prevent their recurrence. Through our survey, we found the following two points. First, many thermal-hydraulics related troubles can be attributed to design faults, since we found some events in foreign countries took place after inadequate facility renovation. To ensure appropriate design verification, it is important to take account of state-of-the-art science and technology and at the same time to pay attention to the compatibility with the initial design concept. Second point, thermal-hydraulic related troubles are common and recurrent to nuclear power plants worldwide. Japanese utilities are planning to introduce some of overseas experiences to their plants, such as power uprate and renovations of aged facilities. It is important to learn lessons from experiences paying close attention continuously to overseas trouble events, including thermal-hydraulics related events, and to use them to improve safety and reliability of nuclear power plants. (author)

  14. Survey of thermal-hydraulic models of commercial nuclear power plants

    International Nuclear Information System (INIS)

    Determan, J.C.; Hendrix, C.E.

    1992-12-01

    A survey of the thermal-hydraulic models of nuclear power plants has been performed to identify the NRC's current analytical capabilities for critical event response. The survey also supports ongoing research for accident management. The results of the survey are presented here. The PC database which records detailed data on each model is described

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

    International Nuclear Information System (INIS)

    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.

    1993-01-01

    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

  16. Thermal-hydraulic calculation and analysis for QNPP (Qinshan Nuclear Power Plant) containment

    International Nuclear Information System (INIS)

    Xie Hui; Zhou Jie; He Yingchao

    1993-01-01

    Three containment thermal-hydraulic codes CONTEMPT-LT/028, CONTEMPT-4/MOD3 and COMPARE are used to compute and analyse the Qinshan Nuclear Power Plant (QNPP) containment response under LOCA or MSLB conditions. An evaluation of the capability of containment of QNPP is given

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

    International Nuclear Information System (INIS)

    Heard, F.J.

    1999-01-01

    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

  18. Steady-state thermal-hydraulic analysis of the pellet-bed reactor for nuclear thermal propulsion

    International Nuclear Information System (INIS)

    El-Genk, M.S.; Morley, N.J.; Yang, J.Y.

    1992-01-01

    The pellet-bed reactor (PBR) for nuclear thermal propulsion is a hydrogen-cooled, BeO-reflected, fast reactor, consisting of an annular core region filled with randomly packed, spherical fuel pellets. The fuel pellets in the PBR are self-supported, eliminating the need for internal core structure, which simplifies the core design and reduces the size and mass of the reactor. Each spherical fuel pellet is composed of hundreds of fuel microspheres embedded in a zirconium carbide (ZrC) matrix. Each fuel microsphere is composed of a UC-NbC fuel kernel surrounded by two consecutive layers of the NbC and ZrC. Gaseous hydrogen serves both as core coolant and as the propellant for the PBR rocket engine. The cold hydrogen flows axially down the inlet channel situated between the core and the external BeO reflector and radially through the orifices in the cold frit, the core, and the orifices in the hot frit. Finally, the hot hydrogen flows axially out the central channel and exits through converging-diverging nozzle. A thermal-hydraulic analysis of the PBR core was performed with an emphasis on optimizing the size and axial distribution of the orifices in the hot and cold frits to ensure that hot spots would not develop in the core during full-power operation. Also investigated was the validity of the assumptions of neglecting the axial conduction and axial cross flow in the core

  19. Optimal thermal-hydraulic performance for helium-cooled divertors

    International Nuclear Information System (INIS)

    Izenson, M.G.; Martin, J.L.

    1996-01-01

    Normal flow heat exchanger (NFHX) technology offers the potential for cooling divertor panels with reduced pressure drops (<0.5% Δp/p), reduced pumping power (<0.75% pumping/thermal power), and smaller duct sizes than conventional helium heat exchangers. Furthermore, the NFHX can easily be fabricated in the large sizes required for divertors in large tokamaks. Recent experimental and computational results from a program to develop NFHX technology for divertor coolings using porous metal heat transfer media are described. We have tested the thermal and flow characteristics of porous metals and identified the optimal heat transfer material for the divertor heat exchanger. Methods have been developed to create highly conductive thermal bonds between the porous material and a solid substrate. Computational fluid dynamics calculations of flow and heat transfer in the porous metal layer have shown the capability of high thermal effectiveness. An 18-kW NFHX, designed to meet specifications for the international Thermonuclear Experimental Reactor divertor, has been fabricated and tested for thermal and flow performance. Preliminary results confirm design and fabrication methods. 11 refs., 12 figs., 1 tab

  20. Oyster Creek fuel thermal margin during core thermal-hydraulic oscillations

    International Nuclear Information System (INIS)

    Dougher, J.D.

    1990-01-01

    The Oyster Creek nuclear facility, a boiling water reactor (BWR)-2 plant type, has never experienced core thermal-hydraulic instability. Power oscillations, however, have been observed in other BWR cores both domestically and internationally. Two modes of oscillations have been observed, core wide and regional half-core. During core wide oscillations, the neutron flux in the core oscillates in the radial fundamental mode. During regional half-core oscillations, higher order harmonics in the radial plane result in out-of-phase oscillations with the neutron flux in one half of the core oscillating 180 deg out-of-phase with the neutron flux in the other half of the core. General Design Criteria 12 requires either prevention or detection and suppression of power oscillations which could result in violations of fuel design limits. Analyses performed by General Electric have demonstrated that for large-magnitude oscillations the potential exists for violation of the safety limit minimum critical power ratio (MCPR). However, for plants with a flow-biased neutron flux scram automatic mitigation of oscillations may be provided at an oscillation magnitude below that at which the safety limit is challenged. Plant-specific analysis for Oyster Creek demonstrates that the existing average power range monitor (APRM) system will sense and suppress power oscillations prior to violation of any safety limits

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

    International Nuclear Information System (INIS)

    Cho, Jaehyun; Park, Jin Hee; Kim, Dong-San; Lim, Ho-Gon

    2017-01-01

    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

  2. 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

  3. Thermal-hydraulic limitations on water-cooled limiters

    International Nuclear Information System (INIS)

    Cha, Y.S.; Misra, B.

    1984-08-01

    An assessment of the cooling requirements for fusion reactor components, such as the first wall and limiter/divertor, was carried out using pressurized water as the coolant. In order to establish the coolant operating conditions, a survey of the literature on departure from nucleate boiling, critical heat flux, asymmetrical heating and heat transfer augmentation techniques was carried out. The experimental data and the empirical correlations indicate that thermal protection for the fusion reactor components based on current design concepts can be provided with an adequate margin of safety without resorting to either high coolant velocities, excessive coolant pressures, or heat transfer augmentation techniques. If, however, the future designs require heat transfer enhancement techniques, experimental verification would be necessary since no data on heat transfer augmentation techniques exist for complex geometries, especially under asymmetrically heated conditions. Since the data presented herein concern primarily thermal protection, the final design should consider other factors such as thermal stresses, temperature limits, and fatigue

  4. Hydraulic investigation on free surface flow of windowless target

    International Nuclear Information System (INIS)

    Hu Chen; Gu Hanyang

    2015-01-01

    The formation and control of free surface are the most essential parts in the studies of windowless target in ACCELERATOR-DRIVEN sub-critical system (ADS). Water model experiments and 360° full scale three dimensional simulations were conducted. The experimental study demonstrates that the free surface is significantly affected by the inlet flow velocity and outlet pressure. The length of free surface decreases in the second order with the increase of inlet flow velocity, while it decreases linearly with the outlet pressure. The structure and feature of flow field were investigated. The results show that the free surface is vulnerable to the vortex movement. Transient simulations were performed with volume of fluid (VOF) method, large eddy simulation (LES) and the pressure implicit with splitting of operators (PISO) algorithm. The simulation results agree qualitatively well with the experimental data related to both free surface flow and flow field. These simulation models and methods are proved to be applicable in the hydraulic simulations of liquid heavy metal target. (authors)

  5. Experimental Investigation of Crack Extension Patterns in Hydraulic Fracturing with Shale, Sandstone and Granite Cores

    Directory of Open Access Journals (Sweden)

    Jianming He

    2016-12-01

    Full Text Available Hydraulic fracturing is an important method of reservoir stimulation in the exploitation of geothermal resources, and conventional and unconventional oil and gas resources. In this article, hydraulic fracturing experiments with shale, sandstone cores (from southern Sichuan Basin, and granite cores (from Inner Mongolia were conducted to investigate the different hydraulic fracture extension patterns in these three reservoir rocks. The different reactions between reservoir lithology and pump pressure can be reflected by the pump pressure monitoring curves of hydraulic fracture experiments. An X-ray computer tomography (CT scanner was employed to obtain the spatial distribution of hydraulic fractures in fractured shale, sandstone, and granite cores. From the microscopic and macroscopic observation of hydraulic fractures, different extension patterns of the hydraulic fracture can be analyzed. In fractured sandstone, symmetrical hydraulic fracture morphology could be formed, while some micro cracks were also induced near the injection hole. Although the macroscopic cracks in fractured granite cores are barely observed by naked eye, the results of X-ray CT scanning obviously show the morphology of hydraulic fractures. It is indicated that the typical bedding planes well developed in shale formation play an important role in the propagation of hydraulic fractures in shale cores. The results also demonstrated that heterogeneity influenced the pathway of the hydraulic fracture in granite cores.

  6. Methodology for thermal hydraulic conceptual design and performance analysis of KALIMER core

    International Nuclear Information System (INIS)

    Young-Gyun Kim; Won-Seok Kim; Young-Jin Kim; Chang-Kue Park

    2000-01-01

    This paper summarizes the methodology for thermal hydraulic conceptual design and performance analysis which is used for KALIMER core, especially the preliminary methodology for flow grouping and peak pin temperature calculation in detail. And the major technical results of the conceptual design for the KALIMER 98.03 core was shown and compared with those of KALIMER 97.07 design core. The KALIMER 98.03 design core is proved to be more optimized compared to the 97.07 design core. The number of flow groups are reduced from 16 to 11, and the equalized peak cladding midwall temperature from 654 deg. C to 628 deg. C. It was achieved from the nuclear and thermal hydraulic design optimization study, i.e. core power flattening and increase of radial blanket power fraction. Coolant flow distribution to the assemblies and core coolant/component temperatures should be determined in core thermal hydraulic analysis. Sodium flow is distributed to core assemblies with the overall goal of equalizing the peak cladding midwall temperatures for the peak temperature pin of each bundle, thus pin cladding damage accumulation and pin reliability. The flow grouping and the peak pin temperature calculation for the preliminary conceptual design is performed with the modules ORFCE-F60 and ORFCE-T60 respectively. The basic subchannel analysis will be performed with the SLTHEN code, and the detailed subchannel analysis will be done with the MATRA-LMR code which is under development for the K-Core system. This methodology was proved practical to KALIMER core thermal hydraulic design from the related benchmark calculation studies, and it is used to KALIMER core thermal hydraulic conceptual design. (author)

  7. Study of thermal hydraulic behavior of supercritical water flowing through fuel rod bundles

    International Nuclear Information System (INIS)

    Thakre, Sachin; Lakshmanan, S.P.; Kulkarni, Vinayak; Pandey, Manmohan

    2009-01-01

    Investigations on thermal-hydraulic behavior in Supercritical Water Reactor (SCWR) fuel assembly have obtained a significant attention in the international SCWR community because of its potential to obtain high thermal efficiency and compact design. Present work deals with CFD analysis to study the flow and heat transfer behavior of supercritical water in 4 metre long 7-pin fuel bundle using commercial CFD package ANSYS CFX for single phase steady state conditions. Considering the symmetric conditions, 1/12th part of the fuel rod bundle is taken as a domain of analysis. RNG K-epsilon model with scalable wall functions is used for modeling the turbulence behavior. Constant heat flux boundary condition is applied at the fuel rod surface. IAPWS equations of state are used to compute thermo-physical properties of supercritical water. Sharp variations in its thermo-physical properties (specific heat, density) are observed near the pseudo-critical temperature causing sharp change in heat transfer coefficient. The pseudo-critical point initially appears in the gaps among heated fuel rods, and then spreads radially outward reaching the adiabatic wall as the flow goes downstream. The enthalpy gain in the centre of the channel is much higher than that in the wall region. Non-uniformity in the circumferential distribution of surface temperature and heat transfer coefficient is observed which is in agreement with published literature. Heat transfer coefficient is high on the rod surface near the tight region and decreases as the distance between rod surfaces increases. (author)

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

    Directory of Open Access Journals (Sweden)

    Anil Kumar

    2016-07-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2017-09-01

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

  10. Thermal-hydraulic analysis of a 600 MW supercritical CFB boiler with low mass flux

    International Nuclear Information System (INIS)

    Pan Jie; Yang Dong; Chen Gongming; Zhou Xu; Bi Qincheng

    2012-01-01

    Supercritical Circulating Fluidized Bed (CFB) boiler becomes an important development trend for coal-fired power plant and thermal-hydraulic analysis is a key factor for the design and operation of water wall. According to the boiler structure and furnace-sided heat flux, the water wall system of a 600 MW supercritical CFB boiler is treated in this paper as a flow network consisting of series-parallel loops, pressure grids and connecting tubes. A mathematical model for predicting the thermal-hydraulic characteristics in boiler heating surface is based on the mass, momentum and energy conservation equations of these components, which introduces numerous empirical correlations available for heat transfer and hydraulic resistance calculation. Mass flux distribution and pressure drop data in the water wall at 30%, 75% and 100% of the boiler maximum continuous rating (BMCR) are obtained by iteratively solving the model. Simultaneity, outlet vapor temperatures and metal temperatures in water wall tubes are estimated. The results show good heat transfer performance and low flow resistance, which implies that the water wall design of supercritical CFB boiler is applicable. - Highlights: → We proposed a model for thermal-hydraulic analysis of boiler heating surface. → The model is applied in a 600 MW supercritical CFB boiler. → We explore the pressure drop, mass flux and temperature distribution in water wall. → The operating safety of boiler is estimated. → The results show good heat transfer performance and low flow resistance.

  11. 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.

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

    International Nuclear Information System (INIS)

    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

  13. Thermal-hydraulic analysis and design improvement for coolant channel of ITER shield block

    International Nuclear Information System (INIS)

    Zhao Ling; Li Huaqi; Zheng Jiantao; Yi Jingwei; Kang Weishan; Chen Jiming

    2013-01-01

    As an important part for ITER, shield block is used to shield the neutron heat. The structure design of shield block, especially the inner coolant channel design will influence its cooling effect and safety significantly. In this study, the thermal-hydraulic analysis for shield block has been performed by the computational fluid dynamics software, some optimization suggestions have been proposed and thermal-hydraulic characteristics of the improved model has been analyzed again. The analysis results for improved model show that pressure drop through flow path near the inlet and outlet region of the shield block has been reduced, and the total pressure drop in cooling path has been reduced too; the uniformity of the mass flowrate distribution and the velocity distribution have been improved in main cooling branches; the local highest temperature of solid domain reduced considerably, which could avoid thermal stress becoming too large because of coolant effect unevenly. (authors)

  14. Thermal-hydraulic calculations for KUHFR with reduced enrichment uranium fuel

    International Nuclear Information System (INIS)

    Mishima, Kaichiro; Shibata, Toshikazu.

    1982-01-01

    This report provides the preliminary results of the thermal-hydraulic calculations to study the safety aspects in fueling the KUHFR with reduced enrichment uranium. The calculations were based on what was outlined in the Safety Analysis Report for the KUHFR and the guidebook for research reactor core conversion, IAEA-TECDOC-233, published by the International Atomic Energy Agency. No significant differences in the thermal-hydraulic operating conditions have been found between HEU and MEU fuels. However, in LEU cases, the combination of three factors - larger power peaking with LEU fuel, smaller thermal conductivity of U 3 O 8 -Al fuel with high uranium densities, and thicker fuel meat - resulted in higher maximum fuel and surface temperatures with the LEU oxide fuel. (author)

  15. Thermal-hydraulic tests on net divertor targets using swirl tubes

    International Nuclear Information System (INIS)

    Schlosser, J.; Chappuis, P.; Deschamps, P.; Massmann, P.; Falter, H.D.; Deschamps, G.H.

    1991-01-01

    Thermal-hydraulic tests have been carried out in collaboration between NET, CEA Cadarache and JET in order to find a cooling method capable of removing the high heat fluxes expected for the NET/ITER divertor. The goal was to evaluate by experiments the critical heat flux (CHF) and heat transfer in the subcooled boiling regime using twisted tapes as turbulence promoters and testing them under relevant thermal-hydraulic conditions. The CEA 200 kW Electron Beam (EB) facility and the 10 MW JET Neutral Beam (NB) test bed have been used to heat up the NET relevant test sections (TS) consisting of rectangular copper elements with circular internal channels. The TS have been exposed to the electron or ion beams under normal incidence. This paper reports the results of the experiments and of thermal analyses performed in support of the tests. The experimental CHF values have been benchmarked with the Tong-75 correlation

  16. Single-channel model for steady thermal-hydraulic analysis in nuclear reactor

    International Nuclear Information System (INIS)

    Zhang Xiaoying; Huang Yuanyuan

    2010-01-01

    This article established a single-channel model for steady analysis in the reactor and an example of thermal-hydraulic analysis was made by using this model, including the Maximum heat flux density of fuel element, enthalpy, Coolant flow, various kinds of pressure drop, enthalpy increase in average tube and thermal tube. I also got the Coolant temperature distribution and the fuel element temperature distribution and analysis of the final result. The results show that some relevant parameters which we got in this paper are well coincide with the actual operating parameters. It is also show that the single-channel model can be used to the steady thermal-hydraulic analysis. (authors)

  17. Advanced methodology to simulate boiling water reactor transient using coupled thermal-hydraulic/neutron-kinetic codes

    Energy Technology Data Exchange (ETDEWEB)

    Hartmann, Christoph Oliver

    2016-06-13

    Coupled Thermal-hydraulic/Neutron-kinetic (TH/NK) simulations of Boiling Water Reactor transients require well validated and accurate simulation tools. The generation of cross-section (XS) libraries, depending on the individual thermal-hydraulic state parameters, is of paramount importance for coupled simulations. Problem-dependent XS-sets for 3D core simulations are being generated mainly by well validated, fast running commercial and user-friendly lattice codes such as CASMO and HELIOS. In this dissertation a computational route, based on the lattice code SCALE6/TRITON, the cross-section interface GenPMAXS, the best-estimate thermal-hydraulic system code TRACE and the core simulator PARCS, for best-estimate simulations of Boiling Water (BWR) transients has been developed and validated. The computational route has been supplemented by a subsequent uncertainty and sensitivity study based on Monte Carlo sampling and propagation of the uncertainties of input parameters to the output (SUSA code). The analysis of a single BWR fuel assembly depletion problem with PARCS using SCALE/TRITON cross-sections has been shown a good agreement with the results obtained with CASMO cross-section sets. However, to compensate the deficiencies of the interface program GenPMAXS, PYTHON scripts had to be developed to incorporate missing data, as the yields of Iodine, Xenon and Promethium, into the cross-section-data sets (PMAXS-format) generated by GenPMAXS from the SCALE/TRITON output. The results of the depletion analysis of a full BWR core with PARCS have indicated the importance of considering history effects, adequate modeling of the reflector region and the control rods, as the PARCS simulations for depleted fuel and all control rods inserted (ARI) differs significantly at the fuel assembly top and bottom. Systematic investigations with the coupled codes TRACE/PARCS have been performed to analyse the core behaviour at different thermal conditions using nuclear data (XS

  18. Flow resistance of orifices and spacers of BWR thermal-hydraulic and neutronic coupling loop

    International Nuclear Information System (INIS)

    Iguchi, Tadashi; Asaka, Hideaki; Nakamura, Hideo

    2002-03-01

    Authors are performing THYNC experiments to study thermal-hydraulic instability under neutronic and thermal-hydraulic coupling. In THYNC experiments, the orifices are installed at the exit of the test section and the spacers are installed in the test section, in order to properly simulate in-core thermal-hydraulics in the reactor core. It is necessary to know the flow resistance of the orifices and spacers for the analysis of THYNC experimental results. Consequently, authors measured the flow resistance of orifice and spacer under single-phase and two-phase flows. Using the experimental results, authors investigated the dependency of the flow resistances on the parameters, such as pressure, mass flux, an geometries. Furthermore, authors investigated the applicability of the basic two-phase flow models, for example the separate flow model, to the two-phase flow multiplier. As the result of the investigation on the single-phase flow experiment, it was found (1) that the effects of pressure and mass flux flow resistance are described by a function of Reynolds number, and (2) that flow resistances of the orifice and the spacer are calculated with the previous prediction methods. However, it was necessary to introduce an empirical coefficient, since it was difficult to predict accurately the flow resistance only with the previous prediction method due to the complicated geometry dependency, for example a flow area blockage ratio. On the other hand, according to the investigation on two-phase flow experiment, the followings were found. (1) Relation between the two-phase flow multiplier and the quality is regarded to be linear under pressure of 2MPa - 7MPa. The relation is dependent on pressure and geometry, and is little dependent on mass flux. (2) Relation between the two-phase flow multiplier and void fraction is little dependent on pressure, mass flux, and geometry under pressure of 0.2MPa - 7MPa and void fraction less than 0.6. The relation is less dependent on

  19. Gas-cooled reactor thermal-hydraulics using CAST3M and CRONOS2 codes

    International Nuclear Information System (INIS)

    Studer, E.; Coulon, N.; Stietel, A.; Damian, F.; Golfier, H.; Raepsaet, X.

    2003-01-01

    The CEA R and D program on advanced Gas Cooled Reactors (GCR) relies on different concepts: modular High Temperature Reactor (HTR), its evolution dedicated to hydrogen production (Very High Temperature Reactor) and Gas Cooled Fast Reactors (GCFR). Some key safety questions are related to decay heat removal during potential accident. This is strongly connected to passive natural convection (including gas injection of Helium, CO 2 , Nitrogen or Argon) or forced convection using active safety systems (gas blowers, heat exchangers). To support this effort, thermal-hydraulics computer codes will be necessary tools to design, enhance the performance and ensure a high safety level of the different reactors. Accurate and efficient modeling of heat transfer by conduction, convection or thermal radiation as well as energy storage are necessary requirements to obtain a high level of confidence in the thermal-hydraulic simulations. To achieve that goal a thorough validation process has to ve conducted. CEA's CAST3M code dedicated to GCR thermal-hydraulics has been validated against different test cases: academic interaction between natural convection and thermal radiation, small scale in-house THERCE experiments and large scale High Temperature Test Reactor benchmarks such as HTTR-VC benchmark. Coupling with neutronics is also an important modeling aspect for the determination of neutronic parameters such as neutronic coefficient (Doppler, moderator,...), critical position of control rods...CEA's CAST3M and CRONOS2 computer codes allow this coupling and a first example of coupled thermal-hydraulics/neutronics calculations has been performed. Comparison with experimental data will be the next step with High Temperature Test Reactor experimental results at nominal power

  20. Techniques for the thermal/hydraulic analysis of LMFBR check valves

    International Nuclear Information System (INIS)

    Cho, S.M.; Kane, R.S.

    1979-01-01

    A thermal/hydraulic analysis of the check valves in liquid sodium service for LMFBR plants is required to provide temperature data for thermal stress analysis of the valves for specified transient conditions. Because of the complex three-dimensional flow pattern within the valve, the heat transfer analysis techniques for less complicated shapes could not be used. This paper discusses the thermal analysis techniques used to assure that the valve stress analysis is conservative. These techniques include a method for evaluating the recirculating flow patterns and for selecting appropriately conservative heat transfer correlations in various regions of the valve

  1. Development and Applications of a General Coupled Thermal-hydraulic/Neutronic Model for the Ringhals-3 Pressurized Water Reactor

    International Nuclear Information System (INIS)

    Staalek, Mathias

    2008-03-01

    Coupled calculations are important for the simulation of nuclear power plants when there is a strong feedback between the neutron kinetics and the thermal-hydraulics. A general coupled model of the Ringhals-3 Pressurized Water Reactor has been developed for this purpose. The development is outlined in the thesis with details given in the appended papers. A PARCS model was developed for the core calculations and a RELAP5 model for the thermal-hydraulic calculations. The RELAP5 model has 157 channels for modelling the flow in the fuel assemblies. This means that there is a one-one correspondence radially between the neutronic and thermal-hydraulic nodalization. This detailed mapping between the neutron kinetics and the thermal-hydraulics makes it possible to use the model for all kinds of transient. To provide realistic material data to the PARCS model, a cross-section interface was developed. With this interface one can import material data from a binary CASMO-4 library file into PARCS. Due to the one-to-one mapping, any any core loading can easily be considered. The PARCS model was benchmarked against measurements of the steady-state power distribution of Ringhals-3. The power shape was well reproduced by the model. Validational work for steady-state conditions of the thermal-hydraulic was also successfully performed. The most challenging part of the validation of a coupled model is for transients. This is much more difficult since the dynamics of the system becomes very important. Two transients that occurred at Ringhals-3 were chosen for the validational work. The first transient was a Load Rejection Transient. In general the model gave good results but some problems were experienced, e.g. the pressurizer pressure turned out to be more difficult to be correctly simulated. The second transient was a Loss of Feed Water transient. A malfunctioning feed water control valve closed, and therefore shut down the feed water supply to the steam generator in one of the

  2. Development and Applications of a General Coupled Thermal-hydraulic/Neutronic Model for the Ringhals-3 Pressurized Water Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Staalek, Mathias

    2008-03-15

    Coupled calculations are important for the simulation of nuclear power plants when there is a strong feedback between the neutron kinetics and the thermal-hydraulics. A general coupled model of the Ringhals-3 Pressurized Water Reactor has been developed for this purpose. The development is outlined in the thesis with details given in the appended papers. A PARCS model was developed for the core calculations and a RELAP5 model for the thermal-hydraulic calculations. The RELAP5 model has 157 channels for modelling the flow in the fuel assemblies. This means that there is a one-one correspondence radially between the neutronic and thermal-hydraulic nodalization. This detailed mapping between the neutron kinetics and the thermal-hydraulics makes it possible to use the model for all kinds of transient. To provide realistic material data to the PARCS model, a cross-section interface was developed. With this interface one can import material data from a binary CASMO-4 library file into PARCS. Due to the one-to-one mapping, any any core loading can easily be considered. The PARCS model was benchmarked against measurements of the steady-state power distribution of Ringhals-3. The power shape was well reproduced by the model. Validational work for steady-state conditions of the thermal-hydraulic was also successfully performed. The most challenging part of the validation of a coupled model is for transients. This is much more difficult since the dynamics of the system becomes very important. Two transients that occurred at Ringhals-3 were chosen for the validational work. The first transient was a Load Rejection Transient. In general the model gave good results but some problems were experienced, e.g. the pressurizer pressure turned out to be more difficult to be correctly simulated. The second transient was a Loss of Feed Water transient. A malfunctioning feed water control valve closed, and therefore shut down the feed water supply to the steam generator in one of the

  3. A Newton-based Jacobian-free approach for neutronic-Monte Carlo/thermal-hydraulic static coupled analysis

    International Nuclear Information System (INIS)

    Mylonakis, Antonios G.; Varvayanni, M.; Catsaros, N.

    2017-01-01

    Highlights: •A Newton-based Jacobian-free Monte Carlo/thermal-hydraulic coupling approach is introduced. •OpenMC is coupled with COBRA-EN with a Newton-based approach. •The introduced coupling approach is tested in numerical experiments. •The performance of the new approach is compared with the traditional “serial” coupling approach. -- Abstract: In the field of nuclear reactor analysis, multi-physics calculations that account for the bonded nature of the neutronic and thermal-hydraulic phenomena are of major importance for both reactor safety and design. So far in the context of Monte-Carlo neutronic analysis a kind of “serial” algorithm has been mainly used for coupling with thermal-hydraulics. The main motivation of this work is the interest for an algorithm that could maintain the distinct treatment of the involved fields within a tight coupling context that could be translated into higher convergence rates and more stable behaviour. This work investigates the possibility of replacing the usually used “serial” iteration with an approximate Newton algorithm. The selected algorithm, called Approximate Block Newton, is actually a version of the Jacobian-free Newton Krylov method suitably modified for coupling mono-disciplinary solvers. Within this Newton scheme the linearised system is solved with a Krylov solver in order to avoid the creation of the Jacobian matrix. A coupling algorithm between Monte-Carlo neutronics and thermal-hydraulics based on the above-mentioned methodology is developed and its performance is analysed. More specifically, OpenMC, a Monte-Carlo neutronics code and COBRA-EN, a thermal-hydraulics code for sub-channel and core analysis, are merged in a coupling scheme using the Approximate Block Newton method aiming to examine the performance of this scheme and compare with that of the “traditional” serial iterative scheme. First results show a clear improvement of the convergence especially in problems where significant

  4. Momentum integral network method for thermal-hydraulic transient analysis

    International Nuclear Information System (INIS)

    Van Tuyle, G.J.

    1983-01-01

    A new momentum integral network method has been developed, and tested in the MINET computer code. The method was developed in order to facilitate the transient analysis of complex fluid flow and heat transfer networks, such as those found in the balance of plant of power generating facilities. The method employed in the MINET code is a major extension of a momentum integral method reported by Meyer. Meyer integrated the momentum equation over several linked nodes, called a segment, and used a segment average pressure, evaluated from the pressures at both ends. Nodal mass and energy conservation determined nodal flows and enthalpies, accounting for fluid compression and thermal expansion

  5. A comprehensive review on the methodologies to simulate the nuclear fuel bundle for the thermal hydraulic experiments

    International Nuclear Information System (INIS)

    Vishnoi, A.K.; Chandraker, D.K.; Pal, A.K.; Vijayan, P.K.; Saha, D.

    2011-01-01

    The designer of a nuclear reactor system has to ensure its safety during normal operation as well as accidental conditions. This requires, among other things, a proper understanding of the various thermal hydraulic phenomena occurring in the reactor core. In a nuclear reactor core the fuel elements are the heat source and highly loaded components of the reactor system. Therefore their behaviour under normal and accidental conditions must be extensively investigated. Data generation for Critical heat flux (CHF) in full scale bundle and parallel channel instability studies with at least two full size channels are required in order to evaluate the thermal margin and stability margin of the reactor. The complex nature of these phenomena calls for exhaustive experimental investigations. Fuel Rod Cluster Simulator (FRCS) is a very important component required for the experimental investigation of the thermal hydraulic behaviour of reactor fuel elements under normal and accidental conditions. This paper brings out a comprehensive review of the FRCS elaborating the challenges and important design aspects of the FRCS. Some of the main features and analysis results on the performance of the developed FRCS with respect to the actual nuclear fuel bundle will be presented in the paper. (author)

  6. PWR boron dilution transients. Thermal-hydraulic analyses of PKL-E experiments

    International Nuclear Information System (INIS)

    Pietro Alessandro Di Maio; Antonino Tomasello; Giuseppe Vella

    2005-01-01

    Full text of publication follows: In the field of PWR reactor safety, one of the topics that is currently of major interest worldwide is that of inadvertent boron dilution events. The safety issue involved in such scenarios is that inadvertent transport into the reactor core of un-borated water - or water having only a low boron concentration - can lead to local recriticality and possibly to power excursions. Studies on various accidental sequences that could initiate boron dilution events revealed that some SBLOCAs, occurring in the primary system, lead to reflux condenser conditions and subsequent re-establishment of natural circulation are of particular significance. In this work, the first field of analysis is related to the investigation of the thermal - hydraulic conditions that could lead to boron dilution events, such as the stop of natural circulation within primary system and the subsequent start of reflux condenser functioning mode. The investigation of the primary thermal - hydraulic conditions has been performed using the experimental results obtained in the PKL test integral facility in which some SBLOCA sequences have been carried out. Particular useful were the PKL III E experiments data whose results have been numerically reproduced using the code Relap5/MOD3.3/Beta code, contributing to understand the complex thermalhydraulic phenomena related to a PWR boron dilution event. The second field of analysis is related to the effects that possible displacements of un-borated water slugs towards the Reactor Pressure Vessel (RPV) could have on the core reactivity. A numerical approach using the Relap5 reactor kinetics model has been adopted to integrate the experimental thermal - hydraulic data obtained in the PKL III E tests. A careful analysis has been performed in order to establish which core conditions at incident start could produce the largest reactivity increase as a consequence of restarting of natural circulation during the primary system

  7. PWR boron dilution transients. Thermal-hydraulic analyses of PKL-E experiments

    Energy Technology Data Exchange (ETDEWEB)

    Pietro Alessandro Di Maio; Antonino Tomasello; Giuseppe Vella [Dipartimento di Ingegneria Nucleare, Viale delle Scienze, 90128 Palermo (Italy)

    2005-07-01

    Full text of publication follows: In the field of PWR reactor safety, one of the topics that is currently of major interest worldwide is that of inadvertent boron dilution events. The safety issue involved in such scenarios is that inadvertent transport into the reactor core of un-borated water - or water having only a low boron concentration - can lead to local recriticality and possibly to power excursions. Studies on various accidental sequences that could initiate boron dilution events revealed that some SBLOCAs, occurring in the primary system, lead to reflux condenser conditions and subsequent re-establishment of natural circulation are of particular significance. In this work, the first field of analysis is related to the investigation of the thermal - hydraulic conditions that could lead to boron dilution events, such as the stop of natural circulation within primary system and the subsequent start of reflux condenser functioning mode. The investigation of the primary thermal - hydraulic conditions has been performed using the experimental results obtained in the PKL test integral facility in which some SBLOCA sequences have been carried out. Particular useful were the PKL III E experiments data whose results have been numerically reproduced using the code Relap5/MOD3.3/Beta code, contributing to understand the complex thermalhydraulic phenomena related to a PWR boron dilution event. The second field of analysis is related to the effects that possible displacements of un-borated water slugs towards the Reactor Pressure Vessel (RPV) could have on the core reactivity. A numerical approach using the Relap5 reactor kinetics model has been adopted to integrate the experimental thermal - hydraulic data obtained in the PKL III E tests. A careful analysis has been performed in order to establish which core conditions at incident start could produce the largest reactivity increase as a consequence of restarting of natural circulation during the primary system

  8. Proceedings of the ANS/ASME/NRC international topical meeting on nuclear reactor thermal-hydraulics: LMFBR and HTGR advanced reactor concepts and analysis methods

    International Nuclear Information System (INIS)

    1980-01-01

    Separate abstracts are included for each of the papers presented concerning the thermal-hydraulics of LMFBR type reactors; mathematical methods in nuclear reactor thermal-hydraulics; heat transfer in gas-cooled reactors; and thermal-hydraulics of pebble-bed reactors. Two papers have been previously abstracted and input to the data base

  9. Overview of the ANL advanced LMR system thermal-hydraulic test program supporting both GE/PRISM and RI/SAFR

    International Nuclear Information System (INIS)

    Oras, J.J.; Kuzay, T.M.; Kasza, K.E.

    1988-01-01

    Descriptions of the ANL thermal-hydraulic water models of both the PRISM and SAFR reactors are presented, together with results from Phases I and II of the thermal-hydraulic test program. Phenomena discovered during these tests and modeling results are presented. Overall, these efforts demonstrate the acceptable thermal-hydraulic performance of both the PRISM and SAFR concepts

  10. Challenges in thermal and hydraulic analysis of ADS target systems

    International Nuclear Information System (INIS)

    Groetzbach, G.; Batta, A.; Lefhalm, C.-H.; Otic, I.

    2004-01-01

    The liquid metal cooled spallation targets of Accelerator Driven nuclear reactor Systems obey high thermal loads; in addition some flow and cooling conditions are of a prototypical character; in contrast the operating conditions for the engaged materials are narrow; thus, the target development requires a very careful analysis by experimental and numerical means. Especially the cooling of the steel window, which is heated by the proton beam, needs special care. Some of the main goals of the experimental and numerical analyses of the thermal dynamics of those systems are discusses. The prediction of locally detached flows and of flows with larger recirculation areas suffers from insufficient turbulence modeling; this has to be compensated by using prototypical model experiments, e.g. with water, to select the adequate models and numerical schemes. The well known problems with the Reynolds analogy in predicting the heat transfer in liquid metals requires always prototypic liquid metal experiments to select and adapt the turbulent heat flux models. The uncertainties in liquid metal experiments cannot be neglected; so it is necessary to perform CFD calculations and experiments always hand in hand and to develop improve turbulent heat flux models. One contribution to an improved 3 or 4-equation model is deduced from recent Direct Numerical Simulation (DNS) data. (author)

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

    International Nuclear Information System (INIS)

    Jimenez, J.; Herrero, J. J.; Cuervo, D.; Aragones, J. M.

    2010-10-01

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

  12. Thermal-hydraulic experiment and analysis for interim dry storage of spent nuclear fuel

    International Nuclear Information System (INIS)

    Yoo, Seung Hun

    2011-02-01

    The experimental and numerical studies of interim storages for nuclear spent fuels have been performed to investigate thermal-hydraulic characteristics of the dry storage systems and to propose new methodologies for the analysis and the design. Three separate researches have been performed in the present study: (a) Development of a scaling methodology and thermal-hydraulic experiment of a single spent fuel assembly simulating a dry storage cask: (b) Full-scope simulation of a dry storage cask by the use of Computational Fluid Dynamics (CFD) code: (c) Thermal-hydraulic design of a tunnel-type interim storage facility. In the first study, a scaling methodology has been developed to design a scaled-down canister. The scaling was performed in two steps. For the first step, the height of a spent fuel assembly was reduced from full height to half height. In order to consider the effect of height reduction on the natural convection, the scaling law of Ishii and Kataoka (1984) was employed. For the second step, the quantity of spent fuel assemblies was reduced from multiple assemblies to a single assembly. The scaling methodology was validated through the comparison with the experiment of the TN24P cask. The Peak Cladding Temperature (PCT), temperature gradients, and the axial and radial temperature distribution in the nondimensional forms were in good agreement with the experimental data. Based on the developed methodology, we have performed a single assembly experiment which was designed to simulate the full scale of the TN24P cask. The experimental data was compared with the CFD calculations. It turns out that their PCTs were less than the maximum allowable temperature for the fuel cladding and that the differences of their PCTs were agreed within 3 .deg. C, which was less than measurement uncertainty. In the second study, the full-scope simulations of the TN24P cask were performed by FLUENT. In order to investigate the sensitivity of the numerical and physical

  13. Thermal-hydraulics design comparisons for the tandem mirror hybrid reactor blanket

    International Nuclear Information System (INIS)

    Wong, C.P.C.; Yang, Y.S.; Schultz, K.R.

    1980-09-01

    The Tandem Mirror Hybrid Reactor (TMHR) is a cylindrical reactor, and the fertile materials and tritium breeding fuel elements can be arranged with radial or axial orientation in the blanket module. Thermal-hydraulics performance comparisons were made between plate, axial rod and radial rod fuel geometrices. The three configurations result in different coolant/void fractions and different clad/structure fractions. The higher void fraction in the two rod designs means that these blankets will have to be thicker than the plate design blanket in order to achieve the same level of nuclear interactions. Their higher structural fractions will degrade the uranium breeding ratio and energy multiplication factor of the design. One difficulty in the thermal-hydraulics analysis of the plate design was caused by the varying energy multiplication of the blanket during the lifetime of the plate which forced the use of designs that operated in the transition flow regime at some point during life. To account for this, an approach was adopted from Gas Cooled Fast Reactor (GCFR) experience for the pressure drop calculation and the corresponding heat transfer coefficient that was used for the film drop thermal calculation. Because of the superior nuclear performance, the acceptable thermal-hydraulic characteristics and the mechanical design feasibility, the plate geometry concept was chosen for the reference gas-cooled TMHR blanket design

  14. Quality assurance of PTS thermal hydraulic calculations at BNL

    International Nuclear Information System (INIS)

    Rohatgi, U.S.; Pu, J.; Jo, J.; Saha, P.

    1983-01-01

    Rapid cooling of the reactor pressure vessel at high pressure has a potential of challenging the vessel integrity. This phenomenon is called overcooling or Pressurized Thermal Shock (PTS). The Nuclear Regulatory Commission (NRC) has selected three plants representing three types of PWRs in use for detailed PTS study. Oconee-1 (B and W), Calvert Cliffs (C.E.), and H.B. Robinson (Westinghouse). The Brookhaven National Laboratory (BNL) has been requested by NRC to review and compare the input decks developed at LANL and INEL, and to compare and explain the differences between the common calculations performed at these two laboratories. However, for the transients that will be computed by only one laboratory, a consistency check will be performed. So far only Oconee-1 calculations have been reviewed at BNL, and the results are presented here

  15. Thermal hydraulics of accelerator driven system: validation and analysis

    International Nuclear Information System (INIS)

    Kumari, I.; Khanna, A.

    2014-01-01

    This paper presents validation of RELAP5/Mod4.0 code modified to incorporate Lead Bismuth Eutectic (LBE)fluid properties for simulation of Accelerator Driven System (ADS) against Barone's NACIE facility.Results of mass flow rates (MFR), Reynolds number, heat transfer coefficients, temperatures and temperature difference for three powers (10.8, 21.7 and 32.5 kW) under natural circulation of LBE match with Barone's values within 7%,18%,37%, 5% and 8% of relative error respectively. After this validation Indian ADS for thermal power of 15 kW has been simulated. Simulated profiles of temperature, MFR and pressure drop LBE and air are reported. Air and LBE temperatures of present work match with literature design values within 5% of relative error. (author)

  16. A two-compartment thermal-hydraulic experiment (LACE-LA4) analyzed by ESCADRE code

    International Nuclear Information System (INIS)

    Passalacqua, R.

    1994-01-01

    Large scale experiments show that whenever a Loss of Coolant Accident (LOCA) occurs, water pools are generated. Stratifications of steam saturated gas develop above water pools causing a two-compartment thermal-hydraulics. The LACE (LWR Advanced Containment Experiment) LA4 experiment, performed at the Hanford Engineering Development Laboratory (HEDL), exhibited a strong stratification, at all times, above a growing water pool. JERICHO and AEROSOLS-B2 are part of the ESCADRE code system (Ensemble de Systemes de Codes d'Analyse d'accident Des Reacteurs A Eau), a tool for evaluating the response of a nuclear plant to severe accidents. These two codes are here used to simulate respectively the thermal-hydraulics and the associated aerosol behavior. Code results have shown that modelling large containment thermal-hydraulics without taking account of the stratification phenomenon leads to large overpredictions of containment pressure and temperature. If the stratification is modelled as a zone with a higher steam condensation rate and a higher thermal resistance, ESCADRE predictions match quite well experimental data. The stratification thermal-hydraulics is controlled by power (heat fluxes) repartition in the lower compartment between the water pool and the nearby walls. Therefore the total, direct heat exchange between the two compartment is reduced. Stratification modelling is believed to be important for its influence on aerosol behavior: aerosol deposition through the inter-face of the two subcompartments is improved by diffusiophoresis and thermophoresis. In addition the aerosol concentration gradient, through the stratification, will cause a driving force for motion of smaller particles towards the pool. (author)

  17. Thermal Hydraulics Analysis for the 3MW TRIGA MARK-II Research Reactor Under Transient Condition

    International Nuclear Information System (INIS)

    Huda, M.Q.; Bhuiyan, S.I.; Mondal, M.A.W.

    1996-12-01

    Some important thermal hydraulic parameters of the 3 MW TRIGA MARK-II research reactor operating under transient condition were investigated using two computer codes PULTRI and TEMPUL. Major transient parameters, such as, peak power and prompt energy released after pulse, maximum fuel and coolant temperature, surface heat flux, time and radial distribution of temperature within fuel element after pulse, fuel, fuel-cladding gap width variation, etc. were computer and compared with the experimental and operational values as reported in the safety Analysis Report (SAR). It was observed that pulsing of the reactor inserting an excess reactivity of $2.00 shoots the reactor power level to 854.353 MW compared to an experimental value of 852 MW; the maximum fuel temperature corresponding to this peak power was found to be 846.76 o C which is much less than the limiting maximum value of fuel temperature of 1150 0 C as reported in SAR. During a pulse if the film boiling occurs for a peak adiabatic fuel temperature of 1000 o C, the calculated outer cladding wall temperature was observed to be 702.39 0 C compared to a value of 760 o C reported in SAR under the same condition. The investigated other results were also found to be in good agreement with the values reported in the SAR. 16 refs., 22 figs. (author)

  18. Parallel Computing Characteristics of Two-Phase Thermal-Hydraulics code, CUPID

    International Nuclear Information System (INIS)

    Lee, Jae Ryong; Yoon, Han Young

    2013-01-01

    Parallelized CUPID code has proved to be able to reproduce multi-dimensional thermal hydraulic analysis by validating with various conceptual problems and experimental data. In this paper, the characteristics of the parallelized CUPID code were investigated. Both single- and two phase simulation are taken into account. Since the scalability of a parallel simulation is known to be better for fine mesh system, two types of mesh system are considered. In addition, the dependency of the preconditioner for matrix solver was also compared. The scalability for the single-phase flow is better than that for two-phase flow due to the less numbers of iterations for solving pressure matrix. The CUPID code was investigated the parallel performance in terms of scalability. The CUPID code was parallelized with domain decomposition method. The MPI library was adopted to communicate the information at the interface cells. As increasing the number of mesh, the scalability is improved. For a given mesh, single-phase flow simulation with diagonal preconditioner shows the best speedup. However, for the two-phase flow simulation, the ILU preconditioner is recommended since it reduces the overall simulation time

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

    Science.gov (United States)

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

    1988-01-01

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

  20. Thermal-hydraulic effects of transition to improved System 80TM fuel

    International Nuclear Information System (INIS)

    Rodack, T.; Joffre, P.F.; Kapoor, R.K.

    2004-01-01

    ABB CE's improved System 80 TM PWR fuel design includes GUARDIAN debris-resistant features and laser-welded Zircaloy grids. The GUARDIAN features include an Inconel grid with debris-filtering features located just above the Lower End Fitting, and a solid fuel rod bottom end cap that extends above the filtering features. Tests and analyses were done to establish the impact of these design improvements on fuel assembly hydraulic performance. Further analysis was done to determine the mixed core thermal-hydraulic performance as the transition is made over two fuel cycles to a full core of the improved System 80 TM fuel. Results confirm that the Thermal-Hydraulic (T-H) effects of the reduction in hydraulic resistance between the improved and resident fuel due to the laser-welded Zircaloy grids offsets the effects of the increased resistance GUARDIAN grid. Therefore, the mechanically improved System 80 TM fuel can be implemented with no net impact on Departure from Nucleate Boiling (DNB) margin in transition cores. (author)

  1. Proceedings of the 10th international topical meeting on nuclear thermal hydraulics, operation and safety (NUTHOS-10)

    International Nuclear Information System (INIS)

    2014-01-01

    The 10th International Topical Meeting on Nuclear Thermal Hydraulics, Operations and Safety (NUTHOS-10) in Okinawa, Japan is sponsored by Atomic Energy Society of Japan, in cooperation with the International Atomic Energy Agency, and co-sponsored by American Nuclear Society Thermal Hydraulics Division among others. Enhanced safety and reducing cost are going together, which can be achieved through continued research and development efforts. NUTHOS keeps you abreast of the most updated information in the advancement of science and technology in nuclear thermal hydraulics, operations and safety, and provides you insights into the future. (J.P.N.)

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

    Directory of Open Access Journals (Sweden)

    Yohan Cha

    2016-09-01

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

  3. Spent-fuel pool thermal hydraulics: The evaporation question

    International Nuclear Information System (INIS)

    Yilmaz, T.P.; Lai, J.C.

    1996-01-01

    Many nuclear power plants are currently using dense fuel arrangements that increase the number of spent fuel elements stored in their spent-fuel pools (SFPs). The denser spent-fuel storage results in higher water temperatures, especially when certain event scenarios are analyzed. In some of these event scenarios, it is conservative to maximize the evaporation rate, while in other circumstances it is required to minimize the evaporation rates for conservatism. Evaporation is such a fundamental phenomenon that many branches of engineering developed various equations based on theory and experiments. The evaporation rates predicted by existing equations present a wide range of variation, especially at water temperatures >40 degrees C. Furthermore, a study on which equations provide the highest and lowest evaporation rates has not been done until now. This study explores the sensitivity of existing evaporation equations to various parameters and recommends the limiting evaporation equations for use in the solution of SFP thermal problems. Note that the results of this study may be applicable to a much wider range of applications from irrigation ponds, cooling lakes, and liquid-waste management to calculating adequate air exchange rate for swimming pools and health spas

  4. Thermal-hydraulic and characteristic models for packed debris beds

    International Nuclear Information System (INIS)

    Mueller, G.E.; Sozer, A.

    1986-12-01

    APRIL is a mechanistic core-wide meltdown and debris relocation computer code for Boiling Water Reactor (BWR) severe accident analyses. The capabilities of the code continue to be increased by the improvement of existing models. This report contains information on theory and models for degraded core packed debris beds. The models, when incorporated into APRIL, will provide new and improved capabilities in predicting BWR debris bed coolability characteristics. These models will allow for a more mechanistic treatment in calculating temperatures in the fluid and solid phases in the debris bed, in determining debris bed dryout, debris bed quenching from either top-flooding or bottom-flooding, single and two-phase pressure drops across the debris bed, debris bed porosity, and in finding the minimum fluidization mass velocity. The inclusion of these models in a debris bed computer module will permit a more accurate prediction of the coolability characteristics of the debris bed and therefore reduce some of the uncertainties in assessing the severe accident characteristics for BWR application. Some of the debris bed theoretical models have been used to develop a FORTRAN 77 subroutine module called DEBRIS. DEBRIS is a driver program that calls other subroutines to analyze the thermal characteristics of a packed debris bed. Fortran 77 listings of each subroutine are provided in the appendix

  5. A Hydraulic Stress Measurement System for Deep Borehole Investigations

    Science.gov (United States)

    Ask, Maria; Ask, Daniel; Cornet, Francois; Nilsson, Tommy

    2017-04-01

    Luleå University of Technology (LTU) is developing and building a wire-line system for hydraulic rock stress measurements, with funding from the Swedish Research Council and Luleå University of Technology. In this project, LTU is collaborating with University of Strasbourg and Geosigma AB. The stress state influences drilling and drillability, as well as rock mass stability and permeability. Therefore, knowledge about the state of in-situ stress (stress magnitudes, and orientations) and its spatial variation with depth is essential for many underground rock engineering projects, for example for underground storage of hazardous material (e.g. nuclear waste, carbon dioxide), deep geothermal exploration, and underground infrastructure (e.g. tunneling, hydropower dams). The system is designed to conduct hydraulic stress testing in slim boreholes. There are three types of test methods: (1) hydraulic fracturing, (2) sleeve fracturing and (3) hydraulic testing of pre-existing fractures. These are robust methods for determining in situ stresses from boreholes. Integration of the three methods allows determination of the three-dimensional stress tensor and its spatial variation with depth in a scientific unambiguously way. The stress system is composed of a downhole and a surface unit. The downhole unit consists of hydraulic fracturing equipment (straddle packers and downhole imaging tool) and their associated data acquisition systems. The testing system is state of the art in several aspects including: (1) Large depth range (3 km), (2) Ability to test three borehole dimensions (N=76 mm, H=96 mm, and P=122 mm), (3) Resistivity imager maps the orientation of tested fracture; (4) Highly stiff and resistive to corrosion downhole testing equipment; and (5) Very detailed control on the injection flow rate and cumulative volume is obtained by a hydraulic injection pump with variable piston rate, and a highly sensitive flow-meter. At EGU General Assembly 2017, we would like to

  6. 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.

  7. Thermal-hydraulic and thermo-mechanical design of plasma facing components for SST-1 tokamak

    International Nuclear Information System (INIS)

    Chaudhuri, Paritosh; Santra, P.; Chenna Reddy, D.; Parashar, S.K.S.

    2014-01-01

    The Plasma Facing Components (PFCs) are one of the major sub-systems of ssT-1 tokamak. PFC of ssT-1 consisting of divertors, passive stabilizers, baffles and limiters are designed to be compatible for steady state operation. The main consideration in the design of the PFC cooling is the steady state heat removal of up to 1 MW/m 2 . The PFC has been designed to withstand the peak heat fluxes and also without significant erosion such that frequent replacement of the armor is not necessary. Design considerations included 2-D steady state and transient tile temperature distribution and resulting thermal loads in PFC during baking, and cooling, coolant parameters necessary to maintain optimum thermal-hydraulic design, and tile fitting mechanism. Finite Element (FE) models using ANSYS have been developed to carry out the heat transfer and stress analyses of the PFC to understand its thermal and mechanical behaviors. The results of the calculation led to a good understanding of the coolant flow behavior and the temperature distribution in the tube wall and the different parts of the PFC. Thermal analysis of the PFC is carried out with the purpose of evaluating the thermal mechanical behavior of PFCs. The detailed thermal-hydraulic and thermo-mechanical designs of PFCs of ssT-1 are discussed in this paper. (authors)

  8. Optimized iteration in coupled Monte-Carlo - Thermal-hydraulics calculations

    International Nuclear Information System (INIS)

    Hoogenboom, J.E.; Dufek, J.

    2013-01-01

    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 methods are also tested and it is concluded that the presented iteration method is near optimal. (authors)

  9. Thermal-hydraulic and aerosol containment phenomena modelling in ASTEC severe accident computer code

    International Nuclear Information System (INIS)

    Kljenak, Ivo; Dapper, Maik; Dienstbier, Jiri; Herranz, Luis E.; Koch, Marco K.; Fontanet, Joan

    2010-01-01

    Transients in containment systems of different scales (Phebus.FP containment, KAEVER vessel, Battelle Model Containment, LACE vessel and VVER-1000 nuclear power plant containment) involving thermal-hydraulic phenomena and aerosol behaviour, were simulated with the computer integral code ASTEC. The results of the simulations in the first four facilities were compared with experimental results, whereas the results of the simulated accident in the VVER-1000 containment were compared to results obtained with the MELCOR code. The main purpose of the simulations was the validation of the CPA module of the ASTEC code. The calculated results support the applicability of the code for predicting in-containment thermal-hydraulic and aerosol phenomena during a severe accident in a nuclear power plant.

  10. A generalized interface module for the coupling of spatial kinetics and thermal-hydraulics codes

    Energy Technology Data Exchange (ETDEWEB)

    Barber, D.A.; Miller, R.M.; Joo, H.G.; Downar, T.J. [Purdue Univ., West Lafayette, IN (United States). Dept. of Nuclear Engineering; Wang, W. [SCIENTECH, Inc., Rockville, MD (United States); Mousseau, V.A.; Ebert, D.D. [Nuclear Regulatory Commission, Washington, DC (United States). Office of Nuclear Regulatory Research

    1999-03-01

    A generalized interface module has been developed for the coupling of any thermal-hydraulics code to any spatial kinetics code. The coupling scheme was designed and implemented with emphasis placed on maximizing flexibility while minimizing modifications to the respective codes. In this design, the thermal-hydraulics, general interface, and spatial kinetics codes function independently and utilize the Parallel Virtual Machine software to manage cross-process communication. Using this interface, the USNRC version of the 3D neutron kinetics code, PARCX, has been coupled to the USNRC system analysis codes RELAP5 and TRAC-M. RELAP5/PARCS assessment results are presented for two NEACRP rod ejection benchmark problems and an NEA/OECD main steam line break benchmark problem. The assessment of TRAC-M/PARCS has only recently been initiated, nonetheless, the capabilities of the coupled code are presented for a typical PWR system/core model.

  11. A generalized interface module for the coupling of spatial kinetics and thermal-hydraulics codes

    International Nuclear Information System (INIS)

    Barber, D.A.; Miller, R.M.; Joo, H.G.; Downar, T.J.; Mousseau, V.A.; Ebert, D.D.

    1999-01-01

    A generalized interface module has been developed for the coupling of any thermal-hydraulics code to any spatial kinetics code. The coupling scheme was designed and implemented with emphasis placed on maximizing flexibility while minimizing modifications to the respective codes. In this design, the thermal-hydraulics, general interface, and spatial kinetics codes function independently and utilize the Parallel Virtual Machine software to manage cross-process communication. Using this interface, the USNRC version of the 3D neutron kinetics code, PARCX, has been coupled to the USNRC system analysis codes RELAP5 and TRAC-M. RELAP5/PARCS assessment results are presented for two NEACRP rod ejection benchmark problems and an NEA/OECD main steam line break benchmark problem. The assessment of TRAC-M/PARCS has only recently been initiated, nonetheless, the capabilities of the coupled code are presented for a typical PWR system/core model

  12. Fuel management service for Tarapur Atomic Power Station core thermal hydraulics

    International Nuclear Information System (INIS)

    Saha, D.; Venkat Raj, V.; Markandeya, S.G.

    1977-01-01

    Core thermal hydraulic analysis forms an integral part of the fuel management service for the Tarapur reactors. A distinguishing feature of boiling water reactors is the dependence of core flow distribution on the power distribution. Because of the changes in the axial and radial power distribution from cycle to cycle as well as during the cycle and also the variations in leakage flow, it is necessary to evaluate the core thermal hydraulic parameters for every cycle. Some of the typical results obtained in the course of analysis for different cycles of both the units at Tarapur are presented. The use of MCPR (Minimum Critical Power Ratio), instead of MCHFR (Minimum Critical Heat Flux Ratio) as a figure of merit for fuel cladding integrity is also discussed. (K.B.)

  13. Thermal hydraulic analyses of LVR-15 research reactor with IRT-M fuel

    International Nuclear Information System (INIS)

    Macek, J.

    1997-01-01

    The LVR-15 pool-type research reactor has been in operation at the Nuclear Research Institute at Rez since 1955. Following a number of reconstructions and redesigning, the current reactor power is 15 MW. Thermal hydraulic analyses to demonstrate that the core heat will be safely removed during operation as well as in accident situations were performed based on methodology which had been specifically developed for the LVR-15 research reactor. This methodology was applied to stationary thermal hydraulic computations, as well as to transients, particularly with reactivity failure and loss of circulation pumps emergencies. The applied methodology and the core configuration as used in the Safety Report are described. The initial and boundary conditions are then considered and the summary of the calculated failures with regard to the defined safety limits is presented. The results of the core configuration analyses are also discussed with respect to meeting the safety limits and to the applicability of the methodology to this purpose

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

    International Nuclear Information System (INIS)

    Yan, B.H.

    2017-01-01

    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.

  15. 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.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2017-08-15

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

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

    International Nuclear Information System (INIS)

    Tanaka, F.; Hibiki, T.; Saito, Y.; Takeda, T.; Mishima, K.

    2001-01-01

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

  18. 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.

  19. 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.

  20. Development of steady thermal-hydraulic analysis code for China advanced research reactor

    International Nuclear Information System (INIS)

    Tian Wenxi; Qiu Suizheng; Guo Yun; Su Guanghui; Jia Dounan; Liu Tiancai; Zhang Jianwei

    2006-01-01

    A multi-channel model steady-state thermal-hydraulic analysis code was developed for China Advanced Research Reactor (CARR). By simulating the whole reactor core, the detailed flow distribution in the core was obtained. The result shows that the structure size plays the most important role in flow distribution and the influence of core power could be neglected under single-phase flow. The temperature field of fuel element under unsymmetrical cooling condition was also obtained, which is necessary for the 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 hot channel was carried out and it is proved that all thermal-hydraulic parameters accord with the Safety Regulation of CARR. (authors)

  1. Current status and future prospects for thermal-hydraulics and safety research

    International Nuclear Information System (INIS)

    Park, G.C.

    2000-01-01

    The present paper is to outline the current activities in Korea for the thermal-hydraulics and safety researches, and furthermore illuminate the future aspect of those field under the umbrella of worldwide nuclear prospect. In Korea, a long-term nuclear research plan has been established since 1992, which was recently funded with a fixed monetary rate of Korean won 1.20 per kWh of electricity produced with nuclear power. 11.5% of the fund is assigned for nuclear safety research in 6 areas. Under this program, 3 axes of research body (KAERI, KINS, University) has been operated with close cooperation. Their role, current activities and long-term plan of each body are introduced in the point of thermal-hydraulics' view. (author)

  2. TEMP-M program for thermal-hydraulic calculation of fast reactor fuel assemblies

    International Nuclear Information System (INIS)

    Bogoslovskaya, C.P.; Sorokin, A.P.; Tikhomirov, B.B.; Titov, P.A.; Ushakov, P.A.

    1983-01-01

    TEMP-M program (Fortran, BESM-6 computer) for thermal-hydraulic calculation of fast reactor fuel assemblies is described. Results of calculation of temperature field in a 127 fuel element assembly of BN-600, reactor accomplished according to TEMP-N program are considered as an example. Algorithm, realized in the program, enables to calculate the distributions of coolant heating, fuel element temperature (over perimeter and length) and assembly shell temperature. The distribution of coolant heating in assembly channels is determined from a solution of the balance equation system which accounts for interchannel exchange, nonadiabatic conditions on the assembly shell. The TEMP-M program gives necessary information for calculation of strength, seviceability of fast reactor core elements, serves an effective instrument for calculations when projecting reactor cores and analyzing thermal-hydraulic characteristics of operating reactor fuel assemblies

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

    International Nuclear Information System (INIS)

    Banerjee, S.; Hassan, Y.A.

    1995-01-01

    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

  4. Thermal Hydraulic Assessment for Loss of SDCS Event During the Outage of CANDU Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Jonghyun [Gnest, Inc. Taejon (Korea, Republic of); Lee, Kwangho; Oh, Haechol; Jun, Hwangyong [KEPRI, Taejon (Korea, Republic of)

    2006-07-01

    During the outage(overhaul) of the nuclear power plant, there are several operating states other than the full power state, that is 'Hot-Zero Power', 'Depressurized-Cooldown', and 'Partially Drained'. Until now safety assessment has not been done much for this operating state of CANDU type reactor worldwide. For the accuracy and confidence of PSA for the CANDU outage, the safety analysis is necessary. At the first stage, we analyzed the thermal hydraulic characteristics and safety of the postulated event of loss of shutdown cooling system (SDCS) during the partially drained state which is the longest one in the middle of outage period. As an analysis tool, this study uses the best estimate thermal hydraulic code, RELAP5/CANDU which was modified according to the CANDU specific characteristics and based on RELAP5.Mod3.

  5. Influence of geometrical and thermal hydraulic parameters on the short term containment system response

    International Nuclear Information System (INIS)

    Krishna Chandran, R.; Ali, Seik Mansoor; Balasubramaniyan, V.

    2014-01-01

    This paper discusses the effect of a number of geometrical and thermal hydraulic parameters on the containment peak pressure following a simulated LOCA. The numerical studies are carried out using an inhouse containment thermal hydraulics program called 'THYCON' with focus only on the short term transient response. In order to highlight the effect of above variables, a geometrically scaled (1:270) model of a typical 220 MWe Indian PHWR containment is considered. The discussions in this paper are limited to explaining the influence of individual parameters by comparing with a base case value. It is essential to mention that the results presented here are not general and should be taken as indicative only. Nevertheless, these numerical studies give insight into short term containment response that would be useful to both the system designer as well as the regulator. (author)

  6. Thermal-hydraulic software development for nuclear waste transportation cask design and analysis

    International Nuclear Information System (INIS)

    Brown, N.N.; Burns, S.P.; Gianoulakis, S.E.; Klein, D.E.

    1991-01-01

    This paper describes the development of a state-of-the-art thermal-hydraulic software package intended for spent fuel and high-level nuclear waste transportation cask design and analysis. The objectives of this software development effort are threefold: (1) to take advantage of advancements in computer hardware and software to provide a more efficient user interface, (2) to provide a tool for reducing inefficient conservatism in spent fuel and high-level waste shipping cask design by including convection as well as conduction and radiation heat transfer modeling capabilities, and (3) to provide a thermal-hydraulic analysis package which is developed under a rigorous quality assurance program established at Sandia National Laboratories. 20 refs., 5 figs., 2 tabs

  7. RAMONA-3B/MINET composite representation of BWR thermal-hydraulic systems

    International Nuclear Information System (INIS)

    Van Tuyle, G.J.; Slovik, G.; Cazzoli, E.G.; Nepsee, T.C.; Guppy, J.G.

    1985-01-01

    The modification and interfacing of two computer codes, RAMONA-3B and MINET, for the thermal hydraulic transient analysis of a Boiling Water Reactor nuclear steam supply system, is described. The RAMONA-3B code provides for multi-channel thermal hydraulics and three-dimensional (or one-dimensional) neutron kinetics analysis of a boiling water reactor core. The RAMONA-3B system representation terminates at the end of the steam line and at the junction of the feedwater line at the vessel inlet. By interfacing RAMONA-3B with MINET, a generic balance-of-plant systems analysis code, a complete BWR systems code with detailed core modeling was obtained. The result is a code of particular importance to the analysis of transients such as ATWS. A comparison between the 3-D and 1-D neutronics representation is provided, along with a test case utilizing the composite RAMONA-3B/MINET code

  8. 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.

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

    International Nuclear Information System (INIS)

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

    1997-01-01

    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

  10. ATHENA [Advanced Thermal Hydraulic Energy Network Analyzer] solutions to developmental assessment problems

    International Nuclear Information System (INIS)

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

    1987-03-01

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

  11. ATWS thermal-hydraulic analysis for Krsko Full Scope Simulator validation

    International Nuclear Information System (INIS)

    Parzer, I.; Kljenak, I.

    2005-01-01

    The purpose of this analysis was to simulate Anticipated Transient without Scram transient for Krsko NPP. The results of these calculations were used for annual ANSI/ANS validation of reactor coolant system thermal-hydraulic response predicted by Krsko Full Scope Simulator. For the thermal-hydraulic analyses the RELAP5/MOD3.3 code and the input model for NPP Krsko, delivered by NPP Krsko, was used. In the presented paper the most severe ATWS scenario has been analyzed, starting with the loss of Main Feedwater at both steam generators. Thus, gradual loss of secondary heat sink occurred. On top of that, control rods were not supposed to scram, leaving the chain reaction to be controlled only by inherent physical properties of the fuel and moderator and eventual actions of the BOP system. The primary system response has been studied assuming AMSAC availability. (author)

  12. Full scale mock-up tests for rod bundle thermal-hydraulics in Japan

    International Nuclear Information System (INIS)

    Sugawara, S.

    1995-01-01

    This poster describes tests aimed at development and validation of principal design methodology of rod bundle thermal-hydraulics correlations. The works are based on domestic data base using the full-scale mock-up test facilities. The scope of the tests comprises DNB heat flux, transient DNB heat flux, post DNB heat transfer, pressure drop and void distribution. The works have been performed under collaboration among electric facilities, NPP vendors, universities, governmental corporations. 1 tab., 14 figs

  13. 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.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2017-07-01

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

  15. Thermal hydraulic analysis of Pb-Bi cooled HYPER fuel assemblies using SLTHEN code

    International Nuclear Information System (INIS)

    Tak, Nam Il; Song, Tae Y.; Park, Won S.; Kim, Chang Hyun

    2002-12-01

    In the present work, the existing SLTHEN code, which had been originally developed for subchannel analysis of sodium cooled fast reactors, was modified and applied to the Pb-Bi cooled HYPER core which consists of 237 fuel assemblies (TRU assemblies). In the analysis of single fuel assembly having chopped cosine power profile, the validation and the assessment of usefulness of the modified SLTHEN were focused. In the quantitative comparison, the results of the modified SLTHEN agreed well with those of analytical calculations and of MATRA. For the qualitative approaches, the sensitivity calculations for intra-assembly gap flow and turbulent mixing parameter were used. The sensitivity analysis results showed that the modified SLTHEN can provide reasonable simulations of subchannel thermal hydraulics. In particular, turbulent mixing parameter which is known as the most uncertain parameter in subchannel analyses did not affect largely the maximum cladding temperature. Therefore, it can be said that the results of single assembly show the usefulness of the modified SLTHEN code for thermal hydraulic analysis and design of HYPER under the conceptual design stage. In order to assess intra-assembly heat transfer, subchannel analyses were implemented for two types of 7 assemblies; 1) artificial 7 fuel assemblies to maximize intra-assembly heat transfer, 2) central 7 fuel assemblies in the HYPER reference core. The results showed that the modified SLTHEN can reasonably simulate intra-heat transfer and the amount of intra-assembly heat transfer is not so large in HYPER conditions. Particularly, intra-heat transfer did not affect the maximum coolant and the maximum cladding temperatures which are major parameters in conceptual core designs. The capability of full core thermal hydraulic analysis was confirmed by the analysis of 45 fuel assemblies in 1/6 HYPER core at the first cycle. The SLTHEN predicted that the reference design parameters are acceptable in terms of thermal

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

    International Nuclear Information System (INIS)

    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

  17. The thermal-hydraulic for the new technologies: the micro-fluidics

    International Nuclear Information System (INIS)

    Crecy, F. de; Gruss, A.; Bricard, A.; Excoffon, J.

    2000-01-01

    The micro-fluidics can be defined as the fluid flow in little canals. This scale offers a great interest for the biotechnology type. In this paper, the authors present this fluids form and detail the researches performed at the Department of Physics and Thermal-hydraulics of the CEA, in the domain of the physical properties characterization and of the numerical two-phase direct simulation. (A.L.B.)

  18. Power transients of Ghana research reactor-1 using PARET/ANL thermal hydraulic code

    International Nuclear Information System (INIS)

    Ampomah-Amoaka, E.; Akaho, E.H.K.; Anim-Sampong, S.; Nyarko, B.J.B.

    2010-01-01

    PARET/ANL(Version 7.3 of 2007) thermal-hydraulic code was used to perform transient analysis of the Ghana Research Reactor-1.The reactivities inserted were 2.1mk and 4mk.The peak power of 5.81kW was obtained for 2.1 mk insertion whereas the peak power for 4mk insertion of reactivity was 92.32kW.These results compare closely with experiments and theoretical studies conducted previously.

  19. Computer code for the thermal-hydraulic analysis of ITU TRIGA Mark-II reactor

    International Nuclear Information System (INIS)

    Ustun, G.; Durmayaz, A.

    2002-01-01

    Istanbul Technical University (ITU) TRIGA Mark-II reactor core consists of ninety vertical cylindrical elements located in five rings. Sixty-nine of them are fuel elements. The reactor is operated and cooled with natural convection by pool water, which is also cooled and purified in external coolant circuits by forced convection. This characteristic leads to consider both the natural and forced convection heat transfer in a 'porous-medium analysis'. The safety analysis of the reactor requires a thermal-hydraulic model of the reactor to determine the thermal-hydraulic parameters in each mode of operation. In this study, a computer code cooled TRIGA-PM (TRIGA - Porous Medium) for the thermal-hydraulic analysis of ITU is considered. TRIGA Mark-II reactor code has been developed to obtain velocity, pressure and temperature distributions in the reactor pool as a function of core design parameters and pool configuration. The code is a transient, thermal-hydraulic code and requires geometric and physical modelling parameters. In the model, although the reactor is considered as only porous medium, the other part of the reactor pool is considered partly as continuum and partly as porous medium. COMMIX-1C code is used for the benchmark purpose of TRIGA-PM code. For the normal operating conditions of the reactor, estimations of TRIGA-PM are in good agreement with those of COMMIX-1C. After some more improvements, this code will be employed for the estimation of LOCA scenario, which can not be analyses by COMMIX-1C and the other multi-purpose codes, considering a break at one of the beam tubes of the reactor

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

    International Nuclear Information System (INIS)

    Perez, Daniel Milian; Lorenzo, Daniel E. Milian; Lira, Carlos A. Brayner de Oliveira; Garcia, Lorena P. Rodríguez; Universidade Federal de Pernambuco

    2017-01-01

    The use of Aqueous Homogenous Reactors (AHR) is one of the most promissory alternatives to produce medical isotopes, mainly "9"9Mo. Compare to multipurpose research reactors, an AHR dedicated for "9"9Mo production has advantages because of their low cost, small critical mass, inherent passive safety, and simplified fuel handling, processing, and purification characteristics. This article presents the current state of research in our working group on this topic. Are presented and discussed the group validation efforts with benchmarking exercises that include neutronic and thermal-hydraulic results of two solution reactors, the SUPO and ARGUS reactors. Neutronic and thermal-hydraulic results of 75 kWth AHR based on the ARGUS reactor LEU configuration are presented. The neutronic studies included the determination of parameters such as reflector thickness, critical height, medical isotopes production and others. Thermal-hydraulics studies were focused on demonstrating that sufficient cooling capacity exists to prevent fuel overheating. In addition, the effects of some calculation parameters on the computational modeling of temperature, velocity and gas volume fraction during steady-state operation of an AHR are discussed. The neutronic and thermal-hydraulics studies have been performed with the MCNPX version 2.6e computational code and the version 14 of ANSYS CFX respectively. Our group studies and the results obtained contribute to demonstrate the feasibility of using AHR for the production of medical isotopes, however additional studies are still necessary to confirm these results and contribute to development and demonstration of their technical, safety, and economic viability. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2017-11-01

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

  2. Loss-of-Fluid Test findings in pressurized water reactor core's thermal-hydraulic behavior

    International Nuclear Information System (INIS)

    Russell, M.

    1983-01-01

    This paper summarizes the pressurized water reactor (PWR) core's thermal-hydraulic behavior findings from experiments performed at the Loss-of-Fluid Test (LOFT) Facility at the Idaho National Engineering Laboratory. The potential impact of these findings on the safety and economics of PWR's generation of electricity is also discussed. Reviews of eight important findings in the core's physical behavior and in experimental methods are presented with supporting evidence

  3. Some neutronics and thermal-hydraulics codes for reactor analysis using personal computers

    International Nuclear Information System (INIS)

    Woodruff, W.L.

    1990-01-01

    Some neutronics and thermal-hydraulics codes formerly available only for main frame computers may now be run on personal computers. Brief descriptions of the codes are provided. Running times for some of the codes are compared for an assortment of personal and main frame computers. With some limitations in detail, personal computer versions of the codes can be used to solve many problems of interest in reactor analyses at very modest costs. 11 refs., 4 tabs

  4. 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.

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

    Science.gov (United States)

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

    2014-05-01

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

  6. 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

  7. 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.

  8. Thermal-Hydraulic Analysis of a Once-Through Steam Generator Considering Performance Degradation

    Energy Technology Data Exchange (ETDEWEB)

    Han, Hun Sik; Kang, Han Ok; Yoon, Ju Hyeon; Kim, Young In; Song, Jae Seung; Kim, Keung Koo [KAERI, Daejeon (Korea, Republic of)

    2016-05-15

    Several countries have entered into a global race for the commercialization of SMRs, and considerable research and development have been implemented. Among the various reactor designs, many SMRs have adopted an integral type pressurized water reactor (PWR) to enhance the nuclear safety and system reliability. In the integral reactor design, a single reactor pressure vessel contains primary system components such as fuel and core, steam generators, pumps, and a pressurizer. For the component integration into a reactor vessel, it is important to design each component as small as possible. Thus, it is a common practice to employ a once-through steam generator in the integral reactor design due to its advantages in compactness. In general, gradual degradation in thermal-hydraulic performance of the steam generator occurs with time, and it changes slowly the operating point of the steam generator during plant lifetime. Numerical solutions are acquired to evaluate the thermal-hydraulic performance of the steam generator at various AUFs. The design results obtained show that the average tube length of the steam generator is augmented with the increase of design margin to compensate for the design uncertainties and heat transfer area reduction by plugging, fouling, etc. A helically coiled tube once-through steam generator with 30% design margin is considered for comparison of thermal-hydraulic performances according to the degradation rate.

  9. The analysis of thermal-hydraulic performances of nuclear ship reactor

    International Nuclear Information System (INIS)

    Wakabayashi, Shinshichi; Hamada, Masao

    1975-01-01

    Thermal-hydraulic performances in the core of nuclear ship reactor was analysed by thermal-hydraulic analyser codes, AMRTC and COBRA-11+DNBCAL. This reactor is of a pressurized water type and incorporates the steam generator within the reactor vessel with the rated power of 330 MWt, which is developed by Nuclear Ship Research Panel Seven (NSR-7) in The Shipbuilding Research Association of Japan. Fuel temperature distributions, coolant temperature distributions, void fractions in coolant and minimum burn out ratio etc. were calculated. Results are as follows; a) The maximum temperature of fuel center is 1,472 0 C that corresponds to 53% as small as the melting point (2,800 0 C). b) Subcooled boiling exists in the core and the maximum void fraction is less than 4%. c) The minimum burn out ratio is not less than the minimum allowable limit of 1.25. It was found from the results of analysis that this reactor was able to be operated wide margin with respect to thermal-hydraulic design limits at the rated power. (auth.)

  10. Real time thermal hydraulic model for high temperature gas-cooled reactor core

    International Nuclear Information System (INIS)

    Sui Zhe; Sun Jun; Ma Yuanle; Zhang Ruipeng

    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. (authors)

  11. Thermal-Hydraulic Analysis of a Once-Through Steam Generator Considering Performance Degradation

    International Nuclear Information System (INIS)

    Han, Hun Sik; Kang, Han Ok; Yoon, Ju Hyeon; Kim, Young In; Song, Jae Seung; Kim, Keung Koo

    2016-01-01

    Several countries have entered into a global race for the commercialization of SMRs, and considerable research and development have been implemented. Among the various reactor designs, many SMRs have adopted an integral type pressurized water reactor (PWR) to enhance the nuclear safety and system reliability. In the integral reactor design, a single reactor pressure vessel contains primary system components such as fuel and core, steam generators, pumps, and a pressurizer. For the component integration into a reactor vessel, it is important to design each component as small as possible. Thus, it is a common practice to employ a once-through steam generator in the integral reactor design due to its advantages in compactness. In general, gradual degradation in thermal-hydraulic performance of the steam generator occurs with time, and it changes slowly the operating point of the steam generator during plant lifetime. Numerical solutions are acquired to evaluate the thermal-hydraulic performance of the steam generator at various AUFs. The design results obtained show that the average tube length of the steam generator is augmented with the increase of design margin to compensate for the design uncertainties and heat transfer area reduction by plugging, fouling, etc. A helically coiled tube once-through steam generator with 30% design margin is considered for comparison of thermal-hydraulic performances according to the degradation rate

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

    International Nuclear Information System (INIS)

    Strizhov, V.; Kanukova, V.; Vinogradova, T.; Askenov, E.; Nikulshin, V.

    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. TRSM-a thermal-hydraulic real-time simulation model for PWR

    International Nuclear Information System (INIS)

    Zhou Weichang

    1997-01-01

    TRSM (a Thermal-hydraulic Real-time Simulation Model) has been developed for PWR real-time simulation and best-estimate prediction of normal operating and abnormal accident conditions. It is a non-equilibrium two phase flow thermal-hydraulic model based on five basic conservation equations. A drift flux model is used to account for the unequal velocities of liquid and gaseous mixture, with or without the presence of the noncondensibles. Critical flow models are applied for break flow and valve flow calculations. A 5-regime two phase heat convection model is applied for clad-to-coolant as well as fluid-to-tubing heat transfer. A rigorous reactor coolant pump model is used to calculate the pressure drop and rise for the suction and discharge ends with complete pump characteristics curves included. The TRSM model has been adapted in the full-scale training simulator of Qinshan Nuclear Power Plant 300 MW unit to simulate the thermal-hydraulic performance of the NSSS. The simulation results of a cold leg LOCA and a steam generator tube rupture (SGTR) accident are presented

  14. Development of core thermal-hydraulics module for intelligent reactor design system (IRDS)

    International Nuclear Information System (INIS)

    Kugo, Teruhiko; Nakagawa, Masayuki; Fujii, Sadao.

    1994-08-01

    We have developed an innovative reactor core thermal-hydraulics module where a designer can easily and efficiently evaluate his design concept of a new type reactor in the thermal-hydraulics field. The main purpose of this module is to decide a feasible range of basic design parameters of a reactor core in a conceptual design stage of a new type reactor. The module is to be implemented in Intelligent Reactor Design System (IRDS). The module has the following characteristics; 1) to deal with several reactor types, 2) four thermal hydraulics and fuel behavior analysis codes are installed to treat different type of reactors and design detail, 3) to follow flexibly modification of a reactor concept, 4) to provide analysis results in an understandable way so that a designer can easily evaluate feasibility of his concept, and so on. The module runs on an engineering workstation (EWS) and has a user-friendly man-machine interface on a pre- and post-processing. And it is equipped with a function to search a feasible range called as Design Window, for two design parameters by artificial intelligence (AI) technique and knowledge engineering. In this report, structure, guidance for users of an usage of the module and instruction of input data for analysis modules are presented. (author)

  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. Evaluation of operational safety at Babcock and Wilcox Plants: Volume 2, Thermal-hydraulic results

    International Nuclear Information System (INIS)

    Wheatley, P.D.; Davis, C.B.; Callow, R.A.; Fletcher, C.D.; Dobbe, C.A.; Beelman, R.J.

    1987-11-01

    The Nuclear Regulatory Commission has initiated a research program to develop a methodology to assess the operational performance of Babcock and Wilcox plants and to apply this methodology on a trial basis. The methodology developed for analyzing Babcock and Wilcox plants integrated methods used in both thermal-hydraulics and human factors and compared results with information used in the assessment of risk. The integrated methodology involved an evaluation of a selected plant for each pressurized water reactor vendor during a limited number of transients. A plant was selected to represent each vendor, and three transients were identified for analysis. The plants were Oconee Unit 1 for Babcock and Wilcox, H.B. Robinson Unit 2 for Westinghouse, and Calvert Cliffs Unit 1 for Combustion Engineering. The three transients were a complete loss of all feedwater, a small-break loss-of-coolant accident, and a steam-generator overfill with auxiliary feedwater. Included in the integrated methodology was an assessment of the thermal-hydraulic behavior, including event timing, of the plants during the three transients. Thermal-hydraulic results are presented in this volume (Volume 2) of the report. 26 refs., 30 figs., 7 tabs

  17. Steady-state thermal-hydraulic design analysis of the Advanced Neutron Source reactor

    International Nuclear Information System (INIS)

    Yoder, G.L. Jr.; Dixon, J.R.; Elkassabgi, Y.; Felde, D.K.; Giles, G.E.; Harrington, R.M.; Morris, D.G.; Nelson, W.R.; Ruggles, A.E.; Siman-Tov, M.; Stovall, T.K.

    1994-05-01

    The Advanced Neutron Source (ANS) is a research reactor that is planned for construction at Oak Ridge National Laboratory. This reactor will be a user facility with the major objective of providing the highest continuous neutron beam intensities of any reactor in the world. Additional objectives for the facility include providing materials irradiation facilities and isotope production facilities as good as, or better than, those in the High Flux Isotope Reactor. To achieve these objectives, the reactor design uses highly subcooled heavy water as both coolant and moderator. Two separate core halves of 67.6-L total volume operate at an average power density of 4.5 MW(t)/L, and the coolant flows upward through the core at 25 m/s. Operating pressure is 3.1 MPa at the core inlet with a 1.4-MPa pressure drop through the core region. Finally, in order to make the resources available for experimentation, the fuel is designed to provide a 17-d fuel cycle with an additional 4 d planned in each cycle for the refueling process. This report examines the codes and models used to develop the thermal-hydraulic design for ANS, as well as the correlations and physical data; evaluates thermal-hydraulic uncertainties; reports on thermal-hydraulic design and safety analysis; describes experimentation in support of the ANS reactor design and safety analysis; and provides an overview of the experimental plan

  18. Methodology for thermal-hydraulics analysis of pool type MTR fuel research reactors

    International Nuclear Information System (INIS)

    Umbehaun, Pedro Ernesto

    2000-01-01

    This work presents a methodology developed for thermal-hydraulic analysis of pool type MTR fuel research reactors. For this methodology a computational program, FLOW, and a model, MTRCR-IEAR1 were developed. FLOW calculates the cooling flow distribution in the fuel elements, control elements, irradiators, and through the channels formed among the fuel elements and among the irradiators and reflectors. This computer program was validated against experimental data for the IEA-R1 research reactor core at IPEN-CNEN/SP. MTRCR-IEAR1 is a model based on the commercial program Engineering Equation Solver (EES). Besides the thermal-hydraulic analyses of the core in steady state accomplished by traditional computational programs like COBRA-3C/RERTR and PARET, this model allows to analyze parallel channels with different cooling flow and/or geometry. Uncertainty factors of the variables from neutronic and thermalhydraulic calculation and also from the fabrication of the fuel element are introduced in the model. For steady state analyses MTRCR-IEAR1 showed good agreement with the results of COBRA-3C/RERTR and PARET. The developed methodology was used for the calculation of the cooling flow distribution and the thermal-hydraulic analysis of a typical configuration of the IEA-R1 research reactor core. (author)

  19. Validation of thermal hydraulic computer codes for advanced light water reactor

    International Nuclear Information System (INIS)

    Macek, J.

    2001-01-01

    The Czech Republic operates 4 WWER-440 units, two WWER-1000 units are being finalised (one of them is undergoing commissioning). Thermal-hydraulics Department of the Nuclear Research Institute Rez performs accident analyses for these plants using a number of computer codes. To model the primary and secondary circuits behaviour the system codes ATHLET, CATHARE, RELAP, TRAC are applied. Containment and pressure-suppressure system are modelled with RALOC and MELCOR codes, the reactor power calculations (point and space-neutron kinetics) are made with DYN3D, NESTLE and CDF codes (FLUENT, TRIO) are used for some specific problems. An integral part of the current Czech project 'New Energy Sources' is selection of a new nuclear source. Within this and the preceding projects financed by the Czech Ministry of Industry and Trade and the EU PHARE, the Department carries and has carried out the systematic validation of thermal-hydraulic and reactor physics computer codes applying data obtained on several experimental facilities as well as the real operational data. The paper provides a concise information on these activities of the NRI and its Thermal-hydraulics Department. A detailed example of the system code validation and the consequent utilisation of the results for a real NPP purposes is included. (author)

  20. Comparative performance of the conjugate gradient and SOR [Successive Over Relaxation] methods for computational thermal hydraulics

    International Nuclear Information System (INIS)

    King, J.B.; Anghaie, S.; Domanus, H.M.

    1987-01-01

    Finite difference approximations to the continuity, momentum, and energy equations in thermal hydraulics codes result in a system of N by N equations for a problem having N field points. In a three dimensional problem, N increases as the problem becomes larger or more complex, and more rapidly as the computational mesh size is reduced. As a consequence, the execution time required to solve the problem increases, which may lead to placing limits on the problem resolution or accuracy. A conventinal method of solution of these systems of equations is the Successive Over Relaxation (SOR) technique. However, for a wide range of problems the execution time may be reduced by using a more efficient linear equation solver. One such method is the conjugate gradient method which was implemented in COMMIX-1B thermal hydraulics code. It was found that the execution time required to solve the resulting system of equations was reduced by a factor of about 2 for some problems. This paper summarizes the characteristics of these iterative solution procedures and compares their performance in modeling of a variety of reactor thermal hydraulic problems, using the COMMIX-1B computer code

  1. The International Experimental Thermal Hydraulic Systems database – TIETHYS: A new NEA validation tool

    Energy Technology Data Exchange (ETDEWEB)

    Rohatgi, Upendra S.

    2018-07-22

    Nuclear reactor codes require validation with appropriate data representing the plant for specific scenarios. The thermal-hydraulic data is scattered in different locations and in different formats. Some of the data is in danger of being lost. A relational database is being developed to organize the international thermal hydraulic test data for various reactor concepts and different scenarios. At the reactor system level, that data is organized to include separate effect tests and integral effect tests for specific scenarios and corresponding phenomena. The database relies on the phenomena identification sections of expert developed PIRTs. The database will provide a summary of appropriate data, review of facility information, test description, instrumentation, references for the experimental data and some examples of application of the data for validation. The current database platform includes scenarios for PWR, BWR, VVER, and specific benchmarks for CFD modelling data and is to be expanded to include references for molten salt reactors. There are place holders for high temperature gas cooled reactors, CANDU and liquid metal reactors. This relational database is called The International Experimental Thermal Hydraulic Systems (TIETHYS) database and currently resides at Nuclear Energy Agency (NEA) of the OECD and is freely open to public access. Going forward the database will be extended to include additional links and data as they become available. https://www.oecd-nea.org/tiethysweb/

  2. Hydraulic and thermal conduction phenomena in soils at the particle-scale: Towards realistic FEM simulations

    International Nuclear Information System (INIS)

    Narsilio, G A; Yun, T S; Kress, J; Evans, T M

    2010-01-01

    This paper summarizes a method to characterize conduction properties in soils at the particle-scale. The method set the bases for an alternative way to estimate conduction parameters such as thermal conductivity and hydraulic conductivity, with the potential application to hard-to-obtain samples, where traditional experimental testing on large enough specimens becomes much more expensive. The technique is exemplified using 3D synthetic grain packings generated with discrete element methods, from which 3D granular images are constructed. Images are then imported into the finite element analyses to solve the corresponding governing partial differential equations of hydraulic and thermal conduction. High performance computing is implemented to meet the demanding 3D numerical calculations of the complex geometrical domains. The effects of void ratio and inter-particle contacts in hydraulic and thermal conduction are explored. Laboratory measurements support the numerically obtained results and validate the viability of the new methods used herein. The integration of imaging with rigorous numerical simulations at the pore-scale also enables fundamental observation of particle-scale mechanisms of macro-scale manifestation.

  3. Thermal-hydraulic performance of convective boiling jet array impingement

    International Nuclear Information System (INIS)

    Jenkins, R; De Brún, C; Kempers, R; Lupoi, R; Robinson, A J

    2016-01-01

    Jet impingement boiling is investigated with regard to heat transfer and pressure drop performance using a novel laser sintered 3D printed jet impingement manifold design. Water was the working fluid at atmospheric pressure with inlet subcooling of 7 o C. The convective boiling performance of the impinging jet system was investigated for a flat copper target surface for 2700≤Re≤5400. The results indicate that the heat transfer performance of the impinging jet is independent of Reynolds number for fully developed boiling. Also, the investigation of nozzle to plate spacing shows that low spacing delays the onset of nucleate boiling causing a superheat overshoot that is not observed with larger gaps. However, no sensitivity to the gap spacing was measured once boiling was fully developed. The assessment of the pressure drop performance showed that the design effectively transfers heat with low pumping power requirements. In particular, owing to the insensitivity of the heat transfer to flow rate during fully developed boiling, the coefficient of performance of jet impingement boiling in the fully developed boiling regime deteriorates with increased flow rate due to the increase in pumping power flux. (paper)

  4. Thermal-hydraulic analyses of pressurized-thermal-shock-induced vessel ruptures

    International Nuclear Information System (INIS)

    Dobranich, D.

    1982-05-01

    A severe overcooling transient was postulated to produce vessel wall temperatures below the nil-ductility transition temperature which in conjunction with system repressurization, led to vessel rupture at the core midplane. Such transients are referred to as pressurized-thermal-shock transients. A wide range of vessel rupture sizes were investigated to assess the emergency system's ability to cool the fuel rods. Ruptures greater than approximately 0.015 m 2 produced flows greater than those of the emergency system and resulted in core uncovery and subsequent core damage

  5. 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.

  6. COBRA/TRAC analysis of two-dimensional thermal-hydraulic behavior in SCTF reflood tests

    International Nuclear Information System (INIS)

    Iwamura, Takamichi; Ohnuki, Akira; Sobajima, Makoto; Adachi, Hiromichi

    1987-01-01

    The effects of radial power distribution and non-uniform upper plenum water accumulation on thermal-hydraulic behavior in the core were observed in the reflood tests with Slab Core Test Facility (SCTF). In order to examine the predictability of these two effects by a multi-dimensional analysis code, the COBRA/TRAC calculations were made. The calculated results indicated that the heat transfer enhancement in high power bundles above quench front was caused by high vapor flow rate in those bundles due to the radial power distribution. On the other hand, the heat transfer degradation in the peripheral bundles under the condition of non-uniform upper plenum water accumulation was caused by the lower flow rates of vapor and entrained liquid above the quench front in those bundles by the reason that vapor concentrated in the center bundles due to the cross flow induced by the horizontal pressure gradient in the core. The above-mentioned two-dimensional heat transfer behaviors calculated with the COBRA/TRAC code is similar to those observed in SCTF tests and therefore those calculations are useful to investigate the mechanism of the two-dimensional effects in SCTF reflood tests. (author)

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

    International Nuclear Information System (INIS)

    Lue, J.W.; Dresner, L.

    1993-01-01

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

  8. A fifth equation to model the relative velocity the 3-D thermal-hydraulic code THYC

    International Nuclear Information System (INIS)

    Jouhanique, T.; Rascle, P.

    1995-11-01

    E.D.F. has developed, since 1986, a general purpose code named THYC (Thermal HYdraulic Code) designed to study three-dimensional single and two-phase flows in rod tube bundles (pressurised water reactor cores, steam generators, condensers, heat exchangers). In these studies, the relative velocity was calculated by a drift-flux correlation. However, the relative velocity between vapor and liquid is an important parameter for the accuracy of a two-phase flow modelling in a three-dimensional code. The range of application of drift-flux correlations is mainly limited by the characteristic of the flow pattern (counter current flow ...) and by large 3-D effects. The purpose of this paper is to describe a numerical scheme which allows the relative velocity to be computed in a general case. Only the methodology is investigated in this paper which is not a validation work. The interfacial drag force is an important factor of stability and accuracy of the results. This force, closely dependent on the flow pattern, is not entirely established yet, so a range of multiplicator of its expression is used to compare the numerical results with the VATICAN test section measurements. (authors). 13 refs., 6 figs

  9. The phenomenology of a small break LOCA in a complex thermal hydraulic loop

    International Nuclear Information System (INIS)

    Di Marzo, M.; Almenas, K.K.; Hsu, Y.Y.; Wang, Z.

    1988-01-01

    A phenomenological description of the thermal hydraulics events that take place during a simulated Small Break Loss of Coolant Accident (SB-LOCA) is presented. The SB-LOCA transient is described in detail and the various mass and energy transport modes are identified. Similar behavior is observed in other facilities designed for the simulation of this type of accidents. Previous investigations suggest a simple modelling of the phenomena based on fluid mechanic considerations. An extensive experimental program conducted at the experimental facility of the University of Maryland reveals that condensation is a dominant driving force for this type of transients. This finding has significant implications in the modelling of enthalpy transport for some of the flow modes which occur during the transient. In particular it affects the Interruption and Resumption Mode (IRM) during which enthalpy is transported by periodic flow of a two phase mixture. The efforts to predict the flow interruption based on fluid mechanic criteria of phase separation in the hot leg are shown to be misdirected since thermodynamic phenomena taking place in the horizontal portion of the cold legs and in the reactor vessel downcomer are mostly responsible for that transition. For flow resumption to occur the liquid-vapor mixture swelling in the vertical portion of the hot leg determines the occurrence of the liquid spill over the top of the candy cane. (orig.)

  10. 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)

  11. 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.

  12. Parametric study on thermal-hydraulic response following as ATWS event

    International Nuclear Information System (INIS)

    Suh, Jeong Kwan; Bang, Young Seok; Kim, Hho Jung

    2000-01-01

    A series of sensitivity calculations for the LOFT L9-3 experiment were performed using RELAP5/MOD3 code to assess parametric effects on thermal-hydraulic response in the event of Anticipated Transient Without Scram (ATWS). The base case calculation was made by the condition which gave a good agreement for the pressure of the reactor coolant system (RCS) with the experimental data. Four parameters of PORV/spray energy loss coefficient, steam generator nodalization and moderator density coefficient (MDC) were selected during the input preparation and investigated by calculating the total discharged energy through relief valves. The energy loss coefficient of the pressurizer spray valve has a significant effect on the behavior of the RCS pressure and the change of the MDC curve within 15 % at the negative region decreased the difference of the coolant temperature between the experiment and the calculation within a range of measurement uncertainty. The finer steam generator nodalization increased the primary to secondary heat transfer rate

  13. Heat Transfer Computations of Internal Duct Flows With Combined Hydraulic and Thermal Developing Length

    Science.gov (United States)

    Wang, C. R.; Towne, C. E.; Hippensteele, S. A.; Poinsatte, P. E.

    1997-01-01

    This study investigated the Navier-Stokes computations of the surface heat transfer coefficients of a transition duct flow. A transition duct from an axisymmetric cross section to a non-axisymmetric cross section, is usually used to connect the turbine exit to the nozzle. As the gas turbine inlet temperature increases, the transition duct is subjected to the high temperature at the gas turbine exit. The transition duct flow has combined development of hydraulic and thermal entry length. The design of the transition duct required accurate surface heat transfer coefficients. The Navier-Stokes computational method could be used to predict the surface heat transfer coefficients of a transition duct flow. The Proteus three-dimensional Navier-Stokes numerical computational code was used in this study. The code was first studied for the computations of the turbulent developing flow properties within a circular duct and a square duct. The code was then used to compute the turbulent flow properties of a transition duct flow. The computational results of the surface pressure, the skin friction factor, and the surface heat transfer coefficient were described and compared with their values obtained from theoretical analyses or experiments. The comparison showed that the Navier-Stokes computation could predict approximately the surface heat transfer coefficients of a transition duct flow.

  14. Experimental observation of thermal-hydraulic behavior in PCCS horizontal heat exchanger

    International Nuclear Information System (INIS)

    Kondo, Masaya; Nakamura, Hideo; Yamamoto, Kazuhiko; Shimada, Rumi; Tokuma, Hideaki

    2003-01-01

    A series of thermal-hydraulic experiments have been performed using a prototypical-scale experimental facility simulating a horizontal heat exchanger of a Passive Containment Cooling System (PCCS) for next generation BWRs. The influences of multi-dimensional boiling flow in secondary water pool on primary flow in parallel tubes are investigated. The experimental results at postulated accident conditions; 0.7 MPa, steam flow rate equivalent to 1% core power with 1% non-condensable gas, show that steam condensation completes in almost the same heat transfer length in all the instrumented tubes. The secondary heat transfer coefficient is relatively small at the lower portion in the tube bundle when the flow regime is bubbly flow, and increases with elevation as the flow regime turns into churn-like flow. The primary steam flow distribution among tubes is rather insensitive to such a variation in the secondary heat transfer coefficient, since the contribution of the secondary heat transfer to the local heat resistance is 30% or less at postulated accident conditions. The influence of steam flow rate is insensitive too, while the contribution of the secondary heat transfer coefficient increases at low pressure conditions. (author)

  15. 300 kWt core conceptual model thermal/hydraulic characteristics

    International Nuclear Information System (INIS)

    Moody, E.

    1971-01-01

    The 300 kW(t), 199 element NASA-Lewis/AEC core conceptual model, has been analyzed to determine it's thermal-hydraulic characteristics using the GEOM-3 code. Stack-ups of tolerances and fuel rod asymmetry patterns were used to ascertain cross element Δ T's. Both zoned and uniform spacing were analyzed with a somewhat lower fuel temperature and cross element ΔT found for zoned spacing. With the models considered, the core design appears adequate to limit thermal gradients to approximately 32 0 F. Bypass flow should be controlled to prevent excessive edge element ΔT's. 11 references. (U.S.)

  16. Thermal hydraulic phenomena in corium pools: the BALI experiment

    International Nuclear Information System (INIS)

    Bonnet, J.M.

    1999-01-01

    In the framework of severe accident studies, the BALI experiment has been designed to create a data base about heat transfer distribution at corium pool boundaries for in-vessel or ex-vessel configurations. The mechanism investigated is natural convection at high internal Rayleigh number (10 15 to 10 17 ) in cavities with volumetric heating. After a description of the facility and a synthesis of results obtained for in-vessel configurations, the purpose of this paper is to present or extend local or average heat transfer correlations in the prototypic range of dimensionless parameters. (author)

  17. Thermal-hydraulic studies on molten core-concrete interactions

    International Nuclear Information System (INIS)

    Greene, G.A.

    1986-10-01

    This report discusses studies carried out in connection with light water power reactor accidents. Recent assessments have indicated that the consequences of molten-core concrete interactions dominate the considerations of severe accidents. The two areas of interest that have been investigated are interlayer heat and mass transfer and liquid-liquid boiling. Interlayer heat and mass transfer refers to processes that occur within a core melt between the stratified, immiscible phases of core oxides and metals. Liquid-liquid boiling refers to processes that occur at the melt-concrete on melt-coolant interface

  18. Thermal investigation of alkali metal hexacyanoruthenate (2)

    International Nuclear Information System (INIS)

    Okorskaya, A.P.; Sergeeva, A.N.; Pavlenko, L.I.; Semenishin, D.I.

    1978-01-01

    Thermal stability of Li, Na, K, Rb and Cs hexacyanoruthenates has been investigated. It has been established, that thermal decomposition of complexes depends upon outer spherical cations; complex compound stability decreasing with the rize of cation ionization potential. According to their thermal stability, alkali metal hexacyanoruthenates can be placed in the following row: Li < Na < K < Rb < Cs. Decomposition of Na, Rb and Cs complexes is accompanied by formation of thermally stable cyanides of these metals

  19. 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.

  20. An improved thermal-hydraulic modeling of the Jules Horowitz Reactor using the CATHARE2 system code

    Energy Technology Data Exchange (ETDEWEB)

    Pegonen, R., E-mail: pegonen@kth.se [KTH Royal Institute of Technology, Roslagstullsbacken 21, SE-10691 Stockholm (Sweden); Bourdon, S.; Gonnier, C. [CEA, DEN, DER, SRJH, CEA Cadarache, 13108 Saint-Paul-lez-Durance Cedex (France); Anglart, H. [KTH Royal Institute of Technology, Roslagstullsbacken 21, SE-10691 Stockholm (Sweden)

    2017-01-15

    Highlights: • An improved thermal-hydraulic modeling of the JHR reactor is described. • Thermal-hydraulics of the JHR is analyzed during loss of flow accident. • The heat exchanger approach gives more realistic and less conservative results. - Abstract: The newest European high performance material testing reactor, the Jules Horowitz Reactor, will support current and future nuclear reactor designs. The reactor is under construction at the CEA Cadarache research center in southern France and is expected to achieve first criticality at the end of this decade. This paper presents an improved thermal-hydraulic modeling of the reactor using solely CATHARE2 system code. Up to now, the CATHARE2 code was simulating the full reactor with a simplified approach for the core and the boundary conditions were transferred into the three-dimensional FLICA4 core simulation. A new more realistic methodology is utilized to analyze the thermal-hydraulic simulation of the reactor during a loss of flow accident.