Nuclear Reactor Thermal-Hydraulics Vol 1 [7th Intl Meeting]

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Wie erhalte ich diesen Titel? Kopie bestellen. Die Open Access Version kann inhaltlich von der lizenzpflichtigen Version abweichen. Helena St. Kitts und Nevis St. Lucia St. Pierre und Miquelon St. Dokumentinformationen Titel:. Nuclear , Heat. Gedruckte Ausgabe. An application of a higher order finite difference method to a natural convection experiment in the hot plenum of an LMFBR.

Impedance based flow reconstruction - A novel flow composition measuring technique for multi-phase-flows. Sensitivity analysis of bubble size and probe geometry on the measurements of interfacial area concentration in gas-liquid two-phase flow. Measurements of turbulent velocity and temperature in axial flow through a heated rod bundle. Fragmentation mechanisms based on single drop steam explosion experiments using flash X-ray radiography. On the fundamental microinteractions that support the propagation of steam explosions.

Steady-state and transient coupling methodologies based upon a tandem structure were devised and implemented. Additional models for nuclear feedback, equilibrium xenon and direct moderator heating were added. TITAN was tested using a boiling water two channel problem and the coupling methodologies were shown to be effective.

Simulated turbine trip transients and several control rod withdrawal transients were analyzed with good results. Sensitivity studies indicated that the time-step size can affect transient results significantly. The steady-state results were compared to a solution produced by MEKIN-B and poor agreement between the horizontal power shapes was found. Calculations with various mesh spacings showed that the mesh spacings in the MEKIN-B analysis were too large to produce accurate results with a finite difference method. It was therefore concluded that point kinetics methods should be used with caution and that three-dimensional codes like TITAN are superior for analyzing PWR control rod ejection transients.

A comprehensive high fidelity reactor core modeling capability has been developed for detailed analysis of current and advanced reactor designs as part of a US-ROK collaborative I-NERI project. High fidelity was accomplished by integrating highly refined solution modules for the coupled neutronic, thermal-hydraulic , and thermo-mechanical phenomena.

Proceedings of the Fifth International Topical Meeting on - Technische Informationsbibliothek (TIB)

Each solution module employs methods and models that are formulated faithfully to the first-principles governing the physics, real geometry, and constituents. The vast computational problem resulting from such highly refined modeling is solved on massively parallel computers, and serves as the 'numerical nuclear reactor '.

Relaxation of modeling parameters were also pursued to make problems run on clusters of workstations and PCs for smaller scale applications as well. Steady-state thermal-hydraulic analysis of the pellet-bed reactor for nuclear thermal propulsion. 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. 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. Thermal hydraulic aspects of uncertainty in power measurement of nuclear reactors.

Power measurement in Nuclear Reactors is carried out through in-core and ex-core neutron monitors which are continuously calibrated against thermal power. In Indian Pressurized Heavy Water Reactors MWe the temperature difference across steam generator hot and cold legs is taken to be a measure of thermal power as the flow through the primary heat transport system is assumed to be constant through out is operation. Gross flow is not measured directly. However, the flow depends on the characteristics of the primary heat transport pumps, which are centrifugal type and are affected by the grid frequency.

The paper quantifies the percentage increase in the reactor power for the sustained allowable frequency. The paper quantifies the percentage increase in the reactor power for the sustained allowable high grid frequency. This uncertainty is in addition to instrument inaccuracy and should be accounted for in safety analysis. In some reactors thermal power is calculated from stem flow rate and pressure, here the location of steam flow measurement is important to avoid leakage related error in thermal power. Neutron absorption cross section in the power measurement instruments and the power production in the fuel varies with neutron energy levels, these aspects are also discussed in the paper.

Detailed channel thermal-hydraulic calculation of nuclear reactor fuel assemblies. The system of equations of mass balance, quantity of motion and energy used in calculation of nuclear reactor fuel assemblies is obtained. The equation system is obtained on the base of integral equations of hydrodynamics interaction in assemblies of smooth fuel elements and fuel elements with wire packing.


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The calculation results of coolant heating distributions by the fast reactor assembly channels are presented. The analysis of the results obtained shows that interchannel exchange essentially uniforms the coolant heating distribution in the peripheral range of the assembly but it does not remove non-uniformity caused by power distribution non-uniformity in the cross section.

Geometry of the peripheral assembly range plays an essential role in the heating distribution. Change of the calculation gap between the peripheral fuel elements and assembly shells can result either in superheating or in subcooling in the peripheral channels relatively to joint internal channels of the assembly. Heat supply to the coolant passing through interassembly gaps decreases temperature in the assembly periphery and results in the increase of temperature non-uniformity by the perimeter of peripheral fuel elements.

It is concluded that the applied method of the channel-by-channel calculation is ef-- fective in thermal-physical calculation of nuclear reactor fuel assemblies and it permits to solve a wide range of problems [ru. Thermal-Hydraulics analysis of pressurized water reactor core by using single heated channel model. Directory of Open Access Journals Sweden. 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. This comparison shows the capability of single heated channel model for predicting thermal hydraulics behavior of reactor core.

Development of CFD software for the simulation of thermal hydraulics in advanced nuclear reactors. Final report. The objectives of the project were: Improvement of the simulation accuracy for nuclear reactor thermo-hydraulics by coupling system codes with three-dimensional CFD software; Extension of CFD software to predict thermo-hydraulics in advanced reactor concepts; Validation of the CFD software by simulation different UPTF TRAM-C test cases and development of best practice guidelines.

The CFD software was extended with material properties for liquid metals, and validated using existing data. This led to better agreement between predictions and data and reduced uncertainties when applying temperature boundary conditions. The meshes for the CHT simulation were also used for a coupled fluid-structure-thermal analysis which was another novelty.

The results of the multi-physics analysis showed plausible results for the mechanical and thermal stresses. The workflow developed as part of the current project can be directly used for industrial nuclear reactor simulations. Finally, simulations for two-phase flows with and without interfacial mass transfer were performed. These showed good agreement with data. However, a persisting problem for the simulation of multi-phase flows are the long simulation times which make use for industrial applications difficult. Multi scale analysis of thermal-hydraulics of nuclear reactors - the neptune project.

The system scale models the whole reactor circuit with 0D, 1D and 3D modules and is generally applied with a coarse meshing including about a thousand meshes. The component scale models components like the reactor Core or Steam Generators with a finer nodalization and is generally applied with 10 4 to 10 5 meshes. Since these components contain rod bundles or tube bundles the physical modelling uses a homogenization technique with a porosity.

For some specific applications it was found necessary to add a two-phase CFD tool able to zoom on a portion of the circuit where small scale phenomena are of importance for design purpose or safety issues. Here the basic equations are still averaged like in RANS approach for single phase, but the space resolution is finer than in component codes and typical application may require 10 5 to 10 7 meshes.

These three scales have to be coupled in order to simulate many reactor transients where both local effects and system effects play a role. In addition, two-phase Direct Numerical Simulation Tools with Interface Tracking Techniques can be used for even smaller scale investigations for a better understanding of basic physical processes and for developing closure relations for averaged models. The main challenges of this project are here presented and some first results are presented.

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

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. Finally, a systematic quality assurance program and experimental data generation system has been constructed in order to increase the reliability of the experimental data.

Modeling transient thermal hydraulic behavior of a thermionic fuel element for nuclear space reactors. A transient code TFETC for determining the temperature distribution throughout the radial and axial positions of a thermionic fuel element TFE during changes in operating conditions has been successfully developed and tested. A fully implicit method is used to solve the system of equations for temperatures at each time step.

Presently, TFETC has the ability to handle the following transients: startup, loss of flow accidents, and shutdown. The code has been applied to the startup of the ATI single cell configuration which appears to start up and shut down in an orderly and reasonable fashion. No unexpected transient features were observed. The TFE also appears to function robustly under loss of flow accident conditions. It appears hat sufficient time is available to shut the reactor down safely without melting point the fuel.

The model shows that during a complete loss of flow accident without shutdown the coolant reaches its boiling point in approximately 35 seconds. The fuel may exceed its melting point after this time as the NaK coolant will boil if the reactor is not shut down. Proceedings of the Meeting on Reactor Physics and Thermal Hydraulics. 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.

Thermal-hydraulic modeling of porous bed reactors. 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.

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. Cross-cutting european thermal-hydraulics research for innovative nuclear systems. 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. This paper will describe the activities in this project which address the main identified thermal hydraulics issues for innovative nuclear systems. Proceedings of the third nuclear thermal hydraulics meeting. The papers presented include: Simulator qualification using engineering codes and Development of thermal hydraulic analysis capabilities for Oyster Creek.

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

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. Nuclear future: thinking for building. Brazilian national meeting on reactor physics and thermal hydraulics ; 8. General congress on nuclear energy; 5. Brazilian national meeting on nuclear applications. These proceedings, for the first time, present jointly the General congress on nuclear energy 8.

CGEN , and 5. Brazilian national meeting on nuclear applications 5. The main theme of discussion was: ' Nuclear Future: thinking for building'. The papers have analysed the progresses of peaceful utilization of nuclear technology and its forecasting for the beginning of the new millennium.

The construction of Angra-3 nuclear power plant have been discussed. System code assessment with thermal-hydraulic experiment to develop helium cooled breeding blanket for nuclear fusion reactor. The test was performed at pressures of 11, 19, and 29 bar, and under various ranges of flow rate from 0. In the present study, a thermal-hydraulic test was performed with the newly constructed helium supplying system, In which the design pressure and temperature were 9 MPa and C, respectively.

In the experiment, the same mock-up was used, and the test was performed under the conditions of 8 MPa pressure, 0. One-side of the mock-up was heated with a constant heat flux of 0. The wall temperatures were measured using installed thermocouples, and they show a strong parity with the code results simulated under the same test conditions. User's manual. Volume II. Program implementation. The volume is divided into main sections which cover: 1 program description, 2 input data, 3 problem initialization, 4 user guidelines, 5 output discussion, 6 source program description, 7 implementation requirements, 8 data files, 9 description of PLOTR4M, 10 description of STH20, 11 summary flowchart, 12 sample problems, 13 problem definition, and 14 problem input.

Volume 1. Mathematical modeling. 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. A series of studies has been performed to investigate the potential impact of the coupling between neutronics and thermal hydraulics on the design and performance assessment of solid core reactors for nuclear thermal space propulsion, using the particle bed reactor PBR concept as an example system.

For a given temperature distribution in the reactor , the k eff and steady-state core power distribution are obtained from three-dimensional, continuous energy Monte Carlo simulations using the MCNP code. For a given core power distribution, determination of the temperature distribution in the core and hydrogen-filled annulus between the reflector and pressure vessel is based on a nonthermal equilibrium analysis. However, it may be possible to estimate the thermal safety margins and propellant exit temperatures based on power distributions obtained from neutronic calculations at room temperature.

The results also show that, while variation of the hydrogen flow rate in the annulus has been proposed as a partial control mechanism for PBRs, such control mechanism may not be feasible for PBRs with high moderator-to-fuel ratios and hence soft core neutron spectra. Thermal-hydraulic behavior of physical quantities at critical velocities in a nuclear research reactor core channel using plate type fuel.

Full Text Available The thermal-hydraulic study presented here relates to a channel of a nuclear reactor core. This channel is defined as being the space between two fuel plates where a coolant fluid flows. The flow velocity of this coolant should not generate vibrations in fuel plates. The aim of this study is to know the distribution of the temperature in the fuel plates, in the cladding and in the coolant fluid at the critical velocities of Miller, of Wambsganss, and of Cekirge and Ural.

The thermal-hydraulic study is made under steady-state; the equation set-up of the thermal problem is made according to El Wakil and to Delhaye. Once the equation set-up is validated, the three critical velocities are calculated and then used in the calculations of the different temperature profiles. The average heat flux and the critical heat flux are evaluated for each critical velocity and their ratio reported. The recommended critical velocity to be used in nuclear channel calculations is that of Wambsganss. The mathematical model used is more precise and all the physical quantities, when using this critical velocity, stay in safe margins.

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.

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. Proceedings of the 5. Brazilian national meeting on nuclear applications; 8.

General congress on nuclear energy; Brazilian national meeting on reactor physics and thermal hydraulics. Brazilian national meeting on reactor physics and thermal hydraulics ENFIR , the 8. CGEN , and the 5. Thermal hydraulics analysis of the Advanced High Temperature Reactor.

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. Transient three-dimensional thermal-hydraulic analysis of nuclear reactor fuel rod arrays: general equations and numerical scheme.

A mathematical model and a numerical solution scheme for thermal-hydraulic analysis of fuel rod arrays are given. The model alleviates the two major deficiencies associated with existing rod array analysis models, that of a correct transverse momentum equation and the capability of handling reversing and circulatory flows. Possible applications of the model include steady state and transient subchannel calculations as well as analysis of flows in heat exchangers, other engineering equipment, and porous media.

Thermal-hydraulic modeling needs for passive reactors. The U. 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.

To assure the applicability of RELAP5 to the analysis of these transients for the AP design, a four year long program of code development and assessment has been undertaken. Validation of containment thermal hydraulic computer codes for VVER reactor. Thermal-hydraulics Department of the Nuclear Research Institute Rez performs accident analyses for these plants using a number of computer codes. 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. In order to apply these codes for VVER 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. An approach to.

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 Main objectives of the present phase Advanced modelling and numerical strategies in nuclear thermal-hydraulics. 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.

Spent nuclear fuel storage pool thermal-hydraulic analysis. Storage methods and requirements for spent nuclear fuel at U. 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, , and node models of the storage pool are compared to each other and to the experimental data. 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. Thermal-hydraulic methods in fast reactor safety. 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. 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 In the second phase In the third phase Proceedings of the US Nuclear Regulatory Commission twentieth water reactor safety information meeting; Volume 2, Severe accident research, Thermal hydraulics.

The papers describe progress and results of programs in nuclear safety research conducted in this country and abroad. Selected papers have been processed separately for inclusion in the Energy Science and Technology Database. Advances in thermal hydraulic and neutronic simulation for reactor analysis and safety. Tentner, A. 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.


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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.

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. Pegonen, R. 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.

A new more realistic methodology is utilized to analyze the thermal-hydraulic simulation of the reactor during a loss of flow accident. The BEACON code is a best-estimate, advanced containment code designed to perform a best-estimate analysis of the flow of a mixture of air, water, and steam in a nuclear reactor containment system under loss-of-coolant accident conditions. The code can simulate two-component, two-phase fluid flow in complex geometries using a combination of two-dimensional, one-dimensional, and lumped-parameter representations for the various parts of the system.

It is suitable for the evaluation of short-term transients in dry-containment systems. It provides application information for input data preparation and for output data interpretation. Light-water- reactor coupled neutronic and thermal-hydraulic codes. 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. Speculation as to future trends with such codes is also presented. One model uses a structured grid system with the porous media approach for the Calandria tubes in the core region. The other uses a unstructured grid system on the real geometry of Calandria tubes, so that the detailed fluid flow between the Calandria tubes can be observed.

Proceedings of the fourth international topical meeting on nuclear thermal hydraulics , operations and safety. More than papers were presented. The study of thermal-hydraulic performance of a fixed bed nuclear reactor FBNR core and the effect of the porosity was studied by the CFD method with 'SolidWorks' software.

The representative sections of three different packed beds arrangements were analyzed: face-centered cubic FCC , body-centered cubic BCC , and a pseudo-random, with values of porosity of 0. The minimum coolant flow required to avoid the phase change for each one of the configurations was determined. The results show that the heat transfer rate increases when the porosity value decreases, and consequently the minimum coolant flow in each configuration. The results of minimum coolant flow were: Finally the pressure drop was calculated, and the results were 0.

This means that with a higher porosity, the fluid can circulate easier because there are fewer obstacles to cross, so there are fewer energy losses. Primary system thermal hydraulics of future Indian fast reactors. Velusamy, K. Arul; Rajendrakumar, M. Partha; Selvaraj, P. Senthil; Jebaraj, C. 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.

Study on safety of a nuclear ship having an integral marine water reactor. Intelligent information database program concerned with thermal-hydraulic characteristics. As a high economical marine reactor with sufficient safety functions, an integrated type marine water reactor has been considered most promising. Since the program was created as a Windows application using the Visual Basic, it is available to the public and can be easily installed in the operating system.

Main functions of the program are as follows: 1 steady state flow boiling analysis and determination of stability limit for any helical-coil type once-through steam generator design. The program will be useful for the design of not only the future integrated type marine water reactor but also the small sized water reactor.

Thermal hydraulic analysis of the encapsulated nuclear heat source. 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. Numerical simulations of subcritical reactor kinetics in thermal hydraulic transient phases. 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.

Thermal-hydraulic tests for reactor safety system. Tests for the safety depressurization system, Sparger adopted for the Korean next generation reactor , APR 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. 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. Neutronics - thermal-hydraulics coupling: application to the helium-cooled fast reactor. 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.


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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. 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.

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

Current research interests

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. 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 a LB-LOCA, the primary side pressure, coolant density and fuel temperature strongly decrease but the cladding temperature experiences a strong peak. 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. Thermal hydraulic feasibility assessment of the spent nuclear fuel project. 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. Nuclear energy and energetic challenges for 21st. A key goal of these joint meetings is to bring together scientists to exchange the latest research and development R and D information in nuclear science and technology. In the INAC technical program, plenary sessions, such as round table discussions and keynote lectures, has held to present to the general public the recent advances of peaceful nuclear energy usage, reducing the global warming.

The XV ENFIR has covered all aspects of interdisciplinary R and D related to nuclear reactors , and the VIII ENAN has offered a forum for discussion on nuclear applications in industry, geology, agriculture, medicine, biology, environmental sciences always highlighting the social, economical and environmental improvements generated by nuclear techniques.

Thermal hydraulic and safety analyses for Pakistan Research Reactor The present core comprises of 29 standard and 5 control fuel elements. Safety analysis has been carried out for various modes of reactivity insertions.

Department of Mechanical Engineering

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. 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. Thermal-hydraulic considerations for particle bed reactors. 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. Application of thermal-hydraulic codes in the nuclear sector. Queral, C. 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.

Nuclear energy reducing global warming; Brazilian national meeting on reactor physics and thermal hydraulics ; 7. European activities on crosscutting thermal-hydraulic phenomena for innovative nuclear systems. Cheng, X. 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. 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.

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. 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 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. 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.

The 'Heat Transfer package' is a subroutine program and is especially developed for research reactors in which plate-type fuel is adopted. 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. Encontro de Fisica de Reatores e Termo-Hidraulica. A review of the current thermal-hydraulic modeling of the Jules Horowitz Reactor : A loss of flow accident analysis. The reactor is under construction at CEA Cadarache research center in France and is expected to start operation at the end of this decade.

R and D and analytical works have already been performed to set-up the methodology for thermal-hydraulic calculations of the reactor. The main objective of the present work is to analyze the thermal-hydraulic calculations of the reactor during the loss of flow accident using CEA methodology.

Possible improvements of the current methodology are shortly discussed and suggested. Development of steady thermal-hydraulic analysis code for China advanced research reactor. 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.

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. Real time thermal hydraulic model for high temperature gas-cooled reactor core.

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. Thermal-hydraulics design comparisons for the tandem mirror hybrid reactor blanket. 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.