Debris Thermal Hydraulics Modeling of QUENCH Experiments
International Nuclear Information System (INIS)
Kisselev, Arcadi E.; Kobelev, Gennadii V.; Strizhov, Valerii F.; Vasiliev, Alexander D.
2006-01-01
Porous debris formation and behavior in QUENCH experiments (QUENCH-02, QUENCH-03) plays a considerable role and its adequate modeling is important for thermal analysis. This work is aimed to the development of a numerical module which is able to model thermal hydraulics and heat transfer phenomena occurring during the high-temperature stage of severe accident with the formation of debris region and molten pool. The original approach for debris evolution is developed from classical principles using a set of parameters including debris porosity; average particle diameter; temperatures and mass fractions of solid, liquid and gas phases; specific interface areas between different phases; effective thermal conductivity of each phase, including radiative heat conductivity; mass and energy fluxes through the interfaces. The debris model is based on the system of continuity, momentum and energy conservation equations, which consider the dynamics of volume-averaged velocities and temperatures of fluid, solid and gaseous phases of porous debris. The different mechanisms of debris formation are considered, including degradation of fuel rods according to temperature criteria, taking into consideration some correlations between rod layers thicknesses; degradation of rod layer structure due to thermal expansion of melted materials inside intact rod cladding; debris formation due to sharp temperature drop of previously melted material due to reflood; and transition to debris of material from elements lying above. The porous debris model was implemented to best estimate numerical code RATEG/SVECHA/HEFEST developed for modeling thermal hydraulics and severe accident phenomena in a reactor. The model is used for calculation of QUENCH experiments. The results obtained by the model are compared to experimental data concerning different aspects of thermal behavior: thermal hydraulics of porous debris, radiative heat transfer in a porous medium, the generalized melting and refreezing
Thermal-hydraulic model in MABEL-2
International Nuclear Information System (INIS)
Bowring, R.W.; Cooper, C.A.
1979-09-01
The thermal-hydraulic subroutines for MABEL-2 have been written and programmed for testing as a stand-alone code MABSA. The model is that of a fuel rod, surrounded by 8 other rods, with a flow region outside representing the reactor geometry. A subchannel model, similar to that in HAMBO, is used to calculate the feed-back effect of cladding strain on heat transfer due to the progressive blocking of the subchannels around the central fuel rod. A number of assumptions was made in building the coolant model. Their use and validation are discussed. (author)
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)
Similarity conditions for investigations of hydraulic-thermal tidal models
International Nuclear Information System (INIS)
Fluegge, G.; Schwarze, H.
1975-01-01
With the construction of nuclear power plants near German tidal estuaries in mind, investigations of mixing and spreading processes which occur during the discharge of heated cooling water in tidal waters were carried out in hydraulic-thermal tidal models of the Lower Weser and Lower Elbe by the Franzius Institute for hydraulic and coastal engineering of the Technical University Hannover. This contribution discusses in detail the problems met and the experience gained in constructing and operating these models. (orig./TK) [de
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
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
Thermal-hydraulic modeling of reactivity accidents in MTR reactors
Directory of Open Access Journals (Sweden)
Khater Hany
2006-01-01
Full Text Available This paper describes the development of a dynamic model for the thermal-hydraulic analysis of MTR research reactors during a reactivity insertion accident. The model is formulated for coupling reactor kinetics with feedback reactivity and reactor core thermal-hydraulics. To represent the reactor core, two types of channels are considered, average and hot channels. The developed computer program is compiled and executed on a personal computer, using the FORTRAN language. The model is validated by safety-related benchmark calculations for MTR-TYPE reactors of IAEA 10 MW generic reactor for both slow and fast reactivity insertion transients. A good agreement is shown between the present model and the benchmark calculations. Then, the model is used for simulating the uncontrolled withdrawal of a control rod of an ETRR-2 reactor in transient with over power scram trip. The model results for ETRR-2 are analyzed and discussed.
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)
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
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
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.
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)
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
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.
HELOKA-HP thermal-hydraulic model validation and calibration
Energy Technology Data Exchange (ETDEWEB)
Jin, Xue Zhou; Ghidersa, Bradut-Eugen; Badea, Aurelian Florin
2016-11-01
Highlights: • The electrical heater in HELOKA-HP has been modeled with RELAP5-3D using experimental data as input. • The model has been validated using novel techniques for assimilating experimental data and the representative model parameters with BEST-EST. • The methodology is successfully used for reducing the model uncertainties and provides a quantitative measure of the consistency between the experimental data and the model. - Abstract: The Helium Loop Karlsruhe High Pressure (HELOKA-HP) is an experimental facility for the testing of various helium-cooled components at high temperature (500 °C) and high pressure (8 MPa) for nuclear fusion applications. For modeling the loop thermal dynamics, a thermal-hydraulic model has been created using the system code RELAP5-3D. Recently, new experimental data covering the behavior of the loop components under relevant operational conditions have been made available giving the possibility of validating and calibrating the existing models in order to reduce the uncertainties of the simulated responses. This paper presents an example where such process has been applied for the HELOKA electrical heater model. Using novel techniques for assimilating experimental data, implemented in the computational module BEST-EST, the representative parameters of the model have been calibrated.
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
submitter Thermal, Hydraulic, and Electromagnetic Modeling of Superconducting Magnet Systems
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...
Thermal-hydraulic models and correlations for the SPACE code
International Nuclear Information System (INIS)
Kim, K. D.; Lee, S. W.; Bae, S. W.; Moon, S. K.; Kim, S. Y.; Lee, Y. H.
2009-01-01
The SPACE code which is based on a multi-dimensional two-fluid, three-field model is under development to be used for licensing future pressurized water reactors. Several research and industrial organizations are participated in the collaboration of the development program, including KAERI, KHNP, KOPEC, KNF, and KEPRI. KAERI has been assigned to develop the thermal-hydraulic models and correlations which are required to solve the field equations as the closure relationships. This task can be categorized into five packages; i) a flow regime selection package, ii) a wall and interfacial friction package, iii) an interfacial heat and mass transfer package iv) a droplet entrainment and de-entrainment package and v) a wall heat transfer package. Since the SPACE code, unlike other major best-estimate nuclear reactor system analysis codes, RELAP5, TRAC-M and CATHARE which consider only liquid and vapor phases, incorporates a dispersed liquid field in addition to vapor and continuous liquid fields, intel facial interaction models between continuous, dispersed liquid phases and vapor phase have to be developed separately. The proper physical models can significantly improve the accuracy of the prediction of a nuclear reactor system behavior under many different transient conditions because those models are composed of the source terms for the governing equations. In this paper, a development program for the physical models and correlations for the SPACE code will be introduced briefly
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.
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.
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
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
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.
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)
Thermally Actuated Hydraulic Pumps
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
Use of operational data for the validation of thermal hydraulic models
International Nuclear Information System (INIS)
Shoukas, L.; Martin, G.; Ho, S.F.
1996-01-01
Thermal hydraulic models used to predict Primary Heat Transport (PHT) System behaviour have traditionally been applied with design conditions to predict transient responses of accident scenarios in safety analyses. Recently, the use of reactor operational data has been integral in the development of thermal hydraulic codes to improve the quality of the predictions. The basis of accurate thermal hydraulic predictions is the use of appropriate models with accurate input data. An operating reactor provides a wealth of information, therefore, the models can be validated against operating conditions specific to the field of application. Thus, agreement between prediction and plant data continue to improve due to constant update of the thermal hydraulic models and/or the input data. The ability to accurately predict thermal hydraulic responses with the code provides the analyst with a powerful tool in reactor performance monitoring. The primary objective of this paper is to describe the validation process of the Mini-SOPHT (Simulation Of Primary Heat Transport) Header to Header Model with the use of reactor operational data. The secondary objective is to illustrate the effectiveness of the code as a performance monitoring tool by discussing the discoveries that were made during the validation process. (author)
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
Validation of ISAAC Thermal Hydraulic Model against the RD-14M Experiment B9401
International Nuclear Information System (INIS)
Kim, Dong Ha; Kim, Hyung Tae; Park, Soo Yong; Kim, Sang Baik
2009-09-01
The thermal hydraulic behavior prior to core damage was compared with the experimental data to validate the ISAAC thermal hydraulic models. B9401 test at RD-14M facility was selected as a benchmark data set, which was open for the international standard problem (ISP). As ISAAC had hard-wired systems and models inside the code, new input parameters were prepared with the minimum model changes. Among 37 parameters measured from the experiment, pressures, flow rates, void fractions, and the fuel temperatures were compared. The simulation showed that the thermal hydraulic behavior estimated from ISAAC was similar to the experimental data or to the detailed code, even though ISAAC was developed mainly for the severe accident analysis. Based on these comparison analyses, it can be concluded that the thermal hydraulic conditions estimated from ISAAC can be used as the boundary conditions for further analysis causing severe core damage. In order to extend the data base for the thermal hydraulic behavior of ISAAC, more RD-14M tests as well as DBA sequences need to be simulated
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
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
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)
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
Modeling thermal stress propagation during hydraulic stimulation of geothermal wells
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
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
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
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.
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
A Thermal Hydraulic Model of Melt Lubrication in Railgun Armatures
National Research Council Canada - National Science Library
Kothmann, R
2003-01-01
... wear of 7075 aluminum sliding against ETP copper for face pressures ranging from 6 to 22 ksi. Discrepancies between calculated and experimental results are attributed to uncertainties in modeling the complex phase change behavior of aluminum alloy 7075 and uncertain conditions at the rail interface.
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)
A General Model for Thermal, Hydraulic and Electric Analysis of Superconducting Cables
Bottura, L; Rosso, C
2000-01-01
In this paper we describe a generic, multi-component and multi-channel model for the analysis of superconducting cables. The aim of the model is to treat in a general and consistent manner simultaneous thermal, electric and hydraulic transients in cables. The model is devised for most general situations, but reduces in limiting cases to most common approximations without loss of efficiency. We discuss here the governing equations, and we write them in a matrix form that is well adapted to numerical treatment. We finally demonstrate the model capability by comparison with published experimental data on current distribution in a two-strand cable.
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
Comparative study of boron transport models in NRC Thermal-Hydraulic Code Trace
Energy Technology Data Exchange (ETDEWEB)
Olmo-Juan, Nicolás; Barrachina, Teresa; Miró, Rafael; Verdú, Gumersindo; Pereira, Claubia, E-mail: nioljua@iqn.upv.es, E-mail: tbarrachina@iqn.upv.es, E-mail: rmiro@iqn.upv.es, E-mail: gverdu@iqn.upv.es, E-mail: claubia@nuclear.ufmg.br [Institute for Industrial, Radiophysical and Environmental Safety (ISIRYM). Universitat Politècnica de València (Spain); Universidade Federal de Minas Gerais (UFMG), Belo Horizonte, MG (Brazil). Departamento de Engenharia Nuclear
2017-07-01
Recently, the interest in the study of various types of transients involving changes in the boron concentration inside the reactor, has led to an increase in the interest of developing and studying new models and tools that allow a correct study of boron transport. Therefore, a significant variety of different boron transport models and spatial difference schemes are available in the thermal-hydraulic codes, as TRACE. According to this interest, in this work it will be compared the results obtained using the different boron transport models implemented in the NRC thermal-hydraulic code TRACE. To do this, a set of models have been created using the different options and configurations that could have influence in boron transport. These models allow to reproduce a simple event of filling or emptying the boron concentration in a long pipe. Moreover, with the aim to compare the differences obtained when one-dimensional or three-dimensional components are chosen, it has modeled many different cases using only pipe components or a mix of pipe and vessel components. In addition, the influence of the void fraction in the boron transport has been studied and compared under close conditions to BWR commercial model. A final collection of the different cases and boron transport models are compared between them and those corresponding to the analytical solution provided by the Burgers equation. From this comparison, important conclusions are drawn that will be the basis of modeling the boron transport in TRACE adequately. (author)
Review of turbulence modelling for numerical simulation of nuclear reactor thermal-hydraulics
International Nuclear Information System (INIS)
Bernard, J.P.; Haapalehto, T.
1996-01-01
The report deals with the modelling of turbulent flows in nuclear reactor thermal-hydraulic applications. The goal is to give tools and knowledge about turbulent flows and their modelling in practical applications for engineers, and especially nuclear engineers. The emphasize is on the theory of turbulence, the existing different turbulence models, the state-of-art of turbulence in research centres, the available models in the commercial code CFD-FLOW3D, and the latest applications of turbulence modelling in nuclear reactor thermal-hydraulics. It turns out that it is difficult to elaborate an universal turbulence model and each model has its advantages and drawbacks in each application. However, the increasing power of computers can permit the emergence of new methods of turbulence modelling such as Direct Numerical Simulation (DNS) and Large Eddy Simulation (LES) which open new horizons in this field. These latter methods are beginning to be available in commercial codes and are used in different nuclear applications such as 3-D modelling of the nuclear reactor cores and the steam generators. (orig.) (22 refs.)
Coupled neutronics and thermal hydraulics modelling in reactor dynamics codes TRAB-3D and HEXTRAN
International Nuclear Information System (INIS)
Kyrki-Rajamaeki, R.; Raety, H.
1999-01-01
The reactor dynamics codes for transient and accident analyses inherently include the coupling of neutronics and thermal hydraulics modelling. In Finland a number of codes with 1D and 3D neutronic models have been developed, which include models also for the cooling circuits. They have been used mainly for the needs of Finnish power plants, but some of the codes have also been utilized elsewhere. The continuous validation, simultaneous development, and experiences obtained in commercial applications have considerably improved the performance and range of application of the codes. The fast operation of the codes has enabled realistic analysis of 3D core combined to a full model of the cooling circuit even in such long reactivity scenarios as ATWS. The reactor dynamics methods are developed further and new more detailed models are created for tasks related to increased safety requirements. For thermal hydraulics calculations, an accurate general flow model based on a new solution method has been developed. Although mainly intended for analysis purposes, the reactor dynamics codes also provide reference solutions for simulator applications. As computer capability increases, these more sophisticated methods can be taken into use also in simulator environments. (author)
Development of the Real-time Core and Thermal-Hydraulic Models for Kori-1 Simulator
International Nuclear Information System (INIS)
Hong, Jin Hyuk; Lee, Myeong Soo; Hwang, Do Hyun; Byon, Soo Jin
2010-01-01
The operation of the Kori-Unit 1 (1723.5MWt) is expanded to additional 10 years with upgrades of the Main Control Room (MCR). Therefore, the revision of the procedures, performance tests and works related with the exchange of the Main Control Board (MCB) are currently carried out. And as a part of it, the fullscope simulator for the Kori-1 is being developed for the purpose of the pre-operation and emergence response capability for the operators. The purpose of this paper is to report on the performance of the developed neutronics and thermal-hydraulic (TH) models of Kori Unit 1 simulator. The neutronics model is based on the NESTLE code and TH model based on the RELAP5/MOD3 thermal-hydraulics analysis code which was funded as FY-93 LDRD Project 7201 and is running on the commercial simulator environment tool (the 3KeyMaster TM of the WSC). As some examples for the verification of the developed neutronics and TH models, some figures are provided. The outputs of the developed neutronics and TH models are in accord with the Nuclear Design Report (NDR) and Final Safety Analysis Report (FSAR) of the reference plant
A model selection support system for numerical simulations of nuclear thermal-hydraulics
International Nuclear Information System (INIS)
Gofuku, Akio; Shimizu, Kenji; Sugano, Keiji; Yoshikawa, Hidekazu; Wakabayashi, Jiro
1990-01-01
In order to execute efficiently a dynamic simulation of a large-scaled engineering system such as a nuclear power plant, it is necessary to develop intelligent simulation support system for all phases of the simulation. This study is concerned with the intelligent support for the program development phase and is engaged in the adequate model selection support method by applying AI (Artificial Intelligence) techniques to execute a simulation consistent with its purpose and conditions. A proto-type expert system to support the model selection for numerical simulations of nuclear thermal-hydraulics in the case of cold leg small break loss-of-coolant accident of PWR plant is now under development on a personal computer. The steps to support the selection of both fluid model and constitutive equations for the drift flux model have been developed. Several cases of model selection were carried out and reasonable model selection results were obtained. (author)
IMPROVEMENT OF THE LOCA PSA MODEL USING A BEST-ESTIMATE THERMAL-HYDRAULIC ANALYSIS
Directory of Open Access Journals (Sweden)
DONG HYUN LEE
2014-08-01
Full Text Available Probabilistic Safety Assessment (PSA has been widely used to estimate the overall safety of nuclear power plants (NPP and it provides base information for risk informed application (RIA and risk informed regulation (RIR. For the effective and correct use of PSA in RIA/RIR related decision making, the risk estimated by a PSA model should be as realistic as possible. In this work, a best-estimate thermal-hydraulic analysis of loss-of-coolant accidents (LOCAs for the Hanul Nuclear Units 3&4 is first carried out in a systematic way. That is, the behaviors of peak cladding temperature (PCT were analyzed with various combinations of break sizes, the operating conditions of safety systems, and the operator's action time for aggressive secondary cooling. Thereafter, the results of the thermal-hydraulic analysis have been reflected in the improvement of the PSA model by changing both accident sequences and success criteria of the event trees for the LOCA scenarios.
Transient thermal hydraulic modeling and analysis of ITER divertor plate system
International Nuclear Information System (INIS)
El-Morshedy, Salah El-Din; Hassanein, Ahmed
2009-01-01
A mathematical model has been developed/updated to simulate the steady state and transient thermal-hydraulics of the International Thermonuclear Experimental Reactor (ITER) divertor module. The model predicts the thermal response of the armour coating, divertor plate structural materials and coolant channels. The selected heat transfer correlations cover all operating conditions of ITER under both normal and off-normal situations. The model also accounts for the melting, vaporization, and solidification of the armour material. The developed model is to provide a quick benchmark of the HEIGHTS multidimensional comprehensive simulation package. The present model divides the coolant channels into a specified axial regions and the divertor plate into a specified radial zones, then a two-dimensional heat conduction calculation is created to predict the temperature distribution for both steady and transient states. The model is benchmarked against experimental data performed at Sandia National Laboratory for both bare and swirl tape coolant channel mockups. The results show very good agreements with the data for steady and transient states. The model is then used to predict the thermal behavior of the ITER plasma facing and structural materials due to plasma instability event where 60 MJ/m 2 plasma energy is deposited over 500 ms. The results for ITER divertor response is analyzed and compared with HEIGHTS results.
Transient thermal hydraulic modeling and analysis of ITER divertor plate system
Energy Technology Data Exchange (ETDEWEB)
El-Morshedy, Salah El-Din [Argonne National Laboratory, Argonne, IL (United States); Atomic Energy Authority, Cairo (Egypt)], E-mail: selmorshedy@etrr2-aea.org.eg; Hassanein, Ahmed [Purdue University, West Lafayette, IN (United States)], E-mail: hassanein@purdue.edu
2009-12-15
A mathematical model has been developed/updated to simulate the steady state and transient thermal-hydraulics of the International Thermonuclear Experimental Reactor (ITER) divertor module. The model predicts the thermal response of the armour coating, divertor plate structural materials and coolant channels. The selected heat transfer correlations cover all operating conditions of ITER under both normal and off-normal situations. The model also accounts for the melting, vaporization, and solidification of the armour material. The developed model is to provide a quick benchmark of the HEIGHTS multidimensional comprehensive simulation package. The present model divides the coolant channels into a specified axial regions and the divertor plate into a specified radial zones, then a two-dimensional heat conduction calculation is created to predict the temperature distribution for both steady and transient states. The model is benchmarked against experimental data performed at Sandia National Laboratory for both bare and swirl tape coolant channel mockups. The results show very good agreements with the data for steady and transient states. The model is then used to predict the thermal behavior of the ITER plasma facing and structural materials due to plasma instability event where 60 MJ/m{sup 2} plasma energy is deposited over 500 ms. The results for ITER divertor response is analyzed and compared with HEIGHTS results.
Neutronic and Thermal-hydraulic Modelling of High Performance Light Water Reactor
International Nuclear Information System (INIS)
Seppaelae, Malla
2008-01-01
High Performance Light Water Reactor (HPLWR), which is studied in EU project 'HPLWR2', uses water at supercritical pressures as coolant and moderator to achieve higher core outlet temperature and thus higher efficiency compared to present reactors. At VTT Technical Research Centre of Finland, functionality of the thermal-hydraulics in the coupled reactor dynamics code TRAB3D/ SMABRE was extended to supercritical pressures for the analyses of HPLWR. Input models for neutronics and thermal-hydraulics were made for TRAB3D/ SMABRE according to the latest HPLWR design. A preliminary analysis was performed in which the capability of SMABRE in the transition from supercritical pressures to subcritical pressures was demonstrated. Parameterized two-group cross sections for TRAB3D neutronics were received from Hungarian Academy of Sciences KFKI Atomic Energy Research Institute together with a subroutine for handling them. PSG, a new Monte Carlo transport code developed at VTT, was also used to generate two-group constants for HPLWR and comparisons were made with the KFKI cross sections and MCNP calculations. (author)
Parametric analyses of DEMO Divertor using two dimensional transient thermal hydraulic modelling
Domalapally, Phani; Di Caro, Marco
2017-11-01
Among the options considered for cooling of the Plasma facing components of the DEMO reactor, water cooling is a conservative option because of its high heat removal capability. In this work a two-dimensional transient thermal hydraulic code is developed to support the design of the divertor for the projected DEMO reactor with water as a coolant. The mathematical model accounts for transient 2D heat conduction in the divertor section. Temperature-dependent properties are used for more accurate analysis. Correlations for single phase flow forced convection, partially developed subcooled nucleate boiling, fully developed subcooled nucleate boiling and film boiling are used to calculate the heat transfer coefficients on the channel side considering the swirl flow, wherein different correlations found in the literature are compared against each other. Correlation for the Critical Heat Flux is used to estimate its limit for a given flow conditions. This paper then investigates the results of the parametric analysis performed, whereby flow velocity, diameter of the coolant channel, thickness of the coolant pipe, thickness of the armor material, inlet temperature and operating pressure affect the behavior of the divertor under steady or transient heat fluxes. This code will help in understanding the basic parameterś effect on the behavior of the divertor, to achieve a better design from a thermal hydraulic point of view.
Neutronic and Thermal-hydraulic Modelling of High Performance Light Water Reactor
Energy Technology Data Exchange (ETDEWEB)
Seppaelae, Malla [VTT Technical Research Centre of Finland, P.O.Box 1000, FI02044 VTT (Finland)
2008-07-01
High Performance Light Water Reactor (HPLWR), which is studied in EU project 'HPLWR2', uses water at supercritical pressures as coolant and moderator to achieve higher core outlet temperature and thus higher efficiency compared to present reactors. At VTT Technical Research Centre of Finland, functionality of the thermal-hydraulics in the coupled reactor dynamics code TRAB3D/ SMABRE was extended to supercritical pressures for the analyses of HPLWR. Input models for neutronics and thermal-hydraulics were made for TRAB3D/ SMABRE according to the latest HPLWR design. A preliminary analysis was performed in which the capability of SMABRE in the transition from supercritical pressures to subcritical pressures was demonstrated. Parameterized two-group cross sections for TRAB3D neutronics were received from Hungarian Academy of Sciences KFKI Atomic Energy Research Institute together with a subroutine for handling them. PSG, a new Monte Carlo transport code developed at VTT, was also used to generate two-group constants for HPLWR and comparisons were made with the KFKI cross sections and MCNP calculations. (author)
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
International Nuclear Information System (INIS)
Guingo, M.; Baudry, C.; Hassanaly, M.; Lavieville, J.; Mechitouna, N.; Merigoux, N.; Mimouni, S.; Bestion, D.; Coste, P.; Morel, C.
2015-01-01
NEPTUNE CFD is a Computational Multi-(Fluid) Dynamics code dedicated to the simulation of multiphase flows, primarily targeting nuclear thermo-hydraulics applications, such as the departure from nuclear boiling (DNB) or the two-phase Pressurized Thermal Shock (PTS). It is co-developed within the joint research/development project NEPTUNE (AREVA, CEA, EDF, IRSN) since 2001. Over the years, to address the aforementioned applications, dedicated physical models and numerical methods have been developed and implemented in the code, including specific sets of models for turbulent boiling flows and two-phase non-adiabatic stratified flows. This paper aims at summarizing the current main modeling capabilities of the code, and gives an overview of the associated validation database. A brief summary of emerging applications of the code, such as containment simulation during a potential severe accident or in-vessel retention, is also provided. (authors)
Energy Technology Data Exchange (ETDEWEB)
Anh Bui; Nam Dinh; Brian Williams
2013-09-01
In addition to validation data plan, development of advanced techniques for calibration and validation of complex multiscale, multiphysics nuclear reactor simulation codes are a main objective of the CASL VUQ plan. Advanced modeling of LWR systems normally involves a range of physico-chemical models describing multiple interacting phenomena, such as thermal hydraulics, reactor physics, coolant chemistry, etc., which occur over a wide range of spatial and temporal scales. To a large extent, the accuracy of (and uncertainty in) overall model predictions is determined by the correctness of various sub-models, which are not conservation-laws based, but empirically derived from measurement data. Such sub-models normally require extensive calibration before the models can be applied to analysis of real reactor problems. This work demonstrates a case study of calibration of a common model of subcooled flow boiling, which is an important multiscale, multiphysics phenomenon in LWR thermal hydraulics. The calibration process is based on a new strategy of model-data integration, in which, all sub-models are simultaneously analyzed and calibrated using multiple sets of data of different types. Specifically, both data on large-scale distributions of void fraction and fluid temperature and data on small-scale physics of wall evaporation were simultaneously used in this work’s calibration. In a departure from traditional (or common-sense) practice of tuning/calibrating complex models, a modern calibration technique based on statistical modeling and Bayesian inference was employed, which allowed simultaneous calibration of multiple sub-models (and related parameters) using different datasets. Quality of data (relevancy, scalability, and uncertainty) could be taken into consideration in the calibration process. This work presents a step forward in the development and realization of the “CIPS Validation Data Plan” at the Consortium for Advanced Simulation of LWRs to enable
Validation of the Thermal-Hydraulic Model in the SACAP Code with the ISP Tests
Energy Technology Data Exchange (ETDEWEB)
Park, Soon-Ho; Kim, Dong-Min; Park, Chang-Hwan [FNC Technology Co., Yongin (Korea, Republic of)
2016-10-15
In safety viewpoint, the pressure of the containment is the important parameter, of course, the local hydrogen concentration is also the parameter of the major concern because of its flammability and the risk of the detonation. In Korea, there have been an extensive efforts to develop the computer code which can analyze the severe accident behavior of the pressurized water reactor. The development has been done in a modularized manner and SACAP(Severe Accident Containment Analysis Package) code is now under final stage of development. SACAP code adopts LP(Lumped Parameter) model and is applicable to analyze the synthetic behavior of the containment during severe accident occurred by thermal-hydraulic transient, combustible gas burn, direct containment heating by high pressure melt ejection, steam explosion and molten core-concrete interaction. The analyses of a number of ISP(International Standard Problem) experiments were done as a part of the SACAP code V and V(verification and validation). In this paper, the SACAP analysis results for ISP-35 NUPEC and ISP-47 TOSQAN are presented including comparison with other existing NPP simulation codes. In this paper, we selected and analyzed ISP-35 NUPEC, ISP-47 TOSQAN in order to confirm the computational performance of SACAP code currently under development. Now the multi-node analysis for the ISP-47 is under process. As a result of simulation, SACAP predicts well the thermal-hydraulic variables such as temperature, pressure, etc. Also, we verify that SACAP code is properly equipped to analyze the gas distribution and condensation.
Simulating High Flux Isotope Reactor Core Thermal-Hydraulics via Interdimensional Model Coupling
Energy Technology Data Exchange (ETDEWEB)
Travis, Adam R [ORNL
2014-05-01
A coupled interdimensional model is presented for the simulation of the thermal-hydraulic characteristics of the High Flux Isotope Reactor core at Oak Ridge National Laboratory. The model consists of two domains a solid involute fuel plate and the surrounding liquid coolant channel. The fuel plate is modeled explicitly in three-dimensions. The coolant channel is approximated as a twodimensional slice oriented perpendicular to the fuel plate s surface. The two dimensionally-inconsistent domains are linked to one another via interdimensional model coupling mechanisms. The coupled model is presented as a simplified alternative to a fully explicit, fully three-dimensional model. Involute geometries were constructed in SolidWorks. Derivations of the involute construction equations are presented. Geometries were then imported into COMSOL Multiphysics for simulation and modeling. Both models are described in detail so as to highlight their respective attributes in the 3D model, the pursuit of an accurate, reliable, and complete solution; in the coupled model, the intent to simplify the modeling domain as much as possible without affecting significant alterations to the solution. The coupled model was created with the goal of permitting larger portions of the reactor core to be modeled at once without a significant sacrifice to solution integrity. As such, particular care is given to validating incorporated model simplifications. To the greatest extent possible, the decrease in solution time as well as computational cost are quantified versus the effects such gains have on the solution quality. A variant of the coupled model which sufficiently balances these three solution characteristics is presented alongside the more comprehensive 3D model for comparison and validation.
Neutronic and thermal-hydraulic coupling for 3D reactor core modeling combining MCB and fluent
Directory of Open Access Journals (Sweden)
Królikowski Igor P.
2015-09-01
Full Text Available Three-dimensional simulations of neutronics and thermal hydraulics of nuclear reactors are a tool used to design nuclear reactors. The coupling of MCB and FLUENT is presented, MCB allows to simulate neutronics, whereas FLUENT is computational fluid dynamics (CFD code. The main purpose of the coupling is to exchange data such as temperature and power profile between both codes. Temperature required as an input parameter for neutronics is significant since cross sections of nuclear reactions depend on temperature. Temperature may be calculated in thermal hydraulics, but this analysis needs as an input the power profile, which is a result from neutronic simulations. Exchange of data between both analyses is required to solve this problem. The coupling is a better solution compared to the assumption of estimated values of the temperatures or the power profiles; therefore the coupled analysis was created. This analysis includes single transient neutronic simulation and several steady-state thermal simulations. The power profile is generated in defined points in time during the neutronic simulation for the thermal analysis to calculate temperature. The coupled simulation gives information about thermal behavior of the reactor, nuclear reactions in the core, and the fuel evolution in time. Results show that there is strong influence of neutronics on thermal hydraulics. This impact is stronger than the impact of thermal hydraulics on neutronics. Influence of the coupling on temperature and neutron multiplication factor is presented. The analysis has been performed for the ELECTRA reactor, which is lead-cooled fast reactor concept, where the coolant fl ow is generated only by natural convection
International Nuclear Information System (INIS)
Guertin, Chantal
1995-01-01
This thesis is part of the validation process of using coupled 3D neutronics and thermal-hydraulics codes for studying accidental situations with boiling. First part is dedicated to a numerical stability analysis of neutronics and thermal-hydraulics coupled schemes. Both explicit and semi-implicit coupling schemes were applied to solve the set of equations describing the linearized neutronics and thermal-hydraulics of point reactor. Point reactor modelling was preferred to obtain analytical expressions of eigenvalues of the discretized Systems. Stability criteria, based on eigenvalues, was calculated as well as neutronic and thermalhydraulic responses of the System following insertion of a reactivity step. Results show no severe restriction of time domain, stability wise. Actual transient calculations using coupled neutronics and thermal-hydraulics codes, like COCCINELLE and THYC developed at Electricite de France, do not show stability problems. Second part introduces surface spline as a new neutronic feedback model. The cross influences of feedback parameters is now taken into account. Moderator temperature and density were modeled. This method, simple and accurate, allows an homogeneous description of cross-sections overall operating reactor situations including accidents with boiling. (author) [fr
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
Atucha II NPP full scope simulator modelling with the thermal hydraulic code TRACRT
International Nuclear Information System (INIS)
Alonso, Pablo Rey; Ruiz, Jose Antonio; Rivero, Norberto
2011-01-01
In February 2010 NA-SA (Nucleoelectrica Argentina S.A.) awarded Tecnatom the Atucha II full scope simulator project. NA-SA is a public company owner of the Argentinean nuclear power plants. Atucha II is due to enter in operation shortly. Atucha II NPP is a PHWR type plant cooled by the water of the Parana River and has the same design as the Atucha I unit, doubling its power capacity. Atucha II will produce 745 MWe utilizing heavy water as coolant and moderator, and natural uranium as fuel. A plant singular feature is the permanent core refueling. TRAC R T is the first real time thermal hydraulic six-equations code used in the training simulation industry for NSSS modeling. It is the result from adapting to real time the best estimate code TRACG. TRAC R T is based on first principle conservation equations for mass, energy and momentum for liquid and steam phases, with two phase flows under non homogeneous and non equilibrium conditions. At present, it has been successfully implemented in twelve full scope replica simulators in different training centers throughout the world. To ease the modeling task, TRAC R T includes a graphical pre-processing tool designed to optimize this process and alleviate the burden of entering alpha numerical data in an input file. (author)
Computational models for thermal-hydraulic assessment of TADSEA and its use for hydrogen production
International Nuclear Information System (INIS)
Rojas, L.; Gonzalez, D.; Garcia, C.; Gamez, A.; Garcia, L.; Lira, C. A. B. O.
2015-01-01
The Transmutation Advanced Device for Sustainable Energy Applications (TADSEA) is a pebble-bed Accelerator Driven System (ADS) with a graphite-gas configuration, designed for nuclear waste transmutation and for obtaining heat at very high temperatures to produce hydrogen. In previous work, the TADSEA's nuclear core was considered as a porous medium performed with a CFD code and thermal-hydraulic studies of the nuclear core were presented. In this paper, three critical fuel elements groups were defined regarding their position inside the core. In this article, the heat transfer from the fuel to the coolant was analyzed for the three core states during normal operation. The heat transfer inside the spherical fuel elements was also studied with a realistic CFD model of the critical elements groups. During the steady state, no critical elements reached the limit temperature of this type of fuel. Also, it is presented a model built in ANSYS for the simulation and optimization of high- temperature electrolysis using the TADSEA as a heat source. A flow diagram of the electrolysis process with the high temperature electrolyzer as the main component using TADSEA as an energy source is finally proposed and discussed. (Author)
International Nuclear Information System (INIS)
1984-08-01
The MELCOR project is developing a new generation of risk assessment computer programs for analysis of severe accidents at nuclear power plants. This document details the findings in the thermal-hydraulic areas for PWRs and areas common to both PWRs and BWRs. The study was performed by specialists from each of the various nuclear plant design or analysis areas where thermal-hydraulics is important. The study found that the current level of thermal-hydraulic modeling that exists for performing risk assessments is typified by the modeling in the MARCH code. This level of modeling was found to be inadequate for MELCOR because of inconsistent or nonexistent modeling of important phenomena, particularly invessel phenomena, systems, and structures. Pressurized water reactor modeling was found to be more complete than modeling for boiling water reactor designs, particularly in containment. Finally, although reactor cavity modeling was considered essential for risk assessments, the lack of adequate modeling found for the cavity was identified as a serious impediment to the development of second-generation risk codes
A time-dependent neutron transport model and its coupling to thermal-hydraulics
International Nuclear Information System (INIS)
Pautz, A.
2001-01-01
A new neutron transport code for time-dependent analyses of nuclear systems has been developed. The code system is based on the well-known Discrete Ordinates code DORT, which solves the steady-state neutron/photon transport equation in two dimensions for an arbitrary number of energy groups and the most common regular geometries. For the implementation of time-dependence a fully implicit first-order scheme was employed to minimize errors due to temporal discretization. This requires various modifications to the transport equation as well as the extensive use of elaborated acceleration mechanisms. The convergence criteria for fluxes, fission rates etc. had to be strongly tightened to ensure the reliability of results. To perform coupled analyses, an interface to the GRS system code ATHLET has been developed. The nodal power densities from the neutron transport code are passed to ATHLET to calculate thermal-hydraulic system parameters, e.g. fuel and coolant temperatures. These are in turn used to generate appropriate nuclear cross sections by interpolation of pre-calculated data sets for each time step. Finally, to demonstrate the transient capabilities of the coupled code system, the research reactor FRM-II has been analysed. Several design basis accidents were modelled, like the loss of off site power, loss of secondary heat sink and unintended control rod withdrawal. (author)
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
International Nuclear Information System (INIS)
Yudov, Y.V.
2001-01-01
The functional part of the KORSAR computer code is based on the computational unit for the reactor system thermal-hydraulics and other thermal power systems with water cooling. The two-phase flow dynamics of the thermal-hydraulic network is modelled by KORSAR in one-dimensional two-fluid (non-equilibrium and nonhomogeneous) approximation with the same pressure of both phases. Each phase is characterized by parameters averaged over the channel sections, and described by the conservation equations for mass, energy and momentum. The KORSAR computer code relies upon a novel approach to mathematical modelling of two-phase dispersed-annular flows. This approach allows a two-fluid model to differentiate the effects of the liquid film and droplets in the gas core on the flow characteristics. A semi-implicit numerical scheme has been chosen for deriving discrete analogs the conservation equations in KORSAR. In the semi-implicit numerical scheme, solution of finite-difference equations is reduced to the problem of determining the pressure field at a new time level. For the one-channel case, the pressure field is found from the solution of a system of linear algebraic equations by using the tri-diagonal matrix method. In the branched network calculation, the matrix of coefficients in the equations describing the pressure field is no longer tri-diagonal but has a sparseness structure. In this case, the system of linear equations for the pressure field can be solved with any of the known classical methods. Such an approach is implemented in the existing best-estimate thermal-hydraulic computer codes (TRAC, RELAP5, etc.) For the KORSAR computer code, we have developed a new non-iterative method for calculating the pressure field in the network of any topology. This method is based on the tri-diagonal matrix method and performs well when solving the thermal-hydraulic network problems. (author)
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
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
International Nuclear Information System (INIS)
Ohshima, Hiroyuki
2001-10-01
A whole core thermal-hydraulic analysis program ACT is being developed for the purpose of evaluating detailed in-core thermal hydraulic phenomena of fast reactors including the effect of the flow between wrapper-tube walls (inter-wrapper flow) under various reactor operation conditions. As appropriate boundary conditions in addition to a detailed modeling of the core are essential for accurate simulations of in-core thermal hydraulics, ACT consists of not only fuel assembly and inter-wrapper flow analysis modules but also a heat transport system analysis module that gives response of the plant dynamics to the core model. This report describes incorporation of a simplified model to the fuel assembly analysis module and program parallelization by a message passing method toward large-scale simulations. ACT has a fuel assembly analysis module which can simulate a whole fuel pin bundle in each fuel assembly of the core and, however, it may take much CPU time for a large-scale core simulation. Therefore, a simplified fuel assembly model that is thermal-hydraulically equivalent to the detailed one has been incorporated in order to save the simulation time and resources. This simplified model is applied to several parts of fuel assemblies in a core where the detailed simulation results are not required. With regard to the program parallelization, the calculation load and the data flow of ACT were analyzed and the optimum parallelization has been done including the improvement of the numerical simulation algorithm of ACT. Message Passing Interface (MPI) is applied to data communication between processes and synchronization in parallel calculations. Parallelized ACT was verified through a comparison simulation with the original one. In addition to the above works, input manuals of the core analysis module and the heat transport system analysis module have been prepared. (author)
International Nuclear Information System (INIS)
Kim, Kyung-Doo; Jeong, Jae-Jun; Lee, Seung-Wook; Lee, Myeong-Soo; Suh, Jae-Seung; Hong, Jin-Hyuk; Lee, Yong-Kwan
2004-01-01
The Nuclear Steam Supply System (NSSS) thermal-hydraulic model adopted in the Korea Nuclear Plant Education Center (KNPEC)-2 simulator was provided in the early 1980s. The reference plant for KNPEC-2 is the Yong Gwang Nuclear Unit 1, which is a Westinghouse-type 3-loop, 950 MW(electric) pressurized water reactor. Because of the limited computational capability at that time, it uses overly simplified physical models and assumptions for a real-time simulation of NSSS thermal-hydraulic transients. This may entail inaccurate results and thus, the possibility of so-called ''negative training,'' especially for complicated two-phase flows in the reactor coolant system. To resolve the problem, we developed a realistic NSSS thermal-hydraulic program (named ARTS code) based on the best-estimate code RETRAN-3D. The systematic assessment of ARTS has been conducted by both a stand-alone test and an integrated test in the simulator environment. The non-integrated stand-alone test (NIST) results were reasonable in terms of accuracy, real-time simulation capability, and robustness. After successful completion of the NIST, ARTS was integrated with a 3-D reactor kinetics model and other system models. The site acceptance test (SAT) has been completed successively and confirmed to comply with the ANSI/ANS-3.5-1998 simulator software performance criteria. This paper presents our efforts for the ARTS development and some test results of the NIST and SAT
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.
Energy Technology Data Exchange (ETDEWEB)
Nithianandan, C.K.; Klingenfus, J.A.; Reilly, S.S. [B& W Nuclear Technologies, Lynchburg, VA (United States)
1995-09-01
Droplet breakup at spacer grids and a cladding swelled and ruptured locations plays an important role in the cooling of nuclear fuel rods during the reflooding period of a loss-of-coolant accident (LOCA) in a pressurized water reactor (PWR). During the reflood phase, a spacer grid affects the thermal-hydraulic system behavior through increased turbulence, droplet breakup due to impact on grid straps, grid rewetting, and liquid holdup due to grid form losses. Recently, models to simulate spacer grid effects and blockage and rupture effects on system thermal hydraulics were added to the B&W Nuclear Technologies (BWNT) version of the RELAP5/MOD2 computer code. Several FLECHT-SEASET forced reflood tests, CCTF Tests C1-19 and C2-6, SCTF Test S3-15, and G2 Test 561 were simulated using RELAP5/MOD2-B&W to verify the applicability of the model at the cladding swelled and rupture locations. The results demonstrate the importance of modeling the thermal-hydraulic effects due to grids, and clad swelling and rupture to correctly predict the clad temperature response during the reflood phase of large break LOCA. The RELAP5 models and the test results are described in this paper.
Energy Technology Data Exchange (ETDEWEB)
Chung, Ji Bum [Institute for Advanced Engineering, Yongin (Korea, Republic of); Park, Jong Woon [Korea Electric Power Research Institute, Taejon (Korea, Republic of)
1998-12-31
In order to enhance the dynamic and interactive simulation capability of a system thermal hydraulic code for nuclear power plant, applicability of flow network models in SINDA/FLUINT{sup TM} has been tested by modeling feedwater system and coupling to DSNP which is one of a system thermal hydraulic simulation code for a pressurized heavy water reactor. The feedwater system is selected since it is one of the most important balance of plant systems with a potential to greatly affect the behavior of nuclear steam supply system. The flow network model of this feedwater system consists of condenser, condensate pumps, low and high pressure heaters, deaerator, feedwater pumps, and control valves. This complicated flow network is modeled and coupled to DSNP and it is tested for several normal and abnormal transient conditions such turbine load maneuvering, turbine trip, and loss of class IV power. The results show reasonable behavior of the coupled code and also gives a good dynamic and interactive simulation capabilities for the several mild transient conditions. It has been found that coupling system thermal hydraulic code with a flow network code is a proper way of upgrading simulation capability of DSNP to mature nuclear plant analyzer (NPA). 5 refs., 10 figs. (Author)
Thermal hydraulics development for CASL
Energy Technology Data Exchange (ETDEWEB)
Lowrie, Robert B [Los Alamos National Laboratory
2010-12-07
This talk will describe the technical direction of the Thermal-Hydraulics (T-H) Project within the Consortium for Advanced Simulation of Light Water Reactors (CASL) Department of Energy Innovation Hub. CASL is focused on developing a 'virtual reactor', that will simulate the physical processes that occur within a light-water reactor. These simulations will address several challenge problems, defined by laboratory, university, and industrial partners that make up CASL. CASL's T-H efforts are encompassed in two sub-projects: (1) Computational Fluid Dynamics (CFD), (2) Interface Treatment Methods (ITM). The CFD subproject will develop non-proprietary, scalable, verified and validated macroscale CFD simulation tools. These tools typically require closures for their turbulence and boiling models, which will be provided by the ITM sub-project, via experiments and microscale (such as DNS) simulation results. The near-term milestones and longer term plans of these two sub-projects will be discussed.
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.
International Nuclear Information System (INIS)
Ohshima, Hiroyuki
2003-03-01
The thermal-hydraulic analysis computer program ACT is under development for the evaluation of detailed flow and temperature fields in a core region of fast breeder reactors under various operation conditions. The purpose of this program development is to contribute not only to clarifying thermal hydraulic characteristics that cannot be revealed by experiments due to measurement difficulty but also to performing rational safety design and assessment. This report describes the incorporation of a three-dimensional upper plenum model to ACT and its verification study as part of the program development. To treat the influence of three-dimensional thermal-hydraulic behavior in a upper plenum on the in-core temperature field, the multi-dimensional general purpose thermal-hydraulic analysis program AQUA, which was developed and validated at JNC, was applied as the base of the upper plenum analysis module of ACT. AQUA enables to model the upper plenum configuration including immersed heat exchangers of the direct reactor auxiliary cooling system (DRACS). In coupling core analysis module that consists of the fuel-assembly and the inter-wrapper gap calculation parts with the upper plenum module, different types of computation mesh systems were jointed using the staggered quarter assembly mesh scheme. A coupling algorithm among core, upper plenum and heat transport system modules, which can keep mass, momentum and energy conservation, was developed and optimized in consideration of parallel computing. ACT was applied to analyzing a sodium experiment (PLANDTL-DHX) performed at JNC, which simulated the natural circulation decay heat removal under DRACS operation conditions for the program verification. From the calculation result, the validity of the improved program was confirmed. (author)
International Nuclear Information System (INIS)
Unal, C.; Nelson, R.
1991-01-01
After completion of the thermal-hydraulic model developed in a companion paper, the authors performed developmental assessment calculation of the model using steady-state and transient post-critical heat flux (CHF) data. This paper discusses the results of those calculations. The overall interfacial drag model predicted reasonable drag coefficients for both the nucleate boiling and the inverted annular flow (IAF) regimes. The predicted pressure drops agreed reasonably well with the measured data of two transient experiments, CCTF Run 14 and a Lehigh reflood test. The thermal-hydraulic model for post-CHF convective heat transfer predicted the rewetting velocities reasonably well for both experiments. The predicted average slope of the wall temperature traces for these tests showed reasonable agreement with the measured data, indicating that the transient-calculated precursory cooling rates agreed with measured data. The hot-patch model, in conjunction with the other thermal-hydraulic models, was capable of modeling the Winfrith post-CHF hot-patch experiments. The hot-patch model kept the wall temperatures at the specified levels in the hot-patch regions and did not allow any quench-front propagation from either the bottom or the top of the test section. The interfacial heat-transfer model tended to slightly underpredict the vapor temperatures. The maximum difference between calculated and measured vapor temperatures was 20%, with a 10% difference for the remainder of the runs considered. The wall-to-fluid heat transfer was predicted reasonably well, and the predicted wall temperatures were in reasonable agreement with measured data with a maximum relative error of less than 13%
Numerical analysis of thermal hydraulics in secondary side steam generator
International Nuclear Information System (INIS)
Yang Zhilin
1997-01-01
The author presents the analysis on the possibility for COBRA-TF to model thermal-hydraulics of SG (Steam Generator) secondary side, numerically analyzes the thermal-hydraulic processes of heater section of SG secondary side of Qinshan Nuclear Power Plant, the results are compared with those from ALBERTINE-2
International Nuclear Information System (INIS)
Ren Zhihao; Kong Xiangyin; Tsai Chiungwen; Ruan Jialei; Li Jinggang; Ma Zhongying; Yan Jianxing; Ma Yinxiang
2015-01-01
A system transient thermal-hydraulic analysis code for PWRs named GINKGO is being developed as part of the indigenous effort of China General Nuclear Power Corp. (CGN) to develop a full-spectrum software package for reactor design and safety analysis. Implemented using the Object-Oriented programming technology, GINKGO is designed to be used for simulating all PWR transients except LBLOCA. The primary physical models and key algorithms applied in GINKGO and also the preliminary validation with the phenomena cases are introduced in the paper. To account for reactor coolant transients, the separated phase model under thermal equilibrium is used in the code. The three governing mixture balance equations augmented with Chexal-Lellouche drift-flux model to determine phase velocities are solved at each time step. Thermal equilibrium between the vapor and liquid phases is assumed with the exception of the upper head volume and pressurizer. And two-region non-equilibrium model and multi-region non-equilibrium model are available for the pressurizer simulation. The reactor point kinetics model with six groups of delayed neutrons, the partial derivative approximation of the DNBR model and decay heat model are combined to give a full description for the reactor core. The additional component model, engineered safety system model and models for other auxiliary systems in GINKGO demonstrate a complete capability for PWR safety analysis and thermal-hydraulic design. A fully implicit solution algorithm involving pressure search is applied in GINKGO to improve the stability of the solution method, especially when two-phase conditions with unequal phase velocities exist. Different phenomena cases are set up to demonstrate the capability of GINKGO used in different boundary conditions, steady state achievement, reverse and branch flow, etc. The GINKGO code uses the C/C++ programming language to take advantage of the language's inherent Object Oriented characteristic and to
Directory of Open Access Journals (Sweden)
Esch Markus
2014-01-01
Full Text Available For future high temperature reactor projects, e. g., for electricity production or nuclear process heat applications, the steam generator is a crucial component. A typical design is a helical coil steam generator consisting of several tubes connected in parallel forming cylinders of different diameters. This type of steam generator was a significant component used at the thorium high temperature reactor. In the work presented the temperature profile is being analyzed by the nodal thermal hydraulics code TRACE for the thorium high temperature reactor steam generator. The influence of the nodalization is being investigated within the scope of this study and compared to experimental results from the past. The results of the standard TRACE code are compared to results using a modified Nusselt number for the primary side. The implemented heat transfer correlation was developed within the past German HTR program. This study shows that both TRACE versions are stable and provides a discussion of the nodalization requirements.
International Nuclear Information System (INIS)
Saghafi, Mahdi; Ghofrani, Mohammad Bagher; D’Auria, Francesco
2016-01-01
Highlights: • A thermal-hydraulic nodalization for PSB-VVER test facility has been developed. • Station blackout accident is modeled with the developed nodalization in MELCOR code. • The developed nodalization is qualified at both steady state and transient levels. • MELCOR predictions are qualitatively and quantitatively in acceptable range. • Fast Fourier Transform Base Method is used to quantify accuracy of code predictions. - Abstract: This paper deals with the development of a qualified thermal-hydraulic nodalization for modeling Station Black-Out (SBO) accident in PSB-VVER Integral Test Facility (ITF). This study has been performed in the framework of a research project, aiming to develop an appropriate accident management support tool for Bushehr nuclear power plant. In this regard, a nodalization has been developed for thermal-hydraulic modeling of the PSB-VVER ITF by MELCOR integrated code. The nodalization is qualitatively and quantitatively qualified at both steady-state and transient levels. The accuracy of the MELCOR predictions is quantified in the transient level using the Fast Fourier Transform Base Method (FFTBM). FFTBM provides an integral representation for quantification of the code accuracy in the frequency domain. It was observed that MELCOR predictions are qualitatively and quantitatively in the acceptable range. In addition, the influence of different nodalizations on MELCOR predictions was evaluated and quantified using FFTBM by developing 8 sensitivity cases with different numbers of control volumes and heat structures in the core region and steam generator U-tubes. The most appropriate case, which provided results with minimum deviations from the experimental data, was then considered as the qualified nodalization for analysis of SBO accident in the PSB-VVER ITF. This qualified nodalization can be used for modeling of VVER-1000 nuclear power plants when performing SBO accident analysis by MELCOR code.
Rais, A.; Siefman, D.; Girardin, G.; Hursin, M.; Pautz, A.
2015-01-01
In order to analyze the steady state and transient behavior of the CROCUS reactor, several methods and models need to be developed in the areas of reactor physics, thermal-hydraulics, and multiphysics coupling. The long-term objectives of this project are to work towards the development of a modern method for the safety analysis of research reactors and to update the Final Safety Analysis Report of the CROCUS reactor. A first part of the paper deals with generation of a core simulator nuclear...
International Nuclear Information System (INIS)
Jahanfarnia, G.; Zarifi, E.; Veysi, F.
2013-01-01
The aim of this study was to perform a thermal-hydraulic analysis of nanofluids as coolant in the Bushehr VVER-1000 reactor core using the porous media approach. Water-based nanofluids containing various volume fractions of Al 2 O 3 and TiO 2 nanoparticles were analyzed. The conservation equations were discretized by the finite volume method and solved by numerical methods. To validate the approaches applied in this study, the results of the model and COBRA-EN code were compared for pure water. The achieved results show that the temperature of the coolant increases with the concentration of the nanoparticles. (authors)
Conversion of the thermal hydraulics components of Almaraz NPP model from RELAP5 into TRAC-M
International Nuclear Information System (INIS)
Queral, C.; Mulas, J.; Collazo, I.; Concejal, A.; Burbano, N.; Gallego, I.; Lopez Lechas, A.
2002-01-01
In the scope of a joint project between the Spanish Nuclear Regulatory Commission (CSN) and the electric energy industry of Spain (UNESA) on the USNRC state-of-the-art thermal hydraulic code, TRAC-M, there is a task relating to the translation of the Spanish NPP models from other TH codes to the new one. As part of this project, our team is working on the translation of Almaraz NPP model from RELAP5/MOD3.2 to TRAC-M. At present, several portions of the input deck have been converted to TRAC-M, and the output data have also been compared with RELAP5 data. This paper refers to the translation of the following thermal hydraulic models: pressurizer, hot and cold legs (including the pumps and the injection systems), and steam generators. The comparison of the results obtained with both codes shows a good agreement. However, some difficulties have been found in the translation of some code components, like pipes, heat structures, pumps, branchs, valves and boundary conditions. In this paper, these translation problems and their solutions are described.(author)
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
Energy Technology Data Exchange (ETDEWEB)
Chijimatsu, M.; Borgesson, L.; Fujita, T.; Jussila, P.; Nguyen, S.; Rutqvist, J.; Jing, L.; Hernelind, J.
2009-02-01
In Task A of the international DECOVALEX-THMC project, five research teams study the influence of thermal-hydro-mechanical (THM) coupling on the safety of a hypothetical geological repository for spent fuel. In order to improve the analyses, the teams calibrated their bentonite models with results from laboratory experiments, including swelling pressure tests, water uptake tests, thermally gradient tests, and the CEA mock-up THM experiment. This paper describes the mathematical models used by the teams, and compares the results of their calibrations with the experimental data.
International Nuclear Information System (INIS)
Wagner, S.G.; Harris, K.P.; Mansur, J.M.
1981-01-01
PWR accident analyses for breaks in the secondary system are characterized by extreme cooling of the core, large inlet temperature gradients, asymmetric stuck control rods and primary coolant flow predictions ranging from a few percent of nominal to full flow. For several years Combustion Engineering has been using the Hermite code to compute reacitvities, power distributions and flow distributions in 3-D for far off nominal thermal-hydraulic and neutronic conditions. These 3-D methods, while costly, have been very successful in analyzing a large class of PWR secondary break accidents. They indicate a much reduced post-scram return-to-power compared to less detailed, bounding core average feedback methods. The availability of a growing number of 3-D calculations has made it possible to develop and benchmark faster running of 2-D methods which account for many of the same effects seen in 3-D. Two such methods are described and evaluated in this paper and one method is shown to be realiable over a wide range of conditions. (orig.) [de
Energy Technology Data Exchange (ETDEWEB)
Reis, Patricia A.L.; Costa, Antonella L.; Hamers, Adolfo R.; Pereira, Claubia; Rodrigues, Thiago D.A.; Mantecon, Javier G.; Veloso, Maria A.F., E-mail: patricialire@yahoo.com.br, E-mail: antonella@nuclear.ufmg.br, E-mail: adolforomerohamers@hotmail.com, E-mail: claubia@nuclear.ufmg.br, E-mail: thiagodanielbh@gmail.com, E-mail: mantecon1987@gmail.com, E-mail: dora@nuclear.ufmg.br [Universidade Federal de Minas Gerais (UFMG), Belo Horizonte, MG (Brazil). Departamento de Engenharia Nuclear; Instituto Nacional de Ciencias e Tecnologia de Reatores Nucleares Inovadores (INCT/CNPq), Belo Horizonte (Brazil); Miro, Rafael; Verdu, Gumersindo, E-mail: rmiro@iqn.upv.es, E-mail: gverdu@iqn.upv.es [Universidad Politecnica de Valencia (Spain). Departamento de Ingenieria Quimica y Nuclear
2015-07-01
The computational advances observed in the last two decades have been provided direct impact on the researches related to nuclear simulations, which use several types of computer codes, including coupled between them, allowing representing with very accuracy the behavior of nuclear plants. Studies of complex scenarios in nuclear reactors have been improved by the use of thermal-hydraulic (TH) and neutron kinetics (NK) coupled codes. This technique consists in incorporating three-dimensional (3D) neutron modeling of the reactor core into codes, mainly to simulate transients that involve asymmetric core spatial power distributions and strong feedback effects between neutronics and reactor thermal-hydraulics. Therefore, this work presents preliminary results of TH RELAP5 and the NK PARCS calculations applied to model of the Angra 2 reactor. The WIMSD-5B code has been used to generate the macroscopic cross sections used in the NK code. The results obtained are satisfactory and represent important part of the development of this methodology. The next step is to couple the codes. (author)
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
International Nuclear Information System (INIS)
McGraw, D.; Oberlander, P.
2007-01-01
The purpose of this study is to report on the results of a preliminary modeling framework to investigate the causes of the large hydraulic gradient north of Yucca Mountain. This study builds on the Saturated Zone Site-Scale Flow and Transport Model (referenced herein as the Site-scale model (Zyvoloski, 2004a)), which is a three-dimensional saturated zone model of the Yucca Mountain area. Groundwater flow was simulated under natural conditions. The model framework and grid design describe the geologic layering and the calibration parameters describe the hydrogeology. The Site-scale model is calibrated to hydraulic heads, fluid temperature, and groundwater flowpaths. One area of interest in the Site-scale model represents the large hydraulic gradient north of Yucca Mountain. Nearby water levels suggest over 200 meters of hydraulic head difference in less than 1,000 meters horizontal distance. Given the geologic conceptual models defined by various hydrogeologic reports (Faunt, 2000, 2001; Zyvoloski, 2004b), no definitive explanation has been found for the cause of the large hydraulic gradient. Luckey et al. (1996) presents several possible explanations for the large hydraulic gradient as provided below: The gradient is simply the result of flow through the upper volcanic confining unit, which is nearly 300 meters thick near the large gradient. The gradient represents a semi-perched system in which flow in the upper and lower aquifers is predominantly horizontal, whereas flow in the upper confining unit would be predominantly vertical. The gradient represents a drain down a buried fault from the volcanic aquifers to the lower Carbonate Aquifer. The gradient represents a spillway in which a fault marks the effective northern limit of the lower volcanic aquifer. The large gradient results from the presence at depth of the Eleana Formation, a part of the Paleozoic upper confining unit, which overlies the lower Carbonate Aquifer in much of the Death Valley region. The
Energy Technology Data Exchange (ETDEWEB)
McGraw, D.; Oberlander, P.
2007-12-18
The purpose of this study is to report on the results of a preliminary modeling framework to investigate the causes of the large hydraulic gradient north of Yucca Mountain. This study builds on the Saturated Zone Site-Scale Flow and Transport Model (referenced herein as the Site-scale model (Zyvoloski, 2004a), which is a three-dimensional saturated zone model of the Yucca Mountain area. Groundwater flow was simulated under natural conditions. The model framework and grid design describe the geologic layering and the calibration parameters describe the hydrogeology. The Site-scale model is calibrated to hydraulic heads, fluid temperature, and groundwater flowpaths. One area of interest in the Site-scale model represents the large hydraulic gradient north of Yucca Mountain. Nearby water levels suggest over 200 meters of hydraulic head difference in less than 1,000 meters horizontal distance. Given the geologic conceptual models defined by various hydrogeologic reports (Faunt, 2000, 2001; Zyvoloski, 2004b), no definitive explanation has been found for the cause of the large hydraulic gradient. Luckey et al. (1996) presents several possible explanations for the large hydraulic gradient as provided below: The gradient is simply the result of flow through the upper volcanic confining unit, which is nearly 300 meters thick near the large gradient. The gradient represents a semi-perched system in which flow in the upper and lower aquifers is predominantly horizontal, whereas flow in the upper confining unit would be predominantly vertical. The gradient represents a drain down a buried fault from the volcanic aquifers to the lower Carbonate Aquifer. The gradient represents a spillway in which a fault marks the effective northern limit of the lower volcanic aquifer. The large gradient results from the presence at depth of the Eleana Formation, a part of the Paleozoic upper confining unit, which overlies the lower Carbonate Aquifer in much of the Death Valley region. The
Two-fluid modeling of thermal-hydraulic phenomena for best-estimate LWR safety analysis
International Nuclear Information System (INIS)
Yadigaroglu, G.; Andreani, M.
1989-01-01
Two-fluid formulation of the conservation equations has allowed modelling of the two-phase flow and heat transfer phenomena and situations involving strong departures in thermal and velocity equilibrium between the phases. The paper reviews the state of the art in modelling critical flows, and certain phase separation phenomena, as well as post-dryout heat transfer situations. Although the two-fluid models and the codes have the potential for correctly modelling such situations, this potential has not always been fully used in practice. (orig.)
DEFF Research Database (Denmark)
Madsen, Henrik; Zhou, Jianjun; Hansen, Lars Henrik
1997-01-01
This paper describes a case study of identifying the physical model (or the grey box model) of a hydraulic test robot. The obtained model is intended to provide a basis for model-based control of the robot. The physical model is formulated in continuous time and is derived by application...
Energy Technology Data Exchange (ETDEWEB)
Badea, Aurelian F., E-mail: aurelian.badea@kit.edu [Karlsruhe Institute of Technology, Vincenz-Prießnitz-Str. 3, 76131 Karlsruhe (Germany); Cacuci, Dan G. [Center for Nuclear Science and Energy/Dept. of ME, University of South Carolina, 300 Main Street, Columbia, SC 29208 (United States)
2017-03-15
Highlights: • BWR Turbine Trip 2 (BWR-TT2) benchmark. • Substantial (up to 50%) reduction of uncertainties in the predicted transient power. • 6660 uncertain model parameters were calibrated. - Abstract: By applying a comprehensive predictive modeling methodology, this work demonstrates a substantial (up to 50%) reduction of uncertainties in the predicted total transient power in the BWR Turbine Trip 2 (BWR-TT2) benchmark while calibrating the numerical simulation of this benchmark, comprising 6090 macroscopic cross sections, and 570 thermal-hydraulics parameters involved in modeling the phase-slip correlation, transient outlet pressure, and total mass flow. The BWR-TT2 benchmark is based on an experiment that was carried out in 1977 in the NPP Peach Bottom 2, involving the closure of the turbine stop valve which caused a pressure wave that propagated with attenuation into the reactor core. The condensation of the steam in the reactor core caused by the pressure increase led to a positive reactivity insertion. The subsequent rise of power was limited by the feedback and the insertion of the control rods. The BWR-TT2 benchmark was modeled with the three-dimensional reactor physics code system DYN3D, by coupling neutron kinetics with two-phase thermal-hydraulics. All 6660 DYN3D model parameters were calibrated by applying a predictive modeling methodology that combines experimental and computational information to produce optimally predicted best-estimate results with reduced predicted uncertainties. Simultaneously, the predictive modeling methodology yields optimally predicted values for the BWR total transient power while reducing significantly the accompanying predicted standard deviations.
Thermal hydraulic stability experiments in rod bundle
International Nuclear Information System (INIS)
Enomoto, T.; Muto, S.; Ishizuka, T.; Tanabe, A.; Mitsutake, T.; Sakurai, M.
1985-01-01
Thermal hydraulic stability tests have been performed on electrically heated bundles to simulate Boiling Water Reactor (BWR) fuels in a parallel channel test-loop. The test facility used is for the study of the steady state and transient characteristics of various thermal hydraulic conditions encountered in BWR operation, such as flow- high power operation, abnormal transient conditions and post boiling transition, including thermal hydraulic stability. Moreover, steady state and transient void behavior can be measured using an additional test section for this facility
Energy Technology Data Exchange (ETDEWEB)
Wilson, G.J.; Kazimi, M.S.
1980-05-01
Several different models and correlations were developed and incorporated in the sodium version of THERMIT, a thermal-hydraulics code written at MIT for the purpose of analyzing transients under LMFBR conditions. This includes: a mechanism for the inclusion of radial heat conduction in the sodium coolant as well as radial heat loss to the structure surrounding the test section. The fuel rod conduction scheme was modified to allow for more flexibility in modelling the gas plenum regions and fuel restructuring. The formulas for mass and momentum exchange between the liquid and vapor phases were improved. The single phase and two phase friction factors were replaced by correlations more appropriate to LMFBR assembly geometry.
CFD model development and data comparison for thermal-hydraulic analysis of HTO pilot scale reactor
International Nuclear Information System (INIS)
Kochan, R.J.; Oh, C.H.
1995-09-01
The DOE Hydrothermal Oxidation (HTO) program is validating computational methods for use in scaling up small HTO systems to production scale. As part of that effort, the computational fluid dynamics code FLUENT is being used to calculate the integrated fluid dynamics and chemical reactions in an HTO vessel reactor designed by MODAR, Inc. Previous validation of the code used data from a benchscale reactor. This reports presents the validation of the code using pilotscale (10 times greater throughput than benchscale) data. The model for the pilotscale reactor has been improved based upon the benchscale data by including better fluid thermal properties, a better solution algorithm, addition of external heat transfer, investigation of the effects of turbulent flow, and, although not built into the computer model, a technique for using the calculated adiabatic oxidation temperatures for selecting initial conditions. Thermal results from this model show very good agreement with the limited test data from MODAR Run 920. In addition to the reactor temperatures, flowfield details, including chemical reaction distribution, and simulated salt particle transport were obtained. This model will be very beneficial in designing and evaluating larger commercial scale units. The results of these calculations indicate that for model validation, more accurate boundary conditions need to be measured in future test runs
Regeneration in an internal combustion engine: Thermal-hydraulic modeling and analysis
International Nuclear Information System (INIS)
Thyageswaran, Sridhar
2016-01-01
Highlights: • An arrangement is proposed for in-cylinder regeneration in a 4-stroke engine. • Thermodynamic models are formulated for overall cycle analysis. • A design procedure is outlined for micro-channel regenerators. • Partial differential equations are solved for flow inside the regenerator. • Regeneration with lean combustion decreases the idealized cycle efficiency. - Abstract: An arrangement is proposed for a four-stroke internal combustion engine to: (a) recover thermal energy from products of combustion during the exhaust stroke; (b) store that energy as sensible heat in a micro-channel regenerator matrix; and (c) transfer the stored heat to compressed fresh charge that flows through the regenerator during the succeeding mechanical cycle. An extra moveable piston that can be locked at preferred positions and a sequence of valve events enable the regenerator to lose heat to the working fluid during one interval of time but gain heat from the fluid during another interval of time. This paper examines whether or not this scheme for in-cylinder regeneration (ICR) improves the cycle thermal efficiency η I . Models for various thermodynamic processes in the cycle and treatments for unsteady compressible flow and heat transfer inside the regenerator are developed. Digital simulations of the cycle are made. Compared to an idealized engine cycle devoid of regeneration, provisions for ICR seem to deteriorate the thermal efficiency. In an 8:1 compression ratio octane engine simulated with an equivalence ratio of 0.75, η I = 0.455 with regeneration and η I = 0.491 without. This study shows that previous claims on efficiency gains via ICR, using highly-simplified models, may be misleading.
THE THREE DIMENSIONAL THERMAL HYDRAULIC CODE BAGIRA.
Energy Technology Data Exchange (ETDEWEB)
KALINICHENKO,S.D.; KOHUT,P.; KROSHILIN,A.E.; KROSHILIN,V.E.; SMIRNOV,A.V.
2003-05-04
BAGIRA - a thermal-hydraulic program complex was primarily developed for using it in nuclear power plant simulator models, but is also used as a best-estimate analytical tool for modeling two-phase mixture flows. The code models allow consideration of phase transients and the treatment of the hydrodynamic behavior of boiling and pressurized water reactor circuits. It provides the capability to explicitly model three-dimensional flow regimes in various regions of the primary and secondary circuits such as, the mixing regions, circular downcomer, pressurizer, reactor core, main primary loops, the steam generators, the separator-reheaters. In addition, it is coupled to a severe-accident module allowing the analysis of core degradation and fuel damage behavior. Section II will present the theoretical basis for development and selected results are presented in Section III. The primary use for the code complex is to realistically model reactor core behavior in power plant simulators providing enhanced training tools for plant operators.
Energy Technology Data Exchange (ETDEWEB)
Cho, Hyoung Kyu; Park, Ik Kyu; Kim, Jong Tae; Yoon, Han Young; Jeong, Jae Jun
2008-09-15
A thermal-hydraulic analysis module, CUPID has been developed for a design and safety analysis of a nuclear reactor component. In the CUPID code, a two-fluid three-field model is adopted and the governing equations are solved on an unstructured grid to make CUPID advantageous for a flow analysis in complicated geometries. As for the numerical solution scheme, the semi-implicit method was adopted, which has proved to be stable and accurate for most applications of a nuclear reactor transient. A thermal analysis of a heat structure in a nuclear reactor component is very important for the safety analysis of a nuclear reactor as well as that of a fluid. Moreover, a conjugate heat transfer analysis between a fluid and a heat structure is indispensable in many transient situations of a nuclear reactor. For this reason, a heat conductor model was implemented into the CUPID module in this study. This report presents the governing equation, numerical method and verification results of the heat conductor model.
An approach to modeling coupled thermal-hydraulic-chemical processes in geothermal systems
Palguta, Jennifer; Williams, Colin F.; Ingebritsen, Steven E.; Hickman, Stephen H.; Sonnenthal, Eric
2011-01-01
Interactions between hydrothermal fluids and rock alter mineralogy, leading to the formation of secondary minerals and potentially significant physical and chemical property changes. Reactive transport simulations are essential for evaluating the coupled processes controlling the geochemical, thermal and hydrological evolution of geothermal systems. The objective of this preliminary investigation is to successfully replicate observations from a series of hydrothermal laboratory experiments [Morrow et al., 2001] using the code TOUGHREACT. The laboratory experiments carried out by Morrow et al. [2001] measure permeability reduction in fractured and intact Westerly granite due to high-temperature fluid flow through core samples. Initial permeability and temperature values used in our simulations reflect these experimental conditions and range from 6.13 × 10−20 to 1.5 × 10−17 m2 and 150 to 300 °C, respectively. The primary mineralogy of the model rock is plagioclase (40 vol.%), K-feldspar (20 vol.%), quartz (30 vol.%), and biotite (10 vol.%). The simulations are constrained by the requirement that permeability, relative mineral abundances, and fluid chemistry agree with experimental observations. In the models, the granite core samples are represented as one-dimensional reaction domains. We find that the mineral abundances, solute concentrations, and permeability evolutions predicted by the models are consistent with those observed in the experiments carried out by Morrow et al. [2001] only if the mineral reactive surface areas decrease with increasing clay mineral abundance. This modeling approach suggests the importance of explicitly incorporating changing mineral surface areas into reactive transport models.
Advanced Multiphysics Thermal-Hydraulics Models for the High Flux Isotope Reactor
Energy Technology Data Exchange (ETDEWEB)
Jain, Prashant K [ORNL; Freels, James D [ORNL
2015-01-01
Engineering design studies to determine the feasibility of converting the High Flux Isotope Reactor (HFIR) from using highly enriched uranium (HEU) to low-enriched uranium (LEU) fuel are ongoing at Oak Ridge National Laboratory (ORNL). This work is part of an effort sponsored by the US Department of Energy (DOE) Reactor Conversion Program. HFIR is a very high flux pressurized light-water-cooled and moderated flux-trap type research reactor. HFIR s current missions are to support neutron scattering experiments, isotope production, and materials irradiation, including neutron activation analysis. Advanced three-dimensional multiphysics models of HFIR fuel were developed in COMSOL software for safety basis (worst case) operating conditions. Several types of physics including multilayer heat conduction, conjugate heat transfer, turbulent flows (RANS model) and structural mechanics were combined and solved for HFIR s inner and outer fuel elements. Alternate design features of the new LEU fuel were evaluated using these multiphysics models. This work led to a new, preliminary reference LEU design that combines a permanent absorber in the lower unfueled region of all of the fuel plates, a burnable absorber in the inner element side plates, and a relocated and reshaped (but still radially contoured) fuel zone. Preliminary results of estimated thermal safety margins are presented. Fuel design studies and model enhancement continue.
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.
3D Coupled Thermal-Hydraulic Model of the Lower Yarmouk Gorge, Jordan Rift Valley
Walther, M.; Magri, F.; Inbar, N.; Möller, P.; Raggad, M.; Rödiger, T.; Rosenthal, E.; Shentsis, I.; Siebert, C.; Volpi, G.
2017-12-01
It is supposed that the Lower Yarmouk Gorge (LYG), in the Jordan Rift Valley acts as the mixing zone of two crossing flow pathways: N-S from the Hermon Mountains and from the Ajlun Dome, and E-W from Jebel al Arab Mountain in Syria (also known as Huran Plateau or Yarmouk drainage basin). As a result, several springs can be found within the gorge. These are characterized by widespread temperatures (20 - 60 °C) which indicate that, beside the complex regional flow, also ascending thermal waters control the hydrologic behavior of the LYG. Previous simulations based on a conceptual simplified 3D model (Magri et al., 2016) showed that crossing flow paths result from the coexistence of convection, that can develop for example along NE-SW oriented faults within the gorge or in permeable aquifers below Maastrichtian aquiclude, and additional flow fields that are induced by the N-S topographic gradients. Here we present the first 3D hydrogeological model of the entire LYG that includes structural features based on actual logs and interpreted seismic lines from both Israeli and Jordanian territories. The model distinguishes seven units from upper Eocene to the Lower Triassic, accounting for major aquifers, aquicludes and deep-cutting faults. Recharges are implemented based on the numerical representation developed by Shentsis (1990) that considers relationships between mean annual rain and topographic elevation. The model reveals that topography-driven N-S and E-W flows strongly control the location of discharge areas while the anomalous spring temperature is not necessarily linked to the presence of fault convection. Local permeability anisotropy due to aquifers folding or facies changes are features sufficient for the rising of hot fluids. Shentsis, I., 1990. Mathematical models for long-term prediction of mountainous river runoff: methods, information and results, Hydrological Sciences Journal, 35:5, 487-500 Magri, F., Möller, S., Inbar, N., Möller, P., Raggad, M., R
Energy Technology Data Exchange (ETDEWEB)
Dussel, M.; Lueschen, E.; Thomas, R. [Leibniz-Institut fuer Angewandte Geophysik, Hannover (DE)] (and others)
2011-10-24
The mutual potential influence of geothermal duplicates and the scientific investigation of the relationship between seismic and hydraulic parameters are investigated in the joint research project 'Geothermal characterization of fractured karst limestone aquifers in the Munich metropolitan area'. Thirteen doublets and triplets being in production or sunk illustrate the great geothermal potential and provide important data on the development of a thermal-hydraulic modeling of the reservoir. 3D seismic Unterhaching, 3D structural model, hydrogeological model and a high-resolution 3D temperature model form the basis of the numerical modeling. Different seismic signatures, seismic attributes and variations in the interval velocities characterize the ground geophysically, and were interpreted under consideration of geological and hydrogeological background information as well as borehole measurements in terms of hydraulically conductive homogeneous areas. For the Munich metropolitan area, the numerical modeling is a decision aid for the future optimized and sustainable hydrothermal utilization of the Malm aquifer.
Directory of Open Access Journals (Sweden)
A. Rais
2015-01-01
Full Text Available In order to analyze the steady state and transient behavior of the CROCUS reactor, several methods and models need to be developed in the areas of reactor physics, thermal-hydraulics, and multiphysics coupling. The long-term objectives of this project are to work towards the development of a modern method for the safety analysis of research reactors and to update the Final Safety Analysis Report of the CROCUS reactor. A first part of the paper deals with generation of a core simulator nuclear data library for the CROCUS reactor using the Serpent 2 Monte Carlo code and also with reactor core modeling using the PARCS code. PARCS eigenvalue, radial power distribution, and control rod reactivity worth results were benchmarked against Serpent 2 full-core model results. Using the Serpent 2 model as reference, PARCS eigenvalue predictions were within 240 pcm, radial power was within 3% in the central region of the core, and control rod reactivity worth was within 2%. A second part reviews the current methodology used for the safety analysis of the CROCUS reactor and presents the envisioned approach for the multiphysics modeling of the reactor.
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.
Directory of Open Access Journals (Sweden)
Douglas A. Fynan
2016-06-01
Full Text Available The Gaussian process model (GPM is a flexible surrogate model that can be used for nonparametric regression for multivariate problems. A unique feature of the GPM is that a prediction variance is automatically provided with the regression function. In this paper, we estimate the safety margin of a nuclear power plant by performing regression on the output of best-estimate simulations of a large-break loss-of-coolant accident with sampling of safety system configuration, sequence timing, technical specifications, and thermal hydraulic parameter uncertainties. The key aspect of our approach is that the GPM regression is only performed on the dominant input variables, the safety injection flow rate and the delay time for AC powered pumps to start representing sequence timing uncertainty, providing a predictive model for the peak clad temperature during a reflood phase. Other uncertainties are interpreted as contributors to the measurement noise of the code output and are implicitly treated in the GPM in the noise variance term, providing local uncertainty bounds for the peak clad temperature. We discuss the applicability of the foregoing method to reduce the use of conservative assumptions in best estimate plus uncertainty (BEPU and Level 1 probabilistic safety assessment (PSA success criteria definitions while dealing with a large number of uncertainties.
International Nuclear Information System (INIS)
Skorek, Tomasz; Crecy, Agnes de
2013-01-01
PREMIUM (Post BEMUSE Reflood Models Input Uncertainty Methods) is an activity launched with the aim to push forward the methods of quantification of physical models uncertainties in thermal-hydraulic codes. It is endorsed by OECD/NEA/CSNI/WGAMA. The benchmark PREMIUM is addressed to all who applies uncertainty evaluation methods based on input uncertainties quantification and propagation. The benchmark is based on a selected case of uncertainty analysis application to the simulation of quench front propagation in an experimental test facility. Application to an experiment enables evaluation and confirmation of the quantified probability distribution functions on the basis of experimental data. The scope of the benchmark comprises a review of the existing methods, selection of potentially important uncertain input parameters, preliminary quantification of the ranges and distributions of the identified parameters, evaluation of the probability density function using experimental results of tests performed on FEBA test facility and confirmation/validation of the performed quantification on the basis of blind calculation of Reflood 2-D PERICLES experiment. (authors)
International Nuclear Information System (INIS)
Nelson, R.A.; Unal, C.
1991-01-01
In this paper, a phenomenological model of the thermal hydraulics of convective boiling in the post-critical-heat-flux (post-CHF) regime is developed and discussed. The model was implemented in the TRAC-PF1/MOD2 computer code (an advanced best-estimate computer program written for the analysis of pressurized water reactor systems). The model was built around the determination of flow regimes downstream of the quench front. The regimes were determined from the flow-regime map suggested by Ishii and his coworkers. Heat transfer in the transition boiling region was formulated as a position-dependent model. The propagation of the CHF point was strongly dependent on the length of the transition boiling region. Wall-to-fluid film boiling heat transfer was considered to consist of two components: first, a wall-to-vapor convective heat-transfer portion and, second, a wall-to-liquid heat transfer representing near-wall effects. Each contribution was considered separately in each of the inverted annular flow (IAF) regimes. The interfacial heat transfer was also formulated as flow-regime dependent. The interfacial drag coefficient model upstream of the CHF point was considered to be similar to flow through a roughened pipe. A free-stream contribution was calculated using Ishii's bubbly flow model for either fully developed subcooled or saturated nucleate boiling. For the drag in the smooth IAF region, a simple smooth-tube correlation for the interfacial friction factor was used. The drag coefficient for the rough-wavy IAF was formulated in the same way as for the smooth IAF model except that the roughness parameter was assumed to be proportional to liquid droplet diameter entrained from the wavy interface. The drag coefficient in the highly dispersed flow regime considered the combined effects of the liquid droplets within the channel and a liquid film on wet unheated walls. 431 refs., 6 figs., 4 tabs
Cacace, Mauro; Jacquey, Antoine B.
2017-09-01
Theory and numerical implementation describing groundwater flow and the transport of heat and solute mass in fully saturated fractured rocks with elasto-plastic mechanical feedbacks are developed. In our formulation, fractures are considered as being of lower dimension than the hosting deformable porous rock and we consider their hydraulic and mechanical apertures as scaling parameters to ensure continuous exchange of fluid mass and energy within the fracture-solid matrix system. The coupled system of equations is implemented in a new simulator code that makes use of a Galerkin finite-element technique. The code builds on a flexible, object-oriented numerical framework (MOOSE, Multiphysics Object Oriented Simulation Environment) which provides an extensive scalable parallel and implicit coupling to solve for the multiphysics problem. The governing equations of groundwater flow, heat and mass transport, and rock deformation are solved in a weak sense (either by classical Newton-Raphson or by free Jacobian inexact Newton-Krylow schemes) on an underlying unstructured mesh. Nonlinear feedbacks among the active processes are enforced by considering evolving fluid and rock properties depending on the thermo-hydro-mechanical state of the system and the local structure, i.e. degree of connectivity, of the fracture system. A suite of applications is presented to illustrate the flexibility and capability of the new simulator to address problems of increasing complexity and occurring at different spatial (from centimetres to tens of kilometres) and temporal scales (from minutes to hundreds of years).
International Nuclear Information System (INIS)
Hoeld, A.
2004-01-01
The thermal-hydraulic theory of single- and especially two-phase flow systems used for plant transient analysis is dominated by separate-phase models. The corresponding mostly very comprehensive codes (TRAC, RELAP, CATHARE, ATHLET etc.) are looked as to be by far more efficient than a 3 eq. mixture-fluid approach and code also if they show deficiencies in describing flow situations within inner loops as for example the distribution into parallel channels (and thus the simulation of 3D thermal-hydraulic phenomena). This may be justified if comparing them to the very simple 'homogeneous equilibrium models (HEM)', but not if looking to the more refined non-homogeneous 'separate-region' mixture-fluid approaches based on appropriate drift-flux correlation packages which can have, on the contrary, enormous advantages with respect to such separate-phase models. Especially if comparing the basic (and starting) eqs. of such theoretical models of both types the differences are remarkable. Single-phase and mixture-fluid models start from genuine conservation eqs. for mass, energy and momentum, demanding (in case of two-phase flow) additionally an adequate drift flux package (in order to get a relation for a fourth independent variable), a heat transfer coefficients package (over the whole range of the possible fields of application) and correlations for single- and two-phase friction. The other types of models are looking at each phase separately with corresponding 'field' eqs. for each phase, connected by exchange (=closure) terms which substitute the classical constitutive packages for drift, heat transfer and friction. That the drift-flux, heat transfer into a coolant channel and friction along a wall and between the phases is described better by a separate-phase approach is at least doubtful. The corresponding mixture-fluid correlations are based over a wide range on a treasure of experience and measurements, their pseudo-stationary treatment can (due to their small time
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.
Single-phase sodium pump model for LMFBR thermal-hydraulic analysis
International Nuclear Information System (INIS)
Madni, I.K.; Cazzoli, E.G.; Agrawal, A.K.
1979-01-01
A single-phase, homologous pump model has been developed for simulation of safety-related transients in LMFBR systems. Pump characteristics are modeled by homologous head and torque relations encompassing all regimes of operation. These relations were derived from independent model test results with a centrifugal pump of specific speed equal to 35 (SI units) or 1800 (gpm units), and are used to analyze the steady-state and transient behavior of sodium pumps in a number of LMFBR plants. Characteristic coefficients for the polynomials in all operational regimes are provided in a tabular form. The speed and flow dependence of head is included through solutions of the impeller and coolant dynamic equations. Results show the model to yield excellent agreement with experimental data in sodium for the FFTF prototype pump, and with vendor calculations for the CRBR pump. A sample pipe rupture calculation is also performed to demonstrate the necessity for modeling the complete pump characteristics
Application of the thermal-hydraulic codes in VVER-440 steam generators modelling
Energy Technology Data Exchange (ETDEWEB)
Matejovic, P.; Vranca, L.; Vaclav, E. [Nuclear Power Plant Research Inst. VUJE (Slovakia)
1995-12-31
Performances with the CATHARE2 V1.3U and RELAP5/MOD3.0 application to the VVER-440 SG modelling during normal conditions and during transient with secondary water lowering are described. Similar recirculation model was chosen for both codes. In the CATHARE calculation, no special measures were taken with the aim to optimize artificially flow rate distribution coefficients for the junction between SG riser and steam dome. Contrary to RELAP code, the CATHARE code is able to predict reasonable the secondary swell level in nominal conditions. Both codes are able to model properly natural phase separation on the SG water level. 6 refs.
International Nuclear Information System (INIS)
Hung, T.C.; Dhir, V.K.; Chang, J.C.; Wang, S.K.
2011-01-01
Research highlights: → The COOLOD/N2 and PARET/ANL codes were used for a steady-state thermal-hydraulic and safety analysis of the 2 MW TRIGA MARK II reactor located at the Nuclear Studies Center of Maamora (CENM), Morocco. → The main objective of this study is to ensure the safety margins of different safety related parameters by steady-state calculations at full power level (2 MW). → The most important conclusion is that all obtained values of DNBR, fuel center and surface temperature, cladding surface temperature and coolant temperature across the hottest channel are largely far to compromise safety of the reactor. - Abstract: In this study, a pool-typed design similar to sodium-cooled fast reactor (SFR) of the fourth generation reactors has been modeled using CFD simulations to investigate the characteristics of a passive mechanism of Shutdown Heat Removal System (SHRS). The main aim is to refine the reactor pool design in terms of temperature safety margin of the sodium pool. Thus, an appropriate protection mechanism is maintained in order to ensure the safety and integrity of the reactor system during a shutdown mode without using any active heat removal system. The impacts on the pool temperature are evaluated based on the following considerations: (1) the aspect ratio of pool diameter to depth, (2) the values of thermal emissivity of the surface materials of reactor and guard vessels, and (3) innerpool liner and core periphery structures. The computational results show that an optimal pool design in geometry can reduce the maximum pool temperature down to ∼551 o C which is substantially lower than ∼627 o C as calculated for the reference case. It is also concluded that the passive Reactor Air Cooling System (RACS) is effective in removing decay heat after shutdown. Furthermore, thermal radiation from the surface of the reactor vessel is found to be important; and thus, the selection of the vessel surface materials with a high emissivity would be a
Using finite mixture models in thermal-hydraulics system code uncertainty analysis
Energy Technology Data Exchange (ETDEWEB)
Carlos, S., E-mail: scarlos@iqn.upv.es [Department d’Enginyeria Química i Nuclear, Universitat Politècnica de València, Camí de Vera s.n, 46022 València (Spain); Sánchez, A. [Department d’Estadística Aplicada i Qualitat, Universitat Politècnica de València, Camí de Vera s.n, 46022 València (Spain); Ginestar, D. [Department de Matemàtica Aplicada, Universitat Politècnica de València, Camí de Vera s.n, 46022 València (Spain); Martorell, S. [Department d’Enginyeria Química i Nuclear, Universitat Politècnica de València, Camí de Vera s.n, 46022 València (Spain)
2013-09-15
Highlights: • Best estimate codes simulation needs uncertainty quantification. • The output variables can present multimodal probability distributions. • The analysis of multimodal distribution is performed using finite mixture models. • Two methods to reconstruct output variable probability distribution are used. -- Abstract: Nuclear Power Plant safety analysis is mainly based on the use of best estimate (BE) codes that predict the plant behavior under normal or accidental conditions. As the BE codes introduce uncertainties due to uncertainty in input parameters and modeling, it is necessary to perform uncertainty assessment (UA), and eventually sensitivity analysis (SA), of the results obtained. These analyses are part of the appropriate treatment of uncertainties imposed by current regulation based on the adoption of the best estimate plus uncertainty (BEPU) approach. The most popular approach for uncertainty assessment, based on Wilks’ method, obtains a tolerance/confidence interval, but it does not completely characterize the output variable behavior, which is required for an extended UA and SA. However, the development of standard UA and SA impose high computational cost due to the large number of simulations needed. In order to obtain more information about the output variable and, at the same time, to keep computational cost as low as possible, there has been a recent shift toward developing metamodels (model of model), or surrogate models, that approximate or emulate complex computer codes. In this way, there exist different techniques to reconstruct the probability distribution using the information provided by a sample of values as, for example, the finite mixture models. In this paper, the Expectation Maximization and the k-means algorithms are used to obtain a finite mixture model that reconstructs the output variable probability distribution from data obtained with RELAP-5 simulations. Both methodologies have been applied to a separated
Application of Subcooled Boiling Model to Thermal-hydraulic Analysis inside a CANDU-6 Fuel Channel
Energy Technology Data Exchange (ETDEWEB)
Kim, Man Woong; Lee, Sang Kyu; Kim, Hyun Koon [Korea Institute of Nuclear Safety, Taejon (Korea, Republic of); Yu, Jin Bok; Yi, Sung Chul [Hangyang Univ., Seoul (Korea, Republic of)
2006-07-01
Forced convection nucleate boiling is encountered in heat exchangers during normal and non-nominal modes of operation in pressurized water or boiling water reactors (PWRs or BWRs). If the wall temperature of the piping is higher than the saturation temperature of the nearby liquid, nucleate boiling occurs. In this regime, bubbles are formed at the wall. Their growth is promoted by the wall superheat (the difference between the wall and saturation temperatures), and they depart from the wall as a result of gravitational and liquid inertia forces. If the bulk liquid is subcooled, condensation at the bubble-liquid interface takes place and the bubble may collapse. This convection nucleate boiling is called as a subcooled nucleate boiling. As for the fuel channel of a CANDU 6 reactor, forced convection nucleate boiling models for flows along fuel elements enclosed inside typical CANDU-6 fuel channel has encountered difficulties due to the modeling of local effects along the horizontal channel. Therefore, the subcooled nucleate boiling has been modeled through temperature driven boiling heat and mass transfer, using a model developed at Rensselaer Polytechnic Institute. Therefore, the objectives of this study are: (i) to investigate a proposed sub-cooled boiling model developed at Rensselaer Polytechnic Institute and (ii) to apply against a experiment and (iii) to predict local distributions of flow fields for the actual fuel channel geometries of CANDU-6 reactors. The numerical implementation is conducted using by the FLUENT 6.2 CFD computer code. The RPI model has been implemented in FLUENT 6.2 via user-defined functions (UDFs) in conjunction with the Eulerian multiphase model.
Energy Technology Data Exchange (ETDEWEB)
Dunn, F.E.
1995-12-31
Detailed steady-state and transient coolant temperatures and flow rates from an operating reactor have been used to validate the multiple pin model in the SASSYS-1 liquid metal reactor systems analysis code. This multiple pin capability can be used for explicit calculations of axial and lateral temperature distributions within individual subassemblies. Thermocouples at a number of axial locations and in a number of different coolant sub-channels m the XXO9 instrumented subassembly in the EBR-II reactor provided temperature data from the Shutdown Heat Removal Test (SHRT) series. Flow meter data for XXO9 and for the overall system are also available from these tests. Results of consistent SASSYS-1 multiple pin analyses for both the SHRT-45 loss-of-flow-without-scram-test and the S14RT-17 protected loss-of-flow test agree well with the experimental data, providing validation of the SASSYS-1 code over a wide range of conditions.
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
International Nuclear Information System (INIS)
Laureau, Axel
2015-01-01
In this PhD thesis, we describe the development of innovative neutronic models for their coupling with thermal hydraulics such that they combine precision and reasonable computational times. One of the main cases where this method is applied is the Molten Salt Fast Reactor (MSFR) whose combines a fast neutron spectrum with a thorium cycle. In this fourth generation reactor, the motion of the delayed neutron precursors and the associated phenomena have to be taken into account due to the liquid fuel circulation. The starting point for these developments was the preliminary design of this type of system where a dedicated multi-physical representation was needed to study the reactor performance in steady and transient conditions. As a first step, a stationary coupling was developed. A neutronic model based on a stochastic approach was associated to a CFD (Computational Fluid Dynamics) code to solve the Navier Stokes equations for turbulent flows and the transport of the delayed neutron precursors. The impact of this precursor motion is taken into account by reconstructing the prompt shower that they generate. This approach, called by shower, views the critical reactor as a prompt subcritical reactor that amplifies a source of delayed neutrons. A second step consisted in developing a neutronic model based on a time dependent version of the fission matrices (Transient Fission Matrix or TFM) so as to enable reactor transient studies. With the TFM model, an initial computation of the matrices with a stochastic code (MCNP, SERPENT) allows the characterization of the global spatial and time dependent neutronic response of the reactor with a precision close to that of a Monte Carlo calculation. The information thus obtained is then used to calculate transients, while retaining the advantage of reduced computational time. The TFM model, which can be used for various system concepts, also allows the evaluation of effective kinetic parameters such as the effective fraction of
Modelling of thermal hydraulics in a KAROLINA calorimeter for its calibration methodology validation
Directory of Open Access Journals (Sweden)
Luks Aleksandra
2016-12-01
Full Text Available Results of numerical calculations of heat exchange in a nuclear heating detector for nuclear reactors are presented in this paper. The gamma radiation is generated in nuclear reactor during fission and radiative capture reactions as well as radioactive decay of its products. A single-cell calorimeter has been designed for application in the MARIA research reactor in the National Centre for Nuclear Research (NCBJ in Świerk near Warsaw, Poland, and can also be used in the Jules Horowitz Reactor (JHR, which is under construction in the research centre in Cadarache, France. It consists of a cylindrical sample, which is surrounded by a gas layer, contained in a cylindrical housing. Additional calculations had to be performed before its insertion into the reactor. Within this analysis, modern computational fluid dynamics (CFD methods have been used for assessing important parameters, for example, mean surface temperature, mean volume temperature, and maximum sample (calorimeter core temperature. Results of an experiment performed at a dedicated out-of-pile calibration bench and results of numerical modelling validation are also included in this paper.
International Nuclear Information System (INIS)
Al-Kheliewi, A.S.; Klein, A.C.
1994-01-01
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. For 1/2, 1/3, and 1/4 pump failures, the fuel temperatures never exceed the fuel melting point even if the reactor is not shut down
Thermal-hydraulic analysis of nuclear reactors
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 ...
Energy Technology Data Exchange (ETDEWEB)
Nishimura, M.; Kamide, H.; Ohshima, H. [Power Reactor and Nuclear Fuel Development Corp., Oarai, Ibaraki (Japan). Oarai Engineering Center
1996-10-01
Temperature distributions in fuel subassemblies of fast reactors interactively affect heat transfer from center to outer region of the core (inter-subassembly heat transfer) and cooling capability of an inter-wrapper flow, as well as maximum cladding temperature. The prediction of temperature distribution in the sub-assembly is, therefore one of the important issues for the reactor safety assessment. To treat the complex phenomena in the core, a multi-dimensional thermal hydraulic analysis is the most promising method. From the studies on the multi-dimensional thermal hydraulic modeling for the fuel sub-assemblies, the modeling have been recommended through the analysis of sodium experiments using driver subassembly test rig PLANDTL-DHX and blanket subassembly test rig CCTL-CFR. Computations of steady states experiments in the test rigs using the above modeling showed quite good agreement to the experimental data. In the present study, the use of this modeling was extended to transient analyses, and its applicability was examined. Firstly, non-dimensional parameters used to determine the mixing factors were modified from the ones based on bundle-averaged values to the ones by local values. Secondly, a new threshold function was derived and introduced to cut off the mixing factor of thermal plumes under inertia force dominant conditions. In the results of this validation, the accuracy was comparable between the modeling and the experimental instrumentation. Thus the present modeling is capable of predicting the thermal hydraulic fields of the wire wrapped fuel pin bundles with inter-subassembly heat transfer under the conditions from rated steady operations to transitions toward natural circulation decay heat removal modes. (J.P.N.)
Neutron kinetics for system thermal-hydraulic codes
International Nuclear Information System (INIS)
Diamond, D.J.
1996-01-01
There is general agreement that for many light water reactor (LWR) calculations for licensing safety analysis, probabilistic risk assessment, operational support, and training, it is necessary to use a multidimensional neutron kinetics model coupled to a thermal-hydraulics model in order to obtain satisfactory results. This need coincides with the fact that in recent years there has been considerable research and development in this field, with modelers taking advantage of the increase in computing power that has become available. This progress has now led to coupling multidimensional neutron kinetics models to the nuclear steam supply system thermal hydraulics. This is not new since some coupled codes have always been available. What is new is that the coupling can now be done with very sophisticated models, and the planning of this coupling and the requisite modeling can take advantage of the experience of many code developers in many countries. The U.S. Nuclear Regulatory Commission and other organizations are in the process of reviewing the state of the art and making recommendations for future development. This paper summarizes one contribution to this review process: a review of the multidimensional neutron kinetics modeling, and ancillary modeling, which would be used in conjunction with system thermal-hydraulic models to perform core dynamics calculations
Hydraulic Arm Modeling via Matlab SimHydraulics
Czech Academy of Sciences Publication Activity Database
Věchet, Stanislav; Krejsa, Jiří
2009-01-01
Roč. 16, č. 4 (2009), s. 287-296 ISSN 1802-1484 Institutional research plan: CEZ:AV0Z20760514 Keywords : simulatin modeling * hydraulics * SimHydraulics Subject RIV: JD - Computer Applications, Robotics
Model for polygonal hydraulic jumps
DEFF Research Database (Denmark)
Martens, Erik Andreas; Watanabe, Shinya; Bohr, Tomas
2012-01-01
We propose a phenomenological model for the polygonal hydraulic jumps discovered by Ellegaard and co-workers [Nature (London) 392, 767 (1998); Nonlinearity 12, 1 (1999); Physica B 228, 1 (1996)], based on the known flow structure for the type-II hydraulic jumps with a "roller" (separation eddy...
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.
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)
Thermal hydraulic simulations of the Angra 2 PWR
Directory of Open Access Journals (Sweden)
González-Mantecón Javier
2015-01-01
Full Text Available Angra 2, the second Brazilian nuclear power plant, began the commercial operation in 2001. The plant is a pressurized water reactor (PWR type with electrical output of about 1350 MW. In the present work, the thermal hydraulic RELAP5-3D code was used to develop a model of this reactor. The model was performed using geometrical and material data from the Angra 2 final safety analysis report (FSAR. Simulations of the reactor behavior during steady state and loss of coolant accident were performed. Results of temperature distribution within the core, inlet and outlet coolant temperatures, coolant mass flow, and other parameters have been compared with the reference data and demonstrated to be in good agreement with each other. This study demonstrates that the developed RELAP5-3D model is capable of reproducing the thermal hydraulic behavior of the Angra 2 PWR and it can contribute to the process of the plant safety analysis.
International Nuclear Information System (INIS)
Fischer, K.; Schall, M.; Wolf, L.
1993-01-01
The present final report comprises the major results of Phase II of the CEC thermal-hydraulic benchmark exercise on Fiploc verification experiment F2 in the Battelle model containment, experimental phases 2, 3 and 4, which was organized and sponsored by the Commission of the European Communities for the purpose of furthering the understanding and analysis of long-term thermal-hydraulic phenomena inside containments during and after severe core accidents. This benchmark exercise received high European attention with eight organizations from six countries participating with eight computer codes during phase 2. Altogether 18 results from computer code runs were supplied by the participants and constitute the basis for comparisons with the experimental data contained in this publication. This reflects both the high technical interest in, as well as the complexity of, this CEC exercise. Major comparison results between computations and data are reported on all important quantities relevant for containment analyses during long-term transients. These comparisons comprise pressure, steam and air content, velocities and their directions, heat transfer coefficients and saturation ratios. Agreements and disagreements are discussed for each participating code/institution, conclusions drawn and recommendations provided. The phase 2 CEC benchmark exercise provided an up-to-date state-of-the-art status review of the thermal-hydraulic capabilities of present computer codes for containment analyses. This exercise has shown that all of the participating codes can simulate the important global features of the experiment correctly, like: temperature stratification, pressure and leakage, heat transfer to structures, relative humidity, collection of sump water. Several weaknesses of individual codes were identified, and this may help to promote their development. As a general conclusion it may be said that while there is still a wide area of necessary extensions and improvements, the
Numerical studies of thermal hydraulic characteristics through rectangular channels
Energy Technology Data Exchange (ETDEWEB)
Jo, Dae Seong; Park, Jong Hark; Chae, Hee Taek [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)
2011-05-15
A typical open pool type research reactor utilized with plate type fuel elements is modeled to analyze steady state thermal hydraulic characteristics and thermal margins for reactor safety and design purposes. The forced convection cooling by the primary cooling pump system (PCS) is a normal operation mode when the reactor is operated at a nominal power. The coolant flowing downward through the core during the forced convection cooling is only considered in this model development. The predictions by the present numerical studies are compared with experimental data taken by Sudo et al. (1984)
International Nuclear Information System (INIS)
Auder, Benjamin
2011-01-01
This research thesis has been made within the frame of a project on nuclear reactor vessel life. It deals with the use of numerical codes aimed at estimating probability densities for every input parameter in order to calculate probability margins at the output level. More precisely, it deals with codes with one-dimensional functional responses. The author studies the numerical simulation of a pressurized thermal shock on a nuclear reactor vessel, i.e. one of the possible accident types. The study of the vessel integrity relies on a thermal-hydraulic analysis and on a mechanical analysis. Algorithms are developed and proposed for each of them. Input-output data are classified using a clustering technique and a graph-based representation. A method for output dimension reduction is proposed, and a regression is applied between inputs and reduced representations. Applications are discussed in the case of modelling and sensitivity analysis for the CATHARE code (a code used at the CEA for the thermal-hydraulic analysis)
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)
Multiphase flow dynamics 5 nuclear thermal hydraulics
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...
Multiphase Flow Dynamics 5 Nuclear Thermal Hydraulics
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...
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.)
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
Superheater hydraulic model test plan
Energy Technology Data Exchange (ETDEWEB)
Gabler, M.; Oliva, R.M.
1973-10-01
The plan for conducting a hydraulic test on a full scale model of the AI Steam Generator Module design is presented. The model will incorporate all items necessary to simulate the hydraulic performance characteristics of the superheater but will utilize materials other than the 2-1/4 Cr - 1 Mo in its construction in order to minimize costs and expedite schedule. Testing will be performed in the Rockwell International Rocketdyne High Flow Test Facility which is capable of flowing up to 32,00 gpm of water at ambient temperatures. All necessary support instrumentation is also available at this facility.
Introduction on the thermal-hydraulic analysis codes for nuclear steam generator
International Nuclear Information System (INIS)
Yao Yangui; Zu Hongbiao; Yao Weida
2015-01-01
This paper describes several typical steam generator thermal-hydraulic analysis programs. Three thermal hydraulic code analysis model, principles and functions of the application are introduced. GENF is a one-dimensional code for steady state analysis, and another one-dimensional code TRANFLOW is used for transients, while ATHOS is a three-dimensional code which can be used to deal with steady as wall as transient analysis. And the code test verification and actual operating parameters verify situations are described. At last, the status and development of SG thermal-hydraulic analysis codes in China are analyzed. (authors)
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
Thermal-hydraulic considerations for particle bed reactors
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.
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
Karriem, Veronica V.
Nuclear reactor design incorporates the study and application of nuclear physics, nuclear thermal hydraulic and nuclear safety. Theoretical models and numerical methods implemented in computer programs are utilized to analyze and design nuclear reactors. The focus of this PhD study's is the development of an advanced high-fidelity multi-physics code system to perform reactor core analysis for design and safety evaluations of research TRIGA-type reactors. The fuel management and design code system TRIGSIMS was further developed to fulfill the function of a reactor design and analysis code system for the Pennsylvania State Breazeale Reactor (PSBR). TRIGSIMS, which is currently in use at the PSBR, is a fuel management tool, which incorporates the depletion code ORIGEN-S (part of SCALE system) and the Monte Carlo neutronics solver MCNP. The diffusion theory code ADMARC-H is used within TRIGSIMS to accelerate the MCNP calculations. It manages the data and fuel isotopic content and stores it for future burnup calculations. The contribution of this work is the development of an improved version of TRIGSIMS, named TRIGSIMS-TH. TRIGSIMS-TH incorporates a thermal hydraulic module based on the advanced sub-channel code COBRA-TF (CTF). CTF provides the temperature feedback needed in the multi-physics calculations as well as the thermal hydraulics modeling capability of the reactor core. The temperature feedback model is using the CTF-provided local moderator and fuel temperatures for the cross-section modeling for ADMARC-H and MCNP calculations. To perform efficient critical control rod calculations, a methodology for applying a control rod position was implemented in TRIGSIMS-TH, making this code system a modeling and design tool for future core loadings. The new TRIGSIMS-TH is a computer program that interlinks various other functional reactor analysis tools. It consists of the MCNP5, ADMARC-H, ORIGEN-S, and CTF. CTF was coupled with both MCNP and ADMARC-H to provide the
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)
Development of a model for the thermal-hydraulic characterization of the He-FUS3 loop
Energy Technology Data Exchange (ETDEWEB)
Barone, G., E-mail: gianluca.barone@for.unipi.it [University of Pisa, Department of Civil and Industrial Engineering (DICI), Pisa (Italy); Coscarelli, E.; Forgione, N.; Martelli, D. [University of Pisa, Department of Civil and Industrial Engineering (DICI), Pisa (Italy); Del Nevo, A.; Tarantino, M.; Utili, M. [ENEA UTIS-TCI, CR Brasimone, Camugnano (Italy); Ricapito, I.; Calderoni, P. [TBM& MD Project, Fusion for Energy, EU Commission, Carrer J. Pla, 2, Building B3, 08019 Barcelona (Spain)
2015-10-15
Highlights: • RELAP5-3D model of He-FUS3 facility and turbo circulator (TC). • Cold and hot facility T/H numerical analysis for the TC operanting range. • Effect of the cold by-pass opening on the facility performances. - Abstract: He-FUS3 is a helium facility designed and realized by ENEA in order to test the thermal-mechanical properties of prototypical breeding blanket module assemblies of a DEMO reactor. The actual facility has been upgraded with a high performance turbo circulator and a water heat exchanger integrating the pre-existent air cooler. In addition, a new test section located in the loop hot zone has been settled down with the objective of investigating safety relevant transient conditions of “In-TBM” LOCA scenarios. A RELAP5-3D{sup ©} model has been developed to perform a set of preliminary simulations on the new He-FUS3 layout. Both cold and hot stationary conditions have been analyzed evaluating the turbo circulator performances for a wide range of helium flow rate. Outcomes have shown that RELAP5-3D{sup ©} is an effective tool in reproducing the most significant phenomena of He-FUS3 system, providing relevant insight supporting future experimental campaigns. The post-test analysis phase will be, of course, fundamental for the qualification of a consistent numerical model.
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.
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
Process management using component thermal-hydraulic function classes
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.
Process management using component thermal-hydraulic function classes
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.
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
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)
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
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)
Energy Technology Data Exchange (ETDEWEB)
Castro, Landy Y.; Rojas, Leorlen Y.; Gamez, Abel; Rosales, Jesus; Gonzalez, Daniel; Garcia, Carlos, E-mail: lcastro@instec.cu, E-mail: leored1984@gmail.com, E-mail: agamezgmf@gmail.com, E-mail: jrosales@instec.cu, E-mail: danielgonro@gmail.com, E-mail: cgh@instec.cu [Instituto Superior de Tecnologias y Ciencias Aplicadas (InSTEC), La Habana (Cuba); Oliveira, Carlos Brayner de, E-mail: cabol@ufpe.br [Universidade Federal de Pernambuco (UFPE), Recife, PE (Brazil); Dominguez, Dany S., E-mail: dsdominguez@gmail.com [Universidade Estadual de Santa Cruz (UESC), Ilheus, BA (Brazil). Pos-Graduacao em Modelagem Computacional
2015-07-01
Chosen as one of six Generation‒IV nuclear-reactor concepts, Supercritical Water-cooled Reactors (SCWRs) are expected to have high thermal efficiencies within the range of 45 - 50% owing to the reactor's high pressures and outlet temperatures. In this reactor, the primary water enters the core under supercritical-pressure condition (25 MPa) at a temperature of 280 deg C and leaves it at a temperature of up to 510 deg C. Due to the significant changes in the physical properties of water at supercritical-pressure, the system is susceptible to local temperature, density and power oscillations. The behavior of supercritical water into the core of the SCWR, need to be sufficiently studied. Most of the methods available to predict the effects of the heat transfer phenomena within the pseudocritical region are based on empirical one-directional correlations, which do not capture the multidimensional effects and do not provide accurate results in regions such as the deteriorated heat transfer regime. In this paper, computational fluid dynamics (CFD) analysis was carried out to study the thermal-hydraulic behavior of supercritical water flows in sub-channels of a typical European High Performance Light Water Reactor (HPLWR) fuel assembly using commercial CFD code CFX-14. It was determined the steady-state equilibrium parameters and calculated the temperature and density distributions. A comparative study for different turbulence models were carried out and the obtained results are discussed. (author)
Issues and future direction of thermal-hydraulics research and development in nuclear power reactors
International Nuclear Information System (INIS)
Saha, P.; Aksan, N.; Andersen, J.; Yan, J.; Simoneau, J.P.; Leung, L.; Bertrand, F.; Aoto, K.; Kamide, H.
2013-01-01
The paper archives the proceedings of an expert panel discussion on the issues and future direction of thermal-hydraulic research and development in nuclear power reactors held at the NURETH-14 conference in Toronto, Canada, in September 2011. Thermal-hydraulic issues related to both operating and advanced reactors are presented. Advances in thermal-hydraulics have significantly improved the performance of operating reactors. Further thermal-hydraulics research and development is continuing in both experimental and computational areas for operating reactors, reactors under construction or ready for near-term deployment, and advanced Generation-IV reactors. As the computing power increases, the fine-scale multi-physics computational models, coupled with the systems analysis code, are expected to provide answers to many challenging problems in both operating and advanced reactor designs
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)
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
International Nuclear Information System (INIS)
Chenu, A.
2011-10-01
Nuclear power is nowadays in the front rank as regards helping to meet the growing worldwide energy demand while avoiding an excessive increase in greenhouse gas emissions. However, the operating nuclear power plants are mainly thermal-neutron reactors and, as such, can not be maintained on the basis of the currently identified uranium resources beyond one century at the present consumption rate. Sustainability of nuclear power thus involves closure of the fuel cycle through breeding. With a uranium-based fuel, breeding can only be achieved using a fast-neutron reactor. Sodium-cooled fast reactor (SFR) technology benefits from 400 reactor-years of accumulated experience and is thus a prime candidate for the implementation of so-called Generation-IV nuclear energy systems. In this context, the safety demonstration of SFRs remains a major Research and Development related issue. The current research aims at the development of a computational tool for the in-depth understanding of SFR core behaviour during accidental transients, particularly those including boiling of the coolant. An accurate modelling of the core physics during such transients requires the coupling between 3D neutron kinetics and thermal-hydraulics in the core, to account for the strong interactions between the two-phase coolant flow and power variations caused by the sodium void effect. The present study is specifically focused upon models for the representation of sodium two-phase flow. The extension of the thermal-hydraulics TRACE code, previously limited to the simulation of single-phase sodium flow, has been carried out through the implementation of equations-of-state and closure relations specific to sodium. The different correlations have then been implemented as options. From the validation study carried out, it has been possible to recommend a set of models which provide satisfactory results, while considering annular flow as the dominant regime up to dryout and a smooth breakdown of the
Hydraulic Redistribution: A Modeling Perspective
Daly, E.; Verma, P.; Loheide, S. P., III
2014-12-01
Roots play a key role in the soil water balance. They extract and transport water for transpiration, which usually represents the most important soil water loss in vegetated areas, and can redistribute soil water, thereby increasing transpiration rates and enhancing root nutrient uptake. We present here a two-dimensional model capable of describing two key aspects of root water uptake: root water compensation and hydraulic redistribution. Root water compensation is the ability of root systems to respond to the reduction of water uptake from areas of the soil with low soil water potential by increasing the water uptake from the roots in soil parts with higher water potential. Hydraulic redistribution is a passive transfer of water through the root system from areas of the soil with greater water potential to areas with lower water potential. Both mechanisms are driven by gradients of water potential in the soil and the roots. The inclusion of root water compensation and hydraulic redistribution in models can be achieved by describing root water uptake as a function of the difference in water potential between soil and root xylem. We use a model comprising the Richards equation for the water flow in variably saturated soils and the Darcy's equation for the water flow in the xylem. The two equations are coupled via a sink term, which is assumed to be proportional to the difference between soil and xylem water potentials. The model is applied in two case studies to describe vertical and horizontal hydraulic redistribution and the interaction between vegetation with different root depths. In the case of horizontal redistribution, the model is used to reproduce the fluxes of water across the root system of a tree subjected to uneven irrigation. This example can be extended to situations when only part of the root system has access to water, such as vegetation near creeks, trees at the edge of forests, and street trees in urban areas. The second case is inspired by recent
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
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.
International Nuclear Information System (INIS)
Liu Zhitao; Wang Binghua; Qin Benke; Xie Heng
2009-01-01
Research and development of thermal-hydraulic system codes for nuclear power plants with pressurized water reactors were analyzed on their history, status and application ranges. The important roles of best-estimate methodology, codes coupling and codes qualification were pointed out. The development models of thermal-hydraulic system codes around the world provide references to China's self-innovation. (authors)
Thermal Hydraulic Integral Effect Tests for Pressurized Water Reactors
International Nuclear Information System (INIS)
Baek, W. P.; Song, C. H.; Kim, Y. S. and others
2005-02-01
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∼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∼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
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.
Study RELAP5 Helium Properties for HTGR Thermal Hydraulic Analysis
Widodo, Surip; Rohanda, Anis; Subekti, Muhammad; Setiadipura, Topan; Bakhri, Syaiful; Sunaryo, Geni Rina
2018-02-01
The system codes non-specific for HTGR such as RELAP5 has been utilized for HTGR thermal hydraulic analysis even helium gas property is not based on KTA 3102.1. However, those RELAP5 applications for HTGR above are merely based on the assumption that RELAP5 helium properties are comparable to the helium properties in the KTA 3102.1. Therefore, the study for comparing the helium properties used in RELAP5 and the helium properties in KTA 3102.1 is required. The objective of this paper is to study the appropriateness’ helium properties in RELAP5 code for high temperature gas reactor (HTGR) thermal hydraulic analysis. There has been an inclined interest in the scientific community in the study of the application RELAP5 for HTGR thermal hydraulic analysis. The KTA 3102.1 provides the helium properties that are the most commonly use for the HTGR thermal hydraulic analysis. For this study, the RELAP5 helium properties are compared with the helium properties in KTA 3102.1. The comparison results showed that the RELAP5 helium properties are satisfactory for the HTGR thermal hydraulic analysis.
International Nuclear Information System (INIS)
Yang, C.I.; Sha, W.T.; Kasza, K.E.
1982-01-01
As a result of the uncertainties in the understanding of the influence of thermal-buoyancy effects on the flow and heat transfer in Liquid Metal Fast Breeder Reactor heat exchangers and steam generators under off-normal operating conditions, an extensive experimental program is being conducted at Argonne National Laboratory to eliminate these uncertainties. Concurrently, a parallel analytical effort is also being pursued to develop a three-dimensional transient computer code (COMMIX-IHX) to study and predict heat exchanger performance under mixed, forced, and free convection conditions. This paper presents computational results from a heat exchanger simulation and compares them with the results from a test case exhibiting strong thermal buoyancy effects. Favorable agreement between experiment and code prediction is obtained
Study on thermal-hydraulics during a PWR reflood phase
International Nuclear Information System (INIS)
Iguchi, Tadashi
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
International Nuclear Information System (INIS)
Haga, Katsuhiro; Terada, Atsuhiko; Kaminaga, Masanori; Hino, Ryutaro
2000-01-01
The flow patterns in the mock-up model of the cross-flow type mercury target were measured using the PIV (particle image velocimetry) technique under water flow conditions at room temperature. The experimental results were compared with the analytical results conducted with the thermal hydraulic analysis code, STAR-CD. As a result, it was confirmed experimentally that the cross-flow could be realized in most of the proton beam path area, where the removal of the high density heat is important, with the proper flow rate distribution along the proton beam path. The analytical result showed the good correspondence to the experimental result. Then the mercury flow field and the temperature distribution were analyzed taking the volumetric heat generation by the spallation reaction into consideration. The volumetric heat generation calculated for the proton beam energy and power of 1.5 GeV and 5 MW were assumed in the analysis. The mercury flow analysis showed that the maximum mercury temperature less than the design criteria of 300degC can be attained with the inlet mercury velocity of more than 1.1 m/s and that the recirculation flow seen in the rear of the proton beam path is considered to cause no excessive temperature rise. (author)
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)
International Nuclear Information System (INIS)
Dorning, J.J.
1993-05-01
All the objectives originally scheduled for the first year of this grant have been achieved. Furthermore, the project is ahead of schedule, in that a substantial amount of work has been completed on two significant objectives originally planned for the second year. This interim report is divided into five parts, summarizing the mathematical development, analysis and results of the project goals -- goals originally planned for the first year and completed, and those on which substantial progress has been made ahead of schedule. Effects of unheated riser sections and the downcomer recirculation loop on the stability characteristics of advanced boiling water reactor designs that incorporate risers or unheated channel extensions are summarized in Part A. Such extensions are incorporated above the heated reactor core channels to enhance buoyancy-driven natural thermal convection both during normal at-power operation and during emergency shutdown. The effects of both, the inclusion of unheated riser sections in the designs (one of the goals substantially completed ahead of schedule), and the inclusion of the recirculation loop in the models (first year goal) were generally found to be destabilizing. In general, as riser lengths were increased equilibria that previously were stable became unstable, and the systems with the taller risers evolved to density-wave limit cycle oscillations. As a building block of the second year goal -- to extend the one dimensional dynamical analysis of reactor thermal-hydraulics/neutron-kinetics to two and three dimensions -- we have carried out, ahead of schedule, the nonlinear dynamical analysis of two-phase flow in multiple parallel heated channels. Some basic aspects of bifurcation phenomena in two-phase flow and the related nonlinear dynamics of single and multiple parallel, uniformly and nonuniformly heated channels are studied
Research on integral thermal-hydraulic test facilities
International Nuclear Information System (INIS)
Liu Yusheng; Zhang Chunming; Ma Shuai; Zhang Pan
2014-01-01
Integral thermal-hydraulic test facilities, which have been necessary experimental platforms during the development of nuclear safety technology, could not only test and validate performance of new designed system, but also provide experimental data for development and validation of nuclear safety analysis codes. Typical integral thermal-hydraulic test facilities in the world are studied in this paper, of which the design parameters, system arrangements and functions are emphatically discussed. The results show that those integral thermal-hydraulic test facilities differ with each other in parameter scope and simulation function. Basing the fact that each facility has its advantages and disadvantages, it is better to take more factors into consideration in design of new facility. What is more, the design scheme could be optimized with new measurement technology and analysis software. (authors)
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
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.
Coupled neutronics - thermal-hydraulics programs for SCWRS
Energy Technology Data Exchange (ETDEWEB)
Reiss, T. [Institute of Nuclear Techniques, Budapest University of Technology and Economics, Muegyetem rkp. 9., 1111 Budapest (Hungary)
2010-07-01
The Supercritical Water Cooled Reactor (SCWR) was chosen as one of the Generation IV reactors by GIF. At the moment, a number of concepts - thermal as well as fast ones - exist. The reference parameters for a thermal SCWR have been taken from the European High Performance Light Water Reactor (HPLWR). Since the pressure is higher than the critical pressure (22.1 MPa) there is no change in the phase of the water in the core. On the other hand, due to the significant changes in the physical properties of water at supercritical pressure, the system is susceptible to local temperature, density and power oscillations. This inclination is increased by the pseudo-critical transformation of the water used as coolant. Thus, for modelling a system of this type coupled neutronics - thermal-hydraulics programs are required. Such a program system has been developed with the following main features: great modularity which allows for easy modifications, thus several SCWR concepts can be studied; detailed assembly calculations (with MCNP) and full-core analysis (with SCALE) are supported; the differential equations of xenon poisoning are implemented to study xenon oscillations. The program system was used to examine the assembly of the HPLWR, to design the assembly and the core of the Simplified Supercritical Water Cooled Reactor (SSCWR) and to model xenon oscillations in SCWRs. (authors)
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.
Energy Technology Data Exchange (ETDEWEB)
Navarro-Valenti, S.; Kim, S.H.; Georgevich, V. [Oak Ridge National Lab., TN (United States)] [and others
1995-09-01
The purpose of this paper is to describe the analysis performed to predict the thermal behavior of fuel miniplates under rapid transient heatup conditions. The possibility of explosive boiling was considered, and it was concluded that the heating rates are not large enough for explosive boiling to occur. However, transient boiling effects were pronounced. Because of the complexity of transient pool boiling and the unavailability of experimental data for the situations studied, an approximation was made that predicted the data very well within the uncertainties present. If pool boiling from the miniplates had been assumed to be steady during the heating pulse, the experimental data would have been greatly overestimated. This fact demonstrates the importance of considering the transient nature of heat transfer in the analysis of reactivity excursion accidents. An additional contribution of the present work is that it provided data on highly subcooled steady nulceate boiling from the cooling portion of the thermocouple traces.
TRETA and TIZONA fast running thermal-hydraulic codes
International Nuclear Information System (INIS)
Izquierdo, J.M.; Sanchez, M.; Hortal, F.J.; Melendez, E.; Herrero, R.; Queral, C.; Exposito, A.; Gonzalez, I.
2007-01-01
TRETA and TIZONA codes have been developed to help analysts with the understanding of probabilistic safety analysis (PSA) event trees involving complex transients taking place in nuclear power plants. The two-phase thermal-hydraulic sections of the TIZONA code convey an original model that hinges on the assumption that one of the phases is in the saturation condition. The simulation extends to virtually all plant systems, including control, protection and balance of plant. However, the codes are not tightly bound to a particular technology and can be used to perform simulations of other physical systems. The design of the codes is modular, their inputs being built as a block diagram in which each block is an instance of a more general entity called module provided with particular data. Other particular-purpose codes can be connected to TRETA and TIZONA to perform a joint simulation in which several tasks may be run in parallel if the problem so allows
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)
Application of computational fluid dynamics methods to improve thermal hydraulic code analysis
Sentell, Dennis Shannon, Jr.
A computational fluid dynamics code is used to model the primary natural circulation loop of a proposed small modular reactor for comparison to experimental data and best-estimate thermal-hydraulic code results. Recent advances in computational fluid dynamics code modeling capabilities make them attractive alternatives to the current conservative approach of coupled best-estimate thermal hydraulic codes and uncertainty evaluations. The results from a computational fluid dynamics analysis are benchmarked against the experimental test results of a 1:3 length, 1:254 volume, full pressure and full temperature scale small modular reactor during steady-state power operations and during a depressurization transient. A comparative evaluation of the experimental data, the thermal hydraulic code results and the computational fluid dynamics code results provides an opportunity to validate the best-estimate thermal hydraulic code's treatment of a natural circulation loop and provide insights into expanded use of the computational fluid dynamics code in future designs and operations. Additionally, a sensitivity analysis is conducted to determine those physical phenomena most impactful on operations of the proposed reactor's natural circulation loop. The combination of the comparative evaluation and sensitivity analysis provides the resources for increased confidence in model developments for natural circulation loops and provides for reliability improvements of the thermal hydraulic code.
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)
Thermal hydraulic analysis of flow inversion in a research reactor with downward core cooling
Energy Technology Data Exchange (ETDEWEB)
Jo, Dae Seong; Park, Jong Hark; Chae, Hee Taek [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)
2011-10-15
Research reactors with forced downward core cooling experience flow inversion if the primary cooling pump (PCP) is failed. If PCP failure occurs, the downward flow decreases into zero flow and eventually turn into upward flow by natural circulation. During flow inversion phenomenon, reactor cores may undergo the most unfavorable thermal hydraulic condition, which results in the highest coolant and fuel temperatures and lowest thermal margins. The transient thermal hydraulic analyses of loss of flow accidents (LOFA) in IAEA 10MW benchmark MTR research reactor have been widely investigated by many institutes. In this study, a transient thermal hydraulic model of flow inversion is developed and applied to IAEA 10MW benchmark MTR research reactor. The results are compared against other analyses
11. international topical meeting on nuclear reactor thermal-hydraulics (NURETH-11)
Energy Technology Data Exchange (ETDEWEB)
Lemonnier, H. (ed.)
2005-07-01
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
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)
Thermal-hydraulics Analysis of a Radioisotope-powered Mars Hopper Propulsion System
Energy Technology Data Exchange (ETDEWEB)
Robert C. O' Brien; Andrew C. Klein; William T. Taitano; Justice Gibson; Brian Myers; Steven D. Howe
2011-02-01
Thermal-hydraulics analyses results produced using a combined suite of computational design and analysis codes are presented for the preliminary design of a concept Radioisotope Thermal Rocket (RTR) propulsion system. Modeling of the transient heating and steady state temperatures of the system is presented. Simulation results for propellant blow down during impulsive operation are also presented. The results from this study validate the feasibility of a practical thermally capacitive RTR propulsion system.
Application of Sub-cooled Boiling Model to Thermal-hydraulic Analysis Inside a CANDU-6 Fuel Channel
Energy Technology Data Exchange (ETDEWEB)
Kim, Man Woong; Lee, Sang Kyu; Kim, Hyun Koon; Yoo, Kun Joong [Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of); Kang, Hyoung Chul; Yoo, Seong Yeon [Chungnam National Univ., Daejeon (Korea, Republic of)
2007-07-01
Forced convection nucleate boiling is encountered in heat exchangers during normal and non-nominal modes of operation in pressurized water or boiling water reactors (PWRs or BWRs). If the wall temperature of the piping is higher than the saturation temperature of the nearby liquid, nucleate boiling occurs. In this regime, bubbles are formed at the wall. Their growth is promoted by the wall superheat (the difference between the wall and saturation temperatures), and they depart from the wall as a result of gravitational and liquid inertia forces. If the bulk liquid is subcooled, condensation at the bubble-liquid interface takes place and the bubble may collapse. This convection nucleate boiling is called as a sub-cooled nucleate boiling. As for the fuel channel of a CANDU 6 reactor, forced convection nucleate boiling models for flows along fuel elements enclosed inside typical CANDU-6 fuel channel has encountered difficulties due to the modeling of local effects along the horizontal channel. Therefore, the subcooled nucleate boiling has been modeled through temperature driven boiling heat and mass transfer, using a model developed at Rensselaer Polytechnic Institute. The objectives of this study are: (i) to investigate a proposed sub-cooled boiling model developed at Rensselaer Polytechnic Institute and (ii) to apply against a experiment and (iii) to predict local distributions of flow fields for the actual fuel channel geometries of CANDU-6 reactors. The numerical implementation is conducted using by the FLUENT 6.2 CFD computer code.
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.
Zagrebaev, A. M.; Trifonenkov, A. V.; Trifonenkov, V. P.
2017-12-01
The effectiveness and the security of RBMK reactor operation depends on the accuracy of the control over reactor's parameters and their limitations. The processing of operational parameters archive helps to adjust different mathematical models and significantly widen their field of use. Pressure differential between common pressure header and steam separator is the sum of calculated pressure differential and friction loss on flow control valve. There is known mathematical software, which allows to adapt such model for each fuel channel using the archive. In this research it is suggested not to replace the regular mechanism with such approach, but to use the adapted mathematical model to calculate corrected values of power and flow, which were measured by regular means. Mathematical expressions and procedures for such approach are given.
Thermal hydraulic design of hydride fueled PWR cores
International Nuclear Information System (INIS)
Malen, J.A.; Todreas, N.E.; Romano, A.
2004-01-01
The neutronic characteristics of hydride fuels permit increased fuel to coolant volume ratios in the core. A parametric study was developed to determine the optimum combination of lattice pitch, rod diameter, and channel shape - further referred to as geometry - for minimizing the total cost of operating existing PWRs loaded with UZrH 1.6 fuel. Results of the thermal hydraulic and fuel performance studies are presented here, and will be integrated into an economic model in the next stage of the research. The thermal hydraulic analysis was used to determine the maximum power that can be achieved by a given geometry, subject to four constraints - MDNBR, pressure drop, fuel temperature, and coolant flow velocity. The fuel performance analysis was used to determine the maximum burnup that can be achieved by a given geometry, subject to three additional constraints - fuel internal pressure and fission gas release, clad oxidation, and clad strain. This methodology was successfully validated by comparison of the predicted power and burnup of the current PWR geometry, with the actual power and burnup of an existing PWR. Assuming a 60 psia pressure drop can be sustained through the fuel bundle, we concluded the following for square channels: the peak achievable power is 5556 MWt for a rod diameter of 6.5 mm and a P/D ratio of 1.43, and the highest power that can be achieved using the existing 12.6 mm pitch and 10.2 mm fuel rods is 4586 MWt. These power levels are significantly higher than the 3800 MWt of the reference PWR. (author)
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
International Nuclear Information System (INIS)
Chung, B. D.; Bae, S. W.; Jeong, J. J.; Lee, S. M.
2005-04-01
A new multi-dimensional component has been developed to allow for more flexible 3D capabilities in the system code, MARS. This component can be applied in the Cartesian and cylindrical coordinates. For the development of this model, the 3D convection and diffusion terms are implemented in the momentum and energy equation. And a simple Prandtl's mixing length model is applied for the turbulent viscosity. The developed multi-dimensional component was assessed against five conceptual problems with analytic solution. And some SETs are calculated and compared with experimental data. With this newly developed multi-dimensional flow module, the MARS code can realistic calculate the flow fields in pools such as those occurring in the core, steam generators and IRWST
Microscopic simulation of nuclear reactor thermal-hydraulics
International Nuclear Information System (INIS)
Ninokata, Hisashi; Aoki, Takayuki; Muramatsu, Toshiharu
2000-01-01
On three-dimensional behavior, multi-phase and multi-component thermal-hydraulics co-existing complex phase change and chemical reaction seen at severe accident, and so on under complex systems such as liquid film and drop behavior in a fuel assembly, inside of reactor pressure vessel, and so on, on aiming at further precise analysis on thermal-hydraulics phenomena in a reactor, researches on simulation focusing at microscopic mechanism of the phenomena have been carried out. And, simulations on microscopic thermal-hydraulics phenomena using numerical analysis such as super fine mesh method based on continuous body dynamic procedure, particle method, lattice Boltzmann method, Automaton method, Monte Carlo method at a view point of molecular dynamics, and so on have also become to be found out. This report was summary of previous survey actions under the research special committee on the 'microscopic simulation on reactor thermal-hydraulics' of the Japanese Society of Atomic Energy started in 1996. Here were elucidated a relationship between previous applicable technologies such as macroscopic simulation focusing a motion at local mean field and microscopic simulation one as well as survey and systematization of present condition of microscopic understanding on various physical phenomena, to scope directivity and subject on future research. (G.K.)
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
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.
Energy Technology Data Exchange (ETDEWEB)
Chen, N.C.J.; Wendel, M.W.; Yoder, G.L.
1996-02-01
In order to utilize reduced enrichment fuel, the three-element-core design has been proposed. The proposed core configuration consists of inner, middle, and outer elements, with the middle element offset axially beneath the inner and outer elements, which are axially aligned. The three-element-core RELAP5 model assumes that the reactor hardware is changed only within the core region, so that the loop piping, heat exchangers, and pumps remain as assumed for the two-element-core configuration. However, the total flow rate through the core is greater and the pressure drop across the core is less so that the primary coolant pumps and heat exchangers are operating at a different point in their performance curves. This report describes the new RELAP5 input for the core components.
Palmer, T S
2003-01-01
In this NEER project, researchers from Oregon State University have investigated the limitations of the treatment of two-phase coolants as a homogeneous mixture in neutron transport calculations. Improved methods of calculating the neutron distribution in binary stochastic mixtures have been developed over the past 10-15 years and are readily available in the transport literature. These methods are computationally more expensive than the homogeneous (or atomic mix) models, but can give much more accurate estimates of ensemble average fluxes and reaction rates provided statistical descriptions of the distributions of the two materials are know. A thorough review of the two-phase flow literature has been completed and the relevant mixture distributions have been identified. Using these distributions, we have performed Monte Carlo criticality calculations of fuel assemblies to assess the accuracy of the atomic mix approximation when compared to a resolved treatment of the two-phase coolant. To understand the ben...
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)
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
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
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)
Kuala Kemaman hydraulic model study
International Nuclear Information System (INIS)
Abdul Kadir Ishak
2005-01-01
There The problems facing the area of Kuala Kemaman are siltation and erosion at shoreline. The objectives of study are to assess the best alignment of the groyne alignment, to ascertain the most stable shoreline regime and to investigate structural measures to overcome the erosion. The scope of study are data collection, wave analysis, hydrodynamic simulation and sediment transport simulation. Numerical models MIKE 21 are used - MIKE 21 NSW, for wind-wave model, which describes the growth, decay and transformation of wind-generated waves and swell in nearshore areas. The study takes into account effects of refraction and shoaling due to varying depth, energy dissipation due to bottom friction and wave breaking, MIKE 21 HD - modelling system for 2D free-surface flow which to stimulate the hydraulics phenomena in estuaries, coastal areas and seas. Predicted tidal elevation and waves (radiation stresses) are considered into study while wind is not considered. MIKE 21 ST - the system that calculates the rates of non-cohesive (sand) sediment transport for both pure content and combined waves and current situation
Thermal-hydraulically corrected neutron cross-sections for PWR reactors
Energy Technology Data Exchange (ETDEWEB)
Santiago, Daniela M.N.; Alvim, Antonio C.M.; Silva, Fernando C., E-mail: dsantiago@con.ufrj.b, E-mail: alvim@con.ufrj.b, E-mail: fernando@con.ufrj.b [Universidade Federal do Rio de Janeiro (PEN/COPPE/UFRJ), RJ (Brazil). Coordenacao dos Programas de Pos-Graduacao de Engenharia. Programa de Engenharia Nuclear
2011-07-01
Reactor core simulation codes ought to have a thermal-hydraulics feedback module. This module calculates, among other effects, the fuel temperature thermal-hydraulics feedback, that corrects neutron cross sections. In the nodal code developed at PEN/COPPE/UFRJ, the fuel temperature is obtained through an empirical model. This work presents a physical model to calculate this temperature. A finite volume technique was used to discretize the equation for temperature distribution, while the moderator coefficient of heat transfer was calculated using ASME routines, appended to the developed code. This model allows calculation of an average radial temperature per node, since the thermal-hydraulic feedback must follow the conditions imposed by the nodal code. The results obtained were compared with the ones obtained by the empirical model. The results show that, for fuel elements near core periphery, the empirical model overestimates the temperature in the fuel, as compared to our model, which may indicate that the physical model is more appropriate to calculate the thermal-hydraulic feedback temperatures. (author)
International Nuclear Information System (INIS)
Chen, Ronghua; Tian, Maolin; Chen, Sen; Tian, Wenxi; Su, G.H.; Qiu, Suizheng
2017-01-01
Highlights: • This study constructed a full CFD model for the RPV of a PWR. • The reactor core was simplified using the porous model in CFX. • The CFX simulation result was in good agreement with the scaled test and design values. • The analysis of the SGTR accident was performed. - Abstract: Thermal-hydraulic performance in the reactor core was an essential factor in the nuclear power plant design. In this study, we analyzed the three-dimensional (3-D) thermal-hydraulic characteristic of reactor core based on porous media method. Firstly, a 3-D rector pressure vessel (RPV) model was built, including the inlet leg nozzle, downcomer, lower plenum, reactor core, upper plenum and outlet leg nozzle. Porous media model was used to simplify the reactor core and upper plenum. The commercial CFD code ANSYS CFX was employed to solve the governing equations and provide the 3-D local velocity, temperature and pressure field. After appropriate parameters and turbulent model being carefully selected, the simulation was validated against the 1:5 scaled steady-state hydraulic test. The predicted hydraulic parameters (normalized flowrate distribution and pressure drop) were in good agreement with the test results. And the predicted thermal parameters agreed well with the designed values. The validation indicated that this method was practicable in analyzing the 3-D thermal-hydraulic phenomena in the RPV. Finally, the thermal-hydraulic features in reactor core were analyzed under the condition of the Steam Generator Tube Rupture (SGTR) accident. The simulation results showed that the coolant temperature increased gradually from the center to the periphery in the reactor core in the accident. But the temperature decreased to safety level rapidly after the reactor shutdown and safety injection operation. The reactor core could keep in a safe state if appropriate safety operations were performed after accidents.
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
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
Light-water-reactor coupled neutronic and thermal-hydraulic codes
Energy Technology Data Exchange (ETDEWEB)
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.
International Nuclear Information System (INIS)
Ahn, S.H.; Aksan, N.; Austregesilo, H.; Bestion, D.; Chung, B.D.; D’Auria, F.; Emonot, P.; Gandrille, J.L.; Hanninen, M.; Horvatović, I.; Kim, K.D.; Kovtonyuk, A.; Petruzzi, A.
2015-01-01
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
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.
Data Collecting and Processing System and Hydraulic Control System of Hydraulic Support Model Test
Hong-Yu LIU; Jun-Qing LIU; Jun-Jie XI
2014-01-01
Hydraulic support is an important equipment of mechanization caving coal in modernization coal mine. Hydraulic support must pass national strength test before it quantity production and use. Hydraulic support model test based on similarity theory is a new effective hydraulic support design and test method. The test information such as displacement, stress, strain and so on can be generalized to hydraulic support prototype, which can prompt hydraulic support design. In order to satisfy the nee...
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.
Development of fuel performance and thermal hydraulic technology
Energy Technology Data Exchange (ETDEWEB)
Jung, Youn Ho; Song, K. N.; Kim, H. K. and others
2000-03-01
Space grid in LWR fuel assembly is a key structural component to support fuel rods and to enhance heat transfer from fuel rod to the coolant. Therefore, the original spacer grid has been developed. In addition, new phenomena in fuel behavior occurs at the high burnup, so that models to analyze those new phenomena were developed. Results of this project can be summarized as follows. - Seven different spacer grid candidates have been invented and submitted for domestic and US patents. Spacer grid test specimen(3x3 array and 5x5 array) were fabricated for each candidate and the mechanical tests were performed. - Basic technologies in the mechanical and thermal hydraulic behavior in the spacer grid development are studied and relevant test facilities were established - Fuel performance analysis models and programs were developed for the high burnup pellet and cladding, and fuel performance data base were compiled - Procedures of fuel characterization and in-/out of-pile tests were prepared - Conceptual design of fuel rod for integral PWR was carried out. (author)
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
Features in thermal-hydraulic design of TOMARI-3
International Nuclear Information System (INIS)
Uchida, J.; Kido, H.; Kasama, T.
2004-01-01
Hokkaido Electric Power Company. Inc is constructing 3-loop PWR named TOMARI-3 at TOMARI-mura, Hokkaido. Although basic design is almost same as existent 3-loop PWR, there are some improvements in thermal hydraulic design for a plant performance. Core bypass flow is increased to keep equipment integrity for a long time. Plant safety is confirmed by DNB analysis with this change. (author)
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.
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)
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)
Thermal-hydraulic analysis of PDS-XADS spallation target
International Nuclear Information System (INIS)
Ai Nisai; Yu Jiyang; Yang Yongwei
2012-01-01
This paper is a study of the thermal-hydraulic analysis of PDS-XADS spallation target for the large (80 MW) core concept. PDS-XADS is a small scale experimental accelerator driven sub-critical system (ADS). The analysis presented in this paper is based on lead bismuth eutectic (LBE) cooled XADS type experimental reactors, which are the de signs of the European experimental (PDS-XADS) project. The spallation target is a very important component of accelerator driven sub-critical system (ADS) because it is responsible to keep the reactor power at the required level by spallation reactions. A high rate of neutron production by spallation reaction creates the problem of decay heat cooling. LBE flow is properly cooled, but the window is not properly cooled because of the stagnation point in the pole of the window. It would be very difficult to keep the window temperature below the design limit, which is an important design limit challenge. Thermal-hydraulic analysis of LBE spallation target has been carried out by using ANSYS CFX 11.0. The detailed CFD analysis, which reveals thermal and hydraulic conditions in the window and spallation region, is carried out for different spallation target designs. Finally, the spallation target design limit is used to choose the best design. (authors)
VISTA : thermal-hydraulic integral test facility for SMART reactor
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.
2003-01-01
Preliminary performance tests were carried out using the thermal-hydraulic integral test facility, VISTA (Experimental Verification by Integral Simulation of Transients and Accidents), which has been constructed to simulate the SMART-P. The VISTA facility is an integral test facility including the primary and secondary systems as well as safety-related Passive Residual Heat Removal (PRHR) systems. Its scaled ratio with respect to the SMART-P is 1/1 in height and 1/96 in volume and heater power. Several steady states and power changing tests have been carried out to verify the overall thermal hydraulic primary and secondary characteristics in the range of 10% to 100% power operation. As for the preliminary results, the steady state conditions were found to coincide with the expected design values of the SMART-P. But the major thermal hydraulic parameters are greatly affected by the initial water level and the nitrogen pressure in the reactor's upper annular cavity. The power step/ramp changing tests are successfully carried out and the system responses are observed. The primary natural circulation operation is achieved, but advanced control logics need to be developed to reach the natural circulation mode without pressure excursion. In the PRHR transient tests, the natural circulation flow rate through the PRHR system was found to be about 10 percent in the early phases of PRHR operation
Thermal-Hydraulic Performance of Scrubbing Nozzle Used for CFVS
International Nuclear Information System (INIS)
Lee, Hyun Chul; Lee, Doo Yong; Jung, Woo Young; Lee, Jong Chan; Kim, Gyu Tae
2016-01-01
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.
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
ENHANCING HSPF MODEL CHANNEL HYDRAULIC REPRESENTATION
The Hydrological Simulation Program - FORTRAN (HSPF) is a comprehensive watershed model, which employs depth-area-volume-flow relationships known as hydraulic function table (FTABLE) to represent stream channel cross-sections and reservoirs. An accurate FTABLE determination for a...
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
Hydraulic assessment of the Buda Thermal Karst area and its vulnerability (Budapest, Hungary)
Czauner, Brigitta; Erőss, Anita; Erhardt, Ildikó; Ötvös, Viktória; Simon, Szilvia; Mádl-Szőnyi, Judit
2017-04-01
Thermal and medicinal water resources of Budapest (Hungary), the "City of Spas", are provided by the Buda Thermal Karst area. Assessment of its vulnerability requires the understanding of the discharge phenomena and thus the groundwater flow conditions in the area. Accordingly, BTK has already been the objective of several hydrogeological investigations, including numerical simulations as well, which led to conceptual models. The aim of the present study was the hydraulic evaluation of the flow systems based on the complex analysis of real, i.e. measured, archival hydraulic data of wells in order to i) get acquainted with the real flow systems, and ii) hydraulically confirm or disprove the previous conceptual models, in particular the applicability of gravity-driven regional groundwater flow concept and hydraulic continuity, separation of the natural discharge zones, and hypogenic karstification. Considering the data distribution, pressure vs. elevation profiles, tomographic fluid-potential maps, and hydraulic cross-sections were constructed for the first time in this area. As a result, gravitational flow systems and the modifying effects of aquitard units and faults were identified. Consequently, the differences in temperature, hydrochemistry, discharge distribution (one and two-components), and related cave forming processes between the Central (Rózsadomb) and Southern (Gellért Hill) natural discharge areas could be explained, as well as the hydraulic behaviour of the Northeastern Margin-fault of the Buda Hills could be determined. Regarding the on-going hypogenic karstification processes, regional upward flow conditions were confirmed along the main discharge zone of the Danube. Identification of gravity as the main fluid flow driving force, as well as the hydraulic effects of heterogeneities can significantly contribute to the recognition of the risk factors regarding the vulnerability of the Buda Thermal Karst area. The research was supported by the
Thermal-hydraulics in recirculating steam generators
International Nuclear Information System (INIS)
Carver, M.B.; Carlucci, L.N.; Inch, W.W.R.
1981-04-01
This manual describes the THIRST code and its use in computing three-dimensional two-phase flow and heat transfer in a steam generator under steady state operation. The manual is intended primarily to facilitate the application of the code to the analysis of steam generators typical of CANDU nuclear stations. Application to other steam generator designs is also discussed. Details of the assumptions used to formulate the model and to implement the numerical solution are also included
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.
A THERMAL-HYDRAULIC SYSTEM FOR THE CONVERSION AND THE STORAGE OF ENERGY
MITRAN Tudor; CHIOREANU Nicolae; ABAITANCAI Horia; RUS Alexandru
2016-01-01
The paper proposes the concept design of a thermal-hydraulic system that converts the thermal energy (from the geothermal water, from the cooling water of power equipment, from exhaust gasses, and so.) in hydrostatic energy, that is stored in a hydraulic accumulator. The hydraulic energy can be converted into electrical energy when needed.
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.
Current status of design technology on core thermal-hydraulic performance in FLWR
International Nuclear Information System (INIS)
Ohnuki, Akira; Kobayashi, Noboru
2008-01-01
R and D project to investigate thermal-hydraulic performance in tight-lattice rod bundles for Innovative Water Reactor for Flexible Fuel Cycle (FLWR) has been progressed at Japan Atomic Energy Agency in collaboration with power utilities, reactor vendors and universities. The FLWR adopts a triangular tight-lattice rod bundle with around 1mm gap width between rods and the thermal-hydraulic performance is being recognized as one of the major subjects. We have performed the R and D using large-scale test facility (37-rod bundle with full-height and full-pressure), model experiments and advanced numerical simulation technology. This paper described the master plan for the development of design technology and showed an executive summary for this project. 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 subchannel and 3-D numerical simulation one to evaluate the effects of geometry/scale on the thermal-hydraulic behaviors. (author)
Thermal-hydraulic Analysis of LOCA to Apply PSA Using MAAP and MARS codes
Energy Technology Data Exchange (ETDEWEB)
Cho, Yun Je; Kim, Tae Jin; Park, Goon Cherl [Seoul National University, Seoul (Korea, Republic of); Lim, Ho Gon [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)
2007-10-15
Thermal-hydraulic analysis in Probabilistic Safety Assessment (PSA) is performed to product basic data, which are needed to analyze accident sequence, construct system fault tree and evaluate operator error. Through the thermal-hydraulic analysis, the reactor power level, the pressure and the temperature of primary side, and the pressure, the temperature and the water level of secondary side are calculated. From these data, system success criteria for construction of event tree and the allowable outrage time for human reliability analysis are determined. Until now, system codes such as MAAP, RELAP, MELCOR, RETRAN have been widely used for thermal-hydraulic analysis in PSA. The adequacy of analysis is dependent on the type of accident and the models of codes. As a first step of 'Study on Best-Estimate Thermal-Hydraulic Analysis Methodology Applicable to Probabilistic Safety Assessment', a part of National Nuclear Technology Program of Ministry of Science and Technology, it is required to compare the result of MARS analysis with that of MAAP analysis previously performed, and to evaluate its applicability to PSA.
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.
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)
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
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)
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
2013-02-05
... Hydraulic Phenomena and Materials, Metallurgy and Reactor Fuels; Notice of Meeting The Joint ACRS Subcommittees on Thermal Hydraulic Phenomena and Materials, Metallurgy and Reactor Fuels will hold a meeting on...
Investigation on coupling characteristics of neutronics/thermal-hydraulics of PWR NPP core
International Nuclear Information System (INIS)
Zheng Yong; Peng Minjun; Xia Genglei; Liu Xinkai
2014-01-01
In this paper, an integrated neutronics/thermal-hydraulic model for the reactor of Qinshan Phase n NPP project was developed, using the RELAP5-HD as core coupled computational code. Based on the coupled model, the steady state calculation and the rod drop transient simulation were performed. The results show that the values obtained from RELAP5-HD calculation agree well with the available measured data, and the calculated accident curves can predict all major parameters trends of the transient with good accuracy. Both steady state and transient calculation results are in accordance with the theoretical analysis from the feedback aspect of coupled reactor neutronics/thermal-hydraulics, this demonstrates that a successful coupled model of Qinshan Phase n NPP core has been developed, and the established model provides a good foundation for further simulation analysis of the nuclear power plant system. (authors)
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.
Coupled fully 3D neutron kinetics thermal-hydraulic computations for DNB analysis on PWRs
International Nuclear Information System (INIS)
Pitot, Samuel; Alborghetti, Nicolas
2007-01-01
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. To perform Main Steam Line Break (MSLB) accident calculations EDF have developed its own numerical tool OSCARD based on: the thermal-hydraulic THYC code for DNB analysis, the neutron kinetics COCCINELLE code for power distribution computations, the thermal-hydraulic CATHARE code to provide boundary conditions analysis with system scale computation. With OSCARD a fully three-dimensional (3D) representation of the core is proposed in conjunction with a two-phase flow porous-body approach (THYC) and two-group diffusion equations in the axial and lateral directions with Doppler and void reactivity feedback effects (COCCINELLE). OSCARD provides EDF with an alternative and independent way of evaluating fuel performance and safety margins. In the licensed approach, the coupled 3D neutron kinetics and thermal-hydraulic part of OSCARD steady computations is used to produce 3D power distribution in the reactor core at the most penalizing moment of the transient. Then this distribution is used as an input for THYC to perform thermal-hydraulic subchannel analysis. This 3 steps approach is used with simple conservative and bounding analysis assumptions, that can not occur in reality. In a prospective approach, OSCARD enables to combine thermal-hydraulic subchannel analysis with the neutron kinetics radial average channel model using a nodalization of one quarter of fuel assembly in order to perform one step DNB analysis. (author)
Thermo-Hydraulic Modelling of Buffer and Backfill
International Nuclear Information System (INIS)
Pintado, X.; Rautioaho, E.
2013-09-01
The temporal evolution of saturation, liquid pressure and temperature in the components of the engineered barrier system was studied using numerical methods. A set of laboratory tests was conducted to calibrate the parameters employed in the models. The modelling consisted of thermal, hydraulic and thermo-hydraulic analysis in which the significant thermo-hydraulic processes, parameters and features were identified. CODE B RIGHT was used for the finite element modelling and supplementary calculations were conducted with analytical methods. The main objective in this report is to improve understanding of the thermo-hydraulic processes and material properties that affect buffer behaviour in the Olkiluoto repository and to determine the parametric requirements of models for the accurate prediction of this behaviour. The analyses consisted of evaluating the influence of initial canister temperature and gaps in the buffer, and the role played by fractures and the rock mass located between fractures in supplying water for buffer and backfill saturation. In the thermo-hydraulic analysis, the primary processes examined were the effects of buffer drying near the canister on temperature evolution and the manner in which heat flow affects the buffer saturation process. Uncertainties in parameters and variations in the boundary conditions, modelling geometry and thermo-hydraulic phenomena were assessed with a sensitivity analysis. The material parameters, constitutive models, and assumptions made were carefully selected for all the modelling cases. The reference parameters selected for the simulations were compared and evaluated against laboratory measurements. The modelling results highlight the importance of understanding groundwater flow through the rock mass and from fractures in the rock in order to achieve reliable predictions regarding buffer saturation, since saturation times could range from a few years to tens of thousands of years depending on the hydrogeological
Functional Analysis and Thermal-Hydraulics Program Plan
International Nuclear Information System (INIS)
Paik, I.K.; Lord, R.; Parks, B.
1992-01-01
The purpose of this document is to set forth the Program Plan for the Functional Analysis and Thermal-Hydraulics (FA ampersand TH) Program (herein after referred to as the open-quotes Programclose quotes) for the 5 year period covering fiscal years 1992 thru 1996. Specifically, the actions planned by the Safety Analysis Group (SAG) of the Reactor Safety Research Section within SRTC will be identified, defined, and a schedule and resource projection presented. This document will be used by the Reactor Safety Research Section management as the baseline definition for the Program's scope, schedule and cost. Annual budget and staffing requests will be submitted based on this approved baseline. Status reporting and progress monitoring will be performed against this approved baseline. This Program plan will be revised as needed to reflect the changes that come about due to Program redirection. The Program's primary mission is to provide further assurance that the Savannah River Site K-Reactor is designed, modified, operated and maintained in a safe, cost-effective manner through application of functional analysis methodology and continued development of thermal hydraulic support capabilities. It is envisioned that the Program will continue throughout the operating life of K-Reactor and have a permanent staff of eight: one lead and seven engineers. The Program has two primary elements; (1) functional analysis, and (2) thermal-hydraulics. Functional analysis is the first element of the formal Systems Engineering Process. Systems Engineering methodology is commonly applied in both commercial and military programs, particularly where the needs of the program involve complex interrelationships between hardware, software, personnel, and support facilities. It has been extensively used in development of military systems, and in the commercial sector in the development of designs for nuclear power reactors
Energy Technology Data Exchange (ETDEWEB)
Corradin, Michael [Univ. of Wisconsin, Madison, WI (United States). Dept. of Engineering Physics; Anderson, M. [Univ. of Wisconsin, Madison, WI (United States). Dept. of Engineering Physics; Muci, M. [Univ. of Wisconsin, Madison, WI (United States). Dept. of Engineering Physics; Hassan, Yassin [Texas A & M Univ., College Station, TX (United States); Dominguez, A. [Texas A & M Univ., College Station, TX (United States); Tokuhiro, Akira [Univ. of Idaho, Moscow, ID (United States); Hamman, K. [Univ. of Idaho, Moscow, ID (United States)
2014-10-15
This experimental study investigates the thermal hydraulic behavior and the heat removal performance for a scaled Reactor Cavity Cooling System (RCCS) with air. A quarter-scale RCCS facility was designed and built based on a full-scale General Atomics (GA) RCCS design concept for the Modular High Temperature Gas Reactor (MHTGR). The GA RCCS is a passive cooling system that draws in air to use as the cooling fluid to remove heat radiated from the reactor pressure vessel to the air-cooled riser tubes and discharged the heated air into the atmosphere. Scaling laws were used to preserve key aspects and to maintain similarity. The scaled air RCCS facility at UW-Madison is a quarter-scale reduced length experiment housing six riser ducts that represent a 9.5° sector slice of the full-scale GA air RCCS concept. Radiant heaters were used to simulate the heat radiation from the reactor pressure vessel. The maximum power that can be achieved with the radiant heaters is 40 kW with a peak heat flux of 25 kW per meter squared. The quarter-scale RCCS was run under different heat loading cases and operated successfully. Instabilities were observed in some experiments in which one of the two exhaust ducts experienced a flow reversal for a period of time. The data and analysis presented show that the RCCS has promising potential to be a decay heat removal system during an accident scenario.
International Nuclear Information System (INIS)
Corradin, Michael; Dominguez, A.; Tokuhiro, Akira; Hamman, K.
2014-01-01
This experimental study investigates the thermal hydraulic behavior and the heat removal performance for a scaled Reactor Cavity Cooling System (RCCS) with air. A quarter-scale RCCS facility was designed and built based on a full-scale General Atomics (GA) RCCS design concept for the Modular High Temperature Gas Reactor (MHTGR). The GA RCCS is a passive cooling system that draws in air to use as the cooling fluid to remove heat radiated from the reactor pressure vessel to the air-cooled riser tubes and discharged the heated air into the atmosphere. Scaling laws were used to preserve key aspects and to maintain similarity. The scaled air RCCS facility at UW-Madison is a quarter-scale reduced length experiment housing six riser ducts that represent a 9.5° sector slice of the full-scale GA air RCCS concept. Radiant heaters were used to simulate the heat radiation from the reactor pressure vessel. The maximum power that can be achieved with the radiant heaters is 40 kW with a peak heat flux of 25 kW per meter squared. The quarter-scale RCCS was run under different heat loading cases and operated successfully. Instabilities were observed in some experiments in which one of the two exhaust ducts experienced a flow reversal for a period of time. The data and analysis presented show that the RCCS has promising potential to be a decay heat removal system during an accident scenario.
Operational studies of thermal-hydraulic unit steam generators
International Nuclear Information System (INIS)
Shejnkman, A.G.; Bel'tyukov, A.I.; Vylomov, V.V.; Karpenko, A.I.; Lyzhin, A.A.; Smirnov, M.V.; Khaletskij, Eh.Eh.
1989-01-01
Results of the complex of tests, experimental studies and operational measurements of thermal-hydraulic characteristics of the PGM-200M steam generators conducted in the period of start-up and bringing-up to full capability at the BN-600 power unit are presented. Analysis of the data obtained shows that the steam generator comprising eight sections ensures heat removal at the nominal power with seven operating sections. The uniformity of coolant flow rate distribution over the steam generator sections practically does not depend on the number of operating sections
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.
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
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.
International Nuclear Information System (INIS)
Sanchez S, R.A.; Morales S, J.B.
2005-01-01
In previous versions of the SUN-RAH, the core of the reactor was simulated starting from the punctual kinetics of neutrons of the same one. Different mathematical models to implement an unidimensional simulation of the thermal hydraulics of the core of the reactor to have a tool but exact were proposed. Of among the different ones modeling, those of Heat Transfer of n nodes and that of a differential equation of heat transfer were chosen. Both present the mathematical derivation of the equations of radial transfer of the heat generated in a bar of fuel, numeric routines for the calculation of the typical thermodynamic properties, calculation of the stationary state and dynamic response of some premature operational occurrences. It was carried out the comparison among both proposals with the purpose of being implemented in the SUN-RAH. This simulator includes all the main components of the thermodynamic cycle, with that the implementation of the one dimension models of the core, will be transform it into a tool but reliable. To make congruent the multidimensional kinetics of neutrons is necessary to have a model of heat transfer congruent with her for that here an analysis is made of that model of transfer it can be used in a great number of neutronic nodes. (Author)
CFD Prediction of Thermal-Hydraulic Characteristics Inside a Containment of a CANDU-6 Reactor
Energy Technology Data Exchange (ETDEWEB)
Kim, Man Woong; Choi, Yong Seog; Kim, Hyun Koon [Korea Institute of Nuclear Safety, Taejon (Korea, Republic of); Nho, Hyun Suk [Advanced Technology Engineering Service, Seoul (Korea, Republic of)
2005-07-01
During the course of accident in a CANDU reactor, large amounts of flow mass, enthalpy and hydrogen could be generated and released into the containment. The integrity of the containment could be challenged by certain hydrogen and hydraulic dynamic load. Therefore, a detailed knowledge of containment thermal-hydraulics is necessary to predict the local distribution of hydrogen, steam and air inside the containment. Considerable international efforts have been undertaken to better understand the associated phenomena by conducting a large number of experiments such as ISP 23, ISP29, ISP 35, etc. and then subjecting the test results to extensive analytical assessment. Moreover, the recent progress in CFD methods has provided opportunities to predict the pressure, temperature and hydrogen distribution under accident conditions reflecting the actual geometry. This capability will lead to a significant improvement of the reliability of accident containment models for full-plant analysis. In this study, the CFD prediction of the thermal-hydraulic characteristics inside a containment of a CANDU-6 reactor is carried out. A MSLB (Main Steam Line Break) scenario was selected to analyze the thermal-hydraulic behavior. The source of vaporized water mass flow rate and enthalpy released for the FLUENT CFD analysis is obtained from a RELAP/CANDU calculation. A comparison between FLUENT CFD and PRESCON results is also performed.
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)
Modeling and stability of electro-hydraulic servo of hydraulic excavator
Jia, Wenhua; Yin, Chenbo; Li, Guo; Sun, Menghui
2017-11-01
The condition of the hydraulic excavator is complicated and the working environment is bad. The safety and stability of the control system is influenced by the external factors. This paper selects hydraulic excavator electro-hydraulic servo system as the research object. A mathematical model and simulation model using AMESIM of servo system is established. Then the pressure and flow characteristics are analyzed. The design and optimization of electro-hydraulic servo system and its application in engineering machinery is provided.
The important role of thermal hydraulics in 50 years of nuclear power applications
International Nuclear Information System (INIS)
Levy, Salomon
1994-01-01
Thermal hydraulics have played a very important role in the safety of nuclear power plants. The purpose of this paper is to provide a summary of our knowledge of thermal hydraulics in the 1950s and of the progress made up to the early 1990s. Important ''lessons learned'' over the past 50 years, and future potential issues in nuclear thermal hydraulics, are discussed. ((orig.))
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)
Hydraulic modelling of the CARA Fuel element
International Nuclear Information System (INIS)
Brasnarof, Daniel O.; Juanico, Luis; Giorgi, M.; Ghiselli, Alberto M.; Zampach, Ruben; Fiori, Jose M.; Yedros, Pablo A.
2004-01-01
The CARA fuel element is been developing by the National Atomic Energy Commission for both Argentinean PHWRs. In order to keep the hydraulic restriction in their fuel channels, one of CARA's goals is to keep its similarity with both present fuel elements. In this paper is presented pressure drop test performed at a low-pressure facility (Reynolds numbers between 5x10 4 and 1,5x10 5 ) and rational base models for their spacer grid and rod assembly. Using these models, we could estimate the CARA hydraulic performance in reactor conditions that have shown to be satisfactory. (author) [es
Thermal-hydraulic codes validation for safety analysis of NPPs with RBMK
International Nuclear Information System (INIS)
Brus, N.A.; Ioussoupov, O.E.
2001-01-01
This work is devoted to validation of western thermal-hydraulic codes (RELAP5/MOD3.2 and ATHLET 1.1 Cycle C) in application to Russia designed light water reactors of RBMK type. 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 in comparison of calculation results obtained with the thermal-hydraulics codes with the experimental data obtained earlier with the thermal-hydraulics test facilities. (authors)
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.)
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.
Multi scale thermal hydraulic analysis of PWRs using the CUPID code
Energy Technology Data Exchange (ETDEWEB)
Yoon, Han Young; Cho, Hyoung Kyu; Lee, Jae Ryong; Park, Ik Kyu [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Jeong, Jae Jun [Pusan National University, Busan (Korea, Republic of)
2012-12-15
KAERI has developed a two-phase CFD code, CUPID, for a refined calculation of transient two-phase flows related to nuclear reactor thermal hydraulics, and its numerical models have been verified in previous studies. In this paper, the CUPID code is validated against experiments on the downcomer boiling and moderator flow in a Calandria vessel. Physical models relevant to the validation are discussed. Thereafter, multi-scale thermal hydraulic analyses using the CUPID code are introduced. At first, a component-scale calculation for the passive condensate cooling tank (PCCT) of the PASCAL experiment is linked to the CFD-scale calculation for LOCA boiling heat transfer outside the heat exchanger tube. Next, the Rossendorf coolant mixing (ROCOM) test is analyzed by using the CUPID code, which is implicitly coupled with a system scale code, MARS.
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 s...... between -45.5 % to 86.4 % were reported for the f-factor. The thermal part of the model was validated with good confidence, while the frictional part of the model was validated with a smaller degree of certainty....
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
Reactor thermal-hydraulic FY 1986 status report for the multimegawatt Space Nuclear Power Program
International Nuclear Information System (INIS)
Krotiuk, W.J.; Antoniak, Z.I.
1986-10-01
PNL's 1986 activities can be divided into three basic areas: code assessment, correlation assessment and experimental activities. The ultimate goal of all these activities is developing computer codes and verifying their use to perform the thermal-hydraulic analysis and design of the reactor core and plenum of the various proposed concepts. To perform this task as assessment is made of existing computer codes, models, correlations, and microgravity experimental data
Thermal-hydraulic simulation and analysis of Research Reactor Cooling Systems
International Nuclear Information System (INIS)
EL Khatib, H.H.A.
2013-01-01
The objective of the present study is to formulate a model to simulate the thermal hydraulic behavior of integrated cooling system in a typical material testing reactor (MTR) under loss of ultimate heat sink, the model involves three interactively coupled sub-models for reactor core, heat exchanger and cooling tower. The developed model predicts the temperature profiles in addition it predicts inlet and outlet temperatures of the hot and cold stream as well as the heat exchangers and cooling tower. The model is validated against PARET code for steady-state operation and also verified by the reactor operational records, and then the model is used to simulate the thermal-hydraulic behavior of the reactor under a loss of ultimate heat sink. The simulation is performed for two operational regimes named regime I of (11 MW) thermal power and three operated cooling tower cells and regime II of (22 MW) thermal power and six operated cooling tower cells. In regime I, the simulation is performed for 1, 2 and 3 cooling tower failed cells while in regime II, it is performed for 1, 2, 3, 4, 5 and 6 cooling tower failed cells. The safety action is conducted by the reactor protection system (RPS) named power reduction safety action, it is triggered to decrease the reactor power by amount of 20% of the present power when the water inlet temperature to the core reaches 43 degree C and a scram (emergency shutdown) is triggered in case of the inlet temperature reaches 44 degree C. The model results are analyzed and discussed. The temperature profiles of fuel, clad and coolant are predicted during transient where its maximum values are far from thermal hydraulic limits.
Modeling of a Hydraulic Braking System
Lundin, Christopher
2015-01-01
The objective of this thesis is to derive an analytical model representing a reduced form of a mine hoist hydraulic braking system. Based primarily on fluid mechanical and mechanical physical modeling, along with a number of simplifying assumptions, the analytical model will be derived and expressed in the form of a system of differential equations including a set of static functions. The obtained model will be suitable for basic simulation and analysis of system dynamics, with the aim to cap...
Thermal Hydraulic and Structural Analysis of Liquid Metal Target System
International Nuclear Information System (INIS)
Lee, Yong Suk; Chung, Chang Hyun
2002-01-01
A subcritical transmutation reactor research is in progress for treatment of spent fuel. The subcritical transmutation reactor needs target system to produce high-energy neutrons. In target system, beam window is subject to high thermal field, because it interacts with high energy proton beam. In this study, target was designed based on thermal-hydraulic analysis, and thermal-structural analysis of window was performed. Preliminary design and mechanical analysis of liquid Pb-Bi target and 9Cr-2WVTa window were performed. Target was designed in a way to decrease window temperature. Installation of diffuse plate which has higher porosity in central zone was considered. Temperature and stress of window were analyzed varying minimum window thickness, beam power, and coolant flow rate. Thermal-bending stress was generated in window because of temperature gradient along the thickness of window. Coolant flow rate had insignificant effect on window stresses. It can be concluded that the target and window can be used in transmutation reactor operating condition (1 GeV, 6.78 mA). In this study, only static analysis has been made. But, accelerator beam trip can frequently occur in accelerator operation, so window and target container dynamic stress analysis will be needed. Furthermore, study about corrosion or irradiation characteristics of window will be needed in designing target and window. (authors)
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
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.
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
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)
An assessment of the CORCON-MOD3 code. Part 1: Thermal-hydraulic calculations
Energy Technology Data Exchange (ETDEWEB)
Strizhov, V.; Kanukova, V.; Vinogradova, T.; Askenov, E. [Russian Academy of Sciences, Moscow (Russian Federation). Inst. of Nuclear Safety; Nikulshin, V. [Russian Research Center, Moscow (Russian Federation). Kurchatov Inst.
1996-09-01
This report deals with the subject of CORCON-Mod3 code validation (thermal-hydraulic modeling capability only) based on MCCI (molten core concrete interaction) experiments conducted under different programs in the past decade. Thermal-hydraulic calculations (i.e., concrete ablation, melt temperature, melt energy, concrete temperature, and condensible and non-condensible gas generation) were performed with the code, and compared with the data from 15 experiments, conducted at different scales using both simulant (metallic and oxidic) and prototypic melt materials, using different concrete types, and with and without an overlying water pool. Sensitivity studies were performed in a few cases involving, for example, heat transfer from melt to concrete, condensed phase chemistry, etc. Further, special analysis was performed using the ACE L8 experimental data to illustrate the differences between the experimental and the reactor conditions, and to demonstrate that with proper corrections made to the code, the calculated results were in better agreement with the experimental data. Generally, in the case of dry cavity and metallic melts, CORCON-Mod3 thermal-hydraulic calculations were in good agreement with the test data. For oxidic melts in a dry cavity, uncertainties in heat transfer models played an important role for two melt configurations--a stratified geometry with segregated metal and oxide layers, and a heterogeneous mixture. Some discrepancies in the gas release data were noted in a few cases.
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.
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
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)
International Nuclear Information System (INIS)
Choi, Sun Rock; Back, Min Ho; Park, Won Seok; Kim, Sang Ji
2012-01-01
Since a fuel cladding failure is the most important parameter in a core thermal-hydraulic design, the conceptual design stage only involves fuel assemblies. However, although non-fuel assemblies such as control rod, reflector, and B4C generate a relatively smaller thermal power compared to fuel assemblies, they also require independent flow allocation to properly cool down each assembly. The thermal power in non-fuel assemblies is produced from both neutron and gamma energy, and thus the core thermal-hydraulic design including non-fuel assemblies should consider an energy redistribution by the gamma energy transport. To design non-fuel assemblies, the design-limiting parameters should be determined considering the thermal failure modes. While fuel assemblies set a limiting factor with cladding creep temperature to prevent a fission product ejection from the fuel rods, non-fuel assemblies restrict their outlet temperature to minimize thermally induced stress on the upper internal structure (UIS). This work employs a heat generation distribution reflecting both neutron and gamma transport. The whole core thermal-hydraulic design including fuel and non-fuel assemblies is then conducted using the SLTHEN (Steady-State LMR Thermal-Hydraulic Analysis Code Based on ENERGY Model) code. The other procedures follow from the previous conceptual design
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)
Thermal-Hydraulic Analysis of a NTD System during Irradiation
Energy Technology Data Exchange (ETDEWEB)
Park, Jong-Hark; Kim, Hak-Sung; Park, Sang-Jun; Kim, Heon-Il [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)
2007-07-01
HANARO has a utility for the neutron transmutation doping(NTD) of silicon(Si), which can produce a high quality semiconductor by neutron absorption of a Si crystal ingot. The Si ingot generates heat during the irradiation process. A NTD system is designed to remove heat by a natural circulation of reactor pool water through a small gap between the silicon ingot and its cylindrical irradiation can. If the heat released from a Si ingot can not be removed sufficiently by natural convection, a boiling occurs on the silicon ingot surface. The boiling disturbs the uniformity of the neutron flux, which directly affects the silicon quality. A thermal-hydraulic analysis using a CFD code and a measurement were carried out to examine the temperature distribution of Si ingot during an irradiation. In a comparison these two results agreed well with each other.
Sensitivity Study on Thermal Hydraulic Parameters of Research Reactor with Plate Type Fuel
Energy Technology Data Exchange (ETDEWEB)
Chae, Hee Taek; Park, Jong Hark; Park, Cheol [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)
2009-05-15
This paper presents the preliminary core thermal hydraulic characteristics and safety margins for various core flow rates, core pressures, core inlet temperatures and fuel channel powers for a plate type fuel core with 47 MW power. These sensitivity studies were performed to determine the design values for the thermal hydraulic parameters.
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.
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.
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)
Solution of neutronic and thermal-hydraulic problems on an engineering work station
International Nuclear Information System (INIS)
Zee, S.K.; Sills, E.D.; Turinsky, P.J.; Doster, J.M.
1986-01-01
Interest is in developing neutronic and thermal-hydraulic computer programs that execute efficiently on advanced engineering work stations. Engineering work stations are characterized by a 32-bit arithmetic processor, graphics capabilities, and networking capabilities. These attributes allow an engineer to solve substantive problems in a graphical interactive environment with shared resources available via networking. An advanced engineering work station is further characterized as having computational capability comparable to a mainframe, achieved via a parallel computer architecture obtained by both multi-central processing units (CPUs) and vector pipelines. In this paper, the authors present timing studies completed on an engineering work station, and then extrapolate performance on an advanced engineering work station using results from a supercomputer with parallel architecture. In this paper, the authors report on two codes, a neutronic code and a LWR system's thermal-hydraulic code. The neutronic code solves the two-group, two-dimensional (x-y) neutron diffusion equations using the finite difference method. The system's thermal-hydraulic codes solves the mixture drift-flux representation of the tube-stream form of the Navier-Stokes equations (four-equation model)
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
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
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/
International Nuclear Information System (INIS)
Choi, Young Joon
2004-02-01
Compact simulators for nuclear power plants can be used as cost-effective training or analysis tools; generally, they demonstrate overall responses of transients or accidents in real time or faster. In the thermal-hydraulic models of compact simulators, governing equations are simplified with reasonable assumptions and empirical correlations, and approximate solutions are obtained by using appropriate numerical schemes. Moreover, many physical control volumes in plant modeling are lumped to reduce the computing time. The simplification of equations and reduction of control volume numbers usually degrade the accuracy of solutions. A hybrid accident simulation methodology is proposed to enhance the capabilities of a compact simulator by introducing artificial neural networks. A simplified thermal-hydraulic program, playing the role of compact simulator, is designed to calculate the overall responses of transients and accidents. Two neural networks are designed and trained with the target values obtained from the analyses of detailed computer codes and trained results are combined with the simplified thermal-hydraulic program to perform the following roles: (I) compensation for inaccuracy of a simplified thermal-hydraulic program occurring from simplified governing equation and small number of physical control volumes: the auto-associative neural network (AANN), trained with the target values obtained from RELAP5/MOD3 code analyses, improves the calculated results of the simplified thermal-hydraulic program, and (II) prediction of the critical parameter usually calculated from the sophisticated computer code: the back propagation neural network (BPN), trained with the target values obtained from COBRA-IV code analyses, predicts the minimum departure from nucleate boiling ratio (DNBR) which is not calculated in simplified thermal-hydraulic program. Simulations for the several accidents are carried out to verify the applicability of the proposed methodology. The
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
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)
Energy Technology Data Exchange (ETDEWEB)
Hirano, Masashi [Japan Atomic Energy Research Institute, Ibaraki-ken (Japan)
1997-07-01
This paper describes the results of a scoping study on seismically induced resonance of nuclear-coupled thermal-hydraulic instability in BWRs, which was conducted by using TRAC-BF1 within a framework of a point kinetics model. As a result of the analysis, it is shown that a reactivity insertion could occur accompanied by in-surge of coolant into the core resulted from the excitation of the nuclear-coupled instability by the external acceleration. In order to analyze this phenomenon more in detail, it is necessary to couple a thermal-hydraulic code with a three-dimensional nuclear kinetics code.
Directory of Open Access Journals (Sweden)
Yangping Zhou
2012-01-01
Full Text Available At present, large water demand and carbon dioxide (CO2 emissions have emerged as challenges of steam injection for oil thermal recovery. This paper proposed a strategy of superheated steam injection by the high-temperature gas-cooled reactor (HTR for thermal recovery of heavy oil, which has less demand of water and emission of CO2. The paper outlines the problems of conventional steam injection and addresses the advantages of superheated steam injection by HTR from the aspects of technology, economy, and environment. A Geographic Information System (GIS embedded with a thermal hydraulic analysis function is designed and developed to analyze the strategy, which can make the analysis work more practical and credible. Thermal hydraulic analysis using this GIS is carried out by applying this strategy to a reference heavy oil field. Two kinds of injection are considered and compared: wet steam injection by conventional boilers and superheated steam injection by HTR. The heat loss, pressure drop, and possible phase transformation are calculated and analyzed when the steam flows through the pipeline and well tube and is finally injected into the oil reservoir. The result shows that the superheated steam injection from HTR is applicable and promising for thermal recovery of heavy oil reservoirs.
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
Multiphase flow models for hydraulic fracturing technology
Osiptsov, Andrei A.
2017-10-01
The technology of hydraulic fracturing of a hydrocarbon-bearing formation is based on pumping a fluid with particles into a well to create fractures in porous medium. After the end of pumping, the fractures filled with closely packed proppant particles create highly conductive channels for hydrocarbon flow from far-field reservoir to the well to surface. The design of the hydraulic fracturing treatment is carried out with a simulator. Those simulators are based on mathematical models, which need to be accurate and close to physical reality. The entire process of fracture placement and flowback/cleanup can be conventionally split into the following four stages: (i) quasi-steady state effectively single-phase suspension flow down the wellbore, (ii) particle transport in an open vertical fracture, (iii) displacement of fracturing fluid by hydrocarbons from the closed fracture filled with a random close pack of proppant particles, and, finally, (iv) highly transient gas-liquid flow in a well during cleanup. The stage (i) is relatively well described by the existing hydralics models, while the models for the other three stages of the process need revisiting and considerable improvement, which was the focus of the author’s research presented in this review paper. For stage (ii), we consider the derivation of a multi-fluid model for suspension flow in a narrow vertical hydraulic fracture at moderate Re on the scale of fracture height and length and also the migration of particles across the flow on the scale of fracture width. At the stage of fracture cleanaup (iii), a novel multi-continua model for suspension filtration is developed. To provide closure relationships for permeability of proppant packings to be used in this model, a 3D direct numerical simulation of single phase flow is carried out using the lattice-Boltzmann method. For wellbore cleanup (iv), we present a combined 1D model for highly-transient gas-liquid flow based on the combination of multi-fluid and
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
Open thermal-hydraulic research issues in western nuclear power reactors
International Nuclear Information System (INIS)
Hicken, E.F.
2004-01-01
codes in the past, basic research is needed to meet these requirements. Reliable models for computer codes as well as code validation can only be achieved with a good instrumentation in related experiments. Selected areas for further thermal-hydraulic research for nuclear reactors are presented. (orig.) [de
DEFF Research Database (Denmark)
Irizar, Victor; Andreasen, Casper Schousboe
2017-01-01
Hydraulic pitch systems provide robust and reliable control of power and speed of modern wind turbines. During emergency stops, where the pitch of the blades has to be taken to a full stop position to avoid over speed situations, hydraulic accumulators play a crucial role. Their efficiency...... and capability of providing enough energy to rotate the blades is affected by thermal processes due to the compression and decompression of the gas chamber. This paper presents an in depth study of the thermodynamical processes involved in an hydraulic accumulator during operation, and how they affect the energy...
French code system for a sodium cooled LMR inter-assembly thermal hydraulic analysis
International Nuclear Information System (INIS)
Kim, Young-Gyun; Lim, Hyun-Jin; Kim, Young-Il
2005-03-01
Sodium cooled LMR core is generally comprised of many ducted assemblies which have no flow exchanges between them. So, the required flow to each assembly corresponding to its power has to be allocated in thermal hydraulic design. Flow allocation facility, which is called orifice, is used for this purpose in an LMR core. In this context, flow grouping, assembly subchannel analysis and inter-assembly flow analysis have to be done in the LMR core thermal hydraulic design and analysis. This report describes this sodium cooled LMR core thermal hydraulic design procedure, in which are flow grouping, subchannel analysis and inter-assembly whole core analysis. And the French whole core analysis code system is described which is used for the domestic whole core thermal hydraulic analysis code system development. Firstly, sodium cooled LMR core thermal hydraulic conceptual design and analysis procedure is explained in chapter 2. Chapter 3 overviews the necessity and methodology of the whole core thermal hydraulic analysis, and the French whole core analysis system is described in chapter 4. Chapter 5 describes the domestic plan of the inter-assembly thermal hydraulic analysis system, and chapter 6 shows the conclusion and the future works
French code system for a sodium cooled LMR inter-assembly thermal hydraulic analysis
Energy Technology Data Exchange (ETDEWEB)
Kim, Young-Gyun; Lim, Hyun-Jin; Kim, Young-Il
2005-03-01
Sodium cooled LMR core is generally comprised of many ducted assemblies which have no flow exchanges between them. So, the required flow to each assembly corresponding to its power has to be allocated in thermal hydraulic design. Flow allocation facility, which is called orifice, is used for this purpose in an LMR core. In this context, flow grouping, assembly subchannel analysis and inter-assembly flow analysis have to be done in the LMR core thermal hydraulic design and analysis. This report describes this sodium cooled LMR core thermal hydraulic design procedure, in which are flow grouping, subchannel analysis and inter-assembly whole core analysis. And the French whole core analysis code system is described which is used for the domestic whole core thermal hydraulic analysis code system development. Firstly, sodium cooled LMR core thermal hydraulic conceptual design and analysis procedure is explained in chapter 2. Chapter 3 overviews the necessity and methodology of the whole core thermal hydraulic analysis, and the French whole core analysis system is described in chapter 4. Chapter 5 describes the domestic plan of the inter-assembly thermal hydraulic analysis system, and chapter 6 shows the conclusion and the future works.
APT target/blanket design and thermal hydraulics
International Nuclear Information System (INIS)
Cappiello, M.; Pitcher, E.; Pasamehmetoglu, K.
1999-01-01
The Accelerator Production of Tritium (APT) Target/Blanket (T/B) system is comprised of an assembly of tritium producing modules supported by control, heat removal, shielding and retargeting systems. The T/B assembly produces tritium using a high-energy proton beam, a tungsten/lead spallation neutron source and 3 He gas as the tritium producing feedstock. For the nominal production mode, protons are accelerated to an energy of 1030 MeV at a current of 100 mA and are directed onto the T/B assembly. The protons are expanded using a raster/expansion system to illuminate a 0.19m by 1.9m beam spot on the front face of a centrally located tungsten neutron source. A surrounding lead blanket produces additional neutrons from scattered high-energy particles. The tungsten neutron source consists of nested, Inconel-718 clad tungsten cylinders assembled in horizontal Inconel-718 tubes. Each tube contains up to 6 cylinders with annular flow channel gaps of 0.102 cm. These horizontal tubes are manifolded into larger diameter vertical inlet and outlet pipes, which provide coolant. The horizontal and vertical tubes make up a structure similar to that of rungs on a ladder. The entire tungsten neutron source consists of 11 such ladders separated into two modules, one containing five ladders and the other six. Ladders are separated by a 0.3 m void region to increase nucleon leakage. The peak thermal-hydraulic conditions in the tungsten neutron source occur in the second ladder from the front. Because tungsten neutron source design has a significant number of parallel flow channels, the limiting thermal-hydraulic parameter is the onset of significant void (OSV) rather than critical heat flux (CHF). A blanket region surrounds the tungsten neutron source. The lateral blanket region is approximately 120 cm thick and 400 cm high. Blanket material consists of lead, 3 He gas, aluminum, and light-water coolant. The blanket region is subdivided into rows based on the local power density in
Directory of Open Access Journals (Sweden)
W.Z. Chen
2017-06-01
behaviour of Boom Clay. Based on the experimental results, a hydraulic conductivity evolution model is proposed and then implemented in ABAQUS. Three-dimensional (3D numerical simulation of the admissible thermal loading for argillaceous storage (ATLAS III in situ heating test has been conducted subsequently, and the numerical results are in good agreement with field measurements.
Contribution to the study of thermal-hydraulic problems in nuclear reactors
International Nuclear Information System (INIS)
Cognet, G.
1998-01-01
In nuclear reactors, whatever the type considered, Pressurized Water Water Reactors (PWRs), Fast Breeder reactors (FBRs)..., thermal-hydraulics, the science of fluid mechanics and thermal behaviour, plays an essential role, both in nominal operating and accidental conditions. Fluid can either be the primary fluid (liquid or gas) or a very specific fluid called corium, which, in case of severe accident, could result from core and environning structure melting. The work reported here represents a 20-year contribution to thermal-hydraulic issues which could occur in FBRs and PWRs. Working on these two types of reactors, both in nominal and severe accident situations, has allowed me to compare the problems and to realize the importance of communication between research teams. The evolution in the complexity of studied problems, unavoidable in order to reduce costs and significantly improve safety, has led me from numerical modelling of single-phase flow turbulence to high temperature real melt experiments. The difficulties encountered in understanding the observed phenomena and in increasing experimental databases for computer code qualification have often entailed my participation in specific measurement device developments or adaptations, in particular non-intrusive devices generally based on optical techniques. Being concerned about the end-use of this research work, I actively participated in 'in-situ' thermalhydraulic experiments in the FBRs: Phenix and Super-Phenix, of which I appreciated their undeniable scientific contribution. In my opinion, the thermal-hydraulic questions related to severe accidents are the most complex as they are at the cross-roads of several scientific specialities. Consequently, they require a multi-disciplinary approach and a continuous see-saw motion between experimentalists and modelling teams. After a brief description of the various problems encountered, the main ones are reported. Finally, the importance for research teams to
Thermal-hydraulic analysis techniques for axisymmetric pebble bed nuclear reactor cores
International Nuclear Information System (INIS)
Stroh, K.R.
1979-03-01
The pebble bed reactor's cylindrical core volume contains a random bed of small, spherical fuel-moderator elements. These graphite spheres, containing a central region of dispersed coated-particle fissile and fertile material, are cooled by high pressure helium flowing through the connected interstitial voids. A mathematical model and numerical solution technique have been developed which allow calculation of macroscopic values of thermal-hydraulic variables in an axisymmetric pebble bed nuclear reactor core. The computer program PEBBLE is based on a mathematical model which treats the bed macroscopically as a generating, conducting porous medium. The steady-state model uses a nonlinear Forchheimer-type relation between the coolant pressure gradient and mass flux, with newly derived coefficients for the linear and quadratic resistance terms. The remaining equations in the model make use of mass continuity, and thermal energy balances for the solid and fluid phases
Energy Technology Data Exchange (ETDEWEB)
Stroh, K.R.
1979-03-01
The pebble bed reactor's cylindrical core volume contains a random bed of small, spherical fuel-moderator elements. These graphite spheres, containing a central region of dispersed coated-particle fissile and fertile material, are cooled by high pressure helium flowing through the connected interstitial voids. A mathematical model and numerical solution technique have been developed which allow calculation of macroscopic values of thermal-hydraulic variables in an axisymmetric pebble bed nuclear reactor core. The computer program PEBBLE is based on a mathematical model which treats the bed macroscopically as a generating, conducting porous medium. The steady-state model uses a nonlinear Forchheimer-type relation between the coolant pressure gradient and mass flux, with newly derived coefficients for the linear and quadratic resistance terms. The remaining equations in the model make use of mass continuity, and thermal energy balances for the solid and fluid phases.
Thermal hydraulic characteristics of a double-walled tube advanced nuclear steam generator
International Nuclear Information System (INIS)
Cho, S.M.; Seltzer, A.H.
1989-01-01
In this paper the thermal hydraulic characteristics of double-walled tube steam generator designed for sodium-cooled nuclear reactors are presented. The double-walled tube construction, along with double-barrier welds for tube-to-tubesheet joints, virtually eliminates the probability of heat transfer tube failure. Considerations are given to the use of the internal core tube, helical vane swirl generator, external protector tube, and variably perforated flow baffles to improve thermal and hydraulic performance of the steam generator. These thermal hydraulic design features with a particular reference to a 432 MW PRISM steam generator are discussed
Directory of Open Access Journals (Sweden)
M. Avramova
2013-01-01
Full Text Available The international OECD/NRC PSBT benchmark has been established to provide a test bed for assessing the capabilities of thermal-hydraulic codes and to encourage advancement in the analysis of fluid flow in rod bundles. The benchmark was based on one of the most valuable databases identified for the thermal-hydraulics modeling developed by NUPEC, Japan. The database includes void fraction and departure from nucleate boiling measurements in a representative PWR fuel assembly. On behalf of the benchmark team, PSU in collaboration with US NRC has performed supporting calculations using the PSU in-house advanced thermal-hydraulic subchannel code CTF and the US NRC system code TRACE. CTF is a version of COBRA-TF whose models have been continuously improved and validated by the RDFMG group at PSU. TRACE is a reactor systems code developed by US NRC to analyze transient and steady-state thermal-hydraulic behavior in LWRs and it has been designed to perform best-estimate analyses of LOCA, operational transients, and other accident scenarios in PWRs and BWRs. The paper presents CTF and TRACE models for the PSBT void distribution exercises. Code-to-code and code-to-data comparisons are provided along with a discussion of the void generation and void distribution models available in the two codes.
A methodology for the coupling of RAMONA-3B neutron kinetics and TRAC-BF1 thermal-hydraulics
International Nuclear Information System (INIS)
Lopez, Arsenio Procopio; Morales Sandoval, Jaime B.
2005-01-01
The initial objective of this project was to directly couple the RAMONA and TRAC codes running on different PCs. The idea was to use the best part of each one and eliminate some of their limitations and widen the applicability of these codes to simulate different BWR and system components. It was required to try to minimize the amount of changes to present code subroutines and calculation procedures. If possible, just substitute values obtained in the parallel code. Preliminary results indicated that using a CHAN component of TRAC to model thermal-hydraulic phenomena for each neutronic channel modeled in RAMONA is rather difficult. Large amounts of CPU time consumption are obtained and lots of PCs would make this solution difficult, besides considerable large transients are introduced by the differences in thermal-hydraulic results of these codes. The substitution of the thermal-hydraulics of RAMONA, by the TRAC channel calculations, is possible but simulation of a null transient on both codes must be planed and a gradual change must be controlled by an additional supervisory subroutine. An indirect coupling of these codes, it is therefore proposed, in order to eliminate most of these limitations. In this indirect coupling, a thermal-hydraulic model of the average tube in a bundle and the global channel cooling fluid dynamics is programmed for each neutronic channel while the global reactor vessel and core is modeled by TRAC with just four channels and four rings. Results are more reliable, coupling is simpler and faster simulations are possible
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)
Sensitivity theory applied to a transient thermal-hydraulics problem
International Nuclear Information System (INIS)
Weber, C.F.; Oblow, E.M.
1979-10-01
A new method for sensitivity analysis of transient nonlinear problems is developed and applied to a reactor thermal-hydraulics problem. The method resembles the differential sensitivity methods currently used in the linear problems of reactor physics, but it is applicable to nonlinear systems as well. The equations governing heat transfer and fluid flow in a fuel pin and surrounding coolant are given and used to derive a second set of equations (commonly known as the adjoint equations) used in the sensitivity analysis. Both systems contain one second-order parabolic and one first-order hyperbolic partial differential equation. Difference equations are derived to approximate both systems and the convergence properties of these discrete systems are evaluated, yielding a useful analysis of the numerical solution. The solution functions are used to derive sensitivity coefficients for any desired integral response. These sensitivity coefficients are used in a first-order perturbation theory to predict changes in a response resulting from changes in parameter values. The results of a test problem are shown, verifying that this procedure is indeed useful for a wide variety of sensitivity calculations
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.)
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
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.
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
REVIEW ON MODELING AND CONTROLLER DESIGN OF HYDRAULIC ACTUATOR SYSTEMS
Salleh, S.; Rahmat, M. F.; Othman, S. M.; Danapalasingam, K. A.
2015-01-01
The purpose of this paper is to review the literature on modeling and previous control strategies of the hydraulic actuator system proposed by most of the researchers around the world. Before comes to the main discussion, some background information related to hydraulic actuator will be presented. This review includes a short summary and conclusion for hydraulic actuator system. The repercussion of this review is for future inventions of a better and robust hydraulic actuator system.
Thermal-hydraulic characteristics in a liquid-metal fast breeder reactor hot plenum
International Nuclear Information System (INIS)
Tanaka, N.; Moriya, S.; Wada, A.
1984-01-01
The most important problem in the thermal-hydraulic designs of the pool-type fast breeder reactor is to estimate the thermal conditions affecting the vessel and/or internal structures during both steady and transient operations. The severity of these conditions in the Japanese pool-type reactor, which will be reinforced and equipped with special devices for seismic demands, is apt to be much greater than for other countries. Water tests and thermal-hydraulic analyses have been performed to study such conditions. The effects of the elevations of upper internal structure discharge and intermediate heat exchanger intakes on flow patterns, free surface disturbance, and thermal stratification in the hot plenum have been estimated. From the results of the experiments, suitable elevations could be recommended by comparing some thermal-hydraulic characteristics. The calculations agreed well with the experimental results for the steady-state flow patterns and thermal transients, with the exception of thermal stratification
ARCADIAR - A New Generation of Coupled Neutronics / Core Thermal- Hydraulics Code System at AREVA NP
International Nuclear Information System (INIS)
Curca-Tivig, Florin; Merk, Stephan; Pautz, Andreas; Thareau, Sebastien
2007-01-01
Anticipating future needs of our customers and willing to concentrate synergies and competences existing in the company for the benefit of our customers, AREVA NP decided in 2002 to develop the next generation of coupled neutronics/ core thermal-hydraulic (TH) code systems for fuel assembly and core design calculations for both, PWR and BWR applications. The global CONVERGENCE project was born: after a feasibility study of one year (2002) and a conceptual phase of another year (2003), development was started at the beginning of 2004. The present paper introduces the CONVERGENCE project, presents the main feature of the new code system ARCADIA R and concludes on customer benefits. ARCADIA R is designed to meet AREVA NP market and customers' requirements worldwide. Besides state-of-the-art physical modeling, numerical performance and industrial functionality, the ARCADIA R system is featuring state-of-the-art software engineering. The new code system will bring a series of benefits for our customers: e.g. improved accuracy for heterogeneous cores (MOX/ UOX, Gd...), better description of nuclide chains, and access to local neutronics/ thermal-hydraulics and possibly thermal-mechanical information (3D pin by pin full core modeling). ARCADIA is a registered trademark of AREVA NP. (authors)
A practical view of the insights from scaling thermal-hydraulic tests
Energy Technology Data Exchange (ETDEWEB)
Levin, A.E.; McPherson, G.D.
1995-09-01
The authors review the broad concept of scaling of thermal-hydraulic test facilities designed to acquire data for application to modeling the behavior of nuclear power plants, especially as applied to the design certification of passive advanced light water reactors. Distortions and uncertainties in the scaling process are described, and the possible impact of these effects on the test data are discussed. A practical approach to the use of data from the facilities is proposed, with emphasis on the insights to be gained from the test results rather than direct application of test results to behavior of a large plant.
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)
Study of NPP core thermal-hydraulics design of ABWR on steady state condition
International Nuclear Information System (INIS)
Isnaini, M. D.
1998-01-01
The core thermal-hydraulics calculation of ABWR on steady state condition using COBRA IV-1 code has been carried out. For simplifying the problem, the calculation was done on a fuel bundle of ABWR as a model. The calculation used several data design as input, such as the reactor power 3926 MWt, the core coolant flowrate 115.1 Mlb/hr and coolant enthalpy at core inlet 527.7 Btu/lb. From this simple calculation was hope that it could be used as an introduction to studi the thermohydraulics design of ABWR
Development of best estimate auditing code for CANDU thermal hydraulic safety analysis
Energy Technology Data Exchange (ETDEWEB)
Chung, B. D.; Lee, W. J.; Lim, H. S. [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)
2000-03-15
The main purpose of this study is to develop a thermal hydraulic auditing code for the CANDU reactor, modifying the model if existing PWR auditing tool, i.e. RELAP5/MOD3. This scope of project is a third step of the whole project, and expand the RELAP5/MOD3/CANDU version for implementation of LOCA analysis. There are three main area of model development, i.e. moody critical flow model, flow regime model of horizontal CANDU bundle, and fuel element heatup model when the stratification occurs. Newly developed version, namely RELAP5/MOD3/CANDU+ is applicable to CANDU plant analysis with keeping the function of light water reactor analysis. The limited validations of model installation were performed. Assessment of CHF model using AECL separated effect test and calculation for Wolsong 2 plant were performed also for the applicability test of the developed version.
Development of a best estimate auditing code for CANDU thermal hydraulic safety analysis
Energy Technology Data Exchange (ETDEWEB)
Chung, B.D.; Lee, W.J.; Lim, H.S. [Korea Atomic Energy Research Institute, Taejeon (Korea)
2000-03-01
The main purpose of this study is to develop a thermal hydraulic auditing code for the CANDU reactor, modifying the model of existing PWR auditing tool , i.e. RELAP5/MOD3. This scope of project is a third step of the whole project, and expand the RELAP5/MOD3/CANDU version for implementation of LOCA Analysis. There are three main area of model development, i.e. Moody critical flow model, flow regime model of horizontal CANDU bundle, and fuel element heatup model when the stratification occurs. Newly developed version, namely RELAP5/MOD3/CANDU+ is applicable to CANDU plant analysis with keeping the function of light water reactor analysis. The limited validations of model installation were performed. Assessment of CHF model using AECL separated effect test and calculation for Wolsong 2 plant were performed also for the applicability test of the developed version. 15 refs., 37 figs., 8 tabs. (Author)
Local chemical and thermal-hydraulic analysis of U-tube steam generators
International Nuclear Information System (INIS)
Lee, J.Y.; No, H.C.
1990-01-01
In order to know how pH distribution affects corrosion in a U-tube steam generator, a study of the combination of water chemistry and thermal-hydraulic conditions is suggested. A two-fluid (unequal velocity and unequal temperature) formulation is proposed to describe the convective transport of volatile species in each phase, and a spherical bubble model is developed on the basis of the penetration theory to describe the interfacial mass transfer. The thermal-hydraulic local conditions are obtained by the U-tube steam generator design analysis code FAUST which is based on the three-dimensional two-fluid model. The results of the present study are compared with dynamic equilibrium model calculations. This study shows that, in contrast with dynamic equilibrium calculations, the pH is lower in the cold-leg side than in the hot-leg side because of liquid recirculation. Just above the tube sheet, however, the lower void fraction in this region than that in the hot-leg region results in higher pH, which agrees with the prediction of the dynamic equilibrium model. (orig.)
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
International Nuclear Information System (INIS)
2001-05-01
An OECD Workshop on Advanced Thermal-Hydraulic and Neutronic Codes Applications was held from 10 to 13 April 2000, in Barcelona, Spain, sponsored by the Committee on the Safety of Nuclear Installations (CSNI) of the OECD Nuclear Energy Agency (NEA). It was organised in collaboration with the Spanish Nuclear Safety Council (CSN) and hosted by CSN and the Polytechnic University of Catalonia (UPC) in collaboration with the Spanish Electricity Association (UNESA). The objectives of the Workshop were to review the developments since the previous CSNI Workshop held in Annapolis [NEA/CSNI/ R(97)4; NUREG/CP-0159], to analyse the present status of maturity and remnant needs of thermal-hydraulic (TH) and neutronic system codes and methods, and finally to evaluate the role of these tools in the evolving regulatory environment. The Technical Sessions and Discussion Sessions covered the following topics: - Regulatory requirements for Best-Estimate (BE) code assessment; - Application of TH and neutronic codes for current safety issues; - Uncertainty analysis; - Needs for integral plant transient and accident analysis; - Simulators and fast running codes; - Advances in next generation TH and neutronic codes; - Future trends in physical modeling; - Long term plans for development of advanced codes. The focus of the Workshop was on system codes. An incursion was made, however, in the new field of applying Computational Fluid Dynamic (CFD) codes to nuclear safety analysis. As a general conclusion, the Barcelona Workshop can be considered representative of the progress towards the targets marked at Annapolis almost four years ago. The Annapolis Workshop had identified areas where further development and specific improvements were needed, among them: multi-field models, transport of interfacial area, 2D and 3D thermal-hydraulics, 3-D neutronics consistent with level of details of thermal-hydraulics. Recommendations issued at Annapolis included: developing small pilot/test codes for
International Nuclear Information System (INIS)
2001-01-01
An OECD Workshop on Advanced Thermal-Hydraulic and Neutronic Codes Applications was held from 10 to 13 April 2000, in Barcelona, Spain, sponsored by the Committee on the Safety of Nuclear Installations (CSNI) of the OECD Nuclear Energy Agency (NEA). It was organised in collaboration with the Spanish Nuclear Safety Council (CSN) and hosted by CSN and the Polytechnic University of Catalonia (UPC) in collaboration with the Spanish Electricity Association (UNESA). The objectives of the Workshop were to review the developments since the previous CSNI Workshop held in Annapolis [NEA/CSNI/ R(97)4; NUREG/CP-0159], to analyse the present status of maturity and remnant needs of thermal-hydraulic (TH) and neutronic system codes and methods, and finally to evaluate the role of these tools in the evolving regulatory environment. The Technical Sessions and Discussion Sessions covered the following topics: - Regulatory requirements for Best-Estimate (BE) code assessment; - Application of TH and neutronic codes for current safety issues; - Uncertainty analysis; - Needs for integral plant transient and accident analysis; - Simulators and fast running codes; - Advances in next generation TH and neutronic codes; - Future trends in physical modeling; - Long term plans for development of advanced codes. The focus of the Workshop was on system codes. An incursion was made, however, in the new field of applying Computational Fluid Dynamic (CFD) codes to nuclear safety analysis. As a general conclusion, the Barcelona Workshop can be considered representative of the progress towards the targets marked at Annapolis almost four years ago. The Annapolis Workshop had identified areas where further development and specific improvements were needed, among them: multi-field models, transport of interfacial area, 2D and 3D thermal-hydraulics, 3-D neutronics consistent with level of details of thermal-hydraulics. Recommendations issued at Annapolis included: developing small pilot/test codes for
Energy Technology Data Exchange (ETDEWEB)
Arai, Kenji; Ebata, Shigeo [Toshiba Corp., Yokohama (Japan)
1997-07-01
This paper summarizes the current and anticipated use of the thermal-hydraulic and neutronic codes for the BWR transient and accident analyses in Japan. The codes may be categorized into the licensing codes and the best estimate codes for the BWR transient and accident analyses. Most of the licensing codes have been originally developed by General Electric. Some codes have been updated based on the technical knowledge obtained in the thermal hydraulic study in Japan, and according to the BWR design changes. The best estimates codes have been used to support the licensing calculations and to obtain the phenomenological understanding of the thermal hydraulic phenomena during a BWR transient or accident. The best estimate codes can be also applied to a design study for a next generation BWR to which the current licensing model may not be directly applied. In order to rationalize the margin included in the current BWR design and develop a next generation reactor with appropriate design margin, it will be required to improve the accuracy of the thermal-hydraulic and neutronic model. In addition, regarding the current best estimate codes, the improvement in the user interface and the numerics will be needed.
Thermal hydraulic behavior of sub-assembly local blockage in China experiment fast reactor
International Nuclear Information System (INIS)
Yang Zhimin
2000-01-01
The geometrical parameter ratio of pitch to diameter of China Experiment Fast Reactor (CEFR) subassembly is 1,167. To address the thermal hydraulic behavior of subassembly local blockage which may be caused by deformation of cladding due to severe swelling and thermal stresses and by space swirl loosening etc., the porous numerical model and SIMPLE-P code used to solve Navier-Stokes and energy equations in porous medium was developed, and the bundle experiment with 19 pins with 24 subchannels blocked in the sodium coolant was carried on in China Institute of Atomic Energy (CIAE). The comparison of code predictions against experiments (including non-blockage and ten blockage conditions) seems well. The thermal hydraulic behavior of fuel subassembly with 61 fuel pins blockage of CEFR is calculated with SIMPLE-P code. The results indicate that the maximum temperature is 815 deg. C when the blockage area is about 37% (54 central subchannels are blocked). In this case the cladding won't be damaged and no sodium coolant boiling takes place. (author)
Development of subchannel analysis code MATRA-LMR for KALIMER subassembly thermal-hydraulics
International Nuclear Information System (INIS)
Won-Seok Kim; Young-Gyun Kim
2000-01-01
In the sodium cooled liquid metal reactors, the design limit are imposed on the maximum temperatures of claddings and fuel pins. Thus an accurate prediction of core coolant/fuel temperature distribution is essential to the LMR core thermal-hydraulic design. The detailed subchannel thermal-hydraulic analysis code MATRA-LMR (Multichannel Analyzer for Steady States and Transients in Rod Arrays for Liquid Metal Reactors) is being developed for KALIMER core design and analysis, based on COBRA-IV-i and MATRA. The major modifications and improvements implemented into MATRA-LMR are as follows: a) nonuniform axial noding capability, b) sodium properties calculation subprogram, c) sodium coolant heat transfer correlations, and d) most recent pressure drop correlations, such as Novendstern, Chiu-Rohsenow-Todreas and Cheng-Todreas. To assess the development status of this code, the benchmark calculations were performed with the ORNL 19 pin tests and EBR-II seven-assembly SLTHEN calculation results. The calculation results of MATRA-LMR for ORNL 19-pin assembly tests and EBR-II 91-pin experiments were compared to the measurements, and to SABRE4 and SLTHEN code calculation results, respectively. In this comparison, the differences are found among the three codes because of the pressure drop and the thermal mixing modellings. Finally, the major technical results of the conceptual design for the KALIMER 98.03 core have been compared with the calculations of MATRA-LMR, SABRE4 and SLTHEN codes. (author)
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
Investigation of Thermal-Hydraulic Characteristics on Aging Effect for CANDU Reactors
Energy Technology Data Exchange (ETDEWEB)
Yoo, Jun Soo; Choi, Yong Won; Park, Chang Hwan; Lee, Un Chul [Seoul National Univ., Seoul (Korea, Republic of); Kim, Man Woong; Lee, Sang Kyu; Kim, Hyun Koon [Korea Institute of Nuclear Safety, Taejon (Korea, Republic of)
2006-07-01
Considering that operating year of Wolsong Unit 1 gets close to the design life, 30 years, the aging effect due to the component degradation takes into consideration as an important safety issue. However, since the thermal hydraulic effect due to the aging did not identify clearly, the safety analysis methodology is not well established so far. Therefore, in this study, the aging effect affected thermal-hydraulic characteristics was investigated and a preliminary safety analysis methodology considering aging effect was proposed.
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.).
The analysis of thermal hydraulics in heater of steam generator secondary side
International Nuclear Information System (INIS)
Yang Zhilin; Li Shixing; Xu Ming
1996-01-01
COBRA-TF code is a multi-dimensional multi-field computer program, used for analysis of the thermal hydraulics in reactor system. According to the characteristics of steam generator in a nuclear power plant, the validation of using of COBRA-TF code to analyze the thermal hydraulics of the heater of the secondary side of steam generator has been made. The results are compared with those from ALBERTINE-2
International Nuclear Information System (INIS)
Okano, Yasushi
2003-03-01
Commercial computational fluid dynamic program is taken up to be employed for nuclear thermal-hydraulic applications due to the advantages in high-speed solution and easy-to-use operation. The principal objective of this report is evaluating the numerical simulation accuracy of the Fluent, on single-phase multi-dimensional thermal hydraulic problems. The evaluation problems are: 1) Laminar flow over a backward-facing step, 2) Turbulent flow over a backward-facing step, 3) Temperature of a inner rectangular rotating flow, 4) Thermal-driven natural convection flow in a square cavity, and 5) Turbulent flow in a cubic cavity, those were selected in supposing nuclear reactor thermal-hydraulic conditions by the technical committee of the Japan atomic energy society. The features on numerical method and accuracy of the Fluent being identified are: 1) Spatial differential schemes for convection term: 1st upwind, power-law, 2nd upwind, and Quick, upgrade the numerical accuracy in this order. Each scheme has the same accuracy as of the existing referenced numerical results. Quick scheme employs numerical stability oriented filtering so that no over- or under-shoots are observed. Yet, 2nd central differential scheme -used in large eddy simulation (LES)- leads numerical instability (i.e. temporal oscillation in pressure, and spatial wavering in velocity) typically when we deal with in low-resolution domains. 2) Turbulent models: (Standard, RNG, Realizable) k-ε, (Standard, SST) k-ω, and, (Standard, Quadratic) RST, necessitate to involve non-equilibrium wall function to take numerical accuracy and stability. The Fluent evaluations on re-attaching points and velocity distributions show nearly the same as -and on several counts more accurate than- those of the existing reference results. The LES turbulent model can be used only for 3-D simulations. 3) The evaluations of thermal-driven natural convection flow, which is one of the heat transfer and fluidics coupling problem, show
Analytical modeling of core hydraulics and flow management in breeder reactors
International Nuclear Information System (INIS)
Carelli, M.D.; Willis, J.M.
1979-01-01
An analytical model representing the hydraulic behavior of the primary system of fast breeder nuclear reactors is discussed. A computer code capable of detailing the core flow distribution and characterizing the flow and pressure drop in each reactor component is presented. Application of this method to the reactor core thermal-hydraulic design has allowed optimization of the flow management with consequent upgrading in performance, reduction of unnecessary conservatism and very substantial cost savings. Typical quantitative examples are presented
Soil hydraulic properties near saturation, an improved conductivity model
DEFF Research Database (Denmark)
Børgesen, Christen Duus; Jacobsen, Ole Hørbye; Hansen, Søren
2006-01-01
of commonly used hydraulic conductivity models and give suggestions for improved models. Water retention and near saturated and saturated hydraulic conductivity were measured for a variety of 81 top and subsoils. The hydraulic conductivity models by van Genuchten [van Genuchten, 1980. A closed-form equation....... Reports and Dissertations 9.] were optimised to describe the unsaturated hydraulic conductivity in the range measured. Different optimisation procedures were tested. Using the measured saturated hydraulic conductivity in the vGM model tends to overestimate the unsaturated hydraulic conductivity....... Optimising a matching factor (k0) improved the fit considerably whereas optimising the l-parameter in the vGM model improved the fit only slightly. The vGM was improved with an empirical scaling function to account for the rapid increase in conductivity near saturation. Using the improved models...
Space nuclear reactor SP-100 thermal-hydraulic simulation
International Nuclear Information System (INIS)
Borges, Eduardo M.; Braz Filho, Francisco A.; Camillo, Giannino P.; Guimaraes, Lamartine N.F.
2009-01-01
Since 1983 it has been under development in the USA the project SP-100 of space nuclear reactors for electric generation in a range of 100 to 1000 KWe. In this project the heat is generated at the core of a fast compact liquid lithium refrigerated reactor. Thermoelectric converters produce direct current electric energy and the primary and secondary loops flow is controlled by electromagnetic thermoelectric pumps (EMTE). In this work it is studied a system with a fast nuclear reactor, with similar characteristics to the SP-100, aiming at generating high electric power in space for a future application on the TERRA (Advanced Fast Reactor Technology) Project of IEAv (Institute for Advanced Studies). It will be presented the working principles, basic structure and operation characteristics of an electromagnetic thermoelectric pump (EMTE) for a liquid metal cooled nuclear reactor refrigeration loops flow control. In order to determine the operating point of the reactor, it is indispensable the simulation of the EMTE pump along with the other components of the system, once all the working parameters are connected. So, it has been developed a computer system, named BEMTE-3 (a FORTRAN micro-computer code), which simulates the primary and secondary refrigeration components of liquid metal cooled fast space reactor. This computer code also simulates the thermoelectric conversion, with the flow being controlled by the EMTE pump with thermoelectric converters, determining the system operation point for a given nominal operating power. The BEMTE-3 is used for the study of the SP-100 primary and secondary loops thermal-hydraulic simulation and for the calculation of the operating point of the system based on data from available projects. (author)
Development of the NSSS thermal-hydraulic program for YGN unit 1 simulator
Energy Technology Data Exchange (ETDEWEB)
Kim, Kyung Doo; Jeong, Jae Jun; Lee, Won Jae; Chung, Bub Dong; Ha, Kwi Seok; Kang, Kyung Ho
2000-09-01
The NSSS thermal-hydraulic programs installed in the domestic full-scope power plant simulators were provided in early 1980s by foreign vendors. Because of limited computational capability at that time, they usually adopt very simplified physical models for a real-time simulation of NSSS thermal-hydraulic phenomena, which entails inaccurate results and the possibility of so-called 'negative training', especially for complicated two-phase flows in the reactor coolant system. To resolve the problem, we developed a realistic NSSS T/H program (named 'ARTS' code) for use in YongGwang Nuclear Unit 1 full-scope simulator. The best-estimate code RETRAN03, developed by EPRI and approved by USNRC, was selected as a reference code of ARTS. For the development of ARTS, the followings have been performed: -Improvement of the robustness of RETRAN - Improvement of the real-time simulation capability of RETRAN - Optimum input data generation for the NSSS simulation - New model development that cannot be efficiently modeled by RETRAN - Assessment of the ARTS code. The systematic assessment of ARTS has been conducted in both personal computers (Windows 98, Visual fortran) and the simulator development environment (Windows NT, GSE simulator development tool). The results were resonable in terms of accuracy, real-time simulation and robustness.
How good are thermal-hydraulics codes for analyses of plant transients
International Nuclear Information System (INIS)
Fabic, S.
1996-01-01
In the early seventies, all thermal-hydraulics codes were based on the Homogeneous Equilibrium Model (HEM), represented by three conservation equations: mixture mass, momentum and energy. Various means were utilized to solve the resulting system of equations: finite differences in FLASH, SATAN, RELAP3 and RELAP4, method of characteristics in BLOWDWN2, loop momentum method in RAMONA and NORCOOL, and others. As the result the world came to regard HEM as too restrictive and the Two-Fluid model came into fashion, first featuring a six and later, a seven-equation model. New codes like KACHINA, TRAC and RELAP5 were developed also. Experience and comparisons with test data have recently forced us to wonder whether the ability to 'compute' while considering great many complexities, ran ahead of the ability to competently define various interactions between fluid phases and components that such complex codes require. The long running times are also a problem that needs to be resolved. More recent trends in the treatment of thermal-hydraulics in Power Plant Simulators and in Plant Analyzers will also be discussed
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)
International Nuclear Information System (INIS)
Sahu, M.K.; John, P.K.; Jayaraj, N.; Chatterjee, P.B.; Das, Sandeep; John, Benny; Sharma, A.K.; Prasad, N.; Singhal, M.; Malhotra, P.K.; Haldar, S.C.; Bhambra, H.S.; Chadda, S.K.; Chandra, Umesh
2006-01-01
Experimental thermal hydraulic facility being set up adjacent to R and D Centre at Tarapur is a 13 MW full-elevation scaled down facility having the key components of PHT System of Pressurised Heavy Water Reactor (PHWR). The objective of the facility is to study thermal hydraulic behaviour of PHT System of PHWR by simulating various transients and accidental scenarios, to conduct safety related and operational transient studies and validation of various thermal hydraulic computer codes developed for analysis. The design of thermal hydraulic facility is based on the process parameters of a large PHWR with respect to fluid mass flux, transit time, flow velocity, pressure, temperature and enthalpy in PHT System. Experiments would be conducted in the facility to gain an improved understanding of the thermal hydraulic behaviour of large size PHWR during loss of coolant accident scenarios with forced and natural thermo-siphoning circulation modes etc. The data collected from the experiments would be used in validating computer codes developed for safety analysis. The facility is extensively instrumented to measure parameters such as temperature, pressure, flow, level, void-fraction at key locations. This paper gives the design philosophy used for scaling, design of major components of primary and secondary circuit of Experimental Thermal Hydraulic Facility and details of simulated experiments to be carried out. (author)
Optimizing a gap conductance model applicable to VVER-1000 thermal–hydraulic model
International Nuclear Information System (INIS)
Rahgoshay, M.; Hashemi-Tilehnoee, M.
2012-01-01
Highlights: ► Two known conductance models for application in VVER-1000 thermal–hydraulic code are examined. ► An optimized gap conductance model is developed which can predict the gap conductance in good agreement with FSAR data. ► The licensed thermal–hydraulic code is coupled with the gap conductance model predictor externally. -- Abstract: The modeling of gap conductance for application in VVER-1000 thermal–hydraulic codes is addressed. Two known models, namely CALZA-BINI and RELAP5 gap conductance models, are examined. By externally linking of gap conductance models and COBRA-EN thermal hydraulic code, the acceptable range of each model is specified. The result of each gap conductance model versus linear heat rate has been compared with FSAR data. A linear heat rate of about 9 kW/m is the boundary for optimization process. Since each gap conductance model has its advantages and limitation, the optimized gap conductance model can predict the gap conductance better than each of the two other models individually.
Energy Technology Data Exchange (ETDEWEB)
Kim, Jongtae; Park, Rae-Joon; Hong, Seong-Wan; Kim, Gun-Hong [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)
2016-10-15
In a containment safety analysis, multi-dimensional characteristics in thermal hydraulics are very important because the flow paths are not confined in a large free volume of the containment. The analysis is difficult because of a difference in length scales between a characteristic length of the flow and representative length of the containment. In order to simulate hydrogen and steam behaviors in a containment during postulated severe accidents, the GASFLOW code as a multi-dimensional analysis tool for NPP containment has been used for years because of its computational efficiency. Though GASFLOW is well developed for a real NPP containment analysis, there exist shortcomings in nodalization, two-phase and turbulence models. It is based on a Cartesian or cylindrical coordinate mesh, so it is impractical to refine a mesh locally in a region with a physical or geometrical complication. In this paper, the importance of the hydrogen safety in an NPP containment and requirements of the analysis tool was described. And physical models necessary for the hydrogen safety analysis code were listed. As a member of international collaborative project HYMERES for containment thermal hydraulics, KAERI is actively participating in an analytic working group. As an analysis tool for blind benchmarkes, the analysis code described in this paper was used. From the blind benchmark analyses, it was found that the code is very promising for hydrogen safety analysis. Currently, it is proposed to develop the code collaboratively in a hydrogen safety community based on an open-source strategy.
Thermal-hydraulics stability of natural circulation BWR under startup. Flashing effects
International Nuclear Information System (INIS)
Hu, Rui; Kazimi, Mujid S.
2009-01-01
To help achieve the necessary natural circulation flow, a fairly long chimney is installed in a boiling natural circulation reactor like the ESBWR. In such systems, thermal-hydraulic stability during low pressure start-up should be examined while considering the flashing induced by the pressure drop in the channel and the chimney due to gravity head. In this work, a BWR stability analysis code in the frequency domain, named FISTAB (Flashing-Induced STability Analysis for BWR), was developed to address the issue of flashing-induced instability. A thermal-hydraulics non-homogeneous equilibrium model (NHEM) based on a drift flux formulation along with a lumped fuel dynamics model is incorporated in the work. The vapor generation rate is derived from the mixture energy conservation equation while considering the effect of flashing. The functionality of the FISTAB code was confirmed by comparison to experimental results from SIRIUS-N facility at CRIEPI, Japan. Both stationary and perturbation results agree well with the experimental results. (author)
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)
Energy Technology Data Exchange (ETDEWEB)
Page, R.; Jones, J.R.
1997-07-01
Ensuring that safety analysis needs are met in the future is likely to lead to the development of new codes and the further development of existing codes. It is therefore advantageous to define standards for data interfaces and to develop software interfacing techniques which can readily accommodate changes when they are made. Defining interface standards is beneficial but is necessarily restricted in application if future requirements are not known in detail. Code interfacing methods are of particular relevance with the move towards automatic grid frequency response operation where the integration of plant dynamic, core follow and fault study calculation tools is considered advantageous. This paper describes the background and features of a new code TALINK (Transient Analysis code LINKage program) used to provide a flexible interface to link the RELAP5 thermal hydraulics code with the PANTHER neutron kinetics and the SIBDYM whole plant dynamic modelling codes used by Nuclear Electric. The complete package enables the codes to be executed in parallel and provides an integrated whole plant thermal-hydraulics and neutron kinetics model. In addition the paper discusses the capabilities and pedigree of the component codes used to form the integrated transient analysis package and the details of the calculation of a postulated Sizewell `B` Loss of offsite power fault transient.
Thermal hydraulic method for whole core design analysis of an HTGR
International Nuclear Information System (INIS)
Huning, A. J.; Garimella, S.
2013-01-01
A new thermal hydraulic method and initial results are presented for core-wide steady state analysis of prismatic High Temperature Gas-Cooled Reactors (HTGR). The method allows for the complete solution of temperature and coolant mass flow distribution by solving quasi-steady energy balances for the discretized core. Assembly blocks are discretized into unit cells for which the average temperature of each unit cell is determined. Convective heat removal is coupled to the unit cell energy balances by a 1-D axial flow model. The flow model uses established correlations for friction factor and Nusselt number. Bypass flow is explicitly calculated by using an initial guess for mass flow distribution and determining the exit pressure of each flow channel. The mass flow distribution is updated until a uniform core exit pressure condition is reached. Results are obtained for the MHTGR-350 with emphasis on the change in thermal hydraulic parameters due to various steady state power profiles and bypass gap widths. Steady state temperature distribution and its variations are discussed. (authors)
SB LOCA thermal-hydraulic analyses for Krsko Full Scope Simulator validation
International Nuclear Information System (INIS)
Prosek, A.; Parzer, I.
2005-01-01
The Krsko nuclear power plant (NPP), which is a two-loop pressurized water reactor, Westinghouse type, before modernization in 2000 obtained plant specific full scope simulator. The purpose of the presented analyses was to perform Small Break Loss of Coolant Accident (SBLOCA) reference calculations for KFSS validation in 2004. In addition, the thermal-hydraulic response of the reactor coolant system (RCS) was studied in detail. For the thermal-hydraulic analysis the RELAP5/MOD3.3 code and input model delivered from Krsko NPP were used. The RELAP5 calculated reference results showed that the plant system response to breaks with small break area is slower compared to breaks with larger break area. The comparison of the KFSS data with calculated results suggest that the simulator validation testing in the year 2004 for this kind of accident was successful. Nevertheless, when comparing the physical phenomena and processes, the RELAP5/MOD3.3 predicted smaller core uncovery compared to the KFSS measurement. One reason is different core cycles. Finally, this finding suggests that even for simulator reference calculations the quantification of model uncertainties would be useful. (author)
Best-estimate plus uncertainty thermal-hydraulic stability analysis of BWRs using TRACG code
International Nuclear Information System (INIS)
Vedovi, J.; Yang, J.; Klebanov, L.; Vreeland, D. G.; Zino, J. F.
2012-01-01
Over the last decade, Boiling Water Reactor (BWR) power up-rates have increased plant rated power output significantly. Subsequent projects have expanded flow domains (e.g. MELLLA+) for operation at these higher power levels. This has resulted in an increase in the power to flow ratio in regions susceptible to reactor thermal-hydraulic instabilities. Since BWRs are susceptible to coupled thermal-hydraulic/nuclear oscillations when operating at these conditions, such oscillations must be prevented or reliably detected and suppressed. The Detect and Suppress Solution - Confirmation Density (DSS-CD) is the most sophisticated GEH BWR instability protection system ever employed. DSS-CD implements algorithms that monitor closely-spaced groups of Local Power Range Monitor (LPRM) detectors to detect periodic behavior typical of reactor instability events. This system is able to detect small, localized power variations in the core, distinguish between true instabilities and plant noise, and trip/scram the reactor while maintaining adequate safety margins. The combination of hardware, software, and system setpoints provides protection against violation of the Safety Limit Minimum Critical Power Ratio (SLMCPR) for anticipated oscillations. To support DSS-CD implementation, the TRACG system code is used to simulate events to confirm the capability of the DSS-CD solution for early oscillation detection and suppression. TRACG is a GEH proprietary version of the Transient Reactor Analysis Code (TRAC). TRACG includes a multi-dimensional, two-fluid model for the reactor thermal-hydraulics and a three-dimensional reactor kinetics model. The models are qualified to simulate a large variety of tests and reactor configurations, including thermal-hydraulic stability events. These features allow for detailed, best-estimate simulation of a wide range of BWR phenomena. A set of integrated TRACG event simulations for reasonably limiting anticipated events can be used to calculate the effect
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)
Application of an accurate thermal hydraulics solver in VTT's reactor dynamics codes
International Nuclear Information System (INIS)
Rajamaeki, M.; Raety, H.; Kyrki-Rajamaeki, R.; Eskola, M.
1998-01-01
VTT's reactor dynamics codes are developed further and new more detailed models are created for tasks related to increased safety requirements. For thermal hydraulics calculations an accurate general flow model based on a new solution method PLIM has been developed. It has been applied in VTT's one-dimensional TRAB and three-dimensional HEXTRAN codes. Results of a demanding international boron dilution benchmark defined by VTT are given and compared against results of other codes with original or improved boron tracking. The new PLIM method not only allows the accurate modelling of a propagating boron dilution front, but also the tracking of a temperature front, which is missed by the special boron tracking models. (orig.)
Transient thermal-hydraulic characteristics analysis software for PWR nuclear power systems
International Nuclear Information System (INIS)
Wu Yingwei; Zhuang Chengjun; Su Guanghui; Qiu Suizheng
2010-01-01
A point reactor neutron kinetics model, a two-phase drift-flow U-tube steam generator model, an advanced non-equilibrium three regions pressurizer model, and a passive emergency core decay heat-removed system model are adopted in the paper to develop the computerized analysis code for PWR transient thermal-hydraulic characteristics, by Compaq Visual Fortran 6.0 language. Visual input, real-time processing and dynamic visualization output are achieved by Microsoft Visual Studio. NET language. The reliability verification of the soft has been conducted by RELAP 5, and the verification results show that the software is with high calculation precision, high calculation speed, modern interface, luxuriant functions and strong operability. The software was applied to calculate the transient accident conditions for QSNP, and the analysis results are significant to the practical engineering applications. (authors)
Spent fuel pool thermal-hydraulic analysis using RELAP5-3D
Energy Technology Data Exchange (ETDEWEB)
Ramos, M. C.; Fernandes, G.H.N.; Costa, A.L.; Pereira, F.; Pereira, C., E-mail: marc5663@gmail.com, E-mail: ghnfernandes@pq.cnpq.br, E-mail: claubia@nuclear.ufmg.br, E-mail: antonella@nuclear.ufmg.br [Universidade Federal de Minas Gerais, Belo Horizonte, MG (Brazil). Departamento de Engenharia Nuclear
2017-07-01
In order to analyze the thermo-hydraulic behavior of spent fuel pools, and taking as reference a hypothetic PWR nuclear plant, a model of RELAP-3D for a spent fuel pool has been built. This model has been used to simulate a loss of coolant in SPF. This study focuses on the loss of coolant flow accident in spent fuel storage pool which is modelled by using RELAP5-3D code to observe the coolant level reduction and fuel uncovery because of decay heat generation of the spent fuel in the pool. The results have been compared with the available data. The developed model demonstrated that the RELAP5-3D is capable of reproduce the thermal behavior of SPF in a transient scenario. (author)
International Nuclear Information System (INIS)
Bolander, M.A.
1985-10-01
Thermal-hydraulic analyses of two pressurized thermal shock sequences for the Oconee-1 pressurized water reactor were performed at the Idaho National Engineering Laboratory using the RELAP5/MOD2 computer code. This report presents the results of these calculations
International Nuclear Information System (INIS)
Peng Muzhang; Zhang Quan; Wang Guoli; Zhang Yuman
1988-01-01
TISKTH-3 is a coupled neutronics/thermal-hydraulics code for the transient analysis. A 3-dimensional neutron kinetics equation solved by the Nodal Green's Function Method is used for the neutronics model of the code. A homogeneous equilibrium model with a complete boiling curve and two numerical solutions of the implicit and explicit scheme is used for the thermal-hydraulics model of the code. A 2-dimensional heat conduction equation with variable conductivity solved by the method of weighted residuals is used for the fuel rod heat transfer model of the code. TISKTH-3 is able to analyze the fast transient process and complicate accident situations in the core. The initative applications have shown that the stability and convergency in the calculations with the code are satisfactory
Uncertainty analysis for results of thermal hydraulic codes of best-estimate-type
International Nuclear Information System (INIS)
Alva N, J.
2010-01-01
In this thesis, some fundamental knowledge is presented about uncertainty analysis and about diverse methodologies applied in the study of nuclear power plant transient event analysis, particularly related to thermal hydraulics phenomena. These concepts and methodologies mentioned in this work come from a wide bibliographical research in the nuclear power subject. Methodologies for uncertainty analysis have been developed by quite diverse institutions, and they have been widely used worldwide for application to results from best-estimate-type computer codes in nuclear reactor thermal hydraulics and safety analysis. Also, the main uncertainty sources, types of uncertainties, and aspects related to best estimate modeling and methods are introduced. Once the main bases of uncertainty analysis have been set, and some of the known methodologies have been introduced, it is presented in detail the CSAU methodology, which will be applied in the analyses. The main objective of this thesis is to compare the results of an uncertainty and sensibility analysis by using the Response Surface Technique to the application of W ilks formula, apply through a loss coolant experiment and an event of rise in a BWR. Both techniques are options in the part of uncertainty and sensibility analysis of the CSAU methodology, which was developed for the analysis of transients and accidents at nuclear power plants, and it is the base of most of the methodologies used in licensing of nuclear power plants practically everywhere. Finally, the results of applying both techniques are compared and discussed. (Author)
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.
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)
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
Development of numerical simulation technology for high resolution thermal hydraulic analysis
International Nuclear Information System (INIS)
Yoon, Han Young; Kim, K. D.; Kim, B. J.; Kim, J. T.; Park, I. K.; Bae, S. W.; Song, C. H.; Lee, S. W.; Lee, S. J.; Lee, J. R.; Chung, S. K.; Chung, B. D.; Cho, H. K.; Choi, S. K.; Ha, K. S.; Hwang, M. K.; Yun, B. J.; Jeong, J. J.; Sul, A. S.; Lee, H. D.; Kim, J. W.
2012-04-01
A realistic simulation of two phase flows is essential for the advanced design and safe operation of a nuclear reactor system. The need for a multi dimensional analysis of thermal hydraulics in nuclear reactor components is further increasing with advanced design features, such as a direct vessel injection system, a gravity driven safety injection system, and a passive secondary cooling system. These features require more detailed analysis with enhanced accuracy. In this regard, KAERI has developed a three dimensional thermal hydraulics code, CUPID, for the analysis of transient, multi dimensional, two phase flows in nuclear reactor components. The code was designed for use as a component scale code, and/or a three dimensional component, which can be coupled with a system code. This report presents an overview of the CUPID code development and preliminary assessment, mainly focusing on the numerical solution method and its verification and validation. It was shown that the CUPID code was successfully verified. The results of the validation calculations show that the CUPID code is very promising, but a systematic approach for the validation and improvement of the physical models is still needed
The Thermal Hydraulics of Tube Support Fouling in Nuclear Steam Generators
International Nuclear Information System (INIS)
Rummens, Helena E.C.; Rogers, J.T.; Turner, C.W.
2004-01-01
It is hypothesized that the thermal-hydraulic environment plays a role in the fouling of tube supports in nuclear steam generators. Experiments were performed to simulate the thermal-hydraulic environment near various designs of supports. Pressure loss, local velocity, turbulence intensity, and local void fraction were measured to characterize the effect of the support. Fouling mechanisms specific to supports were inferred from these experimental data and from actual steam generator inspection results. An analytical model was developed to predict the rate of particulate deposition on the supports, to better understand the complex processes involved.This paper presents the following set of tools for assessing the fouling propensity of a given support design: (1) proposed fouling mechanisms, (2) criteria for support fouling propensity, (3) correlation of fouling with parameters such as mass flux and quality, (4) descriptions of experimental tools such as flow visualization and measurement of pressure-loss profiles, and (5) analytical tools.An important conclusion from this and our previous work is that the fouling propensity is greater with broached support plates, both trefoil and quatrefoil, than with lattice bar supports and formed bar supports, in which significant cross flows occur
Teaching Thermal Hydraulics and Numerical Methods: An Introductory Control Volume Primer
International Nuclear Information System (INIS)
D. S. Lucas
2004-01-01
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
Thermal-hydraulic analysis of research reactor core with different LEU fuel types using RELAP5
Energy Technology Data Exchange (ETDEWEB)
El-Sahlamy, Neama M. [Nuclear and Radiological Regulatory Authority, Cairo (Egypt)
2017-11-15
In the current work, comparisons between the core performances when using different LEU fuels are done. The fuels tested are UA1{sub X}-A1, U{sub 3}O{sub 8}-Al, and U{sub 3}Si{sub 2}-Al fuels with 19.7 % enrichment. Calculations are done using RELAP5 code to evaluate the thermal-hydraulic performance of the IAEA benchmark 10 MW reactor. First, a reassessment of the slow reactivity insertion transient with UA1{sub X}-A1 LEU fuel to compare the results with those reported in the IAEA TECDOC [1]. Then, comparisons between the thermal-hydraulic core performances when using the three LEU fuels are done. The assessment is performed at initial power of 1.0 W. The reactor power is calculated using the RELAP5 point kinetic model. The reactivity feedback, from changes in water density and fuel temperature, is considered for all cases. From the results it is noticed that U{sub 3}Si{sub 2}-Al fuel gives the best fuel performance since it has the minimum value of peak fuel temperature and the minimum peak clad surface temperature, as operating parameters. Also, it gives the maximum value of the Critical Heat Flux Ratio and the lowest tendency to flow instability occurrence.
Program ELM: A tool for rapid thermal-hydraulic analysis of solid-core nuclear rocket fuel elements
Walton, James T.
1992-01-01
This report reviews the state of the art of thermal-hydraulic analysis codes and presents a new code, Program ELM, for analysis of fuel elements. ELM is a concise computational tool for modeling the steady-state thermal-hydraulics of propellant flow through fuel element coolant channels in a nuclear thermal rocket reactor with axial coolant passages. The program was developed as a tool to swiftly evaluate various heat transfer coefficient and friction factor correlations generated for turbulent pipe flow with heat addition which have been used in previous programs. Thus, a consistent comparison of these correlations was performed, as well as a comparison with data from the NRX reactor experiments from the Nuclear Engine for Rocket Vehicle Applications (NERVA) project. This report describes the ELM Program algorithm, input/output, and validation efforts and provides a listing of the code.
International Nuclear Information System (INIS)
Masanori Ohtani; Akito Kozuru; Yasuyuki Kashimoto; Mitsuto Montani; Koutaro Takeda; Yasushi Makino
2006-01-01
Asymmetric thermal-hydraulic conditions among primary loops during a postulated steam line break (SLB) induce a non-uniform temperature distribution at a core inlet. When coolant of lower temperature intrudes into a part of core, it leads to a reactivity insertion and a local power increase. Therefore, an appropriate model for the core inlet temperature distribution is required for a realistic SLB analysis. In this study, numerical experiments were conducted to examine the core inlet temperature distribution under the asymmetric thermal-hydraulic coolant conditions among primary loops. 3D steady-state calculations were carried out for Japanese standard Pressurized Water Reactor (PWR) such as 2, 3, 4 loop types and an advanced PWR. Since the flow in a reactor vessel involves time-dependent velocity fluctuations due to a high Reynolds number condition and a complicated geometry of flow path, the turbulent mixing might be enhanced. Hence, the turbulent thermal diffusivity for the steady-state calculation was examined based on experimental results and another transient calculation. As a result, it was confirmed that (1) the turbulent mixing in a downcomer and a lower plenum were enhanced due to time-dependent velocity fluctuations and therefore the turbulent thermal diffusivity for steady-state calculation was specified to be greater, (2) the core inlet temperature distribution predicted by a steady-state calculation reasonably agreed with a experimental data, (3) the patterns of core inlet temperature distribution were comprehended to be dependent on the plant type, i.e. the number of primary loop and (4) under a low flow rate condition, the coolant of lower temperature appeared on the opposite side of the affected loop due to the effect of a natural convection. (authors)
International Nuclear Information System (INIS)
Plas, Roger.
1975-05-01
This computer program describes the flow and heat transfer in steady and transient state in two-phase flows. It is the present stage of the evolution about FLICA, FLICA II and FLICA II B codes which have been used and developed at CEA for the thermal-hydraulic analysis of reactors and experimental loops with heating rod bundles. In the mathematical model all the significant terms of the fundamental hydrodynamic equations are taken into account with the approximations of turbulent viscosity and conductivity. The two-phase flow is calculated by the homogeneous model with slip. In the flow direction an implicit resolution scheme is available, which make possible to study partial or total flow blockage, with upstream and downstream effects. A special model represents the helical wire effects in out-of pile experimental rod bundles [fr
Comparison of empirical models and laboratory saturated hydraulic ...
African Journals Online (AJOL)
Numerous methods for estimating soil saturated hydraulic conductivity exist, which range from direct measurement in the laboratory to models that use only basic soil properties. A study was conducted to compare laboratory saturated hydraulic conductivity (Ksat) measurement and that estimated from empirical models.
Energy Technology Data Exchange (ETDEWEB)
Myung, H. K.; Yang, S. Y.; Kim, B. H.; Song, J. H.; Oh, J. Z. [Kookmin University, Seoul (Korea)
2002-03-01
The flow through a nuclear rod bundle with mixing vanes is very complex and so required a suitable turbulence model for its accurate prediction. Subchannel flow in a nuclear bundle having vanes to mix flow appears complex turbulent flow. Objective of this study is to investigate performance of prediction about turbulence model contained in STAR-CD code and to develop suitable turbulence model which can predict complex flow in nuclear assembly. For several nonlinear {kappa}-{epsilon} turbulence models, their performance were investigated in the prediction of the flow in nuclear fuel assembly, and also their problems were discussed in detail. The results obtained from the present research would give a help for the development of turbulence model which can accurately predict the flow through the rod bundles with mixing vanes. 19 refs., 32 figs., 3 tabs. (Author)
High fidelity thermal-hydraulic analysis using CFD and massively parallel computers
International Nuclear Information System (INIS)
Weber, D.P.; Wei, T.Y.C.; Brewster, R.A.; Rock, Daniel T.; Rizwan-uddin
2000-01-01
Thermal-hydraulic analyses play an important role in design and reload analysis of nuclear power plants. These analyses have historically relied on early generation computational fluid dynamics capabilities, originally developed in the 1960s and 1970s. Over the last twenty years, however, dramatic improvements in both computational fluid dynamics codes in the commercial sector and in computing power have taken place. These developments offer the possibility of performing large scale, high fidelity, core thermal hydraulics analysis. Such analyses will allow a determination of the conservatism employed in traditional design approaches and possibly justify the operation of nuclear power systems at higher powers without compromising safety margins. The objective of this work is to demonstrate such a large scale analysis approach using a state of the art CFD code, STAR-CD, and the computing power of massively parallel computers, provided by IBM. A high fidelity representation of a current generation PWR was analyzed with the STAR-CD CFD code and the results were compared to traditional analyses based on the VIPRE code. Current design methodology typically involves a simplified representation of the assemblies, where a single average pin is used in each assembly to determine the hot assembly from a whole core analysis. After determining this assembly, increased refinement is used in the hot assembly, and possibly some of its neighbors, to refine the analysis for purposes of calculating DNBR. This latter calculation is performed with sub-channel codes such as VIPRE. The modeling simplifications that are used involve the approximate treatment of surrounding assemblies and coarse representation of the hot assembly, where the subchannel is the lowest level of discretization. In the high fidelity analysis performed in this study, both restrictions have been removed. Within the hot assembly, several hundred thousand to several million computational zones have been used, to
Energy Technology Data Exchange (ETDEWEB)
Jensen, Colby B.; Folsom, Charles P.; Davis, Cliff B.; Woolstenhulme, Nicolas E.; Bess, John D.; O' Brien, Robert C.; Ban, Heng; Wachs, Daniel M.
2016-08-01
Experimental testing in the Multi-Static Environment Rodlet Transient Test Apparatus (SERTTA) will lead the rebirth of transient fuel testing in the United States as part of the Accident Tolerant Fuels (ATF) progam. The Multi-SERTTA is comprised of four isolated pressurized environments capable of a wide variety of working fluids and thermal conditions. Ultimately, the TREAT reactor as well as the Multi-SERTTA test vehicle serve the purpose of providing desired thermal-hydraulic boundary conditions to the test specimen. The initial ATF testing in TREAT will focus on reactivity insertion accident (RIA) events using both gas and water environments including typical PWR operating pressures and temperatures. For the water test environment, a test configuration is envisioned using the expansion tank as part of the gas-filled expansion volume seen by the test to provide additional pressure relief. The heat transfer conditions during the high energy power pulses of RIA events remains a subject of large uncertainty and great importance for fuel performance predictions. To support transient experiments, the Multi-SERTTA vehicle has been modeled using RELAP5 with a baseline test specimen composed of UO2 fuel in zircaloy cladding. The modeling results show the influence of the designs of the specimen, vehicle, and transient power pulses. The primary purpose of this work is to provide input and boundary conditions to fuel performance code BISON. Therefore, studies of parameters having influence on specimen performance during RIA transients are presented including cladding oxidation, power pulse magnitude and width, cladding-to-coolant heat fluxes, fuel-to-cladding gap, transient boiling effects (modified CHF values), etc. The results show the great flexibility and capacity of the TREAT Multi-SERTTA test vehicle to provide testing under a wide range of prototypic thermal-hydraulic conditions as never done before.
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
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)
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 AsociaciÃƒÂ³n Nuclear AscÃƒÂ³-VandellÃƒÂ²s (ANAV. ANAV is the consortium that runs the AscÃƒÂ³ power plants (2 units and the VandellÃƒÂ²s-II power plant. The reactors are Westinghouse-design, 3-loop PWRs with an approximate electrical power of 1000Ã¢Â€Â‰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.
Energy Technology Data Exchange (ETDEWEB)
Nelson, R.A. Jr.; Pimentel, D.A.; Jolly-Woodruff, S.; Spore, J.
1998-04-01
In this report, a phenomenological model of simultaneous bottom-up and top-down quenching is developed and discussed. The model was implemented in the TRAC-PF1/MOD2 computer code. Two sets of closure relationships were compared within the study, the Absolute set and the Conditional set. The Absolute set of correlations is frequently viewed as the pure set because the correlations is frequently viewed as the pure set because the correlations utilize their original coefficients as suggested by the developer. The Conditional set is a modified set of correlations with changes to the correlation coefficient only. Results for these two sets indicate quite similar results. This report also summarizes initial results of an effort to investigate nonlinear optimization techniques applied to the closure model development. Results suggest that such techniques can provide advantages for future model development work, but that extensive expertise is required to utilize such techniques (i.e., the model developer must fully understand both the physics of the process being represented and the computational techniques being employed). The computer may then be used to improve the correlation of computational results with experiments.
International Nuclear Information System (INIS)
Ventura, Mirta A.; Rosso, Ricardo D.
2000-01-01
Time evolution of natural circulation in the Atucha I nuclear power plant (CNA-I), in a main pumps lost incident because of the lost of external power feed, is analyzed. It leads to a strong stop transient, without an important blow down, from a forced nominal flow to a natural circulation one. The results are obtained from RELAP5/3.2 code's modeling. The study is based on the refrigeration condition analysis, during the first minutes of the reactor out of service. Previously to the transient, work had been done to obtain the plant steady state, with design parameters in operation conditions at 100 % of power. The object is that the actual plant state would be represented. In this way, each plant part (steam generators, reactor, pressurizer, pumps) had been modeled in separated form with the appropriate boundary conditions to be used in the whole circuit simulation. The developed model had been validated making use of the comparison between the values obtained to the principal thermodynamic parameters with the plant recorder values, in the same incident. The results are satisfactory in a way. On the other hand, it has suggested some modeling changes. The RELAP5/3.2 capability to model the thermodynamic phenomena in a PHWR plant has been verified when, according to the mentioned incident, the flow pass from a nominal forced flow, to one which is governed by natural circulation, still with the CNA-I untypical design conditions. (authors)
Thermal hydraulic codes for LWR safety analysis - present status and future perspective
Energy Technology Data Exchange (ETDEWEB)
Staedtke, H. [Commission of the European Union, Ispra (Italy)
1997-07-01
The aim of the present paper is to give a review on the current status and future perspective of present best-estimate Thermal Hydraulic codes. Reference is made to internationally well-established codes which have reached a certain state of maturity. The first part of the paper deals with the common basic code features with respect to the physical modelling and their numerical methods used to describe complex two-phase flow and heat transfer processes. The general predictive capabilities are summarized identifying some remaining code deficiencies and their underlying limitations. The second part discusses various areas including physical modelling, numerical techniques and informatic structure where the codes could be substantially improved.
2D Thermal Hydraulic Analysis and Benchmark in Support of HFIR LEU Conversion using COMSOL
Energy Technology Data Exchange (ETDEWEB)
Freels, James D [ORNL; Bodey, Isaac T [ORNL; Lowe, Kirk T [ORNL; Arimilli, Rao V [ORNL
2010-09-01
The research documented herein was funded by a research contract between the Research Reactors Division (RRD) of Oak Ridge National Laboratory (ORNL) and the University of Tennessee, Knoxville (UTK) Mechanical, Aerospace and Biomedical Engineering Department (MABE). The research was governed by a statement of work (SOW) which clearly defines nine specific tasks. This report is outlined to follow and document the results of each of these nine specific tasks. The primary goal of this phase of the research is to demonstrate, through verification and validation methods, that COMSOL is a viable simulation tool for thermal-hydraulic modeling of the High Flux Isotope Reactor (HFIR) core. A secondary goal of this two-dimensional phase of the research is to establish methodology and data base libraries that are also needed in the full three-dimensional COMSOL simulation to follow. COMSOL version 3.5a was used for all of the models presented throughout this report.
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.
Driba, D. L.; De Lucia, M.; Peiffer, S.
2014-12-01
Fluid-rock interactions in geothermal reservoirs are driven by the state of disequilibrium that persists among solid and solutes due to changing temperature and pressure. During operation of enhanced geothermal systems, injection of cooled water back into the reservoir disturbs the initial thermodynamic equilibrium between the reservoir and its geothermal fluid, which may induce modifications in permeability through changes in porosity and pore space geometry, consequently bringing about several impairments to the overall system.Modeling of fluid-rock interactions induced by injection of cold brine into Groß Schönebeck geothermal reservoir system situated in the Rotliegend sandstone at 4200m depth have been done by coupling geochemical modeling Code Phreeqc with OpenGeoSys. Through batch modeling the re-evaluation of the measured hydrochemical composition of the brine has been done using Quintessa databases, the results from the calculation indicate that a mineral phases comprising of K-feldspar, hematite, Barite, Calcite and Dolomite was found to match the hypothesis of equilibrium with the formation fluid, Reducing conditions are presumed in the model (pe = -3.5) in order to match the amount of observed dissolved Fe and thus considered as initial state for the reactive transport modeling. based on a measured composition of formation fluids and the predominant mineralogical assemblage of the host rock, a preliminary 1D Reactive transport modeling (RTM) was run with total time set to 30 years; results obtained for the initial simulation revealed that during this period, no significant change is evident for K-feldspar. Furthermore, the precipitation of calcite along the flow path in the brine results in a drop of pH from 6.2 to a value of 5.2 noticed over the simulated period. The circulation of cooled fluid in the reservoir is predicted to affect the temperature of the reservoir within the first 100 -150m from the injection well. Examination of porosity change in
Peabody, Hume L.
2017-01-01
This presentation is meant to be an overview of the model building process It is based on typical techniques (Monte Carlo Ray Tracing for radiation exchange, Lumped Parameter, Finite Difference for thermal solution) used by the aerospace industry This is not intended to be a "How to Use ThermalDesktop" course. It is intended to be a "How to Build Thermal Models" course and the techniques will be demonstrated using the capabilities of ThermalDesktop (TD). Other codes may or may not have similar capabilities. The General Model Building Process can be broken into four top level steps: 1. Build Model; 2. Check Model; 3. Execute Model; 4. Verify Results.
International Nuclear Information System (INIS)
Haga, Katsuhiro; Terada, Atsuhiko; Kaminaga, Masanori; Hino, Ryutaro
2001-01-01
In this study the effectiveness of the cross-flow type mercury target structure was evaluated experimentally and analytically. The average water flow velocity field in the target mock-up model, which was fabricated with plexiglass, was measured at room temperature using the PIV (Particle Image Velocimetry) technique. The water flow analyses were conducted and the analytical results were compared with the experimental results. The experimental results showed that the cross-flow could be realized in the former part of the proton beam path where the heat load by the spallation reaction is large, and the analytical result of the water flow velocity field showed good correspondence to the experimental result in the case of the Reynolds number of more than 4.83 x 10 5 at the model inlet. With these results, the effectiveness of the cross-flow type mercury target structure and the present analysis code system was demonstrated. Then the mercury flow field and the temperature distribution in the target container were analyzed assuming the proton beam energy and power of 3 GeV and 5 MW. The analytical result showed that the cross-flow field of mercury, which is similar to the water flow field, could also be attained. (author)
Hydraulic root water uptake models: old concerns and new insights
Couvreur, V.; Carminati, A.; Rothfuss, Y.; Meunier, F.; Vanderborght, J.; Javaux, M.
2014-12-01
Root water uptake (RWU) affects underground water dynamics, with consequences on plant water availability and groundwater recharge. Even though hydrological and climate models are sensitive to RWU parameters, no consensus exists on the modelling of this process. Back in the 1940ies, Van Den Honert's catenary approach was the first to investigate the use of connected hydraulic resistances to describe water flow in whole plants. However concerns such as the necessary computing when architectures get complex made this approach premature. Now that computing power increased dramatically, hydraulic RWU models are gaining popularity, notably because they naturally produce observed processes like compensatory RWU and hydraulic redistribution. Yet major concerns remain. Some are more fundamental: according to hydraulic principles, plant water potential should equilibrate with soil water potential when the plant does not transpire, which is not a general observation when using current definitions of bulk or average soil water potential. Other concerns regard the validation process: water uptake distribution is not directly measurable, which makes it hard to demonstrate whether or not hydraulic models are more accurate than other models. Eventually parameterization concerns exist: root hydraulic properties are not easily measurable, and would even fluctuate on an hourly basis due to processes like aquaporin gating. While offering opportunities to validate hydraulic RWU models, newly developed observation techniques also make us realize the increasing complexity of processes involved in soil-plant hydrodynamics, such as the change of rhizosphere hydraulic properties with soil drying. Surprisingly, once implemented into hydraulic models, these processes do not necessarily translate into more complex emerging behavior at plant scale, and might justify the use of simplified representations of the soil-plant hydraulic system.
Thermal hydraulic-severe accident code interfaces for SCDAP/RELAP5/MOD3.2
Energy Technology Data Exchange (ETDEWEB)
Coryell, E.W.; Siefken, L.J.; Harvego, E.A. [Idaho National Engineering Lab., Idaho Falls, ID (United States)] [and others
1997-07-01
The SCDAP/RELAP5 computer code is designed to describe the overall reactor coolant system thermal-hydraulic response, core damage progression, and fission product release during severe accidents. The code is being developed at the Idaho National Engineering Laboratory under the primary sponsorship of the Office of Nuclear Regulatory Research of the U.S. Nuclear Regulatory Commission. The code is the result of merging the RELAP5, SCDAP, and COUPLE codes. The RELAP5 portion of the code calculates the overall reactor coolant system, thermal-hydraulics, and associated reactor system responses. The SCDAP portion of the code describes the response of the core and associated vessel structures. The COUPLE portion of the code describes response of lower plenum structures and debris and the failure of the lower head. The code uses a modular approach with the overall structure, input/output processing, and data structures following the pattern established for RELAP5. The code uses a building block approach to allow the code user to easily represent a wide variety of systems and conditions through a powerful input processor. The user can represent a wide variety of experiments or reactor designs by selecting fuel rods and other assembly structures from a range of representative core component models, and arrange them in a variety of patterns within the thermalhydraulic network. The COUPLE portion of the code uses two-dimensional representations of the lower plenum structures and debris beds. The flow of information between the different portions of the code occurs at each system level time step advancement. The RELAP5 portion of the code describes the fluid transport around the system. These fluid conditions are used as thermal and mass transport boundary conditions for the SCDAP and COUPLE structures and debris beds.
Thermal hydraulic-severe accident code interfaces for SCDAP/RELAP5/MOD3.2
International Nuclear Information System (INIS)
Coryell, E.W.; Siefken, L.J.; Harvego, E.A.
1997-01-01
The SCDAP/RELAP5 computer code is designed to describe the overall reactor coolant system thermal-hydraulic response, core damage progression, and fission product release during severe accidents. The code is being developed at the Idaho National Engineering Laboratory under the primary sponsorship of the Office of Nuclear Regulatory Research of the U.S. Nuclear Regulatory Commission. The code is the result of merging the RELAP5, SCDAP, and COUPLE codes. The RELAP5 portion of the code calculates the overall reactor coolant system, thermal-hydraulics, and associated reactor system responses. The SCDAP portion of the code describes the response of the core and associated vessel structures. The COUPLE portion of the code describes response of lower plenum structures and debris and the failure of the lower head. The code uses a modular approach with the overall structure, input/output processing, and data structures following the pattern established for RELAP5. The code uses a building block approach to allow the code user to easily represent a wide variety of systems and conditions through a powerful input processor. The user can represent a wide variety of experiments or reactor designs by selecting fuel rods and other assembly structures from a range of representative core component models, and arrange them in a variety of patterns within the thermalhydraulic network. The COUPLE portion of the code uses two-dimensional representations of the lower plenum structures and debris beds. The flow of information between the different portions of the code occurs at each system level time step advancement. The RELAP5 portion of the code describes the fluid transport around the system. These fluid conditions are used as thermal and mass transport boundary conditions for the SCDAP and COUPLE structures and debris beds
Development of coupled neutronics/thermal-hydraulics test case for HPLWR
Pham, P.; Gamtsemlidze, I. D.; Bahdanovich, R. B.; Nikonov, S. P.; Smirnov, A. D.
2017-01-01
The High-Performance Light Water Reactor (HPLWR) is the European concept of a supercritical water reactor (SCWR) which is one of the most promising and innovative designs of the Generation IV nuclear reactor concepts. The thermal-hydraulics behavior of supercritical water is significantly different from water at sub-critical pressure because of the difference in the specific heat value. Coupled analysis of HPLWR assembly neutronics and thermal-hydraulics has become important because of the strong influence of the water density on the neutron spectrum and power distribution. Programs MCU (Monte-Carlo Universal) and ATHLET (Analysis of Thermal-hydraulics of Leaks and Transients) were used for better estimation of power and temperature distribution in HPLWR assembly.
Improving Neutron Kinetics and Thermal Hydraulics coupled tools for BEPU calculations
Energy Technology Data Exchange (ETDEWEB)
Pericas, R.; Reventós, F.; Batet, Il.
2015-07-01
The BEPU methodology is capable of providing a solution in terms of increasing the nuclear power production without compromising the safety margins. This study presents different improvements performed using tools available at UPC in the field of Neutron Kinetics and Thermal Hydraulics coupled systems. The paper describes a comparison between the BEPU methodology and the Conservative Bounding methodology within the framework of the Neutron Kinetics and Thermal Hydraulics coupled systems. To perform such comparison the following tools have been selected: TRACE for thermal-hydraulic system calculations, PARCS for reactor kinetics core simulator code. A Main Steam Line Break (MSLB) in a Pressurized Water Reactor (PWR) is the selected simulated transient to show the improvements performed. (Author)
Development of system analysis code for thermal-hydraulic simulation of integral reactor, Rex-10
International Nuclear Information System (INIS)
Lee, Y. G.; Kim, J. W.; Yoon, S. J.; Park, G. C.
2010-10-01
Rex-10 is an environment-friendly and economical small-scale nuclear reactor to provide the energy for district heating as well as the electric power in micro-grid. This integral reactor comprises several innovative concepts supported by advanced primary circuit components, low coolant parameters and natural circulation cooling. To evaluate the system performance and thermal-hydraulic behavior of the reactor, a system analysis code is being developed so that the new designs and technologies adopted in Rex-10 can be reflected. The research efforts are absorbed in programming the simple and fast-running thermal-hydraulic analysis software. The details of hydrodynamic governing equations component models and numerical solution scheme used in this code are presented in this paper. On the basis of one-dimensional momentum integral model, the models of point reactor neutron kinetics for thorium-fueled core, physical processes in the steam-gas pressurizer, and heat transfers in helically coiled steam generator are implemented to the system code. Implicit numerical scheme is employed to momentum and energy equations to assure the numerical stability. The accuracy of simulation is validated by applying the solution method to the Rex-10 test facility. Calculated natural circulation flow rate and coolant temperature at steady-state are compared to the experimental data. The validation is also carried out for the transients in which the sudden reduction in the core power or the feedwater flow takes place. The code's capability to predict the steady-state flow by natural convection and the qualitative behaviour of the primary system in the transients is confirmed. (Author)
Development of system analysis code for thermal-hydraulic simulation of integral reactor, Rex-10
Energy Technology Data Exchange (ETDEWEB)
NONE
2010-10-15
Rex-10 is an environment-friendly and economical small-scale nuclear reactor to provide the energy for district heating as well as the electric power in micro-grid. This integral reactor comprises several innovative concepts supported by advanced primary circuit components, low coolant parameters and natural circulation cooling. To evaluate the system performance and thermal-hydraulic behavior of the reactor, a system analysis code is being developed so that the new designs and technologies adopted in Rex-10 can be reflected. The research efforts are absorbed in programming the simple and fast-running thermal-hydraulic analysis software. The details of hydrodynamic governing equations component models and numerical solution scheme used in this code are presented in this paper. On the basis of one-dimensional momentum integral model, the models of point reactor neutron kinetics for thorium-fueled core, physical processes in the steam-gas pressurizer, and heat transfers in helically coiled steam generator are implemented to the system code. Implicit numerical scheme is employed to momentum and energy equations to assure the numerical stability. The accuracy of simulation is validated by applying the solution method to the Rex-10 test facility. Calculated natural circulation flow rate and coolant temperature at steady-state are compared to the experimental data. The validation is also carried out for the transients in which the sudden reduction in the core power or the feedwater flow takes place. The code's capability to predict the steady-state flow by natural convection and the qualitative behaviour of the primary system in the transients is confirmed. (Author)
International Nuclear Information System (INIS)
Wareing, T.A.
1988-02-01
Two potential undercooling transients are of concern in the design of TIBER-II (Tokamak Ignition/Burn Experimental Reactor), namely loss of coolant and loss of flow accidents. The major area of concern for TIBER-II is the inboard shield, where, due to tungsten material, the decay heat is extremely high. The purpose of this study was to analyze these transients using the thermal-hydraulic code ATHENA (Advanced Thermal-Hydraulic Energy Network Analyzer). The most comprehensive portion of this project involved creating a simple, yet complete, ATHENA model representative of TIBER-II. The completed model represents the case when the plasma is off and contains the inboard shield, the outboard shield, the divertor shields, and the primary loop. The primary loop contains the piping, pump, pressurizer, and heat exchanger. The heat exchanger is at the same elevation as the reactor, the least favorable to establishing natural circulation. The only transient analyzed so far, however, is a loss of flow accident. Results from the loss of flow analysis show that there is sufficient natural circulation in the inboard, outboard, and lower divertor shield to remove the decay heat, assuming that the secondary side flow is at full capacity. Although the upper divertor shield does not have sufficient natural circulation, cooling is provided due to vaporization and re-flood oscillations. However, one must recognize that there may be some local hot sport where the flow geometry inhibits cooling in a LOFA; the ATHENA model would not detect any localized problem. 9 refs., 18 figs., 4 tabs
International Nuclear Information System (INIS)
Gommlich, A.; Kliem, S.; Rohde, U.; Gomez, A.; Sanchez, V.
2010-01-01
Within the FP7 Collaborative Project NURISP (NUclear Reactor Integrated Simulation Project) new and significant steps will be done towards a European Reference Simulation Platform for applications relevant to present PWR and BWR and to future reactors. The first step towards this target has been made during the FP6 NURESIM Integrated Project, where the already common and well-proven NURESIM informatics platform has been developed. This platform is based on the open source software SALOME. The 3D neutron kinetic core model DYN3D developed at Forschungszentrum Dresden-Rossendorf is part of the NURESIM platform. Within the NURESIM project, a SALOME based pre-processor for creation of DYN3D input data sets via GUI has been developed. DYN3D has been implemented into SALOME as black box, which allowed an independent execution. A conversion of the DYN3D result file into SALOME format was developed which opened the possibility using SALOME tools to visualize DYN3D results. (orig.)
Energy Technology Data Exchange (ETDEWEB)
Kasinathan, N.; Rajakumar, A.; Vaidyanathan, G.; Chetal, S.C. [Indira Gandhi Centre for Atomic Research, Kalpakkam (India)
1995-09-01
Post shutdown decay heat removal is an important safety requirement in any nuclear system. In order to improve the reliability of this function, Liquid metal (sodium) cooled fast breeder reactors (LMFBR) are equipped with redundant hot pool dipped immersion coolers connected to natural draught air cooled heat exchangers through intermediate sodium circuits. During decay heat removal, flow through the core, immersion cooler primary side and in the intermediate sodium circuits are also through natural convection. In order to establish the viability and validate computer codes used in making predictions, a 1:20 scale experimental model called RAMONA with water as coolant has been built and experimental simulation of decay heat removal situation has been performed at KfK Karlsruhe. Results of two such experiments have been compiled and published as benchmarks. This paper brings out the results of the numerical simulation of one of the benchmark case through a 1D/2D coupled code system, DHDYN-1D/THYC-2D and the salient features of the comparisons. Brief description of the formulations of the codes are also included.
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.
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)
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.
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)
Implementation of CFD module in the KORSAR thermal-hydraulic system code
Energy Technology Data Exchange (ETDEWEB)
Yudov, Yury V.; Danilov, Ilia G.; Chepilko, Stepan S. [Alexandrov Research Inst. of Technology (NITI), Sosnovy Bor (Russian Federation)
2015-09-15
The Russian KORSAR/GP (hereinafter KORSAR) computer code was developed by a joint team from Alexandrov NITI and OKB ''Gidropress'' for VVER safety analysis and certified by the Rostechnadzor of Russia in 2009. The code functionality is based on a 1D two-fluid model for calculation of two-phase flows. A 3D CFD module in the KORSAR computer code is being developed by Alexandrov NITI for representing 3D effects in the downcomer and lower plenum during asymmetrical loop operation. The CFD module uses Cartesian grid method with cut cell approach. The paper presents a numerical algorithm for coupling 1D and 3D thermal- hydraulic modules in the KORSAR code. The combined pressure field is calculated by the multigrid method. The performance efficiency of the algorithm for coupling 1D and 3D modules was demonstrated by solving the benchmark problem of mixing cold and hot flows in a T-junction.
Energy Technology Data Exchange (ETDEWEB)
Fajardo, Laura Garcia, E-mail: laura.gf@cern.ch [European Organization for Nuclear Research (CERN), Geneva (Switzerland). Technology Department; Hernandez, Carlos Garcia; Mazaira, Leorlen Rojas, E-mail: cgh@instec.cu, E-mail: irojas@instec.cu [Higher Institute of Technologies and Applied Sciences (INSTEC), Habana (Cuba); Castells, Facundo Alberto Escriva, E-mail: aescriva@iqn.upv.es [University of Valencia (UV), Valencia (Spain). Energetic Engineering Institute; Lira, Carlos Brayner de Olivera, E-mail: cabol@ufpe.br [Universidade Federal de Pernambuco (UFPE), Recife, PE (BRazil). Dept. de Engenharia Nuclear
2013-07-01
The Transmutation Advanced Device for Sustainable Energy Applications (TADSEA) is a pebble-bed Accelerator Driven System (ADS) with a graphite-gas configuration, designed for nuclear waste trans- mutation and for obtaining heat at very high temperatures to produce hydrogen. In previous work, the TADSEA's nuclear core was considered as a porous medium performed with a CFD code and thermal-hydraulic studies of the nuclear core were presented. In this paper, the heat transfer from the fuel to the coolant was analyzed for three core states during normal operation. The heat transfer inside the spherical fuel elements was also studied. Three critical fuel elements groups were defined regarding their position inside the core. Results were compared with a realistic CFD model of the critical fuel elements groups. During the steady state, no critical elements reached the limit temperature of this type of fuel. (author)
Feasibility study for objective oriented design of system thermal hydraulic analysis program
International Nuclear Information System (INIS)
Chung, Bub Dong; Jeong, Jae Jun; Hwang, Moon Kyu
2008-01-01
The system safety analysis code, such as RELAP5, TRAC, CATHARE etc. have been developed based on Fortran language during the past few decades. Refactoring of conventional codes has been also performed to improve code readability and maintenance. However the programming paradigm in software technology has been changed to use objects oriented programming (OOP), which is based on several techniques, including encapsulation, modularity, polymorphism, and inheritance. In this work, objective oriented program for system safety analysis code has been tried utilizing modernized C language. The analysis, design, implementation and verification steps for OOP system code development are described with some implementation examples. The system code SYSTF based on three-fluid thermal hydraulic solver has been developed by OOP design. The verifications of feasibility are performed with simple fundamental problems and plant models. (author)
Coupling the severe accident code SCDAP with the system thermal hydraulic code MARS
Energy Technology Data Exchange (ETDEWEB)
Lee, Young Jin; Chung, Bub Dong [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)
2004-07-01
MARS is a best-estimate system thermal hydraulics code with multi-dimensional modeling capability. One of the aims in MARS code development is to make it a multi-functional code system with the analysis capability to cover the entire accident spectrum. For this purpose, MARS code has been coupled with a number of other specialized codes such as CONTEMPT for containment analysis, and MASTER for 3-dimensional kinetics. And in this study, the SCDAP code has been coupled with MARS to endow the MARS code system with severe accident analysis capability. With the SCDAP, MARS code system now has acquired the capability to simulate such severe accident related phenomena as cladding oxidation, melting and slumping of fuel and reactor structures.
Coupling the severe accident code SCDAP with the system thermal hydraulic code MARS
International Nuclear Information System (INIS)
Lee, Young Jin; Chung, Bub Dong
2004-01-01
MARS is a best-estimate system thermal hydraulics code with multi-dimensional modeling capability. One of the aims in MARS code development is to make it a multi-functional code system with the analysis capability to cover the entire accident spectrum. For this purpose, MARS code has been coupled with a number of other specialized codes such as CONTEMPT for containment analysis, and MASTER for 3-dimensional kinetics. And in this study, the SCDAP code has been coupled with MARS to endow the MARS code system with severe accident analysis capability. With the SCDAP, MARS code system now has acquired the capability to simulate such severe accident related phenomena as cladding oxidation, melting and slumping of fuel and reactor structures
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)
Thermal-hydraulic and neutronic analysis of pressurized water reactor cores
International Nuclear Information System (INIS)
Alves, C.H.
1982-01-01
A computational code, named CANAL2, was developed for the simulation of the steady-state and transient behaviour of a Pressurized Water Reactor core. The conservation equations for the control volumes are obtained by area-averaging of the two-fluid model conservation equations and reducing them to the drift-flux model formulation. The resulting equations are aproximated by finite differences and solved by a marching-type numerical scheme. The model takes into account the exchange of mass, momentum and energy between adjacent subchannels of a fuel bundle. Turbulent mixing and diversion crossflow are considered. Correlations are provided for several heat trans and flow regimes and selected according to the local conditons. During transients core power can be evaluated by a point-Kinetics model. Fuel and coolant temperatures are feedback to the neutronics. The heat conduction equation is solved in the fuel using the Crank-Nicolson scheme. Temperature-dependent correlations are provided for the fuel and cladding thermal conductivities. Several runs were made with the code CANAL2 using the available experimental and calculated data in the open literature. Results indicate that CANAL2 is a good calculational tool for the thermal-hydraulics of PWR cores. A few refinements will make the code useful for design. (Author) [pt
Energy Technology Data Exchange (ETDEWEB)
Chung, Bub Dong; Lee, Won Jae; Kim, Hee Cheol; Song, Jin Ho; Sim, Suk Ku [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)
1997-12-31
The work reported in this paper identifies the thermal-hydraulic phenomena that are expected to occur during a number of key transients in SMART (System-integrated Modular Advanced ReacTor) which is under development at KAERI. The result of this effort is based on the current design concept of SMART integral reactor. Although the design is still evolving, the preliminary Phenomena Identification and Ranking Table (PIRT) has been developed based on the experts` knowledge and experience. The preliminary PIRT has been developed by consensus of KAERI expert panelists and AHP (Analytical Hierarchy Process). Preliminary PIRT developed in this paper is intended to be used to identify and integrate development areas of further experimental tests needed, thermal hydraulic models and correlations and code improvements for the safety analysis of the SMART. 8 refs., 4 tabs (Author)
Cheng, Xiaoman; Ma, Xuebin; Jiang, Kecheng; Chen, Lei; Huang, Kai; Liu, Songlin
2015-09-01
The water-cooled ceramic breeder blanket (WCCB) is one of the blanket candidates for China fusion engineering test reactor (CFETR). In order to improve power generation efficiency and tritium breeding ratio, WCCB with superheated steam is under development. The thermal-hydraulic design is the key to achieve the purpose of safe heat removal and efficient power generation under normal and partial loading operation conditions. In this paper, the coolant flow scheme was designed and one self-developed analytical program was developed, based on a theoretical heat transfer model and empirical correlations. Employing this program, the design and analysis of related thermal-hydraulic parameters were performed under different fusion power conditions. The results indicated that the superheated steam water-cooled blanket is feasible. supported by the National Special Project for Magnetic Confined Nuclear Fusion Energy of China (Nos. 2013GB108004, 2014GB122000 and 2014GB119000), and National Natural Science Foundation of China (No. 11175207)
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)
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)
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.)
Current and anticipated uses of the thermal hydraulics codes at the NRC
International Nuclear Information System (INIS)
Caruso, R.
1997-01-01
The focus of Thermal-Hydraulic computer code usage in nuclear regulatory organizations has undergone a considerable shift since the codes were originally conceived. Less work is being done in the area of Design Basis Accidents, , and much more emphasis is being placed on analysis of operational events, probabalistic risk/safety assessment, and maintenance practices. All of these areas need support from Thermal-Hydraulic computer codes to model the behavior of plant fluid systems, and they all need the ability to perform large numbers of analyses quickly. It is therefore important for the T/H codes of the future to be able to support these needs, by providing robust, easy-to-use, tools that produce easy-to understand results for a wider community of nuclear professionals. These tools need to take advantage of the great advances that have occurred recently in computer software, by providing users with graphical user interfaces for both input and output. In addition, reduced costs of computer memory and other hardware have removed the need for excessively complex data structures and numerical schemes, which make the codes more difficult and expensive to modify, maintain, and debug, and which increase problem run-times. Future versions of the T/H codes should also be structured in a modular fashion, to allow for the easy incorporation of new correlations, models, or features, and to simplify maintenance and testing. Finally, it is important that future T/H code developers work closely with the code user community, to ensure that the code meet the needs of those users
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)
Interface requirements to couple thermal-hydraulic codes to severe accident codes: ATHLET-CD
Energy Technology Data Exchange (ETDEWEB)
Trambauer, K. [GRS, Garching (Germany)
1997-07-01
The system code ATHLET-CD is being developed by GRS in cooperation with IKE and IPSN. Its field of application comprises the whole spectrum of leaks and large breaks, as well as operational and abnormal transients for LWRs and VVERs. At present the analyses cover the in-vessel thermal-hydraulics, the early phases of core degradation, as well as fission products and aerosol release from the core and their transport in the Reactor Coolant System. The aim of the code development is to extend the simulation of core degradation up to failure of the reactor pressure vessel and to cover all physically reasonable accident sequences for western and eastern LWRs including RMBKs. The ATHLET-CD structure is highly modular in order to include a manifold spectrum of models and to offer an optimum basis for further development. The code consists of four general modules to describe the reactor coolant system thermal-hydraulics, the core degradation, the fission product core release, and fission product and aerosol transport. Each general module consists of some basic modules which correspond to the process to be simulated or to its specific purpose. Besides the code structure based on the physical modelling, the code follows four strictly separated steps during the course of a calculation: (1) input of structure, geometrical data, initial and boundary condition, (2) initialization of derived quantities, (3) steady state calculation or input of restart data, and (4) transient calculation. In this paper, the transient solution method is briefly presented and the coupling methods are discussed. Three aspects have to be considered for the coupling of different modules in one code system. First is the conservation of masses and energy in the different subsystems as there are fluid, structures, and fission products and aerosols. Second is the convergence of the numerical solution and stability of the calculation. The third aspect is related to the code performance, and running time.
Current and anticipated uses of the thermal hydraulics codes at the NRC
Energy Technology Data Exchange (ETDEWEB)
Caruso, R.
1997-07-01
The focus of Thermal-Hydraulic computer code usage in nuclear regulatory organizations has undergone a considerable shift since the codes were originally conceived. Less work is being done in the area of {open_quotes}Design Basis Accidents,{close_quotes}, and much more emphasis is being placed on analysis of operational events, probabalistic risk/safety assessment, and maintenance practices. All of these areas need support from Thermal-Hydraulic computer codes to model the behavior of plant fluid systems, and they all need the ability to perform large numbers of analyses quickly. It is therefore important for the T/H codes of the future to be able to support these needs, by providing robust, easy-to-use, tools that produce easy-to understand results for a wider community of nuclear professionals. These tools need to take advantage of the great advances that have occurred recently in computer software, by providing users with graphical user interfaces for both input and output. In addition, reduced costs of computer memory and other hardware have removed the need for excessively complex data structures and numerical schemes, which make the codes more difficult and expensive to modify, maintain, and debug, and which increase problem run-times. Future versions of the T/H codes should also be structured in a modular fashion, to allow for the easy incorporation of new correlations, models, or features, and to simplify maintenance and testing. Finally, it is important that future T/H code developers work closely with the code user community, to ensure that the code meet the needs of those users.
International Nuclear Information System (INIS)
Wu Xiaohang; Fu Xiaohua
2004-01-01
The electrically heated rod is usually used as a substitute for fuel rod in thermal-hydraulic experiment. However, the different structure and thermal properties between nuclear fuel rod and electrically heated rod result in different steady-state distribution of temperature and stored energy and different response to thermal-hydraulic in simulation transient experiment. This paper analyses the effect of structure and thermal properties differences between nuclear fuel rod and electrically heated rod on experiment, and then introduce a feasible method, i.e. electric power is controlled by a program, to reduce the differences between the transient responses of nuclear fuel rod and electrically heated rod. At the same time, this paper points out the limits of the method. (authors)
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
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
An analytical model for hydraulic fracturing in shallow bedrock formations.
dos Santos, José Sérgio; Ballestero, Thomas Paul; Pitombeira, Ernesto da Silva
2011-01-01
A theoretical method is proposed to estimate post-fracturing fracture size and transmissivity, and as a test of the methodology, data collected from two wells were used for verification. This method can be employed before hydrofracturing in order to obtain estimates of the potential hydraulic benefits of hydraulic fracturing. Five different pumping test analysis methods were used to evaluate the well hydraulic data. The most effective methods were the Papadopulos-Cooper model (1967), which includes wellbore storage effects, and the Gringarten-Ramey model (1974), known as the single horizontal fracture model. The hydraulic parameters resulting from fitting these models to the field data revealed that as a result of hydraulic fracturing, the transmissivity increased more than 46 times in one well and increased 285 times in the other well. The model developed by dos Santos (2008), which considers horizontal radial fracture propagation from the hydraulically fractured well, was used to estimate potential fracture geometry after hydrofracturing. For the two studied wells, their fractures could have propagated to distances of almost 175 m or more and developed maximum apertures of about 2.20 mm and hydraulic apertures close to 0.30 mm. Fracturing at this site appears to have expanded and propagated existing fractures and not created new fractures. Hydraulic apertures calculated from pumping test analyses closely matched the results obtained from the hydraulic fracturing model. As a result of this model, post-fracturing geometry and resulting post-fracturing well yield can be estimated before the actual hydrofracturing. Copyright © 2010 The Author(s). Journal compilation © 2010 National Ground Water Association.
Thermo-hydraulic-mechanical analysis of the SS-050 sodium loop during a thermal shock of 2000C/s
International Nuclear Information System (INIS)
Jesus Miranda, C.A. de; Gebrin, A.N.
1988-01-01
An analytical thermo-hydraulic model was developed to obtain the temperature of the sodium flowing between the mixing tank TM of constant volume and the drain tank of the SS-050 sodium test facility. The piping connecting these two tanks is considered in the analysis. The sodium enters in the TM through a tube with lateral holes immersed in the TM's sodium. The model and relative computer program were tested and a typical situation was studied: a thermal shock with -200 0 C/s of thermal gradient in the test section. The sodium temperature time-histories along the piping length are presented. For the thermal shock situation, the temperature field in the TM bottom and outlet nozzle was calculated and the stresses were evaluated. The final thermal stresses will allow a detailed verification of the circuit design. (author) [pt
Engineering and thermal-hydraulics design of PFC cooling for SST-1 Tokamak
International Nuclear Information System (INIS)
Chaudhuri, Paritosh; Reddy, D. Chenna; Santra, P.; Khiwadkar, S.; Prakash, N. Rabi; Ramash, G.; Dubey, Santosh; Prakash, Arun; Saxena, Y. C.
2003-01-01
The main consideration in the design of the PFC cooling for SST-1 tokamak is the steady state heat removal of upto 1MW/m2. The PFC also has been design to withstand the peak heat fluxes without significant erosion such that frequent replacement is not necessary. Proper brazing of cooling tube on the copper back plate is necessary for the efficient heat transfer from the tube to the back plate. 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 conduct the heat transfer and stress analyses of the PFC to understand its thermal and mechanical behaviors. The temperature distribution results for different PFC obtained by FE results were assessed by comparison with 2-D Finite Difference code. 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. The contact at the brazed joint of the tube to the backplate 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 is studied under various flow parameters. These results of this study will be presented in details in this paper
International Nuclear Information System (INIS)
Park, Youngjae; Kim, Iljin; Kim, Hyungdae
2015-01-01
Diverse integral/small-modular reactors (SMRs) have been developed. Once-through steam generator (OTSG) which generates superheated steam without steam separator and dryer was used in the SMRs to reduce volume of steam generator. It would be possible to design a new steam generator with best estimate thermal-hydraulic codes such as RELAP and MARS. However, it is not convenience to use the general purpose thermal-hydraulic analysis code to design a specific component of nuclear power plants. A widely used simulation tool for thermal-hydraulic analysis of drum-type steam generators is ATHOS, which allows 3D analysis. On the other hand, a simple 1D thermal-hydraulic analysis code might be accurate enough for the conceptual design of OTSG. In this study, thermal-hydraulic analysis code for conceptual design of OTSG was developed using 1D homogeneous equilibrium model (HEM). A benchmark calculation was also conducted to verify and validate the prediction accuracy of the developed code by comparing with the analysis results with MARS. Finally, conceptual design of OTSG was conducted by the developed code. A simple 1D thermal-hydraulic analysis code was developed for the purpose of conceptual design OTSG for SMRs. A set of benchmark calculations was conducted to verify and validate the analysis accuracy of the developed code by comparing results obtained with a best-estimated thermal-hydraulic analysis code, MARS. Finally, analysis of two different OTSG design concepts with superheating and recirculation was demonstrated using the developed code
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
Investigating hydraulic transport and disposal of coal ash at the Gacko thermal power plant
Energy Technology Data Exchange (ETDEWEB)
Knezevic, D.; Grbovic, M.; Petrovic, M.
1986-01-01
This paper discusses ash transport difficulties at the Gacko thermal power station. Designed dumper transport failed due to violent thermal reactions in water-sprayed ash during transport. An system was designed by the Institute for Ore Processing of Belgrade. Large-scale investigation of ash properties and slurry consolidation were conducted prior to hydraulic transport testing. A semi-industrial hydraulic transport was built and tested for ash disposal. It was found that Gacko power station ash may be safely transported by pipeline and disposed in layers 10 cm thick without danger of operation breaks due to ash caking within the pipeline. A sketch of the hydraulic transport system is presented. 4 refs.
Study on thermal-hydraulic phenomena in porous media. Semiannual report 1997. Nov. to 1998. Mar
International Nuclear Information System (INIS)
Matui, Goichi; Monji, Hideaki; Sakakibara, Jun; Tanaka, Masa-aki; Kobayashi, Jun; Kamide, Hideki
1998-03-01
The objective of this study is to clarify the thermal-hydraulic phenomena in porous media and to develop the analytical method to predict the thermal-hydraulic field, deciding the maximum temperature on the fuel pin surface. FY 1998 is the first year of the 3 years plan. In this year period, based on the correspondence of Reynolds number between experimental facility and FBR, the design and construction of the test rig and experimental parameter examination were performed. The test section has rectangle two sub-channel geometry and is twenty times large-scale model. In this study, we use the Particle Image Velocimetry (PIV) analysis method to visualize the flow field in the porous media. The Pyrex grass spheres were used to construct the porous blockage. The refraction-rate matching between obstacle and fluid is important to measure the velocity field with the optical analysis method. As the working fluid, NaI solution was used. When the concentration of NaI is 56.4wt% in the solution, the refraction-rate is correspond to that of the Pyrex grass. The simple test loop was constructed and the experiment was performed to measure the velocity field with Laser Doppler Anemometer and PIV. The purpose of this experiment using the simple test loop is to develop the experimental method of flow visualization in the porous media used NaI solution as working fluid. As the result of experiment with the simple test loop, the vector field in the porous media was obtained and it is shown that the flow pattern with PIV analysis is qualitatively correct. This visualization method using the NaI solution is applicable to measure the flow field in the porous media. (J.P.N.)
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
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)
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 ...
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
Study on relationship between aging and thermal-hydraulic behaviors of nuclear power plants
International Nuclear Information System (INIS)
Murata, Hiroyuki; Inasaka, Fujio; Adachi, Masaki; Sawada, Ken-ichi; Akiyama, Shigeru; Sakuma, Masaaki; Takahashi, Ichihiko; Ushijima, Michio
2005-01-01
The number of aged nuclear power plants will increase in the future, because operation periods of the existing nuclear power plants are being extended from thirty years of initial supposition to sixty years at the longest. Therefore, it is important to establish the methodology to guarantee integrity of the aged nuclear power plants. Among the reported damages of nuclear power plant components due to the fatigue during long terms, many cases are considered to be related to their thermal-hydraulic behaviors during operation. Thus, quantitative understanding of thermal-hydraulic behaviors in the nuclear power plants is important to estimate many kinds of aging processes accurately. The research project, 'study on relationship between nuclear power plant aging and its thermal-hydraulic behaviors', was conducted from 2001 to 2004 in order to clarify effects of thermal-hydraulic behaviors in the nuclear power plants on structural materials during aging processes. In this project, flow induced vibration of an array of circular cylinders was investigated experimentally and numerically. Rotating bending fatigue tests were also performed for the austenitic stainless steel SUS316L (JIS G4304) and Ni-Cr-Fe alloy NCF690 (JIS G4904, INCONEL alloy 690 equivalent material) in order to examine the fatigue strength in the ultra high cycle fatigue region, namely 10 7 -10 9 cycles, and the notch effects. (author)
A linear stability analysis of supercritical water reactors, (1). Thermal-hydraulic stability
International Nuclear Information System (INIS)
Tin Tin Yi; Koshizuka, Seiichi; Oka, Yoshiaki
2004-01-01
This paper summarizes the analysis results of the thermal-hydraulic stability of a high-temperature reactor cooled and moderated by supercritical-pressure light water (SCLWR-H). A linear stability analysis code in the frequency domain was developed to study the thermal-hydraulic stability of SCLWR-H at constant supercritical pressure. The analysis method is based on linearization by perturbation of numerically-discretized one-dimensional single-channel single-phase conservation equations. The effect of water rods on stability is considered. The thermal-hydraulic stability of SCLWR-H for full-power and partial-power normal operations was investigated by frequency domain method. Our analysis reveals that though SCLWR-H has low coolant flow rate and large density change in the core, the thermal-hydraulic stability can be maintained both at normal operation and during power raising phase of constant pressure startup by applying an orifice pressure drop coefficient an the inlet of the fuel assemblies. A parametric study was also carried out to determine the parameters affecting the stability. (author)
International Nuclear Information System (INIS)
Falikov, A.A.; Vakhrushev, V.V.; Kuul, V.S.; Samoilov, O.B.; Tarasov, G.I.
1997-01-01
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
International Nuclear Information System (INIS)
Kuriyama, Minoru; Morishima, Atsuyoshi
1975-02-01
The thermal-hydraulic data of commercial nuclear power plants in the United States are presented for 85 PWRs and 52 BWRs. Covered are the temperature, pressure, heat flux, flow rate, etc. of coolant and/or fuel cladding; the diffinitions are also given for some of these items. (auth.)
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)
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 ...
A review on the thermal hydraulic characteristics of the air-cooled ...
Indian Academy of Sciences (India)
Abstract. 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 ...
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
Energy Technology Data Exchange (ETDEWEB)
Jaeger, W.; Sanchez, V.; Cheng, X. [Karlsruhe Institute of Technology (KIT), Eggenstein-Leopoldshafen (Germany). Inst. for Neutron Physics and Reactor Technology; Monti, L. [Karlsruhe Institute of Technology (KIT), Eggenstein-Leopoldshafen (Germany). Inst. for Nuclear and Energy Technologies; Hurtado, A. [Technical Univ. of Dresden (Germany). Inst. of Power Engineering
2011-07-01
At the Institute for Neutron Physics and Reactor Technology (INR) at the Karlsruhe Institute of Technology (KIT), the development and validation of coupled codes systems is one major activity. In this paper, a 2-step method is proposed to perform uncertainty and sensitivity analysis of a nuclear fuel bundle. At first, the SUSA package (Software system for Uncertainty and Sensitivity Analysis), 2 is applied to the thermal hydraulic results of the TRACE (TRACE/RELAP Advanced Computational Engine) code to identify crucial thermal hydraulic parameter combinations which are successively used in the TH/NP coupled system TRACEERANOS to account for the neutronic feedbacks. This 2-step method was applied since the TRACE-ERANOS system runs 1 input in approximately 1 day (depending on the computer configurations). Since the uncertainty and sensitivity analysis requires about 100 runs of the thermal hydraulic input (with altered parameters, running within minutes) an integral TRACE-SUSA-ERANOS analysis would need around 100 days. For this analysis a fuel assembly model of the HPLWR (High Performance Light Water Reactor) was selected. Due to the general structure of the coupling and code communication scripts, the system can be used for any kind of reactor/system which can be described with TRACE and ERANOS (e.g., fast systems) while SUSA can be applied to anything. (orig.)
International Nuclear Information System (INIS)
Umminger, K.; Liebert, J.; Kastner, W.
1997-01-01
Two complementary test facilities - the Upper Plenum Test Facility (UPTF) and the Primaerkreislauf test facility (PKL) - were constructed to investigate the thermal-hydraulic response of a pressurized water reactor (PWR) during postulated accidents. The UPTF is a geometrical full-scale simulation of the primary system of a 1300-MW PWR. The upper plenum, the downcomer and the four connected loops as well as the pressurizer are represented on a 1:1 scale. The integral test facility PKL also simulates a 1300-MW PWR, whereby the power and volume is reduced by a factor of 1:145 (elevations 1:1). The PKL test facility models the entire primary system, relevant parts of the secondary side and all important engineered safety and auxiliary systems. Whereas the UPTF was mainly designed to perform separate-effect tests focusing on multidimensional thermal-hydraulic phenomena in full-scale simulated components, the main objective of the PKL tests has been the investigation of the thermal-hydraulic system behavior on the primary and secondary side. So far the program objectives represent a reasonable completion and in summary the experimental results from both test facilities provide an essential contribution for a better understanding of assumed accident sequences in a PWR. Test results which demonstrate the complementary character of the UPTF and the PKL test programs as well as the interaction between the two test facilities are presented in this paper. (author)
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)
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
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)
Energy Technology Data Exchange (ETDEWEB)
Matti, B. [Muttoni e Beffa SA, Faido (Switzerland); Tacher, L. [GEOLEP, Lausanne (Switzerland)
2009-07-01
This article describes the three-dimensional hydrogeological and hydro-thermal models used in the study and realisation of an additional, big scale water-water refrigeration system for the hospital in the city of Sion, Switzerland. This study answers the following questions: Will the introduction of new installations disturb the operation of existing systems? Can the aquifer deliver the required flow rate? Will the performance of the water well and the existing heat pump be modified? Hydraulically, numerical models have shown the great importance of the boundary conditions in the dynamics of the system (e.g. the Rhone river and the relation atmosphere / aquifer). From the thermal viewpoint, the heat transfer from the surface through the unsaturated zone controls this thermal heat system. Seasonal atmospheric fluctuations may clearly be noted (more than half of the heat flowing in and out). Numerical simulations show that the existing wells, operating mainly in summer for cooling, have no significant thermal influence on the aquifer (1% of annual heat recharge). These existing installations modify its water temperature only downstream within a distance of maximum 200 m. Hence, the minimum recommended distance for any new geothermal installation would be 200 m. From a hydraulic and thermal viewpoint, it seems that new installations using the aquifer can be added without any restriction. (author)
International Nuclear Information System (INIS)
Boergesson, L.; Hernelind, J.
1995-12-01
PNCs large scale laboratory test with an artificial deposition hole has been simulated with finite element calculations with the code ABAQUS. The test comprised water uptake from an artificial rock and heating of a canister in a deposition hole with the diameter 1 m during 5 months. Water content, pore pressure, and total pressure in the buffer was measured during the test. The given data of the material properties were supplemented with results from own laboratory tests in order to determine parameters required for the calculation. The vapour flow process, which is not included in ABAQUS, was implemented and added to the code. After calibration of the properties of the buffer material, a completely coupled thermo-hydro-mechanical calculation of the test was done. The calculated thermal and hydraulic results were in good agreement with measured values, while the prediction of the mechanical response was less good. 13 refs, 47 figs, 8 tabs
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.
Energy Technology Data Exchange (ETDEWEB)
Lee, Won Jae; Chung, Bub Dong; Jeong, Jae Jun; Ha, Kwi Seok [Korea Atomic Energy Research Institute, Taejon (Korea)
1998-06-01
A multi-dimensional realistic thermal-hydraulic system analysis code, MARS version 1.3 has been developed. Main purpose of MARS 1.3 development is to have the realistic analysis capability of transient two-phase thermal-hydraulics of Pressurized Water Reactors (PWRs) especially during Large Break Loss of Coolant Accidents (LBLOCAs) where the multi-dimensional phenomena domain the transients. MARS code is a unified version of USNRC developed COBRA-TF, domain the transients. MARS code is a unified version of USNRC developed COBRA-TF, three-dimensional (3D) reactor vessel analysis code, and RELAP5/MOD3.2.1.2, one-dimensional (1D) reactor system analysis code., Developmental requirements for MARS are chosen not only to best utilize the existing capability of the codes but also to have the enhanced capability in code maintenance, user accessibility, user friendliness, code portability, code readability, and code flexibility. For the maintenance of existing codes capability and the enhancement of code maintenance capability, user accessibility and user friendliness, MARS has been unified to be a single code consisting of 1D module (RELAP5) and 3D module (COBRA-TF). This is realized by implicitly integrating the system pressure matrix equations of hydrodynamic models and solving them simultaneously, by modifying the 1D/3D calculation sequence operable under a single Central Processor Unit (CPU) and by unifying the input structure and the light water property routines of both modules. In addition, the code structure of 1D module is completely restructured using the modular data structure of standard FORTRAN 90, which greatly improves the code maintenance capability, readability and portability. For the code flexibility, a dynamic memory management scheme is applied in both modules. MARS 1.3 now runs on PC/Windows and HP/UNIX platforms having a single CPU, and users have the options to select the 3D module to model the 3D thermal-hydraulics in the reactor vessel or other
Modeling Hydraulic Components for Automated FMEA of a Braking System
2014-12-23
causes the diminishing of the wheel brake pressure if the brake pedal is released. When operated under the Anti - lock - braking system (ABS), the valves...Modeling Hydraulic Components for Automated FMEA of a Braking System Peter Struss, Alessandro Fraracci Tech. Univ. of Munich, 85748 Garching...the hydraulic part of a vehicle braking system . We describe the FMEA task and the application problem and outline the foundations for automating the
The relevance of thermal hydraulics pipeline simulation as a regulatory support tool
Energy Technology Data Exchange (ETDEWEB)
Silva, Patricia Mannarino; Santos, Almir Beserra dos [Agencia Nacional do Petroleo, Gas Natural e Biocombustiveis (ANP), Rio de Janeiro, RJ (Brazil)
2009-07-01
The capacity definition of a pipeline, along with its allocation, is very relevant to assure market transparency, nondiscriminatory access, security of supply, and also to give consistent signs for expansion needs. Nevertheless, the capacity definition is a controversial issue, and may widely vary depending on the technical and commercial assumptions made. To calculate a pipeline's nominal capacity, there are a variety of simulation tools, which include steady state, transient and on-line computer programs. It is desirable that the simulation tool is robust enough to predict the pipeline's capacity under different conditions. There are many variables that impact the flow through a pipeline, like gas characteristics, pipe and environmental variables. Designing a thermal model is a time-consuming task that requests understanding the level of detail need, in order to achieve success in its application. This article discusses the capacity definition, its role and calculation guidelines, describes ANP's experience with capacity calculation and further challenges according to the new regulation, and debates the role of thermal hydraulic simulation as a regulatory tool. (author)
Review of thermal-hydraulic calculations for Calvert Cliffs and H.B. Robinson PTS study
International Nuclear Information System (INIS)
Jo, J.H.; Yuelys-Miksis, C.; Rohatgi, U.S.
1984-01-01
Thermal-hydraulic transient calculations performed by LANL using the TRAC-PF1 code and by INEL using the RELAP5 code for the USNRC pressurized thermal shock (PTS) study of the Calvert Cliffs and H.B. Robinson Nuclear Power Plants have been reviewed at BNL including the input decks and steady state calculations. Furthermore, six transients for each plant have been selected for the in-depth review. Simple hand calculations based on the mass and energy balances of the entire reactor system, have been performed to predict the temperature and pressure of the reactor system, and the results have been compared with those obtained by the code calculation. In general, the temperatures and pressures of the primary system calculated by the codes have been very reasonable. The secondary pressures calculated by TRAC appear to indicate that the codes have some difficulty with the condensation model and further work is needed to assess the code calculation of the U-tube steam generator pressure when the cold auxiliary feedwater is introduced to the steam generator. However, it is not expected that this uncertainty would affect the transient calculations significantly
Design of a Resistively Heated Thermal Hydraulic Simulator for Nuclear Rocket Reactor Cores
Litchford, Ron J.; Foote, John P.; Ramachandran, Narayanan; Wang, Ten-See; Anghaie, Samim
2007-01-01
A preliminary design study is presented for a non-nuclear test facility which uses ohmic heating to replicate the thermal hydraulic characteristics of solid core nuclear reactor fuel element passages. The basis for this testing capability is a recently commissioned nuclear thermal rocket environments simulator, which uses a high-power, multi-gas, wall-stabilized constricted arc-heater to produce high-temperature pressurized hydrogen flows representative of reactor core environments, excepting radiation effects. Initially, the baseline test fixture for this non-nuclear environments simulator was configured for long duration hot hydrogen exposure of small cylindrical material specimens as a low cost means of evaluating material compatibility. It became evident, however, that additional functionality enhancements were needed to permit a critical examination of thermal hydraulic effects in fuel element passages. Thus, a design configuration was conceived whereby a short tubular material specimen, representing a fuel element passage segment, is surrounded by a backside resistive tungsten heater element and mounted within a self-contained module that inserts directly into the baseline test fixture assembly. With this configuration, it becomes possible to create an inward directed radial thermal gradient within the tubular material specimen such that the wall-to-gas heat flux characteristics of a typical fuel element passage are effectively simulated. The results of a preliminary engineering study for this innovative concept are fully summarized, including high-fidelity multi-physics thermal hydraulic simulations and detailed design features.
Energy Technology Data Exchange (ETDEWEB)
An, P.; Yao, D. [Science and Tech. on Reactor System Design Tech. Laboratory, Chengdu (China)
2011-07-01
The MCATHAS system of coupled neutronics/Thermal-hydraulics in supercritical water reactor is described, which considers the mutual influence between the obvious axial and radial evolution of material temperature, water density and the relative power distribution. This system can obtain the main neutronics and thermal parameters along with burn-up. MCATHAS system is parallel processing coupling. The MCNP code is used for neutronics analysis with the continuous cross section library at any temperature calculated by interpolation algorithm; The sub-channel code ATHAS is for thermal-hydraulics analysis and the ORIGEN Code for burn-up calculation. We validate the code with the assembly of HPLWR and analyze the assembly SCLWR- H. (author)
Influence of Elevation Data Source on 2D Hydraulic Modelling
Bakuła, Krzysztof; StĘpnik, Mateusz; Kurczyński, Zdzisław
2016-08-01
The aim of this paper is to analyse the influence of the source of various elevation data on hydraulic modelling in open channels. In the research, digital terrain models from different datasets were evaluated and used in two-dimensional hydraulic models. The following aerial and satellite elevation data were used to create the representation of terrain-digital terrain model: airborne laser scanning, image matching, elevation data collected in the LPIS, EuroDEM, and ASTER GDEM. From the results of five 2D hydrodynamic models with different input elevation data, the maximum depth and flow velocity of water were derived and compared with the results of the most accurate ALS data. For such an analysis a statistical evaluation and differences between hydraulic modelling results were prepared. The presented research proved the importance of the quality of elevation data in hydraulic modelling and showed that only ALS and photogrammetric data can be the most reliable elevation data source in accurate 2D hydraulic modelling.
Real-time dynamic hydraulic model for water distribution networks: steady state modelling
CSIR Research Space (South Africa)
Osman, Mohammad S
2016-09-01
Full Text Available steady state hydraulic model that will be used within a real-time dynamic hydraulic model (DHM). The Council for Scientific and Industrial Research (CSIR) water distribution network (WDN) is used as a pilot study for this purpose. A hydraulic analysis...
Summary of papers on current and anticipated uses of thermal-hydraulic codes
Energy Technology Data Exchange (ETDEWEB)
Caruso, R.
1997-07-01
The author reviews a range of recent papers which discuss possible uses and future development needs for thermal/hydraulic codes in the nuclear industry. From this review, eight common recommendations are extracted. They are: improve the user interface so that more people can use the code, so that models are easier and less expensive to prepare and maintain, and so that the results are scrutable; design the code so that it can easily be coupled to other codes, such as core physics, containment, fission product behaviour during severe accidents; improve the numerical methods to make the code more robust and especially faster running, particularly for low pressure transients; ensure that future code development includes assessment of code uncertainties as integral part of code verification and validation; provide extensive user guidelines or structure the code so that the `user effect` is minimized; include the capability to model multiple fluids (gas and liquid phase); design the code in a modular fashion so that new models can be added easily; provide the ability to include detailed or simplified component models; build on work previously done with other codes (RETRAN, RELAP, TRAC, CATHARE) and other code validation efforts (CSAU, CSNI SET and IET matrices).
Development of best estimate auditing code for CANDU thermal-hydraulic safety analysis
Energy Technology Data Exchange (ETDEWEB)
Chung, Bub Dong; Lee, Won Jae; Hwang, Moon Kyu; Lim, Hong Sik [Korea Atomic Energy Research Institute, Taejeon (Korea)
2002-04-01
The main purpose of this study is to develop a thermal hydraulic auditing code for the CANDU reactor, modifying the model of existing PWR auditing tool, i.e. RELAP5/MOD3.The study was performed by reconsideration of the previous code assessment works and phenomena identification for essential accident scenario. Improvement areas of model development for auditing tool were identified based on the code comparison and PIRT results. Nine models have been improved significantly for the analysis of LOCA and Mon LOCA event. Conceptual problem or separate effect assessment have been performed to verify the model improvement. The linking calculation with CONTAIN 2.0 has been also enabled to establish the unified auditing code system. Analysis for the CANDU plant real transient and hypothetical LOCA bas been performed using the improved version. It has been concluded that the developed version can be utilized for the auditing analysis of LOCA and non-LOCA event for the CANDU reactor. 25 refs., 84 figs., 36 tabs. (Author)
Summary of papers on current and anticipated uses of thermal-hydraulic codes
International Nuclear Information System (INIS)
Caruso, R.
1997-01-01
The author reviews a range of recent papers which discuss possible uses and future development needs for thermal/hydraulic codes in the nuclear industry. From this review, eight common recommendations are extracted. They are: improve the user interface so that more people can use the code, so that models are easier and less expensive to prepare and maintain, and so that the results are scrutable; design the code so that it can easily be coupled to other codes, such as core physics, containment, fission product behaviour during severe accidents; improve the numerical methods to make the code more robust and especially faster running, particularly for low pressure transients; ensure that future code development includes assessment of code uncertainties as integral part of code verification and validation; provide extensive user guidelines or structure the code so that the 'user effect' is minimized; include the capability to model multiple fluids (gas and liquid phase); design the code in a modular fashion so that new models can be added easily; provide the ability to include detailed or simplified component models; build on work previously done with other codes (RETRAN, RELAP, TRAC, CATHARE) and other code validation efforts (CSAU, CSNI SET and IET matrices)
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
Modelling the Hydraulic Processes on Constructed Stormwater Wetland
Directory of Open Access Journals (Sweden)
Isri Ronald Mangangka
2017-03-01
Full Text Available Constructed stormwater wetlands are manmade, shallow, and extensively vegetated water bodies which promote runoff volume and peak flow reduction, and also treat stormwater runoff quality. Researchers have noted that treatment processes of runoff in a constructed wetland are influenced by a range of hydraulic factors, which can vary during a rainfall event, and their influence on treatment can also vary as the event progresses. Variation in hydraulic factors during an event can only be generated using a detailed modelling approach, which was adopted in this research by developing a hydraulic conceptual model. The developed model was calibrated using trial and error procedures by comparing the model outflow with the measured field outflow data. The accuracy of the developed model was analyzed using a well-known statistical analysis method developed based on the regression analysis technique. The analysis results show that the developed model is satisfactory.
Energy Technology Data Exchange (ETDEWEB)
Waata, C.L.
2006-07-15
The use of water at supercritical pressure as coolant and moderator introduces a challenge in the design of a High-Performance Light-Water Reactor (HPLWR) fuel assembly. At supercritical pressure condition (P=25 MPa), the thermal-hydraulics behaviour of water differs strongly from that at sub-critical pressure due to a rapid variation of the thermal-physical properties across the pseudo-critical line. Due of the strong link between the water (moderation) and the neutron spectrum and subsequently the power distribution, a coupling of neutronics and thermal-hydraulics has become a necessity for reactor concepts operating at supercritical pressure condition. The effect of neutron moderation on the local parameters of thermal-hydraulics and vice-verse in a fuel assembly has to be considered for an accurate design analysis. In this study, the Monte Carlo N-Particle code (MCNP) and the sub-channel code STAFAS (Sub-channel Thermal-hydraulics Analysis of a Fuel Assembly under Supercritical conditions) have been coupled for the design analysis of a fuel assembly with supercritical water as coolant and moderator. Both codes are well known for complex geometry modelling. The MCNP code is used for neutronics analyses and for the prediction of power profiles of individual fuel rods. The sub-channel code STAFAS for the thermal-hydraulics analyses takes into account the coolant properties beyond the critical point as well as separate moderator channels. The coupling procedure is realized automatically. MCNP calculates the power distribution in each fuel rod, which is then transferred into STAFAS to obtain the corresponding thermal-hydraulic conditions in each sub-channel. The new thermal-hydraulic conditions are used to generate a new input deck for the next MCNP calculation. This procedure is repeated until a converged state is achieved. The coupled code system was tested on a proposed fuel assembly design of a HPLWR. An under-relaxation was introduced to achieve convergence
International Nuclear Information System (INIS)
Waata, C.L.
2006-07-01
The use of water at supercritical pressure as coolant and moderator introduces a challenge in the design of a High-Performance Light-Water Reactor (HPLWR) fuel assembly. At supercritical pressure condition (P=25 MPa), the thermal-hydraulics behaviour of water differs strongly from that at sub-critical pressure due to a rapid variation of the thermal-physical properties across the pseudo-critical line. Due of the strong link between the water (moderation) and the neutron spectrum and subsequently the power distribution, a coupling of neutronics and thermal-hydraulics has become a necessity for reactor concepts operating at supercritical pressure condition. The effect of neutron moderation on the local parameters of thermal-hydraulics and vice-verse in a fuel assembly has to be considered for an accurate design analysis. In this study, the Monte Carlo N-Particle code (MCNP) and the sub-channel code STAFAS (Sub-channel Thermal-hydraulics Analysis of a Fuel Assembly under Supercritical conditions) have been coupled for the design analysis of a fuel assembly with supercritical water as coolant and moderator. Both codes are well known for complex geometry modelling. The MCNP code is used for neutronics analyses and for the prediction of power profiles of individual fuel rods. The sub-channel code STAFAS for the thermal-hydraulics analyses takes into account the coolant properties beyond the critical point as well as separate moderator channels. The coupling procedure is realized automatically. MCNP calculates the power distribution in each fuel rod, which is then transferred into STAFAS to obtain the corresponding thermal-hydraulic conditions in each sub-channel. The new thermal-hydraulic conditions are used to generate a new input deck for the next MCNP calculation. This procedure is repeated until a converged state is achieved. The coupled code system was tested on a proposed fuel assembly design of a HPLWR. An under-relaxation was introduced to achieve convergence
Analysis of the Thermal and Hydraulic Stimulation Program at Raft River, Idaho
Bradford, Jacob; McLennan, John; Moore, Joseph; Podgorney, Robert; Plummer, Mitchell; Nash, Greg
2017-05-01
The Raft River geothermal field, located in southern Idaho, roughly 100 miles northwest of Salt Lake City, is the site of a Department of Energy Enhanced Geothermal System project designed to develop new techniques for enhancing the permeability of geothermal wells. RRG-9 ST1, the target stimulation well, was drilled to a measured depth of 5962 ft. and cased to 5551 ft. The open-hole section of the well penetrates Precambrian quartzite and quartz monzonite. The well encountered a temperature of 282 °F at its base. Thermal and hydraulic stimulation was initiated in June 2013. Several injection strategies have been employed. These strategies have included the continuous injection of water at temperatures ranging from 53 to 115 °F at wellhead pressures of approximately 275 psi and three short-term hydraulic stimulations at pressures up to approximately 1150 psi. Flow rates, wellhead and line pressures and fluid temperatures are measured continuously. These data are being utilized to assess the effectiveness of the stimulation program. As of August 2014, nearly 90 million gallons have been injected. A modified Hall plot has been used to characterize the relationships between the bottom-hole flowing pressure and the cumulative injection fluid volume. The data indicate that the skin factor is decreased, and/or the permeability around the wellbore has increased since the stimulation program was initiated. The injectivity index also indicates a positive improvement with values ranging from 0.15 gal/min psi in July 2013 to 1.73 gal/min psi in February 2015. Absolute flow rates have increased from approximately 20 to 475 gpm by February 2 2015. Geologic, downhole temperature and seismic data suggest the injected fluid enters a fracture zone at 5650 ft and then travels upward to a permeable horizon at the contact between the Precambrian rocks and the overlying Tertiary sedimentary and volcanic deposits. The reservoir simulation program FALCON developed at the Idaho National
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
International standard problem ISP-47 on containment thermal hydraulics - Final report
International Nuclear Information System (INIS)
Allelein, H. J.; Schwarz, S.; Fischer, K.; Vendel, J.; Malet, J.; Bentaib, A.; Studer, E.; Paillere, H.; Houkema, M.
2007-01-01
The main objective of the ISP-47 is to assess the capabilities of Lumped Parameter (LP) and Computational Fluid Dynamics (CFD) codes in the area of containment thermal-hydraulics. Following the recommendations made in the state-of-the-art report on 'Containment Thermal-hydraulics and Hydrogen Distribution' this ISP was based on the application of different complementary experimental facilities and a progressive modelling difficulty. The three experimental facilities TOSQAN, MISTRA and ThAI have shown the quality of the provided experimental data suitable for CFD and LP code benchmarking in steady-state and transient conditions (control of the initial and boundary conditions, and the accuracy of the measurement techniques). This mainly includes pressure transients and gas temperature field as in former exercises. Detailed gas velocity and gas concentration (air, steam and helium) fields were obtained for the first time in such an exercise. ISP-47 was executed in two main steps: Step 1 was dedicated to the validation of the codes in the separate effects facility TOSQAN (7 m 3 ). Wall condensation, steam injection in air or air / helium atmospheres, and buoyancy were addressed under well-controlled initial conditions in a simple geometry. Furthermore, the interactions of phenomena such as condensation / stratification, turbulence / buoyancy, etc. were addressed using the larger scale of MISTRA (100 m 3 ) facility. Both TOSQAN and MISTRA were specifically designed to produce data for CFD codes with state-of-the-art instrumentation. The TOSQAN benchmark was open, whereas the MISTRA benchmark was blind. Step 2 addressed the code assessment using an experiment in the multi-compartment ThAI (60 m 3 ) facility with different steam and helium injection phases, transient stratification and mixing conditions in the atmosphere, development of natural convection, wall condensate distribution, fog formation, and transient thermal response of heat conducting walls. Detailed
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
Thermal-hydraulic experiment and analysis for interim dry storage of spent nuclear fuel
International Nuclear Information System (INIS)
Yoo, Seung Hun
2011-02-01
parameters including grid sizes, radiation model, boundary conditions, a small-scale model was designed and tested which has the main components of a dry storage cask, such as fuel rods, basket and cask. Based on the sensitivity study the full-scope simulations by FLUENT were performed and compared with the experiment data and COBRA-SFS results. FLUENT results for the perdition of temperature distribution were agreed within measurement uncertainty. However, there were larger deviations for prediction of velocity between FLUENT and COBRA-SFS. The velocity predicted by COBRA-SFS was much higher than that predicted by FLUENT. The sensitivity studies with the full-scope simulation model were performed for a grid size and basket gap sizes. It was revealed that the basket gap size was the most sensitive parameter deciding the PCT in the dry storage cask. By this study, we proposed an applicable guideline of CFD use for the full-scope simulation of the dry storage cask. In the third study, a new type of the interim storage, which consists of one tunnel with eighty metal storage casks and one chimney, was proposed and studied. The tunnel-type storage has the following advantages: the reduction of the risk of the air plane crash and the easy control of the capacity of the system through the height of the chimney. The thermal-hydraulic characteristics of the tunnel storage were analyzed by FLUENT. The CFD simulation methodology was validated against the experimental data of a vault storage system (Sakamoto et al., 2000) because of the similar thermal-hydraulic characteristics between the tunnel and the vault. Then, the sensitivity studies of three parameters were done: the tunnel geometry, the cask arrangement and the turbulence. The local peak temperature on the metal casks is mainly governed by mass flow rates in the gaps. V1 case was recommended for the accommodation of the metal storage casks in the tunnel. The staggered cask arrangement was recommended due to the advantageous
An XFEM Model for Hydraulic Fracturing in Partially Saturated Rocks
Directory of Open Access Journals (Sweden)
Salimzadeh Saeed
2016-01-01
Full Text Available Hydraulic fracturing is a complex multi-physics phenomenon. Numerous analytical and numerical models of hydraulic fracturing processes have been proposed. Analytical solutions commonly are able to model the growth of a single hydraulic fracture into an initially intact, homogeneous rock mass. Numerical models are able to analyse complex problems such as multiple hydraulic fractures and fracturing in heterogeneous media. However, majority of available models are restricted to single-phase flow through fracture and permeable porous rock. This is not compatible with actual field conditions where the injected fluid does not have similar properties as the host fluid. In this study we present a fully coupled hydro-poroelastic model which incorporates two fluids i.e. fracturing fluid and host fluid. Flow through fracture is defined based on lubrication assumption, while flow through matrix is defined as Darcy flow. The fracture discontinuity in the mechanical model is captured using eXtended Finite Element Method (XFEM while the fracture propagation criterion is defined through cohesive fracture model. The discontinuous matrix fluid velocity across fracture is modelled using leak-off loading which couples fracture flow and matrix flow. The proposed model has been discretised using standard Galerkin method, implemented in Matlab and verified against several published solutions. Multiple hydraulic fracturing simulations are performed to show the model robustness and to illustrate how problem parameters such as injection rate and rock permeability affect the hydraulic fracturing variables i.e. injection pressure, fracture aperture and fracture length. The results show the impact of partial saturation on leak-off and the fact that single-phase models may underestimate the leak-off.
A Component-Scale Thermal-Hydraulic Analysis of the Flow in a CANDU Moderator Tank
Energy Technology Data Exchange (ETDEWEB)
Choi, Su Ryong; Jeong, Jae Jun [Pusan National University, Busan (Korea, Republic of); Yoon, Han Young [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)
2015-05-15
In this regard, to estimate the local subcooling of the moderator inside the Calandria vessel is very important. However, the local temperature is measured at the inlet and the outlet of the vessel only. Therefore, we need to accurately predict the local temperature inside the Calandria vessel. In this study the component-scale thermal-hydraulic analysis of a real CANDU moderator tank is performed by using the CUPID code, which is based on the lessons from the previous results. The porous media approach is used for an efficient simulation at a component scale. The inlet nozzles are also modeled using a simplified mesh. The validity of the results is checked and effects of the inlet nozzle modeling on the results are discussed. To examine the effects of the inlet nozzle modeling, three inlet nozzle models are proposed. The Case 1 using a uniform velocity at the nozzles presented a thermal stratification in the upper region of the Calandria tank, which is caused by the lack of the flow momentum injected into the inlet nozzles. The results are not realistic. The Case 2 using a more realistic velocity distribution instead of the uniform velocity, predicted an asymmetric mixed flow regime, maintaining with a balance between the momentum and buoyancy force, which was experimentally shown the STERN test facility in the nominal operating conditions. For a further improvement, the Case 3 was attempted, which aims at the preservation of both the mass and momentum flow at the inlet nozzles. The results of the Case 3 seem the best among the three cases. The local maximum temperature was very close to that of other calculations and the flow pattern was also very similar to the experimental and computational results.
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
Neutronic / thermal-hydraulic coupling with the code system Trace / Parcs
International Nuclear Information System (INIS)
Mejia S, D. M.; Del Valle G, E.
2015-09-01
The developed models for Parcs and Trace codes corresponding for the cycle 15 of the Unit 1 of the Laguna Verde nuclear power plant are described. The first focused to the neutronic simulation and the second to thermal hydraulics. The model developed for Parcs consists of a core of 444 fuel assemblies wrapped in a radial reflective layer and two layers, a superior and another inferior, of axial reflector. The core consists of 27 total axial planes. The model for Trace includes the vessel and its internal components as well as various safety systems. The coupling between the two codes is through two maps that allow its intercommunication. Both codes are used in coupled form performing a dynamic simulation that allows obtaining acceptably a stable state from which is carried out the closure of all the main steam isolation valves (MSIVs) followed by the performance of safety relief valves (SRVs) and ECCS. The results for the power and reactivities introduced by the moderator density, the fuel temperature and total temperature are shown. Data are also provided like: the behavior of the pressure in the steam dome, the water level in the downcomer, the flow through the MSIVs and SRVs. The results are explained for the power, the pressure in the steam dome and the water level in the downcomer which show agreement with the actions of the MSIVs, SRVs and ECCS. (Author)
Thermal hydraulic investigations and optimization on the EVC system of a PWR by CFD simulation
Energy Technology Data Exchange (ETDEWEB)
Xi, Mengmeng [Department of Nuclear Science and Technology, State Key Laboratory of Multiphase Flow in Power Engineering, Xi’an Jiaotong University, 710049 Xi’an (China); Zhang, Dalin, E-mail: dlzhang@mail.xjtu.edu.cn [Department of Nuclear Science and Technology, State Key Laboratory of Multiphase Flow in Power Engineering, Xi’an Jiaotong University, 710049 Xi’an (China); Tang, Mao [China Nuclear Power Design Engineering Co., Ltd., 518124 Shenzhen (China); Wang, Chenglong; Zheng, Meiyin; Qiu, Suizheng [Department of Nuclear Science and Technology, State Key Laboratory of Multiphase Flow in Power Engineering, Xi’an Jiaotong University, 710049 Xi’an (China)
2015-08-15
Highlights: • This study constructs a full CFD model for the EVC system of a PWR. • The complex fluid and solid coupling is treated in the computation. • Primary characteristics of the velocity, pressure and temperature distributions in the EVC system are investigated. • The optimization of the EVC system with different inlet boundaries are performed. - Abstract: In order to optimize the design of Reactor Pit Ventilation (EVC) system in a Pressurized Water Reactor (PWR), it is necessary to study the characteristics of the velocity, pressure and temperature fields in the EVC system. A full computational fluid dynamics (CFD) model for the EVC system is constructed by a commercial CFD code, where the complex fluid and solid coupling is treated. The Shear Stress Transport (SST) model is adopted to perform the turbulence calculation. This paper numerically investigates the characteristics of the velocity, pressure and temperature distributions in the EVC system. In particular, the effects of inlet air parameters on the thermal hydraulic characteristics and the reactor pit structure are also discussed for the EVC system optimization. Simulations are carried out with different mesh sizes and boundary conditions for sensitivity analysis. The computational results are important references to optimize the design and verify the rationality of the EVC system.
International Nuclear Information System (INIS)
Tyobeka, Bismark; Reitsma, Frederik; Ivanov, Kostadin
2011-01-01
The continued development of High Temperature Gas Cooled Reactors (HTGRs) requires verification of HTGR design and safety features with reliable high fidelity physics models and robust, efficient, and accurate codes. The predictive capability of coupled neutronics/thermal hydraulics and depletion simulations for reactor design and safety analysis can be assessed with sensitivity analysis and uncertainty analysis methods. In order to benefit from recent advances in modeling and simulation and the availability of new covariance data (nuclear data uncertainties) extensive sensitivity and uncertainty studies are needed for quantification of the impact of different sources of uncertainties on the design and safety parameters of HTGRs. Uncertainty and sensitivity studies are an essential component of any significant effort in data and simulation improvement. In February 2009, the Technical Working Group on Gas-Cooled Reactors recommended that the proposed IAEA Coordinated Research Project (CRP) on the HTGR Uncertainty Analysis in Modeling be implemented. In the paper the current status and plan are presented. The CRP will also benefit from interactions with the currently ongoing OECD/NEA Light Water Reactor (LWR) UAM benchmark activity by taking into consideration the peculiarities of HTGR designs and simulation requirements. (author)
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
A validation of ATR LOCA thermal-hydraulic code with a statistical approach
International Nuclear Information System (INIS)
Mochizuki, Hiroyasu
2000-01-01
When cladding temperatures are measured for a blowdown experiment, cladding temperatures at the same elevation in the fuel bundle have usually some differences due to eccentricity of the fuel bundle and other reasons such as biased two-phase flow. In the present paper, manufacturing tolerances and uncertainties of thermal-hydraulics are incorporated into a LOCA code that is applied with the statistical method. The present method was validated with the results of different blowdown experiments conducted using the 6 MW blowdown facility simulating the Advanced Thermal Reactor (ATR). In the present statistical method, the code was modified to run fast in order to calculate the blowdown thermal-hydraulics a lot of times with the code using different sets of input data. These input data for sizes and empirical correlations are prepared by the effective Monte-Carlo method based on the distribution functions deduced by the measured manufacturing errors and the uncertainties of thermal hydraulics. The calculated curves express uncertainties due to the different input deck. The uncertainty band and tendency of the cladding temperature were dependent on the beak sizes in the experiment. The measured results were traced by the present method. (author)
the Modeling of Hydraulic Jump Generated Partially on Sloping Apron
Directory of Open Access Journals (Sweden)
Shaker Abdulatif Jalil
2017-12-01
Full Text Available Modeling aims to characterize system behavior and achieve simulation close as possible of the reality. The rapid energy exchange in supercritical flow to generate quiet or subcritical flow in hydraulic jump phenomenon is important in design of hydraulic structures. Experimental and numerical modeling is done on type B hydraulic jump which starts first on sloping bed and its end on horizontal bed. Four different apron slopes are used, for each one of these slopes the jump is generated on different locations by controlling the tail water depth. Modelling validation is based on 120 experimental runs which they show that there is reliability. The air volume fraction which creates in through hydraulic jump varied between 0.18 and 0.28. While the energy exchanges process take place within 6.6, 6.1, 5.8, 5.5 of the average relative jump height for apron slopes of 0.18, 0.14, 0.10, 0.07 respectively. Within the limitations of this study, mathematical prediction model for relative hydraulic jump height is suggested.The model having an acceptable coefficient of determination.
Energy Technology Data Exchange (ETDEWEB)
Cai, Lijun, E-mail: cailj@swip.ac.cn [Southwestern Institute of Physics, Chengdu (China); Lin, Tao; Wang, Yingqiao; Wang, Mingxu [Southwestern Institute of Physics, Chengdu (China); Maruyama, So; Yang, Yu; Kiss, Gabor [ITER Organization, Route de Vinon-sur-Verdon, CS 90 046, 13067 St. Paul Lez Durance Cedex (France)
2016-11-01
Highlights: • The plasma facing closure cap has to survive after 30,000 thermal heat load cycles. • 0.35 MW/m2 radiation heat load plus nuclear heat load are very challenging for stainless steel. • Multilayer structure has been designed by using advanced welding and drilling technology to solve the neutron heating problem. • Accurate volumetric load application in analysis model by CFX has been mastered. - Abstract: Glow discharge cleaning (GDC) shall be used on ITER device to reduce and control impurity and hydrogenic fuel out-gassing from in-vessel plasma facing components. After first plasma, permanent electrode (PE) will be used to replace Temporary Electrode (TE) for subsequent operation. Two fundamental scenarios i.e., GDC and Plasma Operation State (POS) should be considered for electrode design, which requires the heat load caused by plasma radiation and neutron heating must be taken away by cooling water flowing inside the electrode. In this paper, multilayer cooling channels inside PE are preliminarily designed, and snakelike route in each layer is adopted to improve the heat exchange. Detailed thermal-hydraulic analyses have been done to validate the design feasibility or rationality. The analysis results show that during GDC the cooling water inlet and outlet temperature difference is far less than the allowable temperature rise under water flow rate 0.15 kg/s compromised by many factors. For POS, the temperature rise and pressure drop are within the design goals, but high thermal stress occurs on the front surface of closure cap of electrode. After several iterations of optimization of the closure cap, the equivalent strain range after 30,000 loading cycles for POS is well below 0.3% design goals.
The 25 kWe solar thermal Stirling hydraulic engine system: Conceptual design
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.
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)
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)
Lower pressurization to increase BWR electric power under thermal hydraulic criteria
International Nuclear Information System (INIS)
Kataoka, Kazuyoshi; Chuman, Kazuto; Mizumachi, Wataru; Yoshioka, Ritsuo; Mori, Michitsugu; Horie, Akira; Machida, Yuzo
1995-01-01
Electric power output versus core size is one of the factors that determine the electricity generation costs of BWRs. The power output is roughly calculated from the thermal power of the BWR core and the thermal efficiency of the BWR turbine system. The thermal power is restricted by the reactor's thermal hydraulic criteria such as the maximum linear heat generation rate, the minimum critical power ratio, the pressure drop in the core and the feedwater temperature at the BWR inlet. The combination of a system pressure of approximately 5.5 MPa and a feedwater temperature of approximately 439 K offers the maximum electric power output for a BWR with 9 x 9 fuel bundles. The amount of electric power generated is about 9% more than that generated by a conventional BWR under the thermal hydraulic criteria. The electric power output increases as the system pressure and the feedwater temperature are decreased from the current design of 7.3 MPa and 488 K, respectively, because the increased critical power of the fuel bundles compensates for the lower thermal efficiency. (author)
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
Energy Technology Data Exchange (ETDEWEB)
Chemerisov, Sergey [Argonne National Lab. (ANL), Argonne, IL (United States); Gromov, Roman [Argonne National Lab. (ANL), Argonne, IL (United States); Makarashvili, Vakho [Argonne National Lab. (ANL), Argonne, IL (United States); Heltemes, Thad [Argonne National Lab. (ANL), Argonne, IL (United States); Sun, Zaijing [Argonne National Lab. (ANL), Argonne, IL (United States); Wardle, Kent E. [Argonne National Lab. (ANL), Argonne, IL (United States); Bailey, James [Argonne National Lab. (ANL), Argonne, IL (United States); Quigley, Kevin [Argonne National Lab. (ANL), Argonne, IL (United States); Stepinski, Dominique [Argonne National Lab. (ANL), Argonne, IL (United States); Vandegrift, George [Argonne National Lab. (ANL), Argonne, IL (United States)
2014-10-01
Argonne is assisting SHINE Medical Technologies in developing SHINE, a system for producing fission-product ^{99}Mo using a D/T-accelerator to produce fission in a non-critical target solution of aqueous uranyl sulfate. We have developed an experimental setup for studying thermal-hydraulics and bubble formation in the uranyl sulfate solution to simulate conditions expected in the SHINE target solution during irradiation. A direct electron beam from the linac accelerator will be used to irradiate a 20 L solution (sector of the solution vessel). Because the solution will undergo radiolytic decomposition, we will be able to study bubble formation and dynamics and effects of convection and temperature on bubble behavior. These experiments will serve as a verification/ validation tool for the thermal-hydraulic model. Utilization of the direct electron beam for irradiation allows homogeneous heating of a large solution volume and simplifies observation of the bubble dynamics simultaneously with thermal-hydraulic data collection, which will complement data collected during operation of the miniSHINE experiment. Irradiation will be conducted using a 30-40 MeV electron beam from the high-power linac accelerator. The total electron-beam power will be 20 kW, which will yield a power density on the order of 1 kW/L. The solution volume will be cooled on the front and back surfaces and central tube to mimic the geometry of the proposed SHINE solution vessel. Also, multiple thermocouples will be inserted into the solution vessel to map thermal profiles. The experimental design is now complete, and installation and testing are in progress.
Scaling for integral simulation of thermal-hydraulic phenomena in SBWR during LOCA
Energy Technology Data Exchange (ETDEWEB)
Ishii, M.; Revankar, S.T.; Dowlati, R [Purdue Univ., West Layfayette, IN (United States)] [and others
1995-09-01
A scaling study has been conducted for simulation of thermal-hydraulic phenomena in the Simplified Boiling Water Reactor (SBWR) during a loss of coolant accident. The scaling method consists of a three-level scaling approach. The integral system scaling (global scaling or top down approach) consists of two levels, the integral response function scaling which forms the first level, and the control volume and boundary flow scaling which forms the second level. The bottom up approach is carried out by local phenomena scaling which forms the third level scaling. Based on this scaling study the design of the model facility called Purdue University Multi-Dimensional Integral Test Assembly (PUMA) has been carried out. The PUMA facility has 1/4 height and 1/100 area ratio scaling, corresponding to the volume scaling of 1/400. The PUMA power scaling based on the integral scaling is 1/200. The present scaling method predicts that PUMA time scale will be one-half that of the SBWR. The system pressure for PUMA is full scale, therefore, a prototypic pressure is maintained. PUMA is designed to operate at and below 1.03 MPa (150 psi), which allows it to simulate the prototypic SBWR accident conditions below 1.03 MPa (150 psi). The facility includes models for all components of importance.
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.)
Investigation of Correlations for the Thermal-hydraulic Analysis of Liquid Metal Reactors
Energy Technology Data Exchange (ETDEWEB)
Chang, Won Pyo; Jeong, Hae Yong; Lee, Yong Bum
2007-08-15
The present investigation is aimed at reducing favorable constitutive correlations from those developed for the thermal-hydraulic analysis of Liquid Metal Reactors (LMR), for reliable safety analyses of KALIMER. It is achieved by analyzing them in a point of their accuracies. The study is particularly important because its outcomes can provide an essential knowledge on their relative errors including their conservatisms to be analyzed in the future KALIMER licensing stage. The predictions of the correlations have been compared with available experimental data on both friction factors for the wired-wrapped rod bundles in the core and the heat transfer coefficients in the system. As a result, the heat transfer coefficient inside pipe currently featured in SSC-K has been found acceptable. It, however, has shown a discrepancy of about 60 % and thus an alternative one has been proposed for improvement. Meanwhile, the friction factor model in the current SSC-K has not shown a prominent discrepancy in prediction trend but it has not backed an enough theoretical basis so that another model has been proposed. A systematic assessment for effects of those factors to the conservatism must be fully understood for the future licensing stage, and systematic calculations must be followed by designing an assessment matrix. Besides, it is essential to conduct experiments under similar conditions for constitutive parts of geometries which represent the KALIMER design.
Thyc, a 3D thermal-hydraulic code for rod bundles. Recent developments and validation tests
International Nuclear Information System (INIS)
Caremoli, C.; Rascle, P.; Aubry, S.; Olive, J.
1993-09-01
PWR or LMFBR cores or fuel assemblies, PWR steam generators, condensers, tubular heat exchangers, are basic components of a nuclear power plant involving two-phase flows in tub