Thermally Actuated Hydraulic Pumps

Science.gov (United States)

Jones, Jack; Ross, Ronald; Chao, Yi

2008-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

Page, R.; Jones, J.R.

1997-07-01

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

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

International Nuclear Information System (INIS)

Page, R.; Jones, J.R.

1997-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2015-08-15

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

Comparison report of open calculations for ATLAS Domestic Standard Problem (DSP 02)

International Nuclear Information System (INIS)

Choi, Ki Yong; Kim, Y. S.; Kang, K. H.; Cho, S.; Park, H. S.; Choi, N. H.; Kim, B. D.; Min, K. H.; Park, J. K.; Chun, H. G.; Yu, Xin Guo; Kim, H. T.; Song, C. H.; Sim, S. K.; Jeon, S. S.; Kim, S. Y.; Kang, D. G.; Choi, T. S.; Kim, Y. M.; Lim, S. G.; Kim, H. S.; Kang, D. H.; Lee, G. H.; Jang, M. J.

2012-09-01

KAERI (Korea Atomic Energy Research Institute) has been operating an integral effect test facility, the Advanced Thermal Hydraulic Test Loop for Accident Simulation (ATLAS) for transient and accident simulations of advanced pressurized water reactors (PWRs). By using the ATLAS, a high quality integral effect test database has been established for major design basis accidents of the APR1400. A Domestic Standard Problem (DSP) exercise using the ATLAS database was promoted in order to transfer the database to domestic nuclear industries and to contribute to improving safety analysis methodology for PWRs. This 2nd ATLAS DSP exercise was led by KAERI in collaboration with KINS since the successful completion of the 1st ATLAS DSP in 2009. This exercise aims at effective utilization of integral effect database obtained from the ATLAS, establishment of cooperation framework among the domestic nuclear industry, better understanding of thermal hydraulic phenomena, and investigation of the possible limitation of the existing best estimate safety analysis codes. A small break loss of coolant accident of 6 inch break at the cold leg was determined as a target scenario by considering its technical importance and by incorporating with interests from participants. Twelve domestic organizations joined this DSP 02 exercise. Finally, eleven out of the joined organizations submitted their calculation results, including universities, government, and nuclear industries. This DSP exercise was performed in an open calculation environment where the integral effect test data was open to participants prior to code calculations. This report includes all information of the 2nd ATLAS DSP (DSP 02) exercise as well as comparison results between the calculations and the experimental data

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

Energy Technology Data Exchange (ETDEWEB)

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

1997-09-01

The work reported in this document identifies the thermal-hydraulic phenomena that are expected to occur during a number of key transients in a 330 MWt SMART integral reactor which is under development at KAERI. The result of this efforts is based on the current design concept of SMART integral reactor. Although the design is still evolving, the preliminary Phenomena Identification and Ranking Table (PIRT) has been developed based on the experts` knowledge and experience. The preliminary PIRT has been developed by the consensus of KAERI expert panelists and AHP (Analytical Hierarchy Process). Preliminary PIRT developed in this report is intended for use to identify and integrate development areas of further experimental tests needed and thermal-hydraulic models and correlations and code improvements for the safety analysis of the SMART integral reactor. (author). 7 refs., 21 tabs., 22 figs.

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

International Nuclear Information System (INIS)

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

1993-01-01

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

Review of computational thermal-hydraulic modeling

International Nuclear Information System (INIS)

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

1995-01-01

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

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

Integrated Validation System for a Thermal-hydraulic System Code, TASS/SMR-S

Energy Technology Data Exchange (ETDEWEB)

Kim, Hee-Kyung; Kim, Hyungjun; Kim, Soo Hyoung; Hwang, Young-Dong [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Kim, Hyeon-Soo [Chungnam National University, Daejeon (Korea, Republic of)

2015-10-15

Development including enhancement and modification of thermal-hydraulic system computer code is indispensable to a new reactor, SMART. Usually, a thermal-hydraulic system code validation is achieved by a comparison with the results of corresponding physical effect tests. In the reactor safety field, a similar concept, referred to as separate effect tests has been used for a long time. But there are so many test data for comparison because a lot of separate effect tests and integral effect tests are required for a code validation. It is not easy to a code developer to validate a computer code whenever a code modification is occurred. IVS produces graphs which shown the comparison the code calculation results with the corresponding test results automatically. IVS was developed for a validation of TASS/SMR-S code. The code validation could be achieved by a comparison code calculation results with corresponding test results. This comparison was represented as a graph for convenience. IVS is useful before release a new code version. The code developer can validate code result easily using IVS. Even during code development, IVS could be used for validation of code modification. The code developer could gain a confidence about his code modification easily and fast and could be free from tedious and long validation work. The popular software introduced in IVS supplies better usability and portability.

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.

Views on the future of thermal hydraulic modeling

International Nuclear Information System (INIS)

Ishii, M.

1997-01-01

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

Thermal Testing and Model Correlation for Advanced Topographic Laser Altimeter Instrument (ATLAS)

Science.gov (United States)

Patel, Deepak

2016-01-01

The Advanced Topographic Laser Altimeter System (ATLAS) part of the Ice Cloud and Land Elevation Satellite 2 (ICESat-2) is an upcoming Earth Science mission focusing on the effects of climate change. The flight instrument passed all environmental testing at GSFC (Goddard Space Flight Center) and is now ready to be shipped to the spacecraft vendor for integration and testing. This topic covers the analysis leading up to the test setup for ATLAS thermal testing as well as model correlation to flight predictions. Test setup analysis section will include areas where ATLAS could not meet flight like conditions and what were the limitations. Model correlation section will walk through changes that had to be made to the thermal model in order to match test results. The correlated model will then be integrated with spacecraft model for on-orbit predictions.

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.

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

Energy Technology Data Exchange (ETDEWEB)

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

2015-08-15

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

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)

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2015-01-15

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

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)

Reactor Thermal Hydraulic Numerical Calculation And Modeling

International Nuclear Information System (INIS)

Duong Ngoc Hai; Dang The Ba

2008-01-01

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

Thermal-Hydraulic Analysis of a Once-Through Steam Generator Considering Performance Degradation

International Nuclear Information System (INIS)

Han, Hun Sik; Kang, Han Ok; Yoon, Ju Hyeon; Kim, Young In; Song, Jae Seung; Kim, Keung Koo

2016-01-01

Several countries have entered into a global race for the commercialization of SMRs, and considerable research and development have been implemented. Among the various reactor designs, many SMRs have adopted an integral type pressurized water reactor (PWR) to enhance the nuclear safety and system reliability. In the integral reactor design, a single reactor pressure vessel contains primary system components such as fuel and core, steam generators, pumps, and a pressurizer. For the component integration into a reactor vessel, it is important to design each component as small as possible. Thus, it is a common practice to employ a once-through steam generator in the integral reactor design due to its advantages in compactness. In general, gradual degradation in thermal-hydraulic performance of the steam generator occurs with time, and it changes slowly the operating point of the steam generator during plant lifetime. Numerical solutions are acquired to evaluate the thermal-hydraulic performance of the steam generator at various AUFs. The design results obtained show that the average tube length of the steam generator is augmented with the increase of design margin to compensate for the design uncertainties and heat transfer area reduction by plugging, fouling, etc. A helically coiled tube once-through steam generator with 30% design margin is considered for comparison of thermal-hydraulic performances according to the degradation rate

Thermal-Hydraulic Analysis of a Once-Through Steam Generator Considering Performance Degradation

Energy Technology Data Exchange (ETDEWEB)

Han, Hun Sik; Kang, Han Ok; Yoon, Ju Hyeon; Kim, Young In; Song, Jae Seung; Kim, Keung Koo [KAERI, Daejeon (Korea, Republic of)

2016-05-15

Several countries have entered into a global race for the commercialization of SMRs, and considerable research and development have been implemented. Among the various reactor designs, many SMRs have adopted an integral type pressurized water reactor (PWR) to enhance the nuclear safety and system reliability. In the integral reactor design, a single reactor pressure vessel contains primary system components such as fuel and core, steam generators, pumps, and a pressurizer. For the component integration into a reactor vessel, it is important to design each component as small as possible. Thus, it is a common practice to employ a once-through steam generator in the integral reactor design due to its advantages in compactness. In general, gradual degradation in thermal-hydraulic performance of the steam generator occurs with time, and it changes slowly the operating point of the steam generator during plant lifetime. Numerical solutions are acquired to evaluate the thermal-hydraulic performance of the steam generator at various AUFs. The design results obtained show that the average tube length of the steam generator is augmented with the increase of design margin to compensate for the design uncertainties and heat transfer area reduction by plugging, fouling, etc. A helically coiled tube once-through steam generator with 30% design margin is considered for comparison of thermal-hydraulic performances according to the degradation rate.

ANTHEM2000TM: Integration of the ANTHEM Thermal Hydraulic Model in the ROSETM Environment

International Nuclear Information System (INIS)

Boire, R.; Nguyen, M; Salim, G.

1999-01-01

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

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.

Thermal-Hydraulic Performance of Scrubbing Nozzle Used for CFVS

Energy Technology Data Exchange (ETDEWEB)

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

2016-05-15

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

Fuel management service for Tarapur Atomic Power Station core thermal hydraulics

International Nuclear Information System (INIS)

Saha, D.; Venkat Raj, V.; Markandeya, S.G.

1977-01-01

Core thermal hydraulic analysis forms an integral part of the fuel management service for the Tarapur reactors. A distinguishing feature of boiling water reactors is the dependence of core flow distribution on the power distribution. Because of the changes in the axial and radial power distribution from cycle to cycle as well as during the cycle and also the variations in leakage flow, it is necessary to evaluate the core thermal hydraulic parameters for every cycle. Some of the typical results obtained in the course of analysis for different cycles of both the units at Tarapur are presented. The use of MCPR (Minimum Critical Power Ratio), instead of MCHFR (Minimum Critical Heat Flux Ratio) as a figure of merit for fuel cladding integrity is also discussed. (K.B.)

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

Characteristics of Core Thermal-Hydraulic Design of SMART-P

International Nuclear Information System (INIS)

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

2006-01-01

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

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.

Double-Ended Break Test of an 8.5 inch Direct Vessel Injection Line using the ATLAS

International Nuclear Information System (INIS)

Choi, Ki Yong; Kang, Kyoung Ho; Kim, Bok Deuk; Kim, Yeon Sik; Min, Kyoung Ho; Park, Choon Kyoung; Park, Hyun Sik; Baek, Won Pil; Cho, Seok; Choi, Nam Hyun

2010-01-01

A thermal-hydraulic integral effect test facility, ATLAS (Advanced Thermal-hydraulic Test Loop for Accident Simulation), has been constructed at the Korea Atomic Energy Research Institute (KAERI). It is a 1/2 reduced height and 1/288-volume scaled test facility with respect to the APR1400, an evolutionary pressurized water reactor developed by the Korean industry. In 2008, an integral effect test for simulating a guillotine break of a DVI line of the APR1400 was carried out as the first DVI test item, named as SB-DVI-03, on May, 2008. With an improvement on the break flow measuring system, the second DVI test for a guillotine break of a DVI line of the APR1400, named as SB-DVI-08, was conducted for repeatability. The present data is the first integral effect test data of its kind for simulating a DVI line break accident. It will help in understanding the thermal hydraulic phenomena occurring during the DVI line break accident. A post-test calculation was performed with a best-estimate safety analysis code MARS 3.1 to examine its prediction capability and to identify any code deficiencies for the thermal hydraulic phenomena occurring during the DVI line break accidents. The present integral effect test data will be used to validate the current safety analysis methodology for the DVI line break accident

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

International Nuclear Information System (INIS)

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

2010-01-01

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

Experimental thermal hydraulics in support of FBR

International Nuclear Information System (INIS)

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

2009-01-01

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

Primary system thermal hydraulics of future Indian fast reactors

Energy Technology Data Exchange (ETDEWEB)

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

2015-12-01

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

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

International Nuclear Information System (INIS)

Lemonnier, H.

2005-01-01

; aerosol transport, deposition and re-entrainment; steam generators thermal-hydraulics; system codes development and assessment; uncertainties analysis; diffuse interface methods and interface tracking methods; C - severe accidents and fires: molten core natural convection and physico-chemical phenomena, modeling and experiments; fuel coolant interaction, modeling and experiments; debris bed cooling; combustion and fires, modeling and experiments; molten corium concrete interaction; D - advanced code developments: fast transient modelling and experiments; multidimensional single-phase or two-phase flow and heat transfer modeling; neutronics and thermal-hydraulics coupling; fluid and structures mechanical interactions; coupled thermal-hydraulics of fluids and structures; thermal-hydraulic dependent corrosion and ablation; E - operation and safety of existing reactors: instabilities and nonlinear dynamics; NPP transients and accidents analysis; RBMK and VVER safety analysis, including the OECD benchmark; F - experimental thermal-hydraulics: boiling heat transfer; CHF and post-CHF heat transfer; condensation heat transfer; integral testing; vibrations, wear and thermal fatigue phenomena; fuel design and performance; G - advanced reactors thermal-hydraulics (gen IV, INPRO, fusion, hydrogen production): accelerator driven reactors; advanced pressurized water reactors thermal-hydraulics; gas cooled fast reactors; gas cooled high temperature reactors; lead and lead-bismuth cooled reactors; future and existing sodium cooled reactors; molten salt reactors; H - waste management thermal-hydraulics: thermal-hydraulics problems related to waste processing and storage; I - thermal-hydraulics of non electricity generating nuclear equipment: sono-fusion (cavitation induced bubble fusion; hydrogen producing nuclear reactors

Experimental Results of A1.2 Test for OECD-ATLAS Project

Energy Technology Data Exchange (ETDEWEB)

Kang, Kyoung-Ho; Bae, Byoung-Uhn; Park, Yu-Sun; Kim, Jong-Rok; Choi, Nam-Hyun; Choi, Ki-Yong [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2015-10-15

In order to meet the international interests in the multiple high-risk design extension conditions (DECs) raised after the Fukushima accident, KAERI (Korea Atomic Energy Research Institute) is operating an OECD/NEA project (hereafter, OECD-ATLAS project) by utilizing a thermal-hydraulic integral effect test facility, ATLAS (Advanced Thermal-hydraulic Test Loop for Accident Simulation). As for a prolonged SBO transient of the OECD-ATLAS project, two tests, named A1.1 and A1.2, were determined to be performed. In particular, passive safety systems are considered as the most promising alternatives to reinforce the safety and reliability of an ultimate heat removal system without any operator actions in the SBO transients. As one of the new safety improvement concepts to mitigate an SBO accident efficiently, a cooling and operational performance of the passive auxiliary feedwater system (PAFS) is investigated in the framework of the OECD-ATLAS project to produce clearer knowledge of the actual phenomena and to provide the best guidelines for accident management. As the second test of the OECD-ATLAS project, the A1.2 test was conducted to simulate a prolonged SBO with asymmetric secondary cooling through the supply of passive auxiliary feedwater only to SG-2. When the collapsed water level of steam generator reached a wide range of 25%, PAFS was actuated. PAFS played a key role in cooling down the primary system by the heat transfer and the natural circulation. With the actuation of PAFS, the fluid temperatures at the core inlet and outlet started to decrease without any excursion of the maximum heater surface temperature in the core. This integral effect test data of A1.2 test can be used to evaluate the prediction capability of existing safety analysis codes and identify any code deficiency for an SBO simulation with an operation of a passive system such as PAFS.

SMITHERS: An object-oriented modular mapping methodology for MCNP-based neutronic–thermal hydraulic multiphysics

International Nuclear Information System (INIS)

Richard, Joshua; Galloway, Jack; Fensin, Michael; Trellue, Holly

2015-01-01

Highlights: • A modular mapping methodogy for neutronic-thermal hydraulic nuclear reactor multiphysics, SMITHERS, has been developed. • Written in Python, SMITHERS takes a novel object-oriented approach for facilitating data transitions between solvers. This approach enables near-instant compatibility with existing MCNP/MONTEBURNS input decks. • It also allows for coupling with thermal-hydraulic solvers of various levels of fidelity. • Two BWR and PWR test problems are presented for verifying correct functionality of the SMITHERS code routines. - Abstract: A novel object-oriented modular mapping methodology for externally coupled neutronics–thermal hydraulics multiphysics simulations was developed. The Simulator using MCNP with Integrated Thermal-Hydraulics for Exploratory Reactor Studies (SMITHERS) code performs on-the-fly mapping of material-wise power distribution tallies implemented by MCNP-based neutron transport/depletion solvers for use in estimating coolant temperature and density distributions with a separate thermal-hydraulic solver. The key development of SMITHERS is that it reconstructs the hierarchical geometry structure of the material-wise power generation tallies from the depletion solver automatically, with only a modicum of additional information required from the user. Additionally, it performs the basis mapping from the combinatorial geometry of the depletion solver to the required geometry of the thermal-hydraulic solver in a generalizable manner, such that it can transparently accommodate varying levels of thermal-hydraulic solver geometric fidelity, from the nodal geometry of multi-channel analysis solvers to the pin-cell level of discretization for sub-channel analysis solvers. The mapping methodology was specifically developed to be flexible enough such that it could successfully integrate preexisting depletion solver case files with different thermal-hydraulic solvers. This approach allows the user to tailor the selection of a

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)

Evaluation of operational safety at Babcock and Wilcox Plants: Volume 2, Thermal-hydraulic results

International Nuclear Information System (INIS)

Wheatley, P.D.; Davis, C.B.; Callow, R.A.; Fletcher, C.D.; Dobbe, C.A.; Beelman, R.J.

1987-11-01

The Nuclear Regulatory Commission has initiated a research program to develop a methodology to assess the operational performance of Babcock and Wilcox plants and to apply this methodology on a trial basis. The methodology developed for analyzing Babcock and Wilcox plants integrated methods used in both thermal-hydraulics and human factors and compared results with information used in the assessment of risk. The integrated methodology involved an evaluation of a selected plant for each pressurized water reactor vendor during a limited number of transients. A plant was selected to represent each vendor, and three transients were identified for analysis. The plants were Oconee Unit 1 for Babcock and Wilcox, H.B. Robinson Unit 2 for Westinghouse, and Calvert Cliffs Unit 1 for Combustion Engineering. The three transients were a complete loss of all feedwater, a small-break loss-of-coolant accident, and a steam-generator overfill with auxiliary feedwater. Included in the integrated methodology was an assessment of the thermal-hydraulic behavior, including event timing, of the plants during the three transients. Thermal-hydraulic results are presented in this volume (Volume 2) of the report. 26 refs., 30 figs., 7 tabs

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

Proceedings of the third nuclear thermal hydraulics meeting

International Nuclear Information System (INIS)

Anon.

1987-01-01

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

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

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)

Process management using component thermal-hydraulic function classes

Science.gov (United States)

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

1999-07-27

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

Process management using component thermal-hydraulic function classes

Science.gov (United States)

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

1999-01-01

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

Thermal-Hydraulic Experiments and Modelling for Advanced Nuclear Reactor Systems

International Nuclear Information System (INIS)

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

2005-04-01

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

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.

Thermal hydraulic feasibility assessment for the Spent Nuclear Fuel Project

International Nuclear Information System (INIS)

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

1996-01-01

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

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)

Thermal-Hydraulic Tests for Reactor Core Safety

International Nuclear Information System (INIS)

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

2005-04-01

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

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)

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

Energy Technology Data Exchange (ETDEWEB)

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

2017-07-01

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

An overview on rod-bundle thermal-hydraulic analyses

International Nuclear Information System (INIS)

Sha, W.T.

1980-01-01

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

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

International Nuclear Information System (INIS)

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

1997-01-01

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

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

International Nuclear Information System (INIS)

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

2000-01-01

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

Thermal-hydraulic analysis for wire-wrapped PWR cores

Energy Technology Data Exchange (ETDEWEB)

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

2009-08-15

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

SPACE code simulation of ATLAS DVI line break accident test (SB DVI 08 Test)

Energy Technology Data Exchange (ETDEWEB)

Lim, Sang Gyu [KHNP, Daejeon (Korea, Republic of)

2012-10-15

APR1400 has adopted new safety design features which are 4 mechanically independent DVI (Direct Vessel Injection) systems and fluidic device in the safety injection tanks (SITs). Hence, DVI line break accident has to be evaluated as one of the small break LOCA (SBLOCA) to ensure the safety of APR1400. KAERI has been performed for DVI line break test (SB DVI 08) using ATLAS (Advanced Thermal Hydraulic Test Loop for Accident Simulation) facility which is an integral effect test facility for APR1400. The test result shows that the core collapsed water level decreased before a loop seal clearance, so that a core uncover occurred. At this time, the peak cladding temperature (PCT) is rapidly increased even though the emergency core cooling (ECC) water is injected from safety injection pump (SIP). This test result is useful for supporting safety analysis using thermal hydraulic safety analysis code and increases the understanding of SBLOCA phenomena in APR1400. The SBLOCA evaluation methodology for APR1400 is now being developed using SPACE code. The object of the development of this methodology is to set up a conservative evaluation methodology in accordance with appendix K of 10 CFR 50. ATLAS SB DVI 08 test is selected for the evaluation of SBLOCA methodology using SPACE code. Before applying the conservative models and correlations, benchmark calculation of the test is performed with the best estimate models and correlations to verify SPACE code capability. This paper deals with benchmark calculations results of ATLAS SB DVI 08 test. Calculation results of the major hydraulics variables are compared with measured data. Finally, this paper carries out the SPACE code performances for simulating the integral effect test of SBLOCA.

HANARO thermal hydraulic accident analysis

Energy Technology Data Exchange (ETDEWEB)

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

1996-06-01

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

Validation of thermal hydraulic computer codes for advanced light water reactor

International Nuclear Information System (INIS)

Macek, J.

2001-01-01

The Czech Republic operates 4 WWER-440 units, two WWER-1000 units are being finalised (one of them is undergoing commissioning). Thermal-hydraulics Department of the Nuclear Research Institute Rez performs accident analyses for these plants using a number of computer codes. To model the primary and secondary circuits behaviour the system codes ATHLET, CATHARE, RELAP, TRAC are applied. Containment and pressure-suppressure system are modelled with RALOC and MELCOR codes, the reactor power calculations (point and space-neutron kinetics) are made with DYN3D, NESTLE and CDF codes (FLUENT, TRIO) are used for some specific problems. An integral part of the current Czech project 'New Energy Sources' is selection of a new nuclear source. Within this and the preceding projects financed by the Czech Ministry of Industry and Trade and the EU PHARE, the Department carries and has carried out the systematic validation of thermal-hydraulic and reactor physics computer codes applying data obtained on several experimental facilities as well as the real operational data. The paper provides a concise information on these activities of the NRI and its Thermal-hydraulics Department. A detailed example of the system code validation and the consequent utilisation of the results for a real NPP purposes is included. (author)

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

Energy Technology Data Exchange (ETDEWEB)

Ebert, D.

1997-07-01

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

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

International Nuclear Information System (INIS)

Ebert, D.

1997-07-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

1998-12-31

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)

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

Energy Technology Data Exchange (ETDEWEB)

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

1997-12-31

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

Thermal-hydraulic and aerosol containment phenomena modelling in ASTEC severe accident computer code

International Nuclear Information System (INIS)

Kljenak, Ivo; Dapper, Maik; Dienstbier, Jiri; Herranz, Luis E.; Koch, Marco K.; Fontanet, Joan

2010-01-01

Transients in containment systems of different scales (Phebus.FP containment, KAEVER vessel, Battelle Model Containment, LACE vessel and VVER-1000 nuclear power plant containment) involving thermal-hydraulic phenomena and aerosol behaviour, were simulated with the computer integral code ASTEC. The results of the simulations in the first four facilities were compared with experimental results, whereas the results of the simulated accident in the VVER-1000 containment were compared to results obtained with the MELCOR code. The main purpose of the simulations was the validation of the CPA module of the ASTEC code. The calculated results support the applicability of the code for predicting in-containment thermal-hydraulic and aerosol phenomena during a severe accident in a nuclear power plant.

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

Energy Technology Data Exchange (ETDEWEB)

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

2002-03-01

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

Study on safety of a nuclear ship having an integral marine water reactor. Intelligent information database program concerned with thermal-hydraulic characteristics

International Nuclear Information System (INIS)

Inasaka, Fujio; Nariai, Hideki; Kobayashi, Michiyuki; Murata, Hiroyuki; Aya, Izuo

2001-01-01

As a high economical marine reactor with sufficient safety functions, an integrated type marine water reactor has been considered most promising. At the National Maritime Research Institute, a series of the experimental studies on the thermal-hydraulic characteristics of an integrated/passive-safety type marine water reactor such as the flow boiling of a helical-coil type steam generator, natural circulation of primary water under a ship rolling motion and flashing-condensation oscillation phenomena in pool water has been conducted. This current study aims at making use of the safety analysis or evaluation of a future marine water reactor by developing an intelligent information database program concerned with the thermal-hydraulic characteristics of an integral/passive-safety reactor on the basis of the above-mentioned valuable experimental knowledge. Since the program was created as a Windows application using the Visual Basic, it is available to the public and can be easily installed in the operating system. Main functions of the program are as follows: (1) steady state flow boiling analysis and determination of stability limit for any helical-coil type once-through steam generator design. (2) analysis and comparison with the flow boiling data, (3) reference and graphic display of the experimental data, (4) indication of the knowledge information such as analysis method and results of the study. The program will be useful for the design of not only the future integrated type marine water reactor but also the small sized water reactor. (author)

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.

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

Science.gov (United States)

Aldaeef, A A; Rayhani, M T

2015-10-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2015-10-15

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

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.

Horizontal steam generator PGV-1000 thermal-hydraulic analysis

International Nuclear Information System (INIS)

Ubra, O.; Doubek, M.

1995-01-01

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

Horizontal steam generator PGV-1000 thermal-hydraulic analysis

Energy Technology Data Exchange (ETDEWEB)

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

1996-12-31

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

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.

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

Energy Technology Data Exchange (ETDEWEB)

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

2011-07-01

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

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)

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.

Experimental Results of A1.1 Test for OECD-ATLAS Project

Energy Technology Data Exchange (ETDEWEB)

Kang, Kyoung-Ho; Bae, Byoung-Uhn; Park, Yu-Sun; Kim, Jong-Rok; Choi, Nam-Hyun; Choi, Ki-Yong [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2015-10-15

KAERI (Korea Atomic Energy Research Institute) is operating an OECD/NEA project (hereafter, OECD-ATLAS project) by utilizing a thermal-hydraulic integral effect test facility, ATLAS (Advanced Thermal-hydraulic Test Loop for Accident Simulation). Considering the importance of the SBO scenario and the related accident mitigation measures, a prolonged SBO scenario was selected as the first test subject worthy of investigation in the OECD-ATLAS project as summarized in Table 1. After the Fukushima accident, design extension conditions (DECs) such as an SBO and a total loss of feed water (TLOFW) attracted wide international attention in that such high-risk multiple failure accidents should be revisited from the viewpoint of the reinforcement of the 'defense in depth' concept. In particular, an SBO is one of the most important DECs because a total loss of heat sink can lead to a core melt-down scenario under high pressure without any proper operator action. As for a prolonged SBO transient of the OECD-ATLAS project, two tests, named A1.1 and A1.2, were determined to be performed. In most nuclear power plants (NPPs), a turbine-driven auxiliary feedwater system was designed to remove the decay heat during the early period of an SBO transient. From a conservative point of view, however, it is necessary to investigate the thermal-hydraulic behaviors of the NPP when a turbine-driven auxiliary feedwater supply is not available during the initial period of an SBO transient and moreover a mobile pump-driven auxiliary feedwater supply can only become realized in the later period of the scenario. In particular, asymmetric heat removal characteristic through the supply of auxiliary feedwater only to one steam generator has its own peculiar importance in terms of safety analysis code validation. With an aim of considering these safety importance, in the A1.1 test, a prolonged SBO transient was simulated with two temporal phases: Phase (I) for a conservative SBO transient

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.

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

Directory of Open Access Journals (Sweden)

W.Z. Chen

2017-06-01

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.

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

International Nuclear Information System (INIS)

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

2016-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

Shadid, J.N., E-mail: jnshadi@sandia.gov [Sandia National Laboratories, Computational Mathematics Department (United States); Department of Mathematics and Statistics, University of New Mexico (United States); Smith, T.M. [Sandia National Laboratories, Multiphysics Applications Department (United States); Cyr, E.C. [Sandia National Laboratories, Computational Mathematics Department (United States); Wildey, T.M. [Sandia National Laboratories, Optimization and UQ Department (United States); Pawlowski, R.P. [Sandia National Laboratories, Multiphysics Applications Department (United States)

2016-09-15

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

Thermal hydraulic issues and challenges for current and new generation FBRs

Energy Technology Data Exchange (ETDEWEB)

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

2015-12-01

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

Thermal and Alignment Analysis of the Instrument-Level ATLAS Thermal Vacuum Test

Science.gov (United States)

Bradshaw, Heather

2012-01-01

This paper describes the thermal analysis and test design performed in preparation for the ATLAS thermal vacuum test. NASA's Advanced Topographic Laser Altimeter System (ATLAS) will be flown as the sole instrument aboard the Ice, Cloud, and land Elevation Satellite-2 (ICESat-2). It will be used to take measurements of topography and ice thickness for Arctic and Antarctic regions, providing crucial data used to predict future changes in worldwide sea levels. Due to the precise measurements ATLAS is taking, the laser altimeter has very tight pointing requirements. Therefore, the instrument is very sensitive to temperature-induced thermal distortions. For this reason, it is necessary to perform a Structural, Thermal, Optical Performance (STOP) analysis not only for flight, but also to ensure performance requirements can be operationally met during instrument-level thermal vacuum testing. This paper describes the thermal model created for the chamber setup, which was used to generate inputs for the environmental STOP analysis. This paper also presents the results of the STOP analysis, which indicate that the test predictions adequately replicate the thermal distortions predicted for flight. This is a new application of an existing process, as STOP analyses are generally performed to predict flight behavior only. Another novel aspect of this test is that it presents the opportunity to verify pointing results of a STOP model, which is not generally done. It is possible in this case, however, because the actual pointing will be measured using flight hardware during thermal vacuum testing and can be compared to STOP predictions.

The 3rd ATLAS Domestic Standard Problem for Improvement of Safety Analysis Technology

International Nuclear Information System (INIS)

Choi, Ki-Yong; Kang, Kyoung-Ho; Park, Yusun; Kim, Jongrok; Bae, Byoung-Uhn; Choi, Nam-Hyun

2014-01-01

The third ATLAS DSP (domestic standard problem exercise) was launched at the end of 2012 in response to the strong need for continuation of the ATLAS DSP. A guillotine break of a main steam line without LOOP at a zero power condition was selected as a target scenario, and it was successfully completed in the beginning of 2014. In the 3 rd ATLAS DSP, comprehensive utilization of the integral effect test data was made by dividing analysis with three topics; 1. scale-up where extrapolation of ATLAS IET data was investigated 2. 3D analysis where how much improvement can be obtained by 3D modeling was studied 3. 1D sensitivity analysis where the key phenomena affecting the SLB simulation were identified and the best modeling guideline was achieved. Through such DSP exercises, it has been possible to effectively utilize high-quality ATLAS experimental data of to enhance thermal-hydraulic understanding and to validate the safety analysis codes. A strong human network and technical expertise sharing among the various nuclear experts are also important outcomes from this program

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

Energy Technology Data Exchange (ETDEWEB)

Park, Joo Hwan; Choi, Hang Bok

2005-03-01

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

Atlas – a data warehouse for integrative bioinformatics

Directory of Open Access Journals (Sweden)

Yuen Macaire MS

2005-02-01

Full Text Available Abstract Background We present a biological data warehouse called Atlas that locally stores and integrates biological sequences, molecular interactions, homology information, functional annotations of genes, and biological ontologies. The goal of the system is to provide data, as well as a software infrastructure for bioinformatics research and development. Description The Atlas system is based on relational data models that we developed for each of the source data types. Data stored within these relational models are managed through Structured Query Language (SQL calls that are implemented in a set of Application Programming Interfaces (APIs. The APIs include three languages: C++, Java, and Perl. The methods in these API libraries are used to construct a set of loader applications, which parse and load the source datasets into the Atlas database, and a set of toolbox applications which facilitate data retrieval. Atlas stores and integrates local instances of GenBank, RefSeq, UniProt, Human Protein Reference Database (HPRD, Biomolecular Interaction Network Database (BIND, Database of Interacting Proteins (DIP, Molecular Interactions Database (MINT, IntAct, NCBI Taxonomy, Gene Ontology (GO, Online Mendelian Inheritance in Man (OMIM, LocusLink, Entrez Gene and HomoloGene. The retrieval APIs and toolbox applications are critical components that offer end-users flexible, easy, integrated access to this data. We present use cases that use Atlas to integrate these sources for genome annotation, inference of molecular interactions across species, and gene-disease associations. Conclusion The Atlas biological data warehouse serves as data infrastructure for bioinformatics research and development. It forms the backbone of the research activities in our laboratory and facilitates the integration of disparate, heterogeneous biological sources of data enabling new scientific inferences. Atlas achieves integration of diverse data sets at two levels. First

Development of thermal hydraulic models for the reliable regulatory auditing code

Energy Technology Data Exchange (ETDEWEB)

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

2003-04-15

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

Development of thermal hydraulic models for the reliable regulatory auditing code

International Nuclear Information System (INIS)

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

2003-04-01

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

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

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

Energy Technology Data Exchange (ETDEWEB)

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

1998-03-01

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

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

International Nuclear Information System (INIS)

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

2004-01-01

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

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

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-hydraulics of the Loviisa reactor pressure vessel overcooling transients

International Nuclear Information System (INIS)

Tuomisto, Harri.

1987-06-01

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

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.

Hydraulic and thermal design of a gas microchannel heat exchanger

International Nuclear Information System (INIS)

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

2012-01-01

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

Hydraulic and thermal conduction phenomena in soils at the particle-scale: Towards realistic FEM simulations

International Nuclear Information System (INIS)

Narsilio, G A; Yun, T S; Kress, J; Evans, T M

2010-01-01

This paper summarizes a method to characterize conduction properties in soils at the particle-scale. The method set the bases for an alternative way to estimate conduction parameters such as thermal conductivity and hydraulic conductivity, with the potential application to hard-to-obtain samples, where traditional experimental testing on large enough specimens becomes much more expensive. The technique is exemplified using 3D synthetic grain packings generated with discrete element methods, from which 3D granular images are constructed. Images are then imported into the finite element analyses to solve the corresponding governing partial differential equations of hydraulic and thermal conduction. High performance computing is implemented to meet the demanding 3D numerical calculations of the complex geometrical domains. The effects of void ratio and inter-particle contacts in hydraulic and thermal conduction are explored. Laboratory measurements support the numerically obtained results and validate the viability of the new methods used herein. The integration of imaging with rigorous numerical simulations at the pore-scale also enables fundamental observation of particle-scale mechanisms of macro-scale manifestation.

Investigation of coupling scheme for neutronic and thermal-hydraulic codes

International Nuclear Information System (INIS)

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

1988-01-01

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

Nuclear power plant thermal-hydraulic performance research program plan

International Nuclear Information System (INIS)

1988-07-01

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

Development of thermal hydraulic evaluation code for CANDU reactors

International Nuclear Information System (INIS)

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

2004-02-01

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

Thermal hydraulic analysis of BWR containment venting system

International Nuclear Information System (INIS)

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

2015-01-01

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

ATLAS Cold Leg Top Slot Break Analysis using RELAP5

Energy Technology Data Exchange (ETDEWEB)

Sung, Haejung; Lee, Sang Ik; Park, Ju-Hyun; Choi, Tong-Soo [KEPCO NF, Daejeon (Korea, Republic of)

2016-10-15

U.S. Nuclear Regulatory Commission (US-NRC) has been reviewing the design certification application for APR1400 submitted by Korea Electric Power Corporation (KEPCO). The main concern about cold leg top slot break is that cladding temperature might be increased by core uncover due to four loop seal reformation following flooding of safety injection water. An integral effect test for cold leg top slot break was performed by KAERI (Korea Atomic Energy Research Institute) using ATLAS (Advanced Thermal-Hydraulic Test Loop for Accident Simulation), which is a scaled down experimental facility for APR1400. In this study, RELAP5/MOD3.3/Patch04 is assessed by experimental result of ATLAS cold leg top slot break. Also, thermal hydraulic phenomena by four loop seals reformation is observed by RELAP5 result. The RELAP5/MOD3.3/Patch04 is assessed by the experimental result of ATLAS cold leg top slot break. The top slot break is described by offtake model, and the mass flow rate is fairly well estimated. The RELAP5 well predicts the correlation between general trend and four loop seal reformation. The pressure of the core region and the cladding temperature tends to increase during four loop seal reformation due to steam path blockage on four loop seals. It is presumed that the code cannot estimate two phase phenomena by loop seal clearing as same as experiments. In terms of cladding temperature, loop seal reformation due to loop seal elevation of APR1400 does not need to be the issue, since the void fraction at the active top core is maintained over 0.4.

Momentum integral network method for thermal-hydraulic transient analysis

International Nuclear Information System (INIS)

Van Tuyle, G.J.

1983-01-01

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

Simulation of total loss of feed water in ATLAS test facility using SPACE code

Energy Technology Data Exchange (ETDEWEB)

Kim, Minhee; Kim, Seyun [Korea Hydro and Nuclear Power Co., Daejeon (Korea, Republic of). Central Research Inst.

2017-08-15

A total loss of feedwater (TLOFW) with additional failures in ATLAS test facility was analyzed using SPACE code, which is an advanced thermal-hydraulic system analysis code developed by the Korea nuclear industry. Partial failure of the safety injection pumps (SIPs) and the pilot-operated safety relief valves (POSRVs) of pressurizer were selected as additional failures. In order to assess the capability of SPACE code, partial failure was modeled, and compared with results of OECD-ATLAS A3.1 results. Reasonably good agreement with major thermal-hydraulic parameters was obtained by analyzing the transient behavior. From the results, this indicated that SPACE code has capabilities to design extension conditions, and feed and bleed operation using POSRVs and SIPs were effective for RCS cooling capability during TLOFW.

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.

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)

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

International Nuclear Information System (INIS)

Ville Lestinen; Timo Toppila

2005-01-01

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

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)

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.

Thermal-hydraulic analysis of nuclear reactors

CERN Document Server

Zohuri, Bahman

2015-01-01

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

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

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

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)

Extension of BEPU methods to Sub-channel Thermal-Hydraulics and to Coupled Three-Dimensional Neutronics/Thermal-Hydraulics Codes

International Nuclear Information System (INIS)

Avramova, M.; Ivanov, K.; Arenas, C.

2013-01-01

The principles that support the risk-informed regulation are to be considered in an integrated decision-making process. Thus, any evaluation of licensing issues supported by a safety analysis would take into account both deterministic and probabilistic aspects of the problem. The deterministic aspects will be addressed using Best Estimate code calculations and considering the associated uncertainties i.e. Plus Uncertainty (BEPU) calculations. In recent years there has been an increasing demand from nuclear research, industry, safety and regulation for best estimate predictions to be provided with their confidence bounds. This applies also to the sub-channel thermal-hydraulic codes, which are used to evaluate local safety parameters. The paper discusses the extension of BEPU methods to the sub-channel thermal-hydraulic codes on the example of the Pennsylvania State University (PSU) version of COBRA-TF (CTF). The use of coupled codes supplemented with uncertainty analysis allows to avoid unnecessary penalties due to incoherent approximations in the traditional decoupled calculations, and to obtain more accurate evaluation of margins regarding licensing limit. This becomes important for licensing power upgrades, improved fuel assembly and control rod designs, higher burn-up and others issues related to operating LWRs as well as to the new Generation 3+ designs being licensed now (ESBWR, AP-1000, EPR-1600 and etc.). The paper presents the application of Generalized Perturbation Theory (GPT) to generate uncertainties associated with the few-group assembly homogenized neutron cross-section data used as input in coupled reactor core calculations. This is followed by a discussion of uncertainty propagation methodologies, being implemented by PSU in cooperation of Technical University of Catalonia (UPC) for reactor core calculations and for comprehensive multi-physics simulations. (authors)

Pre-test analysis of a LBLOCA using the design data of the ATLAS facility, a reduced-height integral effect test loop for PWRs

International Nuclear Information System (INIS)

Hyun-Sik Park; Ki-Yong Choi; Dong-Jin Euh; Tae-Soon Kwon; Won-Pil Baek

2005-01-01

Full text of publication follows: The simulation capability of the KAERI integral effect test facility, ATLAS (Advanced Thermalhydraulic Test Loop for Accident Simulation), has been assessed for a large-break loss-of-coolant accident (LBLOCA) transient. The ATLAS facility is a 1/2 height-scaled, 1/144 area-scaled (1/288 in volume scale), and full-pressure test loop based on the design features of the APR1400, an evolutionary pressurized water reactor that has been developed by Korean industry. The APR1400 has four mechanically separated hydraulic trains for the emergency core cooling system (ECCS) with direct vessel injection (DVI). The APR1400 design features have brought about several new safety issues related to the LBLOCA including the steam-water interaction, ECC bypass, and boiling in the reactor vessel downcomer. The ATLAS facility will be used to investigate the multiple responses between the systems or between the components during various anticipated transients. The ATLAS facility has been designed according to a scaling method that is mainly based on the model suggested by Ishii and Kataoka. The ATLAS facility is being evaluated against the prototype plant APR1400 with the same control logics and accident scenarios using the best-estimated code, MARS. This paper briefly introduces the basic design features of the ATLAS facility and presents the results of pre-test analysis for a postulated LBLOCA of a cold leg. The LBLOCA analyses has been conducted to assess the validity of the applied scaling law and the similarity between the ATLAS facility and the APR1400. As the core simulator of the ATLAS facility has the 10% capability of the scaled full power, the blowdown phase can not be simulated, and the starting point of the accident scenario is around the end of blowdown. So it is an important problem to find the correct initial conditions. For the analyzed LBLOCA scenario, the ATLAS facility showed very similar thermal-hydraulic characteristics to the APR

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)

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)

Thermal hydraulics and mechanics core design programs

International Nuclear Information System (INIS)

Heinecke, J.

1992-10-01

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

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

International Nuclear Information System (INIS)

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

2015-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2016-05-15

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

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

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

Science.gov (United States)

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

2017-07-01

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

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

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.

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.

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

International Nuclear Information System (INIS)

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

2005-01-01

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

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)

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

International Nuclear Information System (INIS)

Chao Yanmeng; Yang Lixin; Zhang Mingqian

2014-01-01

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

CSNI Integral Test Facility Matrices for Validation of Best-Estimate Thermal-Hydraulic Computer Codes

International Nuclear Information System (INIS)

Glaeser, H.

2008-01-01

Internationally agreed Integral Test Facility (ITF) matrices for validation of realistic thermal hydraulic system computer codes were established. ITF development is mainly for Pressurised Water Reactors (PWRs) and Boiling Water Reactors (BWRs). A separate activity was for Russian Pressurised Water-cooled and Water-moderated Energy Reactors (WWER). Firstly, the main physical phenomena that occur during considered accidents are identified, test types are specified, and test facilities suitable for reproducing these aspects are selected. Secondly, a list of selected experiments carried out in these facilities has been set down. The criteria to achieve the objectives are outlined. In this paper some specific examples from the ITF matrices will also be provided. The matrices will be a guide for code validation, will be a basis for comparisons of code predictions performed with different system codes, and will contribute to the quantification of the uncertainty range of code model predictions. In addition to this objective, the construction of such a matrix is an attempt to record information which has been generated around the world over the last years, so that it is more accessible to present and future workers in that field than would otherwise be the case.

Thermal-hydraulic criteria for the APT tungsten neutron source design

International Nuclear Information System (INIS)

Pasamehmetoglu, K.

1998-03-01

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

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)

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

Energy Technology Data Exchange (ETDEWEB)

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

2015-10-15

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

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

International Nuclear Information System (INIS)

Jeon, Byong Guk; Bae, Hwang; Ryu, Sung-Uk; Ryu, Hyobong; Byun, Sun-Joon; Yi, Sung-Jae; Park, Hyun-Sik

2015-01-01

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

Thermal hydraulic model descrition of TASS/SMR

Energy Technology Data Exchange (ETDEWEB)

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

2001-04-01

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

Analysis of uncertainties of thermal hydraulic calculations

International Nuclear Information System (INIS)

Macek, J.; Vavrin, J.

2002-12-01

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

Thermal hydraulics in undergraduate nuclear engineering education

International Nuclear Information System (INIS)

Theofanous, T.G.

1986-01-01

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

The integration and engineering of the ATLAS SemiConductor Tracker Barrel

Energy Technology Data Exchange (ETDEWEB)

Abdesselam, A; Barr, A J [Department of Physics, Oxford University, Denys Wilkinson Building, Keble Road, Oxford OX1 3RH (United Kingdom); Allport, P P; Austin, N [Oliver Lodge Laboratory, University of Liverpool, P.O. Box 147, Oxford Street, Liverpool L69 3BX (United Kingdom); Anastopoulos, C [University of Sheffield, Department of Physics and Astronomy, Hounsfield Road, Sheffield S3 7RH (United Kingdom); Anderson, B; Attree, D J [Department of Physics and Astronomy, University College London (United Kingdom); Andricek, L; Bangert, A [Max-Planck-Institut fuer Physik, (Werner-Heisenberg-Institut), Foehringer Ring 6, 80805 Muenchen (Germany); Anghinolfi, F [CERN, CH - 1211 Geneva 23 (Switzerland); Apsimon, R; Barclay, P; Batchelor, L E [Rutherford Appleton Laboratory, Science and Technology Facilities Council, Harwell Science and Innovation Campus, Didcot OX11 0QX (United Kingdom); Atkinson, T [School of Physics, University of Melbourne, Parkville, Victoria 3010 (Australia); Barbier, G [Universite de Geneve, Section de Physique, 24 rue Ernest Ansermet, CH - 1211 Geneve 4 (Switzerland); Bates, R L; Bell, W H [University of Glasgow, Department of Physics and Astronomy, Glasgow G12 8QQ (United Kingdom); Batley, J R [Cavendish Laboratory, University of Cambridge, J J Thomson Avenue, Cambridge CB3 0HE (United Kingdom); Beck, G A [Department of Physics, Queen Mary, University of London, Mile End Road, London E1 4NS (United Kingdom); Bell, P J [School of Physics and Astronomy, University of Manchester, Manchester M13 9PL (United Kingdom)] (and others)

2008-10-15

The ATLAS SemiConductor Tracker (SCT) was built in three sections: a barrel and two end-caps. This paper describes the design, construction and final integration of the barrel section. The barrel is constructed around four nested cylinders that provide a stable and accurate support structure for the 2112 silicon modules and their associated services. The emphasis of this paper is directed at the aspects of engineering design that turned a concept into a fully-functioning detector, as well as the integration and testing of large sub-sections of the final SCT barrel detector. The paper follows the chronology of the construction. The main steps of the assembly are described with the results of intermediate tests. The barrel service components were developed and fabricated in parallel so that a flow of detector modules, cooling loops, opto-harnesses and Frequency-Scanning-Interferometry (FSI) alignment structures could be assembled onto the four cylinders. Once finished, each cylinder was conveyed to the next site for the mounting of modules to form a complete single barrel. Extensive electrical and thermal function tests were carried out on the completed single barrels. In the next stage, the four single barrels and thermal enclosures were combined into the complete SCT barrel detector so that it could be integrated with the Transition Radiation Tracker (TRT) barrel to form the central part of the ATLAS inner detector. Finally, the completed SCT barrel was tested together with the TRT barrel in noise tests and using cosmic rays.

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

Energy Technology Data Exchange (ETDEWEB)

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

2017-02-15

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

Thermal-hydraulic characteristics of double flat core HCLWR

International Nuclear Information System (INIS)

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

1989-02-01

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

Thermal-Hydraulic Experiment Facility (THEF)

International Nuclear Information System (INIS)

Martinell, J.S.

1982-01-01

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

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

Intelligent information database of the thermal-hydraulic characteristics for a future marine water reactor

International Nuclear Information System (INIS)

Inasaka, Fujio; Nariai, Hideki

2000-01-01

At the Ship Research Institute, a series of the experimental studies on the thermal-hydraulic characteristics of an integrated type marine water reactor has been conducted. This current study aims at developing an intelligent information database program with the thermal-hydraulic characteristics of a future marine water reactor on the basis of the valuably experimental knowledge, which was obtained from the above-mentioned studies. In this paper, the experimental knowledge with the flow boiling of a once-through steam generator and the natural circulation of primary water under a ship rolling motion was converted into an intelligent information database program. The program was created as a Windows application using the Visual Basic. Main functions of the program are as follows: (1) steady state flow boiling analysis and determination of stability for any helical-coil type once-through steam generator design, (2) reference and graphic display of the experimental data, (3) reference of the information such as analysis method and experimental apparatus. The program will be useful for the design of not only the future integrated type marine water reactor but also the small sized reactor with helical-coil type steam generator. (author)

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

Science.gov (United States)

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

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

The International Experimental Thermal Hydraulic Systems database – TIETHYS: A new NEA validation tool

Energy Technology Data Exchange (ETDEWEB)

Rohatgi, Upendra S.

2018-07-22

Nuclear reactor codes require validation with appropriate data representing the plant for specific scenarios. The thermal-hydraulic data is scattered in different locations and in different formats. Some of the data is in danger of being lost. A relational database is being developed to organize the international thermal hydraulic test data for various reactor concepts and different scenarios. At the reactor system level, that data is organized to include separate effect tests and integral effect tests for specific scenarios and corresponding phenomena. The database relies on the phenomena identification sections of expert developed PIRTs. The database will provide a summary of appropriate data, review of facility information, test description, instrumentation, references for the experimental data and some examples of application of the data for validation. The current database platform includes scenarios for PWR, BWR, VVER, and specific benchmarks for CFD modelling data and is to be expanded to include references for molten salt reactors. There are place holders for high temperature gas cooled reactors, CANDU and liquid metal reactors. This relational database is called The International Experimental Thermal Hydraulic Systems (TIETHYS) database and currently resides at Nuclear Energy Agency (NEA) of the OECD and is freely open to public access. Going forward the database will be extended to include additional links and data as they become available. https://www.oecd-nea.org/tiethysweb/

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

International Nuclear Information System (INIS)

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

2017-01-01

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

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

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)

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.).

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

International Nuclear Information System (INIS)

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

1989-01-01

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

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.

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

Energy Technology Data Exchange (ETDEWEB)

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

2000-03-01

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

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

Thermal-hydraulic design of the 200 MW NHR

Energy Technology Data Exchange (ETDEWEB)

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

1997-09-01

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

Multiphase Flow Dynamics 5 Nuclear Thermal Hydraulics

CERN Document Server

Kolev, Nikolay Ivanov

2012-01-01

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

Multiphase flow dynamics 5 nuclear thermal hydraulics

CERN Document Server

Kolev, Nikolay Ivanov

2015-01-01

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

Virginia Power thermal-hydraulics methods

International Nuclear Information System (INIS)

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

1987-01-01

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

Validation of containment thermal hydraulic computer codes for VVER reactor

Energy Technology Data Exchange (ETDEWEB)

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

2005-07-01

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

Thermal-hydraulic design of the 200 MW NHR

International Nuclear Information System (INIS)

Li Jincai; Gao Zuying; Xu Baocheng; He Junxiao

1997-01-01

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

Thermal-hydraulic design of the 200 MW NHR

Energy Technology Data Exchange (ETDEWEB)

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

1997-09-01

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

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

International Nuclear Information System (INIS)

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

2006-01-01

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

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

International Nuclear Information System (INIS)

Kwon, Hyukjoon; Sprengel, Michael; Ivantysynova, Monika

2016-01-01

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

SBWR core thermal hydraulic analysis during startup

International Nuclear Information System (INIS)

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

1993-01-01

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

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.

Thermal-hydraulics technological strategy roadmap for LWR safety improvement and development

International Nuclear Information System (INIS)

Nakamura, Hideo; Arai, Kenji; Oikawa, Hirohide

2015-01-01

New version of the Thermal-Hydraulics Safety Evaluation Fundamental Technology Enhancement Strategy Roadmap (TH-RM) was developed by the Atomic Energy Society of Japan (AESJ) for LWR safety improvement and development. The 1st version of TH-RM was prepared in 2009 under collaboration of utilities, vendors, universities, research institutes and technical support organizations (TSO) for regulatory body. The revision was made by three sub-working groups (SWGs) by considering the lessons learned from the Fukushima Daiichi Accident. The 'safety assessment' SWG pursued development of computer codes for safety assessment. The 'fundamental technology' SWG pursued safety improvement and risk reduction via accident management (AM) measures by referring the technical map for severe accident (SA) established by the 'severe accident' SWG. Phenomena and components for counter-measures and/or proper prediction are identified by going through SA progression in both reactor and spent-fuel pool of PWR and BWR. Twelve important technology development subjects were identified, which include melt coolability enhancement to maintain integrity of containment vessel. Fact Sheet was developed to describe each of identified and selected R and D subjects. External hazards are also considered how to cope with from thermal-hydraulic safety point of view. This paper summarizes the revised TH-RM with several examples and future perspectives. (author)

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

International Nuclear Information System (INIS)

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

1994-07-01

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

The analysis of thermal-hydraulic performances of nuclear ship reactor

International Nuclear Information System (INIS)

Wakabayashi, Shinshichi; Hamada, Masao

1975-01-01

Thermal-hydraulic performances in the core of nuclear ship reactor was analysed by thermal-hydraulic analyser codes, AMRTC and COBRA-11+DNBCAL. This reactor is of a pressurized water type and incorporates the steam generator within the reactor vessel with the rated power of 330 MWt, which is developed by Nuclear Ship Research Panel Seven (NSR-7) in The Shipbuilding Research Association of Japan. Fuel temperature distributions, coolant temperature distributions, void fractions in coolant and minimum burn out ratio etc. were calculated. Results are as follows; a) The maximum temperature of fuel center is 1,472 0 C that corresponds to 53% as small as the melting point (2,800 0 C). b) Subcooled boiling exists in the core and the maximum void fraction is less than 4%. c) The minimum burn out ratio is not less than the minimum allowable limit of 1.25. It was found from the results of analysis that this reactor was able to be operated wide margin with respect to thermal-hydraulic design limits at the rated power. (auth.)

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

Simulation of Thermal Hydraulic at Supercritical Pressures with APROS

Energy Technology Data Exchange (ETDEWEB)

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

2008-07-01

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

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)

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

Energy Technology Data Exchange (ETDEWEB)

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

2016-10-15

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

Thermal-hydraulic calculations for KUHFR with reduced enrichment uranium fuel

International Nuclear Information System (INIS)

Mishima, Kaichiro; Shibata, Toshikazu.

1982-01-01

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

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

International Nuclear Information System (INIS)

Yarlagadda, B.S.

1989-04-01

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

Coupling analysis of deformation and thermal-hydraulics in a FBR fuel pin bundle using BAMBOO and ASFRE-IV Codes

International Nuclear Information System (INIS)

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

2004-03-01

The bundle-duct interaction may occur in sodium cooled wire-wrapped FBR fuel subassemblies in high burn-up conditions. JNC has been developing a bundle deformation analysis code BAMBOO (Behavior Analysis code for Mechanical interaction of fuel Bundle under On-power Operation), a thermal hydraulics analysis code ASFRE-IV (Analysis of Sodium Flow in Reactor Elements - ver. IV) and their coupling method as a simulation system for the evaluation on the integrity of deformed FBR fuel pin bundles. In this study, the simulation system was applied to a coupling analysis of deformation and thermal-hydraulics in the fuel pin-bundle under a steady-state condition just after startup for the purpose of the verification of the simulation system. The iterative calculations of deformation and thermal-hydraulics employed in the coupling analysis provided numerically unstable solutions. From the result, it was found that improvement of the coupling algorithm of BAMBOO and ASFRE-IV is necessary to reduce numerical fluctuations and to obtain better convergence by introducing such computational technique as the optimized under-relaxation method. (author)

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

International Nuclear Information System (INIS)

Zhang Xiaoying; Huang Yuanyuan

2010-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

1998-06-01

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

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

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

International Nuclear Information System (INIS)

Chung, Bub Dong; Lee, Won Jae; Kim, Hee Cheol; Song, Jin Ho; Sim, Suk Ku

1997-01-01

The work reported in this paper identifies the thermal-hydraluic 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 expert's knowledge and experience. The preliminary PIRT has been developed by consensus of KAERI expert panelists and AHP (Analytical Hierachy Process). Preliminary PIRT developed in this paper is intended to be used to identify and integrate development areas of further experimental test needed, thermal hydraulic models and correlations and code improvements for the safety analysis of the SMART

Control and Data Acquisition System of the ATLAS Facility

International Nuclear Information System (INIS)

Choi, Ki-Yong; Kwon, Tae-Soon; Cho, Seok; Park, Hyun-Sik; Baek, Won-Pil; Kim, Jung-Taek

2007-02-01

This report describes the control and data acquisition system of an integral effect test facility, ATLAS (Advanced Thermal-hydraulic Test Loop for Accident Simulation) facility, which recently has been constructed at KAERI (Korea Atomic Energy Research Institute). The control and data acquisition system of the ATLAS is established with the hybrid distributed control system (DCS) by RTP corp. The ARIDES system on a LINUX platform which is provided by BNF Technology Inc. is used for a control software. The IO signals consists of 1995 channels and they are processed at 10Hz. The Human-Machine-Interface (HMI) consists of 43 processing windows and they are classified according to fluid system. All control devices can be controlled by manual, auto, sequence, group, and table control methods. The monitoring system can display the real time trend or historical data of the selected IO signals on LCD monitors in a graphical form. The data logging system can be started or stopped by operator and the logging frequency can be selected among 0.5, 1, 2, 10Hz. The fluid system of the ATLAS facility consists of several systems including a primary system to auxiliary system. Each fluid system has a control similarity to the prototype plant, APR1400/OPR1000

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

accurately determine cross-flow and mixing effects within the subassembly. This analysis also provides detailed thermal profiles for more accurate assessment of thermal conditions such as approach to local boiling. The second restriction was also removed by using similar level high fidelity models in the surrounding subassemblies. Thus, there was no need for using simplified representative pin analyses in the surrounding assemblies and the correct inter-assembly flows were calculated directly. The results of these calculations demonstrate the ability of large scale, high fidelity CFD calculations to perform core design and reload analysis in a reasonable amount of time on massively parallel computers. This approach provides quantitative detail on the flow implications of proposed spacer grids and mixing vanes and provides necessary porous media flow coefficient information for subsequent whole core calculations. The fine detail multiple assembly calculation demonstrates the ability to perform high fidelity design calculations to more accurately determine thermal conditions and safety margins. Such detailed thermal hydraulic analyses with well established commercial CFD codes offer the possibility of building a fully integrated neutronic-thermal hydraulic design capability. (J.P.N.)

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

International Nuclear Information System (INIS)

Misra, B.; Maroni, V.A.

1977-01-01

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

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

International Nuclear Information System (INIS)

Misra, B.; Maroni, V.A.

1978-01-01

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

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

Thermal-hydraulic Analysis of High-temperature Cover Gas Region in STELLA-2

Energy Technology Data Exchange (ETDEWEB)

Jo, Youngchul; Son, Seok-Kwon; Yoon, Jung; Eoh, Jaehyuk; Jeong, Ji-Young [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2016-10-15

The first phase of the program was focused on the key sodium component tests, and the second one has been concentrated on the sodium thermal-hydraulic integral effect test (STELLA-2). Based on its platform, simulation of the PGSFR transient will be made to evaluate plant dynamic behaviors as well as to demonstrate decay heat removal performance. Therefore, most design features of PGSFR have been modeled in STELLA-2 as closely as possible. The similarities of temperature and pressure between the model (STELLA-2) and the prototype (PGSFR) have been well preserved to reflect thermal-hydraulic behavior with natural convection as well as heat transfer between structure and sodium coolant inside the model reactor vessel (RV). For this reason, structural integrity of the entire test section should be confirmed as in the prototype. In particular, since the model reactor head in STELLA-2 supports key components and internal structures, its structural integrity exposed to high-temperature cover gas region should be confirmed. In order to reduce thermal radiation heat transfer from the hot sodium pool during normal operation, a dedicated insulation layer has been installed at the downward surface of the model reactor head to prevent direct heat flux from the sodium free surface at 545 .deg. C. Three-dimensional conjugate heat transfer analyses for the full-shape geometry of the upper part of the model reactor vessel in STELLA-2 have been carried out. Based on the results, steady-state temperature distributions in the cover gas region and the model reactor head itself have been obtained and the design requirement in temperature of the model reactor head has been newly proposed to be 350 .deg. C. For any elevated temperature conditions in STELLA-2, it was confirmed that the model reactor head generally satisfied the requirement. The CFD database constructed from this study will be used to optimize geometric parameters such as thicknesses and/or types of the insulator.

AGIS: Integration of new technologies used in ATLAS Distributed Computing

CERN Document Server

Anisenkov, Alexey; The ATLAS collaboration; Alandes Pradillo, Maria

2016-01-01

AGIS is the information system designed to integrate configuration and status information about resources, services and topology of the computing infrastructure used by ATLAS Distributed Computing (ADC) applications and services. In this note, we describe the evolution and the recent developments of AGIS functionalities, related to integration of new technologies recently become widely used in ATLAS Computing like flexible computing utilization of opportunistic Cloud and HPC resources, ObjectStore services integration for Distributed Data Management (Rucio) and ATLAS workload management (PanDA) systems, unified storage protocols declaration required for PandDA Pilot site movers 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.

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

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)

Tests and final integration of the ATLAS semiconductor tracker

CERN Document Server

Mikulec, Bettina

2005-01-01

The Silicon Tracker (SCT) is part of the Inner Detector at the ATLAS experiment at CERN. Its basic building blocks are 5 different types of silicon strip modules. In total more than 15000 p-on-n single-sided silicon strip sensors of an area of about 61 m2 were used to produce 4088 SCT modules. An overall module production yield of 92% could be achieved, where the silicon modules comply with the tight electrical, thermal and mechanical specifications. The macro-assembly of 2112 barrel modules to the four barrel support cylinders was successfully carried out. The nine disks of one endcap are fully populated with 988 modules, and for the second endcap more than 50% of the modules are already mounted. Test results operating complete barrels will be presented as well as a description of the test setup. The different integration steps of the SCT with the surrounding Transition Radiation Tracker (TRT) will be explained. The installation of SCT and TRT into the ATLAS pit will happen during 2006.

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

International Nuclear Information System (INIS)

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

2000-01-01

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

An atlas of thermal data for biomass and other fuels

Energy Technology Data Exchange (ETDEWEB)

Gaur, S.; Reed, T.B. [Colorado School of Mines, Golden, CO (United States)

1995-06-01

Biomass is recognized as a major source of renewable energy. In order to convert biomass energy to more useful forms, it is necessary to have accurate scientific data on the thermal properties of biomass. This Atlas has been written to supply a uniform source of that information. In the last few decades Thermal analysis (TA) tools such as thermogravimetry, differential thermal analysis, thermo mechanical analysis, etc. have become more important. The data obtained from these techniques can provide useful information in terms of reaction mechanism, kinetic parameters, thermal stability, phase transformation, heat of reaction, etc. for gas-solid and gas-liquid systems. Unfortunately, there are no ASTM standards set for the collection of these types of data using TA techniques and therefore, different investigators use different conditions which suit their requirements for measuring this thermal data. As a result, the information obtained from different laboratories is not comparable. This Atlas provides the ability to compare new laboratory results with a wide variety of related data available in the literature and helps ensure consistency in using these data.

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

International Nuclear Information System (INIS)

Takenaka, Nobuyuki

1996-01-01

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

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

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)

AGIS: Integration of new technologies used in ATLAS Distributed Computing

Science.gov (United States)

Anisenkov, Alexey; Di Girolamo, Alessandro; Alandes Pradillo, Maria

2017-10-01

The variety of the ATLAS Distributed Computing infrastructure requires a central information system to define the topology of computing resources and to store different parameters and configuration data which are needed by various ATLAS software components. The ATLAS Grid Information System (AGIS) is the system designed to integrate configuration and status information about resources, services and topology of the computing infrastructure used by ATLAS Distributed Computing applications and services. Being an intermediate middleware system between clients and external information sources (like central BDII, GOCDB, MyOSG), AGIS defines the relations between experiment specific used resources and physical distributed computing capabilities. Being in production during LHC Runl AGIS became the central information system for Distributed Computing in ATLAS and it is continuously evolving to fulfil new user requests, enable enhanced operations and follow the extension of the ATLAS Computing model. The ATLAS Computing model and data structures used by Distributed Computing applications and services are continuously evolving and trend to fit newer requirements from ADC community. In this note, we describe the evolution and the recent developments of AGIS functionalities, related to integration of new technologies recently become widely used in ATLAS Computing, like flexible computing utilization of opportunistic Cloud and HPC resources, ObjectStore services integration for Distributed Data Management (Rucio) and ATLAS workload management (PanDA) systems, unified storage protocols declaration required for PandDA Pilot site movers and others. The improvements of information model and general updates are also shown, in particular we explain how other collaborations outside ATLAS could benefit the system as a computing resources information catalogue. AGIS is evolving towards a common information system, not coupled to a specific experiment.

Advanced thermal-hydraulic and neutronic codes: current and future applications. Summary and conclusions

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

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.

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

International Nuclear Information System (INIS)

Zhou Kefeng; Zhou Yangping; Sui Zhe; Ma Yuanle

2014-01-01

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

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

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

Energy Technology Data Exchange (ETDEWEB)

NONE

2004-07-01

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

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

International Nuclear Information System (INIS)

2004-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

NONE

2004-07-01

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

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

International Nuclear Information System (INIS)

2004-01-01

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

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.

Thermal hydraulic design of PFBR core

International Nuclear Information System (INIS)

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

2000-01-01

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

Thermal Hydraulic Design of PWT Accelerating Structures

CERN Document Server

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

2005-01-01

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

Creys-Malville nuclear plant. Simulation of the cold plenum thermal-hydraulics. 12 zone model presentation

International Nuclear Information System (INIS)

Faulot, J.P.

1990-05-01

The CRUSIFI code has been developed by SEPTEN (Engineering and Construction Division) with SICLE software during 1983-1985 in order to study the CREYS-MALVILLE dynamic behavior. At the time, the version was based on project data (version 2.3). It includes a 2 zones model for the cold plenum thermal-hydraulics, modelling which does not allow to reproduce accurately dissymetries apt to occur as well in usual operating (hydraulic dissymetries bound to one or many systems out of order), as during incidentally operating (hydraulic dissymetries bound to primary pump working back or thermal dissymetries after a transient on one or many secondary loops). Moreover, a 2 zones model cannot simulate axial temperature gradients which appear during double stratification phenomenon (upper and lower part of the plenum) produced by alternating thermal shock. A 12 zones model (4 sectors with 3 axial zones each) such as model developed by R$DD (Research and Development Division) allows to satisfy correctly these problems. This report is a specification of the chosen modelling. This model is now operational after qualifying with experimental transients on mockup and reactor. It is to-day connected with the EDF general operating code CRUSIFI (calibrating version 3.0). It could be easily integrated in a four loops plant modelling such as the CREYS-MALVILLE simulator in a four loops plant modelling such as the CREYS-MALVILLE simulator under construction at the present time by THOMSON

Overcooling transient selection and thermal hydraulic analyses of the Loviisa PTS assessments

Energy Technology Data Exchange (ETDEWEB)

Tuomisto, H [IVO Power Engineering Ltd, Vantaa (Finland)

1997-09-01

This paper describes transients selection and thermal hydraulic analyses of various PTS assessment studies performed for the pressure vessels of the Loviisa WWER-reactors. Deterministic analyses have been performed in various stages of the PTS studies and they have always made the formal basis for design and licensing of the reactor pressure vessel. The integrated, probabilistic PTS study was carried out to give an overview of the severity of all different PTS sequences, and give a quantitative estimate of the importance of the PTS issues in relation to the overall safety of the plant. Later, the sequences including external flooding of the pressure vessels were added to the PTS assessment. Thermal recovery annealing of the Loviisa 1 reactor pressure vessel took place during refuelling outage in 1996. (author). 10 refs, 4 figs, 3 tabs.

CSNI Integral test facility validation matrix for the assessment of thermal-hydraulic codes for LWR LOCA and transients

International Nuclear Information System (INIS)

1996-07-01

This report deals with an internationally agreed integral test facility (ITF) matrix for the validation of best estimate thermal-hydraulic computer codes. Firstly, the main physical phenomena that occur during the considered accidents are identified, test types are specified, and test facilities suitable for reproducing these aspects are selected. Secondly, a life of selected experiments carried out in these facilities has been set down. The criteria to achieve the objectives are outlined. The construction of such a matrix is an attempt to collect together in a systematic way the best sets of openly available test data for code validation, assessment and improvement, including quantitative assessment of uncertainties in the modelling of phenomena by the codes. In addition to this objective, it is an attempt to record information which has been generated around the world over the last 20 years so that it is more accessible to present and future workers in that field than would otherwise be the case

Thermal Performance of ATLAS Laser Thermal Control System Demonstration Unit

Science.gov (United States)

Ku, Jentung; Robinson, Franklin; Patel, Deepak; Ottenstein, Laura

2013-01-01

The second Ice, Cloud, and Land Elevation Satellite mission currently planned by National Aeronautics and Space Administration will measure global ice topography and canopy height using the Advanced Topographic Laser Altimeter System {ATLAS). The ATLAS comprises two lasers; but only one will be used at a time. Each laser will generate between 125 watts and 250 watts of heat, and each laser has its own optimal operating temperature that must be maintained within plus or minus 1 degree Centigrade accuracy by the Laser Thermal Control System (LTCS) consisting of a constant conductance heat pipe (CCHP), a loop heat pipe (LHP) and a radiator. The heat generated by the laser is acquired by the CCHP and transferred to the LHP, which delivers the heat to the radiator for ultimate rejection. The radiator can be exposed to temperatures between minus 71 degrees Centigrade and minus 93 degrees Centigrade. The two lasers can have different operating temperatures varying between plus 15 degrees Centigrade and plus 30 degrees Centigrade, and their operating temperatures are not known while the LTCS is being designed and built. Major challenges of the LTCS include: 1) A single thermal control system must maintain the ATLAS at 15 degrees Centigrade with 250 watts heat load and minus 71 degrees Centigrade radiator sink temperature, and maintain the ATLAS at plus 30 degrees Centigrade with 125 watts heat load and minus 93 degrees Centigrade radiator sink temperature. Furthermore, the LTCS must be qualification tested to maintain the ATLAS between plus 10 degrees Centigrade and plus 35 degrees Centigrade. 2) The LTCS must be shut down to ensure that the ATLAS can be maintained above its lowest desirable temperature of minus 2 degrees Centigrade during the survival mode. No software control algorithm for LTCS can be activated during survival and only thermostats can be used. 3) The radiator must be kept above minus 65 degrees Centigrade to prevent ammonia from freezing using no more

Thermal-hydraulics of actinide burner reactors

International Nuclear Information System (INIS)

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

1989-07-01

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

ATLAS Solenoid Integration

CERN Multimedia

Ruber, R

Last month the central solenoid was installed in the barrel cryostat, which it shares with the liquid argon calorimeter. Figure 1: Some members of the solenoid and liquid argon teams proudly pose in front of the barrel cryosat, complete with detector and magnet. Some two years ago the central solenoid arrived at CERN after being manufactured and tested in Japan. It was kept in storage until last October when it was finally moved to the barrel cryostat integration area. Here a position survey of the solenoid (with respect to the cryostat's inner warm vessel) was performed. Figure 2: The alignment survey by Dirk Mergelkuhl and Aude Wiart. (EST-SU) At the start of the New Year the solenoid was moved to the cryostat insertion stand. Figure 3: The solenoid on the insertion stand, with Akira Yamamoto the solenoid designer and project leader. Figure 4: Taka Kondo, ATLAS Japan spokesperson, and Shoichi Mizumaki, Toshiba project engineer for the ATLAS solenoid, celebrate the insertion. Aft...

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

International Nuclear Information System (INIS)

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

1974-06-01

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

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.

Analysis of ATLAS Cold Leg SBLOCA Using SPACE Code

International Nuclear Information System (INIS)

Kang, Doo Hyuk; Suh, Jae Seung; Kim, Se Yun

2012-01-01

SPACE Code has been developed to predict the thermal-hydraulic responses of nuclear steam supply system to the anticipated transients and postulated accidents and adopted advanced physical modeling of two-phase flows, mainly two-fluid, three-field models that comprise gas, continuous liquid, and droplet fields and has the capability to simulate 3D effects by the use of structured and/or non-structured meshes. In this paper, a cold-leg SBLOCA which is the experiment, SB-CL-09, of the ATLAS integral effect test facility during the second domestic stand problem (DSP-02) was analyzed. The results were compared with those of MARS-KS code simulations. The SPACE code with a 1.0 version was released by KHNP in 2012. The analysis has been performed in a desktop PC with Windows 7 environment

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)

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)

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

International Nuclear Information System (INIS)

Yin Hao; Zou Yao; Liu Xingmin

2015-01-01

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

Evolution of thermal-hydraulics testing in EBR-II

International Nuclear Information System (INIS)

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

1987-01-01

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

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2016-07-15

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2005-03-15

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

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

International Nuclear Information System (INIS)

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

2005-03-01

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

CRISSUE-S, Neutronics/Thermal-hydraulics Coupling in LWR Technology

International Nuclear Information System (INIS)

D'Auria, Francesco; Bousbia Salah, Anis; Galassi, G.M.; Vedovi, Juswald; Van Goethem, Georges; Hadek, Jan; Macek, Jiri; Rindelhardt, Udo; Rohde, Ulrich; Ahnert Iglesias, Carol; Aragones Beltran, Jose Maria; Reventos, Francesc; Cuadra, Arantxa; Gago, Jose Luis; Verdu, Gumersindo; Miro, Rafael; Ginestar, Damian; Sanchez, Ana Maria; Sjoberg, Anders; Yitbarek, M.; Sandervag, Oddbjoern; Garis, Ninos; Frid, Wiktor; Panayotov, Dobromir; Ivanov, Kostadin; Uddin, Rizwan; Sartori, Enrico

2004-01-01

Description: The CRISSUE-S project was created with the aim of re-evaluating fundamental technical issues in the technology of LWRs. Specifically, the project seeks to address the interactions between neutron kinetics and thermal-hydraulics that affect neutron moderation and influence the accident performance of the NPPs. This is undertaken in the light of the advanced computational tools that are readily available to the scientific community today. Specifically, the CRISSUE-S activity deals with the control of fission power and the use of high burn up fuel; these topics are part of the EC Work Programme as well as that of other international organisations such as the OECD/NEA and the IAEA. The problems of evaluating reactivity induced accident (RIA) consequences and eventually deciding the possibility of NPP prolongation must be addressed and resolved. RIA constitutes one of the most important of the ?less-resolved? safety issues, and treating this problem may have huge positive financial, social and environmental impacts. Public acceptance of nuclear technology implies that problems such as these be satisfactorily resolved. Cross-disciplinary (regulators, industry, utilities and research bodies) interaction and co operation within CRISSUE-S provides results which can directly and immediately be beneficial to EU industry. Co-operation at an international level: the participation of the EU, former Eastern European countries, the USA, and observers from Japan testify to the broad interest these problems engender. Competencies in broad areas such as thermal-hydraulics, neutronics and fuel, overall system design and reactor surveillance are needed to address the problems that are posed here. Excellent expertise is available in specific areas, while limited knowledge exists in the interface zones of those areas, e.g. in the coupling between thermal-hydraulics and neutronics. In general terms, the activities carried out and described here aim at exploiting available

A Roadmap to Continuous Integration for ATLAS Software Development

Science.gov (United States)

Elmsheuser, J.; Krasznahorkay, A.; Obreshkov, E.; Undrus, A.; ATLAS Collaboration

2017-10-01

The ATLAS software infrastructure facilitates efforts of more than 1000 developers working on the code base of 2200 packages with 4 million lines of C++ and 1.4 million lines of python code. The ATLAS offline code management system is the powerful, flexible framework for processing new package versions requests, probing code changes in the Nightly Build System, migration to new platforms and compilers, deployment of production releases for worldwide access and supporting physicists with tools and interfaces for efficient software use. It maintains multi-stream, parallel development environment with about 70 multi-platform branches of nightly releases and provides vast opportunities for testing new packages, for verifying patches to existing software and for migrating to new platforms and compilers. The system evolution is currently aimed on the adoption of modern continuous integration (CI) practices focused on building nightly releases early and often, with rigorous unit and integration testing. This paper describes the CI incorporation program for the ATLAS software infrastructure. It brings modern open source tools such as Jenkins and GitLab into the ATLAS Nightly System, rationalizes hardware resource allocation and administrative operations, provides improved feedback and means to fix broken builds promptly for developers. Once adopted, ATLAS CI practices will improve and accelerate innovation cycles and result in increased confidence in new software deployments. The paper reports the status of Jenkins integration with the ATLAS Nightly System as well as short and long term plans for the incorporation of CI practices.

Production and integration of the ATLAS Insertable B-Layer

Science.gov (United States)

Abbott, B.; Albert, J.; Alberti, F.; Alex, M.; Alimonti, G.; Alkire, S.; Allport, P.; Altenheiner, S.; Ancu, L. S.; Anderssen, E.; Andreani, A.; Andreazza, A.; Axen, B.; Arguin, J.; Backhaus, M.; Balbi, G.; Ballansat, J.; Barbero, M.; Barbier, G.; Bassalat, A.; Bates, R.; Baudin, P.; Battaglia, M.; Beau, T.; Beccherle, R.; Bell, A.; Benoit, M.; Bermgan, A.; Bertsche, C.; Bertsche, D.; Bilbao de Mendizabal, J.; Bindi, F.; Bomben, M.; Borri, M.; Bortolin, C.; Bousson, N.; Boyd, R. G.; Breugnon, P.; Bruni, G.; Brossamer, J.; Bruschi, M.; Buchholz, P.; Budun, E.; Buttar, C.; Cadoux, F.; Calderini, G.; Caminada, L.; Capeans, M.; Carney, R.; Casse, G.; Catinaccio, A.; Cavalli-Sforza, M.; Červ, M.; Cervelli, A.; Chau, C. C.; Chauveau, J.; Chen, S. P.; Chu, M.; Ciapetti, M.; Cindro, V.; Citterio, M.; Clark, A.; Cobal, M.; Coelli, S.; Collot, J.; Crespo-Lopez, O.; Dalla Betta, G. F.; Daly, C.; D'Amen, G.; Dann, N.; Dao, V.; Darbo, G.; DaVia, C.; David, P.; Debieux, S.; Delebecque, P.; De Lorenzi, F.; de Oliveira, R.; Dette, K.; Dietsche, W.; Di Girolamo, B.; Dinu, N.; Dittus, F.; Diyakov, D.; Djama, F.; Dobos, D.; Dondero, P.; Doonan, K.; Dopke, J.; Dorholt, O.; Dube, S.; Dzahini, D.; Egorov, K.; Ehrmann, O.; Einsweiler, K.; Elles, S.; Elsing, M.; Eraud, L.; Ereditato, A.; Eyring, A.; Falchieri, D.; Falou, A.; Fausten, C.; Favareto, A.; Favre, Y.; Feigl, S.; Fernandez Perez, S.; Ferrere, D.; Fleury, J.; Flick, T.; Forshaw, D.; Fougeron, D.; Franconi, L.; Gabrielli, A.; Gaglione, R.; Gallrapp, C.; Gan, K. K.; Garcia-Sciveres, M.; Gariano, G.; Gastaldi, T.; Gavrilenko, I.; Gaudiello, A.; Geffroy, N.; Gemme, C.; Gensolen, F.; George, M.; Ghislain, P.; Giangiacomi, N.; Gibson, S.; Giordani, M. P.; Giugni, D.; Gjersdal, H.; Glitza, K. W.; Gnani, D.; Godlewski, J.; Gonella, L.; Gonzalez-Sevilla, S.; Gorelov, I.; Gorišek, A.; Gössling, C.; Grancagnolo, S.; Gray, H.; Gregor, I.; Grenier, P.; Grinstein, S.; Gris, A.; Gromov, V.; Grondin, D.; Grosse-Knetter, J.; Guescini, F.; Guido, E.; Gutierrez, P.; Hallewell, G.; Hartman, N.; Hauck, S.; Hasi, J.; Hasib, A.; Hegner, F.; Heidbrink, S.; Heim, T.; Heinemann, B.; Hemperek, T.; Hessey, N. P.; Hetmánek, M.; Hinman, R. R.; Hoeferkamp, M.; Holmes, T.; Hostachy, J.; Hsu, S. C.; Hügging, F.; Husi, C.; Iacobucci, G.; Ibragimov, I.; Idarraga, J.; Ikegami, Y.; Ince, T.; Ishmukhametov, R.; Izen, J. M.; Janoška, Z.; Janssen, J.; Jansen, L.; Jeanty, L.; Jensen, F.; Jentzsch, J.; Jezequel, S.; Joseph, J.; Kagan, H.; Kagan, M.; Karagounis, M.; Kass, R.; Kastanas, A.; Kenney, C.; Kersten, S.; Kind, P.; Klein, M.; Klingenberg, R.; Kluit, R.; Kocian, M.; Koffeman, E.; Korchak, O.; Korolkov, I.; Kostyukhina-Visoven, I.; Kovalenko, S.; Kretz, M.; Krieger, N.; Krüger, H.; Kruth, A.; Kugel, A.; Kuykendall, W.; La Rosa, A.; Lai, C.; Lantzsch, K.; Lapoire, C.; Laporte, D.; Lari, T.; Latorre, S.; Leyton, M.; Lindquist, B.; Looper, K.; Lopez, I.; Lounis, A.; Lu, Y.; Lubatti, H. J.; Maeland, S.; Maier, A.; Mallik, U.; Manca, F.; Mandelli, B.; Mandić, I.; Marchand, D.; Marchiori, G.; Marx, M.; Massol, N.; Mättig, P.; Mayer, J.; McGoldrick, G.; Mekkaoui, A.; Menouni, M.; Menu, J.; Meroni, C.; Mesa, J.; Michal, S.; Miglioranzi, S.; Mikuž, M.; Miucci, A.; Mochizuki, K.; Monti, M.; Moore, J.; Morettini, P.; Morley, A.; Moss, J.; Muenstermann, D.; Murray, P.; Nakamura, K.; Nellist, C.; Nelson, D.; Nessi, M.; Nisius, R.; Nordberg, M.; Nuiry, F.; Obermann, T.; Ockenfels, W.; Oide, H.; Oriunno, M.; Ould-Saada, F.; Padilla, C.; Pangaud, P.; Parker, S.; Pelleriti, G.; Pernegger, H.; Piacquadio, G.; Picazio, A.; Pohl, D.; Polini, A.; Pons, X.; Popule, J.; Portell Bueso, X.; Potamianos, K.; Povoli, M.; Puldon, D.; Pylypchenko, Y.; Quadt, A.; Quayle, B.; Rarbi, F.; Ragusa, F.; Rambure, T.; Richards, E.; Riegel, C.; Ristic, B.; Rivière, F.; Rizatdinova, F.; RØhne, O.; Rossi, C.; Rossi, L. P.; Rovani, A.; Rozanov, A.; Rubinskiy, I.; Rudolph, M. S.; Rummler, A.; Ruscino, E.; Sabatini, F.; Salek, D.; Salzburger, A.; Sandaker, H.; Sannino, M.; Sanny, B.; Scanlon, T.; Schipper, J.; Schmidt, U.; Schneider, B.; Schorlemmer, A.; Schroer, N.; Schwemling, P.; Sciuccati, A.; Seidel, S.; Seiden, A.; Šícho, P.; Skubic, P.; Sloboda, M.; Smith, D. S.; Smith, M.; Sood, A.; Spencer, E.; Stramaglia, M.; Strauss, M.; Stucci, S.; Stugu, B.; Stupak, J.; Styles, N.; Su, D.; Takubo, Y.; Tassan, J.; Teng, P.; Teixeira, A.; Terzo, S.; Therry, X.; Todorov, T.; Tomášek, M.; Toms, K.; Travaglini, R.; Trischuk, W.; Troncon, C.; Troska, G.; Tsiskaridze, S.; Tsurin, I.; Tsybychev, D.; Unno, Y.; Vacavant, L.; Verlaat, B.; Vigeolas, E.; Vogt, M.; Vrba, V.; Vuillermet, R.; Wagner, W.; Walkowiak, W.; Wang, R.; Watts, S.; Weber, M. S.; Weber, M.; Weingarten, J.; Welch, S.; Wenig, S.; Wensing, M.; Wermes, N.; Wittig, T.; Wittgen, M.; Yildizkaya, T.; Yang, Y.; Yao, W.; Yi, Y.; Zaman, A.; Zaidan, R.; Zeitnitz, C.; Ziolkowski, M.; Zivkovic, V.; Zoccoli, A.; Zwalinski, L.

2018-05-01

During the shutdown of the CERN Large Hadron Collider in 2013-2014, an additional pixel layer was installed between the existing Pixel detector of the ATLAS experiment and a new, smaller radius beam pipe. The motivation for this new pixel layer, the Insertable B-Layer (IBL), was to maintain or improve the robustness and performance of the ATLAS tracking system, given the higher instantaneous and integrated luminosities realised following the shutdown. Because of the extreme radiation and collision rate environment, several new radiation-tolerant sensor and electronic technologies were utilised for this layer. This paper reports on the IBL construction and integration prior to its operation in the ATLAS detector.

Thermal-hydraulics associated with nuclear education and research

International Nuclear Information System (INIS)

Yokobori, Seiichi

2011-01-01

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

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

MNRSIM: an interactive visual model which links thermal hydraulics, neutron production and other phenomena

International Nuclear Information System (INIS)

Gilbert, D.; Garland, W.J.; Ha, T.

2004-01-01

The goal for the McMaster Nuclear Reactor Simulator (MNRSIM) is a first order visual approximation of the major elements of the reactor including flux calculations, reactor control, thermal hydraulic calculations and eventually fuel management all within a graphical windows environment. The main purpose in the development of this tool is not to provide the staff and researchers at the reactor with a tool for understanding the reactor as an integrated system of simulations. The tool follows an extensible modular program design. (author)

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

AGIS: Integration of new technologies used in ATLAS Distributed Computing

OpenAIRE

Anisenkov, Alexey; Di Girolamo, Alessandro; Alandes Pradillo, Maria

2017-01-01

The variety of the ATLAS Distributed Computing infrastructure requires a central information system to define the topology of computing resources and to store different parameters and configuration data which are needed by various ATLAS software components. The ATLAS Grid Information System (AGIS) is the system designed to integrate configuration and status information about resources, services and topology of the computing infrastructure used by ATLAS Distributed Computing applications and s...

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

International Nuclear Information System (INIS)

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

2001-01-01

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

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

CERN Document Server

Bottura, L

2016-01-01

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

Simulation of thermal-hydraulic process in reactor of HTR-PM based on flow and heat transfer network

International Nuclear Information System (INIS)

Zhou Kefeng; Zhou Yangping; Sui Zhe; Ma Yuanle

2012-01-01

The development of HTR-PM full scale simulator (FSS) is an important part in the project. The simulation of thermal-hydraulic process in reactor is one of the key technologies in the development of FSS. The simulation of thermal-hydraulic process in reactor was studied. According to the geometry structures and the characteristics of thermal-hydraulic process in reactor, the model was setup in components construction way. Based on the established simulation method of flow and heat transfer network, a Fortran code was developed and the simulation of thermal-hydraulic process was achieved. The simulation results of 50% FP steady state, 100% FP steady state and control rod mistakenly ascension accidents were given. The verification of simulation results was carried out by comparing with the design and analysis code THERMIX. The results show that the method and model based on flow and heat transfer network can meet the requirements of FSS and reflect the features of thermal-hydraulic process in HTR-PM. (authors)

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

Science.gov (United States)

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

2014-05-01

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

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

Energy Technology Data Exchange (ETDEWEB)

NONE

1998-03-01

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

PMK-2, the First Integral Thermal-Hydraulics Tests for the Safety Evaluation of VVER-440/213 Nuclear Power Plants

Directory of Open Access Journals (Sweden)

Gy. Ézsöl

2012-01-01

Full Text Available The PMK-2 facility is a full-pressure thermal-hydraulic model of the primary and partly the secondary circuit of the VVER-type units of Paks NPP. The facility was the first integral-type facility for VVERs. The PMK-2 was followed later by the PACTEL (for VVER-440, the ISB, and PSB for VVER-1000. Since the startup of the facility in 1985, 55 experiments have been performed primarily in international frameworks with the participation of experts from 29 European and overseas countries forming a scientific school to better understand VVER system behaviour and reach a high level of modelling of accident sequences. The ATHLET, CATHARE, and RELAP5 codes have been validated including both qualitative and quantitative assessments. The former was almost exclusively applied to the early phase of validation by integral experiments, while the quantitative assessments have been performed by the Fast Fourier Transform Based Method. Paper gives comprehensive information on the design features of PMK-2 facility with a special respect to the representativeness of phenomena, the experiments performed, and the results of the validation of ATHLET, CATHARE, and RELAP5 codes. Safety significance of the PMK-2 projects is also discussed.

PMK-2, the First Integral Thermal-Hydraulics Tests for the Safety Evaluation of VVER-440/213 Nuclear Power Plants

International Nuclear Information System (INIS)

Ezsol, G.; Perneczky, L.; Szabados, L.; Toth, I.

2012-01-01

The PMK-2 facility is a full-pressure thermal-hydraulic model of the primary and partly the secondary circuit of the VVER-type units of Paks NPP. The facility was the first integral-type facility for VVERs. The PMK-2 was followed later by the PACTEL (for VVER-440), the ISB, and PSB for VVER-1000. Since the startup of the facility in 1985, 55 experiments have been performed primarily in international frameworks with the participation of experts from 29 European and overseas countries forming a scientific school to better understand VVER system behaviour and reach a high level of modelling of accident sequences. The ATHLET, CATHARE, and RELAP5 codes have been validated including both qualitative and quantitative assessments. The former was almost exclusively applied to the early phase of validation by integral experiments, while the quantitative assessments have been performed by the Fast Fourier Transform Based Method. Paper gives comprehensive information on the design features of PMK-2 facility with a special respect to the representativeness of phenomena, the experiments performed, and the results of the validation of ATHLET, CATHARE, and RELAP5 codes. Safety significance of the PMK-2 projects is also discussed

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

International Nuclear Information System (INIS)

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

1988-01-01

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

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

Directory of Open Access Journals (Sweden)

L. Batet

2007-11-01

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

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

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2015-02-15

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2014-09-01

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

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

International Nuclear Information System (INIS)

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

2004-01-01

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

Alternative solution algorithm for coupled thermal-hydraulic problems

International Nuclear Information System (INIS)

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

1986-01-01

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

Development of a computer code, PZRTR, for the thermal hydraulic analysis of a multi-cavity cold gas pressurizer for an integral reactor, SMART-P

Energy Technology Data Exchange (ETDEWEB)

Seo, Jae Kwang; Yoon, J

2003-12-01

The concept of a Multi-cavity Cold Gas PressuriZeR (MCGPZR) is applied to the SMART: The pressurizer system includes in-vessel cavities and out-of-vessel gas cylinders holding the gas supply/vent system. The gas cylinders are connected to the one of the in-vessel cavities via piping with valves. A pressurizer is maintained at a cold temperature of less than about 100 .deg. C, which is realized with coolers installed in and with wet thermal insulators installed on one of the cavities located inside the hot reactor vessel, to minimize the contribution of a steam partial pressure and is filled with nitrogen gas as a pressure-absorbing medium. The working medium and working temperature of the MCGPZR is totally different from that of a hot steam pressurizer of the commercial PWR. In addition, the MCGPZR is intended to be designed to meet a pressure transient during normal power operation (by its gas volume capacity) without using an active control system and during plant heatup/cooldown operation by using an active gas control (filling/venting) system. Therefore in order to evaluate the feasibility of the concept of the MCGPZR and its intended design goal, the thermal hydraulic behaviors and controllability of the MCGPZR during transients especially a heatup/cooldown operation must be analyzed. In this study, a thermal hydraulic transient analysis computer code, PZRTR, for the Reactor Coolant System (RCS) of an integral reactor composed of the MCGPZR, modular Once-Through Steam Generators (OTSGs), a core and a reactor coolant loop is developed. The pressurizer module (MCGPZR module) of the PZRTR code is based on a two-fluid, nonhomogeneous, nonequilibrium model for the two-phase system behavior and the OTSG module is based on a homogeneous equilibrium model of the two-phase flow process. The core module is simply based on the axial power distributions and the reactor coolant loop is based on the temperature distributions. The code is currently dedicated for the

Design and thermal/hydraulic characteristics of the ITER-FEAT vacuum vessel

International Nuclear Information System (INIS)

Onozuka, M.; Ioki, K.; Sannazzaro, G.; Utin, Y.; Yoshimura, H.

2001-01-01

Recent progress in structural design and thermal and hydraulic assessment of the vacuum vessel (VV) for ITER-FEAT is presented. Because of the direct attachment of the blanket modules to the VV, the module support structures are recessed into the double-wall VV, partially replacing the stiffening ribs between the VV shells to simplify the VV structure. Structural integrity of the VV is provided by the ribs and the module support structures with local reinforcement ribs. The detailed structural design of the VV taking account of the fabricability and code/standard acceptance is presented. Cost reduction of the VV fabrication using casting or forging is proposed. A high heat removal capability is required for the VV cooling to keep the thermal stress below the allowable. It is expected that natural thermo-gravitational convection due to the heat flux from the vessel wall to the water will enhance heat transfer characteristics even in the low flow velocity region

Design and thermal/hydraulic characteristics of the ITER-FEAT vacuum vessel

Energy Technology Data Exchange (ETDEWEB)

Onozuka, M. E-mail: onozukm@itereu.de; Ioki, K.; Sannazzaro, G.; Utin, Y.; Yoshimura, H

2001-11-01

Recent progress in structural design and thermal and hydraulic assessment of the vacuum vessel (VV) for ITER-FEAT is presented. Because of the direct attachment of the blanket modules to the VV, the module support structures are recessed into the double-wall VV, partially replacing the stiffening ribs between the VV shells to simplify the VV structure. Structural integrity of the VV is provided by the ribs and the module support structures with local reinforcement ribs. The detailed structural design of the VV taking account of the fabricability and code/standard acceptance is presented. Cost reduction of the VV fabrication using casting or forging is proposed. A high heat removal capability is required for the VV cooling to keep the thermal stress below the allowable. It is expected that natural thermo-gravitational convection due to the heat flux from the vessel wall to the water will enhance heat transfer characteristics even in the low flow velocity region.

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

Science.gov (United States)

2013-02-05

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

Thermal-hydraulic tests on net divertor targets using swirl tubes

International Nuclear Information System (INIS)

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

1991-01-01

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

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

International Nuclear Information System (INIS)

Martin, R.P.

1993-05-01

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

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

International Nuclear Information System (INIS)

Andrade Lima, F.R. de

1993-01-01

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

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-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

Optimised Iteration in Coupled Monte Carlo - Thermal-Hydraulics Calculations

Science.gov (United States)

Hoogenboom, J. Eduard; Dufek, Jan

2014-06-01

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

Optimized iteration in coupled Monte-Carlo - Thermal-hydraulics calculations

International Nuclear Information System (INIS)

Hoogenboom, J.E.; Dufek, J.

2013-01-01

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

A study on the thermal hydraulics in rod bundles

International Nuclear Information System (INIS)

Chung, Moon Ki; Yang, Sun Kyu

1989-03-01

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

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

International Nuclear Information System (INIS)

Jimenez, J.; Herrero, J. J.; Cuervo, D.; Aragones, J. M.

2010-10-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2015-08-15

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

Whole Core Thermal-Hydraulic Design of a Sodium Cooled Fast Reactor Considering the Gamma Energy Transport

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 performance of the Atlas SCT forward modules

CERN Document Server

Clark, A; Nasteva, I; Snow, S W; Wallny, R; Wilmut, I

2003-01-01

We describe the thermal design of the Atlas SCT forward modules and their cooling blocks. We report on the performance of the $C_3 F_8$ evaporative cooling system and the blocks alone, then on the performance of an irradiated inner module mounted on two alternative prototype cooling blocks (baseline and PEEK split). Runs are presented at different cooling conditions, representative of those expected to be used in the final experiment. We have also measured thermal runaway, with the module mounted on the PEEK split block and cooled with liquid cooling.

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

International Nuclear Information System (INIS)

Vondy, D.R.

1984-07-01

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

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

Energy Technology Data Exchange (ETDEWEB)

Vondy, D.R.

1984-07-01

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

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

Instrumentation and Control Systems for Sodium thermal hydraulic Experiment Loop for Finned-tube sodium-to-Air heat exchanger (SELFA)

Energy Technology Data Exchange (ETDEWEB)

Kim, Byeong Yeon; Kim, Hyung Mo; Cho, Youn Gil; Kim, Jong Man; Ko, Yung Joo; Kang, Byeong Su; Jung, Min Hwan; Jeong, Ji Young [KAERI, Daejeon (Korea, Republic of)

2016-05-15

A forced-draft sodium-to-air heat exchanger (FHX) is a part of decay heat removal system (DHRS) in Prototype Gen-IV Sodium-cooled fast reactor (PGSFR), which is being developed at Korea Atomic Energy Research Institute (KAERI). Sodium thermal hydraulic Experiment Loop for Finned-tube sodium-to-Air heat exchanger (SELFA) is a test facility for verification and validation of the design code for a forced-draft sodium-to-air heat exchanger (FHX). In this paper, we have provided design and fabrication features for the instrumentation and control systems of SELFA. In general, the instrumentation systems and control systems are coupled for measurement and control of process variables. Instrumentation systems have been designed for investigating thermal-hydraulic characteristics of FHX and control systems have been designed to control the main components (e.g. electromagnetic pumps, heaters, valves etc.) required for test in SELFA. In this paper, we have provided configurations of instrumentation and control systems for Sodium thermal hydraulic Experiment Loop for Finned-tube sodium-to-Air heat exchanger (SELFA). The instrumentation and control systems of SELFA have been implemented based on the expected operation ranges and lesson learned from operational experience of 'Sodium integral effect test loop for safety simulation and assessment-1' (STELLA-1)

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

International Nuclear Information System (INIS)

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

1991-01-01

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

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

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

International Nuclear Information System (INIS)

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

2011-01-01

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

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

International Nuclear Information System (INIS)

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

2011-01-01

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

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

International Nuclear Information System (INIS)

Mur, J.; Meignin, J.C.

1997-07-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

1997-07-01

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

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

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

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.

The Phebus FP thermal-hydraulic analysis with Melcor

International Nuclear Information System (INIS)

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

1995-01-01

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

An improved thermal-hydraulic modeling of the Jules Horowitz Reactor using the CATHARE2 system code

Energy Technology Data Exchange (ETDEWEB)

Pegonen, R., E-mail: pegonen@kth.se [KTH Royal Institute of Technology, Roslagstullsbacken 21, SE-10691 Stockholm (Sweden); Bourdon, S.; Gonnier, C. [CEA, DEN, DER, SRJH, CEA Cadarache, 13108 Saint-Paul-lez-Durance Cedex (France); Anglart, H. [KTH Royal Institute of Technology, Roslagstullsbacken 21, SE-10691 Stockholm (Sweden)

2017-01-15

Highlights: • An improved thermal-hydraulic modeling of the JHR reactor is described. • Thermal-hydraulics of the JHR is analyzed during loss of flow accident. • The heat exchanger approach gives more realistic and less conservative results. - Abstract: The newest European high performance material testing reactor, the Jules Horowitz Reactor, will support current and future nuclear reactor designs. The reactor is under construction at the CEA Cadarache research center in southern France and is expected to achieve first criticality at the end of this decade. This paper presents an improved thermal-hydraulic modeling of the reactor using solely CATHARE2 system code. Up to now, the CATHARE2 code was simulating the full reactor with a simplified approach for the core and the boundary conditions were transferred into the three-dimensional FLICA4 core simulation. A new more realistic methodology is utilized to analyze the thermal-hydraulic simulation of the reactor during a loss of flow accident.

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)

Classification and modelling of functional outputs of computation codes. Application to accidental thermal-hydraulic calculations in pressurized water reactor (PWR)

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)

Analysis of PBMR transients using a coupled neutron transport/thermal-hydraulics code DORT-TD/thermix

International Nuclear Information System (INIS)

Tyobeka, B.; Ivanov, K.; Pautz, A.

2007-01-01

In the advent of increased demand for safety and economics of nuclear power plants, nuclear engineers and designers are called upon to develop advanced computation tools. In these developments, space-time effects in the dynamics of nuclear reactors must be considered within the framework of a full 3-dimensional treatment of both neutron kinetics and thermal hydraulics. In a recent effort at the Pennsylvania State University, a time-dependent version of the discrete ordinates transport code DORT, DORT-TD was coupled to a 2-dimensional core thermal hydraulics code THERMIX-DIREKT. In the coupling process, a feedback model was developed to account for the feedback effects and was implemented into DORT-TD. During the calculation process for each spatial node of the DORT-TD core model, feedback parameters representative of this node are passed to the feedback module. Using these values, cross section tables are then interpolated for the appropriate macroscopic cross section values. The updated macroscopic cross sections are passed back to DORT-TD to perform transport core calculations, and the power distribution is transferred to THERMIX-DIREKT to obtain the relevant thermal-hydraulics data in turn, and this calculation loop continues. In this paper, DORT-TD/THERMIX is used to simulate transients of interest in the PBMR (Pebble Bed Modular Reactor) safety using established benchmark problems: load change from 100% to 40% power and fast control rod ejection (PBMR-268 benchmark problem). The results obtained are compared with those obtained using the diffusion-based module of the code. The results are only preliminary and so far show that diffusion theory is not such a bad approximation for PBMR for the prediction of integral parameters

Establishment of International Cooperative Network and Cooperative Research Strategy Between Korea and USA on Nuclear Thermal Hydraulics

International Nuclear Information System (INIS)

Baek, Won Pil; Song, Chul Hwa; Jeong, Jae Jun; Choi, Ki Yong; Kang, Kyoung Ho

2004-07-01

1. Scope and Objectives of the Project - Successful holding of the NURETH-10 - Analysis of the international trends in technology development and applications for nuclear thermal-hydraulics - Establishment of the international cooperative network and cooperative research strategy between Korea and USA on nuclear thermal-hydraulics 2. Research Results - Successful holding of the NURETH-10 - Analysis of the international trends in technology development and applications for nuclear thermal-hydraulics: - Establishment of international cooperative network and cooperative research strategy focused between Korea and USA on nuclear thermal-hydraulics: 3. Application Plan of the Research Results - Utilization as the basic data/information in establishing the domestic R and D directions and the international cooperative research strategy, - Application of the relevant experiences and data bases of NURETH-10 for holding future international conferences, - Promote more effective and productive research cooperation between Korea and USA

Proceedings of the workshop on advanced thermal-hydraulic and neutronic codes: current and future applications

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

AGIS: Integration of new technologies used in ATLAS Distributed Computing

CERN Document Server

AUTHOR|(INSPIRE)INSPIRE-00291854; The ATLAS collaboration; Di Girolamo, Alessandro; Alandes Pradillo, Maria

2017-01-01

The variety of the ATLAS Distributed Computing infrastructure requires a central information system to define the topology of computing resources and to store different parameters and configuration data which are needed by various ATLAS software components. The ATLAS Grid Information System (AGIS) is the system designed to integrate configuration and status information about resources, services and topology of the computing infrastructure used by ATLAS Distributed Computing applications and services. Being an intermediate middleware system between clients and external information sources (like central BDII, GOCDB, MyOSG), AGIS defines the relations between experiment specific used resources and physical distributed computing capabilities. Being in production during LHC Runl AGIS became the central information system for Distributed Computing in ATLAS and it is continuously evolving to fulfil new user requests, enable enhanced operations and follow the extension of the ATLAS Computing model. The ATLAS Computin...

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

International Nuclear Information System (INIS)

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

1983-04-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

1995-06-01

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

Thermal hydraulics in the hot pool of Fast Breeder Test Reactor

International Nuclear Information System (INIS)

Padmakumar, G.; Pandey, G.K.; Vaidyanathan, G.

2009-01-01

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

ATWS thermal-hydraulic analysis for Krsko Full Scope Simulator validation

International Nuclear Information System (INIS)

Parzer, I.; Kljenak, I.

2005-01-01

The purpose of this analysis was to simulate Anticipated Transient without Scram transient for Krsko NPP. The results of these calculations were used for annual ANSI/ANS validation of reactor coolant system thermal-hydraulic response predicted by Krsko Full Scope Simulator. For the thermal-hydraulic analyses the RELAP5/MOD3.3 code and the input model for NPP Krsko, delivered by NPP Krsko, was used. In the presented paper the most severe ATWS scenario has been analyzed, starting with the loss of Main Feedwater at both steam generators. Thus, gradual loss of secondary heat sink occurred. On top of that, control rods were not supposed to scram, leaving the chain reaction to be controlled only by inherent physical properties of the fuel and moderator and eventual actions of the BOP system. The primary system response has been studied assuming AMSAC availability. (author)

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

International Nuclear Information System (INIS)

Ikeda, K.; Hoshi, M.

2001-01-01

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

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

International Nuclear Information System (INIS)

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

2011-01-01

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

Current status and future prospects for thermal-hydraulics and safety research

International Nuclear Information System (INIS)

Park, G.C.

2000-01-01

The present paper is to outline the current activities in Korea for the thermal-hydraulics and safety researches, and furthermore illuminate the future aspect of those field under the umbrella of worldwide nuclear prospect. In Korea, a long-term nuclear research plan has been established since 1992, which was recently funded with a fixed monetary rate of Korean won 1.20 per kWh of electricity produced with nuclear power. 11.5% of the fund is assigned for nuclear safety research in 6 areas. Under this program, 3 axes of research body (KAERI, KINS, University) has been operated with close cooperation. Their role, current activities and long-term plan of each body are introduced in the point of thermal-hydraulics' view. (author)

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

International Nuclear Information System (INIS)

Ninokata, H.; Kamide, H.

2011-01-01

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

Integration of Flex Nozzle System and Electro Hydraulic Actuators to Solid Rocket Motors

Science.gov (United States)

Nayani, Kishore Nath; Bajaj, Dinesh Kumar

2017-10-01

A rocket motor assembly comprised of solid rocket motor and flex nozzle system. Integration of flex nozzle system and hydraulic actuators to the solid rocket motors are done after transportation to the required place where integration occurred. The flex nozzle system is integrated to the rocket motor in horizontal condition and the electro hydraulic actuators are assembled to the flex nozzle systems. The electro hydraulic actuators are connected to the hydraulic power pack to operate the actuators. The nozzle-motor critical interface are insulation diametrical compression, inhibition resin-28, insulation facial compression, shaft seal `O' ring compression and face seal `O' ring compression.

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

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

Development of fuel performance and thermal hydraulic technology

International Nuclear Information System (INIS)

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

2000-03-01

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

Comparison of the DVI line break LOCA with the equivalent cold leg break with the ATLAS facility

International Nuclear Information System (INIS)

Choi, K. Y.; Cho, S.; Kang, K. H.; Park, H. S.; Kim, Y. S.; Baek, W. P.

2010-01-01

The APR1400 (Advanced Power Reactor, 1400 MWe) adopts a DVI (Direct Vessel Injection) method for ECC (Emergency Core Cooling) water delivery rather than a conventional CLI (Cold Leg Injection) method as an advanced safety feature. The break scenario of one DVI nozzle is taken into account in the small break LOCA analysis. Transient behavior during the DVI line breaks needs to be investigated and compared with the equivalent break on the cold leg. An 8.5-inch double-ended break of one DVI nozzle was simulated with the ATLAS, and a counterpart test for the DVI break was performed at the cold leg with the equivalent break size for comparison. This comparison will contribute to enhancing a comprehensive understanding of the thermal hydraulic behavior during transients. A constructed integral effect database is also used to validate the existing conservative safety analysis methodology and to develop a best-estimate safety analysis methodology for small-break LOCAs. A post-test calculation was performed with a best-estimate safety analysis code, MARS 3.1, in order to examine its prediction capability and to identify any code deficiencies for thermal hydraulic phenomena occurring during the transient. (authors)

Integration of Detectors Into a Large Experiment: Examples From ATLAS and CMS

CERN Document Server

Froidevaux, D

2011-01-01

Integration of Detectors Into a Large Experiment: Examples From ATLAS andCMS, part of 'Landolt-Börnstein - Group I Elementary Particles, Nuclei and Atoms: Numerical Data and Functional Relationships in Science and Technology, Volume 21B2: Detectors for Particles and Radiation. Part 2: Systems and Applications'. This document is part of Part 2 'Principles and Methods' of Subvolume B 'Detectors for Particles and Radiation' of Volume 21 'Elementary Particles' of Landolt-Börnstein - Group I 'Elementary Particles, Nuclei and Atoms'. It contains the Chapter '5 Integration of Detectors Into a Large Experiment: Examples From ATLAS and CMS' with the content: 5 Integration of Detectors Into a Large Experiment: Examples From ATLAS and CMS 5.1 Introduction 5.1.1 The context 5.1.2 The main initial physics goals of ATLAS and CMS at the LHC 5.1.3 A snapshot of the current status of the ATLAS and CMS experiments 5.2 Overall detector concept and magnet systems 5.2.1 Overall detector concept 5.2.2 Magnet systems 5.2.2.1 Rad...

Proceedings of the 10th international topical meeting on nuclear thermal hydraulics, operation and safety (NUTHOS-10)

International Nuclear Information System (INIS)

2014-01-01

The 10th International Topical Meeting on Nuclear Thermal Hydraulics, Operations and Safety (NUTHOS-10) in Okinawa, Japan is sponsored by Atomic Energy Society of Japan, in cooperation with the International Atomic Energy Agency, and co-sponsored by American Nuclear Society Thermal Hydraulics Division among others. Enhanced safety and reducing cost are going together, which can be achieved through continued research and development efforts. NUTHOS keeps you abreast of the most updated information in the advancement of science and technology in nuclear thermal hydraulics, operations and safety, and provides you insights into the future. (J.P.N.)

Cyberinfrastructure for the digital brain: spatial standards for integrating rodent brain atlases.

Science.gov (United States)

Zaslavsky, Ilya; Baldock, Richard A; Boline, Jyl

2014-01-01

Biomedical research entails capture and analysis of massive data volumes and new discoveries arise from data-integration and mining. This is only possible if data can be mapped onto a common framework such as the genome for genomic data. In neuroscience, the framework is intrinsically spatial and based on a number of paper atlases. This cannot meet today's data-intensive analysis and integration challenges. A scalable and extensible software infrastructure that is standards based but open for novel data and resources, is required for integrating information such as signal distributions, gene-expression, neuronal connectivity, electrophysiology, anatomy, and developmental processes. Therefore, the International Neuroinformatics Coordinating Facility (INCF) initiated the development of a spatial framework for neuroscience data integration with an associated Digital Atlasing Infrastructure (DAI). A prototype implementation of this infrastructure for the rodent brain is reported here. The infrastructure is based on a collection of reference spaces to which data is mapped at the required resolution, such as the Waxholm Space (WHS), a 3D reconstruction of the brain generated using high-resolution, multi-channel microMRI. The core standards of the digital atlasing service-oriented infrastructure include Waxholm Markup Language (WaxML): XML schema expressing a uniform information model for key elements such as coordinate systems, transformations, points of interest (POI)s, labels, and annotations; and Atlas Web Services: interfaces for querying and updating atlas data. The services return WaxML-encoded documents with information about capabilities, spatial reference systems (SRSs) and structures, and execute coordinate transformations and POI-based requests. Key elements of INCF-DAI cyberinfrastructure have been prototyped for both mouse and rat brain atlas sources, including the Allen Mouse Brain Atlas, UCSD Cell-Centered Database, and Edinburgh Mouse Atlas Project.

Cyberinfrastructure for the digital brain: spatial standards for integrating rodent brain atlases

Directory of Open Access Journals (Sweden)

Ilya eZaslavsky

2014-09-01

Full Text Available Biomedical research entails capture and analysis of massive data volumes and new discoveries arise from data-integration and mining. This is only possible if data can be mapped onto a common framework such as the genome for genomic data. In neuroscience, the framework is intrinsically spatial and based on a number of paper atlases. This cannot meet today’s data-intensive analysis and integration challenges. A scalable and extensible software infrastructure that is standards based but open for novel data and resources, is required for integrating information such as signal distributions, gene-expression, neuronal connectivity, electrophysiology, anatomy, and developmental processes. Therefore, the International Neuroinformatics Coordinating Facility (INCF initiated the development of a spatial framework for neuroscience data integration with an associated Digital Atlasing Infrastructure (DAI. A prototype implementation of this infrastructure for the rodent brain is reported here. The infrastructure is based on a collection of reference spaces to which data is mapped at the required resolution, such as the Waxholm Space (WHS, a 3D reconstruction of the brain generated using high-resolution, multi-channel microMRI. The core standards of the digital atlasing service-oriented infrastructure include Waxholm Markup Language (WaxML: XML schema expressing a uniform information model for key elements such as coordinate systems, transformations, points of interest (POIs, labels, and annotations; and Atlas Web Services: interfaces for querying and updating atlas data. The services return WaxML-encoded documents with information about capabilities, spatial reference systems and structures, and execute coordinate transformations and POI-based requests. Key elements of INCF-DAI cyberinfrastructure have been prototyped for both mouse and rat brain atlas sources, including the Allen Mouse Brain Atlas, UCSD Cell-Centered Database, and Edinburgh Mouse Atlas

Thermal-hydraulic research plan for Babcock and Wilcox plants

International Nuclear Information System (INIS)

Lee, R.Y.

1988-02-01

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

Integration and test of the ATLAS Semiconductor Tracker

CERN Document Server

Pernegger, H

2007-01-01

The ATLAS Semiconductor Tracker (SCT) will be a central part of the tracking system of the ATLAS experiment and is one of the major new silicon detector systems for LHC. The paper summarizes the system integration of the SCT from individual components to the completed tracker barrel and endcaps ready for installation in the pit. Particular attention will be given to the test results obtained during the different integration steps: from single barrels and disks to the final tests inside the ID before installation in the pit. The tests provided us with operational experience for a significant fraction of the full detector system and showed the very good performance of the final assembled detector.

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

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.

Numerical simulations of subcritical reactor kinetics in thermal hydraulic transient phases

Energy Technology Data Exchange (ETDEWEB)

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

1999-12-31

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

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)

Proceedings of the 7th International Meeting on Nuclear Reactor Thermal-Hydraulics NURETH-7. Sessions 17-24

Energy Technology Data Exchange (ETDEWEB)

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

1995-09-01

Technical papers accepted for presentation at the Seventh International Topical Meeting on Nuclear Reactor Thermal-Hydraulics are included in the present Proceedings. Except for the invited papers in the plenary session, all other papers are contributed papers. The topics of the meeting encompass all major areas of nuclear thermal-hydraulics, including analytical and experimental works on the fundamental mechanisms of fluid flow and heat transfer, the development of advanced mathematical and numerical methods, and the application of advancements in the field in the development of novel reactor concepts. Because of the complex nature of nuclear reactors and power plants, several papers deal with the combined issues of thermal-hydraulics and reactor/power-plant safety, core neutronics and/or radiation. The participation in the conference by the authors from several countries and four continents makes the Proceedings a comprehensive review of the recent progress in the field of nuclear reactor thermal-hydraulics worldwide. Individual papers have been cataloged separately.

Proceedings of the 7th International Meeting on Nuclear Reactor Thermal-Hydraulics NURETH-7. Sessions 17-24

International Nuclear Information System (INIS)

Block, R.C.; Feiner, F.

1995-09-01

Technical papers accepted for presentation at the Seventh International Topical Meeting on Nuclear Reactor Thermal-Hydraulics are included in the present Proceedings. Except for the invited papers in the plenary session, all other papers are contributed papers. The topics of the meeting encompass all major areas of nuclear thermal-hydraulics, including analytical and experimental works on the fundamental mechanisms of fluid flow and heat transfer, the development of advanced mathematical and numerical methods, and the application of advancements in the field in the development of novel reactor concepts. Because of the complex nature of nuclear reactors and power plants, several papers deal with the combined issues of thermal-hydraulics and reactor/power-plant safety, core neutronics and/or radiation. The participation in the conference by the authors from several countries and four continents makes the Proceedings a comprehensive review of the recent progress in the field of nuclear reactor thermal-hydraulics worldwide. Individual papers have been cataloged separately

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

International Nuclear Information System (INIS)

Fujiki, Kazuo; Asaka, Hideaki; Ishida, Toshihisa

1986-01-01

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

Development of a Computer Code, PZRTR rev 1, for the Thermal Hydraulic Analysis of a Multi-Cavity Cold Gas Pressurizer for an Integral Reactor, SMART-P

Energy Technology Data Exchange (ETDEWEB)

Seo, Jae Kwang; Kang, H. O.; Yoon, J.; Kim, K. K

2006-12-15

The concept of a Multi-cavity Cold Gas PressuriZeR(MCGPZR) is applied to the SMART: The pressurizer system includes in-vessel cavities and out-of-vessel gas cylinders holding the gas supply/vent system. The gas cylinders are connected to the one of the in-vessel cavities via piping with valves. A pressurizer is maintained at a cold temperature of less than about 120 .deg. C which is realized with coolers installed in and with wet thermal insulators installed on one of the cavities located inside the hot reactor vessel, to minimize the contribution of a steam partial pressure and is filled with nitrogen gas as a pressure-absorbing medium. The working medium and working temperature of the MCGPZR is totally different from that of a hot steam pressurizer of the commercial PWR. In addition, the MCGPZR is intended to be designed to meet a pressure transient during normal power operation (by its gas volume capacity) without using an active control system and during plant heatup/cooldown operation by using an active gas control (filling/venting) system. Therefore in order to evaluate the feasibility of the concept of the MCGPZR and its intended design goal, the thermal hydraulic behaviors and controllability of the MCGPZR during transients especially a heatup/cooldown operation must be analyzed. In this study, a thermal hydraulic transient analysis computer code, PZRTR rev 1, for the Reactor Coolant System(RCS) of an integral reactor composed of the MCGPZR, modular Once-Through Steam Generators(OTSGs), a core and a reactor coolant loop is developed. The pressurizer module (MCGPZR module) of the PZRTR rev 1 code is based on a two-fluid, nonhomogeneous, nonequilibrium model for the two-phase system behavior and the OTSG module is based on a homogeneous equilibrium model of the two-phase flow process. The core module is simply based on the axial power distributions and the reactor coolant loop is based on the temperature distributions. The code is currently dedicated for the

A new approach to designing reduced scale thermal-hydraulic experiments

International Nuclear Information System (INIS)

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

2004-01-01

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

Applications for coupled core neutronics and thermal-hydraulic models

International Nuclear Information System (INIS)

Eller, J.

1996-01-01

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

arXiv Production and Integration of the ATLAS Insertable B-Layer

CERN Document Server

Abbott, B.; Alberti, F.; Alex, M.; Alimonti, G.; Alkire, S.; Allport, P.; Altenheiner, S.; Ancu, L.S.; Anderssen, E.; Andreani, A.; Andreazza, A.; Axen, B.; Arguin, J.; Backhaus, M.; Balbi, G.; Ballansat, J.; Barbero, M.; Barbier, G.; Bassalat, A.; Bates, R.; Baudin, P.; Battaglia, M.; Beau, T.; Beccherle, R.; Bell, A.; Benoit, M.; Bermgan, A.; Bertsche, C.; Bertsche, D.; Bilbao de Mendizabal, J.; Bindi, F.; Bomben, M.; Borri, M.; Bortolin, C.; Bousson, N.; Boyd, R.G.; Breugnon, P.; Bruni, G.; Brossamer, J.; Bruschi, M.; Buchholz, P.; Budun, E.; Buttar, C.; Cadoux, F.; Calderini, G.; Caminada, L.; Capeans, M.; Carney, R.; Casse, G.; Catinaccio, A.; Cavalli-Sforza, M.; Červ, M.; Cervelli, A.; Chau, C.C.; Chauveau, J.; Chen, S.P.; Chu, M.; Ciapetti, M.; Cindro, V.; Citterio, M.; Clark, A.; Cobal, M.; Coelli, S.; Collot, J.; Crespo-Lopez, O.; Dalla Betta, G.F.; Daly, C.; D'Amen, G.; Dann, N.; Dao, V.; Darbo, G.; DaVia, C.; David, P.; Debieux, S.; Delebecque, P.; De Lorenzi, F.; de Oliveira, R.; Dette, K.; Dietsche, W.; Di Girolamo, B.; Dinu, N.; Dittus, F.; Diyakov, D.; Djama, F.; Dobos, D.; Dondero, P.; Doonan, K.; Dopke, J.; Dorholt, O.; Dube, S.; Dzahini, D.; Egorov, K.; Ehrmann, O.; Einsweiler, K.; Elles, S.; Elsing, M.; Eraud, L.; Ereditato, A.; Eyring, A.; Falchieri, D.; Falou, A.; Fausten, C.; Favareto, A.; Favre, Y.; Feigl, S.; Fernandez Perez, S.; Ferrere, D.; Fleury, J.; Flick, T.; Forshaw, D.; Fougeron, D.; Franconi, L.; Gabrielli, A.; Gaglione, R.; Gallrapp, C.; Gan, K.K.; Garcia-Sciveres, M.; Gariano, G.; Gastaldi, T.; Gavrilenko, I.; Gaudiello, A.; Geffroy, N.; Gemme, C.; Gensolen, F.; George, M.; Ghislain, P.; Giangiacomi, N.; Gibson, S.; Giordani, M.P.; Giugni, D.; Gjersdal, H.; Glitza, K.W.; Gnani, D.; Godlewski, J.; Gonella, L.; Gonzalez-Sevilla, S.; Gorelov, I.; Gorišek, A.; Gössling, C.; Grancagnolo, S.; Gray, H.; Gregor, I.; Grenier, P.; Grinstein, S.; Gris, A.; Gromov, V.; Grondin, D.; Grosse-Knetter, J.; Guescini, F.; Guido, E.; Gutierrez, P.; Hallewell, G.; Hartman, N.; Hauck, S.; Hasi, J.; Hasib, A.; Hegner, F.; Heidbrink, S.; Heim, T.; Heinemann, B.; Hemperek, T.; Hessey, N.P.; Hetmánek, M.; Hinman, R.R.; Hoeferkamp, M.; Holmes, T.; Hostachy, J.; Hsu, S.C.; Hügging, F.; Husi, C.; Iacobucci, G.; Ibragimov, I.; Idarraga, J.; Ikegami, Y.; Ince, T.; Ishmukhametov, R.; Izen, J.M.; Janoška, Z.; Janssen, J.; Jansen, L.; Jeanty, L.; Jensen, F.; Jentzsch, J.; Jezequel, S.; Joseph, J.; Kagan, H.; Kagan, M.; Karagounis, M.; Kass, R.; Kastanas, A.; Kenney, C.; Kersten, S.; Kind, P.; Klein, M.; Klingenberg, R.; Kluit, R.; Kocian, M.; Koffeman, E.; Korchak, O.; Korolkov, I.; Kostyukhina-Visoven, I.; Kovalenko, S.; Kretz, M.; Krieger, N.; Krüger, H.; Kruth, A.; Kugel, A.; Kuykendall, W.; La Rosa, A.; Lai, C.; Lantzsch, K.; Lapoire, C.; Laporte, D.; Lari, T.; Latorre, S.; Leyton, M.; Lindquist, B.; Looper, K.; Lopez, I.; Lounis, A.; Lu, Y.; Lubatti, H.J.; Maeland, S.; Maier, A.; Mallik, U.; Manca, F.; Mandelli, B.; Mandić, I.; Marchand, D.; Marchiori, G.; Marx, M.; Massol, N.; Mättig, P.; Mayer, J.; Mc Goldrick, G.; Mekkaoui, A.; Menouni, M.; Menu, J.; Meroni, C.; Mesa, J.; Michal, S.; Miglioranzi, S.; Mikuž, M.; Miucci, A.; Mochizuki, K.; Monti, M.; Moore, J.; Morettini, P.; Morley, A.; Moss, J.; Muenstermann, D.; Murray, P.; Nakamura, K.; Nellist, C.; Nelson, D.; Nessi, M.; Nisius, R.; Nordberg, M.; Nuiry, F.; Obermann, T.; Ockenfels, W.; Oide, H.; Oriunno, M.; Ould-Saada, F.; Padilla, C.; Pangaud, P.; Parker, S.; Pelleriti, G.; Pernegger, H.; Piacquadio, G.; Picazio, A.; Pohl, D.; Polini, A.; Pons, X.; Popule, J.; Portell Bueso, X.; Potamianos, K.; Povoli, M.; Puldon, D.; Pylypchenko, Y.; Quadt, A.; Quayle, B.; Rarbi, F.; Ragusa, F.; Rambure, T.; Richards, E.; Riegel, C.; Ristic, B.; Rivière, F.; Rizatdinova, F.; Rø hne, O.; Rossi, C.; Rossi, L.P.; Rovani, A.; Rozanov, A.; Rubinskiy, I.; Rudolph, M.S.; Rummier, A.; Ruscino, E.; Sabatini, F.; Salek, D.; Salzburger, A.; Sandaker, H.; Sannino, M.; Sanny, B.; Scanlon, T.; Schipper, J.; Schmidt, U.; Schneider, B.; Schorlemmer, A.; Schroer, N.; Schwemling, P.; Sciuccati, A.; Seidel, S.; Seiden, A.; Šícho, P.; Skubic, P.; Sloboda, M.; Smith, D.S.; Smith, M.; Sood, A.; Spencer, E.; Stramaglia, M.; Strauss, M.; Stucci, S.; Stugu, B.; Stupak, J.; Styles, N.; Su, D.; Takubo, Y.; Tassan, J.; Teng, P.; Teixeira, A.; Terzo, S.; Therry, X.; Todorov, T.; Tomášek, M.; Toms, K.; Travaglini, R.; Trischuk, W.; Troncon, C.; Troska, G.; Tsiskaridze, S.; Tsurin, I.; Tsybychev, D.; Unno, Y.; Vacavant, L.; Verlaat, B.; Vigeolas, E.; Vogt, M.; Vrba, V.; Vuillermet, R.; Wagner, W.; Walkowiak, W.; Wang, R.; Watts, S.; Weber, M.S.; Weber, M.; Weingarten, J.; Welch, S.; Wenig, S.; Wensing, M.; Wermes, N.; Wittig, T.; Wittgen, M.; Yildizkaya, T.; Yang, Y.; Yao, W.; Yi, Y.; Zaman, A.; Zaidan, R.; Zeitnitz, C.; Ziolkowski, M.; Zivkovic, V.; Zoccoli, A.; Zwalinski, L.

2018-05-16

During the shutdown of the CERN Large Hadron Collider in 2013-2014, an additional pixel layer was installed between the existing Pixel detector of the ATLAS experiment and a new, smaller radius beam pipe. The motivation for this new pixel layer, the Insertable B-Layer (IBL), was to maintain or improve the robustness and performance of the ATLAS tracking system, given the higher instantaneous and integrated luminosities realised following the shutdown. Because of the extreme radiation and collision rate environment, several new radiation-tolerant sensor and electronic technologies were utilised for this layer. This paper reports on the IBL construction and integration prior to its operation in the ATLAS detector.

Design and thermal-hydraulic calculation for EAST PFCs' baking

International Nuclear Information System (INIS)

Wan Xiaogang; Yao Damao

2006-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2014-06-01

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

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

International Nuclear Information System (INIS)

Foucher, L.; Cousin, F.; Fleurot, J.; Brethes, S.

2014-01-01

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

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

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

International Nuclear Information System (INIS)

Perez, Daniel Milian; Lorenzo, Daniel E. Milian; Lira, Carlos A. Brayner de Oliveira; Garcia, Lorena P. Rodríguez; Universidade Federal de Pernambuco

2017-01-01

The use of Aqueous Homogenous Reactors (AHR) is one of the most promissory alternatives to produce medical isotopes, mainly "9"9Mo. Compare to multipurpose research reactors, an AHR dedicated for "9"9Mo production has advantages because of their low cost, small critical mass, inherent passive safety, and simplified fuel handling, processing, and purification characteristics. This article presents the current state of research in our working group on this topic. Are presented and discussed the group validation efforts with benchmarking exercises that include neutronic and thermal-hydraulic results of two solution reactors, the SUPO and ARGUS reactors. Neutronic and thermal-hydraulic results of 75 kWth AHR based on the ARGUS reactor LEU configuration are presented. The neutronic studies included the determination of parameters such as reflector thickness, critical height, medical isotopes production and others. Thermal-hydraulics studies were focused on demonstrating that sufficient cooling capacity exists to prevent fuel overheating. In addition, the effects of some calculation parameters on the computational modeling of temperature, velocity and gas volume fraction during steady-state operation of an AHR are discussed. The neutronic and thermal-hydraulics studies have been performed with the MCNPX version 2.6e computational code and the version 14 of ANSYS CFX respectively. Our group studies and the results obtained contribute to demonstrate the feasibility of using AHR for the production of medical isotopes, however additional studies are still necessary to confirm these results and contribute to development and demonstration of their technical, safety, and economic viability. (author)

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

Energy Technology Data Exchange (ETDEWEB)

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

2017-11-01

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

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)

Integration of the Chinese HPC Grid in ATLAS Distributed Computing

CERN Document Server

AUTHOR|(INSPIRE)INSPIRE-00081160; The ATLAS collaboration

2016-01-01

Fifteen Chinese High Performance Computing sites, many of them on the TOP500 list of most powerful supercomputers, are integrated into a common infrastructure providing coherent access to a user through an interface based on a RESTful interface called SCEAPI. These resources have been integrated into the ATLAS Grid production system using a bridge between ATLAS and SCEAPI which translates the authorization and job submission protocols between the two environments. The ARC Computing Element (ARC CE) forms the bridge using an extended batch system interface to allow job submission to SCEAPI. The ARC CE was setup at the Institute for High Energy Physics, Beijing, in order to be as close as possible to the SCEAPI front-end interface at the Computing Network Information Center, also in Beijing. This paper describes the technical details of the integration between ARC CE and SCEAPI and presents results so far with two supercomputer centers, Tianhe-IA and ERA. These two centers have been the pilots for ATLAS Monte C...

Integration of the Chinese HPC Grid in ATLAS Distributed Computing

CERN Document Server

AUTHOR|(INSPIRE)INSPIRE-00081160

2017-01-01

Fifteen Chinese High-Performance Computing sites, many of them on the TOP500 list of most powerful supercomputers, are integrated into a common infrastructure providing coherent access to a user through an interface based on a RESTful interface called SCEAPI. These resources have been integrated into the ATLAS Grid production system using a bridge between ATLAS and SCEAPI which translates the authorization and job submission protocols between the two environments. The ARC Computing Element (ARC-CE) forms the bridge using an extended batch system interface to allow job submission to SCEAPI. The ARC-CE was setup at the Institute for High Energy Physics, Beijing, in order to be as close as possible to the SCEAPI front-end interface at the Computing Network Information Center, also in Beijing. This paper describes the technical details of the integration between ARC-CE and SCEAPI and presents results so far with two supercomputer centers, Tianhe-IA and ERA. These two centers have been the pilots for ATLAS Monte C...

3-D thermal hydraulic analysis of transient heat removal from fast reactor core using immersion coolers

International Nuclear Information System (INIS)

Chvetsov, I.; Volkov, A.

2000-01-01

For advanced fast reactors (EFR, BN-600M, BN-1600, CEFR) the special complementary loop is envisaged in order to ensure the decay heat removal from the core in the case of LOF accidents. This complementary loop includes immersion coolers that are located in the hot reactor plenum. To analyze the transient process in the reactor when immersion coolers come into operation one needs to involve 3-D thermal hydraulics code. Furthermore sometimes the problem becomes more complicated due to necessity of simulation of the thermal hydraulics processes into the core interwrapper space. For example on BN-600M and CEFR reactors it is supposed to ensure the effective removal of decay heat from core subassemblies by specially arranged internal circulation circuit: 'inter-wrapper space'. For thermal hydraulics analysis of the transients in the core and in the whole reactor including hot plenum with immersion coolers and considering heat and mass exchange between the main sodium flow and sodium that moves in the inter-wrapper space the code GRIFIC (the version of GRIF code family) was developed in IPPE. GRIFIC code was tested on experimental data obtained on RAMONA rig under conditions simulating decay heat removal of a reactor with the use of immersion coolers. Comparison has been made of calculated and experimental result, such as integral characteristics (flow rate through the core and water temperature at the core inlet and outlet) and the local temperatures (at thermocouple location) as well. In order to show the capabilities of the code some results of the transient analysis of heat removal from the core of BN-600M - type reactor under loss-of-flow accident are presented. (author)

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

International Nuclear Information System (INIS)

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

1978-10-01

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

Development of thermal hydraulic models for the reliable regulatory auditing code

Energy Technology Data Exchange (ETDEWEB)

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

2004-02-15

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

Review of the nuclear reactor thermal hydraulic research in ocean motions

Energy Technology Data Exchange (ETDEWEB)

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

2017-03-15

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

Review of the nuclear reactor thermal hydraulic research in ocean motions

International Nuclear Information System (INIS)

Yan, B.H.

2017-01-01

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

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.

Thermal-hydraulic interfacing code modules for CANDU reactors

International Nuclear Information System (INIS)

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

1997-01-01

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

A review of the current thermal-hydraulic modeling of the Jules Horowitz Reactor: A loss of flow accident analysis

International Nuclear Information System (INIS)

Pegonen, R.; Bourdon, S.; Gonnier, C.; Anglart, H.

2014-01-01

Highlights: • CEA methodology for thermal-hydraulic calculations in the JHR reactor is described. • Thermal-hydraulics of the JHR is analyzed during LOFA using CATHARE and FLICA4. • Safety criteria, important modeling parameters and correlations are presented. • Possible improvements of the current methodology are discussed and proposed. - Abstract: The newest European high performance material testing reactor, the Jules Horowitz Reactor, will support existing and future nuclear reactor designs. The reactor is under construction at CEA Cadarache research center in France and is expected to start operation at the end of this decade. R and D and analytical works have already been performed to set-up the methodology for thermal-hydraulic calculations of the reactor. This paper presents the off-line coupled thermal-hydraulic modeling of the reactor using the CATHARE system code and the FLICA4 core analysis code. The main objective of the present work is to analyze the thermal-hydraulic calculations of the reactor during the loss of flow accident using CEA methodology. Possible improvements of the current methodology are shortly discussed and suggested

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.)

Thermal-hydraulic software development for nuclear waste transportation cask design and analysis

International Nuclear Information System (INIS)

Brown, N.N.; Burns, S.P.; Gianoulakis, S.E.; Klein, D.E.

1991-01-01

This paper describes the development of a state-of-the-art thermal-hydraulic software package intended for spent fuel and high-level nuclear waste transportation cask design and analysis. The objectives of this software development effort are threefold: (1) to take advantage of advancements in computer hardware and software to provide a more efficient user interface, (2) to provide a tool for reducing inefficient conservatism in spent fuel and high-level waste shipping cask design by including convection as well as conduction and radiation heat transfer modeling capabilities, and (3) to provide a thermal-hydraulic analysis package which is developed under a rigorous quality assurance program established at Sandia National Laboratories. 20 refs., 5 figs., 2 tabs

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

Energy Technology Data Exchange (ETDEWEB)

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

1998-04-15

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

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 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

Techniques for the thermal/hydraulic analysis of LMFBR check valves

International Nuclear Information System (INIS)

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

1979-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2001-07-15

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

Methodology for thermal-hydraulics analysis of pool type MTR fuel research reactors

International Nuclear Information System (INIS)

Umbehaun, Pedro Ernesto

2000-01-01

This work presents a methodology developed for thermal-hydraulic analysis of pool type MTR fuel research reactors. For this methodology a computational program, FLOW, and a model, MTRCR-IEAR1 were developed. FLOW calculates the cooling flow distribution in the fuel elements, control elements, irradiators, and through the channels formed among the fuel elements and among the irradiators and reflectors. This computer program was validated against experimental data for the IEA-R1 research reactor core at IPEN-CNEN/SP. MTRCR-IEAR1 is a model based on the commercial program Engineering Equation Solver (EES). Besides the thermal-hydraulic analyses of the core in steady state accomplished by traditional computational programs like COBRA-3C/RERTR and PARET, this model allows to analyze parallel channels with different cooling flow and/or geometry. Uncertainty factors of the variables from neutronic and thermalhydraulic calculation and also from the fabrication of the fuel element are introduced in the model. For steady state analyses MTRCR-IEAR1 showed good agreement with the results of COBRA-3C/RERTR and PARET. The developed methodology was used for the calculation of the cooling flow distribution and the thermal-hydraulic analysis of a typical configuration of the IEA-R1 research reactor core. (author)

Development of core thermal-hydraulics module for intelligent reactor design system (IRDS)

International Nuclear Information System (INIS)

Kugo, Teruhiko; Nakagawa, Masayuki; Fujii, Sadao.

1994-08-01

We have developed an innovative reactor core thermal-hydraulics module where a designer can easily and efficiently evaluate his design concept of a new type reactor in the thermal-hydraulics field. The main purpose of this module is to decide a feasible range of basic design parameters of a reactor core in a conceptual design stage of a new type reactor. The module is to be implemented in Intelligent Reactor Design System (IRDS). The module has the following characteristics; 1) to deal with several reactor types, 2) four thermal hydraulics and fuel behavior analysis codes are installed to treat different type of reactors and design detail, 3) to follow flexibly modification of a reactor concept, 4) to provide analysis results in an understandable way so that a designer can easily evaluate feasibility of his concept, and so on. The module runs on an engineering workstation (EWS) and has a user-friendly man-machine interface on a pre- and post-processing. And it is equipped with a function to search a feasible range called as Design Window, for two design parameters by artificial intelligence (AI) technique and knowledge engineering. In this report, structure, guidance for users of an usage of the module and instruction of input data for analysis modules are presented. (author)

FIIND: Ferret Interactive Integrated Neurodevelopment Atlas

Directory of Open Access Journals (Sweden)

Roberto Toro

2018-03-01

Full Text Available The first days after birth in ferrets provide a privileged view of the development of a complex mammalian brain. Unlike mice, ferrets develop a rich pattern of deep neocortical folds and cortico- cortical connections. Unlike humans and other primates, whose brains are well differentiated and folded at birth, ferrets are born with a very immature and completely smooth neocortex: folds, neocortical regionalisation and cortico-cortical connectivity develop in ferrets during the first postnatal days. After a period of fast neocortical expansion, during which brain volume increases by up to a factor of 4 in 2 weeks, the ferret brain reaches its adult volume at about 6 weeks of age. Ferrets could thus become a major animal model to investigate the neurobiological correlates of the phenomena observed in human neuroimaging. Many of these phenomena, such as the relationship between brain folding, cortico-cortical connectivity and neocortical regionalisation cannot be investigated in mice, but could be investigated in ferrets. Our aim is to provide the research community with a detailed description of the development of a complex brain, necessary to better understand the nature of human neuroimaging data, create models of brain development, or analyse the relationship between multiple spatial scales. We have already started a project to constitute an open, collaborative atlas of ferret brain development, integrating multi-modal and multi-scale data. We have acquired data for 28 ferrets (4 animals per time point from P0 to adults, using high-resolution MRI and diffusion tensor imaging (DTI. We have developed an open-source pipeline to segment and produce – online – 3D reconstructions of brain MRI data. We propose to process the brains of 16 of our specimens (from P0 to P16 using high-throughput 3D histology, staining for cytoarchitectonic landmarks, neuronal progenitors and neurogenesis. This would allow us to relate the MRI data that we have already

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

Energy Technology Data Exchange (ETDEWEB)

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

1997-07-01

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

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.

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

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

International Nuclear Information System (INIS)

Iguchi, Tadashi; Itoh, Hideo; Kiuchi, Toshio; Watanabe, Hironori; Kimura, Mamoru; Anoda, Yoshinari

2001-03-01

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

Thermal hydraulic evaluation of advanced wire-wrapped assemblies

International Nuclear Information System (INIS)

Wei, J.P.

1975-01-01

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

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.

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

International Nuclear Information System (INIS)

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

1980-08-01

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

Transitioning from interpretive to predictive in thermal hydraulic codes

International Nuclear Information System (INIS)

Mousseau, V.A.

2004-01-01

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

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

International Nuclear Information System (INIS)

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

1998-01-01

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

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

International Nuclear Information System (INIS)

Akimoto, Hajime; Kukita; Ohnuki, Akira

1997-01-01

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

Computer code for the thermal-hydraulic analysis of ITU TRIGA Mark-II reactor

International Nuclear Information System (INIS)

Ustun, G.; Durmayaz, A.

2002-01-01

Istanbul Technical University (ITU) TRIGA Mark-II reactor core consists of ninety vertical cylindrical elements located in five rings. Sixty-nine of them are fuel elements. The reactor is operated and cooled with natural convection by pool water, which is also cooled and purified in external coolant circuits by forced convection. This characteristic leads to consider both the natural and forced convection heat transfer in a 'porous-medium analysis'. The safety analysis of the reactor requires a thermal-hydraulic model of the reactor to determine the thermal-hydraulic parameters in each mode of operation. In this study, a computer code cooled TRIGA-PM (TRIGA - Porous Medium) for the thermal-hydraulic analysis of ITU is considered. TRIGA Mark-II reactor code has been developed to obtain velocity, pressure and temperature distributions in the reactor pool as a function of core design parameters and pool configuration. The code is a transient, thermal-hydraulic code and requires geometric and physical modelling parameters. In the model, although the reactor is considered as only porous medium, the other part of the reactor pool is considered partly as continuum and partly as porous medium. COMMIX-1C code is used for the benchmark purpose of TRIGA-PM code. For the normal operating conditions of the reactor, estimations of TRIGA-PM are in good agreement with those of COMMIX-1C. After some more improvements, this code will be employed for the estimation of LOCA scenario, which can not be analyses by COMMIX-1C and the other multi-purpose codes, considering a break at one of the beam tubes of the reactor

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.

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

300 kWt core conceptual model thermal/hydraulic characteristics

International Nuclear Information System (INIS)

Moody, E.

1971-01-01

The 300 kW(t), 199 element NASA-Lewis/AEC core conceptual model, has been analyzed to determine it's thermal-hydraulic characteristics using the GEOM-3 code. Stack-ups of tolerances and fuel rod asymmetry patterns were used to ascertain cross element Δ T's. Both zoned and uniform spacing were analyzed with a somewhat lower fuel temperature and cross element ΔT found for zoned spacing. With the models considered, the core design appears adequate to limit thermal gradients to approximately 32 0 F. Bypass flow should be controlled to prevent excessive edge element ΔT's. 11 references. (U.S.)

Coupled neutronics/thermal-hydraulics for analysis of molten salt reactor

International Nuclear Information System (INIS)

Guo, Zhangpeng; Zhou, Jianjun; Zhang, Dalin; Chaudri, Khurrum Saleem; Tian, Wenxi; Su, Guanghui; Qiu, Suizheng

2013-01-01

Highlights: ► A multiple-channel analysis code (MAC) is developed to be coupled with MCNP. ► 1/8 of core is simulated in MCNP and thermal-hydraulic code. ► The coupling calculation can achieve stable state after a few iterations. ► The coupling calculation results are in reasonable agreement with the analytic solutions of the ORNL. ► Parametric studies of MSR are performed to provide valuable information for future design MSR. -- Abstract: The Generation IV International Forum (GIF) selected molten salt reactor (MSR) among six advanced reactor types. It is characterized by a liquid circulating fuel that also serves as coolant. In this study, a multiple-channel analysis code (MAC) is developed and it is coupled with MCNP4c to analyze the neutronics/thermal-hydraulics behavior of molten salt reactor experiment (MSRE). The MAC calculates thermal-hydraulic parameters, such as temperature distribution, flow distribution and pressure drop. MCNP4c performs the analysis of effective multiplication factor, neutron flux and power distribution. A linkage code is developed to exchange data between MAC and MCNP to implement coupling iteration process until the power convergence is achieved. The coupling calculation can achieve converged solution after a few iterations. The results are in reasonable agreement with the analytic solutions from the ORNL. For further design analysis, parametric studies are performed to provide valuable information for new design of MSR. The effect of inlet temperature, graphite to molten salt volume ratio (G/Ms) from varying channel diameter and different power levels on the effective multiplication factor, neutron flux, graphite lifetime and temperature distribution are discussed in detail

The U.S. Nuclear Regulatory Commission Thermal-Hydraulic Research Program: Maintaining expertise in a changing environment

International Nuclear Information System (INIS)

Sheron, B.W.; Shotkin, L.M.; Baratta, A.J.

1993-01-01

Throughout the 1970s and early 1980s, the U.S. Nuclear Regulatory Commission's (NRC's) thermal-hydraulic research program enjoyed ample funding, sponsored extensive experimental and analytical development programs, and attracted worldwide expertise. With the completion of the major experimental programs and with the promulgation of the revised emergency core-cooling system rule, both the funding and prominence of thermal-hydraulic research at the NRC have declined in recent years. This has led justifiably to the concern by some that the program may no longer have the minimal elements needed to maintain both expertise and world-class status. The purpose of this article is to describe the NRC's current thermal-hydraulic research program and to show how this program ensures maintenance of a viable, robust research effort and retention of needed expertise and international leadership

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

International Nuclear Information System (INIS)

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

1976-01-01

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

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

Directory of Open Access Journals (Sweden)

Snehasis Tripathy

2017-07-01

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

ATHENA [Advanced Thermal Hydraulic Energy Network Analyzer] solutions to developmental assessment problems

International Nuclear Information System (INIS)

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

1987-03-01

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

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

International Nuclear Information System (INIS)

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

1981-05-01

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

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

International Nuclear Information System (INIS)

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

1988-09-01

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

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

Science.gov (United States)

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

1988-09-01

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

Integration of the Chinese HPC Grid in ATLAS Distributed Computing

Science.gov (United States)

Filipčič, A.; ATLAS Collaboration

2017-10-01

Fifteen Chinese High-Performance Computing sites, many of them on the TOP500 list of most powerful supercomputers, are integrated into a common infrastructure providing coherent access to a user through an interface based on a RESTful interface called SCEAPI. These resources have been integrated into the ATLAS Grid production system using a bridge between ATLAS and SCEAPI which translates the authorization and job submission protocols between the two environments. The ARC Computing Element (ARC-CE) forms the bridge using an extended batch system interface to allow job submission to SCEAPI. The ARC-CE was setup at the Institute for High Energy Physics, Beijing, in order to be as close as possible to the SCEAPI front-end interface at the Computing Network Information Center, also in Beijing. This paper describes the technical details of the integration between ARC-CE and SCEAPI and presents results so far with two supercomputer centers, Tianhe-IA and ERA. These two centers have been the pilots for ATLAS Monte Carlo Simulation in SCEAPI and have been providing CPU power since fall 2015.

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

Interaction between thermal/hydraulics, human factors and system analysis for assessing feed and bleed risk benefits

International Nuclear Information System (INIS)

Lanore, J.M.; Caron, J.L.

1987-11-01

For probabilistic analysis of accident sequences, thermal/hydraulics, human factors and systems operation problems are frequently closely interrelated. This presentation will discuss a typical example which illustrates this interrelation: total loss of feedwater flow. It will present thermal/hydraulic analysises performed, how the T/H analysises are related to human factors and systems operation, and how, based on this, the failure probability of the feed and bleed cooling mode was evaluated

Integral effect test and code analysis on the cooling performance of the PAFS (passive auxiliary feedwater system) during an FLB (feedwater line break) accident

International Nuclear Information System (INIS)

Bae, Byoung-Uhn; Kim, Seok; Park, Yu-Sun; Kang, Kyoung-Ho

2014-01-01

Highlights: • This study focuses on the experimental validation of the operational performance of the PAFS (passive auxiliary feedwater system). • A transient simulation of the FLB (feedwater line break) in the integral effect test facility, ATLAS-PAFS, was performed to investigate thermal hydraulic behavior during the PAFS actuation. • The test result confirmed that the APR+ has the capability of coping with the FLB scenario by adopting the PAFS and proper set-points for its operation. • The experimental result was utilized to evaluate the prediction capability of a thermal hydraulic system analysis code, MARS-KS. - Abstract: APR+ (Advanced Power Reactor Plus), which is a GEN-III+ nuclear power plant developed in Korea, adopts PAFS (passive auxiliary feedwater system) as an advanced safety feature. The PAFS can completely replace an active auxiliary feedwater system by cooling down the secondary side of steam generators with a natural convection mechanism. This study focuses on experimental and analytical investigation for cooling and operational performance of the PAFS during an FLB (feedwater line break) transient with an integral effect test facility, ATLAS-PAFS. To realistically simulate the FLB accident of the APR+, the three-level scaling methodology was taken into account to design the test facility and determine the test condition. From the test result, the PAFS was actuated to successfully cool down the decay heat of the reactor core by the condensation heat transfer at the PCHX (passive condensation heat exchanger), and thus it could be confirmed that the APR+ has the capability of coping with a FLB scenario by adopting the PAFS and proper set-points for its operation. This integral effect test data were used to evaluate the prediction capability of a thermal hydraulic system analysis code, MARS-KS. The code analysis result proved that it could reasonably predict the FLB transient including the actuation of the PAFS and the natural convection

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

Energy Technology Data Exchange (ETDEWEB)

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

1997-07-01

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

Creation of computerized 3D MRI-integrated atlases of the human basal ganglia and thalamus

Directory of Open Access Journals (Sweden)

Abbas F. Sadikot

2011-09-01

Full Text Available Functional brain imaging and neurosurgery in subcortical areas often requires visualization of brain nuclei beyond the resolution of current Magnetic Resonance Imaging (MRI methods. We present techniques used to create: 1 a lower resolution 3D atlas, based on the Schaltenbrand and Wahren print atlas, which was integrated into a stereotactic neurosurgery planning and visualization platform (VIPER; and 2 a higher resolution 3D atlas derived from a single set of manually segmented histological slices containing nuclei of the basal ganglia, thalamus, basal forebrain and medial temporal lobe. Both atlases were integrated to a canonical MRI (Colin27 from a young male participant by manually identifying homologous landmarks. The lower resolution atlas was then warped to fit the MRI based on the identified landmarks. A pseudo-MRI representation of the high-resolution atlas was created, and a nonlinear transformation was calculated in order to match the atlas to the template MRI. The atlas can then be warped to match the anatomy of Parkinson’s disease surgical candidates by using 3D automated nonlinear deformation methods. By way of functional validation of the atlas, the location of the sensory thalamus was correlated with stereotactic intraoperative physiological data. The position of subthalamic electrode positions in patients with Parkinson’s disease was also evaluated in the atlas-integrated MRI space. Finally, probabilistic maps of subthalamic stimulation electrodes were developed, in order to allow group analysis of the location of contacts associated with the best motor outcomes. We have therefore developed, and are continuing to validate, a high-resolution computerized MRI-integrated 3D histological atlas, which is useful in functional neurosurgery, and for functional and anatomical studies of the human basal ganglia, thalamus and basal forebrain.

Thermal-hydraulic studies of the Advanced Neutron Source cold source

International Nuclear Information System (INIS)

Williams, P.T.; Lucas, A.T.

1995-08-01

The Advanced Neutron Source (ANS), in its conceptual design phase at Oak Ridge National Laboratory, was to be a user-oriented neutron research facility producing the most intense steady-state flux of thermal and cold neutrons in the world. Among its many scientific applications, the production of cold neutrons was a significant research mission for the ANS. The cold neutrons come from two independent cold sources positioned near the reactor core. Contained by an aluminum alloy vessel, each cold source is a 410-mm-diam sphere of liquid deuterium that functions both as a neutron moderator and a cryogenic coolant. With nuclear heating of the containment vessel and internal baffling, steady-state operation requires close control of the liquid deuterium flow near the vessel's inner surface. Preliminary thermal-hydraulic analyses supporting the cold source design were performed with heat conduction simulations of the vessel walls and multidimensional computational fluid dynamics simulations of the liquid deuterium flow and heat transfer. This report presents the starting phase of a challenging program and describes the cold source conceptual design, the thermal-hydraulic feasibility studies of the containment vessel, and the future computational and experimental studies that were planned to verify the final design

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

Development of a detailed BWR core thermal-hydraulic analysis method based on the Japanese post-BT standard using a best-estimate code

International Nuclear Information System (INIS)

Ono, H.; Mototani, A.; Kawamura, S.; Abe, N.; Takeuchi, Y.

2004-01-01

The post-BT standard is a new fuel integrity standard or the Atomic Energy Society of Japan that allows temporary boiling transition condition in the evaluation for BWR anticipated operational occurrences. For application of the post-BT standard to BWR anticipated operational occurrences evaluation, it is important to identify which fuel assemblies and which axial, radial positions of fuel rods have temporarily experienced the post-BT condition and to evaluates how high the fuel cladding temperature rise was and how long the dryout duration continued. Therefore, whole bundle simulation, in which each fuel assembly is simulated independently by one thermal-hydraulic component, is considered to be an effective analytical method. In the present study, a best-estimate thermal-hydraulic code, TRACG02, has been modified to extend it predictive capability by implementing the post-BT evaluation model such as the post-BT heat transfer correlation and rewetting correlation and enlarging the number of components used for BWR plant simulation. Based on new evaluation methods, BWR core thermal-hydraulic behavior has been analyzed for typical anticipated operational occurrence conditions. The location where boiling transition occurs and the severity of fuel assembly in the case of boiling transition conditions such as fuel cladding temperature, which are important factors in determining whether the reuse of the fuel assembly can be permitted, were well predicted by the proposed evaluation method. In summary, a new evaluation method for a detailed BWR core thermal-hydraulic analysis based on the post-BT standard of the Atomic Energy Society of Japan has been developed and applied to the evaluation of the post-BT standard during the actual BWR plant anticipated operational occurrences. (author)

State of the art of thermal-hydraulics of BWRs

International Nuclear Information System (INIS)

Rouhani, Z.

1980-10-01

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

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

International Nuclear Information System (INIS)

1980-01-01

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

RAMONA-3B/MINET composite representation of BWR thermal-hydraulic systems

International Nuclear Information System (INIS)

Van Tuyle, G.J.; Slovik, G.; Cazzoli, E.G.; Nepsee, T.C.; Guppy, J.G.

1985-01-01

The modification and interfacing of two computer codes, RAMONA-3B and MINET, for the thermal hydraulic transient analysis of a Boiling Water Reactor nuclear steam supply system, is described. The RAMONA-3B code provides for multi-channel thermal hydraulics and three-dimensional (or one-dimensional) neutron kinetics analysis of a boiling water reactor core. The RAMONA-3B system representation terminates at the end of the steam line and at the junction of the feedwater line at the vessel inlet. By interfacing RAMONA-3B with MINET, a generic balance-of-plant systems analysis code, a complete BWR systems code with detailed core modeling was obtained. The result is a code of particular importance to the analysis of transients such as ATWS. A comparison between the 3-D and 1-D neutronics representation is provided, along with a test case utilizing the composite RAMONA-3B/MINET code

Thermal hydraulic analysis of Pb-Bi cooled HYPER fuel assemblies using SLTHEN code

International Nuclear Information System (INIS)

Tak, Nam Il; Song, Tae Y.; Park, Won S.; Kim, Chang Hyun

2002-12-01

In the present work, the existing SLTHEN code, which had been originally developed for subchannel analysis of sodium cooled fast reactors, was modified and applied to the Pb-Bi cooled HYPER core which consists of 237 fuel assemblies (TRU assemblies). In the analysis of single fuel assembly having chopped cosine power profile, the validation and the assessment of usefulness of the modified SLTHEN were focused. In the quantitative comparison, the results of the modified SLTHEN agreed well with those of analytical calculations and of MATRA. For the qualitative approaches, the sensitivity calculations for intra-assembly gap flow and turbulent mixing parameter were used. The sensitivity analysis results showed that the modified SLTHEN can provide reasonable simulations of subchannel thermal hydraulics. In particular, turbulent mixing parameter which is known as the most uncertain parameter in subchannel analyses did not affect largely the maximum cladding temperature. Therefore, it can be said that the results of single assembly show the usefulness of the modified SLTHEN code for thermal hydraulic analysis and design of HYPER under the conceptual design stage. In order to assess intra-assembly heat transfer, subchannel analyses were implemented for two types of 7 assemblies; 1) artificial 7 fuel assemblies to maximize intra-assembly heat transfer, 2) central 7 fuel assemblies in the HYPER reference core. The results showed that the modified SLTHEN can reasonably simulate intra-heat transfer and the amount of intra-assembly heat transfer is not so large in HYPER conditions. Particularly, intra-heat transfer did not affect the maximum coolant and the maximum cladding temperatures which are major parameters in conceptual core designs. The capability of full core thermal hydraulic analysis was confirmed by the analysis of 45 fuel assemblies in 1/6 HYPER core at the first cycle. The SLTHEN predicted that the reference design parameters are acceptable in terms of thermal

Improvement of computer programs 'BAMBOO' and 'ASFRE-IV' for coupling analysis of deformation and thermal-hydraulics in a high burn-up fuel subassembly of fast reactor

International Nuclear Information System (INIS)

Uwaba, Tomoyuki; Ohshima, Hiroyuki; Imai, Yasutomo

2003-04-01

A simulation system of a deformed fuel subassembly is being developed for the structure integrity of high burn-up wire-spacer-type fuel subassemblies of sodium-cooled fast breeder reactors. This report describes a computer program improvement work for coupling analyses of deformation and thermal-hydraulics in a fuel subassembly as part of the simulation system development. In this work, a function of data conversion as an interface between a bundle deformation analysis program BAMBOO and a thermal hydraulic analysis program ASFRE-IV was incorporated to each program. BAMBOO was improved to accept the coolant temperature data from ASFRE-IV and to offer bundle deformation data to ASFRE-IV. ASFRE-IV was also improved to offer the coolant temperature data to BAMBOO and to obtain the bundle deformation data from BAMBOO. Improved BAMBOO and ASFRE-IV were applied to an analysis of 169-pin bundle for the program verification. It was confirmed that the coupling analysis gave the physically reasonable results on both deformation and thermal hydraulic behaviors in the fuel subassembly. (author)

Thermal hydraulics analysis of LIBRA-SP target chamber

International Nuclear Information System (INIS)

Mogahed, E.A.

1996-01-01

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

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)

Minerve: thermal-hydraulic phenomena simulation and virtual reality

Energy Technology Data Exchange (ETDEWEB)

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

2003-07-01

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

TEMP-M program for thermal-hydraulic calculation of fast reactor fuel assemblies

International Nuclear Information System (INIS)

Bogoslovskaya, C.P.; Sorokin, A.P.; Tikhomirov, B.B.; Titov, P.A.; Ushakov, P.A.

1983-01-01

TEMP-M program (Fortran, BESM-6 computer) for thermal-hydraulic calculation of fast reactor fuel assemblies is described. Results of calculation of temperature field in a 127 fuel element assembly of BN-600, reactor accomplished according to TEMP-N program are considered as an example. Algorithm, realized in the program, enables to calculate the distributions of coolant heating, fuel element temperature (over perimeter and length) and assembly shell temperature. The distribution of coolant heating in assembly channels is determined from a solution of the balance equation system which accounts for interchannel exchange, nonadiabatic conditions on the assembly shell. The TEMP-M program gives necessary information for calculation of strength, seviceability of fast reactor core elements, serves an effective instrument for calculations when projecting reactor cores and analyzing thermal-hydraulic characteristics of operating reactor fuel assemblies

Development of steady thermal-hydraulic analysis code for China advanced research reactor

International Nuclear Information System (INIS)

Tian Wenxi; Qiu Suizheng; Guo Yun; Su Guanghui; Jia Dounan; Liu Tiancai; Zhang Jianwei

2006-01-01

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

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

International Nuclear Information System (INIS)

Tanaka, F.; Hibiki, T.; Saito, Y.; Takeda, T.; Mishima, K.

2001-01-01

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

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

International Nuclear Information System (INIS)

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

1994-01-01

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

Proceedings of the 7th International Meeting on Nuclear Reactor Thermal-Hydraulics NURETH-7. Volume 3, Sessions 12-16

Energy Technology Data Exchange (ETDEWEB)

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

1995-09-01

This document, Volume 3, includes papers presented at the 7th International Meeting on Nuclear Reactor Thermal-Hydraulics (NURETH-7) September 10--15, 1995 at Saratoga Springs, N.Y. The following subjects are discussed: Progress in analytical and experimental work on the fundamentals of nuclear thermal-hydraulics, the development of advanced mathematical and numerical methods, ad the application of advancements in the field in the development of novel reactor concepts. Also combined issues of thermal-hydraulics and reactor/power-plant safety, core neutronics and/or radiation. Selected abstracts have been indexed separately for inclusion in the Energy Science and Technology Database.

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

International Nuclear Information System (INIS)

Block, R.C.; Feiner, F.

1995-09-01

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

Proceedings of the 7th International Meeting on Nuclear Reactor Thermal-Hydraulics NURETH-7. Volume 2, Sessions 6-11

Energy Technology Data Exchange (ETDEWEB)

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

1995-09-01

This document, Volume 2, includes papers presented at the 7th International Meeting on Nuclear Reactor Thermal-Hydraulics (NURETH-7) September 10--15, 1995 at Saratoga Springs, N.Y. The following subjects are discussed: Progress in analytical and experimental work on the fundamentals of nuclear thermal-hydraulics, the development of advanced mathematical and numerical methods, and the application of advancements in the field in the development of novel reactor concepts. Also combined issues of thermal-hydraulics and reactor/power-plant safety, core neutronics and/or radiation. Selected papers are indexed separately for inclusion in the Energy Science and Technology Database.

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

Energy Technology Data Exchange (ETDEWEB)

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

1995-09-01

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

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

Highly-Integrated Hydraulic Smart Actuators and Smart Manifolds for High-Bandwidth Force Control

Directory of Open Access Journals (Sweden)

Victor Barasuol

2018-06-01

Full Text Available Hydraulic actuation is the most widely used alternative to electric motors for legged robots and manipulators. It is often selected for its high power density, robustness and high-bandwidth control performance that allows the implementation of force/impedance control. Force control is crucial for robots that are in contact with the environment, since it enables the implementation of active impedance and whole body control that can lead to a better performance in known and unknown environments. This paper presents the hydraulic Integrated Smart Actuator (ISA developed by Moog in collaboration with IIT, as well as smart manifolds for rotary hydraulic actuators. The ISA consists of an additive-manufactured body containing a hydraulic cylinder, servo valve, pressure/position/load/temperature sensing, overload protection and electronics for control and communication. The ISA v2 and ISA v5 have been specifically designed to fit into the legs of IIT’s hydraulic quadruped robots HyQ and HyQ-REAL, respectively. The key features of these components tackle 3 of today’s main challenges of hydraulic actuation for legged robots through: (1 built-in controllers running inside integrated electronics for high-performance control, (2 low-leakage servo valves for reduced energy losses, and (3 compactness thanks to metal additive manufacturing. The main contributions of this paper are the derivation of the representative dynamic models of these highly integrated hydraulic servo actuators, a control architecture that allows for high-bandwidth force control and their experimental validation with application-specific trajectories and tests. We believe that this is the first work that presents additive-manufactured, highly integrated hydraulic smart actuators for robotics.

Thermal-hydraulic considerations for particle bed reactors

Science.gov (United States)

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

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

Preliminary thermal-hydraulic and structural strength analyses for pre-moderator of cold moderator

International Nuclear Information System (INIS)

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

2001-08-01

A light-water cooled pre-moderator with a thin-walled structure made of aluminum alloy is installed around a liquid hydrogen moderator in order to enhance the neutron performance of a MW-scale spallation target system which is being developed in the Japan Atomic Energy Research Institute (JAERI). Since the pre-moderator is needed to be located close to a target working as a neutron source, it is indispensable to remove nuclear heat deposition in the pre-moderator effectively by means of smooth water flow without flow stagnation. Also, the structural integrity of the thin-walled structure should be kept against the water pressure. Preliminary thermal-hydraulic analytical results showed that the water temperature rise could be suppressed less than 1degC while keeping the smooth water flow, which would assure the expected neutron performance. As for the structural integrity, several measures to meet allowable stress conditions of aluminum alloy were proposed on the basis of the preliminary structural strength analyses. (author)

Nuclear fuel element design and thermal-hydraulic analysis of Wolsung-1, 600 MWe CANDU-PHWR (Part II)

International Nuclear Information System (INIS)

Suk, H.C; Lee, J.C.; Suh, K.S.; Yuk, K.E.; Whang, W.; Park, J.S.; Eim, J.S.; Bang, K.H.; Eim, M.S.; Rim, C.S.

1982-01-01

The main objective of the present thermal hydraulic analysis is to determine the thermal hydraulic characteristics of Wolsung-1 600 MWe CANDU-PHW reactor under normal operation. This is to verify and expedite the development of the nuclear fuel design and fabrication as well as the management. The computer program package developed for the stated objective are DOD81, CANREPP, PLOC81 and COBRA-CANDU. (Author)

Multi-dimensional Analysis for SLB Transient in ATLAS Facility as Activity of DSP (Domestic Standard Problem)

International Nuclear Information System (INIS)

Bae, B. U.; Park, Y. S.; Kim, J. R.; Kang, K. H.; Choi, K. Y.; Sung, H. J.; Hwang, M. J.; Kang, D. H.; Lim, S. G.; Jun, S. S.

2015-01-01

Participants of DSP-03 were divided in three groups and each group has focused on the specific subject related to the enhancement of the code analysis. The group A tried to investigate scaling capability of ATLAS test data by comparing to the code analysis for a prototype, and the group C studied to investigate effect of various models in the one-dimensional codes. This paper briefly summarizes the code analysis result from the group B participants in the DSP-03 of the ATLAS test facility. The code analysis by Group B focuses highly on investigating the multi-dimensional thermal hydraulic phenomena in the ATLAS facility during the SLB transient. Even though the one-dimensional system analysis code cannot simulate the whole system of the ATLAS facility with a nodalization of the CFD (Computational Fluid Dynamics) scale, a reactor pressure vessel can be considered with multi-dimensional components to reflect the thermal mixing phenomena inside a downcomer and a core. Also, the CFD could give useful information for understanding complex phenomena in specific components such as the reactor pressure vessel. From the analysis activity of Group B in ATLAS DSP-03, participants adopted a multi-dimensional approach to the code analysis for the SLB transient in the ATLAS test facility. The main purpose of the analysis was to investigate prediction capability of multi-dimensional analysis tools for the SLB experiment result. In particular, the asymmetric cooling and thermal mixing phenomena in the reactor pressure vessel could be significantly focused for modeling the multi-dimensional components

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

International Nuclear Information System (INIS)

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

2013-01-01

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

Analysis of the Phebus FPT0 containment thermal hydraulics with the Jericho and Trio-VF codes

International Nuclear Information System (INIS)

Layly, V.D.; Spitz, P.; Mailliat, A.

1994-01-01

This paper presents the analysis of the thermal hydraulic behavior of the containment, during the Phebus FPT0 test performed on December 2, 1993, with the Jericho code which deals with the thermal hydraulics of containment in the severe accident field. This code is part of Escadre which is the French system of codes in charge of predicting PWR severe accidents. After summarizing the relevant Jericho code characteristics and the preliminary assessment work for the Phebus conditions, we briefly describe the REPF 502 test facility and report the thermal hydraulic FPT0 experimental protocol. Then, the experiment / Jericho calculation comparisons are analysed. Because the Jericho code assumes a well-mixed atmosphere, some additional 3-D calculations have been carried out in order to get further insight on the convection flow patterns and qualify the well-mixed atmosphere assumption in the Phebus containment. (author). 9 refs., 12 figs

Development of RETRAN-03/MOV code for thermal-hydraulic analysis of nuclear reactor under moving conditions

International Nuclear Information System (INIS)

Kim, Hak Jae; Park, Goon Cherl

1996-01-01

Nuclear ship reactors have several; features different from land-based PWR's. Especially, effects of ship motions on reactor thermal-hydraulics and good load following capability for abrupt load changes are essential characteristics of nuclear ship reactors. This study modified the RETRAN-03 to analyze the thermal-hydraulic transients under three-dimensional ship motions, named RETRAN-03/MOV in order to apply to future marine reactors. First Japanese nuclear ship MUTSU reactor have been analyzed under various ship motions to verify this code. Calculations have been performed under rolling,heaving and stationary inclination conditions during normal operation. Also, the natural circulation has been analyzed, which can provide the decay heat removed to ensure the passive safety of marine reactors. As results, typical thermal-hydraulic characteristics of marine reactors such as flow rate oscillations and S/G water level oscillations have been successfully simulated at various conditions. 7 refs., 11 figs. (author)

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 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.

Three-dimensional thermal hydraulic best estimate code BAGIRA: new results of verification

International Nuclear Information System (INIS)

Peter Kohut; Sergey D Kalinichenko; Alexander E Kroshilin; Vladimir E Kroshilin; Alexander V Smirnov

2005-01-01

Full text of publication follows: BAGIRA is a three-dimensional inhomogeneous two-velocity two-temperature thermal hydraulic code of best estimate, elaborated in VNIIAES for modeling two-phase flows in the primary circuit and steam generators of VVER-type nuclear reactors under various accident, transient or normal operation conditions. In this talk we present verification results of the BAGIRA code, obtained on the basis of different experiments performed on special and integral thermohydraulic experimental facilities as well as on real NPPs. Special attention is paid to the verification of three-dimensional flow models. Besides that we expose new results of the code benchmark analysis made on the basis of two recent LOCA-type experiments - 'Leak 2 x 25% from the hot leg double-side rupture' and 'Leak 3% from the cold leg' - performed on the PSB-VVER integral test facility (Electrogorsk Research and Engineering Center, Electrogorsk, Russia) - the most up-to-date Russian large-scale four-loop unit which has been designed for modelling the primary circuit of VVER-1000 type reactors. (authors)

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

International Nuclear Information System (INIS)

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

2017-01-01

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

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

Science.gov (United States)

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

1988-01-01

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

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)

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

Energy Technology Data Exchange (ETDEWEB)

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

2012-01-15

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

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2015-12-31

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

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

International Nuclear Information System (INIS)

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

1983-01-01

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

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.

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

Energy Technology Data Exchange (ETDEWEB)

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

2017-08-15

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

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

International Nuclear Information System (INIS)

Furuya, Masahiro; Inada, Fumio; Yoneda, Kimitoshi

1997-01-01

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

Advances in Integrated Vehicle Thermal Management and Numerical Simulation

Directory of Open Access Journals (Sweden)

Yan Wang

2017-10-01

Full Text Available With the increasing demands for vehicle dynamic performance, economy, safety and comfort, and with ever stricter laws concerning energy conservation and emissions, vehicle power systems are becoming much more complex. To pursue high efficiency and light weight in automobile design, the power system and its vehicle integrated thermal management (VITM system have attracted widespread attention as the major components of modern vehicle technology. Regarding the internal combustion engine vehicle (ICEV, its integrated thermal management (ITM mainly contains internal combustion engine (ICE cooling, turbo-charged cooling, exhaust gas recirculation (EGR cooling, lubrication cooling and air conditioning (AC or heat pump (HP. As for electric vehicles (EVs, the ITM mainly includes battery cooling/preheating, electric machines (EM cooling and AC or HP. With the rational effective and comprehensive control over the mentioned dynamic devices and thermal components, the modern VITM can realize collaborative optimization of multiple thermodynamic processes from the aspect of system integration. Furthermore, the computer-aided calculation and numerical simulation have been the significant design methods, especially for complex VITM. The 1D programming can correlate multi-thermal components and the 3D simulating can develop structuralized and modularized design. Additionally, co-simulations can virtualize simulation of various thermo-hydraulic behaviors under the vehicle transient operational conditions. This article reviews relevant researching work and current advances in the ever broadening field of modern vehicle thermal management (VTM. Based on the systematic summaries of the design methods and applications of ITM, future tasks and proposals are presented. This article aims to promote innovation of ITM, strengthen the precise control and the performance predictable ability, furthermore, to enhance the level of research and development (R&D.

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

International Nuclear Information System (INIS)

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

1978-01-01

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

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

International Nuclear Information System (INIS)

Yamada, K.; Aksan, S. N.

2012-01-01

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

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

International Nuclear Information System (INIS)

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

2013-01-01

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

The analysis of thermal-hydraulic models in MELCOR code

Energy Technology Data Exchange (ETDEWEB)

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

1996-07-15

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

Study of thermal-hydraulic characteristics in an LMFBR intermediate plenum

International Nuclear Information System (INIS)

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

1985-01-01

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

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)

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 ...

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 behavior evaluation of tank A-101

International Nuclear Information System (INIS)

Ogden, D.M.

1996-01-01

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

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)

Development of whole core thermal-hydraulic analysis program ACT. 3. Coupling core module with primary heat transport system module

International Nuclear Information System (INIS)

Ohtaka, Masahiko; Ohshima, Hiroyuki

1998-10-01

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 inter-wrapper flow under various reactor operation conditions. In this work, the core module as a main part of the ACT developed last year, which simulates thermal-hydraulics in the subassemblies and the inter-subassembly gaps, was coupled with an one dimensional plant system thermal-hydraulic analysis code LEDHER to simulate transients in the primary heat transport system and to give appropriate boundary conditions to the core model. The effective algorithm to couple these two calculation modules was developed, which required minimum modification of them. In order to couple these two calculation modules on the computing system, parallel computing technique using PVM (Parallel Virtual Machine) programming environment was applied. The code system was applied to analyze an out-of-pile sodium experiment simulating core with 7 subassemblies under transient condition for code verification. It was confirmed that the analytical results show a similar tendency of experimental results. (author)

Thermal hydraulic and safety analyses for Pakistan Research Reactor-1

International Nuclear Information System (INIS)

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

1999-01-01

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

Single Tube Rupture at Cold Side of Steam Generator Simulation Test Report Using the ATLAS

International Nuclear Information System (INIS)

Kang, Kyoung Ho; Park, Hyun Sik; Cho, Seok

2010-12-01

In this study, a postulated SGTR event of the APR1400 was experimentally investigated with the ATLAS. In order to simulate a double-ended rupture of a single U-tube in the APR1400, the SGTR-CL-01 test was performed with the ATLAS. The main objectives of this test were not only to provide a physical insight into the system response of the APR1400 during the SGTR but also to produce integral effect experimental data to validate the safety analysis code. In the present report, major thermal-hydraulic phenomena such as the system pressures, the collapsed water levels, and the break flow rate were presented and discussed. Following the reactor trip induced by high steam generator level (HSGL) signal, the primary system pressure decreased and the secondary system pressure increased until the MSSVs was opened. The MSSVs repeated on and off status depending on the secondary system pressure during the whole test period. Due to the break flow, the collapsed water level of the affected steam generator showed milder decrease than that of the intact steam generator. This integral effect test data will be used to evaluate the prediction capability of existing safety analysis codes of the MARS and the RELAP5 as well as the SPACE code. Furthermore, this data can be utilized to identify any code deficiency for a SGTR simulation, especially for DVI-adapted plants

Conceptual design of small-sized HTGR system (3). Core thermal and hydraulic design

International Nuclear Information System (INIS)

Inaba, Yoshitomo; Sato, Hiroyuki; Goto, Minoru; Ohashi, Hirofumi; Tachibana, Yukio

2012-06-01

The Japan Atomic Energy Agency has started the conceptual designs of small-sized High Temperature Gas-cooled Reactor (HTGR) systems, aiming for the 2030s deployment into developing countries. The small-sized HTGR systems can provide power generation by steam turbine, high temperature steam for industry process and/or low temperature steam for district heating. As one of the conceptual designs in the first stage, the core thermal and hydraulic design of the power generation and steam supply small-sized HTGR system with a thermal power of 50 MW (HTR50S), which was a reference reactor system positioned as a first commercial or demonstration reactor system, was carried out. HTR50S in the first stage has the same coated particle fuel as HTTR. The purpose of the design is to make sure that the maximum fuel temperature in normal operation doesn't exceed the design target. Following the design, safety analysis assuming a depressurization accident was carried out. The fuel temperature in the normal operation and the fuel and reactor pressure vessel temperatures in the depressurization accident were evaluated. As a result, it was cleared that the thermal integrity of the fuel and the reactor coolant pressure boundary is not damaged. (author)

Assessment of moderator integrity using realisitc model and parametric studies on thermal-hydraulic characteristics in SPEL

International Nuclear Information System (INIS)

Yoo, S. H.; Kim, M. W.; Kang, S. C.; Kim, H. J.; Min, B. J.; Yoon, C.

2002-01-01

Three-dimensional analyses of fluid flow and heat transfer have been performed to assess thermal-hydraulic characteristics for moderator simulation conducted by SPEL(Sheridan Park Experimental Laboratory) experimental facility. The parametric study has also carried out to investigate the effect of major parameters such as flowrate, temperature, and heat load generated from the heaters on the temperature and flow distribution inside the moderator. In this study, three flow patterns have been identified in the moderator with flowrate, heat generation, or both. As the transition of fluid flow is progressed, it is found that the dimensionless numbers (Ar) and the ratio of buoyancy to inertia forces are constant. Moreover, the behavior of temperature distribution inside calandria has also been investigated, when the flowrate of moderator is changed with time

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

International Nuclear Information System (INIS)

Kopecek, Vit; Zacha, Pavel

2014-01-01

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

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

International Nuclear Information System (INIS)

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

2014-01-01

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

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

International Nuclear Information System (INIS)

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

2012-01-01

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

Design Formulae for Hydraulic Stability and Structural Integrity of Dolos Breakwater Round-Heads

DEFF Research Database (Denmark)

Burcharth, H. F.; Jensen, Jacob Birk; Liu, Z.

1995-01-01

A rational design of Dolos armour unit should incorporate both the hydraulic stability and the structural integrity. The previous tests performed by Aalborg University (AU) resulted in design formulae for the trunk of a 1:1.5 slope Dolos breakwater without superstructure including both...... the hydraulic stability and the structural integrity. The objective of the round-head tests is to produce similar design formulae for Dolos armour in around-head. The tests will also include examinations of the hydraulic stability and run-up for a trunk section adjacent to the round-head. A run-up formula...

The thermal-hydraulic for the new technologies: the micro-fluidics

International Nuclear Information System (INIS)

Crecy, F. de; Gruss, A.; Bricard, A.; Excoffon, J.

2000-01-01

The micro-fluidics can be defined as the fluid flow in little canals. This scale offers a great interest for the biotechnology type. In this paper, the authors present this fluids form and detail the researches performed at the Department of Physics and Thermal-hydraulics of the CEA, in the domain of the physical properties characterization and of the numerical two-phase direct simulation. (A.L.B.)

Challenges in coupled thermal-hydraulics and neutronics simulations for LWR safety analysis

International Nuclear Information System (INIS)

Ivanov, Kostadin; Avramova, Maria

2007-01-01

The simulation of nuclear power plant accident conditions requires three-dimensional (3D) modeling of the reactor core to ensure a realistic description of physical phenomena. The operational flexibility of Light Water Reactor (LWR) plants can be improved by utilizing accurate 3D coupled neutronics/thermal-hydraulics calculations for safety margins evaluations. There are certain requirements to the coupling of thermal-hydraulic system codes and neutron-kinetics codes that ought to be considered. The objective of these requirements is to provide accurate solutions in a reasonable amount of CPU time in coupled simulations of detailed operational transient and accident scenarios. These requirements are met by the development and implementation of six basic components of the coupling methodologies: ways of coupling (internal or external coupling); coupling approach (integration algorithm or parallel processing); spatial mesh overlays; coupled time-step algorithms; coupling numerics (explicit, semi-implicit and implicit schemes); and coupled convergence schemes. These principles of the coupled simulations are discussed in details along with the scientific issues associated with the development of appropriate neutron cross-section libraries for coupled code transient modeling. The current trends in LWR nuclear power generation and regulation as well as the design of next generation LWR reactor concepts along with the continuing computer technology progress stimulate further development of these coupled code systems. These efforts have been focused towards extending the analysis capabilities as well as refining the scale and level of detail of the coupling. This article analyses the coupled phenomena and modeling challenges on both global (assembly-wise) and local (pin-wise) levels. The issues related to the consistent qualification of coupled code systems as well as their application to different types of LWR transients are presented. Finally, the advances in numerical

Development of heat transfer package for core thermal-hydraulic design and analysis of upgraded JRR-3

International Nuclear Information System (INIS)

Sudo, Yukio; Ikawa, Hiromasa; Kaminaga, Masanori

1985-01-01

A heat transfer package was developed for the core thermal-hydraulic design and analysis of the Japan Research Reactor-3 (JRR-3) which is to be remodeled to a 20 MWt pool-type, light water-cooled reactor with 20 % low enriched uranium (LEU) plate-type fuel. This paper presents the constitution of the developed heat transfer package and the applicability of the heat transfer correlations adopted in it, based on the heat transfer experiments in which thermal-hydraulic features of the new JRR-3 core were properly reflected. (author)

Investigation of a Station Blackout Scenario with the ATLAS Test

International Nuclear Information System (INIS)

Kim, Yeon Sik; Yu, Xin Guo; Kang, Kyoung Ho; Park, Hyun Sik; Cho, Seok; Min, Kyeong Ho; Choi, Nam Hyeon; Kim, Bok Deuk; Park, Jong Gook; Choi, Ki Yong

2012-01-01

KAERI (Korea Atomic Energy Research Institute) has been operating an integral effect test facility, ATLAS (Advanced Thermal-Hydraulic Test Loop for Accident Simulation), for accident simulations pertaining to the OPR1000 (Optimized Power Reactor, 1000MWe) and the APR1400 (Advanced Power Reactor, 1400MWe) which are in operation and under construction in Korea, respectively. After the Fukushima accidents due to the combination of an earthquake followed by a tsunami in east Japan on March 11, 2011, the concept of boundary between the design basis and beyond-design basis accidents became obscure. One scenario is the station blackout (SBO), which is defined as 'the loss of all alternating current (AC) power in a nuclear power plant' by the USNRC 10CFR50 Section 50.63, which has adopted a new safety regulation for the SBO in June of 1988. In any case the SBO that occurred in Fukushima seemed to go beyond the definition of the current SBO scenario. In the mean time, numerous researches has been conducted on the safety concern of the SBO for existing and advanced nuclear power plants worldwide. From the internal review of an SBO scenario, it was concluded that the understanding of the thermo-hydraulic phenomena occurred within the reactor coolant system is a prerequisite although seemed to be quite a simple sequence of events. This was the motivation of an SBO test using the ATLAS facility. For the understanding of the physical phenomena within the primary system, an SBO was assumed with simple intial and boundary conditions, e.g. start of an SBO at time zero, no diesel and AC powers, no auxiliary feedwater pumps (motor-driven and turbine driven) etc. In this paper, overview of the SBO test results was described including a result of analytical calculations simulating the SBO test using the MARS code

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

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 analysis on the whole module of water cooled ceramic breeder blanket for CFETR

Energy Technology Data Exchange (ETDEWEB)

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

2016-11-15

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

Thermal-Hydraulic Aspects of Changing the Nuclear Fuel-Cladding Materials from Zircaloy to Silicon Carbides

International Nuclear Information System (INIS)

Niceno, Bojan; Pouchon, Manuel

2014-01-01

The accident in Fukushima has drastically shown the drawbacks of Zircaloy claddings despite their beneficial properties in normal use. The effect of the lack of cooling and the production of hydrogen would not have been so strong if the fuel cladding had not consisted of a zirconium (or metal) alloy. International activities have been started to search for an alternative to Zircaloy, however, still on a limited basis. A project sponsored by Swissnuclear has been conducted at Paul Scherrer Institute (PSI) with the aim to close the gap in knowledge on application of silicon carbides (SiC) as potential replacement for Zircaloys as material for nuclear fuel cladding. The work was interdisciplinary, result of collaboration between different laboratories at PSI, and has focused on SiC cladding material properties, implication of its usage on neutronics and on thermal-hydraulics. This paper summarizes thermal-hydraulic aspects of changing Zircaloy for SiC as the cladding material. The change of cladding material inevitably changes the surface properties thus making a significant impact on boiling curve, and critical heat flux (CHF). Low chemical reactivity of SiC means fewer particles in the flow (less crud), which leads to fewer failures, but also decreases the CHF. Due to differences in physical properties between SiC and Zircaloys, higher brittleness of SiC in particular, might have impact on fuel-rod assembly design, which has direct influence on flow patterns and heat transfer in the fuel assembly. Higher melting (i.e. decomposition) point for SiC means that severe accident management guidelines (SAMG) should have to be re-assessed. Not only would the core degrade later than in the case of conventional fuels, but the production of hydrogen would be quite different as well. All these issues are explored in this work in two steps; first the SiC properties which may have influence on thermal-hydraulics are outlined, then each thermal-hydraulic issues is explained from

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)

The NEPTUN experiments on LOCA thermal-hydraulics for tight-lattice PWRs

International Nuclear Information System (INIS)

Dreier, J.; Chawla, R.; Rouge, N.; Yanar, S.

1990-01-01

The NEPTUN test facility at the Paul Scherrer Institute is currently being used to provide a broad data base for the validation of thermal-hydraulics codes used in predicting the reflooding behaviour of a tight-lattice PWR (light water highb conversion reactor, LWHCR). The present paper gives a description of the facility and the matrix to be covered in the experimental program. Results are presented from a number of forced-feed, bottom-reflooding experiments, comparisons being made with (a) measurements carried out earlier for standard-PWR geometry and (b) the results of a calculational benchmark exercise conducted in the framework of a Swiss/German LWHCR-development agreement. Rewetting for the tight, hexagonal-geometry (p/d = 1.13) NEPTUN-III test bundle has been found to occur in all tests carried out to date, in which reasonably LWHCR-representative values for the various thermal-hydraulics parameters are used. Results of the calculational benchmark exercise have confirmed the need for further code development efforts for achieving reliable predictions of LWHCR reflooding behaviour. (author) 11 figs., 3 tabs., 3 refs

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.)

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

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)