LANSCE steady state unperturbed thermal neutron fluxes at 100 μA

International Nuclear Information System (INIS)

Russell, G.J.

1989-01-01

The ''maximum'' unperturbed, steady state thermal neutron flux for LANSCE is calculated to be 2 /times/ 10 13 n/cm 2 -s for 100 μA of 800-MeV protons. This LANSCE neutron flux is a comparable entity to a steady state reactor thermal neutron flux. LANSCE perturbed steady state thermal neutron fluxes have also been calculated. Because LANSCE is a pulsed neutron source, much higher ''peak'' (in time) neutron fluxes can be generated than at a steady state reactor source. 5 refs., 5 figs

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)

Steady-state thermal-hydraulic design analysis of the Advanced Neutron Source reactor

International Nuclear Information System (INIS)

Yoder, G.L. Jr.; Dixon, J.R.; Elkassabgi, Y.; Felde, D.K.; Giles, G.E.; Harrington, R.M.; Morris, D.G.; Nelson, W.R.; Ruggles, A.E.; Siman-Tov, M.; Stovall, T.K.

1994-05-01

The Advanced Neutron Source (ANS) is a research reactor that is planned for construction at Oak Ridge National Laboratory. This reactor will be a user facility with the major objective of providing the highest continuous neutron beam intensities of any reactor in the world. Additional objectives for the facility include providing materials irradiation facilities and isotope production facilities as good as, or better than, those in the High Flux Isotope Reactor. To achieve these objectives, the reactor design uses highly subcooled heavy water as both coolant and moderator. Two separate core halves of 67.6-L total volume operate at an average power density of 4.5 MW(t)/L, and the coolant flows upward through the core at 25 m/s. Operating pressure is 3.1 MPa at the core inlet with a 1.4-MPa pressure drop through the core region. Finally, in order to make the resources available for experimentation, the fuel is designed to provide a 17-d fuel cycle with an additional 4 d planned in each cycle for the refueling process. This report examines the codes and models used to develop the thermal-hydraulic design for ANS, as well as the correlations and physical data; evaluates thermal-hydraulic uncertainties; reports on thermal-hydraulic design and safety analysis; describes experimentation in support of the ANS reactor design and safety analysis; and provides an overview of the experimental plan

Thermal Hydraulic Fortran Program for Steady State Calculations of Plate Type Fuel Research Reactors

International Nuclear Information System (INIS)

Khedr, H.

2008-01-01

The safety assessment of Research and Power Reactors is a continuous process over their life and that requires verified and validated codes. Power Reactor codes all over the world are well established and qualified against a real measuring data and qualified experimental facilities. These codes are usually sophisticated, require special skills and consume much more running time. On the other hand, most of the Research Reactor codes still requiring more data for validation and qualification. Therefore it is benefit for a regulatory body and the companies working in the area of Research Reactor assessment and design to have their own program that give them a quick judgment. The present paper introduces a simple one dimensional Fortran program called THDSN for steady state best estimate Thermal Hydraulic (TH) calculations of plate type fuel RRs. Beside calculating the fuel and coolant temperature distribution and pressure gradient in an average and hot channel the program calculates the safety limits and margins against the critical phenomena encountered in RR such as the burnout heat flux and the onset of flow instability. Well known TH correlations for calculating the safety parameters are used. THDSN program is verified by comparing its results for 2 and 10 MW benchmark reactors with that published in IAEA publications and good agreement is found. Also the program results are compared with those published for other programs such as PARET and TERMIC. An extension for this program is underway to cover the transient TH calculations

Implications of steady-state operation on divertor design

International Nuclear Information System (INIS)

Sevier, D.L.; Reis, E.E.; Baxi, C.B.; Silke, G.W.; Wong, C.P.C.; Hill, D.N.

1996-01-01

As fusion experiments progress towards long pulse or steady state operation, plasma facing components are undergoing a significant change in their design. This change represents the transition from inertially cooled pulsed systems to steady state designs of significant power handling capacity. A limited number of Plasma Facing Component (PFC) systems are in operation or planning to address this steady state challenge at low heat flux. However in most divertor designs components are required to operate at heat fluxes at 5 MW/m 2 or above. The need for data in this area has resulted in a significant amount of thermal/hydraulic and thermal fatigue testing being done on prototypical elements. Short pulse design solutions are not adequate for longer pulse experiments and the areas of thermal design, structural design, material selection, maintainability, and lifetime prediction are undergoing significant changes. A prudent engineering approach will guide us through the transitional phase of divertor design to steady-state power plant components. This paper reviews the design implications in this transition to steady state machines and the status of the community efforts to meet evolving design requirements. 54 refs., 5 figs., 2 tabs

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

Institute of Scientific and Technical Information of China (English)

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

2007-01-01

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

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)

Description of a stable scheme for steady-state coupled Monte Carlo–thermal–hydraulic calculations

International Nuclear Information System (INIS)

Dufek, Jan; Eduard Hoogenboom, J.

2014-01-01

Highlights: • A stable coupling scheme for steady-state MC–TH calculations is described. • The coupling scheme is based on the stochastic approximation method. • The neutron flux (or power) distribution is relaxed using a variable step-size. - Abstract: We provide a detailed description of a numerically stable and efficient coupling scheme for steady-state Monte Carlo neutronic calculations with thermal–hydraulic feedback. While we have previously derived and published the stochastic approximation based method for coupling the Monte Carlo criticality and thermal–hydraulic calculations, its possible implementation has not been described in a step-by-step manner. As the simple description of the coupling scheme was repeatedly requested from us, we have decided to make it available via this note

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

TITAN: an advanced three-dimensional neutronics/thermal-hydraulics code for light water reactor safety analysis

International Nuclear Information System (INIS)

Griggs, D.P.; Kazimi, M.S.; Henry, A.F.

1982-01-01

The initial development of TITAN, a three-dimensional coupled neutronics/thermal-hydraulics code for LWR safety analysis, has been completed. The transient neutronics code QUANDRY has been joined to the two-fluid thermal-hydraulics code THERMIT with the appropriate feedback mechanisms modeled. A detailed steady-state and transient coupling scheme based on the tandem technique was implemented in accordance with the important structural and operational characteristics of QUANDRY and THERMIT. A two channel sample problem formed the basis for steady-state and transient analyses performed with TITAN. TITAN steady-state results were compared with those obtained with MEKIN and showed good agreement. Null transients, simulated turbine trip transients, and a rod withdrawal transient were analyzed with TITAN and reasonable results were obtained

RAP-3A Computer code for thermal and hydraulic calculations in steady state conditions for fuel element clusters

International Nuclear Information System (INIS)

Popescu, C.; Biro, L.; Iftode, I.; Turcu, I.

1975-10-01

The RAP-3A computer code is designed for calculating the main steady state thermo-hydraulic parameters of multirod fuel clusters with liquid metal cooling. The programme provides a double accuracy computation of temperatures and axial enthalpy distributions of pressure losses and axial heat flux distributions in fuel clusters before boiling conditions occur. Physical and mathematical models as well as a sample problem are presented. The code is written in FORTRAN-4 language and is running on a IBM-370/135 computer

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

Real-time dynamic hydraulic model for water distribution networks: steady state modelling

CSIR Research Space (South Africa)

Osman, Mohammad S

2016-09-01

Full Text Available equipment (pipes, reservoirs, pumps, valves, etc.) was used as a pilot WDN. Further information of the various other DHM components has been published [1]. The steady-state hydraulic model calculates the network hydraulic variables at a particular... from the abstraction point to the two low-level concrete reservoirs. On this pipeline there is a 2” tie-off to an alternate consumer as well as another 2” tie-off (5 m length) to the pump station sump. Water from the pump station is pumped to two...

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)

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.

COOLOD-N2: a computer code, for the analyses of steady-state thermal-hydraulics in research reactors

International Nuclear Information System (INIS)

Kaminaga, Masanori

1994-03-01

The COOLOD-N2 code provides a capability for the analyses of the steady-state thermal-hydraulics of research reactors. This code is revised version of the COOLOD-N code, and is applicable not only for research reactors in which plate-type fuel is adopted, but also for research reactors in which rod-type fuel is adopted. In the code, subroutines to calculate temperature distribution in rod-type fuel have been newly added to the COOLOD-N code. The COOLOD-N2 code can calculate fuel temperatures under both forced convection cooling mode and natural convection cooling mode as well as COOLOD-N code. In the COOLOD-N2 code, a 'Heat Transfer package' is used for calculating heat transfer coefficient, DNB heat flux etc. The 'Heat Transfer package' is subroutine program and is especially developed for research reactors in which plate-type fuel is adopted. In case of rod-type fuel, DNB heat flux is calculated by both the 'Heat Transfer package' and Lund DNB heat flux correlation which is popular for TRIGA reactor. The COOLOD-N2 code also has a capability of calculating ONB temperature, the heat flux at onset of flow instability as well as DNB heat flux. (author)

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

On the theoretical–numerical study of the ITER Upper Port Plug structure hydraulic behaviour under steady state and draining and drying transient conditions

International Nuclear Information System (INIS)

Di Maio, P.A.; Paradiso, D.; Dell’Orco, G.; Pitcher, C.S.; Kalish, M.

2011-01-01

Highlights: ► UPP TS hydraulic behaviour has been investigated under steady state and D and D transient conditions. ► A thermal–hydraulic system code has been adopted and a UPP TS model has been set-up and validated against results of steady state CFD analyses. ► The TS steady state hydraulic characteristic functions have been derived for two coolant flow paths showing that right plate inlet one is the most promising. ► Draining simulations indicate that the 4 MPa injection pressure is high enough to drain almost completely the circuit in a reasonable time (∼6 s). ► Results indicate that right plate inlet flow path allows the TS complete draining, eliminating the need for the drying procedure. - Abstract: The ITER diagnostic Upper Port Plug (UPP) is a water-cooled stainless steel structure aimed to integrate within vacuum vessel the plasma diagnostic systems, shielding them from neutron and photon irradiation. Due to the very intense heat loads expected, a proper cooling circuit has been designed to ensure an adequate UPP cooling with an acceptable thermal rise and an unduly high pumping power and to perform its draining and drying procedure by injection of pressurized nitrogen. A theoretical research activity has been launched at the Department of Nuclear Engineering of the University of Palermo aiming to investigate the hydraulic behaviour of the UPP Trapezoid Section cooling circuit under steady state conditions and during its draining and drying transient procedure. The research activity has been performed following a theoretical–computational approach and adopting the RELAP5 thermal–hydraulic system code. The Trapezoid Section cooling circuit characteristic functions have been derived under steady state conditions at various coolant temperatures for both the coolant flow paths at the present under consideration for this circuit. The distributions of coolant mass flow rates along the channels of the cooling circuit have been calculated too

STEADY-SHIP: a computer code for three-dimensional nuclear and thermal-hydraulic analyses of marine reactors

International Nuclear Information System (INIS)

Itagaki, Masafumi; Naito, Yoshitaka; Tokuno, Yukio; Matsui, Yasushi.

1988-01-01

A code STEADY-SHIP has been developed to calculate three-dimensional distributions of neutron flux, power and coolant temperature in the reactor core of the nuclear ship MUTSU. The code consists of two parts, that is, a few-group three-dimensional neutron diffusion module DIFFUSION-SHIP and a thermal-hydraulic module HYDRO-SHIP: In the DIFFUSION-SHIP the leakage iteration method is used for solving the three-dimensional neutron diffusion equation with small computer core memory and short computing time; The HYDRO-SHIP performs the general thermal-hydraulic calculation for evaluating feedbacks required in the neutronic calculation by the DIFFUSION-SHIP. The macroscopic nuclear constants are generated by a module CROSS-SHIP as functions of xenon poison, fuel temperature, moderator temperature and moderator density. A module LOCAL-FINE has the capability of computing a detailed rod power distribution for each local node in the core, using the boundary conditions on the surface of the node which were supplied by the STEADY-SHIP whole-core calculation. The applicability of this code to marine reactors has been demonstrated by comparing the computed results with the data measured during the MUTSU land-loaded core critical experiments and with the data obtained during the hot-zero-power tests performed for the actual MUTSU plant. (author)

Understanding void fraction in steady state and dynamic environments

International Nuclear Information System (INIS)

Chexal, B.; Maulbetsch, J.; Harrison, J.; Petersen, C.; Jensen, P.; Horowitz, J.

1997-01-01

Understanding void fraction behavior in steady-state and dynamic environments is important to accurately predict the thermal-hydraulic behavior of two-phase or two-component systems. The Chexal-Lellouche (C-L) void fraction mode described herein covers the full range of pressures, flows, void fractions, and fluid types (steam-water, air-water, and refrigerants). A drift flux model formulation is used which covers the complete range of concurrent and countercurrent flows. The (1996) model revises the earlier C-L void fraction correlation, improves the capability of the model in countercurrent flow based on the incorporation of additional data, and improves the characteristics of the correlation that are important in transient programs. The model has been qualified with data from a number of steady state two-phase and two-component tests, and has been incorporated into the transient analysis code RELAP5 and RETRAN-3D and evaluated with a variety of transient and steady state tests. A 'plug-in' module for the void fraction correlation has been developed and implemented in RELAP5 and RETRAN-3D. The module is available as source code for inclusion into other thermal-hydraulic programs and can be used in any program that utilizes the same interface variables

Characterization of the TRIGA Mark II reactor full-power steady state

Energy Technology Data Exchange (ETDEWEB)

Cammi, Antonio, E-mail: antonio.cammi@polimi.it [Politecnico di Milano – Department of Energy, CeSNEF (Enrico Fermi Center for Nuclear Studies), via La Masa 34, 20156 Milano (Italy); Zanetti, Matteo [Politecnico di Milano – Department of Energy, CeSNEF (Enrico Fermi Center for Nuclear Studies), via La Masa 34, 20156 Milano (Italy); Chiesa, Davide; Clemenza, Massimiliano; Pozzi, Stefano; Previtali, Ezio; Sisti, Monica [University of Milano-Bicocca, Physics Department “G. Occhialini” and INFN Section, Piazza dell’Ateneo Nuovo, 20126 Milan (Italy); Magrotti, Giovanni; Prata, Michele; Salvini, Andrea [University of Pavia, Applied Nuclear Energy Laboratory (L.E.N.A.), Via Gaspare Aselli 41, 27100 Pavia (Italy)

2016-04-15

Highlights: • Full-power steady state characterization of the TRIGA Mark II reactor. • Monte Carlo and Multiphysics simulation of the TRIGA Mark II reactor. • Sub-cooled boiling effects in the TRIGA Mark II reactor. • Thermal feedback effects in the TRIGA Mark II reactor. • Experimental data based validation. - Abstract: In this paper, the characterization of the full-power steady state of the TRIGA Mark II nuclear reactor at the University of Pavia is achieved by coupling the Monte Carlo (MC) simulation for neutronics with the “Multiphysics” model for thermal-hydraulics. Neutronic analyses have been carried out with a MCNP5 based MC model of the entire reactor system, already validated in fresh fuel and zero-power configurations (in which thermal effects are negligible) and using all available experimental data as a benchmark. In order to describe the full-power reactor configuration, the temperature distribution in the core must be established. To evaluate this, a thermal-hydraulic model has been developed, using the power distribution results from the MC simulation as input. The thermal-hydraulic model is focused on the core active region and takes into account sub-cooled boiling effects present at full reactor power. The obtained temperature distribution is then entered into the MC model and a benchmark analysis is carried out to validate the model in fresh fuel and full-power configurations. An acceptable correspondence between experimental data and simulation results concerning full-power reactor criticality proves the reliability of the adopted methodology of analysis, both from the perspective of neutronics and thermal-hydraulics.

SYNTH-C, Steady-State and Time-Dependent 3-D Neutron Diffusion with Thermohydraulic Feedback

Energy Technology Data Exchange (ETDEWEB)

Brega, E [ENEL-CRTN, Bastioni di Porta Volta 10, Milan (Italy); Salina, E [A.R.S. Spa, Viale Maino 35, Milan (Italy)

1980-04-01

1 - Description of problem or function: SYNTH-C-STEADY and SYNTH-C- TRANS solve respectively the steady-state and time-dependent few- group neutron diffusion equations in three dimensions x,y,z in the presence of fuel temperature and thermal-hydraulic feedback. The neutron diffusion and delayed precursor equations are approximated by a space-time (z,t) synthesis method with axially discontinuous trial functions. Three thermal-hydraulic and fuel heat transfer models are available viz. COBRA-3C/MIT model, lumped parameter (WIGL) model and adiabatic fuel heat-up model. 2 - Method of solution: The steady-state and time-dependent synthesis equations are solved respectively by the Wielandt's power method and by the theta-difference method (in time), both coupled with a block factorization technique and double precision arithmetic. The thermal-hydraulic model equations are solved by fully implicit finite differences (WIGL) or explicit-implicit difference techniques with iterations (COBRA-EC/MIT). 3 - Restrictions on the complexity of the problem: Except for the few- group limitation, the programs have no other fixed limitation so the ability to run a problem depends only on the available computer storage.

Design and thermal-hydraulic analysis of PFC baking for SST-1 Tokamak

International Nuclear Information System (INIS)

Chaudhuri, Paritosh; Reddy, D. Chenna; Khirwadkar, S.; Prakash, N. Ravi; Santra, P.; Saxena, Y.C.

2001-01-01

The Steady-State Superconducting Tokamak (SST-1) is a medium-size tokamak with super-conducting magnetic field coils. Plasma facing components (PFC) of the SST-1, consisting of divertors, passive stabilisers, baffles, and poloidal limiters, are designed to be compatible for steady-state operation. Except for the poloidal limiters, all other PFC are structurally continuous in the toroidal direction. As SST-1 is designed to run double-null divertor plasmas, these components also have up-down symmetry. A closed divertor configuration is chosen to produce high recycling and high pumping speed in the divertor region. The passive stabilisers are located close to the plasma to provide stability against the vertical instability of the elongated plasma. The main consideration in the design of the PFC is the steady-state heat removal of up to 1 MW/m 2 . In addition to removing high heat fluxes, the PFC are also designed to be compatible for baking at 350 deg. C. Different flow parameters and various tube layouts have been examined to select the optimum thermal-hydraulic parameters and tube layout for different PFC of SST-1. Thermal response of the PFC during baking has been performed analytically (using a Fortran code) and two-dimensional finite element analysis using ANSYS. The detailed thermal hydraulics and thermal responses of PFC baking is presented in this paper

Design and thermal-hydraulic analysis of PFC baking for SST-1 Tokamak

Energy Technology Data Exchange (ETDEWEB)

Chaudhuri, Paritosh E-mail: paritosh@ipr.res.in; Reddy, D. Chenna; Khirwadkar, S.; Prakash, N. Ravi; Santra, P.; Saxena, Y.C

2001-09-01

The Steady-State Superconducting Tokamak (SST-1) is a medium-size tokamak with super-conducting magnetic field coils. Plasma facing components (PFC) of the SST-1, consisting of divertors, passive stabilisers, baffles, and poloidal limiters, are designed to be compatible for steady-state operation. Except for the poloidal limiters, all other PFC are structurally continuous in the toroidal direction. As SST-1 is designed to run double-null divertor plasmas, these components also have up-down symmetry. A closed divertor configuration is chosen to produce high recycling and high pumping speed in the divertor region. The passive stabilisers are located close to the plasma to provide stability against the vertical instability of the elongated plasma. The main consideration in the design of the PFC is the steady-state heat removal of up to 1 MW/m{sup 2}. In addition to removing high heat fluxes, the PFC are also designed to be compatible for baking at 350 deg. C. Different flow parameters and various tube layouts have been examined to select the optimum thermal-hydraulic parameters and tube layout for different PFC of SST-1. Thermal response of the PFC during baking has been performed analytically (using a Fortran code) and two-dimensional finite element analysis using ANSYS. The detailed thermal hydraulics and thermal responses of PFC baking is presented in this paper.

Coupled neutronic/thermal-hydraulic analysis of the HPLWR three pass core

International Nuclear Information System (INIS)

Monti, Lanfranco; Starflinger, Joerg; Schulenberg, Thomas

2008-01-01

The High Performance Light Water Reactor is an innovative Gen-IV reactor cooled and moderated with water at supercritical pressure. The three pass core concept has been proposed to reduce peaking factors, i.e. hot-channel effects, and it further increases the core heterogeneity, which is mainly due to pronounced water density reduction. For this kind of nuclear reactor, the significant feedbacks - which exist between the properties of the components and the power generation rate - can not be neglected and require a coupled Neutronic/Thermal-Hydraulic analysis even for steady state conditions. The main goal of this paper is to present the developed tool for coupled analyses of the HPLWR. Two state-of-the-art codes have been chosen for Thermal-Hydraulic and Neutronic core analyses, namely TRACE and ERANOS, and they have been coupled with in an iterative procedure in which they are run in series until a steady state condition has been reached. In the simplifying assumptions of uniform enrichment distribution, zero burn-up and ignoring the effect of the control rods, the obtained steady state condition will be discussed and a core power map, flow rate redistribution as well as water and fuel temperature variations will be presented. (author)

Numerical method for three dimensional steady-state two-phase flow calculations

International Nuclear Information System (INIS)

Raymond, P.; Toumi, I.

1992-01-01

This paper presents the numerical scheme which was developed for the FLICA-4 computer code to calculate three dimensional steady state two phase flows. This computer code is devoted to steady state and transient thermal hydraulics analysis of nuclear reactor cores 1,3 . The first section briefly describes the FLICA-4 flow modelling. Then in order to introduce the numerical method for steady state computations, some details are given about the implicit numerical scheme based upon an approximate Riemann solver which was developed for calculation of flow transients. The third section deals with the numerical method for steady state computations, which is derived from this previous general scheme and its optimization. We give some numerical results for steady state calculations and comparisons on required CPU time and memory for various meshing and linear system solvers

MARS input data for steady-state calculation of ATLAS

International Nuclear Information System (INIS)

Park, Hyun Sik; Euh, D. J.; Choi, K. Y.; Kwon, T. S.; Jeong, J. J.; Baek, W. P.

2004-12-01

An integral effect test loop for Pressurized Water Reactors (PWRs), the ATLAS (Advanced Thermal-hydraulic Test Loop for Accident Simulation), is under construction by Thermal-Hydraulics Safety Research Division in Korea Atomic Energy Research Institute (KAERI). This report includes calculation sheets of the input for the best-estimate system analysis code, the MARS code, based on the ongoing design features of ATLAS. The ATLAS facility has been designed to have the length scale of 1/2 and area scale of 1/144 compared with the reference plant, APR1400. The contents of this report are divided into three parts: (1) core and reactor vessel, (2) steam generator and steam line, and (3) primary piping, pressurizer and reactor coolant pump. The steady-state analysis for the ATLAS facility will be performed based on these calculation sheets, and its results will be applied to the detailed design of ATLAS. Additionally, the calculation results will contribute to getting optimum test conditions and preliminary operational test conditions for the steady-state and transient experiments

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.

Thermal-hydraulic and thermo-mechanical design of plasma facing components for SST-1 tokamak

International Nuclear Information System (INIS)

Chaudhuri, Paritosh; Santra, P.; Chenna Reddy, D.; Parashar, S.K.S.

2014-01-01

The Plasma Facing Components (PFCs) are one of the major sub-systems of ssT-1 tokamak. PFC of ssT-1 consisting of divertors, passive stabilizers, baffles and limiters are designed to be compatible for steady state operation. The main consideration in the design of the PFC cooling is the steady state heat removal of up to 1 MW/m 2 . The PFC has been designed to withstand the peak heat fluxes and also without significant erosion such that frequent replacement of the armor is not necessary. Design considerations included 2-D steady state and transient tile temperature distribution and resulting thermal loads in PFC during baking, and cooling, coolant parameters necessary to maintain optimum thermal-hydraulic design, and tile fitting mechanism. Finite Element (FE) models using ANSYS have been developed to carry out the heat transfer and stress analyses of the PFC to understand its thermal and mechanical behaviors. The results of the calculation led to a good understanding of the coolant flow behavior and the temperature distribution in the tube wall and the different parts of the PFC. Thermal analysis of the PFC is carried out with the purpose of evaluating the thermal mechanical behavior of PFCs. The detailed thermal-hydraulic and thermo-mechanical designs of PFCs of ssT-1 are discussed in this paper. (authors)

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)

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

International Nuclear Information System (INIS)

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

1995-12-01

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

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.

Thermal-hydraulic analysis of PWR cores in transient condition

International Nuclear Information System (INIS)

Silva Galetti, M.R. da.

1984-01-01

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

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)

The STAT7 Code for Statistical Propagation of Uncertainties In Steady-State Thermal Hydraulics Analysis of Plate-Fueled Reactors

Energy Technology Data Exchange (ETDEWEB)

Dunn, Floyd E. [Argonne National Lab. (ANL), Argonne, IL (United States); Hu, Lin-wen [Massachusetts Inst. of Technology (MIT), Cambridge, MA (United States). Nuclear Reactor Lab.; Wilson, Erik [Argonne National Lab. (ANL), Argonne, IL (United States)

2016-12-01

The STAT code was written to automate many of the steady-state thermal hydraulic safety calculations for the MIT research reactor, both for conversion of the reactor from high enrichment uranium fuel to low enrichment uranium fuel and for future fuel re-loads after the conversion. A Monte-Carlo statistical propagation approach is used to treat uncertainties in important parameters in the analysis. These safety calculations are ultimately intended to protect against high fuel plate temperatures due to critical heat flux or departure from nucleate boiling or onset of flow instability; but additional margin is obtained by basing the limiting safety settings on avoiding onset of nucleate boiling. STAT7 can simultaneously analyze all of the axial nodes of all of the fuel plates and all of the coolant channels for one stripe of a fuel element. The stripes run the length of the fuel, from the bottom to the top. Power splits are calculated for each axial node of each plate to determine how much of the power goes out each face of the plate. By running STAT7 multiple times, full core analysis has been performed by analyzing the margin to ONB for each axial node of each stripe of each plate of each element in the core.

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

Energy Technology Data Exchange (ETDEWEB)

Reis, Patricia A.L.; Costa, Antonella L.; Pereira, Claubia; Silva, Clarysson A.M. da; Veloso, Maria Auxiliadora F.; Soares, Humbero V., E-mail: patricialire@yahoo.com.br, E-mail: antonella@nuclear.ufmg.br, E-mail: claubia@nuclear.ufmg.br, E-mail: clarysson@nuclear.ufmg.br, E-mail: dora@nuclear.ufmg.br, E-mail: betovitor@ig.com.br [Universidade Federal de Minas Gerais (UFMG), Belo Horizonte, MG (Brazil). Departamento de Engenharia Nuclear; Instituto Nacional de Ciencias e Tecnologia de Reatores Nucleares Inovadores (INCT/CNPq Rede), Rio de Janeiro, RJ (Brazil)

2013-07-01

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

First wall thermal hydraulic models for fusion blankets

International Nuclear Information System (INIS)

Fillo, J.A.

1980-01-01

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

Thermal-hydraulic 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

Thermal, thermo-hydraulic and thermo-mechanic analysis for fuel elements of IEA-R1 reactor at 5MW

International Nuclear Information System (INIS)

Teixeira e Silva, A.; Silva Macedo, L.V. da

1989-01-01

In connection with the on going conversion of IEA-R1 Research Reactor, operated by IPEN-CNEN/SP, from the use of highly enriched uranium (HEU) fuel to the use of low enriched uranium (LEU) fuel, steady-state thermal and thermo-hydraulic analysis of both existing HEU and proposed LEU cores under 2 MW operating conditions have been carried out. Keeping in mind the possibility of power upgrading, steady-state thermal, thermo-hydraulic and thermomechanical analysis of proposed LEU core under 5 MW operating conditions have also been carried out. The thermal and thermo-hydraulic analysis at 2 MW show that the conversion of the existing HEU core to be proposed LEU core will not change the reactor safety margins. Although the upgrading of the reactor power to 5 MW will result in safety margins lower than in case of 2MW, these will be still sufficient for optimum operation and safe behaviour. The thermomechanical analysis at 5 MW show that the thermal stresses induced in the fuel element will satisfy the design limits for mechanical strenght and elastic stability. (author) [pt

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

ORNL rod-bundle heat-transfer test data. Volume 7. Thermal-Hydraulic Test Facility experimental data report for test series 3.07.9 - steady-state film boiling in upflow

International Nuclear Information System (INIS)

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

1982-05-01

Thermal-Hydraulic Test Facility (THTF) test series 3.07.9 was conducted by members of the Oak Ridge National Laboratory Pressurized-Water Reactor (ORNL-PWR) Blowdown Heat Transfer (BDHT) Separate-Effects Program on September 11, September 18, and October 1, 1980. The objective of the program is to investigate heat transfer phenomena believed to occur in PWRs during accidents, including small- and large-break loss-of-coolant accidents. Test series 3.07.9 was designed to provide steady-state film boiling data in rod bundle geometry under reactor accident-type conditions. This report presents the reduced instrument responses for THTF test series 3.07.9. Also included are uncertainties in the instrument responses, calculated mass flows, and calculated rod powers

Thermal hydraulic simulation of the CANDU nuclear reactor

Energy Technology Data Exchange (ETDEWEB)

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

2017-07-01

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

Thermal 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.

BR2 reactor core steady state transient modeling

International Nuclear Information System (INIS)

Makarenko, A.; Petrova, T.

2000-01-01

A coupled neutronics/hydraulics/heat-conduction model of the BR2 reactor core is under development at SCK-CEN. The neutron transport phenomenon has been implemented as steady state and time dependent nodal diffusion. The non-linear heat conduction equation in-side fuel elements is solved with a time dependent finite element method. To allow coupling between functional modules and to simulate subcooled regimes, a simple single-phase hydraulics has been introduced, while the two-phase hydraulics is under development. Multiple tests, general benchmark cases as well as calculation/experiment comparisons demonstrated a good accuracy of both neutronic and thermal hydraulic models, numerical reliability and full code portability. A refinement methodology has been developed and tested for better neutronic representation in hexagonal geometry. Much effort is still needed to complete the development of an extended cross section library with kinetic data and two-phase flow representation. (author)

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 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.

Thermal-hydraulic analysis of the improved TOPAZ-II power system using a heat pipe radiator

Energy Technology Data Exchange (ETDEWEB)

Zhang, Wenwen; Zhang, Dalin, E-mail: dlzhang@mail.xjtu.edu.cn; Tian, Wenxi; Qiu, Suizheng; Su, G.H.

2016-10-15

Highlights: • The system thermal-hydraulic model of the improved space thermionic reactor is developed. • The temperature reactivity feedback effects of the moderator, UO2 fuel, electrodes and reflector are considered. • The alkali metal heat pipe radiator is modeled with the two dimensional heat pipe model. • The steady state and the start-up procedure of the system are analyzed. - Abstract: A system analysis code coupled with the heat pipe model is developed to analyze the thermal-hydraulic characteristics of the improved TOPAZ-II reactor power system with a heat pipe radiator. The core thermal-hydraulic model, neutron physics model, and the coolant loop component models (including pump, volume accumulator, pipes and plenums) are established. The designed heat pipe radiator, which replaces the original pumped loop radiator, is also modeled, including two-dimensional heat pipe analysis model, fin model and coolant transport duct model. The system analysis code and the heat pipe model is coupled in the transport duct model. Steady state condition and start-up procedure of the improved TOPAZ-II system are calculated. The results show that the designed radiator can satisfy the waste heat rejection requirement of the improved power system. Meanwhile, the code can be used to obtained the thermal characteristics of the system transients such as the start-up process.

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

Transient thermal hydraulic modeling and analysis of ITER divertor plate system

International Nuclear Information System (INIS)

El-Morshedy, Salah El-Din; Hassanein, Ahmed

2009-01-01

A mathematical model has been developed/updated to simulate the steady state and transient thermal-hydraulics of the International Thermonuclear Experimental Reactor (ITER) divertor module. The model predicts the thermal response of the armour coating, divertor plate structural materials and coolant channels. The selected heat transfer correlations cover all operating conditions of ITER under both normal and off-normal situations. The model also accounts for the melting, vaporization, and solidification of the armour material. The developed model is to provide a quick benchmark of the HEIGHTS multidimensional comprehensive simulation package. The present model divides the coolant channels into a specified axial regions and the divertor plate into a specified radial zones, then a two-dimensional heat conduction calculation is created to predict the temperature distribution for both steady and transient states. The model is benchmarked against experimental data performed at Sandia National Laboratory for both bare and swirl tape coolant channel mockups. The results show very good agreements with the data for steady and transient states. The model is then used to predict the thermal behavior of the ITER plasma facing and structural materials due to plasma instability event where 60 MJ/m 2 plasma energy is deposited over 500 ms. The results for ITER divertor response is analyzed and compared with HEIGHTS results.

Transient thermal hydraulic modeling and analysis of ITER divertor plate system

Energy Technology Data Exchange (ETDEWEB)

El-Morshedy, Salah El-Din [Argonne National Laboratory, Argonne, IL (United States); Atomic Energy Authority, Cairo (Egypt)], E-mail: selmorshedy@etrr2-aea.org.eg; Hassanein, Ahmed [Purdue University, West Lafayette, IN (United States)], E-mail: hassanein@purdue.edu

2009-12-15

A mathematical model has been developed/updated to simulate the steady state and transient thermal-hydraulics of the International Thermonuclear Experimental Reactor (ITER) divertor module. The model predicts the thermal response of the armour coating, divertor plate structural materials and coolant channels. The selected heat transfer correlations cover all operating conditions of ITER under both normal and off-normal situations. The model also accounts for the melting, vaporization, and solidification of the armour material. The developed model is to provide a quick benchmark of the HEIGHTS multidimensional comprehensive simulation package. The present model divides the coolant channels into a specified axial regions and the divertor plate into a specified radial zones, then a two-dimensional heat conduction calculation is created to predict the temperature distribution for both steady and transient states. The model is benchmarked against experimental data performed at Sandia National Laboratory for both bare and swirl tape coolant channel mockups. The results show very good agreements with the data for steady and transient states. The model is then used to predict the thermal behavior of the ITER plasma facing and structural materials due to plasma instability event where 60 MJ/m{sup 2} plasma energy is deposited over 500 ms. The results for ITER divertor response is analyzed and compared with HEIGHTS results.

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

International Nuclear Information System (INIS)

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

1992-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2015-07-15

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

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

International Nuclear Information System (INIS)

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

2017-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2017-07-01

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

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

International Nuclear Information System (INIS)

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

2010-01-01

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

Steady-state entanglement and thermalization of coupled qubits in two common heat baths

Science.gov (United States)

Hu, Li-Zhen; Man, Zhong-Xiao; Xia, Yun-Jie

2018-03-01

In this work, we study the steady-state entanglement and thermalization of two coupled qubits embedded in two common baths with different temperatures. The common bath is relevant when the two qubits are difficult to be isolated to only contact with their local baths. With the quantum master equation constructed in the eigenstate representation of the coupled qubits, we have demonstrated the variations of steady-state entanglement with respect to various parameters of the qubits' system in both equilibrium and nonequilibrium cases of the baths. The coupling strength and energy detuning of the qubits as well as the temperature gradient of the baths are found to be beneficial to the enhancement of the entanglement. We note a dark state of the qubits that is free from time-evolution and its initial population can greatly influence the steady-state entanglement. By virtues of effective temperatures, we also study the thermalization of the coupled qubits and their variations with energy detuning.

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

International Nuclear Information System (INIS)

Lightston, M.F.; Rock, R.

1996-01-01

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

Calculation of the thermal and hydraulic states in rod cluster cores of light-water reactors

International Nuclear Information System (INIS)

Teichel, H.

1977-01-01

For calculating the three-dimensional steady distribution of the thermal and hydraulic states in rod cluster cores of light-water reactors, the subchannel analysis programs COLA 1 and COLA 2 have been developed. Both programs contain a multitude of competing empirical correlations which may be used by choice. The programs COLA 1 and COLA 2 differ in the calculation method and in the treatment of the boundary condition 'equal pressure at the end of all cooling channels' governing the problem. All parts of the programs are identical. By means of recomputed experiments statements on the accuracy of the results to be expected can be made. In addition, the different suitability of both programs for different experimental conditions are shown. (orig.) [de

The ITER divertor cassette. Steady state characterisation and draining and drying transient hydraulic analyses

International Nuclear Information System (INIS)

Pietro Alessandro Di Maio; Valerio Tomarchio; Giuseppe Vella; Irene Zammuto; Giovanni Dell'Orco

2005-01-01

gas bubbles. Due to the complex flow scheme of the hydraulic circuit, a pure theoretical study does not appears sufficient to address all the above mentioned items and an experimental validation of the models is mandatory. In addition to that, the assembly of the PFCs onto the cassette body as well as their integration by welding the coolant connections of the PFCs, also represent a critical step to be investigated. In order to investigate the aforementioned critical issues a theoretical and experimental research activity has been launched by the ENEA-Brasimone labs, in cooperation with the Department of Nuclear Engineering of the University of Palermo, with the specific aim of investigating the thermal-hydraulic behaviour of the whole divertor cassette both in steady state and operational or accidental transient conditions. The theoretical study, based on a computational approach, has been carried out with the RELAP5 code and the results obtained are herewith presented and critically discussed. In particular, the paper presents steady state and transient theoretical analyses intended to characterise the steady state behaviour of the cassette, determining flow distribution, pressure drop and CHF margin for each cooling channels, and to investigate the cassette behaviour during the draining and drying procedure, respectively. (authors)

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)

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

International Nuclear Information System (INIS)

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

2005-01-01

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

Thermal-hydraulic calculations for a fuel assembly in a European Pressurized Reactor using the RELAP5 code

Directory of Open Access Journals (Sweden)

Skrzypek Maciej

2015-09-01

Full Text Available The main object of interest was a typical fuel assembly, which constitutes a core of the nuclear reactor. The aim of the paper is to describe the phenomena and calculate thermal-hydraulic characteristic parameters in the fuel assembly for a European Pressurized Reactor (EPR. To perform thermal-hydraulic calculations, the RELAP5 code was used. This code allows to simulate steady and transient states for reactor applications. It is also an appropriate calculation tool in the event of a loss-of-coolant accident in light water reactors. The fuel assembly model with nodalization in the RELAP5 (Reactor Excursion and Leak Analysis Program code was presented. The calculations of two steady states for the fuel assembly were performed: the nominal steady-state conditions and the coolant flow rate decreased to 60% of the nominal EPR flow rate. The calculation for one transient state for a linearly decreasing flow rate of coolant was simulated until a new level was stabilized and SCRAM occurred. To check the correctness of the obtained results, the authors compared them against the reactor technical documentation available in the bibliography. The obtained results concerning steady states nearly match the design data. The hypothetical transient showed the importance of the need for correct cooling in the reactor during occurrences exceeding normal operation. The performed analysis indicated consequences of the coolant flow rate limitations during the reactor operation.

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

A phenomenological model of the thermal-hydraulics of convective boiling during the quenching of hot rod bundles: Part 2, Assessment of the model with steady-state and transient post-CHF data

International Nuclear Information System (INIS)

Unal, C.; Nelson, R.

1991-01-01

After completing the thermal-hydraulic model developed in a companion paper, we performed assessment calculations of the model using steady-state and transient post-critical heat flux (CHF) data. This paper discusses the results of those calculations. The hot-patch model, in conjunction with the other thermal-hydraulic models, was capable of modeling the Winfrith post-CHF hot-patch experiments. The hot-patch model kept the wall temperatures at the specified levels in the hot-patch regions and did not allow any quench-front propagation from either the bottom or the top of the test section. Among the four Winfrith runs selected to assess the hot-patch model, the average deviation in hot-patch power predictions was 15.4%, indicating reasonable predictions of the amount of energy transferred to the fluid by the hot patch. The interfacial heat-transfer model tended to slightly under-predict the vapor temperatures. The maximum difference between calculated and measured vapor superheats was 20%, with a 10% difference for the remainder of the runs considered. The wall-to-fluid heat transfer was predicted reasonably well, and the predicted wall superheats were in reasonable agreement with measured data with a maximum relative error of less than 13%. The effects of pressure, test section power, and flow rate on the axial variation of tube wall temperature are predicted reasonably well for a large range of operating parameters. A comparison of the predicted and measured local wall. The thermal-hydraulic model in TRAC/PF1-MOD2 was used to predict the axial variation of void fraction as measured in Winfrith post-CHF tests. The predictions for reflood calculations were reasonable. The model correctly predicted the trends in void fraction as a result of the effect of pressure and power, with the effect of pressure being more apparent than that of power. 13 refs

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

Quasi‐steady centrifuge method for unsaturated hydraulic properties

Science.gov (United States)

Caputo, Maria C.; Nimmo, John R.

2005-01-01

We have developed the quasi‐steady centrifuge (QSC) method as a variation of the steady state centrifuge method that can be implemented simply and inexpensively with greater versatility in terms of sample size and other features. It achieves these advantages by somewhat relaxing the criterion for steadiness of flow through the sample. This compromise entails an increase in measurement uncertainty but to a degree that is tolerable in most applications. We have tested this new approach with an easily constructed apparatus to establish a quasi‐steady flow of water in unsaturated porous rock samples spinning in a centrifuge, obtaining measurements of unsaturated hydraulic conductivity and water retention that agree with results of other methods. The QSC method is adaptable to essentially any centrifuge suitable for hydrogeologic applications, over a wide range of sizes and operating speeds. The simplified apparatus and greater adaptability of this method expands the potential for exploring situations that are common in nature but have been the subject of few laboratory investigations.

Quasi-steady centrifuge method for unsaturated hydraulic properties

Science.gov (United States)

Caputo, Maria C.; Nimmo, John R.

2005-11-01

We have developed the quasi-steady centrifuge (QSC) method as a variation of the steady state centrifuge method that can be implemented simply and inexpensively with greater versatility in terms of sample size and other features. It achieves these advantages by somewhat relaxing the criterion for steadiness of flow through the sample. This compromise entails an increase in measurement uncertainty but to a degree that is tolerable in most applications. We have tested this new approach with an easily constructed apparatus to establish a quasi-steady flow of water in unsaturated porous rock samples spinning in a centrifuge, obtaining measurements of unsaturated hydraulic conductivity and water retention that agree with results of other methods. The QSC method is adaptable to essentially any centrifuge suitable for hydrogeologic applications, over a wide range of sizes and operating speeds. The simplified apparatus and greater adaptability of this method expands the potential for exploring situations that are common in nature but have been the subject of few laboratory investigations.

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

International Nuclear Information System (INIS)

Walton, J.T.

1992-11-01

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

Thermal hydraulic analysis of the IPR-R1 TRIGA reactor

International Nuclear Information System (INIS)

Veloso, Marcelo Antonio; Fortini, Maria Auxiliadora

2002-01-01

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

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)

Steady-state and accident analyses of PBMR with the computer code SPECTRA

International Nuclear Information System (INIS)

Stempniewicz, Marek M.

2002-01-01

The SPECTRA code is an accident analysis code developed at NRG. It is designed for thermal-hydraulic analyses of nuclear or conventional power plants. The code is capable of analysing the whole power plant, including reactor vessel, primary system, various control and safety systems, containment and reactor building. The aim of the work presented in this paper was to prepare a preliminary thermal-hydraulic model of PBMR for SPECTRA, and perform steady state and accident analyses. In order to assess SPECTRA capability to model the PBMR reactors, a model of the INCOGEN system has been prepared first. Steady state and accident scenarios were analyzed for INCOGEN configuration. Results were compared to the results obtained earlier with INAS and OCTOPUS/PANTHERMIX. A good agreement was obtained. Results of accident analyses with PBMR model showed qualitatively good results. It is concluded that SPECTRA is a suitable tool for analyzing High Temperature Reactors, such as INCOGEN or for example PBMR (Pebble Bed Modular Reactor). Analyses of INCOGEN and PBMR systems showed that in all analyzed cases the fuel temperatures remained within the acceptable limits. Consequently there is no danger of release of radioactivity to the environment. It may be concluded that those are promising designs for future safe industrial reactors. (author)

Modelling and thermal hydraulic analysis of the Angra-2 nuclear reactor using RELAP5-3D code

International Nuclear Information System (INIS)

González Mantecón, Javier

2015-01-01

The evaluation of Nuclear Power Plants (NPPs) performance during steady-state and accident conditions has been one of the main research subjects in the nuclear field. In order to simulate the behavior of water-cooled reactors, several complex thermal-hydraulic codes systems have been developed. Particularly, the RELAP5 code, developed by the Idaho National Laboratory, is a best-estimate thermal-hydraulic analysis tool and one of the most used in nuclear industry. The RELAP5-3D 3.0.0 code was used to develop a detailed model of Angra 2 nuclear reactor using reference data from the Final Safety Analysis Report. Angra 2 is the second Brazilian NPP, which began commercial operation in 2001. The plant is equipped with a Pressurized Water Reactor (PWR) type with 3771.0 MWt. Simulations of the reactor behavior during normal operation conditions and postulated accident conditions were performed. Results achieved in the reactor steady-state simulation were compared with nominal parameters of the NPP. These results proved to be in good agreement, with relative errors less than 1%. In the transient simulation, the obtained results were coherent and satisfactory. This study demonstrates that the RELAP5-3D model is capable to reproduce the thermal-hydraulic behavior of the Angra-2 PWR during diverse operation conditions and it can contribute for the process of the plant safety analysis. (author)

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

Feasibility study for improved steady-state initialization algorithms for the RELAP5 computer code

International Nuclear Information System (INIS)

Paulsen, M.P.; Peterson, C.E.; Katsma, K.R.

1993-04-01

A design for a new steady-state initialization method is presented that represents an improvement over the current method used in RELAP5. Current initialization methods for RELAP5 solve the transient fluidflow balance equations simulating a transient to achieve steady-state conditions. Because the transient solution is used, the initial conditions may change from the desired values requiring the use of controllers and long transient running times to obtain steady-state conditions for system problems. The new initialization method allows the user to fix thermal-hydraulic values in volumes and junctions where the conditions are best known and have the code compute the initial conditions in other areas of the system. The steady-state balance equations and solution methods are presented. The constitutive, component, and specialpurpose models are reviewed with respect to modifications required for the new steady-state initialization method. The requirements for user input are defined and the feasibility of the method is demonstrated with a testbed code by initializing some simple channel problems. The initialization of the sample problems using, the old and the new methods are compared

Quasi-steady state thermal performances of a solar air heater with ...

African Journals Online (AJOL)

Quasi-steady state thermal performance of a solar air heater with a combined absorber is studied. The whole energy balance equations related to the system were articulated as a linear system of temperature equations. Solutions to this linear system were assessed from program based on an iterative process. The mean ...

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)

Effects of governing parameters on steady-state inter-wrapper flow in an LMFBR

International Nuclear Information System (INIS)

Moriya, Shoichi

2001-01-01

Hydraulic experiments were performed using a 1/8th scale rectangular model, based on a Japanese demonstration fast breeder reactor design, in order to study fundamental characteristics of interwrapper flows occurring under steady state conditions in an LMFBR. The steady state interwrapper flow of which direction was downward in the center region and upward in the peripheral region of a core barrel was observed because of the radial static pressure gradient in the upper part of the core barrel, produced by a core blockage effect resulting from an above core structure with a perforated skirt. Thermal stratification phenomena were moreover observed in the interwrapper region, created by the hot steady state interwrapper flow from an upper plenum and the cold leakage flow through the separated plate of the core barrel. The thermal interface was generated in higher part of the core barrel when the core blockage effect was smaller and Richardson number and the leakage flow rate ratio were larger. Significant temperature fluctuations occurred in the peripheral region of the core barrel, when the difference between the interface elevations in the center and peripheral regions of the core barrel was enough large. (author)

A method for statistical steady state thermal analysis of reactor cores

International Nuclear Information System (INIS)

Whetton, P.A.

1981-01-01

In a previous publication the author presented a method for undertaking statistical steady state thermal analyses of reactor cores. The present paper extends the technique to an assessment of confidence limits for the resulting probability functions which define the probability that a given thermal response value will be exceeded in a reactor core. Establishing such confidence limits is considered an integral part of any statistical thermal analysis and essential if such analysis are to be considered in any regulatory process. In certain applications the use of a best estimate probability function may be justifiable but it is recognised that a demonstrably conservative probability function is required for any regulatory considerations. (orig.)

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)

Thermal hydraulic analysis of the encapsulated nuclear heat source

Energy Technology Data Exchange (ETDEWEB)

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

2001-07-01

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

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

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

International Nuclear Information System (INIS)

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

2004-01-01

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

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)

Steady state thermal hydraulic analysis of a boiling water reactor core, for various power distributions, using computer code THABNA

International Nuclear Information System (INIS)

Venkat Raj, V.; Saha, D.

1976-01-01

The core of a boiling water reactor may see different power distributions during its operational life. How some of the typical power distributions affect some of the thermal hydraulic parameters such as pressure drop minimum critical heat flux ratio, void distribution etc. has been studied using computer code THABNA. The effect of an increase in the leakage flow has also been analysed. (author)

Comprehensive thermal-hydraulic and thermal-mechanical analysis of core and fuel rods for the safety validation of real refueling at the Kozloduy WWER-440

Energy Technology Data Exchange (ETDEWEB)

Stefanova, S; Panajotov, D; Ilieva, B; Vitkova, M; Simeonova, V; Passage, G; Manolova, M [Bylgarska Akademiya na Naukite, Sofia (Bulgaria). Inst. za Yadrena Izsledvaniya i Yadrena Energetika

1996-12-31

Safety analysis aimed at determination of thermal-hydraulic and thermal-mechanical margins of core and fuel rods has been carried out using computer codes COBSOFM and PIN-micro. Thermal-hydraulic calculations for the part of the core with maximum heat flux during steady-state regime show that the coolant, cladding and fuel temperatures are within the design limits. A severe accident with reactor blackout has been simulated. It is found that at 95% probability level there is no boiling crisis anywhere in the core. The thermal-mechanical parameters of working assembly fuel rod with maximum load have been calculated. The assembly linear power reached a maximum of 25 kW/m during the second fuel cycle, the fuel temperature remaining well below 1000{sup o} C. As the fuel assembly with typical power history has enough safety margins, it was proposed to use it for one more cycle. 4 refs., 12 figs.

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.

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.

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

Report on the thermal-hydraulics computational component

International Nuclear Information System (INIS)

Laughton, T.; Jones, B.G.

1996-01-01

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

Development of design technology on thermal-hydraulic performance in tight-lattice rod bundles. II-rod bowing effect on boiling transition

International Nuclear Information System (INIS)

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

2007-01-01

A thermal-hydraulic feasibility project for an Innovative Water Reactor for Flexible fuel cycle (FLWR) has been performed since 2002. In this R and D project, large-scale thermal-hydraulic tests, several model experiments and development of advanced numerical analysis codes have been carried out. In this paper, we will describe the critical power characteristics in a 37-rod tight-lattice bundle with rod-bowing under both steady and transient states. It is observed that no matter it is run under a steady or a transient state, boiling transition (BT) always occurs axially at exit elevation of upper high-heat-flux region and transversely in the central area of the bundle. Steady critical power increases monotonically with the increase of mass velocity, with the decrease of inlet water temperature and with the decrease of exit pressure. These trends are same as those in the base case test without rod-bowing. The steady critical power with rod-bowing is about 10% lower than that without rod-bowing. For the postulated power increase and flow decrease cases that may be possibly met in a normal operation of the FLWR, it is confirmed that no BT occurs when Initial Critical Power Ratio (ICPR) is 1.3. Moreover, when the transitions are run under severer ICPR that causes BT, the transient critical powers are generally same as the steady ones. The experiments are analyzed with TRAC-BF1 code. The TRAC-BF1 code shows good prediction for the occurrence or the non occurrence of the BT and predicts the BT starting time within the accuracy of critical power correlation. Traditional quasi - steady state prediction of the transient BT is confirmed being applicable for the postulated abnormal transient processes in the tight lattice bundle with rod - bowing. (author)

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)

Steady-state analysis of a conceptual offshore wind turbine driven electricity and thermocline energy extraction plant

OpenAIRE

Buhagiar, Daniel; Sant, Tonio

2014-01-01

A system for using offshore wind energy to generate electricity and simultaneously extract thermal energy is proposed. This concept is based on an offshore wind turbine driven hydraulic pump supplying deep seawater under high pressure to a land based plant consisting of a hydroelectric power generation unit and heat exchanger. A steady-state system model is developed using empirical formulae. The mathematical model comprises the fundamental system sub-models that are categoris...

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

International Nuclear Information System (INIS)

Wnek, W.J.; Ramshaw, J.D.; Trapp, J.A.; Hughes, E.D.; Solbrig, C.W.

1975-11-01

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

Fundamental approaches for analysis thermal hydraulic parameter for Puspati Research Reactor

International Nuclear Information System (INIS)

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

2016-01-01

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

Fundamental approaches for analysis thermal hydraulic parameter for Puspati Research Reactor

Science.gov (United States)

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

2016-01-01

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

Fundamental approaches for analysis thermal hydraulic parameter for Puspati Research Reactor

Energy Technology Data Exchange (ETDEWEB)

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

2016-01-22

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

A fission-fusion hybrid reactor in steady-state L-mode tokamak configuration with natural uranium

International Nuclear Information System (INIS)

Reed, Mark; Parker, Ronald R.; Forget, Benoit

2012-01-01

This work develops a conceptual design for a fusion-fission hybrid reactor operating in steady-state L-mode tokamak configuration with a subcritical natural or depleted uranium pebble bed blanket. A liquid lithium-lead alloy breeds enough tritium to replenish that consumed by the D-T fusion reaction. The fission blanket augments the fusion power such that the fusion core itself need not have a high power gain, thus allowing for fully non-inductive (steady-state) low confinement mode (L-mode) operation at relatively small physical dimensions. A neutron transport Monte Carlo code models the natural uranium fission blanket. Maximizing the fission power gain while breeding sufficient tritium allows for the selection of an optimal set of blanket parameters, which yields a maximum prudent fission power gain of approximately 7. A 0-D tokamak model suffices to analyze approximate tokamak operating conditions. This fission blanket would allow the fusion component of a hybrid reactor with the same dimensions as ITER to operate in steady-state L-mode very comfortably with a fusion power gain of 6.7 and a thermal fusion power of 2.1 GW. Taking this further can determine the approximate minimum scale for a steady-state L-mode tokamak hybrid reactor, which is a major radius of 5.2 m and an aspect ratio of 2.8. This minimum scale device operates barely within the steady-state L-mode realm with a thermal fusion power of 1.7 GW. Basic thermal hydraulic analysis demonstrates that pressurized helium could cool the pebble bed fission blanket with a flow rate below 10 m/s. The Brayton cycle thermal efficiency is 41%. This reactor, dubbed the Steady-state L-mode non-Enriched Uranium Tokamak Hybrid (SLEUTH), with its very fast neutron spectrum, could be superior to pure fission reactors in terms of breeding fissile fuel and transmuting deleterious fission products. It would likely function best as a prolific plutonium breeder, and the plutonium it produces could actually be more

Transient thermal, hydraulic, and mechanical analysis of a counter flow offset strip fin intermediate heat exchanger using an effective porous media approach

Science.gov (United States)

Urquiza, Eugenio

This work presents a comprehensive thermal hydraulic analysis of a compact heat exchanger using offset strip fins. The thermal hydraulics analysis in this work is followed by a finite element analysis (FEA) to predict the mechanical stresses experienced by an intermediate heat exchanger (IHX) during steady-state operation and selected flow transients. In particular, the scenario analyzed involves a gas-to-liquid IHX operating between high pressure helium and liquid or molten salt. In order to estimate the stresses in compact heat exchangers a comprehensive thermal and hydraulic analysis is needed. Compact heat exchangers require very small flow channels and fins to achieve high heat transfer rates and thermal effectiveness. However, studying such small features computationally contributes little to the understanding of component level phenomena and requires prohibitive computational effort using computational fluid dynamics (CFD). To address this issue, the analysis developed here uses an effective porous media (EPM) approach; this greatly reduces the computation time and produces results with the appropriate resolution [1]. This EPM fluid dynamics and heat transfer computational code has been named the Compact Heat Exchanger Explicit Thermal and Hydraulics (CHEETAH) code. CHEETAH solves for the two-dimensional steady-state and transient temperature and flow distributions in the IHX including the complicating effects of temperature-dependent fluid thermo-physical properties. Temperature- and pressure-dependent fluid properties are evaluated by CHEETAH and the thermal effectiveness of the IHX is also calculated. Furthermore, the temperature distribution can then be imported into a finite element analysis (FEA) code for mechanical stress analysis using the EPM methods developed earlier by the University of California, Berkeley, for global and local stress analysis [2]. These simulation tools will also allow the heat exchanger design to be improved through an

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

International Nuclear Information System (INIS)

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

2004-01-01

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

Thermal-hydraulic analysis techniques for axisymmetric pebble bed nuclear reactor cores

International Nuclear Information System (INIS)

Stroh, K.R.

1979-03-01

The pebble bed reactor's cylindrical core volume contains a random bed of small, spherical fuel-moderator elements. These graphite spheres, containing a central region of dispersed coated-particle fissile and fertile material, are cooled by high pressure helium flowing through the connected interstitial voids. A mathematical model and numerical solution technique have been developed which allow calculation of macroscopic values of thermal-hydraulic variables in an axisymmetric pebble bed nuclear reactor core. The computer program PEBBLE is based on a mathematical model which treats the bed macroscopically as a generating, conducting porous medium. The steady-state model uses a nonlinear Forchheimer-type relation between the coolant pressure gradient and mass flux, with newly derived coefficients for the linear and quadratic resistance terms. The remaining equations in the model make use of mass continuity, and thermal energy balances for the solid and fluid phases

The analysis of the annular fuel performance in steady state condition by using AFPAC code

International Nuclear Information System (INIS)

He Xiaojun; Ji Songtao; Zhang Yingchao

2012-01-01

The fuel performance code AFPAC v1.0 is used to analyze annular fuel's behavior under steady state conditions, including neutronics, thermal hydraulic, rod deformation, fission gas release and rod internal pressure. The calculation results show that: 1) Annular fuel has a good steady irradiation performance at 150% power level as current LWRs' with burnup up to 50 GWd/t, and all parameters, such as temperature, rod internal pressure and rod deformation, are meet the rod design criteria for current fuel of PWRs: 2) Compared to the solid fuel under the same irradiation condition. annular fuel has lower temperature, smaller deformation, lower fission gas release and lower pressure at EOL. From the point of view of steady irradiation performance, the safety of reactors can significantly improved by u sing the annular fuel. (authors)

Engineering design and thermal hydraulics of plasma facing components of SST-1

International Nuclear Information System (INIS)

Pragash, N. Ravi; Chaudhuri, P.; Santra, P.; Chenna Reddy, D.; Khirwadkar, S.; Saxena, Y.C.

2001-01-01

SST-1 is a medium size tokamak with super conducting magnetic field coils. All the subsystems of SST-1 are designed for quasi steady state (∼1000 s) operation. Plasma Facing Components (PFCs) of SST-1 consisting of divertors, passive stabilizers, baffles and poloidal limiters are also designed to be compatible for steady state operation. As SST-1 is designed to run double null divertor plasmas, these components also have up-down symmetry. A closed divertor configuration is chosen to produce high recycling and high pumping speed in the divertor region. All the PFC are made of copper alloys (CuCrZr and CuZr) on which graphite tiles are mechanically attached. These copper alloy back plates are actively cooled with water flowing in the channels grooved on them with the main consideration in the design of PFCs as the steady state heat removal of about 1.0 MW/m 2 . In addition to be able to remove high heat fluxes, the PFCs are also designed to be compatible for baking at 350 degree sign C. Extensive studies, involving different flow parameters and various cooling layouts, have been done to select the final cooling parameters and layout. Thermal response of the PFCs and vacuum vessel during baking, has been calculated using a FORTRAN code and a 2-D finite element analysis. The PFCs and their supports are also designed to withstand large electro-magnetic forces. Finite element analysis using ANSYS software package is used in this and other PFCs design. The engineering design including thermal hydraulics for cooling and baking of all the PFCs is completed. Poloidal limiters are being fabricated. The remaining PFCs, viz. divertors, stabilizers and baffles are likely to go for fabrication in the next few months. The detailed engineering design, the finite element calculations in the structural and thermal designs are presented in this paper

HEATHYD, Steady-State Thermal Hydraulic Analysis of Low-Enriched U Fuel Reactor

International Nuclear Information System (INIS)

NABBI, R.

1989-01-01

1 - Description of program or function: HEATHYD is a code for the steady-state heat transfer calculation of research nuclear reactors with forced convection. It models heat transfer and coolant flow for assemblies of parallel fuel plates of MTR type with any axial power distribution. The thermodynamic model accounts for single phase cooling and sub- cooled boiling condition using the transition criterion of Bergeles-Rosenow. In addition to the calculation of the channel flow velocities and coolant pressure drops, HEATHYD calculates axial distribution of the coolant and clad-surface temperatures. Safety margins to the critical heat flux as a result of burnout condition or flow instability are determined. 2 - Method of solution: Applying the finite difference method, HEATHYD solves the equations of heat conduction and heat transfer to the coolant. For the physical properties of the coolant as a function of the coolant temperature polynomials of degree 6 are used. Depending on the coolant condition, different correlations for the heat transfer coefficient can be applied. The analysis of the critical cooling conditions resulting in burnout or flow instability, is performed according to the correlations developed by Mirshak/ Labuntsov and Forgan/Whittle

A method for statistical steady state thermal analysis of reactor cores

International Nuclear Information System (INIS)

Whetton, P.A.

1980-01-01

This paper presents a method for performing a statistical steady state thermal analysis of a reactor core. The technique is only outlined here since detailed thermal equations are dependent on the core geometry. The method has been applied to a pressurised water reactor core and the results are presented for illustration purposes. Random hypothetical cores are generated using the Monte-Carlo method. The technique shows that by splitting the parameters into two types, denoted core-wise and in-core, the Monte Carlo method may be used inexpensively. The idea of using extremal statistics to characterise the low probability events (i.e. the tails of a distribution) is introduced together with a method of forming the final probability distribution. After establishing an acceptable probability of exceeding a thermal design criterion, the final probability distribution may be used to determine the corresponding thermal response value. If statistical and deterministic (i.e. conservative) thermal response values are compared, information on the degree of pessimism in the deterministic method of analysis may be inferred and the restrictive performance limitations imposed by this method relieved. (orig.)

FRAPCON-2: A Computer Code for the Calculation of Steady State Thermal-Mechanical Behavior of Oxide Fuel Rods

Energy Technology Data Exchange (ETDEWEB)

Berna, G. A; Bohn, M. P.; Rausch, W. N.; Williford, R. E.; Lanning, D. D.

1981-01-01

FRAPCON-2 is a FORTRAN IV computer code that calculates the steady state response of light Mater reactor fuel rods during long-term burnup. The code calculates the temperature, pressure, deformation, and tai lure histories of a fuel rod as functions of time-dependent fuel rod power and coolant boundary conditions. The phenomena modeled by the code include (a) heat conduction through the fuel and cladding, (b) cladding elastic and plastic deformation, (c) fuel-cladding mechanical interaction, (d) fission gas release, (e} fuel rod internal gas pressure, (f) heat transfer between fuel and cladding, (g) cladding oxidation, and (h) heat transfer from cladding to coolant. The code contains necessary material properties, water properties, and heat transfer correlations. FRAPCON-2 is programmed for use on the CDC Cyber 175 and 176 computers. The FRAPCON-2 code Is designed to generate initial conditions for transient fuel rod analysis by either the FRAP-T6 computer code or the thermal-hydraulic code, RELAP4/MOD7 Version 2.

TITAN: an advanced three-dimensional coupled neutronic/thermal-hydraulics code for light water nuclear reactor core analysis

International Nuclear Information System (INIS)

Griggs, D.P.; Kazimi, M.S.; Henry, A.F.

1984-06-01

The three-dimensional nodal neutronics code QUANDRY and the three-dimensional two-fluid thermal-hydraulics code THERMIT are combined into TITAN. Steady-state and transient coupling methodologies based upon a tandem structure were devised and implemented. Additional models for nuclear feedback, equilibrium xenon and direct moderator heating were added. TITAN was tested using a boiling water two channel problem and the coupling methodologies were shown to be effective. Simulated turbine trip transients and several control rod withdrawal transients were analyzed with good results. Sensitivity studies indicated that the time-step size can affect transient results significantly. TITAN was also applied to a quarter core PWR problem based on a real reactor geometry. The steady-state results were compared to a solution produced by MEKIN-B and poor agreement between the horizontal power shapes was found. Calculations with various mesh spacings showed that the mesh spacings in the MEKIN-B analysis were too large to produce accurate results with a finite difference method. The TITAN results were shown to be reasonable. A pair of control rod ejection accidents were also analyzed with TITAN. A comparison of the TITAN PWR control rod ejection results with results from coupled point kinetics/thermal-hydraulics analyses showed that the point kinetics method used (adiabatic method for control rod reactivities, steady-state flux shape for core-averaged reactivity feedback) underpredicted the power excursion in one case and overpredicted it in the other. It was therefore concluded that point kinetics methods should be used with caution and that three-dimensional codes like TITAN are superior for analyzing PWR control rod ejection transients

Thermal Conductivity of EB-PVD Thermal Barrier Coatings Evaluated by a Steady-State Laser Heat Flux Technique

Science.gov (United States)

Zhu, Dongming; Miller, Robert A.; Nagaraj, Ben A.; Bruce, Robert W.

2000-01-01

The thermal conductivity of electron beam-physical vapor deposited (EB-PVD) Zr02-8wt%Y2O3 thermal barrier coatings was determined by a steady-state heat flux laser technique. Thermal conductivity change kinetics of the EB-PVD ceramic coatings were also obtained in real time, at high temperatures, under the laser high heat flux, long term test conditions. The thermal conductivity increase due to micro-pore sintering and the decrease due to coating micro-delaminations in the EB-PVD coatings were evaluated for grooved and non-grooved EB-PVD coating systems under isothermal and thermal cycling conditions. The coating failure modes under the high heat flux test conditions were also investigated. The test technique provides a viable means for obtaining coating thermal conductivity data for use in design, development, and life prediction for engine applications.

Steady-State Thermal-Hydraulics Analyses for the Conversion of the BR2 Reactor to LEU

Energy Technology Data Exchange (ETDEWEB)

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

2016-09-01

BR2 is a research reactor used for radioisotope production and materials testing. It’s a tank-in-pool type reactor cooled by light water and moderated by beryllium and light water. The reactor core consists of a beryllium moderator forming a matrix of 79 hexagonal prisms in a hyperboloid configuration; each having a central bore that can contain a variety of different components such as a fuel assembly, a control or regulating rod, an experimental device, or a beryllium or aluminum plug. Based on a series of tests, the BR2 operation is currently limited to a maximum allowable heat flux of 470 W/cm² to ensure fuel plate integrity during steady-state operation and after a loss-of-flow/loss-of-pressure accident. A feasibility study for the conversion of the BR2 reactor from highly-enriched uranium (HEU) to low-enriched uranium (LEU) fuel was previously performed to verify it can operate safely at the same maximum nominal steady-state heat flux. An assessment was also performed to quantify the heat fluxes at which the onset of flow instability and critical heat flux occur for each fuel type. This document updates and expands these results for the current representative core configuration (assuming a fresh beryllium matrix) by evaluating the onset of nucleate boiling (ONB), onset of fully developed nucleate boiling (FDNB), onset of flow instability (OFI) and critical heat flux (CHF).

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)

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)

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

International Nuclear Information System (INIS)

Staalek, Mathias

2008-03-01

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

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

Energy Technology Data Exchange (ETDEWEB)

Staalek, Mathias

2008-03-15

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

Estimating the saturated soil hydraulic conductivity by the near steady-state phase of a beerkan infiltration run

Science.gov (United States)

Di Prima, Simone; Bagarello, Vincenzo; Iovino, Massimo

2017-04-01

Simple infiltration experiments carried out in the field allow an easy and inexpensive way of characterizing soil hydraulic behavior, maintaining the functional connection of the sampled soil volume with the surrounding soil. The beerkan method consists of a three-dimensional (3D) infiltration experiment at zero pressure head (Haverkamp et al., 1996). It uses a simple annular ring inserted to a depth of about 0.01 m to avoid lateral loss of the ponded water. Soil disturbance is minimized by the limited ring insertion depth. Infiltration time of small volumes of water repeatedly poured on the confined soil are measured to determine the cumulative infiltration. Different algorithms based on this methodology (the so-called BEST family of algorithms) were developed for the determination of soil hydraulic characteristic parameters (Bagarello et al., 2014a; Lassabatere et al., 2006; Yilmaz et al., 2010). Recently, Bagarello et al. (2014b) developed a Simplified method based on a Beerkan Infiltration run (SBI method) to determine saturated soil hydraulic conductivity, Ks, by only the transient phase of a beerkan infiltration run and an estimate of the α* parameter, expressing the relative importance of gravity and capillary forces during an infiltration process (Reynolds and Elrick, 1990). However, several problems yet arise with the existing BEST-algorithms and the SBI method, including (i) the need of supplementary field and laboratory measurements (Bagarello et al., 2013); (ii) the difficulty to detect a linear relationship between I / √t and √t in the early stage of the infiltration process (Bagarello et al., 2014b); (iii) estimation of negative Ks values for hydrophobic soils (Di Prima et al., 2016). In this investigation, a new Simplified method based on the analysis of the Steady-state Beerkan Infiltration run (SSBI method) was proposed and tested. In particular, analytical data were generated to simulate beerkan infiltration experiments for six contrasting

Steady thermal stress and strain rates in a rotating circular cylinder under steady state temperature

Directory of Open Access Journals (Sweden)

Pankaj Thakur

2014-01-01

Full Text Available Thermal stress and strain rates in a thick walled rotating cylinder under steady state temperature has been derived by using Seth’s transition theory. For elastic-plastic stage, it is seen that with the increase of temperature, the cylinder having smaller radii ratios requires lesser angular velocity to become fully plastic as compared to cylinder having higher radii ratios The circumferential stress becomes larger and larger with the increase in temperature. With increase in thickness ratio stresses must be decrease. For the creep stage, it is seen that circumferential stresses for incompressible materials maximum at the internal surface as compared to compressible material, which increase with the increase in temperature and measure n.

A simple differential steady-state method to measure the thermal conductivity of solid bulk materials with high accuracy.

Science.gov (United States)

Kraemer, D; Chen, G

2014-02-01

Accurate measurements of thermal conductivity are of great importance for materials research and development. Steady-state methods determine thermal conductivity directly from the proportionality between heat flow and an applied temperature difference (Fourier Law). Although theoretically simple, in practice, achieving high accuracies with steady-state methods is challenging and requires rather complex experimental setups due to temperature sensor uncertainties and parasitic heat loss. We developed a simple differential steady-state method in which the sample is mounted between an electric heater and a temperature-controlled heat sink. Our method calibrates for parasitic heat losses from the electric heater during the measurement by maintaining a constant heater temperature close to the environmental temperature while varying the heat sink temperature. This enables a large signal-to-noise ratio which permits accurate measurements of samples with small thermal conductance values without an additional heater calibration measurement or sophisticated heater guards to eliminate parasitic heater losses. Additionally, the differential nature of the method largely eliminates the uncertainties of the temperature sensors, permitting measurements with small temperature differences, which is advantageous for samples with high thermal conductance values and/or with strongly temperature-dependent thermal conductivities. In order to accelerate measurements of more than one sample, the proposed method allows for measuring several samples consecutively at each temperature measurement point without adding significant error. We demonstrate the method by performing thermal conductivity measurements on commercial bulk thermoelectric Bi2Te3 samples in the temperature range of 30-150 °C with an error below 3%.

Steady State Thermal Analyses of SCEPTOR X-57 Wingtip Propulsion

Science.gov (United States)

Schnulo, Sydney L.; Chin, Jeffrey C.; Smith, Andrew D.; Dubois, Arthur

2017-01-01

Electric aircraft concepts enable advanced propulsion airframe integration approaches that promise increased efficiency as well as reduced emissions and noise. NASA's fully electric Maxwell X-57, developed under the SCEPTOR program, features distributed propulsion across a high aspect ratio wing. There are 14 propulsors in all: 12 high lift motor that are only active during take off and climb, and 2 larger motors positioned on the wingtips that operate over the entire mission. The power electronics involved in the wingtip propulsion are temperature sensitive and therefore require thermal management. This work focuses on the high and low fidelity heat transfer analysis methods performed to ensure that the wingtip motor inverters do not reach their temperature limits. It also explores different geometry configurations involved in the X-57 development and any thermal concerns. All analyses presented are performed at steady state under stressful operating conditions, therefore predicting temperatures which are considered the worst-case scenario to remain conservative.

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.

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

Engineering and thermal-hydraulics design of PFC cooling for SST-1 Tokamak

International Nuclear Information System (INIS)

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

2003-01-01

The main consideration in the design of the PFC cooling for SST-1 tokamak is the steady state heat removal of upto 1MW/m2. The PFC also has been design to withstand the peak heat fluxes without significant erosion such that frequent replacement is not necessary. Proper brazing of cooling tube on the copper back plate is necessary for the efficient heat transfer from the tube to the back plate. Design considerations included 2-D steady state and transient tile temperature distribution and resulting thermal loads in PFC during baking, and cooling, coolant parameters necessary to maintain optimum thermal-hydraulic design, and tile fitting mechanism. Finite Element (FE) models using ANSYS have been developed to conduct the heat transfer and stress analyses of the PFC to understand its thermal and mechanical behaviors. The temperature distribution results for different PFC obtained by FE results were assessed by comparison with 2-D Finite Difference code. The results of the calculation led to a good understanding of the coolant flow behavior and the temperature distribution in the tube wall and the different parts of the PFC. The contact at the brazed joint of the tube to the backplate is critical for the above application. The manufactured modules need to be evaluated for the quality of brazed joint. Using an infra-red-camera, spatial and temporal evaluation of the temperature profile is studied under various flow parameters. These results of this study will be presented in details in this paper

3D thermal-hydraulic analysis on core of PWR nuclear power station

International Nuclear Information System (INIS)

Yao Zhaohui; Wang Xuefang; Shen Mengyu

1997-01-01

Thermal hydraulic analysis of core is of great importance in reactor safety analysis. A computer code, thermal hydraulic analysis porous medium analysis (THAPMA), has been developed to simulate the flow and heat transfer characteristics of reactor components. It has been proved reliable by several numerical tests. In the THAPMA code, a new difference scheme and solution method have been studied in developing the computer software. For the difference scheme, a second order accurate, high resolution scheme, called WSUC scheme, has been proposed. This scheme is total variation bounded and unconditionally stable in convective numeral stability. Numerical tests show that the WSUC is better in accuracy and resolution than the 1-st order upwind, 2-nd order upwind, SOUCUP by Zhu and Rodi. In solution method, a modified PISO algorithm is used, which is not only simpler but also more accurate and more rapid in convergence than the original PISO algorithm. Moreover, the modified PISO algorithm can effectively solve steady and transient state problem. Besides, with the THAPMA code, the flow and heat transfer phenomena in reactor core have been numerically simulated in the light of the design condition of Qinshan PWR nuclear power station (the second-term project). The simulation results supply a theoretical basis for the core design

Steady and unsteady calculations on thermal striping phenomena in triple-parallel jet

Energy Technology Data Exchange (ETDEWEB)

Yu, Y.Q., E-mail: yyu@anl.gov [Nuclear Engineering Division, Argonne National Laboratory, Lemont, IL 60439 (United States); Merzari, E.; Thomas, J.W. [Nuclear Engineering Division, Argonne National Laboratory, Lemont, IL 60439 (United States); Obabko, A. [Mathematics and Computer Science Division, Argonne National Laboratory, Lemont, IL 60439 (United States); Aithal, S.M. [Computing, Environment and Life Sciences Directorate, Argonne National Laboratory, Lemont, IL 60439 (United States)

2017-02-15

Highlights: • Both steady (RANS) and unsteady (URANS, LES) methods were applied to study thermal striping. • The unsteady results exhibited reasonably good agreement with experimental results. • The parametric studies on the effects of mesh density and boundary conditions on the accuracy of the overall solutions were also conducted. - Abstract: The phenomenon of thermal striping is encountered in liquid metal cooled fast reactors (LMFR), in which temperature fluctuation due to convective mixing between hot and cold fluids can lead to a possibility of crack initiation and propagation in the structure due to high cycle thermal fatigue. Using sodium experiments of parallel triple jets configuration performed by Japan Atomic Energy Agency (JAEA) as benchmark, numerical simulations were carried out to evaluate the temperature fluctuation characteristics in fluid and the transfer characteristics of temperature fluctuation from fluid to structure, which is important to assess the potential thermal fatigue damage. In this study, both steady (RANS) and unsteady (URANS, LES) methods were applied to predict the temperature fluctuations of thermal striping. The parametric studies on the effects of mesh density and boundary conditions on the accuracy of the overall solutions were also conducted. The velocity, temperature and temperature fluctuation intensity distribution were compared with the experimental data. As expected, steady calculation has limited success in predicting the thermal–hydraulic characteristics of the thermal striping, highlighting the limitations of the RANS approach in unsteady heat transfer simulations. The unsteady results exhibited reasonably good agreement with experimental results for temperature fluctuation intensity, as well as the average temperature and velocity components at the measurement locations.

Steady and unsteady calculations on thermal striping phenomena in triple-parallel jet

International Nuclear Information System (INIS)

Yu, Y.Q.; Merzari, E.; Thomas, J.W.; Obabko, A.; Aithal, S.M.

2017-01-01

Highlights: • Both steady (RANS) and unsteady (URANS, LES) methods were applied to study thermal striping. • The unsteady results exhibited reasonably good agreement with experimental results. • The parametric studies on the effects of mesh density and boundary conditions on the accuracy of the overall solutions were also conducted. - Abstract: The phenomenon of thermal striping is encountered in liquid metal cooled fast reactors (LMFR), in which temperature fluctuation due to convective mixing between hot and cold fluids can lead to a possibility of crack initiation and propagation in the structure due to high cycle thermal fatigue. Using sodium experiments of parallel triple jets configuration performed by Japan Atomic Energy Agency (JAEA) as benchmark, numerical simulations were carried out to evaluate the temperature fluctuation characteristics in fluid and the transfer characteristics of temperature fluctuation from fluid to structure, which is important to assess the potential thermal fatigue damage. In this study, both steady (RANS) and unsteady (URANS, LES) methods were applied to predict the temperature fluctuations of thermal striping. The parametric studies on the effects of mesh density and boundary conditions on the accuracy of the overall solutions were also conducted. The velocity, temperature and temperature fluctuation intensity distribution were compared with the experimental data. As expected, steady calculation has limited success in predicting the thermal–hydraulic characteristics of the thermal striping, highlighting the limitations of the RANS approach in unsteady heat transfer simulations. The unsteady results exhibited reasonably good agreement with experimental results for temperature fluctuation intensity, as well as the average temperature and velocity components at the measurement locations.

RELAP5 capabilities in thermal-hydraulic prediction of SBWR containment behaviour: PANDA steady state and transient tests evaluation

International Nuclear Information System (INIS)

Faluomi, V.; Aksan, S.N.

2000-01-01

This paper summarizes the results of the qualification activity of RELAP5/Mod3.2 code performed using PANDA steady state and integral test experimental data. The steady state tests evaluate the PCC performances in removing decay heat power in presence and in absence of non-condensable gases, while the considered integral test (M3) simulates the transient following a break in the main steam line of the SBWR, using, as nominal initial conditions, those calculated for the SBWR under SSAR assumptions at one hour into the LOCA. The results obtained simulating both types of tests show a rather good and robust overall code behavior both in the simulation of steady state test and in the representation of the integral test considered: most of the main experimental results (WW/DW pressures, PCC heat exchange) were well represented by the code. The different studies performed indicated that: Different models of PCC pool lead a different trend of system pressure, and sometimes to an opening of vacuum breaker valves, that does not occur in the transient; The code underestimate the heat exchanged between PCC pool and tubes: n the considered test the system pressure is slightly overestimated (maximum 2% more than the experimental value). This fact is also proved by the differences in the temperature of the condensing mixture in the PCC, quite large in all the performed studies; The treatment of the non condensable gases, as implemented in the code, lead some errors in the calculation of the heat transfer coefficient in the PCC components and generally slow down the overall calculation. In general terms, the RELAP5/Mod3.2 was found to be suitable to represent the SBWR containment behavior under the conditions specified in the experimental side. (author)

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.

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)

Coupled MCNP - SAS-SFR calculations for sodium fast reactor core at steady-state - 15460

International Nuclear Information System (INIS)

Ponomarev, A.; Travleev, A.; Pfrang, W.; Sanchez, V.

2015-01-01

The prediction of core parameters at steady state is the first step when studying core accident transient behaviour. At this step thermal hydraulics (TH) and core geometry parameters are calculated corresponding to initial operating conditions. In this study we present the coupling of the SAS-SFR code to the Monte-Carlo neutron transport code MCNP at steady state together with application to the European Sodium Fast Reactor (ESFR). The SAS-SFR code employs a multi-channel core representation where each channel represents subassemblies with similar power, thermal-hydraulics and pin mechanics conditions. For every axial node of every channel the individual geometry and material compositions parameters are calculated in accord with power and cooling conditions. This requires supplying the SAS-SFR-code with nodal power values which should be calculated by neutron physics code with given realistic core parameters. In the conventional approach the neutron physics model employs some core averaged TH and geometry data (fuel temperature, coolant density, core axial and radial expansion). In this study we organize a new approach coupling the MCNP neutron physics models and the SAS-SFR models, so that calculations of power can be improved by using distributed core parameters (TH and geometry) taken from SAS-SFR. The MCNP code is capable to describe cores with distributed TH parameters and even to model non-uniform axial expansion of fuel subassemblies. In this way, core TH and geometrical data calculated by SAS-SFR are taken into account accurately in the neutronics model. The coupling implementation is done by data exchange between two codes with help of processing routines managed by driver routine. Currently it is model-specific and realized for the ESFR 'Reference Oxide' core. The Beginning-Of-Life core state is considered with 10 channel representation for fuel subassemblies. For this model several sets of coupled calculations are performed, in which different

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

single assembly geometry with well-controlled boundary conditions simplified interpretation of results. Two different arrangements of ducting were used to mimic conditions for aboveground and belowground storage configurations for vertical, dry cask systems with canisters. Transverse and axial temperature profiles were measured throughout the test assembly. The induced air mass flow rate was measured for both the aboveground and belowground configurations. In addition, the impact of cross-wind conditions on the belowground configuration was quantified. Over 40 unique data sets were collected and analyzed for these efforts. Fourteen data sets for the aboveground configuration were recorded for powers and internal pressures ranging from 0.5 to 5.0 kW and 0.3 to 800 kPa absolute, respectively. Similarly, fourteen data sets were logged for the belowground configuration starting at ambient conditions and concluding with thermal-hydraulic steady state. Over thirteen tests were conducted using a custom-built wind machine. The results documented in this report highlight a small, but representative, subset of the available data from this test series. This addition to the dry cask experimental database signifies a substantial addition of first-of-a-kind, high-fidelity transient and steady-state thermal-hydraulic data sets suitable for CFD model validation.

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.

Development of subchannel analysis code MATRA-LMR for KALIMER subassembly thermal-hydraulics

International Nuclear Information System (INIS)

Won-Seok Kim; Young-Gyun Kim

2000-01-01

In the sodium cooled liquid metal reactors, the design limit are imposed on the maximum temperatures of claddings and fuel pins. Thus an accurate prediction of core coolant/fuel temperature distribution is essential to the LMR core thermal-hydraulic design. The detailed subchannel thermal-hydraulic analysis code MATRA-LMR (Multichannel Analyzer for Steady States and Transients in Rod Arrays for Liquid Metal Reactors) is being developed for KALIMER core design and analysis, based on COBRA-IV-i and MATRA. The major modifications and improvements implemented into MATRA-LMR are as follows: a) nonuniform axial noding capability, b) sodium properties calculation subprogram, c) sodium coolant heat transfer correlations, and d) most recent pressure drop correlations, such as Novendstern, Chiu-Rohsenow-Todreas and Cheng-Todreas. To assess the development status of this code, the benchmark calculations were performed with the ORNL 19 pin tests and EBR-II seven-assembly SLTHEN calculation results. The calculation results of MATRA-LMR for ORNL 19-pin assembly tests and EBR-II 91-pin experiments were compared to the measurements, and to SABRE4 and SLTHEN code calculation results, respectively. In this comparison, the differences are found among the three codes because of the pressure drop and the thermal mixing modellings. Finally, the major technical results of the conceptual design for the KALIMER 98.03 core have been compared with the calculations of MATRA-LMR, SABRE4 and SLTHEN codes. (author)

Steady-State Thermal-Hydraulics Analyses for the Conversion of the BR2 Reactor to LEU

Energy Technology Data Exchange (ETDEWEB)

Licht, J. R. [Argonne National Lab. (ANL), Argonne, IL (United States); Bergeron, A. [Argonne National Lab. (ANL), Argonne, IL (United States); Dionne, B. [Argonne National Lab. (ANL), Argonne, IL (United States); Van den Branden, G. [SCK CEN (Belgium); Kalcheva, S. [SCK CEN (Belgium); Sikik, E. [SCK CEN (Belgium); Koonen, E. [SCK CEN (Belgium)

2015-12-01

BR2 is a research reactor used for radioisotope production and materials testing. It’s a tank-in-pool type reactor cooled by light water and moderated by beryllium and light water (Figure 1). The reactor core consists of a beryllium moderator forming a matrix of 79 hexagonal prisms in a hyperboloid configuration; each having a central bore that can contain a variety of different components such as a fuel assembly, a control or regulating rod, an experimental device, or a beryllium or aluminum plug. Based on a series of tests, the BR2 operation is currently limited to a maximum allowable heat flux of 470 W/cm2 to ensure fuel plate integrity during steady-state operation and after a loss-of-flow/loss-of-pressure accident.

Long term thermo-hydro-mechanical interaction behavior study of the saturated, discontinuous granitic rock mass around the radwaste repository using a steady state flow algorithm

Energy Technology Data Exchange (ETDEWEB)

Kim, Jhin Wung; Bae, Dae Suk; Kang, Chul Hyung; Choi, Jong Won [Korea Atomic Energy Research Institute, Taejeon (Korea)

2002-02-01

The objective of the present study is to understand the long term (500 years) thermo-hydro-mechanical interaction behavior of the 500 m depth underground radwaste repository in the saturated, discontinuous granitic rock mass using a steady state flow algorithm. The numerical model includes a saturated granitic rock mass with joints around the repository and a 45 .deg. C fault passing through the tunnel roof-wall intersection, and a canister with PWR spent fuels surrounded by the compacted bentonite and mixed-bentonite. Barton-Bandis joint constitutive model from the UDEC code is used for the joints. For the hydraulic analysis, a steady state flow algorithm is used for the groundwater flow through the rock joints. For the thermal analysis, heat transfer is modeled as isotropic conduction and heat decays exponentially with time. The results show that the variations of the hydraulic aperture, hydraulic conductivity, normal stress, normal displacements, and shear displacements of the joints are high in the vicinity of the repository and stay fairly constant on the region away from the repository. 14 refs., 15 figs., 11 tabs. (Author)

Analysis of the OECD/NRC BWR Turbine Trip Transient Benchmark with the Coupled Thermal-Hydraulics and Neutronics Code TRAC-M/PARCS

International Nuclear Information System (INIS)

Lee, Deokjung; Downar, Thomas J.; Ulses, Anthony; Akdeniz, Bedirhan; Ivanov, Kostadin N.

2004-01-01

An analysis of the Peach Bottom Unit 2 Turbine Trip 2 (TT2) experiment has been performed using the U.S. Nuclear Regulatory Commission coupled thermal-hydraulics and neutronics code TRAC-M/PARCS. The objective of the analysis was to assess the performance of TRAC-M/PARCS on a BWR transient with significance in two-phase flow and spatial variations of the neutron flux. TRAC-M/PARCS results are found to be in good agreement with measured plant data for both steady-state and transient phases of the benchmark. Additional analyses of four fictitious extreme scenarios are performed to provide a basis for code-to-code comparisons and comprehensive testing of the thermal-hydraulics/neutronics coupling. The obtained results of sensitivity studies on the effect of direct moderator heating on transient simulation indicate the importance of this modeling aspect

Thermal-hydraulics analysis of a PWR reactor using zircaloy and carbide silicon reinforced with type S fibers as fuel claddings: Simulation of a channel blockage transient

Energy Technology Data Exchange (ETDEWEB)

Matuck, Vinicius; Ramos, Mario C.; Faria, Rochkhudson B.; Reis, Patricia A.L.; Costa, Antonella L.; Pereira, Claubia, E-mail: rochkdefaria@yahoo.com.br, E-mail: matuck747@gmail.com, E-mail: patricialire@yahoo.com.br, E-mail: marc5663@gmail.com, E-mail: antonella@nuclear.ufmg.br, E-mail: claubia@nuclear.ufmg.br [Universidade Federal de Minas Gerais (UFMG), Belo Horizonte (Brazil). Departamento de Engenharia Nuclear

2017-11-01

A detailed thermal-hydraulic reactor model using as reference data from the Angra 2 Final Safety Analysis Report (FSAR) has been developed and SiC reinforced with Hi-Nicalon type S fibers (SiC HNS) was used as fuel cladding. The goal is to compare its behavior from the thermal viewpoint with the Zircaloy, at the steady- state and transient conditions. The RELAP-3D was used to perform the thermal-hydraulic analysis and a blockage transient has been investigated at full power operation. The transient considered is related to total obstruction of a core cooling channel of one fuel assembly. The calculations were performed using a point kinetic model. The reactor behavior after this transient was analyzed and the time evolution of cladding and coolant temperatures mass flow and void fraction are presented. (author)

A steady state thermal duct model derived by fin-theory approach and applied on an unglazed solar collector

Energy Technology Data Exchange (ETDEWEB)

Stojanovic, B.; Hallberg, D.; Akander, J. [Building Materials Technology, KTH Research School, Centre for Built Environment, University of Gaevle, SE-801 76 Gaevle (Sweden)

2010-10-15

This paper presents the thermal modelling of an unglazed solar collector (USC) flat panel, with the aim of producing a detailed yet swift thermal steady-state model. The model is analytical, one-dimensional (1D) and derived by a fin-theory approach. It represents the thermal performance of an arbitrary duct with applied boundary conditions equal to those of a flat panel collector. The derived model is meant to be used for efficient optimisation and design of USC flat panels (or similar applications), as well as detailed thermal analysis of temperature fields and heat transfer distributions/variations at steady-state conditions; without requiring a large amount of computational power and time. Detailed surface temperatures are necessary features for durability studies of the surface coating, hence the effect of coating degradation on USC and system performance. The model accuracy and proficiency has been benchmarked against a detailed three-dimensional Finite Difference Model (3D FDM) and two simpler 1D analytical models. Results from the benchmarking test show that the fin-theory model has excellent capabilities of calculating energy performances and fluid temperature profiles, as well as detailed material temperature fields and heat transfer distributions/variations (at steady-state conditions), while still being suitable for component analysis in junction to system simulations as the model is analytical. The accuracy of the model is high in comparison to the 3D FDM (the prime benchmark), as long as the fin-theory assumption prevails (no 'or negligible' temperature gradient in the fin perpendicularly to the fin length). Comparison with the other models also shows that when the USC duct material has a high thermal conductivity, the cross-sectional material temperature adopts an isothermal state (for the assessed USC duct geometry), which makes the 1D isothermal model valid. When the USC duct material has a low thermal conductivity, the heat transfer

SAFE: A computer code for the steady-state and transient thermal analysis of LMR fuel elements

International Nuclear Information System (INIS)

Hayes, S.L.

1993-12-01

SAFE is a computer code developed for both the steady-state and transient thermal analysis of single LMR fuel elements. The code employs a two-dimensional control-volume based finite difference methodology with fully implicit time marching to calculate the temperatures throughout a fuel element and its associated coolant channel for both the steady-state and transient events. The code makes no structural calculations or predictions whatsoever. It does, however, accept as input structural parameters within the fuel such as the distributions of porosity and fuel composition, as well as heat generation, to allow a thermal analysis to be performed on a user-specified fuel structure. The code was developed with ease of use in mind. An interactive input file generator and material property correlations internal to the code are available to expedite analyses using SAFE. This report serves as a complete design description of the code as well as a user's manual. A sample calculation made with SAFE is included to highlight some of the code's features. Complete input and output files for the sample problem are provided

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

Numerical investigation of steady-state thermal behavior of an infrared detector cryo chamber

Directory of Open Access Journals (Sweden)

Singhal Mayank

2017-01-01

Full Text Available An infrared (IR detector is simply a transducer of radiant energy, converting radiant energy into a measurable form. Since radiation does not rely on visible light, it offers the possibility of seeing in the dark or through obscured conditions, by detecting the IR energy emitted by objects. One of the prime applications of IR detector systems for military use is in target acquisition and tracking of projectile systems. The IR detectors also have great potential in commercial market. Typically, IR detectors perform best when cooled to cryogenic temperatures in the range of nearly 120 K. However, the necessity to operate in such cryogenic regimes makes the application of IR detectors extremely complex. Further, prior to proceeding on to a full blown transient thermal analysis it is worthwhile to perform a steady-state numerical analysis for ascertaining the effect of variation in viz., material, gas conduction coefficient, h, emissivity, ε, on the temperature profile along the cryo chamber length. This would enable understanding the interaction between the cryo chamber and its environment. Hence, the present work focuses on the development of steady-state numerical models for thermal analysis of IR cryo chamber using MATLAB. The numerical results show that gas conduction coefficient has marked influence on the temperature profile of the cryo chamber whereas the emissivity has a weak effect. The experimental validation of numerical results has also been presented.

Efficient steady-state solver for hierarchical quantum master equations

Science.gov (United States)

Zhang, Hou-Dao; Qiao, Qin; Xu, Rui-Xue; Zheng, Xiao; Yan, YiJing

2017-07-01

Steady states play pivotal roles in many equilibrium and non-equilibrium open system studies. Their accurate evaluations call for exact theories with rigorous treatment of system-bath interactions. Therein, the hierarchical equations-of-motion (HEOM) formalism is a nonperturbative and non-Markovian quantum dissipation theory, which can faithfully describe the dissipative dynamics and nonlinear response of open systems. Nevertheless, solving the steady states of open quantum systems via HEOM is often a challenging task, due to the vast number of dynamical quantities involved. In this work, we propose a self-consistent iteration approach that quickly solves the HEOM steady states. We demonstrate its high efficiency with accurate and fast evaluations of low-temperature thermal equilibrium of a model Fenna-Matthews-Olson pigment-protein complex. Numerically exact evaluation of thermal equilibrium Rényi entropies and stationary emission line shapes is presented with detailed discussion.

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

International Nuclear Information System (INIS)

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

2002-01-01

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

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.

A field assessment of the value of steady shape hydraulic tomography for characterization of aquifer heterogeneities

Science.gov (United States)

Bohling, Geoffrey C.; Butler, James J.; Zhan, Xiaoyong; Knoll, Michael D.

2007-01-01

Hydraulic tomography is a promising approach for obtaining information on variations in hydraulic conductivity on the scale of relevance for contaminant transport investigations. This approach involves performing a series of pumping tests in a format similar to tomography. We present a field‐scale assessment of hydraulic tomography in a porous aquifer, with an emphasis on the steady shape analysis methodology. The hydraulic conductivity (K) estimates from steady shape and transient analyses of the tomographic data compare well with those from a tracer test and direct‐push permeameter tests, providing a field validation of the method. Zonations based on equal‐thickness layers and cross‐hole radar surveys are used to regularize the inverse problem. The results indicate that the radar surveys provide some useful information regarding the geometry of the K field. The steady shape analysis provides results similar to the transient analysis at a fraction of the computational burden. This study clearly demonstrates the advantages of hydraulic tomography over conventional pumping tests, which provide only large‐scale averages, and small‐scale hydraulic tests (e.g., slug tests), which cannot assess strata connectivity and may fail to sample the most important pathways or barriers to flow.

International standard problem ISP-47 on containment thermal hydraulics - Final report

International Nuclear Information System (INIS)

Allelein, H. J.; Schwarz, S.; Fischer, K.; Vendel, J.; Malet, J.; Bentaib, A.; Studer, E.; Paillere, H.; Houkema, M.

2007-01-01

The main objective of the ISP-47 is to assess the capabilities of Lumped Parameter (LP) and Computational Fluid Dynamics (CFD) codes in the area of containment thermal-hydraulics. Following the recommendations made in the state-of-the-art report on 'Containment Thermal-hydraulics and Hydrogen Distribution' this ISP was based on the application of different complementary experimental facilities and a progressive modelling difficulty. The three experimental facilities TOSQAN, MISTRA and ThAI have shown the quality of the provided experimental data suitable for CFD and LP code benchmarking in steady-state and transient conditions (control of the initial and boundary conditions, and the accuracy of the measurement techniques). This mainly includes pressure transients and gas temperature field as in former exercises. Detailed gas velocity and gas concentration (air, steam and helium) fields were obtained for the first time in such an exercise. ISP-47 was executed in two main steps: Step 1 was dedicated to the validation of the codes in the separate effects facility TOSQAN (7 m 3 ). Wall condensation, steam injection in air or air / helium atmospheres, and buoyancy were addressed under well-controlled initial conditions in a simple geometry. Furthermore, the interactions of phenomena such as condensation / stratification, turbulence / buoyancy, etc. were addressed using the larger scale of MISTRA (100 m 3 ) facility. Both TOSQAN and MISTRA were specifically designed to produce data for CFD codes with state-of-the-art instrumentation. The TOSQAN benchmark was open, whereas the MISTRA benchmark was blind. Step 2 addressed the code assessment using an experiment in the multi-compartment ThAI (60 m 3 ) facility with different steam and helium injection phases, transient stratification and mixing conditions in the atmosphere, development of natural convection, wall condensate distribution, fog formation, and transient thermal response of heat conducting walls. Detailed

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2002-07-01

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

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)

Steady-state entanglement activation in optomechanical cavities

Science.gov (United States)

Farace, Alessandro; Ciccarello, Francesco; Fazio, Rosario; Giovannetti, Vittorio

2014-02-01

Quantum discord, and related indicators, are raising a relentless interest as a novel paradigm of nonclassical correlations beyond entanglement. Here, we discover a discord-activated mechanism yielding steady-state entanglement production in a realistic continuous-variable setup. This comprises two coupled optomechanical cavities, where the optical modes (OMs) communicate through a fiber. We first use a simplified model to highlight the creation of steady-state discord between the OMs. We show next that such discord improves the level of stationary optomechanical entanglement attainable in the system, making it more robust against temperature and thermal noise.

Steady-State Thermal Properties of Rectangular Straw-Bales (RSB for Building

Directory of Open Access Journals (Sweden)

Leonardo Conti

2016-10-01

Full Text Available Straw is an inevitable product of cereal production and is available in huge quantities in the world. In order to use straw-bales as a building material, the characteristic values of the thermal performances should be determined. To not lose the benefits of the cheapness and sustainability of the material, the characteristics must be determined with simple and inexpensive means and procedures. This research aims to implement tools and methods focused at the determination of the thermal properties of straw-bales. For this study, the guidelines dictated by ASTM and ISO were followed. A measurement system consisting of a Metering Chamber (MC was realized. The MC was placed inside a Climate Chamber (CC. During the test, a known quantity of energy is introduced inside MC. When the steady-state is reached, all the energy put into MC passes through its walls in CC, where it is absorbed by the air-conditioner. A series of thermopiles detect the temperature of the surfaces of the measurement system and of the specimen. Determining the amount of energy transmitted by the various parts of MC and by the specimen, it is possible to apply Fourier’s law to calculate the thermal conductivity of the specimen.

Advances in thermal-hydraulic studies of a transmutation advanced device for sustainable energy applications

International Nuclear Information System (INIS)

Fajardo, Laura Garcia; Castells, Facundo Alberto Escriva; Lira, Carlos Brayner de Olivera

2013-01-01

The Transmutation Advanced Device for Sustainable Energy Applications (TADSEA) is a pebble-bed Accelerator Driven System (ADS) with a graphite-gas configuration, designed for nuclear waste trans- mutation and for obtaining heat at very high temperatures to produce hydrogen. In previous work, the TADSEA's nuclear core was considered as a porous medium performed with a CFD code and thermal-hydraulic studies of the nuclear core were presented. In this paper, the heat transfer from the fuel to the coolant was analyzed for three core states during normal operation. The heat transfer inside the spherical fuel elements was also studied. Three critical fuel elements groups were defined regarding their position inside the core. Results were compared with a realistic CFD model of the critical fuel elements groups. During the steady state, no critical elements reached the limit temperature of this type of fuel. (author)

IBIS, FBR 3-D Steady-State and Kinetics with Thermohydraulic Feedback

International Nuclear Information System (INIS)

Konomura, Mamoru; Tada, Nobuo; Oka, Yoshiaki; An, Shigehiro

1987-01-01

1 - Description of program or function: The IBIS code performs steady state and kinetics calculations based on a three-dimensional nuclear diffusion kinetics with thermal hydraulic feedback. It can calculate the following values in hexagonal-Z geometry of a fast breeder reactor core through the progress of transient: (1) Net reactivity; (2) Total and group-wise delayed neutron fraction; (3) Group-wise delayed neutron precursor concentration; (4) Total power and energy; (5) Space dependent neutron flux in each energy group; (6) Space dependent temperature of each material; (7) Maximum temperature of each material and its location. 2 - Method of solution: The quasi-static method is adopted to solve the three-dimensional nuclear diffusion kinetics problem. The method is the same as employed in the code QX1. The shape function equation is solved with the finite difference treatment as used in the codes CITATION and HONEYCOMB. One-dimensional thermo-hydraulics is solved with a model similar to that given in the code SASLA. Sodium boiling can be taken into account. 3 - Restrictions on the complexity of the problem: The number of neutron energy groups is fixed to 3 groups in the present version of the code

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

Energy Technology Data Exchange (ETDEWEB)

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

2013-06-15

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

Steady-State and Transient Analysis for Design Validation of SMART-ITL Secondary System

Energy Technology Data Exchange (ETDEWEB)

Yun, Eunkoo; Bae, Hwang; Ryu, Sung Uk; Jeon, Byong-Guk; Yang, Jin-Hwa; Yi, Sung-Jae; Park, Hyun-Sik [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2016-10-15

SMART can prevent large-break loss of coolant accident (LBLOCA) inherently. SMART-ITL is an experimental simulation facility designed to perform integral effect tests for the SMART plant. In terms of the secondary system of SMART-ITL, the design has been simplified from that of reference plant by replacing several components, such as expansion device and condenser, with an appropriate device to be functional as the alternatives. In this paper, in order to understand the operational characteristics as well as design concept, the secondary system of SMRAT-ITL is analyzed in steady-state and transient aspects, and the results are compared with relevant experimental results. This study focuses on the understanding of thermal-hydraulic behavior of SMART-ITL secondary system, which is simplified from that of reference plant. To identify the behaviors of the secondary system, the steady-state and transient analysis were conducted based on experimental results. In steady-state analysis, the results clearly showed that the system pressure is related to the temperature of condensation tank which varies depending on mixture enthalpy. In transient analysis, the dynamic behavior during heat-up process has been investigated. The results reveal that we can reasonably assume the fluid filled in TK-CD-01 be in a saturated condition. The results showed that the design of SMART-ITL secondary system is appropriate, and the system is being properly operated to match the design intent.

Quantum thermodynamics of nanoscale steady states far from equilibrium

Science.gov (United States)

Taniguchi, Nobuhiko

2018-04-01

We develop an exact quantum thermodynamic description for a noninteracting nanoscale steady state that couples strongly with multiple reservoirs. We demonstrate that there exists a steady-state extension of the thermodynamic function that correctly accounts for the multiterminal Landauer-Büttiker formula of quantum transport of charge, energy, or heat via the nonequilibrium thermodynamic relations. Its explicit form is obtained for a single bosonic or fermionic level in the wide-band limit, and corresponding thermodynamic forces (affinities) are identified. Nonlinear generalization of the Onsager reciprocity relations are derived. We suggest that the steady-state thermodynamic function is also capable of characterizing the heat current fluctuations of the critical transport where the thermal fluctuations dominate. Also, the suggested nonequilibrium steady-state thermodynamic relations seemingly persist for a spin-degenerate single level with local interaction.

Thermal-hydraulic code qualification: ATHOS2 and data from Bugey 4 and Tricastin 1. Final report

International Nuclear Information System (INIS)

Masiello, P.J.

1983-02-01

Measured data from steam generators at the Bugey 4 and Tricastin 1 nuclear power plants operated by Electricite de France (EdF) have been used in the qualification of the ATHOS2 computer code. ATHOS2 is a three-dimensional, two-phase thermal-hydraulic code for the steady-state and transient analysis of recirculating-type steam generators. Predicted data for circulation ratio and secondary fluid temperature just above the tube sheet have been compared with corresponding data measured by EdF during on-site testing of Westinghouse Model 51A (Bugey 4) and 51M (Tricastin 1) steam generators. Comparative analyses have been performed for steady-state operating conditions at five power levels for each plant installation. The transient capabilities of the ATHOS2 code were examined in the simulation of an open-grid (load reject from 100% power) test conducted at Bugey 4. Results show that predicted data for secondary fluid temperature at eight locations just above the tube sheet are typically within 1.5 0 C of measured data

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

International Nuclear Information System (INIS)

Qiu, Suizheng; Zhang, Dalin; Liu, Minghao; Liu, Limin; Xu, Rongshuan; Gong, Cheng; Su, Guanghui

2016-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2016-05-15

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

The relevance of thermal hydraulics pipeline simulation as a regulatory support tool

Energy Technology Data Exchange (ETDEWEB)

Silva, Patricia Mannarino; Santos, Almir Beserra dos [Agencia Nacional do Petroleo, Gas Natural e Biocombustiveis (ANP), Rio de Janeiro, RJ (Brazil)

2009-07-01

The capacity definition of a pipeline, along with its allocation, is very relevant to assure market transparency, nondiscriminatory access, security of supply, and also to give consistent signs for expansion needs. Nevertheless, the capacity definition is a controversial issue, and may widely vary depending on the technical and commercial assumptions made. To calculate a pipeline's nominal capacity, there are a variety of simulation tools, which include steady state, transient and on-line computer programs. It is desirable that the simulation tool is robust enough to predict the pipeline's capacity under different conditions. There are many variables that impact the flow through a pipeline, like gas characteristics, pipe and environmental variables. Designing a thermal model is a time-consuming task that requests understanding the level of detail need, in order to achieve success in its application. This article discusses the capacity definition, its role and calculation guidelines, describes ANP's experience with capacity calculation and further challenges according to the new regulation, and debates the role of thermal hydraulic simulation as a regulatory tool. (author)

Thermal hydraulic analysis of the JMTR improved LEU-core

Energy Technology Data Exchange (ETDEWEB)

Tabata, Toshio; Nagao, Yoshiharu; Komukai, Bunsaku; Naka, Michihiro; Fujiki, Kazuo [Japan Atomic Energy Research Inst., Oarai, Ibaraki (Japan). Oarai Research Establishment; Takeda, Takashi [Radioactive Waste Management and Nuclear Facility Decommissioning Technology Center, Tokai, Ibaraki (Japan)

2003-01-01

After the investigation of the new core arrangement for the JMTR reactor in order to enhance the fuel burn-up and consequently extend the operation period, the ''improved LEU core'' that utilized 2 additional fuel elements instead of formerly installed reflector elements, was adopted. This report describes the results of the thermal-hydraulic analysis of the improved LEU core as a part of safety analysis for the licensing. The analysis covers steady state, abnormal operational transients and accidents, which were described in the annexes of the licensing documents as design bases events. Calculation conditions for the computer codes were conservatively determined based on the neutronic analysis results and others. The results of the analysis, that revealed the safety criteria were satisfied on the fuel temperature, DNBR and primary coolant temperature, were used in the licensing. The operation license of the JMTR with the improved LEU core was granted in March 2001, and the reactor operation with new core started in November 2001 as 142nd operation cycle. (author)

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)

A phenomenological model of thermal-hydraulics of convective boiling during the quenching of hot rod bundles

International Nuclear Information System (INIS)

Unal, C.; Nelson, R.

1991-01-01

After completion of the thermal-hydraulic model developed in a companion paper, the authors performed developmental assessment calculation of the model using steady-state and transient post-critical heat flux (CHF) data. This paper discusses the results of those calculations. The overall interfacial drag model predicted reasonable drag coefficients for both the nucleate boiling and the inverted annular flow (IAF) regimes. The predicted pressure drops agreed reasonably well with the measured data of two transient experiments, CCTF Run 14 and a Lehigh reflood test. The thermal-hydraulic model for post-CHF convective heat transfer predicted the rewetting velocities reasonably well for both experiments. The predicted average slope of the wall temperature traces for these tests showed reasonable agreement with the measured data, indicating that the transient-calculated precursory cooling rates agreed with measured data. The hot-patch model, in conjunction with the other thermal-hydraulic models, was capable of modeling the Winfrith post-CHF hot-patch experiments. The hot-patch model kept the wall temperatures at the specified levels in the hot-patch regions and did not allow any quench-front propagation from either the bottom or the top of the test section. The interfacial heat-transfer model tended to slightly underpredict the vapor temperatures. The maximum difference between calculated and measured vapor temperatures was 20%, with a 10% difference for the remainder of the runs considered. The wall-to-fluid heat transfer was predicted reasonably well, and the predicted wall temperatures were in reasonable agreement with measured data with a maximum relative error of less than 13%

2-D CFD time-dependent thermal-hydraulic simulations of CANDU-6 moderator flows

Energy Technology Data Exchange (ETDEWEB)

Mehdi Zadeh, Foad [Department of Engineering Physics/Polytechnique Montréal, Montréal, QC (Canada); Étienne, Stéphane [Department of Mechanical Engineering/Polytechnique Montréal, Montréal, QC (Canada); Teyssedou, Alberto, E-mail: alberto.teyssedou@polymtl.ca [Department of Engineering Physics/Polytechnique Montréal, Montréal, QC (Canada)

2016-12-01

Highlights: • 2-D time-dependent CFD simulations of CANDU-6 moderator flows are presented. • A thermal-hydraulic code using thermal physical fluid properties is used. • The numerical approach and convergence is validated against available data. • Flow configurations are correlated using Richardson’s number. • Frequency components indicate moderator flow oscillations vs. Richardson numbers. - Abstract: The distribution of the fluid temperature and mass density of the moderator flow in CANDU-6 nuclear power reactors may affect the reactivity coefficient. For this reason, any possible moderator flow configuration and consequently the corresponding temperature distributions must be studied. In particular, the variations of the reactivity may result in major safety issues. For instance, excessive temperature excursions in the vicinity of the calandria tubes nearby local flow stagnation zones, may bring about partial boiling. Moreover, steady-state simulations have shown that for operating condition, intense buoyancy forces may be dominant, which can trigger a thermal stratification. Therefore, the numerical study of the time-dependent flow transition to such a condition, is of fundamental safety concern. Within this framework, this paper presents detailed time-dependent numerical simulations of CANDU-6 moderator flow for a wide range of flow conditions. To get a better insight of the thermal-hydraulic phenomena, the simulations were performed by covering long physical-time periods using an open-source code (Code-Saturne V3) developed by Électricité de France. The results show not only a region where the flow is characterized by coherent structures of flow fluctuations but also the existence of two limit cases where fluid oscillations disappear almost completely.

Measurement of tissue-radiation dosage using a thermal steady-state elastic shear wave.

Science.gov (United States)

Chang, Sheng-Yi; Hsieh, Tung-Sheng; Chen, Wei-Ru; Chen, Jin-Chung; Chou, Chien

2017-08-01

A biodosimeter based on thermal-induced elastic shear wave (TIESW) in silicone acellular porcine dermis (SAPD) at thermal steady state has been proposed and demonstrated. A square slab SAPD treated with ionizing radiation was tested. The SAPD becomes a continuous homogeneous and isotropic viscoelastic medium due to the generation of randomly coiled collagen fibers formed from their bundle-like structure in the dermis. A harmonic TIESW then propagates on the surface of the SAPD as measured by a nanometer-scaled strain-stress response under thermal equilibrium conditions at room temperature. TIESW oscillation frequency was noninvasively measured in real time by monitoring the transverse displacement of the TIESW on the SAPD surface. Because the elastic shear modulus is highly sensitive to absorbed doses of ionizing radiation, this proposed biodosimeter can become a highly sensitive and noninvasive method for quantitatively determining tissue-absorbed dosage in terms of TIESW’s oscillation frequency. Detection sensitivity at 1 cGy and dynamic ranges covering 1 to 40 cGy and 80 to 500 cGy were demonstrated.

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)

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)

A steady-state stomatal model of balanced leaf gas exchange, hydraulics and maximal source-sink flux.

Science.gov (United States)

Hölttä, Teemu; Lintunen, Anna; Chan, Tommy; Mäkelä, Annikki; Nikinmaa, Eero

2017-07-01

Trees must simultaneously balance their CO2 uptake rate via stomata, photosynthesis, the transport rate of sugars and rate of sugar utilization in sinks while maintaining a favourable water and carbon balance. We demonstrate using a numerical model that it is possible to understand stomatal functioning from the viewpoint of maximizing the simultaneous photosynthetic production, phloem transport and sink sugar utilization rate under the limitation that the transpiration-driven hydrostatic pressure gradient sets for those processes. A key feature in our model is that non-stomatal limitations to photosynthesis increase with decreasing leaf water potential and/or increasing leaf sugar concentration and are thus coupled to stomatal conductance. Maximizing the photosynthetic production rate using a numerical steady-state model leads to stomatal behaviour that is able to reproduce the well-known trends of stomatal behaviour in response to, e.g., light, vapour concentration difference, ambient CO2 concentration, soil water status, sink strength and xylem and phloem hydraulic conductance. We show that our results for stomatal behaviour are very similar to the solutions given by the earlier models of stomatal conductance derived solely from gas exchange considerations. Our modelling results also demonstrate how the 'marginal cost of water' in the unified stomatal conductance model and the optimal stomatal model could be related to plant structural and physiological traits, most importantly, the soil-to-leaf hydraulic conductance and soil moisture. © The Author 2017. Published by Oxford University Press. All rights reserved. For Permissions, please email: journals.permissions@oup.com.

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

Steady state ensembles of thermal radiation in a layered media with a constant heat flux

International Nuclear Information System (INIS)

Budaev, Bair V.; Bogy, David B.

2013-01-01

This paper describes steady-state ensembles of thermally excited electromagnetic radiation in nano-scale layered media with a constant non-vanishing heat flux across the layers. It is shown that Planck's law of thermal radiation, the principle of equivalence, and the laws of wave propagation in layered media, imply that in order for the ensemble of thermally excited electromagnetic fields to exist in a medium consisting of a stack of layers between two half-space, the net heat flux across the layers must exceed a certain threshold that is determined by the temperatures of the half spaces and by the reflective properties of the entire structure. The obtained results provide a way for estimating the radiative heat transfer coefficient of nano-scale layered structures. (copyright 2013 by WILEY-VCH Verlag GmbH and Co. KGaA, Weinheim)

Thermal-hydraulic and neutronic analysis of pressurized water reactor cores

International Nuclear Information System (INIS)

Alves, C.H.

1982-01-01

A computational code, named CANAL2, was developed for the simulation of the steady-state and transient behaviour of a Pressurized Water Reactor core. The conservation equations for the control volumes are obtained by area-averaging of the two-fluid model conservation equations and reducing them to the drift-flux model formulation. The resulting equations are aproximated by finite differences and solved by a marching-type numerical scheme. The model takes into account the exchange of mass, momentum and energy between adjacent subchannels of a fuel bundle. Turbulent mixing and diversion crossflow are considered. Correlations are provided for several heat trans and flow regimes and selected according to the local conditons. During transients core power can be evaluated by a point-Kinetics model. Fuel and coolant temperatures are feedback to the neutronics. The heat conduction equation is solved in the fuel using the Crank-Nicolson scheme. Temperature-dependent correlations are provided for the fuel and cladding thermal conductivities. Several runs were made with the code CANAL2 using the available experimental and calculated data in the open literature. Results indicate that CANAL2 is a good calculational tool for the thermal-hydraulics of PWR cores. A few refinements will make the code useful for design. (Author) [pt

Experiments and analyses in support of the US ALMR thermal-hydraulic design

International Nuclear Information System (INIS)

Hunsbedt, A.

1993-01-01

The U.S. Advanced Liquid Metal Reactor (ALMR) which is based on the modular PRISM concept utilizes passive safety characteristics to simplify the reactor design and enhance its safety performance. The relatively small size of each reactor facilitates the use of strong negative feedback with rising temperature for inherent reactivity control and direct, natural air cooling for decay heat removal. The tall, slender reactor geometry of the ALMR enhances uniformity and stability of internal flow distribution during steady state operation and natural circulation flow during transient conditions. The flow uniformity and low operating pressure and temperature of the reactor contributes to high structural margins. A number of experiments and associated analyses have been performed to evaluate natural convection and thermal-hydraulic phenomena experienced under decay heat removal conditions. This paper summarizes these various efforts as described separately below and presents the main results. (author)

Effect of vadose zone on the steady-state leakage rates from landfill barrier systems

International Nuclear Information System (INIS)

Celik, B.; Rowe, R.K.; Unlue, K.

2009-01-01

Leakage rates are evaluated for a landfill barrier system having a compacted clay liner (CCL) underlain by a vadose zone of variable thickness. A numerical unsaturated flow model SEEP/W is used to simulate the moisture flow regime and steady-state leakage rates for the cases of unsaturated zones with different soil types and thicknesses. The results of the simulations demonstrate that harmonic mean hydraulic conductivity of coarse textured vadose zones is 3-4 orders of magnitude less than saturated hydraulic conductivity; whereas, the difference is only one order of magnitude for fine textured vadose zones. For both coarse and fine textured vadose zones, the effective hydraulic conductivity of the barrier system and the leakage rate to an underlying aquifer increases with increasing thickness of the vadose zone and ultimately reaches an asymptotic value for a coarse textured vadose zone thickness of about 10 m and a fine textured vadose zone thickness of about 5 m. Therefore, the fine and coarse textured vadose zones thicker than about 5 m and 10 m, respectively, act as an effective part of the barrier systems examined. Although the thickness of vadose zone affects the effective hydraulic conductivity of the overall barrier system, the results demonstrated that the hydraulic conductivity of the CCL is the dominant factor controlling the steady-state leakage rates through barrier systems having single low permeability clay layers

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)

Effect of hydraulic hysteresis on the stability of infinite slopes under steady infiltration

Science.gov (United States)

Chen, Pan; Mirus, Benjamin B.; Lu, Ning; Godt, Jonathan W.

2017-01-01

Hydraulic hysteresis, including capillary soil water retention (SWR), air entrapment SWR, and hydraulic conductivity, is a common phenomenon in unsaturated soils. However, the influence of hydraulic hysteresis on suction stress, and subsequently slope stability, is generally ignored. This paper examines the influence of each of these three types of hysteresis on slope stability using an infinite slope stability analysis under steady infiltration conditions. First, hypothetical slopes for representative silty and sandy soils are examined. Then a monitored hillslope in the San Francisco Bay Area, California is assessed, using observed rainfall conditions and measured hydraulic and geotechnical properties of the colluvial soil. Results show that profiles of suction stress and the corresponding factor of safety are generally strongly affected by hydraulic hysteresis. Results suggest that each of the three types of hydraulic hysteresis may play a major role in the occurrence of slope failure, indicating that ignoring hydraulic hysteresis will likely lead to underestimates of failure potential and hence to inaccurate slope stability analysis.

COBRA-SFS [Spent Fuel Storage]: A thermal-hydraulic analysis computer code: Volume 2, User's manual

International Nuclear Information System (INIS)

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

1986-11-01

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

Transitioning from interpretive to predictive in thermal hydraulic codes

International Nuclear Information System (INIS)

Mousseau, V.A.

2004-01-01

accuracy of thermal hydraulic codes. The first step was a detailed modified equation analysis of a simplified form of the gas momentum equation. This analysis indicated the source of some of the errors in the current thermal hydraulic codes such as RELAP and TRAC. This analysis also showed the benefit of employing an implicitly balanced solution method and particularly a second order in time implicitly balanced method. Even though the analysis was highly simplified, the simulation results on the coupled system of seven equations demonstrated many of the features of the simplified analysis. In particular, the under-relaxation did not seem to consistently have any positive effect on the simulations. The under-relaxation only effected the simulation when the time step was large and the solution was of poor quality. Once the time step was made small enough to provide a quality solution, the under-relaxation had very little effect. The first order in time implicitly balanced solution, apparently because of its smaller truncation error, provided a more accurate solution than the traditional RELAP solution. The second order in time implicitly balanced solution was clearly the most accurate and deserves future study. For the more violent transient, the CFL number for the RELAP solution to obtain the same level of accuracy as the second order in time implicitly balanced was approximately one hundred times smaller. For the gentler transient this ratio goes up to about one thousand. This indicates that the second order in time implicitly balanced solution may be of great value to the long slow transients of modern passive reactor designs. The implicitly balanced method should also be considered for steady state calculations

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.

Steady-state Operational Characteristics of Ghana Research ...

African Journals Online (AJOL)

Steady state operational characteristics of the 30 kW tank-in-pool type reactor named Ghana Research Reactor-1 were investigated after a successful on-site zero power critical experiments. The steadystate operational character-istics determined were the thermal neutron fluxes, maximum period of operation at nominal ...

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.

STEADY-STATE HEAT REJECTION RATES FOR A COAXIAL BOREHOLE HEAT EXCHANGER DURING PASSIVE AND ACTIVE COOLING DETERMINED WITH THE NOVEL STEP THERMAL RESPONSE TEST METHOD

Directory of Open Access Journals (Sweden)

Marija Macenić

2018-01-01

Full Text Available At three locations in Zagreb, classical and extended thermal response test (TRT was conducted on installed coaxial heat exchangers. With classic TR test, thermogeological properties of the ground and thermal resistance of the borehole were determined at each location. It is seen that thermal conductivity of the ground varies, due to difference in geological profile of the sites. In addition, experimental research of steady-state thermal response step test (SSTRST was carried out to determine heat rejection rates for passive and active cooling in steady state regime. Results showed that heat rejection rate is only between 8-11 W/m, which indicates that coaxial system is not suitable for passive cooling demands. Furthermore, the heat pump in passive cooling mode uses additional plate heat exchanger where there is additional temperature drop of working fluid by approximately 1,5 °C. Therefore, steady-state rejection rate for passive cooling is even lower for a real case project. Coaxial heat exchanger should be always designed for an active cooling regime with an operation of a heat pump compressor in a classical vapour compression refrigeration cycle.

Assessment of the impact of neutronic/thermal-hydraulic coupling on the design and performance of nuclear reactors for space propulsion

International Nuclear Information System (INIS)

Aithal, S.M.; Aldemir, T.; Vafai, K.

1994-01-01

A series of studies has been performed to investigate the potential impact of the coupling between neutronics and thermal hydraulics on the design and performance assessment of solid core reactors for nuclear thermal space propulsion, using the particle bed reactor (PBR) concept as an example system. For a given temperature distribution in the reactor, the k eff and steady-state core power distribution are obtained from three-dimensional, continuous energy Monte Carlo simulations using the MCNP code. For a given core power distribution, determination of the temperature distribution in the core and hydrogen-filled annulus between the reflector and pressure vessel is based on a nonthermal equilibrium analysis. The results show that a realistic estimation of fuel, core size, and control requirements for PBRs using hydrogenous moderators, as well as optimization of the overall engine design, may require coupled neutronic/thermal-hydraulic studies. However, it may be possible to estimate the thermal safety margins and propellant exit temperatures based on power distributions obtained from neutronic calculations at room temperature. The results also show that, while variation of the hydrogen flow rate in the annulus has been proposed as a partial control mechanism for PBRs, such control mechanism may not be feasible for PBRs with high moderator-to-fuel ratios and hence soft core neutron spectra

Measurement of the Thermal Conductivity of Unfrozen and Frozen Food Materials by a Steady State Method with Coaxial Dual-cylinder Apparatus.

Science.gov (United States)

Pongsawatmanit, R; Miyawaki, O; Yano, T

1993-01-01

Coaxial dual-cylinder apparatus was used to measure the effective thermal conductivity of aqueous solutions of glucose, sucrose, gelatin and egg albumin over a temperature range from -20° to 20°C by the steady state method. The accuracy of the apparatus was confirmed by testing with water and ice. The effective thermal conductivity decreased with an increase in the total solid content in both the frozen and unfrozen states. In the unfrozen state, the effective thermal conductivity was slightly dependent on temperature. In the frozen state, however, the effective thermal conductivity was strongly dependent on temperature; lower temperatures gave higher effective thermal conductivity, reflecting the increase in the ice fration. For the unfrozen samples, the intrinsic thermal conductivity of each solid component was calculated by heat transfer models. All the models tested, series, parallel and Maxwell-Eucken, were equally applicable to describe the heat conduction in the unfrozen state. In the frozen state, however, the strong temperature dependency of the effective thermal conductivity suggests that the effect of the temperature dependency of the ice fraction should be incorporated into theoretical models.

Transient cases analyses of the TRIGA IPR-R1 using thermal hydraulic and neutron kinetic coupled codes

Energy Technology Data Exchange (ETDEWEB)

Reis, Patricia A.L.; Costa, Antonella L.; Pereira, Claubia; Veloso, Maria A.F.; Scari, Maria E., E-mail: patricialire@yahoo.com.br, E-mail: antonella@nuclear.ufmg.br, E-mail: claubia@nuclear.ufmg.br, E-mail: dora@nuclear.ufmg.br, E-mail: melizabethscari@yahoo.com [Universidade Federal de Minas Gerais (UFMG), Belo Horizonte, MG (Brazil). Departamento de Engenharia Nuclear; Instituto Nacional de Ciencias e Tecnologia de Reatores Nucleares Inovadores (INCT/CNPq), Belo Horizonte (Brazil); Miro, Rafael; Verdu, Gumersindo, E-mail: rmiro@iqn.upv.es, E-mail: gverdu@iqn.upv.es [Universidad Politecnica de Valencia (Spain). Departamento de Ingenieria Quimica y Nuclear

2015-07-01

Simulations and analyses of nuclear reactors have been improved by utilization of coupled thermal-hydraulic (TH) and neutron kinetics (NK) system codes especially to simulate transients that involve strong feedback effects between NK and TH. The TH-NK coupling technique was initially developed and used to simulate the behavior of power reactors; however, several coupling methodologies are now being applied for research reactors. This work presents the coupling methodology application between RELAP5 and PARCS codes using as a model the TRIGA IPR-R1 research reactor. Analyses of steady state and transient conditions and comparisons with results from simulations using only the RELAP5 code are being presented in this paper. (author)

Performance of tungsten-based materials and components under ITER and DEMO relevant steady-state thermal loads

Energy Technology Data Exchange (ETDEWEB)

Ritz, Guillaume Henri

2011-07-01

In nuclear fusion devices the surfaces directly facing the plasma are irradiated with high energy fluxes. The most intense loads are deposited on the divertor located at the bottom of the plasma chamber, which has to withstand continuous heat loads with a power density of several MW . m{sup -2} as well as transient events. These are much shorter (in the millisecond and sub-millisecond regime) but deposit a higher power densities of a few GW . m{sup -2}. The search for materials that can survive to those severe loading conditions led to the choice of tungsten which possesses advantageous attributes such as a high melting point, high thermal conductivity, low thermal expansion and an acceptable activation rate. These properties made it an attractive and promising candidate as armor material for divertors of future fusion devices such as ITER and DEMO. For the DEMO divertor, conceptual studies on helium-cooled tungsten plasma-facing components were performed. The concept was realized and tested under DEMO specific cyclic thermal loads. The examination of the plasma-facing components by microstructural analyses before and after thermal loading enabled to determine the mechanisms for components failure. Among others, it clearly showed the impact of the tungsten grade and the thermal stress induced crack formation on the performance of the armor material and in general of the plasma-facing component under high heat loads. A tungsten qualification program was launched to study the behaviour of various tungsten grades, in particular the crack formation, under fusion relevant steady-state thermal loads. In total, seven commercially available materials from two industrial suppliers were investigated. As the material's thermal response is strongly related to its microstructure, this program comprised different material geometries and manufacturing technologies. It also included the utilization of an actively cooled specimen holder which has been designed to perform

Performance of tungsten-based materials and components under ITER and DEMO relevant steady-state thermal loads

International Nuclear Information System (INIS)

Ritz, Guillaume Henri

2011-01-01

In nuclear fusion devices the surfaces directly facing the plasma are irradiated with high energy fluxes. The most intense loads are deposited on the divertor located at the bottom of the plasma chamber, which has to withstand continuous heat loads with a power density of several MW . m -2 as well as transient events. These are much shorter (in the millisecond and sub-millisecond regime) but deposit a higher power densities of a few GW . m -2 . The search for materials that can survive to those severe loading conditions led to the choice of tungsten which possesses advantageous attributes such as a high melting point, high thermal conductivity, low thermal expansion and an acceptable activation rate. These properties made it an attractive and promising candidate as armor material for divertors of future fusion devices such as ITER and DEMO. For the DEMO divertor, conceptual studies on helium-cooled tungsten plasma-facing components were performed. The concept was realized and tested under DEMO specific cyclic thermal loads. The examination of the plasma-facing components by microstructural analyses before and after thermal loading enabled to determine the mechanisms for components failure. Among others, it clearly showed the impact of the tungsten grade and the thermal stress induced crack formation on the performance of the armor material and in general of the plasma-facing component under high heat loads. A tungsten qualification program was launched to study the behaviour of various tungsten grades, in particular the crack formation, under fusion relevant steady-state thermal loads. In total, seven commercially available materials from two industrial suppliers were investigated. As the material's thermal response is strongly related to its microstructure, this program comprised different material geometries and manufacturing technologies. It also included the utilization of an actively cooled specimen holder which has been designed to perform sophisticated

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

International Nuclear Information System (INIS)

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

2017-01-01

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

Providing thermal-hydraulic boundary conditions to the reactor code TINTE through a Flownex-TINTE coupling - HTR2008-58110

International Nuclear Information System (INIS)

Marais, D.; Greyvenstein, G. P.

2008-01-01

TINTE is a well established reactor analysis code which models the transient behaviour of pebble bed reactor cores but it does not include the capabilities to model a power conversion unit (PCU). This raises the issue that TINTE cannot model full system transients. One way to overcome this problem is to supply TINTE with time-dependant thermal-hydraulic boundary conditions which are obtained from PCU simulations. This study investigates a method to provide boundary conditions for the nuclear code TINTE during full system transients. This was accomplished by creating a high level interface between the systems CFD code Flownex and TINTE. An indirect coupling method is explored whereby characteristics of the PCU are matched to characteristics of the nuclear core. This method eliminates the need to iterate between the two codes. A number of transients are simulated using the coupled code and then compared against stand-alone Flownex simulations. The coupling method introduces relatively small errors when reproducing mass flow, temperature and pressure in steady state analysis, but become more pronounced when dealing with fast thermal-hydraulic transients. Decreasing the maximum time step length of TINTE reduces this problem, but increases the computational time. Copyright ASME 2008. (authors)

Steady- and transient-state analyses of fully ceramic microencapsulated fuel loaded reactor core via two-temperature homogenized thermal-conductivity model

International Nuclear Information System (INIS)

Lee, Yoonhee; Cho, Nam Zin

2015-01-01

Highlights: • Fully ceramic microencapsulated fuel-loaded core is analyzed via a two-temperature homogenized thermal-conductivity model. • The model is compared to harmonic- and volumetric-average thermal conductivity models. • The three thermal analysis models show ∼100 pcm differences in the k eff eigenvalue. • The three thermal analysis models show more than 70 K differences in the maximum temperature. • There occur more than 3 times differences in the maximum power for a control rod ejection accident. - Abstract: Fully ceramic microencapsulated (FCM) fuel, a type of accident-tolerant fuel (ATF), consists of TRISO particles randomly dispersed in a SiC matrix. In this study, for a thermal analysis of the FCM fuel with such a high heterogeneity, a two-temperature homogenized thermal-conductivity model was applied by the authors. This model provides separate temperatures for the fuel-kernels and the SiC matrix. It also provides more realistic temperature profiles than those of harmonic- and volumetric-average thermal conductivity models, which are used for thermal analysis of a fuel element in VHTRs having a composition similar to the FCM fuel, because such models are unable to provide the fuel-kernel and graphite matrix temperatures separately. In this study, coupled with a neutron diffusion model, a FCM fuel-loaded reactor core is analyzed via a two-temperature homogenized thermal-conductivity model at steady- and transient-states. The results are compared to those from harmonic- and volumetric-average thermal conductivity models, i.e., we compare k eff eigenvalues, power distributions, and temperature profiles in the hottest single-channel at steady-state. At transient-state, we compare total powers, reactivity, and maximum temperatures in the hottest single-channel obtained by the different thermal analysis models. The different thermal analysis models and the availability of fuel-kernel temperatures in the two-temperature homogenized thermal

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

The Preliminary GAMMA Code Thermal hydraulic Analysis for the Steady State of HTR-10 Initial Core

Energy Technology Data Exchange (ETDEWEB)

Jun, Ji Su; Lim, Hong Sik; Lee, Won Jae

2006-07-15

This report describes the preliminary thermalhydraulic analysis of HTR-10 steady state full power initial core to provide a benchmark calculation of VHTGR(Very High-Temperature Gas-Cooled Reactors) safety analysis code of GAMMA(GAs Multicomponent Mixture Analysis). The input data of GAMMA code are produced for the models of fluid block, wall block, radiation heat transfer and each component material properties in HTR-10 reactor. The temperature and flow distributions of HTR-10 steady state 10 MW{sub th} full power initial core are calculated by GAMMA code with boundary conditions of total reactor inlet flow rate of 4.32 kg/s, inlet temperature of 250 .deg. C, inlet pressure of 3 MPa, outlet pressure of 2.992 MPa and the fixed temperature at RCCS water cooling tube of 50 .deg C. The calculation results are compared with the measured solid material temperatures at 22 fixed instrumentation positions in HTR-10. The wall temperature distribution in pebble bed core shows that the minimum temperature of 358 .deg. C is located at upper core, a higher temperature zone than 829 .deg. C is located at the inner region of 0.45 m radius at the bottom of core centre, and the maximum wall temperature is 897 .deg. C. The wall temperatures linearly decreases at radially and axially farther side from the bottom of core centre. The maximum temperature of RPV is 230 .deg. C, and the maximum values of fuel average temperature and TRISO centreline temperature are 907 .deg. C and 929 .deg. C, respectively and they are much lower than the fuel temperature limitation of 1230 .deg. C. The comparsion between the GAMMA code predictions and the measured temperature data shows that the calculation results are very close to the measured values in top and side reflector region, but a great difference is appeared in bottom reflector region. Some measured data are abnormally high in bottom reflector region, and so the confirmation of data is necessary in future. Fifteen of twenty two data have a

FLICA III. A digital computer program for thermal-hydraulic analysis of reactors and experimental loops

International Nuclear Information System (INIS)

Plas, Roger.

1975-05-01

This computer program describes the flow and heat transfer in steady and transient state in two-phase flows. It is the present stage of the evolution about FLICA, FLICA II and FLICA II B codes which have been used and developed at CEA for the thermal-hydraulic analysis of reactors and experimental loops with heating rod bundles. In the mathematical model all the significant terms of the fundamental hydrodynamic equations are taken into account with the approximations of turbulent viscosity and conductivity. The two-phase flow is calculated by the homogeneous model with slip. In the flow direction an implicit resolution scheme is available, which make possible to study partial or total flow blockage, with upstream and downstream effects. A special model represents the helical wire effects in out-of pile experimental rod bundles [fr

Three-dimensional coupled kinetics/thermal- hydraulic benchmark TRIGA experiments

International Nuclear Information System (INIS)

Feltus, Madeline Anne; Miller, William Scott

2000-01-01

This research project provides separate effects tests in order to benchmark neutron kinetics models coupled with thermal-hydraulic (T/H) models used in best-estimate codes such as the Nuclear Regulatory Commission's (NRC) RELAP and TRAC code series and industrial codes such as RETRAN. Before this research project was initiated, no adequate experimental data existed for reactivity initiated transients that could be used to assess coupled three-dimensional (3D) kinetics and 3D T/H codes which have been, or are being developed around the world. Using various Test Reactor Isotope General Atomic (TRIGA) reactor core configurations at the Penn State Breazeale Reactor (PSBR), it is possible to determine the level of neutronics modeling required to describe kinetics and T/H feedback interactions. This research demonstrates that the small compact PSBR TRIGA core does not necessarily behave as a point kinetics reactor, but that this TRIGA can provide actual test results for 3D kinetics code benchmark efforts. This research focused on developing in-reactor tests that exhibited 3D neutronics effects coupled with 3D T/H feedback. A variety of pulses were used to evaluate the level of kinetics modeling needed for prompt temperature feedback in the fuel. Ramps and square waves were used to evaluate the detail of modeling needed for the delayed T/H feedback of the coolant. A stepped ramp was performed to evaluate and verify the derived thermal constants for the specific PSBR TRIGA core loading pattern. As part of the analytical benchmark research, the STAR 3D kinetics code (, STAR: Space and time analysis of reactors, Version 5, Level 3, Users Guide, Yankee Atomic Electric Company, YEAC 1758, Bolton, MA) was used to model the transient experiments. The STAR models were coupled with the one-dimensional (1D) WIGL and LRA and 3D COBRA (, COBRA IIIC: A digital computer program for steady-state and transient thermal-hydraulic analysis of rod bundle nuclear fuel elements, Battelle

Thermal hydraulics of accelerator driven system windowless targets

Directory of Open Access Journals (Sweden)

Bruno ePanella

2015-07-01

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

The coupled code system DORT-TD/THERMIX and its application to the OECD/NEA/NSC PBMR400 MW coupled neutronics thermal hydraulics transient benchmark

International Nuclear Information System (INIS)

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

2009-01-01

In new reactor designs that are still under review such as the Pebble Bed Modular Reactor (PBMR), not much experimental data exists to benchmark newly developed computer codes against. Such a situation requires that nuclear engineers and designers of this novel reactor design must resort to the validation of a newly developed code through a code-to-code benchmarking exercise because there are validated codes that are currently in use to analyze this reactor design, albeit very few of them. There are numerous HTR core physics benchmarks that are currently being pursued by different organizations, for different purposes. One such benchmark exercise is the PBMR-400MW OECD/NEA coupled neutronics/thermal hydraulics transient benchmark. In this paper, a newly developed coupled neutronics thermal hydraulics code system, DORT-TD/THERMIX with both transport and diffusion theory options, is used to simulate both the steady-state as well as several transient scenarios in this benchmark problem. (orig.)

Experimental and theoretical comparison of fuel temperature and bulk coolant characteristics in the Oregon State TRIGA reactor during steady state operation

Energy Technology Data Exchange (ETDEWEB)

Marcum, W.R., E-mail: marcumw@engr.orst.ed [Oregon State University, Department of Nuclear Engineering and Radiation Health Physics, 116 Radiation Center, Corvallis, OR 97330 (United States); Woods, B.G.; Reese, S.R. [Oregon State University, Department of Nuclear Engineering and Radiation Health Physics, 116 Radiation Center, Corvallis, OR 97330 (United States)

2010-01-15

In September of 2008 Oregon State University (OSU) completed its core conversion analysis as part of the Reduced Enrichment for Research and Test Reactors (RERTR) Program. Experimental bulk coolant temperatures were collected in various locations throughout the Oregon State TRIGA Reactor (OSTR) core in order to supplement the validity of the numerical thermal hydraulic results produced in RELAP5-3D Version 2.4.2. Axial bulk coolant temperature distributions were collected by acquiring discrete thermocouple measurements in individual subchannel locations during steady state operation at 1.0 MW{sub th}. The experimental axial temperature distribution collected was compared to one-channel, two-channel, and eight-channel RELAP5-3D models and found to match within 11.94%, 11.69%, and 8.78%, respectively, on average. Comparisons to similar studies were made based on a dimensional analysis of fluid body forces in the discrete core locations, indicating that the chosen approach produces conservative results for use in the OSTR safety analysis.

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

Transient and steady-state analyses of an electrically heated Topaz-II Thermionic Fuel Element

International Nuclear Information System (INIS)

El-Genk, M.S.; Xue, H.

1992-01-01

Transient and steady-state analyses of electrically heated, Thermionic Fuel Elements (TFEs) for Topaz-II space power system are performed. The calculated emitter and collector temperatures, load electric power and conversion efficiency are in good agreement with reported data. In this paper the effects or Cs pressure, thermal power input, and load resistance on the steady-state performance of the TFE are also investigated. In addition, the thermal response of the ZrH moderator during a startup transient and following a change in the thermal power input is examined

Nonequilibrium Distribution of the Microscopic Thermal Current in Steady Thermal Transport Systems

KAUST Repository

Yukawa, Satoshi; Ogushi, Fumiko; Shimada, Takashi; Ito, Nobuyasu

2010-01-01

Nonequilibrium distribution of the microscopic thermal current is investigated by direct molecular dynamics simulations. The microscopic thermal current in this study is defined by a flow of kinetic energy carried by a single particle. Asymptotic parallel and antiparallel tails of the nonequilibrium distribution to an average thermal current are identical to ones of equilibrium distribution with different temperatures. These temperatures characterizing the tails are dependent on a characteristic length in which a memory of dynamics is completely erased by several particle collisions. This property of the tails of nonequilibrium distribution is confirmed in other thermal transport systems. In addition, statistical properties of a particle trapped by a harmonic potential in a steady thermal conducting state are also studied. This particle feels a finite force parallel to the average thermal current as a consequence of the skewness of the distribution of the current. This force is interpreted as the microscopic origin of thermophoresis.

Uncertainty and Sensitivity Analysis Applied to the Validation of BWR Bundle Thermal-Hydraulic Calculations

International Nuclear Information System (INIS)

Hernandez-Solis, Augusto

2010-04-01

This work has two main objectives. The first one is to enhance the validation process of the thermal-hydraulic features of the Westinghouse code POLCA-T. This is achieved by computing a quantitative validation limit based on statistical uncertainty analysis. This validation theory is applied to some of the benchmark cases of the following macroscopic BFBT exercises: 1) Single and two phase bundle pressure drops, 2) Steady-state cross-sectional averaged void fraction, 3) Transient cross-sectional averaged void fraction and 4) Steady-state critical power tests. Sensitivity analysis is also performed to identify the most important uncertain parameters for each exercise. The second objective consists in showing the clear advantages of using the quasi-random Latin Hypercube Sampling (LHS) strategy over simple random sampling (SRS). LHS allows a much better coverage of the input uncertainties than SRS because it densely stratifies across the range of each input probability distribution. The aim here is to compare both uncertainty analyses on the BWR assembly void axial profile prediction in steady-state, and on the transient void fraction prediction at a certain axial level coming from a simulated re-circulation pump trip scenario. It is shown that the replicated void fraction mean (either in steady-state or transient conditions) has less variability when using LHS than SRS for the same number of calculations (i.e. same input space sample size) even if the resulting void fraction axial profiles are non-monotonic. It is also shown that the void fraction uncertainty limits achieved with SRS by running 458 calculations (sample size required to cover 95% of 8 uncertain input parameters with a 95% confidence), result in the same uncertainty limits achieved by LHS with only 100 calculations. These are thus clear indications on the advantages of using LHS. Finally, the present study contributes to a realistic analysis of nuclear reactors, in the sense that the uncertainties of

ZZ-PBMR-400, OECD/NEA PBMR Coupled Neutronics/Thermal Hydraulics Transient Benchmark - The PBMR-400 Core Design

International Nuclear Information System (INIS)

Reitsma, Frederik

2007-01-01

efficiency: ≥ 41%; Emergency planning zone: 400 meters. The PBMR functions under a direct Brayton cycle with primary coolant helium flowing downward through the core and exiting at 900 C. The helium then enters the turbine relinquishing energy to drive the electric generator and compressors. After leaving the turbine, the helium then passes consecutively through the LP primary side of the recuperator, then the pre-cooler, the low pressure compressor, inter-cooler, high pressure compressor and then on to the HP secondary side of the recuperator before re-entering the reactor vessel at 500 C. Power is adjusted by regulating the mass flow rate of gas inside the primary circuit. This is achieved by a combination of compressor bypass and system pressure changes. Increasing the pressure results in an increase in mass flow rate, which results in an increase in the power removed from the core. Power reduction is achieved by removing gas from the circuit. A Helium Inventory Control System is used to provide an increase or decrease in system pressure. The benchmark is divided into Phases and Exercises as follows: Phase I: Steady State Benchmark Calculational Cases; Exercise 1: Neutronics Solution with Fixed Cross Sections ; Exercise 2: Thermal Hydraulic solution with given power / heat sources; Exercise 3: Combined neutronics thermal hydraulics calculation - starting condition for the transients. Phase II: Transient benchmark: Exercise 1: De-pressurised Loss of Forced Cooling (DLOFC) without SCRAM; Exercise 2 : De-pressurised Loss of Forced Cooling (DLOFC) with SCRAM; Exercise 3: Pressurised Loss of Forced Cooling (PLOFC) with SCRAM; Exercise 4 : 100-40-100 Load Follow; Exercise 5: Fast Reactivity Insertion - Control Rod Withdrawal (CRW) and Control Rod Ejection (CRE) scenarios at hot full power conditions; Exercise 6 : Cold Helium Inlet

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

Directory of Open Access Journals (Sweden)

Sidi Ali Kamel

2012-01-01

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

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)

COBRA-SFS [Spent Fuel Storage]: A thermal-hydraulic analysis computer code: Volume 1, Mathematical models and solution method

International Nuclear Information System (INIS)

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

1986-11-01

COBRA-SFS (Spent Fuel Storage) is a general thermal-hydraulic analysis computer code used to predict temperatures and velocities in a wide variety of systems. The code was refined and specialized for spent fuel storage system analyses for the US Department of Energy's Commercial Spent Fuel Management Program. The finite-volume equations governing mass, momentum, and energy conservation are written for an incompressible, single-phase fluid. The flow equations model a wide range of conditions including natural circulation. The energy equations include the effects of solid and fluid conduction, natural convection, and thermal radiation. The COBRA-SFS code is structured to perform both steady-state and transient calculations: however, the transient capability has not yet been validated. This volume describes the finite-volume equations and the method used to solve these equations. It is directed toward the user who is interested in gaining a more complete understanding of these methods

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.)

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 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)

Ignition phase and steady-state structures of a non-thermal air plasma

CERN Document Server

Lu Xin Pei

2003-01-01

An AC-driven, non-thermal, atmospheric pressure air plasma is generated within the gap separating a disc-shaped metal electrode and a water electrode. The ignition phase and the steady-state are studied by a high-speed CCD camera. It is found that the plasma always initiates at the surface of the water electrode. The plasma exhibits different structures depending on the polarity of the water electrode: when the water electrode plays the role of cathode, a relatively wide but visibly dim plasma column is generated. At the maximum driving voltage, the gas temperature is between 800 and 900 K, and the peak current is 67 mA; when the water electrode is anode, the plasma column narrows but increases its light emission. The gas temperature in this case is measured to be in the 1400-1500 K range, and the peak current is 81 mA.

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

International Nuclear Information System (INIS)

Basu, Dipankar N.; Bhattacharyya, Souvik; Das, P.K.

2014-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2014-12-15

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

Thermal hydraulic feasibility assessment of the spent nuclear fuel project

International Nuclear Information System (INIS)

Heard, F.J.

1996-01-01

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

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)

Analysis of physical properties controlling steady-state infiltration rates on tropical savannah soils

International Nuclear Information System (INIS)

Mbagwu, J.S.C.

1993-10-01

A knowledge of physical properties influencing the steady-state infiltration rates (ic) of soils is needed for the hydrologic modelling of the infiltration process. In this study evidence is provided to show that effective porosity (Pe) (i.e. the proportion of macro pore spaces with equivalent radius of > 15 μm) and dry bulk density are the most important soil physical properties controlling the steady-state infiltration rates on a tropical savannah with varying land use histories. At a macro porosity value of ≤ 5.0% the steady-state infiltration rate is zero. Total porosity and the proportion of water-retaining pores explained only a small fraction of the variation in this property. Steady-state infiltration rates can also be estimated from either the saturated hydraulic conductivity (Ks) by the equation, i c = 31.1 + 1.06 (Ks), (R 2 = 0.8104, p ≤ 0.001) or the soil water transmissivity (A) by the equation, i c = 30.0 + 29.9(A), (R 2 = 0.8228, ρ ≤ 0.001). The Philip two-parameter model under predicted steady-state infiltration rates generally. Considering the ease of determination and reliability it is suggested that effective porosity be used to estimate the steady-state infiltration rates of these other soils with similar characteristics. The model is, i c 388.7(Pe) - 10.8(R 2 = 0.7265, p ≤ 0.001) where i c is in (cm/hr) and Pe in (cm 3 /cm 3 ). (author). 20 refs, 3 figs, 4 tabs

TRIO-EF a general thermal hydraulics computer code applied to the Avlis process

International Nuclear Information System (INIS)

Magnaud, J.P.; Claveau, M.; Coulon, N.; Yala, P.; Guilbaud, D.; Mejane, A.

1993-01-01

TRIO(EF is a general purpose Fluid Mechanics 3D Finite Element Code. The system capabilities cover areas such as steady state or transient, laminar or turbulent, isothermal or temperature dependent fluid flows; it is applicable to the study of coupled thermo-fluid problems involving heat conduction and possibly radiative heat transfer. It has been used to study the thermal behaviour of the AVLIS process separation module. In this process, a linear electron beam impinges the free surface of a uranium ingot, generating a two dimensional curtain emission of vapour from a water-cooled crucible. The energy transferred to the metal causes its partial melting, forming a pool where strong convective motion increases heat transfer towards the crucible. In the upper part of the Separation Module, the internal structures are devoted to two main functions: vapor containment and reflux, irradiation and physical separation. They are subjected to very high temperature levels and heat transfer occurs mainly by radiation. Moreover, special attention has to be paid to electron backscattering. These two major points have been simulated numerically with TRIO-EF and the paper presents and comments the results of such a computation, for each of them. After a brief overview of the computer code, two examples of the TRIO-EF capabilities are given: a crucible thermal hydraulics model, a thermal analysis of the internal structures

Review of thermal-hydraulic calculations for Calvert Cliffs and H.B. Robinson PTS study

International Nuclear Information System (INIS)

Jo, J.H.; Yuelys-Miksis, C.; Rohatgi, U.S.

1984-01-01

Thermal-hydraulic transient calculations performed by LANL using the TRAC-PF1 code and by INEL using the RELAP5 code for the USNRC pressurized thermal shock (PTS) study of the Calvert Cliffs and H.B. Robinson Nuclear Power Plants have been reviewed at BNL including the input decks and steady state calculations. Furthermore, six transients for each plant have been selected for the in-depth review. Simple hand calculations based on the mass and energy balances of the entire reactor system, have been performed to predict the temperature and pressure of the reactor system, and the results have been compared with those obtained by the code calculation. In general, the temperatures and pressures of the primary system calculated by the codes have been very reasonable. The secondary pressures calculated by TRAC appear to indicate that the codes have some difficulty with the condensation model and further work is needed to assess the code calculation of the U-tube steam generator pressure when the cold auxiliary feedwater is introduced to the steam generator. However, it is not expected that this uncertainty would affect the transient calculations significantly

Thermal Hydraulic Integral Effect Tests for Pressurized Water Reactors

Energy Technology Data Exchange (ETDEWEB)

Baek, W. P.; Song, C. H.; Kim, Y. S. and others

2005-02-15

The objectives of the project are to construct a thermal-hydraulic integral effect test facility and to perform various integral effect tests for design, operation, and safety regulation of pressurized water reactors. During the first phase of this project (1997.8{approx}2002.3), the basic technology for thermal-hydraulic integral effect tests was established and the basic design of the test facility was accomplished: a full-height, 1/300-volume-scaled full pressure facility for APR1400, an evolutionary pressurized water reactor that was developed by Korean industry. Main objectives of the present phase (2002.4{approx}2005.2), was to optimize the facility design and to construct the experimental facility. We have performed following researches: 1) Optimization of the basic design of the thermal-hydraulic integral effect test facility for PWRs - ATLAS (Advanced Thermal-hydraulic Test Loop for Accident Simulation) - Reduced height design for APR1400 (+ specific design features of KSNP safety injection systems) - Thermal-hydraulic scaling based on three-level scaling methodology by Ishii et al. 2) Construction of the ATLAS facility - Detailed design of the test facility - Manufacturing and procurement of components - Installation of the facility 3) Development of supporting technology for integral effect tests - Development and application of advanced instrumentation technology - Preliminary analysis of test scenarios - Development of experimental procedures - Establishment and implementation of QA system/procedure.

Advances of study on thermal-hydraulic performance in tight-lattice rod bundles for reduced-moderation water reactors

International Nuclear Information System (INIS)

Akira Ohnuki; Kazuyuki Takase; Masatoshi Kureta; Hiroyuki Yoshida; Hidesada Tamai; Wei Liu; Toru Nakatsuka; Hajime Akimoto

2005-01-01

R and D project to investigate thermal-hydraulic performance in tight-lattice rod bundles for Reduced-Moderation Water Reactor (RMWR) is started at Japan Atomic Energy Research Institute in collaboration with power company, reactor vendors, universities since 2002. The RMWR can attain the favorable characteristics such as effective utilization of uranium resources, multiple recycling of plutonium, high burn-up and long operation cycle, based on matured LWR technologies. MOX fuel assemblies with tight lattice arrangement are used to increase the conversion ratio by reducing the moderation of neutron. Increasing the in-core void fraction also contributes to the reduction of neutron moderation. The confirmation of thermal-hydraulic feasibility is one of the most important R and D items for the RMWR because of the tight-lattice configuration. In this paper, we will show the R and D plan and describe some advances on experimental and analytical studies. The experimental study is performed mainly using large-scale (37-rod bundle) test facility and the analytical one aims to develop a predictable technology for geometry effects such as gap between rods, grid spacer configuration etc. using advanced 3-D two-phase flow simulation methods. Steady-state and transient critical power experiments are conducted with the test facility (Gap width between rods: 1.0 mm) and the experimental data reveal the feasibility of RMWR. (authors)

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.

The steady-state tokamak program

International Nuclear Information System (INIS)

Politzer, D.A.; Nevins, W.M.

1992-01-01

This paper reports on a steady-state tokamak experiment (STE) needed to develop the technology and physics data base required for construction of a steady-state fusion power demonstration reactor in the early 21st century. The STE will provide an integrated facility for the development and demonstration of steady-state and particle handling, low-activation high-heat-flux components and materials, efficient current drive, and continuous plasma performance in steady-state, with reactor-like plasma conditions under severe conditions of heat and particle bombardment of the wall. The STE facility will also be used to develop operation and control scenarios for ITER

Thermal Hydraulic Assessment for Loss of SDCS Event During the Outage of CANDU Reactor

Energy Technology Data Exchange (ETDEWEB)

Kim, Jonghyun [Gnest, Inc. Taejon (Korea, Republic of); Lee, Kwangho; Oh, Haechol; Jun, Hwangyong [KEPRI, Taejon (Korea, Republic of)

2006-07-01

During the outage(overhaul) of the nuclear power plant, there are several operating states other than the full power state, that is 'Hot-Zero Power', 'Depressurized-Cooldown', and 'Partially Drained'. Until now safety assessment has not been done much for this operating state of CANDU type reactor worldwide. For the accuracy and confidence of PSA for the CANDU outage, the safety analysis is necessary. At the first stage, we analyzed the thermal hydraulic characteristics and safety of the postulated event of loss of shutdown cooling system (SDCS) during the partially drained state which is the longest one in the middle of outage period. As an analysis tool, this study uses the best estimate thermal hydraulic code, RELAP5/CANDU which was modified according to the CANDU specific characteristics and based on RELAP5.Mod3.

Steady state and LOCA analysis of Kartini reactor using RELAP5/SCDAP code: The role of passive system

Science.gov (United States)

Antariksawan, Anhar R.; Wahyono, Puradwi I.; Taxwim

2018-02-01

Safety is the priority for nuclear installations, including research reactors. On the other hand, many studies have been done to validate the applicability of nuclear power plant based best estimate computer codes to the research reactor. This study aims to assess the applicability of the RELAP5/SCDAP code to Kartini research reactor. The model development, steady state and transient due to LOCA calculations have been conducted by using RELAP5/SCDAP. The calculation results are compared with available measurements data from Kartini research reactor. The results show that the RELAP5/SCDAP model steady state calculation agrees quite well with the available measurement data. While, in the case of LOCA transient simulations, the model could result in reasonable physical phenomena during the transient showing the characteristics and performances of the reactor against the LOCA transient. The role of siphon breaker hole and natural circulation in the reactor tank as passive system was important to keep reactor in safe condition. It concludes that the RELAP/SCDAP could be use as one of the tool to analyse the thermal-hydraulic safety of Kartini reactor. However, further assessment to improve the model is still needed.

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

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)

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.

Thermal and Hydraulic Conditions Supporting the Recruitment of Asian Carp in Seiche Affected Rivers of Lake Erie Basin - A Case Study of the Lower Sandusky River in Ohio

Science.gov (United States)

Soong, D. T.; Santacruz, S.; Jones, L.; Garcia, T.; Kočovský, P. M.; Embke, H.

2017-12-01

Grass Carp Ctenopharyngodon idella (Cyprinidae) is an invasive fish species that spawns in rivers during high-flow events. In their native range, it is believed eggs must hatch within the riverine environment in order to eventually result in production of adult fish. The lower Sandusky River is approximately 26 km long extending from its confluence with Sandusky Bay upstream to the Ballville Dam, which is impassible for Grass Carp. Grass Carp are known to have spawned in the Sandusky River, a tributary to Lake Erie, in 2011, 2013, 2015, and 2017. This study characterizes the thermal and hydraulic conditions under which these eggs could hatch in the lower Sandusky River, a relatively short river reach for egg hatching. Grass Carp eggs collected in 2015 were previously analyzed for hatching locations using a one-dimensional steady-state HEC-RAS hydraulic model. In this study we refine estimates of hatching locations by incorporating the influence of fluctuating water levels downstream due to seiches in Lake Erie and overland and tributary inflows using an unsteady 1D/2D HEC-RAS hydraulic model. Additionally, conditions conducive to successful hatching, which occurs when eggs reach the hatching stage within the river, were analyzed from nine high-flow events between 2011 and 2015. Simulated hydraulic and water temperature data were used as inputs to the Fluvial Egg Drift Simulator (FluEgg) model, which was used to analyze the transport and dispersal of Grass carp eggs until hatching. We will describe the differences in steady- and unsteady-state hydraulic modeling in predicting hatching locations of Grass Carp eggs for the 2015 spawning events. Results will also include hydraulic and temperature variables that contribute to the successful/unsuccessful in-river hatching for the nine flow events simulated.

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.

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

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)

LAVENDER: A steady-state core analysis code for design studies of accelerator driven subcritical reactors

Energy Technology Data Exchange (ETDEWEB)

Zhou, Shengcheng; Wu, Hongchun; Cao, Liangzhi; Zheng, Youqi, E-mail: yqzheng@mail.xjtu.edu.cn; Huang, Kai; He, Mingtao; Li, Xunzhao

2014-10-15

Highlights: • A new code system for design studies of accelerator driven subcritical reactors (ADSRs) is developed. • S{sub N} transport solver in triangular-z meshes, fine deletion analysis and multi-channel thermal-hydraulics analysis are coupled in the code. • Numerical results indicate that the code is reliable and efficient for design studies of ADSRs. - Abstract: Accelerator driven subcritical reactors (ADSRs) have been proposed and widely investigated for the transmutation of transuranics (TRUs). ADSRs have several special characteristics, such as the subcritical core driven by spallation neutrons, anisotropic neutron flux distribution and complex geometry etc. These bring up requirements for development or extension of analysis codes to perform design studies. A code system named LAVENDER has been developed in this paper. It couples the modules for spallation target simulation and subcritical core analysis. The neutron transport-depletion calculation scheme is used based on the homogenized cross section from assembly calculations. A three-dimensional S{sub N} nodal transport code based on triangular-z meshes is employed and a multi-channel thermal-hydraulics analysis model is integrated. In the depletion calculation, the evolution of isotopic composition in the core is evaluated using the transmutation trajectory analysis algorithm (TTA) and fine depletion chains. The new code is verified by several benchmarks and code-to-code comparisons. Numerical results indicate that LAVENDER is reliable and efficient to be applied for the steady-state analysis and reactor core design of ADSRs.

Thermal Hydraulic Integral Effect Tests for Pressurized Water Reactors

International Nuclear Information System (INIS)

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

2007-02-01

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

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

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

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

International Nuclear Information System (INIS)

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

2009-01-01

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

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.

Thermal Margin Calculation of the CAREM-25 Core

International Nuclear Information System (INIS)

Mazufri, C.M

2000-01-01

During the operation in steady state and anticipated operational transient of a nuclear reactor it is necessary to avoid the damage in the fuel elements induced by thermal or hydraulic effects.To satisfy that design bases safety limits are established and calculation methodologies are defined to verify them.In the particular case of the reactor CAREM-25 reactor where the core is cooled by natural circulation, it is not adequate to use directly the same calculation methodologies from typical PWR and BWR.The low cooling flow rate and not having channels in the fuel elements (open-channel fuels) produce that most of the models and computer programs typically used must be carefully validated.As result of that process, an adequate calculation procedure for this reactor type was developed.In the present work, the thermal-hydraulic design criteria of the core and the design bases, the uncertainties factors, and the thermal margin results of the core are described.Despite that the methodology of DNBR calculation is under a validation process and considering the available calculation tools, it is possible to assure that the core fulfills the safety regulations in steady state conditions

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

International Nuclear Information System (INIS)

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

2013-01-01

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

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

Experimental Observations of Natural Circulation Flow in the NSTF at Steady-State Conditions

International Nuclear Information System (INIS)

Lisowski, Darius D.; Farmer, Mitch T.

2014-01-01

A ½ scale test facility has been constructed at Argonne National Laboratory (ANL) to study the heat removal performance and natural circulation flow patterns in a Reactor Cavity Cooling System (RCCS). Our test facility, the Natural convection Shutdown heat removal Test Facility (NSTF), supports the broader goal of developing an inherently safe and fully passive ex-vessel decay heat removal for advanced reactor designs. The project, initiated in 2010 to support the Advanced Reactor Concepts (ARC), Small Modular Reactor (SMR), and Next Generation Nuclear Plant (NGNP) programs, has been conducting experimental operations since early 2014. The following paper provides a summary of some primary design features of the 26-m tall test facility along with a description of the data acquisition suite that guides our experimental practices. Specifics of the distributed fiber optic temperature measurements will be discussed, which introduces an unparalleled level of data density that has never before been implemented in a large scale natural circulation test facility. Results from our first test series will then be presented, which provide insight into the thermal hydraulic behavior at steady-state conditions for varying heat flux levels and exhaust chimney configuration states. (author)

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.

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

International Nuclear Information System (INIS)

Lemonnier, H.

2005-01-01

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

Computational models for thermal-hydraulic assessment of TADSEA and its use for hydrogen production

International Nuclear Information System (INIS)

Rojas, L.; Gonzalez, D.; Garcia, C.; Gamez, A.; Garcia, L.; Lira, C. A. B. O.

2015-01-01

The Transmutation Advanced Device for Sustainable Energy Applications (TADSEA) is a pebble-bed Accelerator Driven System (ADS) with a graphite-gas configuration, designed for nuclear waste transmutation and for obtaining heat at very high temperatures to produce hydrogen. In previous work, the TADSEA's nuclear core was considered as a porous medium performed with a CFD code and thermal-hydraulic studies of the nuclear core were presented. In this paper, three critical fuel elements groups were defined regarding their position inside the core. In this article, the heat transfer from the fuel to the coolant was analyzed for the three core states during normal operation. The heat transfer inside the spherical fuel elements was also studied with a realistic CFD model of the critical elements groups. During the steady state, no critical elements reached the limit temperature of this type of fuel. Also, it is presented a model built in ANSYS for the simulation and optimization of high- temperature electrolysis using the TADSEA as a heat source. A flow diagram of the electrolysis process with the high temperature electrolyzer as the main component using TADSEA as an energy source is finally proposed and discussed. (Author)

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)

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

Science.gov (United States)

Domalapally, Phani; Di Caro, Marco

2018-05-01

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

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)

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.

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

Preparation of a thermal-hydraulic design method for driver core fuel pins of a new in-pile experimental reactor for FBR safety research

International Nuclear Information System (INIS)

Mizuno, Masahiro; Yamaguchi, Katsuhisa; Uto, Nariaki

1999-07-01

A design study of a new in-pile experimental reactor, SERAPH (Safety Engineering Reactor for Accident PHenomenology), for FBR safety research has progressed at JNC (Japan Nuclear Cycle Development Institute). SERAPH is intended for various in-pile experiments to be performed under quasi-steady state and various transient operation modes. In order to evaluate the driver core performance in conducting such experiments, clarify the relating design issues to be resolved and refine the experimental needs, it is indispensable to comprehend the allowable margin for the thermal-hydraulic fuel pin design since it largely affects the strategy for the driver core design. This report presents a thermal-hydraulic design method for the driver core fuel pins, which is a combination of a two-dimensional time-dependent heat transfer analysis code TAC-2D and a general non-linear finite-element structural analysis code FINAS. In TAC-2D, the allowable spatial mesh and the time step sizes are evaluated. The code is modified so as to treat time-dependent thermal properties, include an improved gap heat-transfer model and treat the change of intra-pin gap width under transient modes, for the purpose of improving the accuracy of evaluating heat transfer characteristics which gives a significant impact on the thermal-hydraulic design. As for FINAS, the number of element nodes and spatial meshes required to obtain adequate accuracy for the thermal stress characteristics of a fuel pellet during transient modes are investigated. In addition, post-processing tools are newly developed to process the calculation results obtained from these codes. The results of this work contribute to advancing the fuel pin design study for SERAPH as well with the investigation on the technique of manufacturing fuel pins. (author)

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

1-D thermal-hydraulic analysis of the high temperature superconducting current leads for the ITER magnet system from 5 K to 300 K

Energy Technology Data Exchange (ETDEWEB)

Rizzo, Enrico, E-mail: enrico.rizzo@kit.edu [Institute for Technical Physics, Karlsruhe Institute of Technology, 76344 Eggenstein-Leopoldshafen (Germany); Bauer, Pierre [ITER Organization, Cadarache (France); Heller, Reinhard [Institute for Technical Physics, Karlsruhe Institute of Technology, 76344 Eggenstein-Leopoldshafen (Germany); Richard, Laura Savoldi; Zanino, Roberto [Dipartimento Energia, Politecnico di Torino, 10129 Torino (Italy)

2013-12-15

Highlights: • A global, predictive picture of the ITER HTS current lead is not yet available. • A predictive 1-D, steady state thermal-hydraulic analysis of the full length HTS current leads has been performed. • For the heat exchanger, correlations previously derived by the same authors have been used. • The results have been compared with the ITER relevant requirements. • According to our results, the ITER HTS current leads will fulfill the requirements. -- Abstract: The magnet system of ITER includes high temperature superconducting (HTS) current leads with a maximum current of 68 kA for the toroidal field (TF) coils, 55 kA for the poloidal field (PF)/central solenoid (CS) coils and 10 kA for the control coils (CC), respectively. Although different in terms of size and operative conditions, the ITER HTS current leads have been all designed on the basis of an established concept, which was successfully developed for the LHC at CERN and proven by the so-called 70 kA “demonstrator” lead made by KIT and by the ITER pre-prototypes made by ASIPP in China. A broad R and D campaign has been undertaken by ASIPP and CERN in order to find optimized designs for each component of the leads. Nevertheless, a comprehensive picture of the performance of the entire HTS current leads is not yet available. In this paper, a steady state, full length, thermal-hydraulic 1-D modeling is applied to the study of the three types (TF, PF/CS, CC) of ITER HTS current leads. The results of this predictive analysis are then compared with relevant ITER requirements. It was found that the present design of the HTS current leads will fulfill these specifications.

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.

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

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)

X-Ray Spectral Analysis of the Steady States of GRS1915+105

Science.gov (United States)

Peris, Charith S.; Remillard, Ronald A.; Steiner, James F.; Vrtilek, Saeqa D.; Varnière, Peggy; Rodriguez, Jerome; Pooley, Guy

2016-05-01

We report on the X-ray spectral behavior within the steady states of GRS1915+105. Our work is based on the full data set of the source obtained using the Proportional Counter Array (PCA) on the Rossi X-ray Timing Explorer (RXTE) and 15 GHz radio data obtained using the Ryle Telescope. The steady observations within the X-ray data set naturally separated into two regions in the color-color diagram and we refer to these regions as steady-soft and steady-hard. GRS1915+105 displays significant curvature in the coronal component in both the soft and hard data within the RXTE/PCA bandpass. A majority of the steady-soft observations displays a roughly constant inner disk radius ({R}{{in}}), while the steady-hard observations display an evolving disk truncation which is correlated to the mass accretion rate through the disk. The disk flux and coronal flux are strongly correlated in steady-hard observations and very weakly correlated in the steady-soft observations. Within the steady-hard observations, we observe two particular circumstances when there are correlations between the coronal X-ray flux and the radio flux with log slopes η ˜ 0.68+/- 0.35 and η ˜ 1.12+/- 0.13. They are consistent with the upper and lower tracks of Gallo et al. (2012), respectively. A comparison of the model parameters to the state definitions shows that almost all of the steady-soft observations match the criteria of either a thermal or steep power-law state, while a large portion of the steady-hard observations match the hard-state criteria when the disk fraction constraint is neglected.

A simplified method for evaluating thermal performance of unglazed transpired solar collectors under steady state

International Nuclear Information System (INIS)

Wang, Xiaoliang; Lei, Bo; Bi, Haiquan; Yu, Tao

2017-01-01

Highlights: • A simplified method for evaluating thermal performance of UTC is developed. • Experiments, numerical simulations, dimensional analysis and data fitting are used. • The correlation of absorber plate temperature for UTC is established. • The empirical correlation of heat exchange effectiveness for UTC is proposed. - Abstract: Due to the advantages of low investment and high energy efficiency, unglazed transpired solar collectors (UTC) have been widely used for heating in buildings. However, it is difficult for designers to quickly evaluate the thermal performance of UTC based on the conventional methods such as experiments and numerical simulations. Therefore, a simple and fast method to determine the thermal performance of UTC is indispensable. The objective of this work is to provide a simplified calculation method to easily evaluate the thermal performance of UTC under steady state. Different parameters are considered in the simplified method, including pitch, perforation diameter, solar radiation, solar absorptivity, approach velocity, ambient air temperature, absorber plate temperature, and so on. Based on existing design parameters and operating conditions, correlations for the absorber plate temperature and the heat exchange effectiveness are developed using dimensional analysis and data fitting, respectively. Results show that the proposed simplified method has a high accuracy and can be employed to evaluate the collector efficiency, the heat exchange effectiveness and the air temperature rise. The proposed method in this paper is beneficial to directly determine design parameters and operating status for UTC.

The OECD/NEA/NSC PBMR coupled neutronics/thermal hydraulics transient benchmark: The PBMR-400 core design

International Nuclear Information System (INIS)

Reitsma, F.; Ivanov, K.; Downar, T.; De Haas, H.; Gougar, H. D.

2006-01-01

The Pebble Bed Modular Reactor (PBMR) is a High-Temperature Gas-cooled Reactor (HTGR) concept to be built in South Africa. As part of the verification and validation program the definition and execution of code-to-code benchmark exercises are important. The Nuclear Energy Agency (NEA) of the Organisation for Economic Cooperation and Development (OECD) has accepted, through the Nuclear Science Committee (NSC), the inclusion of the Pebble-Bed Modular Reactor (PBMR) coupled neutronics/thermal hydraulics transient benchmark problem in its program. The OECD benchmark defines steady-state and transients cases, including reactivity insertion transients. It makes use of a common set of cross sections (to eliminate uncertainties between different codes) and includes specific simplifications to the design to limit the need for participants to introduce approximations in their models. In this paper the detailed specification is explained, including the test cases to be calculated and the results required from participants. (authors)

Neutronic and thermal-hydraulic coupling using Milonga and OpenFOAM codes: an approach using free software

International Nuclear Information System (INIS)

Silva, Vitor Vasconcelos Araújo

2016-01-01

The development of a fine mesh coupled neutronics/thermal-hydraulics framework mainly using open source software is presented. The contributions proposed go in two different directions: one, is the focus on the open software development, a concept widely spread in many fields of knowledge but rarely explored in the nuclear engineering field; the second, is the use of operating system shared memory as a fast and reliable storage area to couple the computational fluid dynamics (CFD) software OpenFOAM to the free and flexible reactor core analysis code Milonga. This concept was applied to simulate the behavior of the TRIGA Mark 1 IPR-R1 reactor fuel pin in steady-state mode. The macroscopic cross-sections for the model, a set of two-group cross-sections data, were generated using WIMSD-5B code. The results show that this innovative coupled system gives consistent results, encouraging system further development and its use for complex nuclear systems. (author)

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)

Experimental research in neutron physic and thermal-hydraulic at the CDTN Triga reactor

Energy Technology Data Exchange (ETDEWEB)

Mesquita, Amir Z.; Souza, Rose Mary G.P.; Ferreira, Andrea V.; Pinto, Antonio J.; Costa, Antonio C.L.; Rezende, Hugo C., E-mail: amir@cdtn.b, E-mail: souzarm@cdtn.b, E-mail: avf@cdtn.b, E-mail: ajp@cdtn.b, E-mail: aclc@cdtn.b, E-mail: hcr@cdtn.b [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN/CNEN-MG), Belo Horizonte, MG (Brazil)

2011-07-01

The IPR-R1 TRIGA (Training, Research, Isotopes production, General Atomics) at Nuclear Technology Development Center (CDTN) is a pool type reactor cooled by natural circulation of light water and an open surface. TRIGA reactors, developed by General Atomics (GA), are the most widely used research reactor in the world and characterized by inherent safety. The IPR-R1 is the only Brazilian nuclear research reactor available and able to perform experiments in which interaction between neutronic and thermal-hydraulic areas occurs. The IPR-R1 has started up on November 11th, 1960. At that time the maximum thermal power was 30 kW. The present forced cooling system was built in the 70th and the power was upgraded to 100 kW. Recently the core configuration and instrumentation was upgraded again to 250 kW at steady state, and is awaiting the license of CNEN to operate definitely at this new power. This paper describes the experimental research project carried out in the IPR-R1 reactor that has as objective evaluate the behaviour of the reactor operational parameters, and mainly to investigate the influence of temperature on the neutronic variables. The research was supported by Research Support Foundation of the State of Minas Gerais (FAPEMIG) and Brazilian Council for Scientific and Technological Development (CNPq). The research project meets the recommendations of the IAEA, for safety, modernization and development of strategic plan for research reactors utilization. This work is in line with the strategic objectives of Brazil, which aims to design and construct the Brazilian Multipurpose research Reactor (RMB). (author)

Experimental research in neutron physic and thermal-hydraulic at the CDTN Triga reactor

International Nuclear Information System (INIS)

Mesquita, Amir Z.; Souza, Rose Mary G.P.; Ferreira, Andrea V.; Pinto, Antonio J.; Costa, Antonio C.L.; Rezende, Hugo C.

2011-01-01

The IPR-R1 TRIGA (Training, Research, Isotopes production, General Atomics) at Nuclear Technology Development Center (CDTN) is a pool type reactor cooled by natural circulation of light water and an open surface. TRIGA reactors, developed by General Atomics (GA), are the most widely used research reactor in the world and characterized by inherent safety. The IPR-R1 is the only Brazilian nuclear research reactor available and able to perform experiments in which interaction between neutronic and thermal-hydraulic areas occurs. The IPR-R1 has started up on November 11th, 1960. At that time the maximum thermal power was 30 kW. The present forced cooling system was built in the 70th and the power was upgraded to 100 kW. Recently the core configuration and instrumentation was upgraded again to 250 kW at steady state, and is awaiting the license of CNEN to operate definitely at this new power. This paper describes the experimental research project carried out in the IPR-R1 reactor that has as objective evaluate the behaviour of the reactor operational parameters, and mainly to investigate the influence of temperature on the neutronic variables. The research was supported by Research Support Foundation of the State of Minas Gerais (FAPEMIG) and Brazilian Council for Scientific and Technological Development (CNPq). The research project meets the recommendations of the IAEA, for safety, modernization and development of strategic plan for research reactors utilization. This work is in line with the strategic objectives of Brazil, which aims to design and construct the Brazilian Multipurpose research Reactor (RMB). (author)

DORT-TD/THERMIX solutions for the OECD/NEA/NSC PBMR400 MW coupled neutronics thermal hydraulics transient benchmark

International Nuclear Information System (INIS)

Tyobeka, Bismark; Pautz, Andreas; Ivanov, Kostadin

2008-01-01

In new reactor designs that are still under review such as the PBMR, not much experimental data exists to benchmark newly developed computer codes against. Such a situation requires that nuclear engineers and designers of this novel reactor design must resort to the validation of a newly developed code through a code-to-code benchmarking exercise because there are validated codes that are currently in use to analyze this reactor design, albeit very few of them. There are numerous HTR core physics benchmarks that are currently being pursued by different organizations, for different purposes. One such benchmark exercise is the PBMR-400 MW OECD/NEA/NSC coupled neutronics/thermal hydraulics transient benchmark. In this paper, a newly developed coupled neutronics thermal hydraulics code system, DORT-TD/THERMIX with both transport and diffusion theory options, is used to simulate the transient scenarios in the above-mentioned benchmark problem. Steady-state calculations results are compared with selected participants' results as well as transient models in which the diffusion and transport theory solutions of the same code system are directly compared. Several sensitivity studies are also shown in order to determine how much the change in certain parameters influences the overall behaviour of a given transient. It is shown in this paper that DORT-TD/THERMIX is a versatile tool which can be deployed for design and safety analyses of high temperature reactors of pebble-bed type. (authors)

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 ...

TH3D, a three-dimensional thermal hydraulic tool, for design and safety analysis of HTRS - HTR2008-58178

International Nuclear Information System (INIS)

Hossain, K.; Buck, M.; Bernnat, W.; Lohnert, G.

2008-01-01

The institute of nuclear engineering and energy systems (IKE), Univ. of Stuttgart (Germany)) has developed a new thermal hydraulic tool which can be used for three-dimensional thermal hydraulic analysis of pebble bed as well as block type HTRs. During nominal operation, the flow inside the gas-cooled High Temperature Reactor is essentially single-phase, impressible, and non-isothermal. So, at least one gas phase has to be considered beside the solid phase for thermal hydraulic analysis of HTRs. Each phase (e.g. solid, gas) is considered as a continuum which occupies only its respective fraction of. the control volume. Thermal non-equilibrium is considered between phases and time dependent energy conservation equations for solid and gas phases are solved. Simplified momentum conservation equation for gas obtained from porous media approximation is solved along with the time dependent mass conservation equation. Pro visions for simulating more than one gas component are available in this newly developed code TH3D which could be required for simulating some accident situations (e.g air / water ingress by pipes break). The interaction between phases is made by a set of constitutive equations which re/v on semi-empirical correlations obtained from different experiments. Finite volume method with a staggered grid approach is used for spatial discretization and a fully implicit, time adaptive, multi step method is used for time-dependent discretization. A benchmark calculation which is oriented to the pebble i fuel reactor PBMR-400 and a 3D calculation were presented in HTR -2006 conference and will also be published in Nuclear Engineering and Design (NED) journal. In order to demonstrate the capabilities of TH3D for simulating all block type HTRs. A benchmark calculation which is proposed by IAEA CRP-3 and oriented to the Gas Turbine Modular Helium Reactor (GT-MHR) is performed. calculations are performed for the steady state case (nominal operation) as well as for Loss

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

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

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

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

expertise and findings and gathering together expert scientists from various areas relevant to the issues addressed. Added value for the CRISSUE-S activity consists of proposing and making available a list of transients to be analysed by coupled neutron kinetics/thermal-hydraulic techniques and of defining ?acceptability? (or required precision) thresholds for the results of the analyses. The list of transients is specific to the different NPP types such as PWR, BWR and VVER. The acceptability thresholds for calculation precision are general in nature and are applicable to all LWRs. The creation of a database including the main results from coupled 3-D neutron kinetics/thermal-hydraulic calculations and their analysis should also be noted. The CRISSUE-S project is organised into three work packages (WPs). The first WP includes activities related to obtaining and documenting relevant data. The second WP is responsible for the state-of-the-art report (SOAR), while the third WP concerns the evaluation of the findings from the SOAR and includes outcomes of the entire project formulated as recommendations, mainly to the nuclear power industry and to the regulatory authorities. The present report is the result of the first WP and discusses the type of transients that are of interest in relation to reactivity-initiated accidents in LWR NPPs and elaborates on the data required for coupled 3-D neutron kinetics/thermal hydraulic analysis and also on data needed to perform associated validations. The reports have been written to accomplish the objectives established in the contract between the EU and its partners. Expected beneficiaries include institutions and organisations involved with nuclear technology (e.g. utilities, regulators, research, fuel industry). In addition, specific expected beneficiaries are junior- or senior-level researchers and technologists working in the considered field of research and development and application of coupled neutron kinetics/thermal-hydraulics

HYTRAN: hydraulic transient code for investigating channel flow stability

International Nuclear Information System (INIS)

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

1976-01-01

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

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)

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)

Assimilation of temperature and hydraulic gradients for quantifying the spatial variability of streambed hydraulics

Science.gov (United States)

Huang, Xiang; Andrews, Charles B.; Liu, Jie; Yao, Yingying; Liu, Chuankun; Tyler, Scott W.; Selker, John S.; Zheng, Chunmiao

2016-08-01

Understanding the spatial and temporal characteristics of water flux into or out of shallow aquifers is imperative for water resources management and eco-environmental conservation. In this study, the spatial variability in the vertical specific fluxes and hydraulic conductivities in a streambed were evaluated by integrating distributed temperature sensing (DTS) data and vertical hydraulic gradients into an ensemble Kalman filter (EnKF) and smoother (EnKS) and an empirical thermal-mixing model. The formulation of the EnKF/EnKS assimilation scheme is based on a discretized 1D advection-conduction equation of heat transfer in the streambed. We first systematically tested a synthetic case and performed quantitative and statistical analyses to evaluate the performance of the assimilation schemes. Then a real-world case was evaluated to calculate assimilated specific flux. An initial estimate of the spatial distributions of the vertical hydraulic gradients was obtained from an empirical thermal-mixing model under steady-state conditions using a constant vertical hydraulic conductivity. Then, this initial estimate was updated by repeatedly dividing the assimilated specific flux by estimates of the vertical hydraulic gradients to obtain a refined spatial distribution of vertical hydraulic gradients and vertical hydraulic conductivities. Our results indicate that optimal parameters can be derived with fewer iterations but greater simulation effort using the EnKS compared with the EnKF. For the field application in a stream segment of the Heihe River Basin in northwest China, the average vertical hydraulic conductivities in the streambed varied over three orders of magnitude (5 × 10-1 to 5 × 102 m/d). The specific fluxes ranged from near zero (qz < ±0.05 m/d) to ±1.0 m/d, while the vertical hydraulic gradients were within the range of -0.2 to 0.15 m/m. The highest and most variable fluxes occurred adjacent to a debris-dam and bridge pier. This phenomenon is very likely

Thermal response of a pin-type fusion reactor blanket during steady and transient reactor operation

International Nuclear Information System (INIS)

Grotz, S.; Ghoniem, N.M.

1986-02-01

The thermal analysis of the blanket examines both the steady-state and transient reactor operations. The steady-state analysis covers full power and fractional power operation whereas the transient analysis examines the effects of power ramps and blanket preheat. The blanket configuration chosen for this study is a helium cooled solid breeder design. We first discuss the full power, steady-state temperature fields in the first wall, beryllium rods, and breeder rods. Next we examine the effects of fractional power on coolant flow and temperature field distributions. This includes power plateaus of 10%, 20%, 50%, 80%, and 100% of full power. Also examined are the restrictions on the rates of power ramping between plateaus. Finally we discuss the power and time requirements for pre-heating the primary from cold iron conditions up to startup temperature (250 0 C)

Output regulation of large-scale hydraulic networks with minimal steady state power consumption

NARCIS (Netherlands)

Jensen, Tom Nørgaard; Wisniewski, Rafał; De Persis, Claudio; Kallesøe, Carsten Skovmose

2014-01-01

An industrial case study involving a large-scale hydraulic network is examined. The hydraulic network underlies a district heating system, with an arbitrary number of end-users. The problem of output regulation is addressed along with a optimization criterion for the control. The fact that the

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

steady – state performance of induction and transfer state

African Journals Online (AJOL)

eobe

This paper presents paper presents paper presents the steady the steady the steady–state performance state performance state performance comparison comparison comparison between polyphase induction motor and polyphase between polyphase induction motor and polyphase. TF motor operating in. TF motor ...

A Physics-Based Rock Friction Constitutive Law: Steady State Friction

Science.gov (United States)

Aharonov, Einat; Scholz, Christopher H.

2018-02-01

Experiments measuring friction over a wide range of sliding velocities find that the value of the friction coefficient varies widely: friction is high and behaves according to the rate and state constitutive law during slow sliding, yet markedly weakens as the sliding velocity approaches seismic slip speeds. We introduce a physics-based theory to explain this behavior. Using conventional microphysics of creep, we calculate the velocity and temperature dependence of contact stresses during sliding, including the thermal effects of shear heating. Contacts are assumed to reach a coupled thermal and mechanical steady state, and friction is calculated for steady sliding. Results from theory provide good quantitative agreement with reported experimental results for quartz and granite friction over 11 orders of magnitude in velocity. The new model elucidates the physics of friction and predicts the connection between friction laws to independently determined material parameters. It predicts four frictional regimes as function of slip rate: at slow velocity friction is either velocity strengthening or weakening, depending on material parameters, and follows the rate and state friction law. Differences between surface and volume activation energies are the main control on velocity dependence. At intermediate velocity, for some material parameters, a distinct velocity strengthening regime emerges. At fast sliding, shear heating produces thermal softening of friction. At the fastest sliding, melting causes further weakening. This theory, with its four frictional regimes, fits well previously published experimental results under low temperature and normal stress.

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

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.

Two-Dimensional Steady-State Boundary Shape Inversion of CGM-SPSO Algorithm on Temperature Information

Directory of Open Access Journals (Sweden)

Shoubin Wang

2017-01-01

Full Text Available Addressing the problem of two-dimensional steady-state thermal boundary recognition, a hybrid algorithm of conjugate gradient method and social particle swarm optimization (CGM-SPSO algorithm is proposed. The global search ability of particle swarm optimization algorithm and local search ability of gradient algorithm are effectively combined, which overcomes the shortcoming that the conjugate gradient method tends to converge to the local solution and relies heavily on the initial approximation of the iterative process. The hybrid algorithm also avoids the problem that the particle swarm optimization algorithm requires a large number of iterative steps and a lot of time. The experimental results show that the proposed algorithm is feasible and effective in solving the problem of two-dimensional steady-state thermal boundary shape.

Validation of a CATHENA fuel channel model for the post blowdown analysis of the high temperature thermal-chemical experiment CS28-1, I - Steady state

International Nuclear Information System (INIS)

Rhee, Bo Wook; Kim, Hyoung Tae; Park, Joo Hwan

2008-01-01

To form a licensing basis for the new methodology of the fuel channel safety analysis code system for CANDU-6, a CATHENA model for the post-blowdown fuel channel analysis for a Large Break LOCA has been developed, and tested for the steady state of a high temperature thermal-chemical experiment CS28-1. As the major concerns of the post-blowdown fuel channel analysis of the current CANDU-6 design are how much of the decay heat can be discharged to the moderator via a radiation and a convective heat transfer at the expected accident conditions, and how much zirconium sheath would be oxidized to generate H 2 at how high a fuel temperature, this study has focused on understanding these phenomena, their interrelations, and a way to maintain a good accuracy in the prediction of the fuel and the pressure tube temperatures without losing the important physics of the involved phenomena throughout the post-blowdown phase of a LBLOCA. For a better prediction, those factors that may significantly contribute to the prediction accuracy of the steady state of the test bundles were sought. The result shows that once the pressure tube temperature is predicted correctly by the CATHENA heat transfer model between the pressure tube and the calandria tube through a gap thermal resistance adjustment, all the remaining temperatures of the inner ring, middle ring and outer ring FES temperatures can be predicted quite satisfactorily, say to within an accuracy range of 20-25 deg. C, which is comparable to the reported accuracy of the temperature measurement, ±2%. Also the analysis shows the choice of the emissivity of the solid structures (typically, 0.80, 0.34, 0.34 for FES, PT, CT), and the thermal resistance across the CO 2 annulus are factors that significantly affect the steady state prediction accuracy. A question on the legitimacy of using 'transparent' assumption for the CO 2 gas annulus for the radiation heat transfer between the pressure tube and the calandria tube in CATHENA

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)

Comparison of thermo-hydraulic analysis with measurements for HELIOS. The scaled integral test loop for PEACER

International Nuclear Information System (INIS)

Cho, Jae Hyun; Lim, Jun; Kim, Ji Hak; Hwang, Il Soon

2009-01-01

A scaled-down Lead-Bismuth Eutectic circulating integral test loop named as HELIOS (Heavy Eutectic liquid metal Loop for Integral test of Operability and Safety of PEACER) has been employed to characterize steady-state isothermal forced circulation behavior and non-isothermal natural circulation capability of the lead and lead-alloy cooled advanced nuclear energy systems (LACANES). In this time, thermal-hydraulic experiments have been carried out using HELIOS following rigorous calibration campaigns on sensors for temperature and pressure, especially isothermal steady-state forced convection using by the pump. The isothermal steady-state forced convection test was performed to obtain the pressure loss information including friction loss coefficients and form loss coefficients. Then its data were compared with multi-approaching analysis including hand calculation results and computer simulation code results. (MARS-LBE, CFX). We report the results of comparisons between the analysis and measurements together. (author)

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

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)

Steady-state and pre-steady-state kinetic analysis of halopropane conversion by a Rhodococcus haloalkane dehalogenase

NARCIS (Netherlands)

Bosma, T; Pikkemaat, MG; Kingma, Jacob; Dijk, J; Janssen, DB

2003-01-01

Haloalkane dehalogenase from Rhodococcus rhodochrous NCIMB 13064 (DhaA) catalyzes the hydrolysis of carbon-halogen bonds in a wide range of haloalkanes. We examined the steady-state and pre-steady-state kinetics of halopropane conversion by DhaA to illuminate mechanistic details of the

Pseudo Steady-State Free Precession for MR-Fingerprinting.

Science.gov (United States)

Assländer, Jakob; Glaser, Steffen J; Hennig, Jürgen

2017-03-01

This article discusses the signal behavior in the case the flip angle in steady-state free precession sequences is continuously varied as suggested for MR-fingerprinting sequences. Flip angle variations prevent the establishment of a steady state and introduce instabilities regarding to magnetic field inhomogeneities and intravoxel dephasing. We show how a pseudo steady state can be achieved, which restores the spin echo nature of steady-state free precession. Based on geometrical considerations, relationships between the flip angle, repetition and echo time are derived that suffice to the establishment of a pseudo steady state. The theory is tested with Bloch simulations as well as phantom and in vivo experiments. A typical steady-state free precession passband can be restored with the proposed conditions. The stability of the pseudo steady state is demonstrated by comparing the evolution of the signal of a single isochromat to one resulting from a spin ensemble. As confirmed by experiments, magnetization in a pseudo steady state can be described with fewer degrees of freedom compared to the original fingerprinting and the pseudo steady state results in more reliable parameter maps. The proposed conditions restore the spin-echo-like signal behavior typical for steady-state free precession in fingerprinting sequences, making this approach more robust to B 0 variations. Magn Reson Med 77:1151-1161, 2017. © 2016 International Society for Magnetic Resonance in Medicine. © 2016 International Society for Magnetic Resonance in Medicine.

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)

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

Multimode optical fibers: steady state mode exciter.

Science.gov (United States)

Ikeda, M; Sugimura, A; Ikegami, T

1976-09-01

The steady state mode power distribution of the multimode graded index fiber was measured. A simple and effective steady state mode exciter was fabricated by an etching technique. Its insertion loss was 0.5 dB for an injection laser. Deviation in transmission characteristics of multimode graded index fibers can be avoided by using the steady state mode exciter.

Interactions of district electricity and heating systems considering time-scale characteristics based on quasi-steady multi-energy flow

International Nuclear Information System (INIS)

Pan, Zhaoguang; Guo, Qinglai; Sun, Hongbin

2016-01-01

Highlights: • Interaction mechanisms of district electricity and heating systems are analyzed. • The interaction process is divided into four quasi-steady stages. • A quasi-steady multi-energy flow model is proposed and calculated. • A heating network node type transformation technique is developed. • Attention should be paid on the fast hydraulic process and slow thermal process. - Abstract: Integrated energy systems (IESs) are under development for a variety of benefits. District electricity and heating systems (DEHSs) deliver electricity and heat, the most common energy demands, to end-users. This paper studies the interactions in a DEHS considering the time-scale characteristics. Interaction mechanisms of a DEHS are analyzed. A disturbance in one system influences another system through coupling components, depending on the disturbance, operating characteristics, and control strategies. A model of the main components in DEHSs is presented. The time scale characteristics are studied based on a dynamic comparison of the different components. Then the interaction process is divided into four stages; each is a quasi-steady state. A quasi-steady multi-energy flow model is proposed and calculated, with a heating network node type transformation technique developed. A case study with detailed results and discussion of 3 types of disturbance is presented to verify the methods. The results present the interactions between the electricity and the system. It is suggested that attention should be paid both on the fast hydraulic process and slow thermal process for system security and economic operation.

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

Theoretical and experimental studies of heavy liquid metal thermal hydraulics. Proceedings of a technical meeting

International Nuclear Information System (INIS)

2006-10-01

Through the Nuclear Energy Department's Technical Working Group on Fast Reactors (TWG-FR), the IAEA provides a forum for exchange of information on national programmes, collaborative assessments, knowledge preservation, and cooperative research in areas agreed by the Member States with fast reactor and partitioning and transmutation development programmes (e.g. accelerator driven systems (ADS)). Trends in advanced fast reactor and ADS designs and technology development are periodically summarized in status reports, symposia, and seminar proceedings prepared by the IAEA to provide all interested IAEA Member States with balanced and objective information. The use of heavy liquid metals (HLM) is rapidly diffusing in different research and industrial fields. The detailed knowledge of the basic thermal hydraulics phenomena associated with their use is a necessary step for the development of the numerical codes to be used in the engineering design of HLM components. This is particularly true in the case of lead or lead-bismuth eutectic alloy cooled fast reactors, high power particle beam targets and in the case of the cooling of accelerator driven sub-critical cores where the use of computational fluid dynamic (CFD) design codes is mandatory. Periodic information exchange within the frame of the TWG-FR has lead to the conclusion that the experience in HLM thermal fluid dynamics with regard to both the theoretical/numerical and experimental fields was limited and somehow dispersed. This is the case, e.g. when considering turbulent exchange phenomena, free-surface problems, and two-phase flows. Consequently, Member States representatives participating in the 35th Annual Meeting of the TWG-FR (Karlsruhe, Germany, 22-26 April 2002) recommended holding a technical meeting (TM) on Theoretical and Experimental Studies of Heavy Liquid Metal Thermal Hydraulics. Following this recommendation, the IAEA has convened the Technical Meeting on Theoretical and Experimental Studies of

Thermal-hydraulic calculation and water hammer analysis on CEFR loop system

International Nuclear Information System (INIS)

Hao Pengfei; Zhang Xiwen; Cai Weidong; Wang Xuefang

1997-01-01

China Experimental Fast Reactor (CEFR) is one of the '863' High-technical Projects. It is necessary to study the hydraulic and thermal Characteristic of CEFR loop system in order to guarantee the safety of operation. The results of the thermal-hydraulic calculation have been given. The main points are as follows: 1. The simplified model is built according to the loop system of CEFR, and the calculation method which is called 'NODE'-'BRANCH' is applied. This method includes two aspects, one is the theoretical analysis that is based on fluid mechanics and heat transfer theory. The other is the engineering calculation. These two aspects are connected in the computation. On the basis of the work mentioned above, the stable state computation is presented. In order to prevent serious damage caused by power failure accident, the courses of surplus reactor heat removing through two different systems have been simulated in the computation. 2. By using the fluid dynamics theory, the simplified model and the equipment boundary conditions of loop system are given. The water hammer computation is processed during the valve closing and pump stopping accidents. Some pictures of water hammer wave are presented, and the most dangerous state in the accident is also given

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.

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

Interface requirements to couple thermal-hydraulic codes to severe accident codes: ATHLET-CD

Energy Technology Data Exchange (ETDEWEB)

Trambauer, K. [GRS, Garching (Germany)

1997-07-01

The system code ATHLET-CD is being developed by GRS in cooperation with IKE and IPSN. Its field of application comprises the whole spectrum of leaks and large breaks, as well as operational and abnormal transients for LWRs and VVERs. At present the analyses cover the in-vessel thermal-hydraulics, the early phases of core degradation, as well as fission products and aerosol release from the core and their transport in the Reactor Coolant System. The aim of the code development is to extend the simulation of core degradation up to failure of the reactor pressure vessel and to cover all physically reasonable accident sequences for western and eastern LWRs including RMBKs. The ATHLET-CD structure is highly modular in order to include a manifold spectrum of models and to offer an optimum basis for further development. The code consists of four general modules to describe the reactor coolant system thermal-hydraulics, the core degradation, the fission product core release, and fission product and aerosol transport. Each general module consists of some basic modules which correspond to the process to be simulated or to its specific purpose. Besides the code structure based on the physical modelling, the code follows four strictly separated steps during the course of a calculation: (1) input of structure, geometrical data, initial and boundary condition, (2) initialization of derived quantities, (3) steady state calculation or input of restart data, and (4) transient calculation. In this paper, the transient solution method is briefly presented and the coupling methods are discussed. Three aspects have to be considered for the coupling of different modules in one code system. First is the conservation of masses and energy in the different subsystems as there are fluid, structures, and fission products and aerosols. Second is the convergence of the numerical solution and stability of the calculation. The third aspect is related to the code performance, and running time.

Lower bounds for ballistic current and noise in non-equilibrium quantum steady states

Directory of Open Access Journals (Sweden)

Benjamin Doyon

2015-03-01

Full Text Available Let an infinite, homogeneous, many-body quantum system be unitarily evolved for a long time from a state where two halves are independently thermalized. One says that a non-equilibrium steady state emerges if there are nonzero steady currents in the central region. In particular, their presence is a signature of ballistic transport. We analyze the consequences of the current observable being a conserved density; near equilibrium this is known to give rise to linear wave propagation and a nonzero Drude peak. Using the Lieb–Robinson bound, we derive, under a certain regularity condition, a lower bound for the non-equilibrium steady-state current determined by equilibrium averages. This shows and quantifies the presence of ballistic transport far from equilibrium. The inequality suggests the definition of “nonlinear sound velocities”, which specialize to the sound velocity near equilibrium in non-integrable models, and “generalized sound velocities”, which encode generalized Gibbs thermalization in integrable models. These are bounded by the Lieb–Robinson velocity. The inequality also gives rise to a bound on the energy current noise in the case of pure energy transport. We show that the inequality is satisfied in many models where exact results are available, and that it is saturated at one-dimensional criticality.

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

Thermal hydraulic reactor safety analyses and experiments

International Nuclear Information System (INIS)

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

1989-04-01

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

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)

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)

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.

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.).

Experimental study on thermal-hydraulic characteristics of secondary-side passive ECRHR system of AC600

International Nuclear Information System (INIS)

Chen Bingde; Xiao Zejun; Zhuo Wenbin

1998-10-01

An integral system experimental test rig was designed to study the thermal-hydraulic characteristics of secondary-side passive emergency core residual heat removal (ECRHR) system of AC600. Total sets of 166 experimental data on steady state have been obtained and main influence factors on heat removal capacity are identified. The experimental data shows that core residual heat can be removed through natural circulation under the condition of circumstance temperature 6∼23 degree C and L th of 11 m, provided wind-speed higher than 1.5 m/s in the test parameter scope. Experimental results illustrate that short disturbance of wind-speed, power and valve opening had no significant impact on natural circulation. The system seems to tolerate to these disturbance. A correlation of two-phase natural circulation flow rate is derived from constitutive equations by using of specific system parameters and several semi-empirical formulas are obtained. Compared with the measuring data, the deviation of 98.8% data points is within +-15%. The objectives of transient experiment study is to investigate the start-up modes and transition behaviors. The results of 46 runs show the natural circulation of system can be set up by means of all three start-up modes, and its transient characteristics, especially, the system stability and transit time are determined by not only start-up modes, but also system's configuration, initial and boundary conditions, mainly. The calculating results of MISAP-PRHR code, a computer code developed for passive safety system, reveal that it can predict quantitatively the steady state characteristics, such as, pressure, flowrate, heat removal capacity and the stable transient process can be well simulated, too

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

Steady-State Creep of Asphalt Concrete

Directory of Open Access Journals (Sweden)

Alibai Iskakbayev

2017-02-01

Full Text Available This paper reports the experimental investigation of the steady-state creep process for fine-grained asphalt concrete at a temperature of 20 ± 2 °С and under stress from 0.055 to 0.311 MPa under direct tension and was found to occur at a constant rate. The experimental results also determined the start, the end point, and the duration of the steady-state creep process. The dependence of these factors, in addition to the steady-state creep rate and viscosity of the asphalt concrete on stress is satisfactorily described by a power function. Furthermore, it showed that stress has a great impact on the specific characteristics of asphalt concrete: stress variation by one order causes their variation by 3–4.5 orders. The described relations are formulated for the steady-state of asphalt concrete in a complex stressed condition. The dependence is determined between stress intensity and strain rate intensity.

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.

Effects of design variables predicted by a steady - state thermal performance analysis model of a loop heat pipe

International Nuclear Information System (INIS)

Jung, Eui Guk; Boo, Joon Hong

2008-01-01

This study deals with a mathematical modeling for the steady-state temperature characteristics of an entire loop heat pipe. The lumped layer model was applied to each node for temperature analysis. The flat type evaporator and condenser in the model had planar dimensions of 40 mm (W) x 50 mm (L). The wick material was a sintered metal and the working fluid was methanol. The molecular kinetic theory was employed to model the phase change phenomena in the evaporator and the condenser. Liquid-vapor interface configuration was expressed by the thin film theories available in the literature. Effects of design factors of loop heat pipe on the thermal performance were investigated by the modeling proposed in this study

HORUS3D/TH: thermal-hydraulic modelling of the Jules Horowitz reactor core with FLICA4

International Nuclear Information System (INIS)

Royer, E.; Gregoire, O.; Magnaud, J.P.; Roux, L.; Masson, X.

2007-01-01

Cea is planning to build a new pool type reactor as irradiation facility in Cadarache, France: the Jules Horowitz Reactor (JHR). For this purpose, a simulation program is carried out at Cea: HORUS3D, aimed at modeling neutronics, radio-protection and thermal-hydraulics. Advanced features of the thermal-hydraulics component of this simulation program (HORUS3D/TH) are presented in the paper. HORUS3D/TH is based on the FLICA4 thermalhydraulic code. Numerically the main features of HORUS3D/TH are unstructured mesh grids and non-conform mappings. From a phenomenological point of view, flows under study range from high velocity forced convection to natural convection regimes. Steady and transient regimes have been simulated. The validation of physical models used is an important part of HORUS3D project. For thermohydraulics, this validation relies on the SULTAN-RJH experimental facility and fine scale CFD simulations. We have shown in this paper that it is possible to calibrate the macroscopic heat exchange correlation in the forced convection regime and under very high heat fluxes thanks to low Reynolds fine scale calculations. We particularly underline how to cope with the difficulties due to the complex geometry (cylindrical fuel assemblies, made of curved plates) and very high pressure drops and heat fluxes

Effect of stacking fault energy on steady-state creep rate of face ...

African Journals Online (AJOL)

Continuum elastic theory was used to establish the relationships between the force of interaction required to constrict dislocation partials, energy of constriction and climb velocity of the constricted thermal jogs, in order to examine the effect of stacking fault energy (SFE) on steady state creep rate of face centered cubic ...

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.

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

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

An accelerator based steady state neutron source

International Nuclear Information System (INIS)

Burke, R.J.; Johnson, D.L.

1985-01-01

Using high current, c.w. linear accelerator technology, a spallation neutron source can achieve much higher average intensities than existing or proposed pulsed spallation sources. With about 100 mA of 300 MeV protons or deuterons, the Accelerator Based Neutron Research Facility (ABNR) would initially achieve the 10 16 n/cm 2 .s thermal flux goal of the advanced steady state neutron source, and upgrading could provide higher steady state fluxes. The relatively low ion energy compared to other spallation sources has an important impact on R and D requirements as well as capital cost, for which a range of $300-450M is estimated by comparison to other accelerator-based neutron source facilities. The source is similar to a reactor source in most respects. It has some higher energy neutrons but fewer gamma rays, and the moderator region is free of many of the design constraints of a reactor, which helps to implement sources for various neutron energy spectra, many beam tubes, etc. With the development of multi-beam concept and the basis for currents greater than 100 mA that is assumed in the R and D plan, the ABNR would serve many additional uses, such as fusion materials development, production of proton-rich isotopes, and other energy and defense program needs

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.

Steady-State Characterization of Bacteriorhodopsin-D85N Photocycle

Science.gov (United States)

Timucin, Dogan A.; Downie, John D.; Norvig, Peter (Technical Monitor)

1999-01-01

An operational characterization of the photocycle of the genetic mutant D85N of bacteriorhodopsin, BR-D85N, is presented. Steady-state bleach spectra and pump-probe absorbance data are obtained with thick hydrated films containing BR-D85N embedded in a gelatin host. Simple two- and three-state models are used to analyze the photocycle dynamics and extract relevant information such as pure-state absorption spectra, photochemical-transition quantum efficiencies, and thermal lifetimes of dominant states appearing in the photocycle, the knowledge of which should aid in the analysis of optical recording and retrieval of data in films incorporating this photochromic material. The remarkable characteristics of this material and their implications from the viewpoint of optical data storage and processing are discussed.

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

Effect of non-condensable gas on steady-state operation of a loop thermosyphon

International Nuclear Information System (INIS)

He, Jiang; Lin, Guiping; Bai, Lizhan; Miao, Jianyin; Zhang, Hongxing; Wang, Lu

2014-01-01

Non-condensable gas (NCG) generated inside two-phase heat transfer devices can adversely affect the thermal performance and limit the lifetime of such devices. In this work, extensive experimental investigation of the effect of NCG on the steady-state operation of an ammonia-stainless steel loop thermosyphon was conducted. In the experiments, nitrogen was injected into the loop thermosyphon as NCG, and the thermal performance of the loop thermosyphon was tested at different NCG inventories, heat loads applied to the evaporator and condenser cooling conditions, i.e. natural air cooling or circulating ethanol cooling. Experimental results reveal that NCG elevates the steady-state operating temperature of the evaporator, especially when the loop thermosyphon is operating in the low temperature range; meanwhile, the more NCG exists in the loop thermosyphon, the higher the operating temperature of the evaporator, and the lower the reservoir temperature. In addition, the existence of NCG results in the decrease of the overall thermal conductance of the loop thermosyphon, and the overall thermal conductance under the ethanol cooling condition may be even lower than that under the air cooling condition when the heat load is smaller than a certain value. Finally, the experimental results are theoretically analysed and explained. (authors)

Stirling/hydraulic artificial heart power source

International Nuclear Information System (INIS)

Johnston, R.P.; Bennett, A.; Emigh, S.G.; Griffith, W.R.; Noble, J.E.; Perrone, R.E.; White, M.A.; Martini, W.R.; Alexander, J.E.

1977-01-01

The REL power source combines the high efficiency of Stirling engines with the reliability, efficiency, and flexibility of hydraulic power transfer and control to ensure long system life and physiological effectiveness. Extended life testing has been achieved with an engine (2.6 years) and hydraulic actuator/controller (1.6 years). Peak power source efficiency is 15.5 percent on 5 to 10 watts delivered to the blood pump push plate with 33 watts steady thermal input. Planned incorporation of power source output control is expected to reduce daily average thermal input to 18 watts. Animal in-vivo tests with an assist heart have consistently demonstrated required performance by biological synchronization and effective ventricle relief. Volume and weight are 0.93 liter and 2.4 kg (excluding blood pump) with an additional 0.4 liter of low temperature foam insulation required to preclude tissue thermal damage. Carefully planned development of System 7 is expected to produce major reductions in size

Thermo-Hydraulic behaviour of dual-channel superconducting Cable-In-Conduit Conductors for ITER

International Nuclear Information System (INIS)

Renard, B.

2006-09-01

In an effort to optimise the cryogenics of large superconducting coils for fusion applications (ITER), dual channel Cable-In-Conduit Conductors (CICC) are designed with a central channel spiral to provide low hydraulic resistance and faster helium circulation. The qualitative and economic rationale of the conductor central channel is here justified to limit the superconductor temperature increase, but brings more complexity to the conductor cooling characteristics. The pressure drop of spirals is experimentally evaluated in nitrogen and water and an explicit hydraulic friction model is proposed. Temperatures in the cable must be quantified to guarantee superconductor margin during coil operation under heat disturbance and set adequate inlet temperature. Analytical one-dimensional thermal models, in steady state and in transient, allow to better understand the thermal coupling of CICC central and annular channels. The measurement of a heat transfer characteristic space and time constants provides cross-checking experimental estimations of the internal thermal homogenization. A simple explicit model of global inter-channel heat exchange coefficient is proposed. The risk of thermosyphon between the two channels is considered since vertical portions of fusion coils are subject to gravity. The new hydraulic model, heat exchange model and gravitational risk ratio allow the thermohydraulic improvement of CICC central spirals. (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

Data report of tight-lattice rod bundle thermal-hydraulic tests (2). Gap-width effect test using 37-rod bundle simulated water-cooled breeder reactor (Contract research)

International Nuclear Information System (INIS)

Tamai, Hidesada; Kureta, Masatoshi; Liu, Wei; Akimoto, Hajime; Sato, Takashi; Watanabe, Hironori; Ohnuki, Akira

2006-11-01

Japan Atomic Energy Agency has been performing tight-lattice rod bundle thermal-hydraulic tests to realize essential technologies for the technological and engineering feasibility of super high burn-up water-cooled breeder reactor featured by a high breeding ratio and super high burn-up by reducing the core water volume in water-cooled reactor. The tests are performing to make clear the fundamental subjects related to the boiling transition (BT) (Subjects: BT criteria under a highly tight-lattice rod bundle, effects of gap-width between rods and of rod-bowing) using 37-rod bundles (Base case test section (1.3mm gap-width), Two parameter effect test sections (Gap-width effect one (1.0mm) and Rod-bowing one)). In the present report, we summarize the test results from the gap-width effect test section. The thermal-hydraulic characteristics were obtained for the critical power under the steady-state and transient conditions, the pressure drop and the wall heat transfer within a wide range of pressure, flow rate, etc. including normal operational conditions of the designed reactor. Then the gap-width effects were also obtained from the comparison between the results using the base case test section and the gap-width effect one. (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 and hydraulic analyses of TFTR cooling water system and magnetic field coils

International Nuclear Information System (INIS)

Lee, A.Y.

1975-10-01

The TFTR toroidal field coils, ohmic heating, hybrid and equilibrium field coils are cooled by water from the machine area cooling water system. The system has the following major equipment and capacities: flow rate of 3600 gpm; ballast tank volume of 5500 gal; pumps of 70.4 m head; chiller refrigeration rating of 3300 tons and connecting pipe of 45.7 cm I.D. The performance of the closed loop system was analyzed and found to be adequate for the thermal loads. The field coils were analyzed with detailed thermal and hydraulic models, including a simulation of the complete water cooling loop. Under the nominal operating mode of one second of toroidal field flat top time and 300 seconds of pulse cycle time, the maximum temperature for the TF coils is 53 0 C; for the OH coils 46 0 C and for the EF coils 39 0 C, which are well below the coil design limit of 120 0 C. The maximum TF coil coolant temperature is 33 0 C which is below the coolant design limit of 100 0 C. The overall pressure loss of the system is below 6.89 x 10 5 Pa (100 psi). With the given chiller refrigeration capacity, the TF coils can be operated to yield up to 4 seconds of flat top time. The TF coils can be operated on a steady state basis at up to 20% of the pulsed duty design current rating of 7.32 kA/coil

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)

Practical steady-state enzyme kinetics.

Science.gov (United States)

Lorsch, Jon R

2014-01-01

Enzymes are key components of most biological processes. Characterization of enzymes is therefore frequently required during the study of biological systems. Steady-state kinetics provides a simple and rapid means of assessing the substrate specificity of an enzyme. When combined with site-directed mutagenesis (see Site-Directed Mutagenesis), it can be used to probe the roles of particular amino acids in the enzyme in substrate recognition and catalysis. Effects of interaction partners and posttranslational modifications can also be assessed using steady-state kinetics. This overview explains the general principles of steady-state enzyme kinetics experiments in a practical, rather than theoretical, way. Any biochemistry textbook will have a section on the theory of Michaelis-Menten kinetics, including derivations of the relevant equations. No specific enzymatic assay is described here, although a method for monitoring product formation or substrate consumption over time (an assay) is required to perform the experiments described. © 2014 Elsevier Inc. All rights reserved.

A time-dependent neutron transport model and its coupling to thermal-hydraulics

International Nuclear Information System (INIS)

Pautz, A.

2001-01-01

A new neutron transport code for time-dependent analyses of nuclear systems has been developed. The code system is based on the well-known Discrete Ordinates code DORT, which solves the steady-state neutron/photon transport equation in two dimensions for an arbitrary number of energy groups and the most common regular geometries. For the implementation of time-dependence a fully implicit first-order scheme was employed to minimize errors due to temporal discretization. This requires various modifications to the transport equation as well as the extensive use of elaborated acceleration mechanisms. The convergence criteria for fluxes, fission rates etc. had to be strongly tightened to ensure the reliability of results. To perform coupled analyses, an interface to the GRS system code ATHLET has been developed. The nodal power densities from the neutron transport code are passed to ATHLET to calculate thermal-hydraulic system parameters, e.g. fuel and coolant temperatures. These are in turn used to generate appropriate nuclear cross sections by interpolation of pre-calculated data sets for each time step. Finally, to demonstrate the transient capabilities of the coupled code system, the research reactor FRM-II has been analysed. Several design basis accidents were modelled, like the loss of off site power, loss of secondary heat sink and unintended control rod withdrawal. (author)

Molybdenum-99-producing 37-element fuel bundle neutronically and thermal-hydraulically equivalent to a standard CANDU fuel bundle

Energy Technology Data Exchange (ETDEWEB)

Nichita, E., E-mail: Eleodor.Nichita@uoit.ca; Haroon, J., E-mail: Jawad.Haroon@uoit.ca

2016-10-15

Highlights: • A 37-element fuel bundle modified for {sup 99}Mo production in CANDU reactors is presented. • The modified bundle is neutronically and thermal-hydraulically equivalent to the standard bundle. • The modified bundle satisfies all safety criteria satisfied by the standard bundle. - Abstract: {sup 99m}Tc, the most commonly used radioisotope in diagnostic nuclear medicine, results from the radioactive decay of {sup 99}Mo which is currently being produced at various research reactors around the globe. In this study, the potential use of CANDU power reactors for the production of {sup 99}Mo is investigated. A modified 37-element fuel bundle, suitable for the production of {sup 99}Mo in existing CANDU-type reactors is proposed. The new bundle is specifically designed to be neutronically and thermal-hydraulically equivalent to the standard 37-element CANDU fuel bundle in normal, steady-state operation and, at the same time, be able to produce significant quantities of {sup 99}Mo when irradiated in a CANDU reactor. The proposed bundle design uses fuel pins consisting of a depleted-uranium centre surrounded by a thin layer of low-enriched uranium. The new molybdenum-producing bundle is analyzed using the lattice transport code DRAGON and the diffusion code DONJON. The proposed design is shown to produce 4081 six-day Curies of {sup 99}Mo activity per bundle when irradiated in the peak-power channel of a CANDU core, while maintaining the necessary reactivity and power rating limits. The calculated {sup 99}Mo yield corresponds to approximately one third of the world weekly demand. A production rate of ∼3 bundles per week can meet the global demand of {sup 99}Mo.

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

Unsteady-state human-body exergy consumption rate and its relation to subjective assessment of dynamic thermal environments

DEFF Research Database (Denmark)

Schweiker, Marcel; Kolarik, Jakub; Dovjak, Mateja

2016-01-01

of the present study confirmed previously indicated trends that lowest human body exergy consumption rate is associated with thermal sensation close to neutrality. Moreover, higher acceptability was in general associated with lower human body exergy consumption rate. (C) 2016 Elsevier B.V. All rights reserved.......Few examples studied applicability of exergy analysis on human thermal comfort. These examples relate the human-body exergy consumption rate with subjectively obtained thermal sensation votes and had been based on steady-state calculation methods. However, humans are rarely exposed to steady...... between the human-body exergy consumption rate and subjective assessment of thermal environment represented by thermal sensation as well as to extend the investigation towards thermal acceptability votes. Comparison of steady-state and unsteady-state model showed that results from both models were...

Steady-State Performance of Kalman Filter for DPLL

Institute of Scientific and Technical Information of China (English)

QIAN Yi; CUI Xiaowei; LU Mingquan; FENG Zhenming

2009-01-01

For certain system models, the structure of the Kalman filter is equivalent to a second-order vari-able gain digital phase-locked loop (DPLL). To apply the knowledge of DPLLs to the design of Kalman filters, this paper studies the steady-state performance of Kalman filters for these system models. The results show that the steady-state Kalman gain has the same form as the DPLL gain. An approximate simple form for the steady-state Kalman gain is used to derive an expression for the equivalent loop bandwidth of the Kalman filter as a function of the process and observation noise variances. These results can be used to analyze the steady-state performance of a Kalman filter with DPLL theory or to design a Kalman filter model with the same steady-state performance as a given DPLL.

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

Review of turbulence modelling for numerical simulation of nuclear reactor thermal-hydraulics

International Nuclear Information System (INIS)

Bernard, J.P.; Haapalehto, T.

1996-01-01

The report deals with the modelling of turbulent flows in nuclear reactor thermal-hydraulic applications. The goal is to give tools and knowledge about turbulent flows and their modelling in practical applications for engineers, and especially nuclear engineers. The emphasize is on the theory of turbulence, the existing different turbulence models, the state-of-art of turbulence in research centres, the available models in the commercial code CFD-FLOW3D, and the latest applications of turbulence modelling in nuclear reactor thermal-hydraulics. It turns out that it is difficult to elaborate an universal turbulence model and each model has its advantages and drawbacks in each application. However, the increasing power of computers can permit the emergence of new methods of turbulence modelling such as Direct Numerical Simulation (DNS) and Large Eddy Simulation (LES) which open new horizons in this field. These latter methods are beginning to be available in commercial codes and are used in different nuclear applications such as 3-D modelling of the nuclear reactor cores and the steam generators. (orig.) (22 refs.)

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

Transient and steady-state flows in shock tunnels

Energy Technology Data Exchange (ETDEWEB)

Hannemann, K. [Deutsche Forschungsanstalt fuer Luft- und Raumfahrt e.V. (DLR), Goettingen (Germany); Jacobs, P.A. [Queensland Univ., Brisbane (Australia). Dept. of Mechanical Engineering; Thomas, A.; McIntyre, T.J. [Queensland Univ., Brisbane, QLD. (Australia). Dept. of Physics

1999-12-01

Due to the difficulty of measuring all necessary flow quantities in the nozzle reservoir and the test section of high enthalpy shock tunnels, indirect computational methods are necessary to estimate the required flow parameters. In addition to steady state flow computations of the nozzle flow and the flow past wind tunnel models it is necessary to investigate the transient flow in the facility in order to achieve a better understanding of its performance. These transient effects include the nozzle starting flow, the interaction of the shock tube boundary layers and the reflected shock, thermal losses in the shock reflection region and the developing boundary layers in the expanding section of the nozzle. Additionally, the nonequilibrium chemical and thermal relaxation models which are used to compute high enthalpy flows have to be validated with appropriate experimental data. (orig.)

Transient and steady-state flows in shock tunnels

Energy Technology Data Exchange (ETDEWEB)

Hannemann, K. (Deutsche Forschungsanstalt fuer Luft- und Raumfahrt e.V. (DLR), Goettingen (Germany)); Jacobs, P.A. (Queensland Univ., Brisbane (Australia). Dept. of Mechanical Engineering); Thomas, A.; McIntyre, T.J. (Queensland Univ., Brisbane, QLD. (Australia). Dept. of Physics)

1999-01-01

Due to the difficulty of measuring all necessary flow quantities in the nozzle reservoir and the test section of high enthalpy shock tunnels, indirect computational methods are necessary to estimate the required flow parameters. In addition to steady state flow computations of the nozzle flow and the flow past wind tunnel models it is necessary to investigate the transient flow in the facility in order to achieve a better understanding of its performance. These transient effects include the nozzle starting flow, the interaction of the shock tube boundary layers and the reflected shock, thermal losses in the shock reflection region and the developing boundary layers in the expanding section of the nozzle. Additionally, the nonequilibrium chemical and thermal relaxation models which are used to compute high enthalpy flows have to be validated with appropriate experimental data. (orig.)

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)

Numerical Simulation of Steady State Conduction Heat Transfer During the Solidification of Aluminum Casting in Green Sand Mould

Directory of Open Access Journals (Sweden)

Victor ANJO

2012-08-01

Full Text Available The solidification of molten metal during the casting process involves heat transfer from the molten metal to the mould, then to the atmosphere. The mechanical properties and grain size of metals are determined by the heat transfer process during solidification. The aim of this study is to numerically stimulate the steady conduction heat transfer during the solidification of aluminum in green sand mould using finite difference analysis 2D. The properties of materials used are industrial AI 50/60 AFS green sand mould, pure aluminum and MATLAB 7.0.1. for the numerical simulation. The method includes; the finite difference analysis of the heat conduction equation in steady (Laplace’s and transient states and using MATLAB to numerically stimulate the thermal flow and cooling curve. The results obtained are: the steady state thermal flow in 2D and transient state cooling curve of casting. The results obtain were consider relevant in the control of the grain size and mechanical properties of the casting.

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)

Steady state investigation on neutronics of a molten salt reactor considering the flow effect of fuel salt

Institute of Scientific and Technical Information of China (English)

ZHANG Da-Lin; QIU Sui-Zheng; LIU Chang-Liang; SU Guang-Hui

2008-01-01

The Molten Salt Reactor (MSR),one of the‘Generation Ⅳ'concepts,is a liquid-fuel reactor,which is different from the conventional reactors using solid fissile materials due to the flow effect of fuel salt.The study on its neutronice considering the fuel salt flow,which is the base of the thermal-hydraulic calculation and safety analysis,must be done.In this paper,the theoretical model on neutronics under steady condition for a single-liquid-fueled MSR is conducted and calculated by numerical method.The neutronics model consists of two group neutron diffusion equations for fast and thermal neutron fluxes,and balance equations for six-group delayed neutron precursors considering the flow effect of fuel salt. The spatial discretization of the above models is based on the finite volume method,and the discretization equations are computed by the source iteration method.The distributions of neutron fluxes and the distributions of the delayed neutron precursors in the core are obtained.The numerical calculated results show that,the fuel salt flow has little effect on the distribution of fast and thermal neutron fluxes and the effective multiplication factor;however,it affects the distribution of the delayed neutron precursors significantly,especially the long-lived one.In addition,it could be found that the delayed neutron precursors influence the nentronics slightly under the steady condition.

Steady state investigation on neutronics of a molten salt reactor considering the flow effect of fuel salt

International Nuclear Information System (INIS)

Zhang Dalin; Qiu Suizheng; Su Guanghui; Liu Changliang

2008-01-01

The Molten Salt Reactor (MSR), one of the 'Generation IV' concepts, is a liquid-fuel reactor, which is different from the conventional reactors using solid fissile materials due to the flow effect of fuel salt. The study on its neutronics considering the fuel salt flow, which is the base of the thermal-hydraulic calculation and safety analysis, must be done. In this paper, the theoretical model on neutronics under steady condition for a single-liquid-fueled MSR is conducted and calculated by numerical method. The neutronics model consists of two group neutron diffusion equations for fast and thermal neutron fluxes, and balance equations for six-group delayed neutron precursors considering the flow effect of fuel salt. The spatial discretization of the above models is based on the finite volume method, and the discretization equations are computed by the source iteration method. The distributions of neutron fluxes and the distributions of the delayed neutron precursors in the core are obtained. The numerical calculated results show that, the fuel salt flow has little effect on the distribution of fast and thermal neutron fluxes and the effective multiplication factor; however, it affects the distribution of the delayed neutron precursors significantly, especially the long-lived one. In addition, it could be found that the delayed neutron precursors influence the neutronics slightly under the steady condition. (authors)

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)

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)

Safety analysis of RSG-GAS Silicide core using one line cooling system

International Nuclear Information System (INIS)

Endiah-Puji-Hastuti

2003-01-01

In the frame of minimizing the operation-cost, operation mode using one line cooling system is being evaluated. Maximum reactor has been determined and to continuing this program, steady state and transient analysis were done. The analysis was done by means of a core thermal hydraulic code, COOLOD-N, and PARET. The codes solves core thermal hydraulic equation at steady state conditions and transient, respectively. By using silicide core data and coast down flow rate as the input, thermal hydraulics parameters such as fuel cladding and fuel meat temperatures as well as safety margin against flow instability were calculated. Imposing the safety criteria to the results of steady state and transient analysis, maximum permissible power for this operation was obtained as much as 17.1 MW

The code DYN3DR for steady-state and transient analyses of light water reactor cores with Cartesian geometry

International Nuclear Information System (INIS)

Grundmann, U.

1995-11-01

The code DYN3D/M2 was developed for 3-dimensional steady-state and transient analyses of reactor cores with hexagonal fuel assemblies. The neutron kinetics of the new version DYN3DR is based on a nodal method for the solution of the 3-dimensional 2-group neutron diffusion equation for Cartesian geometry. The thermal-hydraulic model FLOCAL simulating the two phase flow of coolant and the fuel rod behaviour is used in the two versions. The fundamentals for the solution of the neutron diffusion equations in DYN3DR are described. The 3-dimensional NEACRP benchmarks for rod ejections in LWR with quadratic fuel assemblies were calculated and the results were compared with the published solutions. The developed algorithm for neutron kinetics are suitable for using parallel processing. The behaviour of speed-up versus the number of processors is demonstrated for calculations of a static neutron flux distribution using a workstation with 4 processors. (orig.) [de

New Tore Supra steady state operating scenario

International Nuclear Information System (INIS)

Martin, G.; Parlange, F.; van Houtte, D.; Wijnands, T.

1995-01-01

This document deals with plasma control in steady state conditions. A new plasma control systems enabling feedback control of global plasma equilibrium parameters has been developed. It also enables to operate plasma discharge in steady state regime. (TEC). 4 refs., 5 figs

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.

Steady state neutral beam injector

International Nuclear Information System (INIS)

Mattoo, S.K.; Bandyopadhyay, M.; Baruah, U.K.; Bisai, N.; Chakbraborty, A.K.; Chakrapani, Ch.; Jana, M.R.; Bajpai, M.; Jaykumar, P.K.; Patel, D.; Patel, G.; Patel, P.J.; Prahlad, V.; Rao, N.V.M.; Rotti, C.; Singh, N.P.; Sridhar, B.

2000-01-01

Learning from operational reliability of neutral beam injectors in particular and various heating schemes including RF in general on TFTR, JET, JT-60, it has become clear that neutral beam injectors may find a greater role assigned to them for maintaining the plasma in steady state devices under construction. Many technological solutions, integrated in the present day generation of injectors have given rise to capability of producing multimegawatt power at many tens of kV. They have already operated for integrated time >10 5 S without deterioration in the performance. However, a new generation of injectors for steady state devices have to address to some basic issues. They stem from material erosion under particle bombardment, heat transfer > 10 MW/m 2 , frequent regeneration of cryopanels, inertial power supplies, data acquisition and control of large volume of data. Some of these engineering issues have been addressed to in the proposed neutral beam injector for SST-1 at our institute; the remaining shall have to wait for the inputs of the database generated from the actual experience with steady state injectors. (author)

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

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.

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-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)

Measurement of non-steady-state free fatty acid turnover

International Nuclear Information System (INIS)

Jensen, M.D.; Heiling, V.; Miles, J.M.

1990-01-01

The accuracy of non-steady-state equations for measuring changes in free fatty acid rate of appearance (Ra) is unknown. In the present study, endogenous lipolysis (traced with [ 14 C]-linoleate) was pharmacologically suppressed in six conscious mongrel dogs. A computer-responsive infusion pump was then used to deliver an intravenous oleic acid emulsion in both constant and linear gradient infusion modes. Both non-steady-state equations with various effective volumes of distribution (V) and steady-state equations were used to measure oleate Ra [( 14 C]oleate). Endogenous lipolysis did not change during the experiment. When oleate Ra increased in a linear gradient fashion, only non-steady-state equations with a large (150 ml/kg) V resulted in erroneous values (9% overestimate, P less than 0.05). In contrast, when oleate Ra decreased in a similar fashion, steady-state and standard non-steady-state equations (V = plasma volume = 50 ml/kg) overestimated total oleate Ra (18 and 7%, P less than 0.001 and P less than 0.05, respectively). Overall, non-steady-state equations with an effective V of 90 ml/kg (1.8 x plasma volume) allowed the most accurate estimates of oleate Ra

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.

ARCADIAR - A New Generation of Coupled Neutronics / Core Thermal- Hydraulics Code System at AREVA NP

International Nuclear Information System (INIS)

Curca-Tivig, Florin; Merk, Stephan; Pautz, Andreas; Thareau, Sebastien

2007-01-01

Anticipating future needs of our customers and willing to concentrate synergies and competences existing in the company for the benefit of our customers, AREVA NP decided in 2002 to develop the next generation of coupled neutronics/ core thermal-hydraulic (TH) code systems for fuel assembly and core design calculations for both, PWR and BWR applications. The global CONVERGENCE project was born: after a feasibility study of one year (2002) and a conceptual phase of another year (2003), development was started at the beginning of 2004. The present paper introduces the CONVERGENCE project, presents the main feature of the new code system ARCADIA R and concludes on customer benefits. ARCADIA R is designed to meet AREVA NP market and customers' requirements worldwide. Besides state-of-the-art physical modeling, numerical performance and industrial functionality, the ARCADIA R system is featuring state-of-the-art software engineering. The new code system will bring a series of benefits for our customers: e.g. improved accuracy for heterogeneous cores (MOX/ UOX, Gd...), better description of nuclide chains, and access to local neutronics/ thermal-hydraulics and possibly thermal-mechanical information (3D pin by pin full core modeling). ARCADIA is a registered trademark of AREVA NP. (authors)

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)

Reactor kinetics - pulse and steady state

Energy Technology Data Exchange (ETDEWEB)

Estes, B F; Morris, F M [Sandia Laboratories (United States)

1974-07-01

An analytical model has been developed which couples the nuclear and thermal characteristics of the Annular Core Pulse Reactor (ACPR) into a solution which describes both the neutron kinetics of the reactor and the temperature behavior of a fuel-moderator element. The model describes both pulse and steady state operations. This paper describes the important aspects of the reactor, the fuel- moderator elements, the neutron kinetic equations of the reactor, and the time-temperature behavior of a fuel-moderator element that is being subjected to the maximum power density in the core. The parameters which are utilized in the equations are divided into two classes, those that can be measured directly and those that are assumed to be known (each is described briefly). Some of the solutions which demonstrate the versatility of the analytical model are described. (author)

On the validity of travel-time based nonlinear bioreactive transport models in steady-state flow.

Science.gov (United States)

Sanz-Prat, Alicia; Lu, Chuanhe; Finkel, Michael; Cirpka, Olaf A

2015-01-01

Travel-time based models simplify the description of reactive transport by replacing the spatial coordinates with the groundwater travel time, posing a quasi one-dimensional (1-D) problem and potentially rendering the determination of multidimensional parameter fields unnecessary. While the approach is exact for strictly advective transport in steady-state flow if the reactive properties of the porous medium are uniform, its validity is unclear when local-scale mixing affects the reactive behavior. We compare a two-dimensional (2-D), spatially explicit, bioreactive, advective-dispersive transport model, considered as "virtual truth", with three 1-D travel-time based models which differ in the conceptualization of longitudinal dispersion: (i) neglecting dispersive mixing altogether, (ii) introducing a local-scale longitudinal dispersivity constant in time and space, and (iii) using an effective longitudinal dispersivity that increases linearly with distance. The reactive system considers biodegradation of dissolved organic carbon, which is introduced into a hydraulically heterogeneous domain together with oxygen and nitrate. Aerobic and denitrifying bacteria use the energy of the microbial transformations for growth. We analyze six scenarios differing in the variance of log-hydraulic conductivity and in the inflow boundary conditions (constant versus time-varying concentration). The concentrations of the 1-D models are mapped to the 2-D domain by means of the kinematic (for case i), and mean groundwater age (for cases ii & iii), respectively. The comparison between concentrations of the "virtual truth" and the 1-D approaches indicates extremely good agreement when using an effective, linearly increasing longitudinal dispersivity in the majority of the scenarios, while the other two 1-D approaches reproduce at least the concentration tendencies well. At late times, all 1-D models give valid approximations of two-dimensional transport. We conclude that the

CFD modeling and thermal-hydraulic analysis for the passive decay heat removal of a sodium-cooled fast reactor

International Nuclear Information System (INIS)

Hung, T.C.; Dhir, V.K.; Chang, J.C.; Wang, S.K.

2011-01-01

Research highlights: → The COOLOD/N2 and PARET/ANL codes were used for a steady-state thermal-hydraulic and safety analysis of the 2 MW TRIGA MARK II reactor located at the Nuclear Studies Center of Maamora (CENM), Morocco. → The main objective of this study is to ensure the safety margins of different safety related parameters by steady-state calculations at full power level (2 MW). → The most important conclusion is that all obtained values of DNBR, fuel center and surface temperature, cladding surface temperature and coolant temperature across the hottest channel are largely far to compromise safety of the reactor. - Abstract: In this study, a pool-typed design similar to sodium-cooled fast reactor (SFR) of the fourth generation reactors has been modeled using CFD simulations to investigate the characteristics of a passive mechanism of Shutdown Heat Removal System (SHRS). The main aim is to refine the reactor pool design in terms of temperature safety margin of the sodium pool. Thus, an appropriate protection mechanism is maintained in order to ensure the safety and integrity of the reactor system during a shutdown mode without using any active heat removal system. The impacts on the pool temperature are evaluated based on the following considerations: (1) the aspect ratio of pool diameter to depth, (2) the values of thermal emissivity of the surface materials of reactor and guard vessels, and (3) innerpool liner and core periphery structures. The computational results show that an optimal pool design in geometry can reduce the maximum pool temperature down to ∼551 o C which is substantially lower than ∼627 o C as calculated for the reference case. It is also concluded that the passive Reactor Air Cooling System (RACS) is effective in removing decay heat after shutdown. Furthermore, thermal radiation from the surface of the reactor vessel is found to be important; and thus, the selection of the vessel surface materials with a high emissivity would be a

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)

Steady state method to determine unsaturated hydraulic conductivity at the ambient water potential

Science.gov (United States)

HUbbell, Joel M.

2014-08-19

The present invention relates to a new laboratory apparatus for measuring the unsaturated hydraulic conductivity at a single water potential. One or more embodiments of the invented apparatus can be used over a wide range of water potential values within the tensiometric range, requires minimal laboratory preparation, and operates unattended for extended periods with minimal supervision. The present invention relates to a new laboratory apparatus for measuring the unsaturated hydraulic conductivity at a single water potential. One or more embodiments of the invented apparatus can be used over a wide range of water potential values within the tensiometric range, requires minimal laboratory preparation, and operates unattended for extended periods with minimal supervision.

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

High fidelity thermal-hydraulic analysis using CFD and massively parallel computers

International Nuclear Information System (INIS)

Weber, D.P.; Wei, T.Y.C.; Brewster, R.A.; Rock, Daniel T.; Rizwan-uddin

2000-01-01

Thermal-hydraulic analyses play an important role in design and reload analysis of nuclear power plants. These analyses have historically relied on early generation computational fluid dynamics capabilities, originally developed in the 1960s and 1970s. Over the last twenty years, however, dramatic improvements in both computational fluid dynamics codes in the commercial sector and in computing power have taken place. These developments offer the possibility of performing large scale, high fidelity, core thermal hydraulics analysis. Such analyses will allow a determination of the conservatism employed in traditional design approaches and possibly justify the operation of nuclear power systems at higher powers without compromising safety margins. The objective of this work is to demonstrate such a large scale analysis approach using a state of the art CFD code, STAR-CD, and the computing power of massively parallel computers, provided by IBM. A high fidelity representation of a current generation PWR was analyzed with the STAR-CD CFD code and the results were compared to traditional analyses based on the VIPRE code. Current design methodology typically involves a simplified representation of the assemblies, where a single average pin is used in each assembly to determine the hot assembly from a whole core analysis. After determining this assembly, increased refinement is used in the hot assembly, and possibly some of its neighbors, to refine the analysis for purposes of calculating DNBR. This latter calculation is performed with sub-channel codes such as VIPRE. The modeling simplifications that are used involve the approximate treatment of surrounding assemblies and coarse representation of the hot assembly, where the subchannel is the lowest level of discretization. In the high fidelity analysis performed in this study, both restrictions have been removed. Within the hot assembly, several hundred thousand to several million computational zones have been used, to

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

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

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)

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.

KIM, Steady-State Transport for Fixed Source in 2-D Thermal Reactor by Monte-Carlo

International Nuclear Information System (INIS)

Cupini, E.; De Matteis, A.; Simonini, R.

1980-01-01

1 - Description of problem or function: KIM (K-infinite Monte Carlo) is a program which solves the steady-state linear transport equation for a fixed-source problem (or, by successive fixed-source runs, for the eigenvalue problem) in a two-dimensional infinite thermal reactor lattice. The main quantities computed in some broad energy groups are the following: - Fluxes and cross sections averaged over the region (i.e. a space portion that can be unconnected but contains everywhere the same homogeneous material), grouping of regions, the whole element. - Average absorption and fission rates per nuclide. - Average flux, absorption and production distributions versus energy. 2 - Method of solution: Monte Carlo simulation is used by tracing particle histories from fission birth down through the resonance region until absorption in the thermal range. The program is organised in three sections for fast, epithermal and thermal simulation, respectively; each section implements a particular model for both numerical techniques and cross section representation (energy groups in the fast section, groups or resonance parameters in the epithermal section, points in the thermal section). During slowing down (energy above 1 eV) nuclei are considered as stationary, with the exception of some resonance nuclei whose spacing between resonances is much greater than the resonance width. The Doppler broadening of s-wave resonances of these nuclides is taken into account by computing cross sections at the current neutron energy and at the temperature of the nucleus hit. During thermalization (energy below 1 eV) the thermal motion of some nuclides is also considered, by exploiting scattering kernels provided by the library for light water, heavy water and oxygen at several temperatures. KIM includes splitting and Russian roulette. A characteristic feature of the program is its approach to the lattice geometry. In fact, besides the usual continuous treatment of the geometry using the well

The effect of time-dependent coupling on non-equilibrium steady states

DEFF Research Database (Denmark)

Cornean, Horia; Neidhardt, Hagen; Zagrebnov, Valentin

Consider (for simplicity) two one-dimensional semi-infinite leads coupled to a quantum well via time dependent point interactions. In the remote past the system is decoupled, and each of its components is at thermal equilibrium. In the remote future the system is fully coupled. We define...... and compute the non equilibrium steady state (NESS) generated by this evolution. We show that when restricted to the subspace of absolute continuity of the fully coupled system, the state does not depend at all on the switching. Moreover, we show that the stationary charge current has the same invariant...

The effect of time-dependent coupling on non-equilibrium steady states

DEFF Research Database (Denmark)

Cornean, Horia; Neidhardt, Hagen; Zagrebnov, Valentin A.

2009-01-01

Consider (for simplicity) two one-dimensional semi-infinite leads coupled to a quantum well via time dependent point interactions. In the remote past the system is decoupled, and each of its components is at thermal equilibrium. In the remote future the system is fully coupled. We define...... and compute the non equilibrium steady state (NESS) generated by this evolution. We show that when restricted to the subspace of absolute continuity of the fully coupled system, the state does not depend at all on the switching. Moreover, we show that the stationary charge current has the same invariant...

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 coupled neutronic/thermal-hydraulic tool with multi-scale capabilities and applications to HPLWR core analysis

International Nuclear Information System (INIS)

Monti, Lanfranco; Starflinger, Joerg; Schulenberg, Thomas

2011-01-01

Highlights: → Advanced analysis and design techniques for innovative reactors are addressed. → Detailed investigation of a 3 pass core design with a multi-physics-scales tool. → Coupled 40-group neutron transport/equivalent channels TH core analyses methods. → Multi-scale capabilities: from equivalent channels to sub-channel pin-by-pin study. → High fidelity approach: reduction of conservatism involved in core simulations. - Abstract: The High Performance Light Water Reactor (HPLWR) is a thermal spectrum nuclear reactor cooled and moderated with light water operated at supercritical pressure. It is an innovative reactor concept, which requires developing and applying advanced analysis tools as described in the paper. The relevant water density reduction associated with the heat-up, together with the multi-pass core design, results in a pronounced coupling between neutronic and thermal-hydraulic analyses, which takes into account the strong natural influence of the in-core distribution of power generation and water properties. The neutron flux gradients within the multi-pass core, together with the pronounced dependence of water properties on the temperature, require to consider a fine spatial resolution in which the individual fuel pins are resolved to provide precise evaluation of the clad temperature, currently considered as one of the crucial design criteria. These goals have been achieved considering an advanced analysis method based on the usage of existing codes which have been coupled with developed interfaces. Initially neutronic and thermal-hydraulic full core calculations have been iterated until a consistent solution is found to determine the steady state full power condition of the HPLWR core. Results of few group neutronic analyses might be less reliable in case of HPLWR 3-pass core than for conventional LWRs because of considerable changes of the neutron spectrum within the core, hence 40 groups transport theory has been preferred to the

Methods and Models for the Coupled Neutronics and Thermal-Hydraulics Analysis of the CROCUS Reactor at EFPL

Directory of Open Access Journals (Sweden)

A. Rais

2015-01-01

Full Text Available In order to analyze the steady state and transient behavior of the CROCUS reactor, several methods and models need to be developed in the areas of reactor physics, thermal-hydraulics, and multiphysics coupling. The long-term objectives of this project are to work towards the development of a modern method for the safety analysis of research reactors and to update the Final Safety Analysis Report of the CROCUS reactor. A first part of the paper deals with generation of a core simulator nuclear data library for the CROCUS reactor using the Serpent 2 Monte Carlo code and also with reactor core modeling using the PARCS code. PARCS eigenvalue, radial power distribution, and control rod reactivity worth results were benchmarked against Serpent 2 full-core model results. Using the Serpent 2 model as reference, PARCS eigenvalue predictions were within 240 pcm, radial power was within 3% in the central region of the core, and control rod reactivity worth was within 2%. A second part reviews the current methodology used for the safety analysis of the CROCUS reactor and presents the envisioned approach for the multiphysics modeling of the reactor.

Steady-state fission gas behavior in uranium-plutonium-zirconium metal fuel elements

International Nuclear Information System (INIS)

Steele, W.G.; Wazzan, A.R.; Okrent, D.

1989-01-01

An analysis of fission gas release and induced swelling in steady state irradiated U-Pu-Zr metal fuels is developed and computer coded. The code is used to simulate, with fair success, some gas release and induced swelling data obtained under the IFR program. It is determined that fuel microstructural changes resulting from zirconium migration, anisotropic swelling, and thermal variations are major factors affecting swelling and gas release behavior. (orig.)

The primary results for the mixed carbon material used for high flux steady-state tokamak operation in China

International Nuclear Information System (INIS)

Guo, Q.G.; Li, J.G.; Zhai, G.T.; Liu, L.; Song, J.R.; Zhang, L.F.; He, Y.X.; Chen, J.L.

2001-01-01

Several types of carbon mixed materials have been developed in China to be used for high flux steady-state tokamak operation. Performance evaluation of these materials is necessary to determine their applicability as PFCs for high flux steady state. This paper describes the primary results of carbon mixed materials and the effects of dopants on properties are primarily discussed. Test results reveal that bulk boronized graphite has excellent physical and mechanical properties while their thermal conductivity is no more than 73 W/m K due to the formation of a uniform boron-carbon solid solution. In case of multi-element doped graphite, titanium dopant or a decreased boron content is favorable to enhance thermal conductivity. A kind of doped graphite has been developed with thermal conductivity as high as 278 W/m K by optimizing the compositions. Correlations among compositions, microstructure and properties of such doped graphite are discussed

Modular first wall concept for steady state operation

International Nuclear Information System (INIS)

Kotzlowski, H.E.

1981-01-01

On the basis of the limiter design proposed for ZEPHYR a first wall concept has been developed which can also be used as a large area limiter, heat shield or beam pump. Its specific feature is the thermal contact of the wall armour elements with the water-cooled base plates. The combination of radiation and contact cooling, compared with radiation only, helps to lower the steady state temperatures of the first wall by approximately 50 % and to reduce the cooling-time between discharges. Particulary the lower wall temperature give a larger margin for additional heating of the wall by plasma disruption or neutral beams until excessive erosion or damage of the armour takes place

Steady-State Process Modelling

DEFF Research Database (Denmark)

Cameron, Ian; Gani, Rafiqul

2011-01-01

illustrate the “equation oriented” approach as well as the “sequential modular” approach to solving complex flowsheets for steady state applications. The applications include the Williams-Otto plant, the hydrodealkylation (HDA) of toluene, conversion of ethylene to ethanol and a bio-ethanol process....

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