Icone11-36407 Thermohydraulic Research for the Core of the Brest-Od-300 Reactor

Smirnov, V. P.; Filin, A. I.; Sila-Novitskiy, A. G.; Leonov, V. P.; Жуков, А. В.; Efanov, A. D.; Сорокин, А. П.; Kuzina, Ju

doi:10.1299/jsmeicone.2003.172

Cited by 7 publications

(5 citation statements)

References 0 publications

Supporting

Mentioning

Contrasting

Order By: Relevance

“…Under normal conditions (NO), the cladding temperature limit was 620 °C. Under anticipated operational occurences (AOO) and design basis accidents (DBA), the cladding temperature limit was 750 °C, considering the international experience, especially the research of BREST-OD-300 (Smirnov et al, 2003). Under design extension conditions (DEC), the cladding temperature limit was 1,500 °C, which is the melting point of T91 (Shi, 2017;Shen et al, 2019).…”

Section: Safety Criteriamentioning

confidence: 99%

Safety analysis of representative accidental transients for a 100 MWe small modular LBE-cooled reactor

Chen¹,

Tang²,

Qian³

et al. 2023

Front. Energy Res.

View full text Add to dashboard Cite

Nuclear energy plays an important role in the green low-carbon energy. The lead-based reactor is one of the most popular small modular reactor types. The pre-conceptual design of a 100 MW(e) small modular lead-bismuth eutectic cooled reactor was developed by Shanghai Nuclear Engineering Research & Design Institute CO., LTD. to meet the market demand of advanced small-scale nuclear plant. Representative accidents were analyzed for the 100 MW(e) small modular lead-bismuth reactor using ATHLET/MOD 3. The result shows that under the design basis conditions, the performance of the reactor can meet the safety criteria with the protection system working normally. The coolant solidification in the primary circuit under these transients should be prevented. Under the design extension condition, the highest cladding temperature is much lower than the melting point. In the unprotected reactivity insertion condition, there may be a short time of local melting of the fuel, which also meet the acceptance criteria. The safety analysis demonstrated the passive safety of the design with the negative temperature coefficient, strong natural circulation capability, and the passive residual heat removal system.

show abstract

Section: Safety Criteriamentioning

confidence: 99%

Safety analysis of representative accidental transients for a 100 MWe small modular LBE-cooled reactor

Chen¹,

Tang²,

Qian³

et al. 2023

Front. Energy Res.

View full text Add to dashboard Cite

show abstract

“…In recent years, many research programs on breeder reactors have been running, including European Synchrotron Radiation Facility (ESRF) European program whose aim is to build a reactor for fast neutrons, in which coolant is sodium or Indian PFBR program, which aims to build a breeder reactor in which the coolant are other heavy metals. Lead is used as a coolant in the breeder reactors in the European research program ELSY and the Russian program, BREST-OD-300 (Smirnov et al , 2003). As the coolants in fast fission reactors or accelerator driven systems in use are liquid metals or alloys, such as potassium, sodium, lithium, lead, mercury, the eutectic alloy sodium–potassium (22 per cent Na + 78 per cent K) or lead-bismuth eutectic.…”

Section: Introductionmentioning

confidence: 99%

“…Many authors studied heat transfer to liquid metals in rod bundles of various tube arrangements. Smirnov et al (2003) carried out experiments and numerical simulations of thermal and hydraulic processes in the core of the lead-cooled BREST-OD-300 reactor. Data and correlations for tube bundles were examined by Mikityuk (2009).…”

Section: Introductionmentioning

confidence: 99%

Semi-empirical heat transfer correlations for turbulent tube flow of liquid metals

Taler¹

2018

HFF

View full text Add to dashboard Cite

Purpose The purpose of this paper is to develop new semi-empirical heat transfer correlations for turbulent flow of liquid metals in the tubes, and then to compare these correlations with the experimental data. The Prandtl and Reynolds numbers can vary in the ranges: 0.0001 ≤ Pr ≤ 0.1 and 3000 ≤ Re ≤ 106. Design/methodology/approach The energy conservation equation averaged by Reynolds was integrated using the universal velocity profile determined experimentally by Reichardt for the turbulent tube flow and four different models for the turbulent Prandtl number. Turbulent heat transfer in the circular tube was analyzed for a constant heat flux at the inner surface. Some constants in different models for the turbulent Prandtl number were adjusted to obtain good agreement between calculated and experimentally obtained Nusselt numbers. Subsequently, new correlations for the Nusselt number as a function of a Peclet number was proposed for different models of the turbulent Prandtl number. Findings The inclusion of turbulent Prandtl number greater than one and the experimentally determined velocity profile of the fluid in the tube while solving the energy conservation equation improved the compatibility of calculated Nusselt numbers, with Nusselt numbers determined experimentally. The correlations proposed in the paper have a sound theoretical basis and give Nusselt number values that are in good agreement with the experimental data. Research limitations/implications Heat transfer correlations proposed in this paper were derived assuming a constant heat flux at the inner surface of the tube. However, they can also be used for a constant wall temperature, as for the turbulent flow (Re > 3,000), the relative difference between the Nusselt number for uniform wall heat flux and uniform wall temperature is very low. Originality/value Unified, systematic approach to derive correlations for the Nusselt number for liquid metals was proposed in the paper. The Nusselt number was obtained from the solution of the energy conservation equation using the universal velocity profile and eddy diffusivity determined experimentally, and various models for the turbulent Prandtl number. Four different relationships for the Nusselt number proposed in the paper were compared with the experimental data.

show abstract

“…The conference was held on July 5-9, 2004 in Obninsk. Specialists from the laboratories of the Physics and Power-Engineering Institute and the Scientific-Research and Design Institute of Electrical Technology prepared the specifications for the standard problem on the basis of experiments designed to validate the thermohydraulics of fast reactors with inherent safety [1][2][3][4][5][6][7][8][9].The goal of the benchmark experiment was to analyze the thermohydraulic characteristics of a model assembly, which is nonuniform with respect to geometry and heat release, in a flow of a sodium-potassium alloy (22% sodium + 78% potassium) with the simulators arranged in a square in the presence of a spacing lattice in regimes with variable energyrelease subzones and with an assessment of the reliability and accuracy of the computational results obtained using thermohydraulic codes. The following experimental and computational parameters had to be compared: the coolant temperature in channels with nonuniform geometric and thermal conditions in an assembly, the surface temperature of a measuring fuel-element simulator on the heated section with nonuniform geometric and thermal conditions in the assembly, and the coolant velocity in the assembly cells surrounding the measuring simulator.…”

mentioning

confidence: 99%

mentioning

confidence: 99%

Analysis of a Benchmark Experiment on the Hydraulics and Heat Transfer in a Liquid-Metal-Cooled Assembly of Fuel-Element Simulators

2005

Self Cite

View full text Add to dashboard Cite

A comparative analysis of the experimental data, which were obtained in a benchmark experiment on the thermohydraulics of a model assembly of fuel-element simulators in a flow of sodium-potassium alloy, and calculations performed by specialists, using thermohydraulic codes, from different countries is performed. The model assembly consisted of 25 fuel-element simulators arranged in a square. Russian specialists used the BRS-TVS.R code to perform calculations of the benchmark experiment, Japanese specialists used SPIRAL and AQUA, Spanish specialists used FLUENT, Dutch specialists used STAR-CD, and South Korean specialists used MATRA and CFX. The following experimental and computational parameters were compared: the coolant temperature in the channels under nonuniform geometric and thermal conditions in the assembly, the surface temperature of the measuring fuel-element simulator on the heated section, and the coolant velocity in the assembly cells around the measuring simulator. Special attention was given to investigating the influence of the spacing lattice on the coolant velocity.A benchmark experiment on hydraulics and heat transfer in assemblies of fuel-element simulators cooled by a liquid metal was performed as part of the conference of the international working group on the thermohydraulics of advanced nuclear reactors of the International Association of Hydraulic Research. The conference was held on July 5-9, 2004 in Obninsk. Specialists from the laboratories of the Physics and Power-Engineering Institute and the Scientific-Research and Design Institute of Electrical Technology prepared the specifications for the standard problem on the basis of experiments designed to validate the thermohydraulics of fast reactors with inherent safety [1][2][3][4][5][6][7][8][9].The goal of the benchmark experiment was to analyze the thermohydraulic characteristics of a model assembly, which is nonuniform with respect to geometry and heat release, in a flow of a sodium-potassium alloy (22% sodium + 78% potassium) with the simulators arranged in a square in the presence of a spacing lattice in regimes with variable energyrelease subzones and with an assessment of the reliability and accuracy of the computational results obtained using thermohydraulic codes. The following experimental and computational parameters had to be compared: the coolant temperature in channels with nonuniform geometric and thermal conditions in an assembly, the surface temperature of a measuring fuel-element simulator on the heated section with nonuniform geometric and thermal conditions in the assembly, and the coolant velocity in the assembly cells surrounding the measuring simulator.The standard problem was open, i.e., both the initial and boundary conditions necessary for performing calculations and the results of experimental investigations were presented in the specification, the latter in tabular and graphical form [10].

show abstract

Icone11-36407 Thermohydraulic Research for the Core of the Brest-Od-300 Reactor

Cited by 7 publications

References 0 publications

Safety analysis of representative accidental transients for a 100 MWe small modular LBE-cooled reactor

Safety analysis of representative accidental transients for a 100 MWe small modular LBE-cooled reactor

Semi-empirical heat transfer correlations for turbulent tube flow of liquid metals

Analysis of a Benchmark Experiment on the Hydraulics and Heat Transfer in a Liquid-Metal-Cooled Assembly of Fuel-Element Simulators

Contact Info

Product

Resources

About