Calculated Substantiation of the Fixed-Bed Nuclear Reactor Core Thermal Reliability

Vestnik MEI ◽  
2021 ◽  
pp. 44-50
Author(s):  
Artem A. Berezin ◽  
◽  
Aleksandr A. Sataev ◽  
Denis I. Novikov ◽  
Oleg V. Khvoinov ◽  
...  
Keyword(s):  
Heat Transfer ◽  
Nuclear Reactor ◽  
Fixed Bed ◽  
Thermal Reliability ◽  
Flow Area ◽  
Pressurized Water ◽  
Fuel Rods ◽  
Spherical Packing ◽  
Study Results ◽  
Fuel Elements

The study is aimed at performing calculated substantiation of the thermal reliability of a pressurized water reactor (PWR) with spherical fuel elements, commonly known as a fixed-bed nuclear reactor, and determining the most advantageous design of the spherical packing from the view point of thermal hydraulics. Spherical fuel elements have a number of advantages over cylindrical fuel rods; in particular, they feature better retention of fission products, enhanced nuclear safety (due to a high melting temperature of ceramic materials), more intense heat transfer due to increased coolant flow turbulence, and a reduced influence of thermal cyclic loads on fuel elements. To confirm these statements, a thermohydraulic calculation of the KLT-40S type reactor with a modified intra-channel filling of fuel assemblies (FAs) consisting of spherical fuel elements was carried out. To determine the optimal spherical filling, two types of spherical packing were calculated, with three different diameters of spherical fuel elements. In the course of the calculation, solutions to the following issues were proposed: how to take into account the channels flow area variability, how to calculate the Reynolds number for channels of a given shape, what formulas should be used to determine the Nusselt number, and how to determine the hydraulic resistance in the channels. As a result of the calculation, data on the following main thermohydraulic characteristics have been obtained: surface heat flux density, heat transfer coefficient, maximum fuel temperature, and hydraulic losses. These results were compared with the results of calculations for cylindrical fuel rods. The obtained results demonstrate the advantage of spherical fuel elements over cylindrical fuel rods in a number of basic parameters, which gives prospects for further study of the use of spherical fuel elements in reactors of this type. The obtained study results can be applied in designing reactor plants of low and medium capacity, as well as in modernizing the existing reactor plants.

scholarly journals Effect of Flow Blockage on the Coolability during Reflood in a 2×2 Rod Bundle

2014 ◽  
Vol 2014 ◽  
pp. 1-9 ◽  
Author(s):  
Kihwan Kim ◽  
Byung-Jae Kim ◽  
Young-Jung Youn ◽  
Hae-Seob Choi ◽  
Sang-Ki Moon ◽  
...  
Keyword(s):  
Heat Transfer ◽  
Test Facility ◽  
Pressurized Water Reactor ◽  
Two Phase ◽  
Flow Area ◽  
Pressurized Water ◽  
Fuel Rods ◽  
Rod Bundle ◽  
Loss Of Coolant Accident ◽  
Two Phase Heat Transfer

During the reflood phase of a large-break loss-of-coolant accident (LBLOCA) in a pressurized-water reactor (PWR), the fuel rods can be ballooned or rearranged owing to an increase in the temperature and internal pressure of the fuel rods. In this study, an experimental study was performed to understand the thermal behavior and effect of the ballooned region on the coolability using a 2 × 2 rod bundle test facility. The electrically heated rod bundle was used and the ballooning shape of the rods was simulated by superimposing hollow sleeves, which have a 90% blockage ratio. Forced reflood tests were performed to examine the transient two-phase heat transfer behavior for different reflood rates and rod powers. The droplet behaviors were also investigated by measuring the velocity and size of droplets near the blockage region. The results showed that the heat transfer was enhanced in the downstream of the blockage region, owing to the reduced flow area of the subchannel, intensification of turbulence, and deposition of the droplet.

Study on the Viability of the Recycling by Electric Arc Melting of Zirconium Alloys Scraps Aiming the Scalability of the Process

2018 ◽  
Vol 930 ◽  
pp. 495-500
Author(s):  
Cristiano Stefano Mucsi ◽  
L.A.M. dos Reis ◽  
Maurilio Pereira Gomes ◽  
L.A.T. Pereira ◽  
Jesualdo Luiz Rossi
Keyword(s):  
Nuclear Reactor ◽  
Zirconium Alloys ◽  
Vacuum Arc ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Vacuum Arc Remelting ◽  
Arc Melting ◽  
Vacuum Arc Furnace ◽  
Fuel Elements

Turning chips of zirconium alloys are produced in large quantities during the machining of alloy rods for the fabrication of the end plugs for the Pressurized Water Reactor (PWR) fuel elements parts of Angra II nuclear reactor (Brazil – Rio de Janeiro). This paper presents a study on the search for an efficient way for the cleaning, quality control and Vacuum Arc Remelting (VAR) of pressed zirconium alloys chips to produce a material viable to be used in the production of the fuel rod end plugs. The process starts with cutting oil clean out. The first step in this process consists in soaking a bunch of chips in clean water, to remove soluble cutting oils, followed by an alkaline degreasing bath and a wash with a high-pressure flow of water. Drying is performed by a flux of warm air. The oil free chips are then subjected to a magnet in order to detect and collect any magnetic material, essentially ferrous, that may be present in the original chips. Samples of the material are collected and then melted in a small non consumable electrode vacuum arc furnace for evaluation by Energy Dispersive X-ray Fluorescence Spectrometry (EDXRFS) in order to define the quality of the chips. The next step consists in the 15 ton hydraulic pressing the chips in a die with 40 mm square section and 500 mm long, producing an electrode with 20% of the Zircaloy bulk density. The electrode was finally melted in a laboratory scale modified VAR furnace located at the CCTM–IPEN, producing 0.8 kg ingots. The authors conclude that the samples obtained from the fuel element industry can be melting in a VAR furnace, modified to accommodate low density electrodes, allowing a reduction up to 40 times the original storage volume, however, it is necessary to remelt the ingots to correct their composition in order to recycle the original zirconium alloys chips. in a process to reduce volume and allow the reutilization of valuable Zircaloy scraps.

Heat-Transfer and Flow-Friction Characteristics of Crossed-Rod Matrices

1960 ◽  
Vol 82 (3) ◽  
pp. 199-213 ◽  
Author(s):  
A. L. London ◽  
J. W. Mitchell ◽  
W. A. Sutherland
Keyword(s):  
Heat Transfer ◽  
Electrical Resistance ◽  
Heat Exchangers ◽  
Air Conditioning ◽  
Nuclear Reactor ◽  
Heat Transfer Surface ◽  
Friction Behavior ◽  
Flow Friction ◽  
Nuclear Reactor Fuel ◽  
Fuel Elements

The paper presents a continuation of the program on porous media heat-transfer and flow-friction behavior previously covered in References [2b] and [3b]. All the previous results of interest to the designer on woven-screen matrices and crossed-rod matrices of a random configuration are summarized here. In addition, new design results for the regular in-line and regular staggered crossed-rod-matrix configurations are reported. Matrices of the type considered here may find application as heat-transfer surface geometries for nuclear-reactor fuel elements, for electrical resistance heaters and for periodic-flow-type heat exchangers used for gas-turbine regenerators, and some air-conditioning applications.

scholarly journals Powering Down

2003 ◽  
Vol 125 (04) ◽  
pp. 46-48
Author(s):  
Harry Hutchinson
Keyword(s):  
Heat Transfer ◽  
Nuclear Reactor ◽  
Reactor Core ◽  
The United States ◽  
Nuclear Industry ◽  
Significant Feature ◽  
Pressurized Water ◽  
Safety Testing ◽  
Fuel Assemblies ◽  
Enriched Uranium

This article reviews that after a half century of safety testing for the nuclear industry, a key heat-transfer lab is losing its home. Columbia University’s Heat Transfer Research Facility has been the only place to go for key safety testing. Since the days of the Atoms for Peace program during the Eisenhower years, the lab has tested generations of nuclear reactor fuel assemblies. The lab’s clients over the years have included all the designers of pressurized water reactors in the United States and others from much of the world. The tests are primarily concerned with one small, but significant feature of a reactor core. A core contains as many as 3000 fuel assemblies, bundles of long, slender rods containing enriched uranium. Controlled fission among the bundles heats water to begin the series of heat-transfer cycles that send steam to the turbines that will drive generators.

scholarly journals EFFECT OF PERMEABILITY OF PEBLE BED ON HEAT TRANSFER IN THE CORE OF NUCLEAR REACTOR WITH HELIUM COOLANT

2017 ◽  
Vol 39 (4) ◽  
pp. 55-60
Author(s):  
A. A. Avramenko ◽  
N. P. Dmitrenko ◽  
М. M. Kovetskaya ◽  
Yu. Yu. Kovetskaya
Keyword(s):  
Heat Transfer ◽  
Mass Transfer ◽  
Temperature Distribution ◽  
Heat And Mass Transfer ◽  
Nuclear Reactor ◽  
Reactor Core ◽  
Geometric Parameters ◽  
The Core ◽  
Fuel Elements ◽  
Helium Coolant

Heat and mass transfer in a model of the core of a nuclear reactor with spherical fuel elements and a helium coolant was studied. The effect of permeability of the pebble bed zone and geometric parameters on the temperature distribution of the coolant in the reactor core is analyzed.  

A Review on Specific Features of Small and Medium Sized Nuclear Power Plants

2010 ◽  
Author(s):  
Salah Ud-din Khan ◽  
Minjun Peng ◽  
Muhammad Zubair ◽  
Shaowu Wang
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Reactor ◽  
Nuclear Reactors ◽  
Fixed Bed ◽  
Nuclear Industry ◽  
Energy Agency ◽  
Pressurized Water ◽  
Water Reactor ◽  
Energy Supply System

Due to global warming and high oil prices nuclear power is the most feasible solution for generating electricity. For the fledging nuclear power industry small and medium sized nuclear reactors (SMR’s) are instrumental for the development and demonstration of nuclear reactor technology. Due to the enhanced and outstanding safety features, these reactors have been considered globally. In this paper, first we have summarized the reactor design by considering some of the large nuclear reactor including advanced and theoretical nuclear reactor. Secondly, comparison between large nuclear reactors and SMR’s have been discussed under the criteria led by International Atomic Energy Agency (IAEA). Thirdly, a brief review about the design and safety aspects of some of SMR’s have been carried out. We have considered the specifications and parametric analysis of the reactors like: ABV which is the floating type integral Pressurized water reactor; Long life, Safe, Simple Small Portable Proliferation Resistance Reactor (LSPR) concept; Multi-Application Small Light Water Reactor (MASLWR) concept; Fixed Bed Nuclear Reactor (FBNR); Marine Reactor (MR-X) & Deep Sea Reactor (DR-X); Space Reactor (SP-100); Passive Safe Small Reactor for Distributed energy supply system (PSRD); System integrated Modular Advanced Reactor (SMART); Super, Safe, Small and Simple Reactor (4S); International Reactor Innovative and Secure (IRIS); Nu-Scale Reactor; Next generation nuclear power plant (NGNP); Small, Secure Transportable Autonomous Reactor (SSTAR); Power Reactor Inherently Safe Module (PRISM) and Hyperion Reactor concept. Finally we have point out some challenges that must be resolved in order to play an effective role in Nuclear industry.

Preliminary Evaluation of the Fixed Bed Nuclear Reactor Concept Using IAEA-INPRO Methodology

2004 ◽  
Author(s):  
Farhang Sefidvash
Keyword(s):  
Nuclear Reactor ◽  
Nuclear Reactors ◽  
Fixed Bed ◽  
Reactor Core ◽  
Primary Circuit ◽  
Preliminary Evaluation ◽  
Spent Fuel ◽  
Energy Agency ◽  
Reactor Technology ◽  
Fuel Elements

The Atomic Energy Agency (IAEA) through its INPRO Project has developed a methodology to evaluate the innovative nuclear reactors. The main objectives of INPRO are to help to ensure that nuclear energy is available in the 21st century in a sustainable manner; and bring together both technology holders and technology users to achieve desired innovations in nuclear reactors and nuclear fuel cycles which are to be acceptable to the public because they are economic, safe, proliferation resistant, sustainable, and having reduced environmental impact. Here is a preliminary application of this methodology (IAEA-TECDOC-1362) to evaluate the Fixed Bed Nuclear Reactor Concept (FBNR). Some of the characteristics of the proposed reactor are: The FBNR is based on pressurized light water reactor technology. It is a small, modular, and integrated primary circuit reactor. The fuel elements of FBNR are 8 mm diameter spherical uranium dioxide pellets cladded by zircaloy or made of compacted TRISO type fuel particles. The reactor core is suspended by the flow of water coolant. The stop in flow causes the fuel elements leave the reactor core by the force of gravity and fall into a passively cooled fuel chamber or even leave the reactor completely and become deposited in the spent fuel pool. It is an inherently safe and passively cooled reactor concept. FBNR in its advanced versions can use supercritical steam or helium gas as coolant, and utilize MOX or thorium fuel.

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