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A Review of Inverted Annular and Low Quality Film Boiling
Published in G. F. Hewitt, J. M. Delhaye, N. Zuber, Post-Dryout Heat Transfer, 2017
Costigan and Wade (1984) were able to visualize the inverted annular flow boiling phenomenon using dynamic neutron radiography. A low neutron energy beam from a materials testing reactor was used in conjunction with an internally cooled tube, which was directly heated to 600°C, and quenched by introducing water from the bottom or top. The neutron imaging system provided video recordings of the post-CHF flow regimes. It was noted that at low flows (Vin < 5 cm/s) the liquid downstream of the quench front moves as a liquid filament which breaks down into large drops or slugs. At higher velocities the more typical inverted annular film boiling regime was observed for upflow, while for downflow a falling film/liquid jet type of flow regime was observed. The measured downflow heat transfer coefficient was considerably lower than the value predicted from Bromley’s (1950) model.
Nuclear Fission Reactor
Published in C. K. Gupta, Materials in Nuclear Energy Applications, 1989
The first group is based on enriched uranium, water moderated and water cooled. As illustrated, mention may be made of the Materials Testing Reactor (MTR) at the National Reactor Testing station in Arco, ID. This reactor is perhaps the first one constructed primarily for the purpose of studying nuclear radiation effects on materials. The MTR fuel elements are heterogeneous plate-type elements, in contrast to cylindrical slugs in the graphite reactors. The fuel elements consist of slightly curved plates of an alloy (20% by weight) of highly enriched uranium in aluminum with a thin cladding of aluminum on both sides, leading to a sandwich type of construction. There are 18 such sandwich plates, about 0.29 cm apart, held together in a 60.96-cm long, box-like assembly, forming a fuel element. Both ends of the element are open to allow the cooling water to flow downward between the plates. This water also serves as the moderator, just as graphite does in the graphite reactors. The inner reflector is of beryllium, which is surrounded by a graphite reflector placed outside the water tank. The MTR provides numerous experimental spaces. Some go right up to the reactor through the beryllium reflector, whereas other terminate in the graphite zone. A thermal column of graphite blocks, penetrating one of the faces, provides low-energy neutrons for experimental purposes.
Control and Shielding Materials
Published in C. K. Gupta, Materials in Nuclear Energy Applications, 1989
The attenuation of gamma radiation and neutron leaks is achieved by shielding materials that surround the core. A distinction can be made between thermal shielding and biological shielding. Thermal shielding functions to avoid exposure of the entire shielding to the heat generated in the reactor and to attenuate radiations nearest to the core. These are made of materials of high density, good thermal conductivity, and high melting temperatures. Thermal shielding absorbs the high-energy gamma radiations and reduces the energy of the fast neutrons by inelastic collisions. These two types of radiation carry most of the energy leaking from the reactor core. The absorption of the radiations accompanying the capture of slowed-down neutrons produces a considerable amount of heat in the thermal shield. In power reactors the heat is removed by the coolant and so contributes to the available energy. In research reactors, on the other hand, the heat is not utilized. In a materials testing reactor, for example, the thermal shield consists of two thick layers of steel, separated by a space through which air is passed to serve as the coolant. The air is then discharged into the atmosphere. For low-power reactors the heat liberated in the shield may not be high and in such cases the introduction of a thermal shield may be unnecessary.
Numerical Simulations of Flow-Induced Deflections in MITR LEU Fuel Plate Due to Channel Size Disparity
Published in Nuclear Technology, 2023
Guanyi Wang, Cezary Bojanowski, Akshay Dave, David Jaluvka, Lin-Wen Hu, Erik Wilson
In the case of coolant velocity significantly lower than Miller’s critical velocity, the nonuniformity in the shape of the plate and channel boundaries may have an impact on the magnitude of the plate’s deflection. Once the velocity approaches the critical threshold, the plate may start to vibrate, assuming one or a combination of its modal shapes. The amplitude of these vibrations will become significantly larger than the assembly tolerances, and at some high velocity (significantly higher than the nominal operating velocity), the plate will fail plastically, assuming one or a combination of these modal shapes. This behavior has been confirmed recently by the Generic Test Plate Assembly experiments.[8] In general, Materials Testing Reactor (MTR) plate-type reactors operate at far less than Miller’s critical velocity, and thus vibrations of the fuel plate at significant magnitudes or collapse are not expected. Thus, characterization of plate deflection and the impact on coolant channels is analyzed.
Dimensional changes and thermal conductivity by annealing and its relation to the defect concentration and stored energy release of neutron-irradiated graphite
Published in Journal of Nuclear Science and Technology, 2019
The description of specimens and neutron irradiations, and the method for measurement of the dimensions and thermal conductivity have been provided in the literature [7]. The specimens are fine-grained isotropic graphite IG-110U (Toyo Tanso, Osaka, Japan) and ETP-10 (Ibiden, Gifu, Japan) with an average grain size of 20 µm and 40 µm and a room temperature thermal conductivity of 120 and 104 Wm−1K−1, respectively. The specimens were irradiated in the Japan Materials Testing Reactor (JMTR) at a design temperature of <200°C to the fluences 0.02 and 0.25 dpa (1.38 x 1023 and 1.92 × 1024 n/m2: E > 1 MeV), respectively. The specimens for the irradiation experiments had the shape of a rectangular bar 2 × 2 x 25 mm3 in size and a circular plate 8 mm in diameter by 2 mm in thickness. The rectangular bars were used for dimensional measurement and the circular plates for thermal conductivity measurement. The dimensional changes were obtained by measuring the specimen length before and after neutron irradiation using a conventional micrometer at room temperature with an accuracy of ± 1 μm. The dimensional change was determined by taking the average of three to ten specimens for each irradiation condition.
Evaluation of Carbide Fuel Property and Model Using Measurement Data from Early Experiments
Published in Nuclear Technology, 2018
In the United States, more than 470 MC fuel rods were irradiated in EBR-II using sodium or helium bonds to stainless steel cladding, and a peak fuel burnup of 20 at. % was achieved.7–11 The U.S.-Switzerland joint research program on MC fuel fabricated 66 pellet-type fuel pins and 25 sphere-pac pins, conducted irradiation tests in the U.S. Fast Flux Test Facility (FFTF), and reached a fuel burnup of over 8% with a peak linear power of ~80 kW/m (Refs. 12 and 13). The test fuel had a smeared density of 75% to 80% TD. In Japan, nine encapsulated carbide fuel pins were irradiated at the Japan Research Reactor No. 2 (JRR-2) and the Japan Materials Testing Reactor (JMTR) to a peak burnup of 4.7% fissions per initial metal atom (FIMA) at 42 to 64 kW/m without failure.14 Currently, India is operating the Fast Breeder Test Reactor with hyperstoichiometric plutonium-rich (70%) mixed uranium-plutonium carbide as the driver fuel. The fuel performance has been assessed by periodic postirradiation examination (PIE) of the fuel subassemblies for extension of life. More than 900 fuel pins of MK-I carbide fuel (0.7PuC + 0.3UC) have reached 155 GWd/t and one subassembly (61 pins) was taken to 165 GWd/t without a single failure.15,16