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Thermal Energy Production in Nuclear Power Plants
Published in Robert E. Masterson, Nuclear Reactor Thermal Hydraulics, 2019
Finally, because neutron cross sections are energy dependent, their values depend on the kinetic energy E of the incoming neutron. Hence when referring to a particular cross section, the values for σt, σs, σc, and σf are generally written as σt(E), σs(E) σc(E), and σf(E) when E is measured in eV. Normally, elastic scattering cross sections are small compared to fission and capture cross sections for most nuclear fuels. However, they are larger for carbon, light water, and heavy water. These materials, which are sometimes used to slow down fission neutrons, are also called neutron moderators. In many reactors, the neutron moderator also serves as the coolant.
Nuclear Energy Security
Published in Maria G. Burns, Managing Energy Security, 2019
The PHWR uses natural uranium in its unenriched form as fuel. It is mainly used in Canada and India, and represents about 12% of reaction worldwide. Heavy water, i.e., a type of water that comprises deuterium, a hydrogen isotope, is used for its cooling and neutron-moderation operations. One of its benefits is the ability to continue operations and fuel replenishment operations without the need to be shut down.
Nuclear Particles, Processes, and Reactions
Published in Robert E. Masterson, Introduction to Nuclear Reactor Physics, 2017
Most reactors (with the exception of fast reactors) require a material called a neutron moderator to slow down the fission neutrons. At lower energies, these neutrons can be absorbed more easily by the uranium and plutonium atoms, and this allows additional fission neutrons to be produced to sustain the chain reaction. However, neutron moderators can also absorb some of the incoming neutrons that are scattering off of them, and when neutron moderation becomes necessary, the material selected for the moderator should have a low absorption rate as well as a high scattering one. Hence, elemental hydrogen (which is found in the cooling water used in thermal water reactors) is an attractive moderating material because it requires fewer collisions than other materials to achieve a specific neutron energy loss (see Table 3.7). However, it also has a larger absorption cross section than another popular moderator (deuterium or heavy hydrogen) for thermal neutrons. Because of this, deuterium absorbs fewer neutrons than ordinary hydrogen does, and this is one of the reasons why heavy water or D2O (which reactors such as the Canadian CANDU reactor use) is sometimes preferred to light water or H2O when the enrichment of the nuclear fuel is low. Hence the ability of a material to scatter neutrons elastically must be compared to the ability of the same material to absorb these neutrons when selecting the optimum moderator for a specific fuel.* The physics of neutron moderation is discussed in Chapter 8 where the nuances of neutron slowing down theory are also discussed.
Reactor Performance and Safety Characteristics of Beryllium-Based Composite Moderators as Replacements for Graphite in mHTGRs
Published in Nuclear Technology, 2022
Veronica Karriem, Edward M. Duchnowski, Bin Cheng, Lance L. Snead, Jason R. Trelewicz, Nicholas R. Brown
In thermal spectrum nuclear reactors, the neutron moderator reduces the energies of neutrons generated from fission, which increases their probability of achieving a sustainable fission reaction. Modular high-temperature gas-cooled reactors (mHTGRs) using advanced gas reactor fuel are flexible systems that can drive high-efficiency electricity production and process heat applications.1 The moderator for the mHTGRs must have a high moderating power, low neutron absorption, and good thermal properties to withstand high temperatures. These systems use graphite as a moderator because of its high moderating ratio and reasonable thermal properties.2 However, graphite has unfavorable properties under irradiation, such as anisotropic dimensional changes with directional swelling, causing carbon atom displacements attributing to the radiation damage of the graphite.3 This damage can limit the reactor core lifetime. Additionally, irradiated graphite contributes to radioactive waste, which is exacerbated in the event of component replacement due to lifetime issues. These challenges have led to the investigation of advanced two-phase composite moderators as an alternative to graphite.
Tritium Content and Chemical Form in Nuclear Graphite from Molten Fluoride Salt Irradiations
Published in Fusion Science and Technology, 2020
Kieran Dolan, Guiqiu Zheng, David Carpenter, Steven Huang, Lin-Wen Hu
Lithium- and beryllium-bearing molten fluoride salts are under investigation as candidate coolant options for liquid- and solid-fueled advanced fission reactor designs,1,2 as well as a combined coolant and tritium breeding medium in a high-field fusion device.3 In both applications, mitigating environmental release of tritium generated from neutron irradiation of the molten salt is an important aspect of a commercial power plant’s safety case. Developing a strategy to prevent tritium release in the proposed designs first requires a thorough understanding of the expected tritium transport mechanisms and resulting tritium distributions in molten fluoride salt and other key materials. For the fission applications with a thermal neutron spectrum, nuclear graphite is a commonly proposed neutron moderator because of its extensive experience base and chemical compatibility with molten fluorides. Therefore, it is important to examine the interaction between tritium and graphite in the molten salt environment to successfully predict tritium transport behavior in future reactor designs.
Development and potential of composite moderators for elevated temperature nuclear applications
Published in Journal of Asian Ceramic Societies, 2022
Lance L Snead, David Sprouster, Bin Cheng, Nick Brown, Caen Ang, Edward M Duchnowski, Xunxiang Hu, Jason Trelewicz
As the in service thermal and mechanical property requirement of an entrained phase are relatively unimportant (see Figure 6) material selection is relatively straight forward focusing on materials processing issues, benefit to overall moderation with minimal neutron scavenging, and susceptibility to environmental degradation. In this application environment effects include long-term thermal stability under irradiation and the potential interaction of the entrained phase with the host matrix. An abbreviated list of candidate entrained phases are provided in Table 1 below, focusing on thermally stable compounds of high density, low-atomic numbered constituents. As neutron moderation is physically described by simple elastic “billiard-ball-type” collision, the most efficient entrained phase moderators possess high densities of elements with mass approaching that of a neutron: H, Li, Be, B, C … Along with the logarithmic energy decrement ξ, the atomic mass and number density of elements in a material determined the slowing down power, a primary metric for comparing the effectiveness of materials as moderators: Slowing Down Power = ξ Nd σs, where Nd is the atomic number density and σs is the cross-section for neutron scattering. In Table 1 and Figure 7 the slowing down power of potential entrained phases are compared to the two most common moderators, graphite, and water. As seen, the slowing down power for water far exceeds that of graphite, owing the very high density of low atomic number hydrogen. For this reason, light-water-reactors (LWR’s) enable relatively small and very high-power density cores, especially as compared to the higher temperature graphite-moderator gas-cooled systems for which the materials in this work are being developed for.