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Neutrons and Other Important Particles of Reactor Physics
Published in Robert E. Masterson, Introduction to Nuclear Reactor Physics, 2017
Some relatively common elements like Boron and Cadmium, which we will study in more detail in future chapters, also have the ability to absorb free neutrons in the event that we generate too many of them. Hence, these elements can be used to create the control rods in a nuclear reactor core—which are used to control the power output of the reactor and the scope of the nuclear chain reaction. Even if the neutrons are moving very quickly (at a speed of several thousand meters per second), an element like Boron or Cadmium can “reach out” and capture one of these neutrons if they get too close to the nucleus. Other isotopes of the elements such as Xenon and Samarium are even more adept at capturing free neutrons that happen to come in contact with them. We will discuss exactly how they do this in Chapter 4. Neutrons are members of a family of very heavy nuclear particles called Hadrons. The probability that they will be absorbed by an atomic nucleus depends in part on the speed at which they are moving. Fast neutrons, which can move at speeds between 10,000 and 20,000 km/s, have a very low probability of being absorbed because they are simply not around the nucleus long enough for the nucleus to reach out and “grab” them. Slow neutrons, on the other hand, are much easier to absorb because they move much more slowly (between speeds of 2 and 3 km/s). Because they move so slowly, they are much easier, figuratively, for a nucleus to reach out and “grab them.” Hence, they are much easier to absorb.
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Published in Philip A. Laplante, Comprehensive Dictionary of Electrical Engineering, 2018
principal component analysis (PCA) performance criteria; these local criteria are then minimized with respect to the local decisions and the results are passed to the coordinator. The coordinator iterates the values of prices until the interaction equations and, if needed, any other coupling constraints are satisfied. primal sketch a hierarchical image representation that makes explicit the amount and disposition of the intensity changes in the image. At the lowest levels of the representation, primitives just describe raw intensity changes and their local geometrical structure; at the higher levels they represent groupings and alignments of lower-level ones. primary (1) the source-side winding. (2) refers to the portion of a nuclear power plant containing the reactor and its equipment. primary coolant the medium used to remove energy, in the form of heat, from a nuclear reactor core, e.g., water, helium, or liquid metal. primary feeder See feeder primitive polynomial a polynomial p(x) of degree m that gives a complete table of 2m distinct symbols containing 0 and 1. The reciprocal of the primitive polynomial is also primitive. See also irreducible polynomial. primitive-based coding any scheme to detect edges, lines, and other local features of images, then use them to code the image. For example, edges may be used to segment the image into regions which are then independently coded as simple surfaces, while the boundaries are compressed with a chain code. Princeton architecture a computer architecture in which the same memory holds both data and instructions. This is contrasted with the Harvard architecture, in which the program and data are held in separate memories. principal axis the optical axis of a lens or camera, usually normal to the image plane. principal component notionally, the direction of greatest variability of a random vector or among a set of sample vectors. More specifically, the principal component is the direction of the eigenvector associated with the largest eigenvalue of the covariance matrix of the random vector (or the sample covariance of a sample set). More generally, the n principal components of a distribution are the eigenvectors corresponding to the n largest eigenvalues. Principal components are frequently used for data clustering, pattern analysis, and compression. principal component analysis (PCA) a technique applied to n-dimensional data vectors with the aim of finding a set of m-orthogonal vectors (typically m n) that account as much as possible for the data's variance. PCA can be carried out in an unsupervised fashion by implementing a normalized version of Hebb's rule on m-linear neural units.
Nonchemical Rocket Engine
Published in D.P. Mishra, Fundamentals of Rocket Propulsion, 2017
A typical solid-core fission thermal reactor is shown in the Figure 11.13, which is similar to that of electrothermal engine. It consists of nuclear reactor, neutron reflector, propellant feed system, and CD nozzle. The nuclear reactor core contains the nuclear fuel components where the nuclear reactions take place and the heat is generated. The enriched uranium is contained in the fuel elements, which can be in rod or pebble bed form. Generally, uranium in metal form is not used as a nuclear fuel in the fuel element, as its melting temperature is around 1400 K and dangerous for rocket engine applications. Rather, uranium compounds, namely, uranium carbide (UC2), uranium dioxide (UO2), and uranium nitride (UN) are being used as nuclear fuel, as interaction between neutrons and uranium nuclei is not influenced by chemical combination of uranium with other elements in its compound form. Of course, dimensions of the fuel elements for same power level will increase as compared to pure uranium, because the uranium compound contains fewer uranium nuclei per volume. The most preferred compound of uranium is uranium dioxide, as it is quite stable with melting temperature of 3075 K. Besides this, it is quite compatible with hydrogen gas from the stability point of view. As discussed earlier, the chained nuclear reactions must be contained to release heat in a controlled manner. The most common material for moderator is graphite, which can scatter away neutrons, which have low molecular mass, and arrest the multiplication process further. Besides this, graphite is being preferred as moderator material as it can maintain structural and dimensional ingenuity at high temperature and pressure due its higher sublimation temperature (3990 K) at reasonable pressure. But it can react with hydrogen at high temperature to produce hydrocarbon. Hence, the fuel element must be coated with certain refractory materials to prevent the graphite of the fuel matrix to come in contact with hydrogen at high temperature. Generally, niobium carbide (NbC) or zirconium carbide (ZrC) are being used as protective coating of fuel elements, as these two refractory materials do not react chemically even at high temperature and remain neutral in the neutron environment. As a result, the lifetime of fuel elements is increased with coating of carbides even when they are operated at high-temperature and high-pressure condition. Let us look at the cross section of fuel element matrix used in the KIWI reactor core, as shown in Figure 11.13b [6]. The uranium oxide in the forms of small spheres is dispersed in the graphite matrix. These graphite uranium oxide matrices are used to fabricate the hexagonal-shaped rods with 19 holes along its length for hydrogen flow, which can be used as fuel elements in the nuclear core. But both inner and outer surfaces of these fuel elements are coated with niobium carbide. All these six hexagonal rods are locked by a stainless steel tie rod to form a fuel assembly. Several such fuel assemblies can be mounted together to form the reactor core.
Computational Analysis of Thermal Striping in Primary Sodium System of Liquid Metal Fast Breeder Reactor Using Finite Volume Method
Published in Nuclear Science and Engineering, 2023
S. Suyambazhahan, T. Sundararajan, Sarit K. Das
Hot and cold sodium pools make up the primary sodium system of the Prototype Fast Breeder Reactor (PFBR). Both pools are separated by a thin-walled inner vessel. The nuclear reactor core comprises the fuel storage, blanket, shielding subassemblies (SAs), and reflector, which are mounted on a grid plate where heat is generated by nuclear fission. Sodium from the cold pool is circulated to the hot pool through the reactor core by operating two centrifugal pumps in parallel. After exchanging heat with the secondary sodium circuit, sodium flows to the cold pool through four intermediate heat exchangers (IHXs). The control plug is a hollow cylindrical shell structure having four compartments that are located right above the core outlet in the hot pool. The control plug is an important component in the reactor assembly. It consists of thermocouples for measuring the temperatures of sodium coming out of various fuel subassemblies, absorber rod drive mechanisms (ARDMs), core power monitoring instrumentation, and sampling tubes. A control plug has a porous plate at the bottom, known as the lattice plate (LP). Structural support is provided for the core-monitoring thermowells by the LP. The thermowells are mounted in the perforated solid plate above the LP and are called the core cover plate (CCP).
Experimental and Numerical Analysis of Heat Transfer in a Tall Vertical Concentric Annular Thermo-siphon at Constant Heat Flux Condition
Published in Heat Transfer Engineering, 2019
Jawed Mustafa, Mohammad Altamush Siddiqui, Sayed Fahad Anwer
From the studies so far reported in the open literature, most of the numerical work done on natural convection in a vertical annulus is limited to the highest aspect ratio of 210 and the working fluid as air. The highest aspect considered for the numerical analysis on water is only up to 71. The experimental work of Chun et al. [7] on flow boiling of water in annulus of aspect ratio 373.25 seems to be of great importance in the field of Nuclear reactors. As explained by Husain and Siddiqui [19], a typical nuclear reactor core is made of several hundred fuel assemblies that consist of bundles of fuel elements. Each fuel element is a long and thin tube containing cylindrical fuel pellets placed centrally that are surrounded by a metal tube called the cladding, surrounded by the coolant flowing through an annular gap outside the cladding. The fuel pellets are typically uranium oxide (UO2) or some others like thorium-bearing materials. Thus, each fuel element represents an annular structure with an internally heated cladded fuel rod (that generates heat due to fission) in the center and a coolant flowing outside through the annulus. The fuel pellets are cladded to prevent the fission products from escaping into the coolant. Since the fuel elements are quite long compared with the annular gap for the coolant, the aspect ratio may go as high as 500. Therefore, numerical analysis on natural convection of water in an annulus of aspect ratio 352 has been taken up and compared with experimental results on a similar system. In the present study, a comparative analysis of the numerical and experimental results for the wall and liquid temperatures, liquid mass flow rates, and Nusselt numbers have been made for different operating conditions.