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Slow Neutron Detectors
Published in Douglas S. McGregor, J. Kenneth Shultis, Radiation Detection, 2020
Douglas S. McGregor, J. Kenneth Shultis
The NIST uses a combination of a calibrated neutron current and calibrated neutron detectors to determine the unknown efficiency of a neutron detector [McGregor and Shultis 2011]. The calibration consists of using a 252Cf neutron source placed in a heavy-water moderator pool that approximates the thermal-neutron environment inside a light-water reactor. However, the preferred method for testing a thermal-neutron detector is with a single-crystal diffracted beam, whose spatial profile is determined from activation analysis of a Dy foil, or is measured with a neutron-imaging detector so that the neutron beam profile is documented within the workspace. The neutron flux is then measured with a calibrated fission chamber. Hence, the basic method is to (1) profile the relative thermal-neutron beam intensity as a function of position and then (2) measure the neutron current (or fluence) with a calibrated standard fission chamber. The count rate of the unknown detector is then directly compared to the measured neutron current. Note that fission chambers, typically fabricated with uranium as the neutron reactive material, deviate slightly from the ideal 1/v behavior. Hence, knowledge of the diffraction angle, and resulting neutron energy, is important for making thermal corrections to the NIST calibration method.
Preliminary considerations: reactor types and characteristics
Published in Peter R. Mounfield, World Nuclear Power, 2017
There are two main sources of radio-isotopes in the heat transport system of a nuclear reactor: first, those produced from the uranium dioxide fuel by the fission process (nuclear fission products) and, second, those produced by neutron activation of materials other than fuel which constitute the reactor core. Fission products normally remain within the metallic sheath which encloses the uranium fuel, but during the course of normal operation a proportion of the fuel elements develop holes or cracks in the sheath through which gaseous products can emerge. Some of them have half-lives measured in minutes; longer-lived products are removed from the heat transport system by purification. Neutron activation products are formed by the effect of neutron flux in the reactor core on any substances residing in or passing through the core. Thus the fuel, any pressure tubes and the calandria all become radioactive as their materials absorb neutrons. Water passing through the reactor core becomes activated, as does the circulating water because of traces of impurities from corrosion of the materials used to construct the heat transport system (Neil 1974).
Nuclear Fission Reactor
Published in C. K. Gupta, Materials in Nuclear Energy Applications, 1989
Classification of reactors on the basis of their purpose falls basically into two categories. One is the power reactor which, of course, represents the major effort in the nuclear field. Into the other category falls the research reactors which are in use in all parts of the world for one or more of such purposes as research on physical, chemical, or biological processes; materials and components testing under irradiation; teaching and training; and radioisotope production. The power produced in research reactors in the form of heat is an undersirable by-product which should be kept to a minimum in order to eliminate the need for elaborate cooling arrangements. The reactor is regarded principally as a source of neutrons. Openings leading into the core or into the lattice provided neutron flux composed of the entire reactor spectrum, and those into the moderator region, where no fuel is present, provide neutron flux which is predominantly thermal, with some fast neutrons. For any specific requirement of well-thermalized neutrons one uses what is called a thermal column, which is essentially an extension of the moderator against a portion of the side of the reactor from which the reactor shielding has been removed. Based on neutron flux, it is convenient to classify research reactors into high-, medium-, and low flux.
Transient Studies on Low-Enriched-Uranium Core of Ghana Research Reactor–1 (GHARR-1)
Published in Nuclear Technology, 2020
Prince Amoah, Edward Shitsi, Emmanuel Ampomah-Amoako, Henry Cecil Odoi
The variation of the neutron flux has effects on the main components of the reactor including the coolant and the fuel. The variation of the neutron flux in the main reactor components is quantified in terms of reactivity and reactivity coefficients including moderator temperature coefficient, void coefficient, and fuel temperature coefficient. These reactivity coefficients were determined from the Monte Carlo N-Particle code version 5 (MCNP5) input deck. The power profiles and temperature distributions obtained in this work are the cumulative effects of these reactivity coefficients and reactivity insertions used as input for the Program for the Analysis of Reactor Transients (PARET)/Argonne National Laboratory (ANL) code.
The Inventory and Source Term Simulation of the Argonaut Nuclear Reactor Inside a Severe Accident
Published in Nuclear Technology, 2021
Anderson M. S. Alves, Adino Heimlich, Fernando Lamego, Celso M. F. Lapa
Nuclear parameters such as radioactive decay, neutron capture probability, fission, etc., are characteristic of each nuclide. These data with the neutron flux allow us to model the inventory variation of each nuclide over time. The Argonaut reactor has a poor historical record of operations containing date, power, and critical time. However, these data are sufficient to determine an approximation of the neutron flux using the reaction rate (RR). The scalar neutron flux represents the number of neutrons that cross an arbitrary area in all directions per unit of time. The fission RR is given by the product between scalar neutron flux and the fission macroscopic section.
Pinwise Diffusion Solution of Partially MOX-Loaded PWRs with the GPS (GET PLUS SPH) Method
Published in Nuclear Science and Engineering, 2019
Hwanyeal Yu, Jaeha Kim, Yonghee Kim
In nuclear reactor analyses, the neutron flux distribution is required to accurately predict the behavior of a reactor. In terms of accuracy, heterogeneous calculation gives the best estimation for the neutron flux distribution. Although computing power has dramatically increased, this direct whole-core transport calculation is still too expensive since more than several tens of thousands of reactor calculations are needed for a practical reactor design. In this sense, transport-corrected diffusion methods in the conventional two-step procedure are still widely used for analyzing commercial reactors.