![]() The size and cost of these neutron sources are comparable to spontaneous fission sources. The useful lifetime for such sources depends on the half-life of the radioisotope. An alpha-beryllium neutron source may produce about 30 neutrons per 10 6 alpha particles. Alpha neutron sources typically produce ~10 6–10 8 neutrons per second. Thus, one can make a neutron source by mixing an alpha-emitter such as radium, polonium, or americium with a low-atomic-weight isotope, usually by blending powders of the two materials. ![]() Neutrons are produced when alpha particles hit any of several light isotopes including isotopes of beryllium, carbon, or oxygen. Radioisotopes which alpha decay mixed with a light element A typical 252Cf neutron source emits 10 7 to 10 9 neutrons per second when new but with a half-life of 2.6 years, neutron output drops by half in 2.6 years. 252Cf neutron sources are typically 1/4" to 1/2" in diameter and 1" to 2" in length. 252Cf and all other SF neutron sources are made by irradiating uranium or a transuranic element in a nuclear reactor, where neutrons are absorbed in the starting material and its subsequent reaction products, transmuting the starting material into the SF isotope. The most common spontaneous fission source is the isotope californium-252. Some isotopes undergo SF with emission of neutrons. Neutron source variables include the energy of the neutrons emitted by the source, the rate of neutrons emitted by the source, the size of the source, the cost of owning and maintaining the source, and government regulations related to the source. Neutron sources are used in physics, engineering, medicine, nuclear weapons, petroleum exploration, biology, chemistry, and nuclear power. Fundamental research with neutrons: Ultracold neutrons Ī neutron source is any device that emits neutrons, irrespective of the mechanism used to produce the neutrons.It can provide an accurate method of determining the neutron emission rate of Cf-252 spontaneous fission, and also an approach to calibrating the detection efficiency of neutron coincidence counter while the source strength is unknown.For neutron sources used in nuclear weapons, see modulated neutron initiator. Furthermore, the detection efficiency is inversed by dividing the total neutron counting rate with the neutron emission rate when the source is placed at the central axis, which accords with the result of Monte Carlo simulation by using the MCNPX code well. It is shown that the values measured by the two regression methods are consistent with the corrected results of the nominal value within 2% deviation. The verification experiments are carried out by the JCC-51 neutron coincidence counter. On the assumption that the average neutron die-away time is constant in the sensitive range of detection system, therefore the characteristic coefficient from the changing process can be extracted, and two kinds of methods of measuring the neutron strength are established, which are independent of the efficiency variation. On the basis of the measurement equations under the point model assumption, the neutron coincidence counting rate is correlated with the total neutron counting rate, and then the regression analyses with different coincidence gates and different source locations in the counter are performed. In order to develop a more portable measurement method for larger suitable dynamic range, the comprehensive algorithms based on the neutron multiplicity counting are studied in this paper. In addition, the indirect measurement method by manganese bath activation needs a long period more than 8 hours and it will have a large uncertainty while the source strength is lower than 10(4)n/s. As the source age increases the contributions from Cf-250 and Cm-248 spontaneous fission become more significant, thus the neutron emission rate cannot be calculated simply according to the Cf-252 decay law. However, it is often necessary to correct the neutron emission rate due to its short half-life of 2.645 years. The Cf-252 isotope sources have a recommended standard neutron spectrum of spontaneous fission, and have been widely used in scientific researches, such as the detection efficiency calibration of neutron detectors, the characterization of neutron dose equivalent meters, the active analysis of special nuclear materials, etc.
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