Total Neutron Cross Section Measurements at the Energy 14.1 MeV for 12 C, 19 F, 32 S, 115 In 128 Te, 208 Pb
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al Neutron Cross Section Measurements at the Energy 14.1 MeV for 12C, 19F, 32S, 115In 128Te, 208Pb S. V. Artemova, *, F. Kh. Ergasheva, M. A. Kajumova, A. A. Karakhodzhaeva, O. R. Tojiboeva, G. A. Abdullaevaa, G. A. Kulabdullaeva, E. T. Ruzieva, V. A. Tatarchuka, and B. S. Yuldasheva aInstitute
of Nuclear Physics, Academy of Sciences Uzbekistan, Tashkent, 100047 Uzbekistan *e-mail: [email protected] Received March 2, 2020; revised April 15, 2020; accepted April 27, 2020
Abstract—To measure the total neutron cross sections, the authors developed a convenient technique using silicon semiconductor detectors with the simple and compact detecting system. The created technique was used to measure A + n total cross sections at the energy En = 14.1 MeV for several nuclei in different mass region. The neutron generator NG-150 in D + T regime of neutrons generation is used as fast neutrons source. The results of measurements and their comparison with the appropriate literature values are also presented. DOI: 10.3103/S1062873820080079
Emerging nuclear technologies require the development of efficient neutron sources with high output fluxes. Their presence will expand the experimental base for nuclear physics research and create opportunities for the introduction of new technologies in nuclear energy, production of radionuclides, etc. Much attention is currently paid to the development of systems of fissile materials in a subcritical state, controlled by an external powerful neutron source (accelerator-driven systems (ADS)) [1, 2]. The assembly of a well-optimized subcritical system and an “external” source implemented by the accelerator can provide a neutron flux comparable to the fluxes of typical research reactors, and in the future it will be the basis for nuclear energy. The indisputable advantages of ADS include not only the ability to multiply neutrons from an “external” source, but also to use them as nuclear fuel stores, based in particular on the large world reserves of thorium. If a mixture of natural or weakly enriched 235U uranium is used as the initial fuel, then when the reactor is irradiated with neutrons obtained with the help of an accelerator, there are two main processes. Firstly, at neutron capture by the 232Th nucleus, there is a chain of β-decays: β
β
233 233 + n → γ + 233 90Th → 91Pa → 92 U. And on the main isotope of uranium, 238U there is a process: 232 90Th
β
β
239 239 + n → γ + 239 92 U → 93 Np → 94 Pu. In the interaction of neutrons with both the developing isotopes, 233U and 239Pu, fission occurs, which is the energy source. Each fission reaction leads to the 238 92 U
loss of one 233U or 239Pu nucleus, and each of the above reactions leads to the appearance of such a nucleus. If the probabilities of the processes of fission and the formation of fissile nuclei are equal, then the number of 233U or 239Pu isotopes during reactor operation reaches a certain equilibrium value, that is, the fuel is reproduced automatically. This technology ensures nuclear safety and thwarts the unauthorized distribu
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