[en] Evaluations of electron accelerators of different types, having maximum energy of bremsstrahlung of up to 30 MeV, as neutron pulse sources, are given. It is shown that it is more advantageous to use tantalum, tungsten, uranium and lead as targets which permits to considerably increase photoneutron yield. It is also shown that electron accelerators may be successfully used to determine concentrations of many elements by the neutron activation method, analysis techniques possessing a high sensitivity

[en] The possibility of an analytical description of an increase in the neutron flux from ²³⁹Pu-Be(α, n)- sources, in which no information on the isotopic composition is used and the results of three or more measurements during a known period are used instead, is cosidered. A method of corrections for the increase in the neutron flux in the interval between measurements is given

[en] Results of evaluations of intense thermal neutron pulses generation in superhigh accelerator targets are given. One of the methods for thermal neutron pulse production in practice is one-turn beam extraction out of the accelerator to a circular target made of an element with a high atomic number equipped with inner moderator of hydrogen-containing material or beryllium. It is shown that thermal neutron pulse generation with 10²⁰ neutr/cm²xs peak density and approximately 10¹⁶ neutr/cm² fluence is possible at the HEPI accelerating-storage complex of 5000 geV proton energy

[en] A method is given to increase the efficiency of data acquisition getting about σsub(t) received by the measurements with energy resolution ΔE/E<=5x10^-4 on continuous action accelerators. The dependence of neutron energy upon the angle of emission from the target is used for simultaneous measurements of transparency of a single-sample by a few detectors. Simultaneous use of several neutron detectors located at different angles of thetasub(i) reaction permits to use increase rate of statistical data collection proportionally to the number of used detectors

[en] Application of a small selections method for spectrometry of pulsed neutron fluxes emitted with short-living hot plasma is described. The essence of the method consists in that that from the whole flux carried out are a simple random selection of small quantities of neutrons, then energy measurement of each of them by the time-of-flight method with subsequent restoration of parameters of spectral line and plasma ion temperature. Owing to modelling of spectrometry of pulsed neutron fluxes obtained were values of the spectral line parameters depending on a selection volume. The method is realized by means of a time of flight spectrometer. It is established that for measuring plasma ion temperature of the order of 1 keV with an error not worse than 50%, it is necessary to provide 6.2 m path length at Esub(n)=2.45 MeV at that the time-of-flight spectrometer consists of 100 plastic scintillation detectors, FEhU-63 photomultiplier and electronics of 0.5 ns time resolution

[en] An expression for calculation of photoneutron yield from targets irradiated with electrons has been derived. The expression for calculating photoneutron yield represents a single integral. It was suggested during the formula derivation that electron losses energy continuously and the whole bremsstrahlung is directed forward and cross section is inversely proportional to photon energy. An expression for calculating track length has been obtained as well. Results of the track length calculation for tantalum and tungsten using the formula and the Monte Carlo method are presented. Good agreement of the results for thickness>0.5R₀ (where R₀ - mean electron path) is observed. Comparison of calculated photoneutron yields and experimental data also showed their good agreement. It the range of 10-40 MeV electron energies for photoneutron yield derived was a formula including giant resonance parameters. Results of calculations of photoneutron yield using the formula and the Monte-Carlo method are practically coincided for electron energies of 10, 15, 20 MeV in the thickness range of 0.5-12 g/cm²

[en] Main stages of development of polarizing neutron guide equipment in LIYaF of the USSR Academy of Sciences are described. To carry out experiments on solid-state physics constructed was a working mock-up of a polarizing neutron guide having 1570 mm length of a mirror channel. Successful application of polarizing mirrors to the working mock-up permitted to develop and fabricate five-meter polarizing neutron guide with output flux equal to 1.5x10⁷ neutr/cm²xs. The following stage of development of polarizing neutron guides was the construction of four-meter neutron guide at the WWR-M reactor with output flux equal to the highest possible. Improvement of optical sections geometry made it possible to produce integral flux of 6.0x10⁷ neutr/cm²xs in this neutron guide at 15 MW reactor power. The results obtained testify to prospects of the mirror method for polarization of thermal neutrons of a wave length lambda >= A. Neutron guides-polarizators permit to produce high fluxes of polarized thermal neutrons in the wide interval of wave length

[en] Simple and reliable time-to-code converter has been developed for using together with AU-128.2 amplitude analyzer. Basic conver-- ter flowsheet is given and principle of its operation is considered. The converter has the following characteristics: 4V amplitude of input pulses of negative polarity, 1 μms duration of input pulses. The number of channels is 128. Value of the channel changes from 0.12 ms to 4.09 ms minimum resolution time amounts to 0.12 ms, output pulses correspond to TTL logic. The converter was used in experiments on study of delayed neutrons during U²³⁵ fission

[en] Possibility for production of considerable quantities of ²³⁷Np in a hard neutron spectrum of the first layers of thermonuclear reactor uranium blanket is considered. Calculation of the neutron spectrum and dynamics of nuclear transformations in the blanket of depleted uranium was carried out according to the BURNEL program. It is established that ²³⁷Np production practically completely results from the (n, 2n) reaction on ²³⁸U. Contribution from the (n, 3n) nuclear reaction on ²³⁸U nuclei constitutes less than 0.5%. Rate of ²³⁸Np production in the blanket of a hybrid thermonuclear reactor designed for ²³⁸Pu production amounts to approximately 460 kg ²³⁷Np per year. Volume-average content of ²³⁷Np in spent fuel constitutes approximately 1.1 kg ²³⁷Np per 1t fuel at a run Tsub(k)=2.65 year. So, using in the hard spectrum of the hybrid thermonuclear reactor blanket depleted or natural uranium as a source material for ²³⁷Np production will cause significant increase of the rate of ²³⁹Pu commercial production

[en] Neutron concentration and neutron output from twolayer loading target during deuteron beam braking in first layer of target is studied theoretically. Struggling effect on neutron output is discussed