High-Intensity Neutron Fields at the Center of the Metallic Core of Fast Reactors as the Primary Standard Neutron Fields

It is proposed to use high-intensity neutron fields with an average neutron energy of 0.7–1.5 MeV at the center of the metallic core of fast reactors as the primary standard neutron fields along with the known standard fields of fission neutrons with average energies of 1.93 and 2.13 MeV, respectively, from ²³⁵U and ²⁵²Cf nuclei, which are used in the world practice of taking neutron measurements. It is also proposed to create standard fields of nuclear fission neutrons with an average energy of 2 MeV by the bombardment of ²³⁵U converters with a beam of fast monoenergetic neutrons with energies of 2.5 and 14 MeV and 50-MeV protons.     

A method for measuring the energy spectrum of neutrons with energies of 1–15 MeV

The characteristics of the detector prototype for 1–15 MeV neutrons are described. The prototype is a full absorption detector consisting of interlaced plastic scintillators and ⁶Li doped glasses. A neutron incident on the detector deposits all its energy in the plastic scintillator, is moderated to thermal energies, and detected in the lithium glass. The measured time of complete neutron moderation is ～60 μs. Recording two signals in this time interval from the first event of neutron scattering in the plastic scintillator and from the neutron absorption by a lithium atom in the glass, it is possible to effectively suppress background thermal neutrons and γ rays and, therefore, detect low-intensity neutron sources. Owing to the proposed detector design, the direction toward the neutron source can be determined.     

Studying the dependence of the ratio of the prompt neutron yields in the symmetric and asymmetric fission of <Superscript>235</Superscript>U and <Superscript>238</Superscript>U nuclei on the energy of neutrons inducing this fission

The ratios have been measured of the prompt neutron yields in the symmetric and asymmetric fission of ²³⁵U nuclei induced by thermal neutrons and fast neutrons of pulsed nuclear reactors, as well as in the ²³⁸U fission induced by 14.7-MeV neutrons. The well-known measured integrated cross sections of the system of dosimetric nuclear reactions are compared to the cross sections calculated using the differential cross sections from the data libraries known worldwide, the three-component representation of fission neutron spectra from ²³⁵U and ²⁵²Cf nuclei, and the KASKAD spectrum reconstruction program.     

A digital method for pulse-shape discrimination between neutrons and γ rays at a high counting rate and a low energy of detected radiation

A digital method for pulse-shape discrimination between neutrons and γ rays was used in measurements at counting rates of up to ～10⁶ counts/s in the energy range of ～2–800 keV. Pulses produced by neutrons and γ rays in a stilbene-based scintillation detector were digitized by a digital oscilloscope and transmitted to a computer for carrying out particle identification. Identification was performed for radiation of radionuclide sources and a pulsed neutron generator operating in a repeated triggering mode. Amplitude spectra of pulses identified as neutrons and γ rays of radiation from the generator were measured. At a detector counting rate of ～8.5 × 10⁵ counts/s, ～90% of all recorded pulses were recognized as neutrons. In the energy range of ～30–800 keV, the γ-ray suppression factor was ～10⁴–10³ at counting rates of ～1.5×10⁵–5 × 10⁵ counts/s, while the efficiency of identifying neutrons was >0.9. The suppression factor for γ rays with an energy of ～10 keV was ～300, and the neutron identification efficiency was ～0.75.     

A method for decreasing the loading on scintillation channels in facilities with pulsed neutron sources

A method is proposed for decreasing the loading on scintillation channels in nuclear material detection and monitoring facilities comprising pulsed neutron sources, neutron moderators, and scintillators with a system for pulse shape discrimination between neutrons and photons. This method is based on the use of composite scintillators containing cylindrical shells of a thermal-neutron absorbing material. Selecting the sizes of zones in a composite scintillator and the absorber type (cadmium and lithium carbonate with different 6 Li content), it is possible to considerably increase the decay time constant for thermal neutrons in a composite scintillator and thereby reduce its load as compared to a homogeneous scintillator over the same time period after a pulse from the neutron source. The sizes of the scintillator component parts and the absorber material are optimized, which provides a means for decreasing the load on the scintillation channel several-fold while maintaining constant detection efficiency for fast fission neutrons. The capabilities of this method for decreasing the load are demonstrated by the example of the operational prototype of the fissile material detection and monitoring facility with a graphite moderator and LS-13 scintillators. The efficiency of the facility used in this method is compared to that of a facility with a deuterium-containing scintillator, in which no radiative capture photons are produced.     

A neutron semiconductor detector based on TIInSe<Subscript>2</Subscript>

A new semiconductor detector of neutron radiation based on a TIInSe₂ crystal has been investigated. The detector is produced from a homogeneous semiconductor sample with two electric contacts and operates in an integrating mode. It is shown that, owing to its high sensitivity (～10^?13 A/(neutron cm^?2 s^?1)) and small size (the volume of the sensitive crystal element is ～7 mm³), the detector is capable of monitoring spatial, time, and intensity distributions of γ rays and neutrons in pulse research reactors.     

Monitoring of <Superscript>235</Superscript>U content of fuel assemblies for nuclear reactors based on neutron and γ-ray scintillation detector systems

A model of the setup for determining the ²³⁵U content of a fuel assembly for WWER-440 reactors is described. The setup operates based on the digital technology for discriminating neutron and photon scintillation responses. An AmLi source is used as an external neutron source. A three-channel system for detecting fission particle coincidences is based on EJ-309 scintillators. A set of digital codes for obtaining and analyzing neutron-neutron (nn), neutron-γ (nγ), and γ-γ (γγ) coincidences has been developed. Dependences of double coincidences have been experimentally obtained and analyzed with the aim of determining the ²³⁵U content of the fuel assembly. It is shown that fission neutrons and coincidences of the nn and nn + γn + nγ types are the best informative parameters for the ²³⁵U content of the fuel assembly.     

A sectioned spectrometer of fast neutrons

Development of a capture gated spectrometer on the basis of a liquid organic scintillator doped with enriched ⁶Li is discussed. Particular interest is evoked by the good pulse height resolution of the spectrometer for 14-MeV neutrons, which is expected to be very high, ～10–15%. This resolution is attained by compensating for the nonlinearity of the light yield in the scintillator owing to the use of separate optically isolated sections, which independently detect scintillations from each recoil proton. The detector is sensitive to fluence rates ranging from 10^?4 to 10² cm^?2 s^?1 above a threshold of 500 keV under conditions of uncorrelated γ-ray background at a level of up to 10² s^?1 (E > 100 keV). A pilot model of the detector based on a scintillator without a lithium dopant has been produced and tested. The detector efficiency is governed by the scintillator volume (～1.2 l); for 3-MeV neutrons, its value is 0.2–0.5%. The response of the pilot detector to neutrons from a Pu-α-Be source with energies of up to 10 MeV has been measured. Initial testing indicates a low threshold at an ～600-keV energy of a recoil proton. A good spectral response is obtained using the criterion that three optical sections of the detector operate at a time. This spectrometer can find application in low-background experiments in basic physics research, as well as in space research and nuclear medicine for measuring the parameters of the neutron flux.     

Asymmetric fission of <Superscript>235</Superscript>U nuclei induced by thermal neutrons and fast neutrons of pulsed nuclear reactors

The contributions that neutrons emitted from the neck of excited dumbbell-shaped fissioning ²³⁵U nuclei at the instant of their fission make to the prompt fission neutron spectrum, as well as the ratios of these contributions, have been measured for nuclear fission induced by thermal neutrons and fast neutrons of pulsed nuclear reactors.     

A digital method for pulse-shape discrimination between particles

Results from experimental and theoretical investigations of a digital technique for pulse-shape discrimination are presented. Pulses from a scintillation (stilbene-based) detector are recorded by a digital oscilloscope and transmitted to a computer for carrying out particle identification. For a stilbene scintillation detector irradiated by neutrons and γ rays, the γ-ray suppression factor is ～0.5 × 10³?10⁴ and the neutron identification efficiency is ～0.70-0.90 over an energy range of ～30–500 keV at a counting rate up to ～10⁵ pulse/s. It is shown that this technique can form the basis for developing a detector unit that comprises a fast ADC and a digital signal processor and provides online identification of particles by scintillation pulse shape.     

A scintillation strip detector for detection of thermal neutrons

A scintillation strip detector designed for precise experiments on the reflectometry of thermal neutrons is described. Test results obtained on the beam of a time-of-flight spectrometer have shown that the detector has a detection efficiency for thermal neutrons close to 100%, a lowered sensitivity to the neutron background, and a low level of intrinsic noise. The coordinate resolution of the detector can be varied in the range 1.5–0.1 mm. The strip is manufactured from a mixture of a ZnS(Ag) scintillator and a ⁶LiF neutron converter.     

Testing the neutron attenuator based on <Superscript>6</Superscript>LiH for γ-ray diagnostics of plasmas in the JET tokamak

A ⁶LiH attenuator of a neutron flux incident on a detector is used to reduce the γ-ray background induced by neutrons in the detector material. This attenuator has been tested during experiments with deuterium (DD) plasmas on the JET tokamak. A specimen of the neutron attenuator with dimensions of ?30 × 300 mm has been developed by the Russian Academy of Sciences’ Ioffe Physico-Technical Institute and inserted into a vertical collimator used for γ spectrometry of plasmas. To compare γ-ray spectra recorded with and without the ⁶LiH attenuator being mounted, identical discharges with heating of the DD plasma by a neutral particle beam have been selected. For γ rays with energies of <3 MeV, which are induced by neutrons in the detector material, the suppression factor is found to be ～100. A low attenuation (～2) observed at energies of >3 MeV can be attributed to the transparency of the ⁶LiH attenuator for γ rays. This portion of the spectrum is due to γ radiation of the plasma and γ rays induced by neutrons in the constructional materials of the tokamak. To estimate the efficiency of the ⁶LiH attenuators as a mandatory component of the ITER γ-ray diagnostic system, it is necessary that measurements be taken in deuterium-tritium (DT) discharges.     

Energy spectra of neutrons escaping from light and heavy fragments with asymmetric <Superscript>235</Superscript>U and <Superscript>252</Superscript>Cf nuclear fission

The neutron-activation analysis is used to determine the energy spectra of neutrons escaping from light and heavy fragments from nuclear fission of ²³⁵U by thermal neutrons and spontaneous fission of ²⁵²Cf nuclei. It is shown that the difference of the spectra obtained is slight, ∼90.6 keV. The 28-group energy spectra of instantaneous neutrons from fission of ²³⁵U and ²⁵²Cf nuclei, which were reconstructed using four-component a priori spectra, are presented.     

A method for determining the intensity of concomitant neutron source D(d, n)3He when studying the characteristics of delayed neutrons from nuclear fission induced by neutrons from reaction T(d, n)4He

A factor that may be responsible for the discrepancies in the total delay neutron yields, the relative yields of separate groups of delayed neutrons, and half-lives of their precursors has been investigated. These discrepancies are shown to be attributable to the effect of a concomitant neutron source—reaction D(d, n)³He. Such a source is unavoidably present in an experiment where reaction T(d, n)⁴He on a solid-state target is used as a neutron source. A method has been developed to calculate the contribution of neutrons from the reaction D(d, n)³He to the measured total and relative yields of delayed neutrons and the half-lives of their precursors in heavy nuclei fission induced by neutrons with energies of 14–18 MeV. The energy dependence of the parameters of delayed neutron groups from the neutron-induced fission of ²³⁸U nuclei in the energy range of 14.2–17.9 MeV is presented.     

A Hemispherical Fission Chamber for the Neutron-Flux Monitoring

A fission ionization chamber has been designed to monitor neutron fluxes produced by the proton beam of the Van de Graaf accelerator in the ⁷Li(p, n)⁷Be nuclear reaction. A target producing neutrons is placed at the center of the chamber, which has a hemispherical shape. The neutrons escaping from the target are detected within the limits of a space angle a little larger than 2 sr. A ²³⁵U isotope is used as a fissile material. A U₃O₈ layer with a thickness of 300 g/cm² and an area of 80 cm² has already been deposited on the surface of the inner hemisphere using electrolysis. The chamber is filled with a gas mixture of 97% Ar + 3% CO₂. The detection efficiency of the chamber for neutrons with 1- to 150-keV energies is found to be 1.6 × 10^–5. Using this chamber, it is possible to measure (with a statistical accuracy of 1%) the integrated neutron yield in a thick target exposed for 20 min to a proton beam with a current of 3 A and an energy of 1942 keV, which is 60 keV above the threshold of the ⁷Li(p, n)⁷Be reaction. The design of the fission chamber is described, and the test results are presented.     

Neutron detectors based on silicon photomultipliers

Neutron detectors comprising a new type of photosensors—silicon photomultipliers—coupled to a single-crystal LiI(Eu) scintillator and a ⁶LiF/ZnS(Ag) scintillation screen are described. These detectors are compact, which improves their usability. Some parameters of the detectors are presented, their sensitivity to thermal neutrons and γ rays is estimated, and their possible applications are proposed.     

A Prototype of a Neutron and γ-Ray Spectrometer for Studying Solar Activity at Distances of 0.5 Astronomical Units to 25 Solar Radii

The InterSONG instrument now being developed at the Skobel’tsyn Institute of Nuclear Physics will allow the sensitivity of experiments on solar neutrons to be significantly raised by performing measurements in the immediate neighborhood of the Sun. It is expected that this instrument will be used on the Inter-HELIOS automatic interplanetary station operating at distances of as many as 25 solar radii. The instrument is based on a LiI scintillating crystal enriched with ⁶Li, and a boron-containing plastic scintillator that also acts as a fast neutron moderator. The instrument is designed to detect neutrons with energies of 0.1–100 MeV and electromagnetic radiation over the range 0.03–10 MeV.__________Translated from Pribory i Tekhnika Eksperimenta, No. 3, 2005, pp. 24–36.Original Russian Text Copyright © 2005 by Bogomolov, Kuznetsov, Lishnevskii, Rubinshtein, Ryumin, Nemchenok, Sobolevskii, Ufimtsev.     

Slow-neutron detector based on thin CVD diamond film

A slow-neutron detector based on thin CVD diamond film is discussed. A deep graphitized layer in the diamond film acts as one of the electrodes in the detector. The detector demonstrates low sensitivity to γ radiation, at the background level. The detector is tested with a laboratory neutron source based on the isotope ²⁵²Cf in a moderator. Slow neutrons are detected by means of the boron isotope ¹⁰В and lithium isotope ⁶Li. In both cases, slow neutrons are recorded with about 3% efficiency. The efficiency of slow-neutron recording is also assessed theoretically.     

An experimental setup for studying spectra of β particles emitted by mixtures of 235U and 239Pu thermal neutron fission products

An experimental setup on the thermal neutron beam of the IR-8 reactor at the National Research Centre Kurchatov Institute is described. This setup has been designed to precisely measure the ratio of spectra of β particles emitted by mixtures of ²³⁵U and ²³⁹Pu fission products. The experiment is based on simultaneous measurements of β spectra from fissile isotopes ²³⁵U and ²³⁹Pu and the background spectrum in the same neutron beam. Measurements of the β spectra are taken by a high-selectivity β spectrometer in the time interval when the neutron beam is intercepted by a mechanical chopper. The evolution of the measurement technique is considered.     

Studying kinetics of scintillation process in a CaMoO<Subscript>4</Subscript> crystal

The kinetics of scintillation process in Czochralski-grown CaMoO₄ crystals excited by α particles and γ rays from ²³⁸Pu and ¹³⁷Cs radioactive isotopes, respectively, is investigated. Kinetics of scintillations is shown to be complex and multicomponent in character. Fast and slow scintillation components are observed, and their relative contributions to the total light yield in the scintillator are determined. The possibility is demonstrated of discriminating between scintillator signals due to α particles and γ rays.