Effect of temperature and neutron irradiation on the plastic deformation of α-uranium single crystals

The plastic deformation of -uranium single crystals was studied for various crystallographic orientations and temperatures. Explanations are given for the form of the strain curves in terms of the principal systems of plastic deformation. The single crystals were irradiated by neutrons in a reactor at temperatures up to 100°C with integral fluxes up to 5.5·10¹⁷ neutrons/cm² and a flux of 4·10²⁰ neutrons/cm². The irradiation increased the critical shearing stress by factors of 3 to 5 for slip in the (010) plane and reduced the relative elongation and the region of easy slip. On annealing single crystals irradiated by fluxes up to 5.5·10¹⁷ neutrons/cm² the mechanical characteristics were restored.Translated from Atomnaya Énergiya Vol. 19, No. 4, pp. 372–380, October, 1965     

Effect of neutron irradiation on some properties of heat-resistant concretes

tn contrast to the structural materials of nuclear reactors, the radiation resistances of concretes used in biological shielding have not been sufficiently studied. A tendency has recently arisen for the preferential use of heat-resistant concretes in biological shielding instead of materials such as steel, cast iron, graphite, boron, etc., which are costly and in relatively short supply. In this paper we shall indicate the effect of reactor neutron irradiation on certain properties of Porland-cement and liquid-glass heat-resistant chromite concretes. The integral neutron flux used in this investigation was (2–2.4) x 10²¹ neutrons/cm² and the irradiation temperature up to 550° C.It was found experimentally that these concretes:retain quite high strength and elastic properties. The thermal conductivity and thermal expansion coefficient change very littte. It is concluded that such concretes may be recommended for use in the biological shielding of nuclear reactors.Translated from Atomnaya Énergiya, Vol. 21, No. 2, pp. 108–112, August, 1966.     

A pulsing reactor

A proposal for the design of a pulsing reactor operating on light water and enriched uranium is given in this article, and the method of calculating the physical and thermal parameters of such a reactor are indicated. It is shown that when certain definite conditions are established very stable heat generation during the neutron pulses may be obtained. In the version of a pulsating reactor discussed here, the calculations indicate that it is possible to obtain over 5 Mw power with an average thermal neutron flux greater than 10¹⁴ neutrons/cm²· sec and a maximum neutron flux (during neutron pulse) of approximately 10¹⁷ neutrons/cm² · sec in the reactor core.The author takes this opportunity to express gratitude to Academician I. V. Kurchatov for his interest in this work. The author is also grateful to U. N. Zankov for his discussion of the conclusions, and to A. K. Sokolov for his participation in a number of calculations.     

The effect of neutron irradiation on the structure and mechanical properties of alloy steels

We have studied the effect of neutron irradiation at temperatures of 200–500°C with various integral doses (1.5·10²⁰·10²¹ neutrons/cm²) on the properties and microstructure of some steels with different chemical compositions and initial structures. We have shown the effect of alloying by various elements on the sensitivity of the steel to irradiation and the temperature of annealing of radiation defects of hardening.Translated from Atomnaya Énergiya, Vol. 15, No. 1, pp. 30–37, July, 1963     

Attenuation of ambient dose equivalent from neutrons by thick concrete, cast iron and composite shields for high energy proton, He, Ca and U ions on Cu targets for shielding design

Data on neutron dose attenuation by thick concrete, cast iron, and cast iron plus concrete composite shields for heavy ions and protons having high energies (200-1000 MeV/u) are necessary for shielding designs of high-powered heavy ion accelerator facilities. Neutron production source terms, shield material attenuation lengths, and neutron dose rate reduction effectiveness of the bulk shielding in the angular range from 0° to 125° were determined by the Particle and Heavy Ion Transport Code (PHITS) for beams of 300 and 550 MeV/u ⁴⁸Ca ions, 200 and 400 MeV/u ²³⁸U ions, 800 MeV/u ³He and 1 GeV protons. Calculated results of interaction lengths of concrete and cast iron were also compared with similar work performed by Agosteo et al., and to experimental and other calculated data on interaction lengths. The agreement can be regarded as acceptable.     

Neutron irradiation of sapphire for compressive strengthening.: I. Processing conditions and compressive strength

Sapphire suffers a dramatic loss of c-axis compression strength at elevated temperatures. Irradiation of sapphire with fission-spectrum neutrons to an exposure of ∼10²² neutrons/m² in the core of a 1 MW fission reactor increased the c-axis compression strength by a factor of ∼3 at 600 °C. Strength was similarly improved when 99% of slow neutrons (?0.1 eV) were removed by 10B and Cd shields during irradiation. Annealing at 600 °C for 10 min changed the yellow-brown color of irradiated sapphire to pale yellow, but had no effect on compressive strength. Annealing irradiated sapphire at 1200 °C for 24 h reduced the compressive strength to its baseline value. Transmission electron microscopy suggests that fast-neutron-induced displacement damage inhibits the propagation of r-plane twins which are responsible for the low compressive strength. When irradiated with 10B and Cd shielding, sapphire that was not grown in iridium crucibles is safe for unrestricted handling after 1 month.     

Attenuation of pile radiations in serpentinite sand

Results of an experimental study of the attenuation of pile radiations in serpentine sand having a bulk density of 1.62 g/cm³ are reported. The sand investigated contains about 11.5% chemically bound water liberated at temperatures upwards of 450°C.The attenuation of fast flux and thermal fux, attenuation of neutron and gamma dose rate, and fast-neutron spectra in serpentinite sand were measured. Relaxation lengths of fast neutrons computed from experimental data are compared to relaxation tengths of fast neutrons in boron carbide, serpentinite concrete, and iron-ore concentrate.Translated from Atomnaya Énergiya, Vol. 19, No. 4, pp. 354–359, October, 1965     

Attenuation of reactor thermal neutrons in a bulk shield of ordinary concrete

This work is concerned with the study of the distribution and attenuation of doses of thermal neutrons emitted directly from the core of ²³⁵U research reactor in ordinary concrete shields. In practice it is not possible to identify the reactor thermal neutrons in the emitted continous neutron spectrum, therefore, measurements were carried out by using a direct and cadmium filtered beam of reactor neutrons. All measurements were performed using Li₂B₄O₇:Mn thermoluminescent dosimeters.The data obtained were analysed and the dose distributions of reactor thermal neutrons were evaluated. A group of isodose curves were constructed which give directly the shape and thickness of the shield required to attenuate the intensity of doses of reactor thermal neutrons to specific values. In addition, the thermal neutron relaxation lengths in ordinary concrete were derived for disc collimated beam and infinite plane monodirectional sources.     

Measurements of neutron induced activation of concrete at 64.5 MeV

The present paper presents the measurement of neutron induced activations on concrete using the 64.5 MeV quasimonoenergetic neutrons produced at the intense ⁷Li(p, n) neutron source at Cyclotron and Radioisotope Center, Tohoku Univeristy (CYRIC). The data were corrected for the effect of continuous neutrons in the source. The neutron energy, neutron yields and the spectrum of continuous neutrons were confirmed with the neutron time-of-flight method and the neutron activation measurement of the ²⁰⁹Bi(n, Xn) reactions having various threshold energy values. The nuclides produced by thermalized source neutrons are negligible. New data were obtained for concrete activation.     

Gamma ray and neutron shielding properties of some concrete materials

Shielding of gamma-rays and neutrons by 12 concrete samples with and without mineral additives has been studied. The total mass attenuation and linear attenuation coefficients, half-value thicknesses, effective atomic numbers, effective electron densities and atomic cross-sections at photons energies of 59.5 and 661 keV have been measured and calculated. The measured and calculated values were compared and a reasonable agreement has been observed. Also the recorded values showed a change with energy and composition of the concrete samples. In addition, neutron shielding has been treated in terms of macroscopic removal cross-section (Σ_R, cm⁻¹) concept. The WinXCom and NXcom programs were employed to calculate the attenuation coefficients of gamma-rays and neutrons, respectively.     

Radiation shielding of concretes containing different lime/silica ratios

The shielding of γ-rays and fast neutrons by concrete has been studied for concretes containing different lime/silica ratios. Calculations were carried out for six different concrete samples. The total mass attenuation coefficients (μ/ρ, cm² g⁻¹) have been computed at photon energies of 1 keV to 100 GeV using the personal computer software package WinXCom. Also the macroscopic effective fast neutron removal cross-sections (Σ_R, cm⁻¹) have been calculated using MERCSF-N program and the removal cross-section database for all required elements. The obtained results showed that the lime/silica ratio of concrete has significant and insignificant effects on μ/ρ and Σ_R values, respectively.     

Evaluation of gamma-ray and neutron energy for area monitoring system in the IFMIF/EVEDA accelerator building

The engineering validation of the IFMIF/EVEDA prototype accelerator, up to 9 MeV by supplying the deuteron beam of 125 mA, will be performed at the BA site in Rokkasho. A design of this area monitoring system, comprising of Si semiconductors and ionization chambers for covering wide energy spectrum of gamma-rays and ³He counters for neutrons, is now in progress. To establish an applicability of this monitoring system, photon and neutron energies have to be suppressed to the detector ranges of 1.5 MeV and 15 MeV, respectively. For this purpose, the reduction of neutron and photon energies throughout shield of water in a beam dump and concrete layer is evaluated by PHITS code, using the experimental data of neutron source spectra. In this article, a similar model using the beam dump structure and the position with a degree of leaning for concrete wall in the accelerator vault is used, and their energy reduction including the air is evaluated. It is found that the neutron and photon flux are decreased by 10⁴-order by employing the local shields using concrete and polyethylene around beam dump, and the photon energy can be suppressed in the low energy.     

Flux of secondary neutrons produced by 660 MeV protons in shielding

With the help of a set of threshold and resonance detectors, measurements were made of the spatial and energy distribution of secondary neutrons in graphite and nickel blocks. Absolute values of the neutron flux as a function of depth in an infinite slab were obtained for a plane, monodireetionat proton source. The energy distribution of the secondary neutrons in the energy range 2.5·10^–8 to 6.6·10² Mev was represented by seven groups. The magnitude of the dose behind plane nickel and graphite shielding as a function of thickness was also determined. The results are discussed.Translated from Atomnaya Énergiya, Vol. 18, No. 6, pp. 573–578, June, 1965     

Passage of neutrons through narrow gaps in the blanket and shielding of a thermonuclear reactor

Conclusions Straight, through cracks in the blanket and shielding of a thermonuclear reactor even with a width of about 5 mm lead to a significant local increase in the fluence of fast and intermediate neutrons in the region of the magnetic coils of the toroidal field. In spite of the sharp drop in the fluxes of 14-MeV neutrons away from the axis, the region where the gap has an appreciable effect is large and equals 5–15 cm, depending on the gap width.The use of a stepped gap significantly increases the efficiency of the shielding when the axes of the gaps are dipslaceed by a distance greater than two gap widths. For smaller displacements the effect of the gaps is appreciable, which sharply reduces the efficiency of the shielding. Even with displacements exceeding two gap widths, however, the coefficient of variation of the fluence of 14-MeV neutrons is large and equals 3–7 for gap widths ranging from 0.5 to 1 cm.Preliminary calculations show that over a period of five years of operation of a test thermonuclear reactor with a neutron load of 1 MW/m² on the first wall the dose absorbed in the layer of insulation of the magnetic coils of the toroidal field, which is adjacent to the cryostat, can reach (5–6)·10⁹ rad, and the number of displacements in the copper stabilizing conductor can reach (1–5)·10^–4 per atom. These values fall at the limit of admissibility, and for this reason even a local increase in the values owing to the gap could be critical. To make a more accurate evaluation of the effect the construction of the blanket and the magnetic coils of the test thermonuclear reactor as well as the criteria adopted for the efficiency of the shielding of the thermonuclear reactor must be specified more accurately.Translated from Atomnaya Énergiya, Vol. 65, pp. 339–343, November, 1988     

Fission Spectrum Averaged Cross Sections of Some Threshold Reactions Measured with Fast Reactor “YAYOI”

Abstract

Fission spectrum averaged cross sections of twenty one threshold reactions were measured in the core center of YAYOI which was a fast neutron source reactor. Fast neutron spectrum in the core was experimentally determined by using a set of activation foils and micro-fission counters, prior to the cross section measurement. It was found that the shape of the fast neutron spectrum was approximately the same as that of fission neutrons above about 2MeV. This fact was also supported by theoretical calculation.

Since this neutron field has scarce thermal and epithermal neutrons, measurement of nuclei produced by threshold reactions is not affected by (n, γ) reactions which are induced by thermal and epithermal neutrons. Moreover, considerably high fast neutron flux (about 5 x 10¹¹n/cm²·sec) enables to measure cross sections of small values.

The results in general agreed with the previous values obtained in a reactor core or with a fission plate within an experimental error, while they were systematically smaller by about 10% than those recommended by Fabry. The measured values are also compared with the results calculated by Pearlstein based on a statistical model.     

Prospects for the development of neutron-activation analytical facilities based on powerful antimony-beryllium sources

Conclusions When using antimony-beryllium neutron sources for activation analysis of the composition of a substance, the problem of the analytical monitoring of many elements can be solved with a limit of determination of 10^–3–10^–5%. Based on the data obtained by the authors concerning the spatial distribution of the neutrons from a¹²⁴Sb–Be-source in different moderators, a beryllium-graphite assembly with a powerful¹²⁴Sb source has been designed, manufactured, and introduced into operation, for neutron-activation analysis.As applicable to the¹²⁴Sb–Be-graphite assembly, a procedure has been developed for the neutron-activation determination of gold, with a limit of determination of 2·10^–5%. The possible limits of the neutron-activation determination of certain other elements have been estimated. In order to ensure operation of the facility with recharging of the source once in 6 months, it is advisable to carry out the preparation of a source in a neutron flux with a density of 3·10^–13 neutrons/ (cm²·sec) during 30–50 days (for a mass of metallic antimony of 500 g) with subsequent two-week cooling in order to reduce the¹²²Sb activity. Initial data have been obtained for the design of a transportation container for powerful¹²⁴Sb sources.Translated from Atomnaya Énergiya, Vol. 53, No. 4, pp. 255–260, October, 1982.     

Simulation of e–γ–n targets by FLUKA and measurement of neutron flux at various angles for accelerator based neutron source

A 6 MeV Race track Microtron (an electron accelerator) based pulsed neutron source has been designed specifically for the elemental analysis of short lived activation products where the low neutron flux requirement is desirable. The bremsstrahlung radiation emitted by impinging 6 MeV electron on the e–γ primary target, was made to fall on the γ–n secondary target to produce neutrons. The optimisation of bremsstrahlung and neutron producing target along with their spectra were estimated using FLUKA code. The measurement of neutron flux was carried out by activation of vanadium and the measured fluxes were 1.1878 × 10⁵, 0.9403 × 10⁵, 0.7428 × 10⁵, 0.6274 × 10⁵, 0.5659 × 10⁵, 0.5210 × 10⁵ n/cm²/s at 0°, 30°, 60°, 90°, 115°, 140° respectively. The results indicate that the neutron flux was found to be decreased as increase in the angle and in good agreement with the FLUKA simulation.     

Measurements and Calculations of Neutron Energy Spectra Behind Polyethylene Shields Bombarded by 40- and 65-MeV Quasi-Monoenergetic Neutron Sources

Measurements of neutron energy spectra behind 30.5-, 61.0-, 122.0-, 183.0-cm-thick polyethylene shields bombarded by 40- and 65-MeV quasi-monoenergetic neutrons are performed at the 90-MeV AVF cyclotron of the TIARA (Takasaki Ion Accelerator for Advanced Radiation Application) at JAERI (Japan Atomic Energy Research Institute). Source neutrons are produced at 3.6- and 5.2-mm-thick⁷ Li targets bombarded by 43- and 68-MeV protons, respectively. A BC501A organic liquid scintillator and multi-moderator spectrometer with a ³He counter (Bonner ball) are used for spectrometry of transmitted neutrons and their energy spectra are obtained with the unfolding technique. The energy spectra from a few MeV up to a peak energy are obtained by the BC501A scintillator measurement and those below a few MeV down to thermal energy are obtained by the Bonner ball measurement. The measurements are performed on the neutron beam axis and at off-center positions, and attenuation profiles of neutron fluxes along the beam axis are obtained. The MORSE Monte Carlo calculations are performed with the DLC119/HIL086 multi-group cross section library for comparison with the measured data. The calculation generally gives a little overestimated fluxes, and a few % longer attenuation lengths of peak flux and dose equivalent.     

In-pile dimensional changes in neutron irradiated zirconium base alloys

A model is developed to predict in-pile growth in zirconium base alloys as a function of neutron flux, neutron fluence, temperature, dislocation density, and texture. The model is based on vacancy and interstitial behavior with respect to straight dislocations, dislocation loops, depleted zones and grain or sub-grain boundaries. Results indicate very little growth dependence on temperature or neutron flux at temperatures below ~320°C, at fluxes above ~10¹³ n/cm² sec, and at fluences below 10²¹ n/cm². As the flux is lowered and the temperature and fluence are raised, the temperature and flux dependencies increase. Comparison between theory and data is given for both growth and dislocation loop size.