Transmission of neutron radiation from a reactor through hydrogen-free media

The results of experimental and theoretical investigations into the transmission of neutrons through hydrogen-free, uniform materials and mixtures are discussed. The relaxation lengths obtained for neutrons of different energy groups are discussed. It is shown that under certain assumptions the mag 7 nitude of the flux attenuation for neutrons of the upper energy group (E_n 3 MeV) in a hydrogen- free medium can be obtained from very simple empirical expressions by using well-known constants: the removal cross section Ore m, the cross section obtained from the reciprocal of the relaxation length (₁/) and the asymptotic cross section _as. Values are derived for these constants.Translated from Atomnaya Énergiya, Vol. 18, No. 4, pp. 350–357, April, 1965     

γ-Ray yield from radiative capture in iron

A technique is described for measuring the yield of capture radiation from iron by the use of a single crystal scintillation -spectrometer. Results are reported for the yield of capture radiation from iron, normalized to the number of neutrons either entering shielding or escaping from it. Measurements were made for the neutron spectrum from a Po-Be source. The dependence of capture y-radiation yield on shielding thickness is given. The effect of boron-containing impurities on the yield of capture radiation was investigated.The authors wish to take this opportunity to thank I. I. Bondarenko and S. G. Tsypin for much valuable advice and comment in discussions of the experimental results.     

Stationary transport of monoenergetic neutrons in inhomogeneous media

The passage of monoenergetic neutrons from a stationary source in an inhomogeneous medium is considered; this may be described by means of a Boltzmann equation with fluctuating coefficients. An equation is derived for the flow of neutrons (neutron flux) averaged over the fluctuations in the medium. A solution is obtained for this equation in limiting cases in which the characteristic dimension of the inhomogeneity 1_s ⁰ (free path in relation to scattering in a homogeneous medium) and 1_s ⁰. In both cases 1L₀ (diffusion length of neutrons in a homogeneous medium). By way of an example of an inhomogeneous medium, the periodic lattice of a reactor is considered.Translated from Atomnaya Énergiya, Vol. 22, No. 4, pp. 271–276, April, 1967.     

14 MeV neutron yields from D-T operation of the MIT Cockcroft-Walton accelerator

As part of our fusion-product diagnostic development program, we have begun a series of experiments with 14 MeV neutrons generated in a Cockcroft-Walton accelerator. Two different detectors have been used to measure the neutron yield: a silicon SBD and a Cu foil. The energy of the emitted neutrons has been determined by using two spectrometers: the SBD and a³He proportional counter. The reaction rate is monitored, with about ±5% accuracy, by detecting the particles from D + T n +. The neutron yields obtained from the Si detector and the Cu activation had associated uncertainties of about ±15% and agreed well with the predicted values from measurements.     

Initial high beta operation of the HBT-EP Tokamak

HBT-EP is a new research tokamak designed and built to investigate passive and active feedback techniques to control MHD instabilities. In particular, HBT-EP will be able to test techniques to control fast MHD instabilities occurring at high Troyon-normalized beta, _N Ba/I_p [Tm/MA], since it is equipped with a thick, close-fitting, and adjustable conducting shell. The major goals of the initial operation of HBT-EP have been the achievement of high beta operation (_N 3) using only ohmic heating and the observation of MHD instabilities. By using a unique fast startup technique, we have successfully achieved these goals. A variety of MHD phenomena were observed during the high beta operation of HBT-EP. At modest beta (_N 2), discharges have been maintained for more than 10 msec, and these discharges exhibit saturated resistive instabilities. When _N approaches 3, major disruptions occur preceded by oscillating, growing precursors. During start-up, one or more minor disruptions are usually observed. A 1-D transport code has been used to simulate the evolution of the current profile, and these early minor instabilities are predicted to be double tearing modes. The simulation also reproduces the observed high beta operation when saturated neo-Alcator energy confinement scaling is assumed.     

Distribution of the ratio of radiation capture and fission cross sections for Pu239 with respect to height in the BR-5 reactor

A study was made of the distribution of the ratio of radiation capture and fission cross sections for Pu²³⁹ with respect to height in the BR-5 reactor. The distribution of neutron capture reactions was measured by determining the concentration of Pu²⁴⁰ from the rate of spontaneous fissions in plutonium samples irradiated in the reactor with an integral flux of 10²¹ to 10²² neutrons/cm². As initial material for irradiation, almost isotopically pure Pu²³⁹ (Pu²⁴⁰ content 5·10^–3% was used. The values of obtained rose from 0.1 to 0.8 on moving away from the center of the reactor. The data corresponding to equilibrium neutron spectra in the active zone and the outer region of the reflector agree with the results of measuring and _f in monoenergetic neutrons.Translated from Atomnaya Énergiya, Vol. 16, No. 6, pp. 497–500, June, 1964     

Effect of a decontamination system on the buildup of active corrosion products in pressurized water reactors

The process of formation, buildup and transfer of active corrosion products in pressurized water reactors and the effect of purification systems are discussed. Solutions are obtained for the differential equations describing these processes with a decontamination factor of 0 and 0. An analysis of these solutions is made.It is shown that the efficiency of the decontamination system has a considerable effect on the activity of the water and the corrosion deposits in the reactor system.Translated from Atomnaya Énergiya, Vol. 20, No. 3, pp. 247–252, March, 1966     

Radiation Conditions during Decontamination of an Experimental Facility for Reprocessing Nuclear Fuel

The basic characteristics of the changes occuring in the radiation conditions in an experimental facility for reprocessing spent nuclear fuel during decontamination of the facility after reprocessing irradiated uranium and uranium–plutonium BOR-60 reactor fuel and plutonium tetrafluoride into dioxide by the pyroelectrochemical method are examined. An expression is obtained experimentally for calculating the decrease in the power of photon radiation from contaminated surfaces as a function of the number of decontamination cycles. It is shown that for one-time processing of the surfaces of stainless steel equipment by the two-bath method the decontamination coefficient varies over the range 2.5–25 for emitters and 5–30 for and , emitters.     

Measurement of the effective cross section in Th232 for thermal neutrons and the resonance integral for neutron absorption

The effective cross section in thorium for thermal neutrons (_ther = 7.31 ± 0.10 barns) and the resonance integral for thorium have been-measured in a heavy water reactor. The measurements were made by the activation method. Gold, indium, and uranium were used as comparison standards. The precise value of the effective cross section for indium for thermal neutrons is (_ther = 162 ± 10 barns) and the resonance integral in indium is (RI=2340 ± 200 barns).     

Corrections for multiple scattering in spherical and circular geometries

In this paper a method is considered of introducing corrections for multiple scattering into the results of measurements of angular distributions of elastically scattered neutrons. It is assumed that the mean path of the neutron in the sample in which scattering takes place is comparable with the neutron free path.In the first part, using direct calculation of the integrals, we find the probability for double scattering and estimate the probabilities for triple and higher-order scattering for a sphere and for a ring of circular and rectangular cross section in the case of isotropic neutron scattering. In the case of anisotropic neutron scattering, at neutron energies of the order of several million electron volts the cross section may be given as a sum () = ₁ () + ₂ (), where ₁() is the forward peak and ₂() is more or less isotropic. Using this representation all elastic scattering events may be provisionally divided into two groups while all double scattering events can be divided into four groups. The probabilities of double scattering for all four are calculated on the basis of results obtained for isotropic scattering. Triple and higher-order scattering are evaluated in similar fashion.     

Calculating the transient temperature field in a reactor tube and the thermoelastic stresses in the fuel element cladding

Approximate analytic methods are given for calculating the transient temperature field in the fuel elements and the coolant temperatures at any point along the reactor tube, as well as the transient thermoelastic stresses in the cladding of a cylindrical fuel element. The coolant temperature at the input to the tube is constant, and the coolant undergoes no changes in state of aggregation. The approximate methods are illustrated by examples.Results are given, for comparison, of accurate calculations of the same examples made with a rapid calculating machine.List of symbols time - r; z coordinates (radius, distance along tube) - r₁; r₂ internal and external radii of fuel element cladding respectively - H total active length of fuel element - a₁; ₁;c _{1 1} coefficients of temperature conductivity, heat conductivity, specific heat capacity and specific gravity of fissionable material respectively - a₂; ₂; C_p2; ₂ cladding parameters - a; ; c_p; coolant parameters - mean cladding radius - f:f₂ cross-sectional area of tube for coolant and cladding respectively - w coolant velocity - coefficient of heat release to coolant - t (r, ); (); () fuel temperature, mean temperature over cross section of cladding, and coolant temperature at pointz. along tube respectively - qv() specific volume of coolant at pointz - values averaged overz - quantities at the initial instant of time - ₃ delay time - _n time required for coolant to go from z=0 to the point in question     

Prompt γ rays from fission

The prompt -ray spectrum from the fragments of U²³⁵ thermal-neutron fission was calculated exactly within the framework of the basic assumptions of statistical theory for the emission of neutrons and rays from excited nuclei. The shape of the spectrum corresponded roughly to that obtained experimentally; however, the computed total energy carried away by the rays (6.2 MeV) was considerably below the experimental value (8–9 MeV). Conclusions are drawn as to the possibility of eliminating this disparity if the thinness of nuclear levels close to the ground state (0–2 MeV) in the fragments is taken into account in calculating residual excitation.Translated from Atomnaya Énergiya, Vol. 19, No. 2, pp. 116–119, August, 1965     

The reversed-field-pinch (RFP) fusion neutron source: A conceptual design

The conceptual design of an ohmically heated, reversed-field pinch (RFP) operating at 5-MW/m² steady-state DT fusion neutron wall loading and 124-MW total fusion power is presented. These results are useful in projecting the development of a cost effective, low-input-power (206 MW) source of DT neutrons for large-volume (10 m³), high-fluence (3.4 MW yr/m²) fusion nuclear materials and technology testing.Work supported by U.S. DOE.     

Measurement of neutron emission from Ti and Pd in pressurized D2 gas and D2O electrolysis cells

Experiments using high-efficiency neutron detectors have detected neutron emission from various forms of Pd and Ti metal in pressurized D₂ gas cells and D₂O electrolysis cells. Four independent neutron detectors based on³He gas tubes were used. Both random neutrons (0.05–0.2 n/s) and time-correlated neutron bursts (10–280 n) of 100-s duration were measured using time-correlation counting techniques. The majority of the neutron burst events occurred at –30°C as the samples were warming up from the liquid nitrogen temperature.     

An investigation of γ-emission associated with thermal neutron capture

A number of spectra of -rays from the (n, ) reaction brought about by irradiation by thermal neutrons were investigated with the use of a magnetic Compton spectrometer at the VVR reactor of AN SSSR. The conditions of the experiment are described, as well as the results of measurements made on the -spectra of tin and antimony.In conclusion, we wish to thank A. S. Volkov for constructing the electronic apparatus and providing for its stable operation; also, we thank the maintenance organization of VVR AN SSSR.     

Calculation of neutron and gamma-ray energy spectra for fusion reactor shield design: Comparison with experiment-II

Measured and calculated neutron and gamma-ray energy spectra resulting from the transport of 14 MeV neutrons through a 0.30-m-thick lithium hydride slab and through a 0.05-m-thick lead slab followed by 0.30 m of lithium hydride are compared. Also reported are comparisons of the measured and calculated neutron energy spectra behind an 0.80-m-thick assembly comprised of stainless steel type 304 and borated polyethylene. The spatial dependence of the gamma-ray energy deposition rate measured using thermoluminescent detectors is compared with calculated data. The calculated data obtained using two-dimensional radiation transport methods and ENDF/B-IV cross-section data are in good agreement for all of the experimental configurations. Calculated integral neutron energy spectra agree with the measured data within 5 to 20% depending on neutron energy for the LiH and Pb plus LiH assemblies. The gamma-ray spectra agree within 20% for these slabs. The measured and calculated neutron energy spectra behind the SS-304-borated polyethylene assembly agree within 5% except at neutron energies below 5 MeV where background radiation influences the measured spectra. The gamma-ray energy deposition rates as a function of depth agree within a factor of two at all detector locations.     

Neutron yield in inelastic collisions of 14-Mev neutrons with nuclei and cross section for the (n, 2n) reaction

A measurement has been made of , the number of neutrons produced in one inelastic scattering event between a neutron and a number of elements of natural isotopic composition: Fe, Cu, Mo, Cd, Sn, Sb, Hg, Pb, Bi and U. The measurements were performed by determining the relative change in the total neutron flux and the attenuation of the primary neutrons after passage through samples of the materials being investigated. It was also possible to obtain data on the cross section _in for inelastic collisions of neutrons with the above-mentioned nuclei. The values of and _in, in conjunction with the known cross sections for neutron capture, were then used to compute the cross section for the (n, 2n) reaction (averaged over the isotopic composition) in nonfissile nuclei.This work was completed in 1952.The authors wish to thank A.A. Malinkin for comparing the neutron yields from sources used to measure the dependence of counter sensitivity on neutron energy.     

Neutron yield of the reactions Li6(t,n) and Li7(t,n)

The differential cross section is determined for the formation of neutrons at an angle of 0 ° in the reactions Li⁶(t,n) and Li⁷(t, n) in the 0.175–2.4 Mev triton energy range. In the reaction Li⁶(t,n) resonance in the neutron yield is observed at E_t = 1.875 Mev, corresponding to a Be⁹ excitation level of 18.936 Mev.Two resonances are detected in the neutron yield of the reaction Li⁷(t, n): at E_t = 0.765 Mev and E_t = 1.735 Mev, which indicates the presence of Be¹⁰ nucleus excitation levels of 17.78 and 18.46 Mev, respectively. The cross section of the reaction Li⁶(t, n) in resonance is d/d(0 °)_res = 37.2 mb/sr; for the reaction Li⁷ (t, n) the cross section at the first resonance is d/d(0 °)_res = 95 mb/sr; at the second resonance d/d(0 °)_res = 159 mb/sr.The angular distributions of neutrons are given for the interval of angles 0–135 ° (every 15 °) for triton energies of 0.358, 0.559, 1.006, 1.218, 1.370, 1.572, 2.123 Mev.The total cross section for the formation of neutrons at E_t = 2.123 Mev in the reaction Li⁶ (t,n) is equal to 324±32.3 mb in the reaction Li⁷ (t, n) to 1332±83.3 mb.     

Cross section for the formation of gamma-quanta during the interaction of fast neutrons with carbon,lead, and rhenium nuclei

Conclusions The values obtained for the cross sections for formation of -lines (E=4.44 MeV) in carbon agree with the data obtained by other researchers, within experimental error.The results of measurements for lead agree qualitatively with our understanding of the deexcitation of levels in the reaction (n, n) for E_n<4 MeV.The energy distribution of -quanta for rhenium decreases smoothly with increasing E. No irregularities, related to separate intense -transitions in the excited nucleus, are noticeable.Translated from Atomnaya Énergiya, Vol. 49, No. 4, pp. 236–239, October, 1980.     

Measurement of neutron spectra in nickel,iron, and stainless steel

Measurements of neutron spectra in various media, obtained by the transit-time method in the fast pulse reactor IBR with a resolution of 0.04 sec/m, are presented. The spectra of neutrons emerging from iron and nickel prisms of various thickness and also those from stainless-steel prisms are studied. The fine structure due to the resonance character of the cross sections of the media studied was clearly seen in the spectra measured. The experimental neutron spectra are compared with calculations made with a multigroup system of constants allowing for resonance self-screening of the cross sections. The reasons for the slight discrepancies found are analyzed.Translated from Atomnaya Énergiya, Vol. 18, No. 6, pp. 593–601, June, 1965