Journal of Nuclear Energy最新文献

The neutron capture cross-section of ¹³⁵Xe has been measured by the activity depletion method using mass separated-ion implanted samples and high resolution gamma-ray spectrometry. An effective cross-section value was measured at a neutron temperature of 318 ± 5°K as 3·09 ± 0·07 Mbarns, relative to a ⁵⁹(Co(n,γ) cross section of 37·09 barns. A calculated value for the ¹³⁵Xe 2200 m/sec capture cross-section is discussed.

The averaged number of prompt neutrons per ³³⁵sU and ²³⁹Pu fission have been measured in fast and thermalized neutron fluxes at the fast zero-power reactor (FRO) at Studsvik, Sweden. The measurements were made with three different reactor cores giving median fission energies for ²³⁵U (core centre) of 50, 170 and 240 keV. The spectrum averaged v data were compared with the corresponding thermal $\text{ν}$ data. The $\text{ν}\text{-}\text{value}$ of the spontaneous fission of ²⁵²Cf was used as an intermediate standard. The fission samples were placed in neutron beams extracted from the reactor cores and a large liquid scintillator was used as the fission neutron detector. The experiment also included relative fission rate measurements, ²³⁸U/²³⁵U, for the three reactor cores.

The results are compared with calculated relative $\text{ν}\text{-}\text{values}$ deduced from different sets of recently evaluated fission cross section and $\text{ν}\text{-}\text{data}$ and from different assumptions concerning the reactor neutron spectra. Within the experimental one standard deviation errors of 1·5 and 2 per cent for ²³⁵U and ²³⁹Pu a good agreement was obtained between the experimental and calculated data assuming $\text{ν}\text{-}\text{values}$ in the fission resonance region equal to $\text{ν}$ at thermal energy.

First flight transport kernels are derived for a spherical shell region where the central zone is empty. Assuming a uniformly distributed, isotropic source throughout the annulus the uncollided angular flux distribution is obtained enabling the formulation of the current at any point in the medium. It is shown that the current at the outer radius of the shell, and therefore also the escape probability P_e of the shell, can be put in a simple analytical form. Numerical values of P_e are given for different inner radii and different shell thicknesses and comparison is made with the escape probabilities for spherical shells with a black central region. Finally, an upper limit is given for the relative error introduced in a shell collision probability by the common assumption of a perfectly absorbing inner zone.

The angular distributions and energy spectra of keV neutrons scattered from a typical target assembly employing the ⁷Li (p, n) ⁷Be reaction have been calculated by the Monte Carlo method for several proton energies, target backing and beam tube wall thicknesses and target radii. A brief discussion is given of possible errors due to scattering in the experimental measurements of neutron activation capture cross-sections, dose equivalents and angular distributions.

A direct numerical inversion of the Laplace transform of the time-variable is used to produce the time-dependence of an infinite medium neutron field. Comparisons with exact solutions show that excellent agreement can be obtained by suitable scaling of the time variable. The examples chosen are slowing down in natural iron and slowing down and up-scattering in light water.

Neutron induced fission events emitting 2, 3, 4 or 5 neutrons have different contributions to fission yields recorded with double and triple neutron coincidence signals as a function of incoming neutron energy. This fact can be used to search for variations of neutron multiplicities in fission resonances, in particular in resonances of different channel spins.

In the case of ²³⁵U no difference in fission neutron numbers for 3⁻ and 4⁻ levels has been detected above the experimental error of ± 1·5 %.