Journal of Nuclear Energy最新文献

If the epithermal activation C_ep of a foil detector is derived from its epicadmium activation C_Cd, the knowledge of the cadmium correction factor $\text{F}{}_{\mathrm{<mtext>Cd</mtext>}}\text{=}\text{C}{}_{\mathrm{<mtext>ep</mtext>}}\text{C}{}_{\mathrm{<mtext>Cd</mtext>}}$ is usually required. The cadmium correction factors of V, B, ¹⁰B, In, Au, Mn, Co, and Dy foil detectors of thickness 0·0·25 mm and cadmium cover thicknesses of 0·5, 0·8, 1·0, and 1·2 mm have been numerically evaluated according to the most recent cross-section values and resonance parameters. F_cd values are listed for both monodirectional and isotropic neutron incidence. While for most of these materials no results have been published so far, for gold and indium the results could be compared with previous calculations and measurements and a discussion of the results is given. Since the F_Cd values, particularly for Mn, Co, and Dy, exceed unity considerably, neglecting this correction is not tolerable with precise measurements of the thermal neutron flux density. Comments on the use of F_Cd values in the Westcott convention are added.

Radial diffusion coefficient for a cylindrical cell is calculated by the P₃-approximation in one-group theory. Williams calculated the anisotropic diffusion coefficient in a cylindrical cell by the P₃-approximation. His calculation was, however, limited to the axial diffusion coefficient. Therefore we calculate the radial diffusion coefficient by using the cylindrical approximation introduced by Williams and by following the concise procedure of the P_N-method by Davison. As an example of numerical calculation we treat a cell consisting of two different regions, a fuel rod and a surrounding moderator. Results from the present method are compared with those from the diffusion theory and from the collision probability method.

A new precise measurement of the 2200 m/s fission cross-section of ²³⁵U was performed with a slow chopper at the BR2 reactor (energy-range from 0·002 to 0·15 eV). Simultaneous recordings of pulse-height and time-of-flight spectra of successively compared ²³⁵UF₄-layers and “standard” elemental boron-layers were made. The detection of the ¹⁰B(n, α)⁷Li- and ²³⁵U(n, f) reaction fragments is made with excellent pulse-height resolution in low geometry with a surface-barrier detector. The absolute ¹⁰B contents were determined from direct weighing in vacuum combined with careful chemical and isotopic analysis of witness foils. The relative ¹⁰B contents of the standard boron-foils were checked by their neutron induced reaction rates in the geometry of the fission experiment.

The amount of uranium on the fissile foils is determined by precise low geometry α-counting making use of the most recent and accurate α half-life values. The fission foils were also checked by α- and fission fragment counting in the geometry of the experiment.

The fission cross-section was calculated in the energy-range from 0·002 to 0·15 eV and the absolute value deduced at 2200 m/s was 587·6 ± 2·6 barn. The value at 25 × 10⁻³ eV is 591·8 ± 2·7 barn. These fission cross-section values were used to calculate the Westcott-g_φ-factor for a 20·44°C Maxwellian-distribution with as result: g_f = 0·9780 ± 0·0010.

The decay of a neutron burst in a plane lattice with symmetric cells is computed by perturbation calculus up to the second order. The perturbation is shown to be negative when the unperturbed system is the homogeneous system obtained by volume averaging of the cross-section in the cell. As an illustration, generalized Nelkin's expansion is computed for plane water-beryllium lattices.

In this paper we present some new results concerning the β + γ residual power following the ²³⁵U thermal fission for cooling times going from 70 to 7 × 10⁴ s.

We give a brief description of the experimental technique and of the methods used to obtain the residual power from the measurements. Then we examine the different sources of errors. The residual power curves for three different irradiation times as well as a curve due to one fission (elementary fission curve) are given.

These experimental results are compared with other values coming from β and γ measurements and from theoretical calculations based on fission products systematics. For the considered cooling times the discrepancies between our experimental results and the theoretical calculations are less than 12% for cooling times <700 s and inside the experimental uncertainties for cooling times > 700 s.

Using heterogeneous formalism we calculated Selengut equivalent diffusion coefficient and absorption cross-section for a heterogeneous cell, containing absorbing media. The observed deviation between Benoist and Selengut diffusion coefficients and also between average and Selengut absorption cross-sections are almost exactly compensated by cell-edge flux normalisation in the thermal group.

In this paper, discrete ordinate neutron transport equations which are equivalent to the P_L approximation are derived in order to eliminate the current problem in the numerical solutions for the transport equation. The number of the unknowns which remain to be solved in the discrete ordinate equations is consistent with that of the independent basic functions appearing in the P_L solution.

First, for even P_L approximation in one-dimensional slab geometry, a system of equations which are satisfied by the L unknowns is derived by eliminating the L-th Legendre moment.

For multi-dimensional geometry, discrete ordinate equations of the second order which are equivalent to the P_L approximation are proposed by following Davis' method for the derivation of the even-parity second-order form of the odd P_L equations. In this case, the total number of the quadrature points is just equal to that of the independent basic functions.

Finally, the boundary conditions of the P_L approximation in x-y geometry are transformed into those for the corresponding discrete ordinate equations. As for the boundary conditions, material interface, reflecting and vacuum boundary conditions are considered.

Analytical descriptions of fast neutron elastic scattering cross-sections and generalized optical model parameter sets are of great value in connection with reactor shielding calculations as well as for predictions of the nuclear properties of fission products. Such descriptions and generalizations have previously been presented on the basis of scattered experimental information, contrary to the present evaluation which is founded on a relatively large set of neutron data. The neutron elastic scattering data collected at our laboratory for a number of elements and at several neutron energies have been used in an attempt to get generalized analytical expressions for some parameters of the spherical optical model potential, i.e. the real and imaginary potential depths and the radii of these potentials. Even if the experimental information is comparatively extensive it is still somewhat too meagre in some atomic mass ranges to be used for such a purpose. Thus the present work does not exclude future cross-section measurements for several elements but it makes it possible to estimate cross-sections for elements where experimental information is lacking.